IR 05000219/2002005
| ML021480417 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/28/2002 |
| From: | Rogge J NRC/RGN-I/DRP/PB7 |
| To: | Skolds J Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-02-005 | |
| Download: ML021480417 (14) | |
Text
SUBJECT:
OYSTER CREEK GENERATING STATION - NRC INSPECTION REPORT 50-219/02-05
Dear Mr. Skolds:
On May 11, 2002, the NRC completed an integrated inspection at your Oyster Creek reactor facility. The enclosed report presents the results of that inspection. The results of this inspection were discussed on May 15, 2002, with Mr. Ernie Harkness and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of this inspection no findings of significance were identified.
Immediately following the terrorist attacks on the World Trade Center and the Pentagon, the NRC issued an advisory recommending that nuclear power plant licensees go to the highest level of security, and all promptly did so. With continued uncertainty about the possibility of additional terrorist activities, the Nation's nuclear power plants remain at the highest level of security and the NRC continues to monitor the situation. This advisory was followed by additional advisories, and although the specific actions are not releasable to the public, they generally include increased patrols, augmented security forces and capabilities, additional security posts, heightened coordination with law enforcement and military authorities, and more limited access of personnel and vehicles to the sites. The NRC has conducted various audits of your response to these advisories and your ability to respond to terrorist attacks with the capabilities of the current design basis threat (DBT). On February 25, 2002, the NRC issued an Order to all nuclear power plant licensees, requiring them to take certain additional interim compensatory measures to address the generalized high-level threat environment. With the issuance of the Order, we will evaluate AmerGens compliance with these interim requirements.
Mr. Jack Skolds 2 In accordance with 10 CFR 2.790 of the NRCs "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/pdr.html (the Public Electronic Reading Room).
We appreciate your cooperation. Please contact me at 610 337-5146 if you have any questions regarding this letter.
Sincerely,
/RA/
John F. Rogge, Chief Projects Branch 7 Division of Reactor Projects Docket No. 50-219 License No. DPR-16
Enclosure:
Inspection Report 50-219/02-05 Attachment: Supplemental Information
REGION I==
Docket No.: 50-219 License No.: DPR-16 Report No.: 50-219/02-05 Licensee: AmerGen Energy Company, LLC (AmerGen)
Facility: Oyster Creek Generating Station Location: Forked River, New Jersey Dates: March 31, 2002 - May 11, 2002 Inspectors: Laura Dudes, Senior Resident Inspector Robert Summers, Senior Resident Inspector Steve Dennis, Resident Inspector Frank Arner, Reactor Inspector, April 4, 2002 - April 11, 2002 Approved By: John F. Rogge, Chief Projects Branch 7 Division of Reactor Projects
SUMMARY OF FINDINGS IR 05000219-02-05; on 03/31-05/11/2002; AmerGen Energy Company, LLC; Oyster Creek Generating Station; integrated report.
This report covered a seven week period of resident inspection and an announced inspection of the independent spent fuel storage installation operation by both resident and a region-based specialist inspectors. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described at its Reactor Oversight Process website at:
http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/index.html.
A. Inspector Identified Findings No findings of significance were identified.
B. Licensee Identified Violations No licensee identified violations were identified.
ii
Report Details Summary of Plant Status:
Oyster Creek (OC) began the inspection period at full power. On April 2, 2002, at 0120, power was reduced to 98 percent for turbine valve testing and a control rod/recirculation flow adjustment. Full power was restored April 2, 2002, at 0555. On April 12, 2002, at 2301, power was reduced to 40 percent for condenser leak repair and control rod drive (CRD) hydraulic control unit planned maintenance. Full power was restored on April 15, 2002, at 0126. On April 16, 2002, at 0122, power was reduced to 88 percent for a control rod pattern adjustment. Full power was restored on April 16, 2002, at 0543. OC remained at full power for the remainder of the inspection period.
1. REACTOR SAFETY Initiating Events, Mitigating Systems, Barrier Integrity (REACTOR-R)
1R04 Equipment Alignment 4160V System a. Inspection Scope The inspectors conducted a complete alignment verification of the 4160V system. The inspectors reviewed operating and surveillance procedures associated with the system and performed a walkdown to verify normal system alignment was maintained in accordance with procedural checklists. In addition, the inspectors reviewed and evaluated the potential impact on the 4160V system operation from open work orders, design modifications, engineering change requests and corrective action process (CAP)
reports. The inspectors also reviewed and discussed system health reports with the system engineer. Documents reviewed included the following:
2000-ADM-3061.01 OC Surveillance Test Program
2000-ABN-3200.44 Loss of Bus 1A1
2000-ABN-3200.45 Loss of Bus 1A2
2000-ABN-3200.46 Loss of Bus 1A3
2000-ABN-3200.47 Loss of Bus 1B1
2000-ABN-3200.48 Loss of Bus 1B2
2000-ABN-3200.49 Loss of Bus 1A3
635.2.001 4160V Switchgear Bus Protective Relay Test
CAP 02002-0714 4160V Fire Protection - CO2 issue
CAP 02002-0291 1E1 4160V bus - alternate feed configuration issue
CAP 02001-1718 1B2 4160V cable failure
337 4160V Electrical System
200-OPS-3024.10A 4160V Diagnostic and Restoration Actions
Operations Workaround List b. Findings No findings of significance were identified.
1R05 Fire Protection a. Inspection Scope The inspectors conducted fire protection inspection activities consisting of plant walkdowns, discussions with fire protection personnel, and reviews of procedure 333, Plant Fire Protection System, and the OC Fire Hazards Analysis Report to verify that the fire program was implemented in accordance with all conditions stated in the facility license. Plant walkdowns included observations of combustible material control, fire detection and suppression equipment availability, and compensatory measures. The inspectors conducted fire protection inspections in the following areas due to the potential to impact mitigating systems:
RB-FZ-1A - Reactor Building 119' elev.
RB-FZ-1E - Reactor Building 23' elev.
OB-FZ-22A - New Cable Spread Room, 63'6 elev.
b. Findings No findings of significance were identified.
1R11 Licensed Operator Requalification a. Inspection Scope The inspectors observed licensed operator requalification training on May 1, 2002. The training/testing exercise was reviewed against criteria listed in NRC Inspection Procedure 71111.11. The inspector reviewed the critical tasks associated with the simulated control room exercise, observed the operators performance during the exercise and observed the post exercise critique. The inspector also reviewed procedure 2611-PGD-2612, OC Licensed Operator Requalification Training Program.
b. Findings No findings of significance were identified.
1R12 Maintenance Rule Implementation a. Inspection Scope The inspectors conducted maintenance rule implementation inspection activities to verify that: (1) failed structures, systems and components (SSCs) were properly characterized in the OC Maintenance Rule Performance Reports, (2) goals and performance criteria were appropriate, (3) corrective action plans were appropriate, and (4) performance was being effectively monitored in accordance with OC procedure 2000-ADM-1220.01, Implementation of the Maintenance Rule. The inspectors selected the following safety significant systems in (a)(1) and (a)(2) status:
A and B Control Room Heating, Ventilation and Air Conditioning System
CRD Hydraulic System
Spent Fuel Cooling System
b. Findings No findings of significance were identified.
1R13 Maintenance Risk Assessment and Emergent Work Evaluation
- 1 Fire Diesel Control Cable Failure a. Inspection Scope On April 15, 2002, the licensee determined that the #1 Fire Diesel engine control cable had failed during post maintenance testing. The inspector reviewed the risk assessment performed by the licensee in accordance with procedure ER-AA-600-1042, On-Line Risk Management, and verified that compensatory actions associated with the #2 Fire Diesel and repair of the #1 Fire Diesel control cable were established. The inspectors also reviewed the OC extent of condition review (CAP 2002-0596) and verified cable repairs were evaluated in accordance with OC procedure 2000-ADM 7210.03, Quality Verification Inspection Program.
b. Findings No findings of significance were identified.
1R15 Operability Evaluations a. Inspection Scope The inspectors reviewed operability evaluations in order to determine that proper operability justifications were performed for the following items. In addition, where a component was determined to be inoperable, the inspectors verified the Technical Specification limiting condition for operation implications were properly addressed.
During the inspection of the 1-1 Reactor Building Closed Cooling Water (RBCCW) Heat Exchanger on April 4, 2002 (WO R0805744), the licensee discovered a degraded baffle plate support beam inside the heat exchanger.
The inspectors reviewed the operability determination prepared by OC engineering including the material nonconformance resolution form, associated structural stress calculations, planned extent of condition reviews, and operational procedural changes recommended to ensure continued operability of the RBCCW system. (CAP-2002-0534)
On April 7, 2002, an adverse trend associated with the offgas radiation monitors was documented by operations personnel. The original operability discussion was documented in CAP-2002-0542, further information revealed degraded heat tracing.
b. Findings No findings of significance were identified.
1R19 Post-Maintenance Testing a. Inspection Scope The inspector reviewed and observed portions of the post-maintenance testing (PMT)
associated with the following maintenance activities because of their function as mitigating systems and their potential role in increasing plant transient frequency. The inspectors reviewed the PMT documents to verify that they were in accordance with the licensees procedures and that the equipment was restored to an operable state.
B CRD Pump motor breaker maintenance ( WO R200773101). Performed procedure 617.4.001, CRD Pump Operability Test as the PMT.
B Control Room Ventilation damper inspections and 480V breaker maintenance (WOs R080718701, 02300769RC). Performed procedure 654.4.003, Control Room HVAC System B Test as the PMT.
b. Findings No findings of significance were identified.
1R22 Surveillance Testing a. Inspection Scope The inspector observed pre-test briefings and portions of the surveillance test (ST)
performance for procedural adherence, and verified that the resulting data associated with the test met the requirements of the plant technical specifications. The inspector also reviewed the results of past performances of the selected STs to verify that degraded or non-conforming conditions were identified and corrected. The following STs were observed:
Procedure 619.4.025, Reactor Protective System (RPS) Auto Scram Contactor Test.
Procedure 602.4.004, Main Steam Isolation Valve (MSIV) 10% Closure Test.
b. Findings No findings of significance were identified.
EP6 Drill Evaluation a. Inspection Scope On May 1, 2002, the inspector observed a licensed operator training assessment that included an emergency activation level classification. The inspector verified that the appropriate emergency classification was identified and external notifications to responsible parties were made in a timely manner.
b. Findings No findings of significance were identified.
4. OTHER ACTIVITIES (OA)
4OA5 Other Independent Spent Fuel Storage Installation Operation a. Inspection Scope The inspectors observed selected activities related to transferring spent fuel from the spent fuel pool to the dry cask storage system (DCSS), preparing the DCSS for storage, and moving the dry shielded canister (DSC) to the Independent Spent Fuel Storage Installation (ISFSI). The inspectors reviewed revisions made to previously approved ISFSI procedures to ensure operational commitments contained in the Safety Analysis Report, Safety Evaluation Report, and the Certificate of Compliance (CoC) No. 1004, remained in the applicable procedures.
The inspectors observed the pre-job briefings associated with selected activities such as spent fuel assembly movements, DSC sealing activities, and control of heavy loads.
The inspectors evaluated the quality of the briefings to ensure they included adequate discussions of planned activities including contingency plans and radiation safety issues.
Procedures associated with the selection and verification of fuel assemblies were reviewed to verify that fuel stored in a DSC met the conditions for canister use specified in the CoC. The inspectors observed the movement of 20 of the 61 spent fuel assemblies transferred from the spent fuel pool storage racks to the DSC to determine if the fuel moves were performed consistent with the requirements of Procedure 205.10, Fuel Assembly Removal/Insertion in Fuel Pool/Dry Shielded Canister.
Welding preparations for the DSC inner cover were observed to verify they were consistent with approved procedures. This included the monitoring of hydrogen levels and purging operations performed to ensure safe conditions existed prior to initial welding operations. The inspector observed welding operations performed on the DSC outer lid along with the performance of subsequent dye penetrant examinations. The inspector observed the final helium leak test and reviewed the results to ensure that CoC limitations for leakage were met.
The inspectors reviewed radiation work permits, the radiological hazards identified, and the controls implemented for the dry cask loading and transferring activities. The inspectors evaluated the effectiveness of health physics personnel in anticipating radiation conditions and providing appropriate guidance to the ISFSI work staff.
Additionally, cask contamination levels were reviewed to ensure compliance with the CoC requirements.
The inspector observed the transfer of the DSC to the Horizontal Storage Module (HSM)
and the insertion of the 61BT DSC into the HSM. These activities were observed to ensure they were performed in accordance with approved procedures. After the insertion of the DSC, the inspector observed the response of the local HSM temperature monitoring instrumentation, which had been installed to measure the air differential temperature between the inlet and outlet HSM vents. This review was performed to verify proper operation of the instrumentation and to ensure that the measured temperatures indicated heat removal capability consistent with the HSM design (i.e.,
measured temperatures were below expected temperatures for the calculated canister decay heat level).
b. Findings No findings of significance were identified.
4OA6 Meetings, including Exit
.1 Exit Meeting Summary On May 15, 2002, the resident inspectors presented the inspection results to Mr. Ernie Harkness and other members of licensee management. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
.2 Annual Assessment Letter Public Meeting On April 11, 2002, the NRC met with the AmerGen staff to present the conclusions associated with the NRCs Annual Assessment of Oyster Creek issued in a letter dated March 4, 2002. The meeting was open for public observation. NRC presentation was placed into ADAMS under ML020630501 and made available for public access.
ATTACHMENT 1 SUPPLEMENTAL INFORMATION a. Key Points of Contact Licensee (in alphabetical order)
V. Aggarwal, Director, Engineering R. DeGregorio, Vice President E. Harkness, Plant Manager R. Hillman, Manager, Chemistry & Radwaste J. Magee, Director, Maintenance M. Massaro, Director, Work Management D. McMillan, Director, Training M. Newcomer, Senior Manager, Design D. Slear, Manager, Regulatory Assurance C. Wilson, Senior Manager, Operations b. List of Documents Reviewed (ISFSI) Project Calculations and Engineering Analyses 6630-ADM-4010.02 Radiological Engineering Calculation Summary Procedures 205.10 Fuel Assembly removal/insertion in fuel pool/dry shielded canister, Rev. 14 205.13 Dry fuel storage monitoring, Rev. 3 6630-ADM-4110.04 Radiological work process, Rev. 8 1000-ADM-3890.01 Lifting and Rigging, Rev. 0 2400-SMM-3891.04 Operation of the Reactor Building Overhead Crane, Rev. 7 NF-OC-624 Independent Spent Fuel Storage Activities During a Plant Emergency, Rev. 0 NF-OC-626 Fuel Loading/Unloading of a DSC, Rev. 0 NF-OC-629 Transport and Preparation of Transfer Cask and 61BT Dry Shielded Canister for Loading Fuel, Rev. 1 NF-OC-630 Transport and Loading of Transport Cask and DSC, Rev. 1 NF-OC-631 Transport of Loaded Transfer Cask and 61BT Dry Shielded Canister to Transfer Trailer, Rev. 2 NF-OC-632 Dry Shielded Canister (61BT) Vacuum Drying and Helium Fill, Rev. 2 NF-OC-633 Loaded Dry Shielded Canister Welding, Rev. 1 NF-OC-634 Transportation, Alignment and Insertion of the 61BT Dry Shielded Canister into the Horizontal Storage Module, Rev. 1 NF-OC-638 Fuel Bundle Selection Process for Loading NUHOMS 61 BT DSC, Rev. 0 RM-AA-101 Management of Records, Rev. 4
Attachment 1 (contd) 8 Corrective Action Program (CAP)
02002-0528 02002-0306 02002-0544 02002-0294 02002-0226 Miscellaneous
- Final Safety Analysis Report (FSAR) for Standardized NUHOMS System, Model No. NUHOMS-61BT for Boiling Water Reactor Fuel, CoC 1004, Amendment No.
- Radiation Work Permit No. OC-1-02-00023, Rev. 01, ISFSI Project (All Areas)
c. List of Acronyms ADAMS Agencywide Documents Access and Management System AmerGen AmerGen Energy Company, LLC CAP Corrective Action Process CFR Code of Federal Regulations CoC Certificate of Compliance CRD Control Rod Drive DBT Design Basis Threat DCSS Dry Cask Storage System DSC Dry Shielded Canister FSAR Final Safety Analysis Report HSM Horizontal Storage Module HVAC Heating, Ventilation and Air Conditioning IAW In Accordance With ISFSI Independent Spent Fuel Storage Installation LORT Licensed Operator Requalification Training MSIV Main Steam Isolation Valve NRC Nuclear Regulatory Commission NUHOMS Nuclear Horizontal Modular Storage OC Oyster Creek PMT Post-Maintenance Test RBCCW Reactor Building Closed Cooling Water RPS Reactor Protective System SSCs Structures, Systems & Components ST Surveillance Testing TS Technical Specification WO Work Order