ML043630019
| ML043630019 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/06/2005 |
| From: | Margie Kotzalas NRC/NRR/DLPM/LPD3 |
| To: | Thomas J. Palmisano Nuclear Management Co |
| Padovan L, NRR/DLPM, 415-1423 | |
| References | |
| TAC MC2222 | |
| Download: ML043630019 (9) | |
Text
January 6, 2005 Mr. Thomas J. Palmisano Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL REQUEST FOR RELIEF NO. 4 (TAC NO. MC2222)
Dear Mr. Palmisano:
The Nuclear Management Company, LLCs (NMCs) letter of February 16, 2004, as supplemented August 18, 2004, proposed an alternative to the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) at Monticello Nuclear Generating Plant (MNGP). Table IWB-2500-1, Category B-K, Item B10.10, of Section XI of the ASME Code requires NMC to perform a surface examination of the reactor pressure vessel stabilizer bracket welds each inspection interval. Instead, NMC proposed performing a surface examination on the stabilizer brackets only if jet reaction forces or seismic design loads are experienced. In addition, NMC proposed performing a one-time visual examination on the accessible surfaces of each bracket attachment weld and adjacent areas during the current inservice inspection (ISI) interval.
The U.S. Nuclear Regulatory Commission (NRC) staff evaluated NMCs request and concludes that compliance with the ASME Code-required examinations would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff further finds that the proposed alternative provides reasonable assurance of the structural integrity of the component. Therefore, the NRC staff authorizes NMCs proposed alternative pursuant to 10 CFR 50.55a(a)(3)(ii) for the fourth 10-year ISI interval at MNGP. Enclosed is our safety evaluation.
Sincerely,
/RA/
Margie Kotzalas, Acting Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
Safety Evaluation cc w/encl: See next page
ML043630019 OFFICE PDIII-1/PM PDIII-1/LA EMCB/SC OGC PDIII-1/(A)SC NAME LPadovan THarris SCoffin MHiggins(nlo)
MKotzalas DATE 12/30/04 12/30/04 11/30/04 1/3/05 1/6/05
Monticello Nuclear Generating Plant cc:
Jonathan Rogoff, Esquire Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 U.S. Nuclear Regulatory Commission Resident Inspector's Office 2807 W. County Road 75 Monticello, MN 55362 Manager, Regulatory Affairs Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637 Robert Nelson, President Minnesota Environmental Control Citizens Association (MECCA) 1051 South McKnight Road St. Paul, MN 55119 Commissioner Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, MN 55155-4194 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Commissioner Minnesota Department of Health 717 Delaware Street, S. E.
Minneapolis, MN 55440 Douglas M. Gruber, Auditor/Treasurer Wright County Government Center 10 NW Second Street Buffalo, MN 55313 Commissioner Minnesota Department of Commerce 85 7th Place East, Suite 500 St. Paul, MN 55101-2198 Manager - Environmental Protection Division Minnesota Attorney Generals Office 445 Minnesota St., Suite 900 St. Paul, MN 55101-2127 John Paul Cowan Executive Vice President & Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Nuclear Asset Manager Xcel Energy, Inc.
414 Nicollet Mall, R.S. 8 Minneapolis, MN 55401 November 2004
ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOURTH 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM RELIEF REQUEST NO. 4 NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263
1.0 INTRODUCTION
The Nuclear Management Company, LLCs (NMCs) letter of February 16, 2004, as supplemented August 18, 2004, proposed an alternative to the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) at Monticello Nuclear Generating Plant (MNGP). Table IWB-2500-1, Category B-K, Item B10.10, of Section XI of the ASME Code requires NMC to perform a surface examination of the reactor pressure vessel stabilizer bracket welds each inspection interval. Instead, NMC proposed performing a surface examination on the stabilizer brackets only if jet reaction forces or seismic design loads are experienced. In addition, NMC proposed performing a one-time visual examination on the accessible surfaces of each bracket attachment weld and adjacent areas during the current inservice inspection (ISI) interval.
2.0 REGULATORY REQUIREMENT ISI of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (NRC), if: (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection (ISI) of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable Code of record for the fourth 10-year ISI for MNGP is the 1995 edition, with 1996 addenda, of the ASME Code,Section XI.
3.0 TECHNICAL EVALUATION
3.1 Component for Which Relief Is Requested Code Class:
Class 1 Examination Categories:
B-K (Welded Attachments for Vessels, Piping, Pumps and Valves)
Item Number:
B10.10 (Welded Attachments for Pressure Vessels)
Component
Description:
Reactor pressure vessel (RPV) stabilizer bracket welds 3.2 ASME Code Requirement ASME Code,Section XI, 1995 edition with 1996 addenda, Table IWB-2500-1, Category B-K, Item Number B10.10 requires each welded attachment to be examined with a surface examination method each inspection interval.
3.3 NMCs Proposed Alternative As an alternative to the requirements of the ASME Code,Section XI, Table IWB-2500-1, Category B-K, Item B10.10, NMC proposes to perform a surface examination on the stabilizer brackets if local (jet reaction forces) or seismic design loads are experienced. In addition, NMC proposes performing a one-time visual examination on the accessible surfaces of each bracket attachment weld and adjacent areas. This one-time visual examination will be performed during the current ISI interval.
3.4 NMCs Basis for Relief (as stated):
Monticello Nuclear Generating Plant is a General Electric Type 3 Boiling Water Reactor (BWR-3) with a Mark I Containment. The reactor vessel was designed and built to the 1965 Edition of ASME Section III with Summer 1966 Addenda.
Piping systems were designed in accordance with the 1967 Edition of USA Standard (USAS) Code for Pressure Piping B31.1.0. "Power Piping."
Construction Permit CPPR-31 was issued on June 19, 1967 and full commercial operation began on June 30, 1971.
Plants of this vintage were designed and erected prior to the examination access requirements of ASME Section XI. The Atomic Energy Commission (AEC) mandated the rules of ASME Section XI in 1971 for all nuclear plants with construction permits issued after January 1, 1971, and in 1976, they mandated use of ASME Section XI for all nuclear plants.
10 CFR 50.55a(g)(1) states, "For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued before January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (g)(5) of this section to the extent practical."
10 CFR 50.55a(g)(4) states, "... components (including supports) which are classified as ASME Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda..." [Emphasis Added]
At MNGP, four RPV stabilizer brackets are attached to the Class 1 RPV with full penetration fillet welds at 0°, 90°, 180°, and 270° RPV azimuth at an elevation of 994' 2". The RPV stabilizers are connected with flexible couplings to the brackets on the RPV and also to the biological shield wall. The RPV stabilizers, brackets, and their attachment welds are designed to withstand and resist local loads (jet reaction forces) and seismic loads while allowing axial and radial movement due to normal thermal growth. The RPV stabilizer brackets do not provide structural support during normal operation. The MNGP RPV has never experienced jet reaction forces or seismic events, therefore, the stabilizers, brackets, and attachment welds have not experienced the loads for which they are designed.
The MNGP Mark I primary containment structure, or drywell, is shaped somewhat like an upside-down light bulb. The RPV stabilizer brackets are located in the higher, necked-down elevations of the drywell. This region of the drywell is a very limited access area; it was not designed with the intention of providing access and accommodations normally considered necessary for a general work area.
The area around the stabilizers is extremely congested. The RPV stabilizer brackets are surrounded by mirror insulation that is secured by cable hangers and buckles, ventilation ductwork with support bracing, and electrical installations, such as thermocouples. All of this equipment must be relocated and restored to provide access to the stabilizers for examination of the welds.
Additionally, due to the location of the stabilizer brackets and the lack of a working platform at the stabilizer location, a complex scaffold installation is required to provide access to the examination location.
The photos and reference drawings attached to the request show the physical obstacles imposed by the design and construction of the primary containment, RPV, ventilation ducting, RPV stabilizers, containment supports, and other interferences. Combined, these obstacles create an unusually difficult hardship to overcome to provide access for the examination of the stabilizer bracket attachment welds that is specified by the Code.
In the course of scaffold installation and removal, interference removal and restoration, insulation removal and restoration, weld preparation, performance of the examination, and health physics monitoring, NMC personnel would be subjected to significant radiation doses found in the drywell. Dose survey maps taken from the recent refueling outages at this region of containment indicate dose rates in the general area to be 5 -140 millirem per hour (mrem/hr). It is reasonable to expect that the contact dose rates at the bracket welds would be similar to those experienced at the nearby feedwater (986'-7" elevation) and main steam nozzles (999'-0" elevation). These dose rates range from 20 - 80 mrem/hr in the general nozzle area and 20 - 800 mrem/hr in contact with the components.
NMC estimates indicate that radiation exposure to personnel involved in the activities associated with examination of the four RPV stabilizer bracket welds would result in 21.675 person-rem.
In summary, NMC has determined that:
MNGP is not subject to the access requirements of ASME code as described in 10 CFR 50.55a(g)(4) due to its vintage and design.
MNGP design makes access to the RPV stabilizer brackets difficult due to their location, interferences, and surrounding equipment.
Radiological dose rate estimates for work activities necessary to perform Code required examinations on the RPV stabilizer bracket welds are high due to the general and localized radiation conditions encountered within the MNGP containment.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), NMC has determined that compliance with the Code requirement would result in hardship or unusual difficulty without a compensating increase in quality or safety.
NMCs letter of August 18, 2004, provided the following supplemental information.
Other design and operational features of the MNGP make degradation to the stabilizer bracket attachment location unlikely... :
The original MNGP reactor vessel stress analysis report and the report provided in support of the MNGP power uprate concluded that the cumulative fatigue usage factor for the stabilizer brackets was extremely low and did not need to be considered for analysis for cyclic operation. Therefore, it is reasonable to expect that inservice cracking from fatigue would not occur.
The drywell at MNGP is maintained in an essentially inerted atmosphere during operation, thus reducing the opportunities for corrosion.
MNGP, being a BWR design, does not use boric acid or a borated solution as a moderator in the reactor coolant system; water is both the reactivity moderator and the coolant. Therefore, general corrosion or degradation due to boric acid does not occur.
The vessel flange to top head joint seals are monitored for leakage to ensure the seals maintain the joint integrity.
In the event that leakage should occur, the refueling bellows skirt provides a physical barrier between the location of the stabilizer brackets and all mechanical joints associated with the vessel top head, including the vessel flange to top head joint.
NMC further said the following:
Visual examinations are capable of determining the general mechanical and structural condition of [the] component. Discontinuities and imperfections such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear or erosion that could affect the operability or functional adequacy of the component can be detected. Visual examinations can be performed remotely with essentially the same capabilities as a direct examination when conditions exist that limit reasonable accessibility. Visual examinations can be performed without the degree of hardship imposed by the Code required surface examination. The visual examination will still provide reasonable assurance that the general mechanical and structural condition has not been affected by degradation and that the components are able to perform their design function.
3.5 Duration of Proposed Alternative:
NMC is requesting relief for the fourth 10-year Interval of the ISI Program for MNGP.
3.6 NRC Staffs Evaluation The ASME Code,Section XI, requires NMC to examine the RPV stabilizer brackets with a surface examination method each inspection interval. Radiological dose rate estimates for work activities necessary to perform the ASME Code-required examinations on the RPV stabilizer bracket welds are high due to the general and localized radiation conditions encountered within the MNGP containment. NMC has provided adequate rationale for the hardships that would be encountered in performing the inspections.
The MNGP RPV has never experienced jet reaction forces or seismic events. Therefore, the stabilizers, brackets, and attachment welds have not experienced the loads for which they are designed. In addition, NMC has shown that design and operational features of the MNGP, in addition to the benign operating environment, make degradation to the stabilizer bracket attachment location unlikely.
NMCs proposed one-time visual examination will detect any discontinuities that could affect the operability or functional adequacy of the component, such as the following:
loss of integrity at bolted or welded connections loose or missing parts debris corrosion wear erosion A visual examination assures safety without the undue hardship described above.
In summary, there is no loading during normal operation. The stabilizers, brackets, and attachment welds have never experienced the loads for which they were designed.
Degradation of the component is unlikely. These determinations, along with NMCs alternative proposal to visually examine the stabilizer bracket welds during the current ISI interval and then again with a surface examination if local or seismic loads are experienced, ensure the continued structural integrity of the component without the logistical hardship and the high dose rates associated with the ASME Code examination. The NRC staff has determined that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety. Therefore, NMCs proposed alternative is authorized in accordance with 10 CFR 50.55a(a)(3)(ii).
4.0 CONCLUSION
The NRC staff concludes that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.
The NRC staff further concludes that the proposed alternative provides reasonable assurance of the structural integrity of the component. Therefore, NMCs proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the fourth 10-year ISI interval at MNGP. All other requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: A. Black Date: January 6, 2005