ML20236V457
| ML20236V457 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/19/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236V440 | List: |
| References | |
| NUDOCS 8712040316 | |
| Download: ML20236V457 (6) | |
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NUCLEAft REGULATORY COMMISSION
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WASHI9sGTON, D. C. 20S55 r
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%,..... f' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELAlED TO AMENDMENT NO. 51_llFAC]LITY OP,EjlQNy LICENSE NO. DPR-22 l
NORTHERN STATES POWER COMPAM h0NTICELLO NUCLEAR GENERATI6MANT 7 0CKET NO. 50-263
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1.0 INTRODUCTION
By letter dated February 14, '1987, Northern States Power Company (the licensee) submitted a request for an amendment to the MonticeUo Nuclear Generating Plant Facility Operating Lkense No, DPR-22.
The amendmpt weuld extend the expiration date of the license from June 19, 2007, to Septemter 8, 2010.
2.0 DISCLSSION Section.103.C of the Atomic Energy Act of 1954 provides for a license to be 4.ssued for a period not exceeding 40 years.
Paragraph 50.51 of 10 CFR Part 50 rtates that the Commission may issue an operating license for the term requested by the applicant, or for the estimated useful life of the facility, should that be less than the term requested, but in no case greater than 40 years from the issuance date. The opercting license for the Monticello Nuclear Generating Plant was issued to expire June 19, 2007, 40 years from the issuance date of the construction permit.
Because approximately 39 months were required to construct the facility to the point of fuel loading and startup ter, ting, the effective perfed of the license is 36 years and 9 months.
Purnant to 10 CFR 50.90, Northern States Power Company by letter dated February 14, W86, proposed to amend the operating license to extend its duration for a full wm of %0 years, starting nn the effective date of the issuance of the operating li ce'nse. This request would extend the operating life of the Monticello Nuclear i
Ge:nerating Plant by 381 months. The requested date for the expiration of the operating license is September 8, 2010.
3.0 EVALUATION 3.1 Mechanical Ecuiment The components of the reactor coolant pressure houndary of the Monticello Nucleer Generatiag Plant were designed, built and tested to the appropriate ASME Boiler and Pressure Vessel Codes, Regulatory standards, and supplemental 8712040316 871119 PDR ADOCK 05000263 p
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.2-criteria in compliance with the requirements of 10 CFR Part 50, Section 50.55a.
" Codes and Standards." The inservice inssection program was described in the Technical Specifications and complies wit 1 the requirements of Section 50.55a(g),
except where specific relief was granted by the Comission pursuant to paragraph 50.55a(g)(6)(i).
1 The inspections conducted at several boiling water reactors (BWR's) indicated intergranular stress corrosion cracking (IGSCC) in large-diameter stainless steel pipe. The staff considered this a generic problem and as a result, the Comission issued Generic Letter 84-11 requiring a reinspection program at all BWR's, involving stainless steel welds in pipes greater than 4-in. diameter, in l
systems that are part of or connected to the reactor coolant pressure boundary, t
I out to the second isolation valve.
If IGSCC is discovered, repair, analysis and additional surveillance may be required to ensure the continued integrity of the affected pipe.
Further, the Comission has issued for public comment a draft of the proposed revision to NUREG-0313. " Technical Report on Material and Processing Guidelines for BWR Piping," including the Generic Letter to implement the staff's position.
i NUREG-0313. Rev. 2, contains the relevant recommendations of the Piping Review Comittee Task Group on Pipe Cracking issued as NUREG-1061 Volume 1.
NUREG-0313, Rev. 2, describes methods acceptable to the staff to control the susceptibility of BWR ASME Boiler and Pressure Vessel Code Class 1, 2, and 3 pressure boundary piping and safe ends to intergranular stress corrosion cracking.
The revision describes the technical bases for the staff's positions on the following items: materials of construction; processes to minimize or control IGSCC; water chemistry; reinforcement by weld overlay; replacement of piping; l
stress improvements; clamping devices crack characterization and repair criteria; inspection methods, schedules, and personnel; and limits on number of cracked weldments in piping. For piping that does not confom to the staff positions, varying degrees of inservice inspection is required to ensure structural integrity of the pressure boundary piping system, pursuant to paragraph 50.55a(g)(6)(ii) of 10 CFR Part 50.
In response to Generic Letter 84-11 and during the 1984 refueling outage, the Monticello Nuclear Generating Plant replaced Type 304 stainless steel pipe in the recirculation system and several connected branch systems susceptible to IGSCC with more corrosion resistant Type 316 NG stainless steel pipe. After the replacement was completed, the staff reviewed the Northern States Power Company submittal, including the esgmented inspection plan, and concluded that the Monticello Nuclear Generating Flant met the requirements and guidelines of Generic letter 84-11.
We conclude from our evaluation that compliance with the codes, standards and regulatory requirements to which the mechanical equipment for the Monticello Nuclear Generating Plant was analyzed, constructed, repaired, and inspected, including the inservice inspection programs in compliance to Section XI of the
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i ASME Boiler and pressure Vessel Code and the augnented inspections of austentic stainless steel piping required by the Commission, provide adequate assurance that the structural integrity of components important.
to safety will be maintained for the authorized operating period including i
i the extension until September 2010.
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3.2 Structures In evaluating the ' design of Category I structures for the Monticello Nuclear Generating Plant, the staff considered the (a) geology and nature of the i
foundation, (b) criteria for design loads, load combinations and design 1
stresses, and (c) seismic design criteria and method of analysis. Consideration was also given to the fact that the Dresden Units 2 and 3, Quad Cities Units l' and 2, and Millstone boiling water reactors under licensing review were designed i
by the General Electric Company and were essentially similar to the Monticello Nuclear Generating Plant.
The general requirements for the design of Category I structures'and equipment include provisions of resisting dead, live and operating combi-nation loads within the allowable stress requirements of local and state building codes, the Uniform Building Code, the ASME Boiler and Pressure Vessel Code, USAS B31.1-1967 Code for Pressure Piping, the American Institute of Steel Construction Code and the American Concrete Institute Code. A number of consultants were engaged by the staff in the review of the Category I structures and equipment.
These were identified in the SER.
In addition, the staff compared the proposed design requirements to the Generic Design Criteria, published for comment by the Commission i
on November 22, 1966.
Industrial experience with Category I structures to these standards confirm that a service life in excess of 40 years may i
be anticipated.
The use of the indicated codes, standards, and specifications in the design, analysis, and construction, and the identified testing and inservice surveillance requirements, provide reasonable assurance that the Category I structures would i
withstand service without loss of function for an extended period of 381 months-at the Monticello Nuclear Generating Plant.
3.3 Reactor Vessel The FSAR states that the reactor vessel for the Monticello Nuclear Generating Plant was designed and fabricated for a service life of 40 years at 80% plant capacity. The vessel was designed, fabricated, inspected and tested in accord-ance to Section III of the ASME Soiler and Pressure Vessel Code,1965 Edition,
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4 including Summer 1966 Addenda. The vessel.was field-erected by the Chicago Bridge and Iron Company (CB&I) under strict supervision of the General Electric I
Company to requirements more stringent than those required by the ASME Code.
The details of the vessel fabrication and inspection are recorded in Volume VII
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of the FSAR, " Pressure Vessel Design Report."
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Operation limitations on temperature and pressure were established using Appendix G of Section III of the ASME Boiler and Pressure Vessel Code and Appendix G of 10 CFR Part 50. The inservice inspection program is periodi-cally upgraded to comply with the recommendations of Section 50.55a(g),10 CFR Part 50, that incorporatesSection XI of the ASME Boiler and Pressure Vessel i
Code.
The integrity and performance capability.of the ferritic materials in the recctor vessel for the Monticello Nuclear Generating Plant is assured because the fracture toughness is monitored with a surveillance program in conformance to the extent practical to the recommendations of Appendix H,10 CFR Part 50,
" Reactor Vessel Materials Surveillance Program Requirements," and ASTM E185,
" Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.* The ferritic materials must meet the fracture toughness properties of Section III'of the ASME Boiler and Pressure Vessel Code and Appendix G,10 CFR Part 50, " Fracture Toughness Properties."
The final report on " Examination, Testing, and Evaluation of Irradiated j
Pressure Vessel Surveillance Specimens from the Monticello Nuclear Generating Plant," BCL-585-84-2, Rev. 1, November 5, 1984, was submitted by Northern States Power Company for staff review.
Surveillancedagwerepresented to show end-of-life (E0L) fluence (E-1MeV) of 3.02 x 10 nyt at IT position in the beltline region of the reactor vessel, assuming a use of 40-years at 80% full power operation of 1670 MWt.
The limiting reactor vessel beltline material was identified as plate, containing 0.17% Cu, 0.65% Ni, and 0.010%P. The unirradiated reference nil-ductility temperature of the material was 14 F and the upper-shelf was 109 ft-lbs. The adjusted reference nil-ductility temperature at j
E0L using the guidelines of Regulatory Guide 1.99, Revisions 1 and 2, was estimated at 91'F and 99*F, respectively. The E0L upper-shelf was estimated at 87 ft-lbs.
Paragraph IV B of Appendix G,10 CFR Part 50, sets minimum limits on the EOL estimated fracture toughness properties at 200*F and 50 f t-lbs, for the adjusted reference nil-ductility temperature and upper-shelf energy, respectively. The fracture toughness properties for the Monticello reactor vessel are estimated at E0L to be above the minimum limits set by Appendix G.
We conclude that there are no special considerations to indicate reactor vessel degradation for the Monticello Nuclear Generating Plant by increasing the useful life for an additional 381 months. The structural integrity of the reactor vessel is assured because it was originally designed and constructed for 32 EFPY usage as a minimum; it is monitored, inspected and
. tested to detect degradation processes at an early stage of their development; and it is operated with procedures to assure that design conditions are not exceeded.
3.a _ Conclusion on Mechanical Equipment, Structures and Reactor Vessel The staff concludes from its evaluation of the design, operation, testing and monitoring of the mechanical equipment, structures, and the reactor vessel that an extension of the operating license for the Monticello Nuclear Generating Plant'to a 40-year service life is consistent with the FSAR, SER and submittals made by the licensee. The plant is operated in compliance with the Comission's regulations, and issues associated with plant degradation have been adequately addressed. The staff recommends extending Operating License DPR-22 for the Monticello Nuclear Generating Plant to a 40 year period, commencing on the effective date of the issuance of the operating license.
3.5 ALARA The following evaluation was conducted to assure that the licensee's "as low as reascnably achievable" (ALARA) measures and dose projections are applicable for the additional years of plant operation and are in accordance with 10 CFR Part 20, " Standards For Protection Against Radiation" and Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Reasonably Achieveable" (Revision 3).
The licensee stated that operating and maintenance personnel will follow l
specific plans and procedures to ensure that ALARA goals are achieved in the extended years of operation. High radiation exposure operations will be planned and carried out by personnel trained in radiation protection and who I
will be using proper equipment. During such activities, personnel will be l
monitored for exposure to radiation and contamination. When major maintenance, i
repair, surveillance, and refueling tasks are completod, the experience gained from these activities will be factored into the radiation protection procedures i
and enhance future job proceoures and techniques to reduce personnel exposures.
The licensee anticipates improvements in robotics, remote surveillance, remote tooling, decontamination, improved computer resources, etc., to be factors in the future toward achieving ALARA doses.
l The staff concludes that the licensee has an adequate health physics organization I
and radiation protection program, and that personnel are trained for the addi-tional years of operation. The staff further concludes that the updated Final Safety Analysis Report (FSAR) for Monticello (0perational Radiation Protection) is in accordance with 10 CFR Part 20 and is consistent with the criteria of Regulatory Guide 8.8.
Thus, the staff finds the ALARA program and practices to be acceptable.
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! 3.6 Dose Assessment The licensee has provided the total occupational dose by year for the past 1
10 years (1977-1986). Special modification work was done in 1981, 1984 and 1986, such as feedwater nozzle safe end improvements, replacement of recircu-lation piping, and replacement of core spray piping resulting in higher than normal exposures ~for those years. Adjusting for the years when special modi-fications were made, the average exposure since 1977 was less than 490 person-rem.
Including those years, the average exposure would be 741 person-rem.
The staff has audited the licensee's dose assessment for the extended years of operation. The licensee based the estimate on 10 years.of operating i
experience engineering judgment. The licensee expects the' additional years-of operation of Monticello to result in an average of 741 person-rem per-year. Currently, operating boiling water reactors (BWR's) average 981 person-rem per unit annually (1980-1986).
1 3.7 Conclusion on Radiation Protection Based on the above, the staff concludes that the licensee's dos'e assessnent I
is acceptable and the Monticello radiation protection program is adequate for ensuring that occupational radiation exposures will be maintained in accordance with ALARA guidelines and in compliance with 10 CFR Part 20 requirements.
4.0 ENVIRONMENTAL CONSIDERATION
An Environmental Assessment and Finding of No Significant Impact relating to the proposed extension of the facility Operating License temination dates was published in the Federal Register on October 22, 1987 (52 FR 39575).
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public l
will not be endangered by operation in the proposed manner; and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the connon defense i
and security or to the health and safety of the public.
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Principal Contributor:
F. Litton, D. Scaletti Dated: November 19, 1987 l
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