IR 05000261/2005004
| ML053040012 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 10/28/2005 |
| From: | Fredrickson P NRC/RGN-II/DRP/RPB4 |
| To: | Moyer J Carolina Power & Light Co |
| References | |
| IR-05-004 | |
| Download: ML053040012 (35) | |
Text
October 28, 2005
SUBJECT:
H.B. ROBINSON STEAM ELECTRIC PLANT - NRC INTEGRATED INSPECTION REPORT 05000261/2005004
Dear Mr. Moyer:
On September 30, the US Nuclear Regulatory Commission (NRC) completed an inspection at your H.B. Robinson reactor facility. The enclosed integrated inspection report documents the inspection findings, which were discussed on October 12, 2005, with C. Church, E. Kapopoulos, J. Lucas and other members of your staff, and with you on October 27, 2005.
The 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.
This report documents one NRC-identified finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements. However, because of its very low safety significance and because it had been entered into your corrective action program, the NRC is treating this issue as a non-cited violation, in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you contest this non-cited violation, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Robinson facility.
In accordance with 10 CFR 2.390 of the NRC's "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 NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Paul E. Fredrickson, Chief Reactor Projects Branch 4 Division of Reactor Projects Docket No.:
50-261 License No.:
Enclosure:
Inspection Report 05000261/2005004 w/Attachment: Supplemental Information
REGION II==
Docket No:
50-261 License No:
DPR-23 Report No:
05000261/2005004 Facility:
H. B. Robinson Steam Electric Plant, Unit 2 Location:
3581 West Entrance Road Hartsville, SC 29550 Dates:
July 1, 2005 - September 30, 2005 Inspectors:
R. Hagar, Senior Resident Inspector D. Jones, Resident Inspector M. Scott, Senior Reactor Inspector (Section 1R12.2)
B. Crowley, Senior Reactor Inspector, Consultant (Sections 4OA5.3, 4OA5.4, & 4OA5.5)
J. Rivera-Ortiz, Reactor Inspector (Section 4OA5.3, 4OA5.4, &
4OA5.5)
R. Chou, Reactor Inspector (Section 4OA5.2)
H. Gepford, Health Physicist (Section 2OS3)
Approved by:
P. Fredrickson, Chief Reactor Projects Branch 4 Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000261/2005-004, 07/01/2005-09/30/2005; H.B. Robinson Steam Electric Plant, Unit 2;
Other Activities.
The report covered a three-month period of inspection by resident inspectors and announced inspections by two Senior Reactor Inspectors, two Reactor Inspectors, and a Health Physicist.
One Green non-cited violation (NCV) was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649,
Reactor Oversight Process, Revision 3, dated July 2000.
NRC-Identified and Self-Revealing Findings
Cornerstone: Barrier Integrity
- Green.
The inspectors identified a green non-cited violation of 10 CFR 50, Appendix B,
Criterion V for two procedures which included instructions for restoring reactor coolant pump seal cooling but did not include any requirement or precaution regarding the time at which seal cooling is restored, even though information provided by the Westinghouse Owners Group indicated that restoration of RCP seal cooling was time-critical.
This finding was more than minor because it affected the procedure quality attribute of the Barrier Integrity Cornerstone objective of providing reasonable assurance that the reactor coolant system protects the public from radionuclide releases caused by accidents or events. The finding was evaluated using Appendix A to Manual Chapter 0609, Significance Determination Process. Because the finding affects a Barrier Integrity Cornerstone objective, the Phase 1 worksheet requires a Phase 3 risk evaluation be completed. A Phase 3 screening analysis was conducted and determined that because of the low likelihood of a station blackout, and the probable recovery of an offsite or onsite alternating-current power source prior to core damage, the finding was determined to be of very low safety significance. (Section 4OA5.6)
Licensee-Identified Violations
None
REPORT DETAILS
Summary of Plant Status The unit began the inspection period at full rated thermal power. On September 17, the unit was shut down for a refueling outage. That refueling outage extended through the end of the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
==1R04 Equipment Alignment
a. Inspection Scope
Partial System Walkdowns==
The inspectors performed the following three partial system walkdowns, while the indicated structures, systems, and/or components (SSCs) were out-of-service for maintenance and testing:
System Walked Down SSC Out of Service Date Inspected A emergency diesel generator B emergency diesel generator August 2 Service water train A D service water pump August 24 B emergency diesel generator A emergency diesel generator September 6 To evaluate the operability of the selected trains or systems under these conditions, the inspectors compared observed positions of valves, switches, and electrical power breakers to the procedures and drawings listed in the Attachment.
Complete System Walkdown The inspectors conducted a detailed review of the alignment and condition of the B motor-driven train of the auxiliary feedwater system to verify that the existing alignment of the system was consistent with the correct alignment. To determine the correct system alignment, the inspectors reviewed the procedures, drawings, and the Updated Final Safety Analysis Report (UFSAR) section listed in the Attachment. The inspectors also walked down the system. During the walkdown, the inspectors reviewed the following:
- Valves were correctly positioned and did not exhibit leakage that would impact the functions of any given valve.
- Electrical power was available as required.
- Major system components were correctly labeled, lubricated, cooled, ventilated, etc.
- Hangers and supports were correctly installed and functional.
- Essential support systems were operational.
- Ancillary equipment or debris did not interfere with system performance.
- Tagging clearances were appropriate.
- Valves were locked as required by the locked valve program.
The inspectors reviewed the documents listed in the Attachment to verify that the ability of the system to perform its functions could not be affected by outstanding design issues, temporary modifications, operator workarounds, adverse conditions, and other system-related issues tracked by the engineering department.
b. Findings
No findings of significance were identified.
==1R05 Fire Protection
a. Inspection Scope
==
For the six areas identified below, the inspectors reviewed the control of transient combustible material and ignition sources, fire detection and suppression capabilities, fire barriers, and any related compensatory measures to verify that those items were consistent with UFSAR Section 9.5.1, Fire Protection System, and UFSAR Appendix 9.5.A, Fire Hazards
Analysis.
The inspectors walked down accessible portions of each area and reviewed results from related surveillance tests to verify that conditions in these areas were consistent with descriptions of the areas in the UFSAR. Documents reviewed are listed in the Attachment.
The following areas were inspected:
Fire Zone Description 25A/B Turbine building east and west ground floor
[Heating, ventilation, and air conditioning] equipment for control room 25F/G Turbine building east/west mezzanine and operating deck 25D Dedicated shutdown diesel generator Battery room Cable spreading room
b. Findings
No findings of significance were identified.
==1R06 Flood Protection Measures
a. Inspection Scope
Internal Flooding==
Because the safety injection pump room contains risk-significant SSCs which are susceptible to flooding from postulated pipe breaks, the inspectors walked down that room to verify that the area configuration, features, and equipment functions were consistent with the descriptions and assumptions used in Calculation RNP-F/PSA-0009, Assessment of Internally Initiated Flooding Events and in the supporting basis documents listed in the Attachment. The inspectors reviewed the operator actions credited in the analysis to verify that the desired results could be achieved using the plant procedures listed in the Attachment.
b. Findings
No findings of significance were identified.
==1R11 Licensed Operator Requalification
a. Inspection Scope
==
The inspectors observed licensed-operator performance during requalification simulator training for crew 2 to verify that operator performance was consistent with expected operator performance, as described in the Continuing Training Simulator Option form dated 6/23/05. This training tested the operators ability to respond to a loss of reactor coolant system inventory and a subsequent loss-of-coolant accident during shutdown conditions. The inspectors focused on clarity and formality of communication, the use of procedures, alarm response, control board manipulations, group dynamics, and supervisory oversight.
The inspectors observed the post-exercise critique to verify that the licensee identified deficiencies and discrepancies that occurred during the simulator training.
b. Findings
No findings of significance were identified.
==1R12 Maintenance Effectiveness
==
.1 Routine Inspection of Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the two degraded SSC/function performance problems or conditions listed below to verify the appropriate handling of these performance problems or conditions in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, and 10 CFR 50.65, Maintenance Rule. Documents reviewed are listed in the
.
The problems/conditions and their corresponding action requests (ARs) were:
Performance Problem/Condition AR Containment isolation valve V12-11 failed to open 143554 Letdown orifice isolation valve regulator setpoint changed 135101 During the reviews, the inspectors focused on the following:
- Appropriate work practices.
- Identifying and addressing common cause failures.
- Scoping in accordance with 10 CFR 50.65(b).
- Characterizing reliability issues (performance).
- Charging unavailability (performance).
- Trending key parameters (condition monitoring).
- 10 CFR 50.65(a)(1) or (a)(2) classification and reclassification.
- Appropriateness of performance criteria for SSCs/functions classified (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified (a)(1).
b. Findings
No findings of significance were identified.
.2 Periodic Evaluation (Biennial)
a. Inspection Scope
The inspectors reviewed the licensees Maintenance Rule (MR) periodic assessment, Self-Assessment Report 147347, dates of assessment May 2-5, 2005. This report was issued to satisfy paragraph (a)(3) of 10 CFR 50.65, and covered the 18 month period of October 1, 2003 to March 31, 2005. The inspection was to determine the effectiveness of the assessment and that it was issued in accordance with the time requirement of the MR and included evaluation of: balancing reliability and unavailability, (a)(1) activities, (a)(2) activities, and use of industry operating experience. To verify compliance with 10 CFR 50.65, the inspectors reviewed selected MR activities covered by the assessment period for the following maintenance rule components and systems: radiation monitors, reactor protection system, Regulatory Guide 1.97 instrumentation, switchyard components, and auxiliary feedwater system. Additionally, the inspectors reviewed a section of the partially completed structural inspection report (containment building) and inspected select plant structures. Specific procedures and documents reviewed are listed in the Attachment.
During the inspection, the inspectors reviewed selected plant work order data, assessments, modifications, the site guidance implementing procedures, discussed and reviewed relevant corrective action issues, reviewed generic operations event data, attendant MR related meeting minutes, probabilistic risk reports, and discussed issues with system engineers. Operational event information was evaluated by the inspectors in its use in MR functions. The inspectors selected work orders and other corrective action documents on systems recently removed from 10 CFR 50.65 a(1) status and those in a(2) status for some period to assess the justification for their status. The inspectors toured and inspected repaired component locations. The documents were compared to the sites MR program criteria, and the MR a(1) evaluations and rule related data bases.
b. Findings
No findings of significance were identified.
==1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
==
For the four time periods listed below, the inspectors reviewed risk assessments and related activities to verify that the licensee performed adequate risk assessments and implemented appropriate risk-management actions when required by 10 CFR 50.65(a)(4). For emergent work, the inspectors also verified that any increase in risk was promptly assessed, and that appropriate risk-management actions were promptly implemented. Documents reviewed are listed in the Attachment. Those periods included the following:
- The work week of July 30 - August 5, including emergent work which rendered the B emergency diesel generator unavailable
- The work week of August 6 - August 12, including emergent work which included unavailability of the steam-driven auxiliary feedwater pump
- The work week of August 13 - August 19, including emergent work which included unavailability of the C safety injection pump
- The work week of August 26 - September 2, including emergent work which included unavailability of the A emergency diesel generator
b. Findings
No findings of significance were identified.
==1R15 Operability Evaluations
a. Inspection Scope
==
The inspectors reviewed the operability determination associated with AR 164063164063 This AR addressed the operability of the B auxiliary feedwater pump when an associated flow indicating controller was degraded. The inspectors assessed the accuracy of the evaluation, the use and control of any necessary compensatory measures, and compliance with the Technical Specifications (TS). The inspectors verified that the operability determination was made as specified by procedure PLP-102, Operability Determinations. The inspectors compared the justifications provided in the determination to the requirements from the TS, the UFSAR, and associated design-basis documents, to verify that operability was properly justified and the auxiliary feedwater system remained available, such that no unrecognized increase in risk occurred.
Documents reviewed are listed in the Attachment.
b. Findings
No findings of significance were identified.
==1R19 Post-Maintenance Testing
a. Inspection Scope
==
For the six post-maintenance tests listed below, the inspectors witnessed the test and/or reviewed the test data to verify that test results adequately demonstrated restoration of the affected safety functions described in the UFSAR and TS. Documents reviewed are listed in the Attachment.
The following tests were witnessed/reviewed:
Test Procedure Title Related Maintenance Activity Date Inspected OST-603 Motor Driven Fire Water and Engine Driven Fire Water Pump Test (Weekly)
Routine maintenance on the engine driven fire pump July 11 OST-101-1
[Chemical and Volume Control System]
Component Test, Charging Pump A Repair the start/stop switch and calibrate the fluid drive oil pressure gauge July 26 OST-401-2
[Emergency Diesel Generator] B Slow Speed Start Replace air start solenoid valve, DA-19B August 2 OST-302-2 Service Water Pumps C and D Inservice Test Breaker replacement and electrical testing August 24 OP-604 Diesel Generators A and B Replace coil for air start solenoid valve, DA-19A September 6 OST-636 Flow Test for [Reactor Coolant Pump] B Pre-Action Sprinkler System (Refueling)
Replace pressure switch PS-7008 September 22 The inspectors reviewed the following ARs associated with this area to verify that the licensee identified and implemented appropriate corrective actions:
Pump A
- AR 160935160935 [Post Maintenance Test] Requirement Not Identified
b. Findings
No findings of significance were identified.
==1R20 Refueling and Outage Activities
a. Inspection Scope
==
For the outage that began on September 17, the inspectors evaluated licensee outage activities as described below to verify that licensee considered risk in developing outage schedules, adhered to administrative risk reduction methodologies they developed to control plant configuration, and adhered to operating license and technical specification requirements that maintained defense-in-depth. The inspectors also verified that the licensee developed mitigation strategies for losses of the following key safety functions:
- inventory control
- power availability
- reactivity control
- containment Documents reviewed are listed in the Attachment.
Review of Outage Plan Prior to the outage, the inspectors reviewed the outage risk control plan to verify that the licensee had performed adequate risk assessments, and had implemented appropriate risk-management strategies when required by 10 CFR 50.65(a)(4).
Monitoring of Shutdown Activities
The inspectors observed portions of the plant shutdown and cooldown process to verify that technical specification cooldown restrictions were followed.
Licensee Control of Outage Activities Periodically during the outage, the inspectors observed the items or activities described below to verify that the licensee maintained defense-in-depth commensurate with the outage risk-control plan for key safety functions and applicable technical specifications when taking equipment out of service.
- Clearance Activities
- Reactor Coolant System Instrumentation
- Electrical Power
- Inventory Control
- Reactivity Control
- Containment Closure The inspectors also reviewed responses to emergent work and unexpected conditions to verify that resulting configuration changes were controlled in accordance with the outage risk control plan, and to verify that control-room operators were kept cognizant of the plant configuration.
Refueling Activities The inspectors observed fuel handling operations (removal) and related activities to verify that those operations and activities were being performed in accordance with technical specifications and approved procedures. Also, the inspectors verified that the locations of the fuel assemblies were tracked during core offload.
b. Findings
No findings of significance were identified.
==1R22 Surveillance Testing
a. Inspection Scope
==
For the six surveillance tests listed below, the inspectors witnessed testing and/or reviewed the test data to verify that the systems, structures, and components involved in these tests satisfied the requirements described in the TS, the UFSAR, and applicable licensee procedures, and that the tests demonstrated that the SSCs were capable of performing their intended safety functions. Documents reviewed are listed in the
.
Test Procedure Title Date Inspected OST-251-1*
[Residual Heat Removal] Pump A and Components Test July 14 OST-151-3 Safety Injection System Components Test -
Pump C July 28 OST-151-1 Safety Injection System Components Test -
Pump A August 3 OST-202 Steam Driven Auxiliary Feedwater System Component Test August 8 OST-409-2
[Emergency Diesel Generator] B Fast Speed Start August 16 OST-154 Safety Injection System High Head Check Valve Test September 27
- This procedure included inservice testing requirements.
The inspectors reviewed AR 165893165893 Steam Driven [Auxiliary Feedwater] Pump Trip to verify that the licensee identified and implemented appropriate corrective actions.
b. Findings
No findings of significance were identified.
==1R23 Temporary Plant Modifications
a. Inspection Scope
==
The inspectors reviewed the temporary modification described in Engineering Change 60359, Temporary Connection of [Radiation Monitor] 14 - Control Room Alarm Circuits During Engineering Change 52464, to verify that the modification did not affect the safety functions of important safety systems, and to verify that the modification satisfied the requirements of procedure EGR-NGGC-005, Engineering Change, and 10 CFR 50, Appendix B, Criterion III, Design Control. Documents reviewed are listed in the
.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS3 Radiation Monitoring Instrumentation and Protective Equipment
a. Inspection Scope
Portable Radiation Monitoring Instrumentation During the week of August 1, 2005, the inspectors evaluated completion and adequacy of radiation survey instrument calibrations performed by the licensees central calibration facility located at the Shearon Harris Nuclear Plant. Availability of portable instruments for licensee use was evaluated through discussion with licensee personnel regarding inventory, logistics, and transfer/receipt of instruments. Calibration data for portable instruments staged or recently used for coverage of field tasks were reviewed. Records associated with the annual certifications of the gamma irradiator and neutron source used for performing calibrations and routine response checks were reviewed in detail.
In addition, the inspectors observed the calibration facility for neutron instrument calibrations and discussed its adequacy for performing instrument calibrations with cognizant licensee personnel. The inspectors discussed techniques and technical bases applied to the calibration of portable survey instruments, including the use of a 25% grace period, with licensee personnel. Two corrective action program (CAP)nuclear condition documents associated with the instrument calibration activities were reviewed and discussed with responsible licensee representatives.
Final Safety Analysis Report (FSAR) Chapter 12; ANSI N323-1978, Radiation Protection Instrumentation Test and Calibration; and applicable licensee procedures. The licensees ability to characterize, prioritize, and resolve identified CAP issues was reviewed against CAP-0200, Corrective Action Program, Rev. 14 and associated guideline documents.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
.1 Routine Review of ARs
To aid in the identification of repetitive equipment failures or specific human performance issues for followup, the inspectors performed frequent screenings of items entered into the CAP. The review was accomplished by reviewing daily AR reports.
Documents reviewed are listed in the Attachment.
.2 Annual Sample Review
a. Inspection Scope
The inspectors selected AR 158738158738 [Self Assessment] 147347 Issue #1: Repetitive Functional Failures, for detailed review. The inspectors selected this AR because it relates generally to the Mitigating Systems Cornerstone, and involved failures of steam dump valves and spurious actuations of reactor protection system components. The inspectors reviewed this report to verify:
- complete and accurate identification of the problem in a timely manner;
- evaluation and disposition of performance issues;
- evaluation and disposition of operability and reportability issues;
- consideration of extent of condition, generic implications, common cause, and previous occurrences;
- appropriate classification and prioritization of the problem;
- identification of root and contributing causes of the problem;
- identification of corrective actions which were appropriately focused to correct the problem; and
- completion of corrective actions in a timely manner.
The inspectors also reviewed this AR to verify compliance with the requirements of the CAP as delineated in Procedure CAP-NGGC-0200, Corrective Action Program, and 10 CFR 50, Appendix B. Documents reviewed are listed in the Attachment.
b. Observations and Findings
No findings of significance were identified.
4OA5 Other Activities
.1 Operation of an Independent Spent Fuel Storage Installation (ISFSI) (IP 60855.1)
a. Inspection Scope
During the first fuel from the storage pool into the new ISFSI, the inspectors observed selected activities associated with inspecting spent fuel; record-keeping; loading spent fuel into storage canisters; and canister welding, vacuum-drying, transport, and insertion into the storage location, to verify that the activities were performed in a safe manner and in accordance with approved procedures. Documents reviewed are listed in the Attachment.
b. Findings
No findings of significance were identified.
.2 Review of Reactor Vessel Closure Head Replacement Project (RVCHRP) Lifting and
Transportation Program Activities
a. Inspection Scope
The inspectors reviewed the adequacy of the RVCHRP lifting program as described in Engineering Change Package 56266, Rev. 2, Reactor Closure Head & Service Structure Replacement, to assure that it was prepared in accordance with regulatory requirements, appropriate industrial codes and standards, and to verify that the maximum anticipated loads to be lifted would not exceed the capacity of the lifting equipment and supporting structures.
The inspectors reviewed and partially examined the RVCHRP lifting equipment including a crane, a down/up-ender, a head lifting tripod and rigs, runway skid systems, load paths, and a transporter, for material condition and adequacy. In addition, the inspectors reviewed the adequacy of the transport programs, procedures, work packages and load test records, to assure that they had been prepared and tested in accordance with regulatory requirements, appropriate industrial codes, and standards.
The inspectors also reviewed the adequacy of the licensee's analyses and calculations for handling loads during the rigging and lifting, the load path analysis, polar crane maintenance records, and lifting and rigging drawings.
b. Findings
No findings of significance were identified.
.3 Reactor Pressure Vessel Head (RPVH) Replacement (IP 71007)
a. Inspection Scope
The inspectors observed/reviewed the activities detailed below for the replacement RPVH to verify compliance with applicable construction and inspection Codes (ASME Boiler and Pressure Vessel Code,Section III, 1998 Edition through 2000 Addenda and Section XI, 1995 Edition through 1996 Addenda) as defined in the Engineering Change (EC) document EC 56266R1, Reactor Head and Service Structure Replacement.
RPVH and Control Rod Drive Mechanism (CRDM) Housing Fabrication Records The inspectors reviewed fabrication records for the RPVH and the CRDM housings including Certified Material Test Reports, Non-Destructive Examination (NDE) reports, hydrostatic testing and dimensional examinations to verify compliance with the applicable construction and inspection codes. The records reviewed are listed in the
. For the following welds, the inspectors reviewed: fabrication process sheets, welding work record sheets, fitup inspection records, PT examination records, RT examination records, and UT examination records, as applicable, to verify compliance with the applicable construction and inspection codes. Records reviewed for the selected welds are listed in the Attachment.
- RPVH J-groove Butter welds WO-S107-1A, 8A, 18A, 25A, and 35A
- RPVH J-groove welds WC-S109-1A, 6A, 16A, 28A, and 62A
- RPVH Clad welds WO-S103-1, and -2
- CRDM Latch Housing to RPVH Adapter welds WC-L202-1A, 14A, 24A, 34A, 62A, and 65A Preservice Inspection (PSI) and Baseline Inspections The inspectors reviewed selected NDE records, which documented the ASME Section XI PSI and baseline inspections performed to provide baseline conditions for future inspections in accordance with NRC Order EA-03-09.
Relative to ASME Section XI PSI of the replacement RPVH, the inspectors reviewed the completed PT records of :
- 28 peripheral Category B-O CRDM Latch Housing to RPVH Adapter welds In addition, the inspectors reviewed personnel certification records of two
In order to support future inspections required by NRC Order EA-03-09, the baseline NDE inspections consisted of:
- Automated open bore inside diameter UT and eddy current (ET) examination of the CRDM penetrations covering from the bottom of the penetrations to a minimum of 2" above the J-groove welds
- J-groove surface ET examination and inside diameter UT and ET examination of the Reactor Vessel Level Indication System (RVLIS) line and vent nozzle penetrations
- ET examination of the outside diameter of the penetrations below the J-groove welds and the surface of the J-groove welds
- Under head PT inspection of all penetration to head J-groove welds using PT white acceptance criteria (performed to obtain the best surface possible to prevent future primary stress corrosion cracking)
The inspectors observed in-process NDE baseline examinations and reviewed the results of a sample UT and ET data. Specifically, the inspectors conducted the following activities:
- Partial observation of open bore scanning UT/ET examination for penetration No. 62
- Partial observation of outside diameter ET examination of penetration No. 37
- Partial observation of ET examination on J-groove weld surface for penetration Nos.
29 and 44
- Review of completed UT and ET reports for penetration Nos. 7, 39, and 63, including saved computer data
- Review of under head PT examination reports of all penetration to head J-groove welds using PT white acceptance criteria, including personnel qualifications and PT material certifications
- Visual inspection on top of the RPVH, specifically penetration Nos. 62, 38, 30, 22, 14, and 26
b. Findings
No findings of significance were identified.
.4 Review of 10 CFR 50.59 Screening/Evaluation for the Replacement RPVH
a. Inspection Scope
The inspectors reviewed EC 56266, Reactor Head and Service Structure Replacement, Rev. 1, including the associated 10 CFR 50.59 screening to verify that changes between the original RPVH and the replacement RPVH, and modifications resulting from installation of the replacement RPVH were properly evaluated in accordance with 10 CFR 50.59. Specifically, the inspectors reviewed the impact of the replacement RPVH weight on the reactor vessel supports and seismic analysis. The inspectors verified that the weight of the replacement RPVH assembly, as described in EC 56266R1, was bounded within the safety margin of the design stress calculations for static and seismic loads.
b. Findings
No findings of significance were identified.
.5 Review of Quality Assurance (QA) Activities for the fabrication of the RPVH
a. Inspection Scope
The inspectors reviewed surveillance reports of licensee QA activities at the vendor facilities to verify that the licensee evaluated the QA activities performed by the fabricator (Mitsubishi Heavy Industries, MHI) and the contractor for RPVH fabrication (Westinghouse) at the MHI facilities. The inspectors performed a review to verify that the licensee conducted QA surveillance such as:
- (1) independent review of fabricator procedures and fabrication records,
- (2) independent verification of RPVH dimensions,
- (3) witnessing NDE examinations, material testing, and manufacturing processes, (4)verification of fabricator personnel knowledge through interviews,
- (5) verification that non-conformance conditions and deviations were identified and dispositioned, and (6)assessment of Westinghouse QA surveillance on MHI QA program.
b. Findings
No findings of significance were identified.
.6 (Closed) URI 05000261/2005003-01:
Failure of Two Procedures to Have Appropriate Acceptance Criteria for Restoration of Reactor Coolant Pump (RCP) Seal Cooling In NRC Inspection Report 05000261/2005003, a URI was identified involving two examples of a violation of 10 CFR 50, Appendix B, Criterion V, Procedures, for failure to include appropriate acceptance criteria in two procedures for restoration of cooling to the reactor coolant pump seals following a loss of all seal cooling. The finding was unresolved because it had potential safety significance greater than very low significance and required the completion of a significance determination process Phase 3 review, prior to finalizing the findings significance.
As discussed in Section 1R17 of Inspection Report 05000261/2005003, the inspectors identified that Procedures EPP-22, Revision 20, Energizing Plant Equipment Using Dedicated Shutdown Diesel Generator, and DSP-002, Revision 30, Hot Shutdown Using the Dedicated/Alternate Shutdown System, both included instructions for restoring RCP seal cooling but did not include any requirement or precaution regarding the time at which RCP seal cooling is restored, even though information provided by the Westinghouse Owners Group indicated that restoration of RCP seal cooling was time-critical.
During this inspection period, a Phase 3 screening analysis was completed by using a bounding calculation to determine the potential limits of the risk associated with the finding. That analysis included consideration of results from licensee-completed thermal hydraulic calculations which determined the time to core uncovery following a loss-of-coolant accident from the RCP seals induced by restoring RCP seal cooling later than would be advisable. Because of the low llikelihood of a station blackout, and the probable recovery of an offsite or onsite alternating-current power source prior to core damage, the finding was determined to be of very low safety significance (GREEN).
10 CFR 50, Appendix B, Criterion V, Procedures, requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances. It further requires that these procedures include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, Procedures EPP-22, Rev. 20 and DSP-002, Rev. 30, do not include appropriate acceptance criteria for restoration of cooling to the RCP seals following a loss of all seal cooling. However, because of the very low safety significance and because this issue was entered into the corrective action program (AR 160357160357, this finding is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy, and has been designated NCV 05000261/2005004-01, Failure of Two Procedures to Have Appropriate Acceptance Criteria for Restoration of RCP Seal Cooling. URI 05000261/2005003-01 is closed.
.7 Operational Readiness of Offsite Power (Temporary Instruction (TI) 2515/163)
Completion of this TI was documented in NRC Inspection Report 05000261/2005003.
However, after an NRC headquarters review of the data provided, additional information related to the TI was requested. The inspectors collected this information from licensee discussions, site procedures and licensee documentation. The information was subsequently provided to the headquarters staff for further analysis.
4OA6 Meetings, Including Exit
On October 12, the resident inspectors presented the inspection results to C. Church, E.
Kapopoulos, J. Lucas, and other members of the Robinson staff. In addition a supplemental exit was conducted with J. Moyer on October 27, 2005. The inspectors confirmed that proprietary information was not provided or examined during the inspection.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- M. Blew, Inservice Inspection Coordinator
- E. Caba, Engineering Superintendent
- C. Castell, Licensing
- A. Cheatham, Radiation Protection Superintendent
- C. Church, Engineering Manager
- B. Clark, Nuclear Assurance Manager
- R. Cline, Non-Destructive Examination Level III Examiner
- D. Etheridge, Lead Engineer RPVH Replacement Project
- W. Farmer, Engineering Superintendent
- J. Huegel, Maintenance Manager
- R. Ivey, Operations Manager
- E. Kapopoulos, Outage and Scheduling Manager
- J. Lucas, Manager, Support Services - Nuclear
- G. Ludlum, Training Manager
- J. Moyer, Vice President, Robinson Nuclear Plant
- W. Noll, Director of Site Operations
- D. Stoddard, Plant General Manager
- V. Wagoner, Reactor Head Replacement Project Manager
- S. Wheeler, Supervisor, Regulatory Support
Contractor Personnel
- R. Driscoll, Westinghouse Principal Engineer
- P. Lancaster, Westdyne Non-Destructive Examination Level III Examiner
- R. Vestovich, Westdyne Project Manager
NRC personnel
- P. Fredrickson, Chief, Reactor Projects Branch 4
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
NONE
Opened and Closed
- 05000261/2005004-01 NCV Failure of Two Procedures to Have Appropriate Acceptance Criteria for Restoration of RCP Seal Cooling (Section 4OA5.6)
Closed
- 05000261/2005003-01 URI Failure of Two Procedures to Have Appropriate Acceptance Criteria for Restoration of Reactor Coolant Pump (RCP) Seal Cooling(Section 4OA5.6)
Discussed
NONE