IR 05000336/2005007
| ML051590242 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/08/2005 |
| From: | Jason White NRC/RGN-I/DRS/PSB2 |
| To: | Christian D Dominion Resources |
| White J | |
| References | |
| IR-05-007 | |
| Download: ML051590242 (15) | |
Text
June 8, 2005
SUBJECT:
MILLSTONE POWER STATION UNIT 2 - NRC INSPECTION REPORT NO.
Dear Mr. Christian:
On May 18, 2005, the U.S. Nuclear Regulatory Commission completed its inspection of your reactor vessel head replacement project at your Millstone Power Station Unit 2. The enclosed inspection report documents the inspection findings, which were discussed on May 18, 2005, with Mr. Alan Price, and other members of your staff.
The inspections 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.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
John R. White, Chief Plant Support Branch 2 Division of Reactor Safety Docket No.
50-336 License No.
DPR-65 Enclosure: Inspection Report 05000336/2005007
Mr.
SUMMARY OF FINDINGS
IR 05000336/2005007; 01/10/05 - 05/18/05; Millstone Power Station Unit 2; Reactor Vessel
Head Replacement IP71007.
This inspection was conducted by two Region I inspectors. No findings of significance were identified.
NRC-Identified and Self-Revealing Findings
No findings of significance were identified.
Licensee-Identified Violations
None.
REPORT DETAILS
OTHER ACTIVITIES
[OA]
==5R01 Reactor Vessel Head Replacement Inspection (71007 - 1 Sample)
a. Inspection Scope
DESIGN==
The inspector verified that key reactor vessel head design aspects, head modifications, and the designs of other related significant modifications were reviewed and approved, by the licensee or authorized inspector, in accordance with procedures and that replacement materials and components meet the appropriate design technical requirements. The technical requirements reviewed included the applicable codes and standards, NRC requirements, and other commitments made by the licensee in the Updated Final Safety Analysis Report (UFSAR). Applicable 10 CFR 50.59 evaluations, and screening for such evaluations, for selected modifications related to head replacement, were reviewed using the guidance contained in NRC inspection procedure IP 71111.02, Evaluation of Changes, Tests, or Experiments, dated July 7, 2003.
Key design aspects and modifications for the replacement reactor vessel head and other modifications associated with reactor vessel head replacement were reviewed.
Adherence to and reconciliation of code requirements were reviewed. These reviews were performed to determine the licensee confirmed that the replacement reactor vessel head conforms to original design requirements and there were no fabrication deviations from design that were not reconciled against current requirements and regulations.
FABRICATION Fabrication in Japan. Fabrication, completed at the vendor facility in Japan, was performed under the jurisdiction of an Authorized Nuclear Inspector. The NRC inspector verified the Authorized Nuclear Inspector performed required hold-point inspections by reviewing fabrication documentation. The reports generated during the process of manufacturing the head contained the signature of the Authorized Nuclear Inspector at the appropriate times and for the appropriate reasons.
Fabrication at CONUS Vendor. The NRC inspector evaluated the ultrasonic and eddy current inspection of the control element drive mechanisms by direct observation at the vendor facility in New Hampshire. The inspector verified the examinations were conducted by qualified personnel, using qualified procedures. Activities related to this phase of the fabrication and examination were in accordance with the licensees Quality Assurance program.
The licensee had detailed procedure specifications for the control element motor housings welds, Type 316 full penetration joints and canopy seal welds of the control element drive mechanisms. The licensee implemented a procedure, reviewed by the inspector, for visually examining accessible welds throughout the fabrication and testing process. The inspector also checked the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI, Appendix VIII, qualification records for technicians performing manual ultrasonic examination of pressure housing welds in the control element drive mechanisms at the Westinghouse vendor facility. The inspector viewed the underside of the head and the J-groove welds and the numerical marking of the penetrations. The inspector observed installation of the upper pressure boundary portion of the control element drive mechanisms onto the new reactor head. The inspector also observed argon gas pressure testing of the pressure boundary control element drive mechanism penetrations after the canopy seal welds were in place. The inspector verified that proper procedural steps and sign-offs were completed during the control element drive mechanism installation. For example, that torque wrenches used on the upper pressure housings were calibrated according to the procedure. The inspector reviewed quality control field observation checklists used during installation of the control element drive mechanisms, as well as during shipment of the head from the State Pier in New London to the site warehouse building.
Fabrication at Millstone. The examinations implemented were for the purpose of establishing a preservice inspection in order to verify the integrity of the penetration base materials and each penetration weld to the reactor pressure vessel head. For these examinations, all indications were reported for evaluation regardless of size, depth or location. The examinations included the control element drive, head vent, and the incore instrumentation penetrations. The inspector assessed the inspection activities using the criteria specified in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI. The inspector directly observed the examination of penetrations 43, 45, 48, 52, 55, 63, 66 and 69. For the samples selected, the inspector verified:
A.
Procedures were qualified in accordance with the Code and had been reviewed and approved by licensee qualified nondestructive personnel.
B.
The procedures being used were supported with appropriate qualification documentation.
C.
The procedures contained sufficient instructions indicating the type of apparatus and specified the acceptable ranges of essential variables to be used.
D.
All data acquisition and analysis personnel were qualified in accordance with the Code requirements.
E.
The equipment used for the examination had been calibrated and was within the current calibration period.
F.
All indications observed were recorded.
Training and qualifications of licensee and contract quality control/assurance inspectors, and nondestructive evaluation examiners were reviewed to verify personnel met site and code qualification requirements and were adequately prepared for the site specific tasks.
For the selected welds, weld procedures and welder qualification records were reviewed to confirm that the Code required essential and supplemental essential welding variables for the welding processes used were met. Non-conformance reports were reviewed to confirm that welding deficiencies were dispositioned in accordance with Code requirements.
LIFTING Engineering design, modification, and analysis associated with reactor vessel (RV) head lifting and rigging including:
- (1) crane, and rigging equipment, and full load testing,
- (2) RV head component drop analysis,
- (3) safe load paths, and
- (4) lay-down areas, were reviewed prior to any head movement. Because the reactor was completely defueled prior to load handling activities, there was no impact on the reactor core or spent fuel nor on cooling and plant support systems for the reactor unit and common systems for the other operating unit at the site. The inspector verified there were no design changes and modifications to systems, structures, and components described in the FSAR for transporting the new and old reactor vessel heads in and out of the plant respectively.
Activities associated with lifting and rigging were reviewed including: preparations and procedures for rigging and heavy lifting including any required crane and rigging inspections, testing, equipment modifications, lay-down area preparations, and training of crane and rigging personnel. The inspector verified the capability of the lifting equipment, including fixtures and rigging, to handle the load were established by analysis.
The licensee's plans and analysis for lifting and rigging of heavy loads were reviewed to verify that lifting equipment and rigging are capable of lifting and moving the reactor head, and the safe load path for component removal and reinstallation was technically sound. The component drop analysis was reviewed to verify, in general, that the potential offsite releases at the exclusion area boundary were within 10 CFR Part 100 limits and equipment to maintain safe shutdown was unaffected.
RADIATION PROTECTION The radiation protection program controls, planning, and preparation was reviewed in the following areas utilizing applicable portions of the baseline inspection procedures IP 71121.01 and 71121.02 using as guidance the following parameters:
G.
As Low As Reasonably Achievable (ALARA) planning.
H.
Job dose estimates and dose tracking.
I.
Exposure controls including temporary shielding.
J.
Airborne and Contamination controls.
K.
Radioactive material controls and management.
L.
Radiological work plans and controls.
M.
Emergency contingencies.
N.
Project staffing and training plans.
Radiological source term was evaluated. The presence of hard-to-detect radionuclides including transuranics was specifically reviewed.
SECURITY Security considerations associated with vital and protected area barriers that were affected during replacement activities were reviewed by qualified NRC security inspection personnel. Security inspections were performed to assure that intended breaches in the perimeter protection, made for the purpose of head movement, were adequately compensated for.
POST-INSTALLATION TESTING Post-installation verification and testing inspections were reviewed in the following areas:
A.
The licensee's post-installation inspections and verifications program and its implementation.
B.
The conduct of reactor coolant system leakage testing and review the test results.
C.
The procedures for equipment performance testing required to confirm the design and to establish baseline measurements and the conduct of testing.
D.
Compliance with regulatory requirements including the incorporation of inservice inspection requirements of 10 CFR 50.55a (g).
E.
Adherence to and reconciliation of Code requirements.
The inspector reviewed the licensee's post-installation verification and testing program to verify that modifications were completed in accordance with the design; that drawings, procedures, and training were already updated or where immediate attention was not required plans were in place to update as appropriate; that post-installation walkdowns and inspections were performed to ensure equipment was restored and temporary services were removed; that equipment cleanliness was verified; that pre-service inspection of welds to establish baseline data was performed; and that deficiencies were properly dispositioned. The inspector also verified that changes in performance of the reactor vessel head and its associated parameters, such as flow rates, pressures, and temperatures were appropriately included in design documents and plant procedures.
The inspector verified testing satisfied the Code and applicable regulatory requirements, that testing was conducted according to procedures, and that results were satisfactory or properly resolved.
STORAGE The inspector reviewed radiological safety plans for storage of the old reactor vessel head on site. The storage facility, consisting of a one inch thick steel vessel placed over the reactor head, was reviewed to verify that access was properly controlled, did not create the potential for an unmonitored effluent release pathway, and that external radiation levels at the perimeter would be below applicable limits, and dose rates at the perimeter were below applicable limits. The storage of the old reactor vessel head met the criteria contained in Generic Letter 81-38, "Storage of Low-Level Radioactive Wastes at Power Reactor Sites."
The old reactor head is stored on-site in a one inch thick container. The old head is located next to Building 100. Assuming a point source of radiation consisting of 5 Ci of Co-60 the total exposure at 1700 feet is less then 1.1E-4 mR/hour. Except for a small portion of the bay, which is not contained in a restricted zone, a circle with a radius of 1700 feet is contained within the owner controlled area. Occasional fisherman would conservatively receive an exposure far less than the 25 mrem/year limit of 40 CFR 190.
b. Findings
No findings of significance were identified.
4OA6 Meetings, including Exit
On May 18, 2005, the inspectors presented inspection results to Mr. Michael Wilson and other members of the Millstone staff during a final exit meeting. No proprietary information was reviewed during the inspection.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- A. Price, Site Vice-President
- M. Doucette, Lead Engineer
- R. Schriener, Engineering
- T. Petit, Project Manager
- R. McIntosh, Licensing
- D. Dodson, Licensing Supervisor
- L. Picarazzi, Scheduling
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None.
Closed
None.