ML20070B826
| ML20070B826 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/29/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070B823 | List: |
| References | |
| NUDOCS 9102040126 | |
| Download: ML20070B826 (8) | |
Text
.. -
- ~ - -.
f' J
p R,00 e$
- UNITED STATES a-8 NUCL EAR REGULATORY COMMISSION a
h WASHWGTON, D. C. 20666 k,*..../
SAFETY EVALUA_ TION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j
SUPPORTING AMEN 0 MENT NO.134 TO FACILITY OPERATING LICENSE NO. DPR-35 BOSTON EDISON COMPANY
_ PILGRIM NUCLEAR f0WER STATION DOCKET NO. 50-293 INTRODUCTION By application dated February 28, 1986, theBostonEdisonCompany-(thelicensee)
J requested an amendment to Facility Operating License No. OPR-35 for the Pilgrim huclear Pvwer Station (PNPS). The proposed amendment would change the expiration date for the license from August 26, 2008 to June 8, 2012, an exterision of 3 years, 9-1/2 months.
Additional infonnation relevant to the environmental impact and addressing the provisions of the National Historic Preservation Act was provided by letter dated July 13,-1989.
DISCUSSION Section103.coftheAtomicEnergyAct(Act)of1954providesthatalicense is to be issued for a specified period-not exceeding 40 years.
The Comission's regulation in 10 CFR 50.51 specifies that each license will be issued for a fixed period of. time, to be specified in the license not to exceed 40 years from date of issuance, 10 CFR 50.57 ellows the issuance of an operating license pursuant to 10 CFR 50.50 for the full term specified in 10 CFR 50.51 in conformity with the construction permit-(CP) and when other provisions specified in 10 CFR'50.57 are met. The current term of_ the license for. the PNPS is 40 years commencing with the issuance of the CP. -This represents an effective operating term of 36 years and 2-1/2 months, not 40-years. Consistent with the Act and our rules, as noted
~above, the licensee seeks an extension of the OL term for PNPS such that the fixed perico of the license woula be 40 years from the date of issuance.of the OL.
i Current' NRC policy is to issue operating licenses for a 40-year term commencing with the date of issuance of the OL.
For PNPS this date was September 15, 1972. Thus, a 40-year term would change the expiration date from August 26,
-2008, for an extension of 3 years, 9-1/2 moraths, the interval between issuance of the CP.and OL.
EVALUATION-The NRC staff has uvaluatta ths environmental impact and safety issues associated with issuance of the proposed license amendment which would allow approximately l
four additional years of operation.
In addressing the environmentel impact the following was considered:
the need for proposed action; radiological impact; l
i 9102040126 910129 PDR ADOCK 05000293
.P.
~.. -
e i
2 nonradiological impact; alternatives; alternative use of resources; other a9enCies and persons contactedi and the basis for not preparing an environmental impact statement.
This information is provided in the NRC staff's Environmental Assessment dated November 27, 1990. The following addresses the safety issues associated with the proposed amendment.
Mechanical Equip 5ent.
The components of the reactor coolant pressure boundary of the Pilgrim Nuclear Power Station were designed, built and tested to the appropriate ASME Boiler and Pressure Vessel Codes, regulatory standards, and supplemental criteria in conipliance with the requirements of 10 CFR Part 50, Section 50.55a. " Codes and Standa rds. " The inservice inspection prograr' was cewribt.c in the Technical Specifications and complied with the requiren'ents of Secticn 50.55a(g), except where specific relief was granted by the Connission pursuant to paragraph 50.55(g)(6)(1).
The inspections conducted at several boiling water reactors (BWRs) indicated it.tergranular stress corrosion cracking (IGSCC) in large-diameter stainless
-steel pipes. ihe stati consicered this a generic problem and, as'a result, the Connission issued Generic letter 84-11 requiring a reinspection program to all BWRs involving stainless steel welds in pipes greater than 4-in. diameter, in systems that are part of or connected to the reactor coolant pressure boundary, 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.
The Pilgrim Nuclear Power Station was shut down on December 10, 1983, in compliance with the Connission's confirmatory order issued on August 15, 1983, to inspect stainless steel piping systems that were susceptible to IGSCC.
When IGSCC was observed in the reactor coolant system, pursuant to 10 CFR 50.59, the Boston Edison Company elected to replace Type 304 stainless steel piping with the more resistant Type 316 NG stainless steel piping in the following systems:
a.
Recirculation system, b.
Residual heat removal system, inside containment, Core spray system, inside containment plus pipe section containing c.
weld No. 14-B-21 outside containment, and d.
-Reactor water clean-up system, suction pipe.
After the repairs were completed Region 1 personnel concluded that the replacement was properly made, and that all applicable staff and Sections III, IX and XI ASME Boiler and pressure Vessel Code requirements were met.
The staff reviewed the reports and other information provided by Boston Edison Company and found that the actions required by the confirmatory order were satisfactorily completed.
The staf f prepared a Safety Evaluation Report which was -forwarded to the-licensee in a letter dated December 4,1984, to William D. Harrington from H. R. Denton, authorizing the Pilgrim Station to return to full power.
-.m
.m
i I
<- further, the staff concluded that crack growth in the unrepaired recirculation nozzle thermal sleeves would not affect the integrity of the reactor coolant pressure boundary during the next eighteen (18) month operating period.
In addition, the staff required that a plan tot either mitigation or repair of the thermal sleeves be submitted for review at least one month prior to the start of the next refueling outage.
In a letter dated January 2, 1987 Boston Edison provided the NRC with the results of-this thermal sleeve study that indicated IGSCC as the most likely cracking mechanism anc that hycrugen water chemistry wcs plannec as the methoc to mitigate IGSCC at Pilgrim.
On January 25, 1988, the Comission issued Generic Letter 88-01 "NRC Position 4
on IGSCC in 8WR Austenitic Stainless Steel Piping," which enclosed Revision 2 to NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Piping." hUREG-031?, Revision 2, contains the relevant recom-i mendations of the Piping Review Comittee Task Group on Pipe Cracking issued as NUREG-1061, Volume 1.
j NUREG-0313. Revision 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 cescribes the technical bases for the staff's position on the following items: material of construction; process to minimize or control IGSCC; water chemistry; reinforcement by weld overlay; replacement of piping; stress improvements; clamping devices; crack characterization and repair criteria; inspection methods, schecules, and personnel; and limits on the staff pos;tions.
Varying degrees of inservice inspection is required to ensure structural integrity of the pressure boundary piping system, pursuant to paragraph 50.55(g)(6)(ii) of 10 CFR Part 50.
Revision 3 to the Inservice Inspection Program for the Pilgrim Station for the second ten-year interval was subn.itted by the Boston Edison Company to the Comission in a letter dated December 12, 1986, and amended in a letter datec March 2,1988. The Comission, in a letter dated September 28, 1988, granted the requirements of Section XI of the ASME Boiler and pressure Vessel Code, 1980 Edition, including Winter 1980 Addendum.
The 151 Program for Pilgrim is supplemented by augmented inspections requires by the NRC and by inspections recomended by engineering judgement based on 4
industry experience.
For example, the program includes the following inspections:
a.-
Visual examination of the core spray-sparger, b.
Ultrasonic examination of the shroud head bolts, c.
Visual examination of SRM and IRM dry tubes, and d.
Examinations pursuant to Generic Letter 88-01.
The Inservice Test Program for pumps and valves was submitted for review on July 11, 1983 A Safety Evaluation Report was prepared on the program by EG&G Idaho, Inc., in February 1985.
The major portion of the program was in compliance l
-4 with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code. However, there were a number of open relief requests which are under discussion with Boston Edison Company pending their resolution.
Following a number of meetings to resolve the Comission's concerns, a revised IST program was submitted by BECo cn January 4, 1990.
In a July 20, 1990, letter to LEco, the hhc proviced comments on the revised program.
These coments were resolved and final revisions to the program were sent to the NRC on October 25, 1990. This revision is now in NRC review.
We conclude from our evaluation that compliance with the codes, standards, and regulatory requirements to which the mechanical eqvipment for the Pilgrim Stuticr. wu, cnubzeo, constructeo, repaired ano inspected, including the inservice inspection programs in compliance with Section XI of the ASME Boiler and Pressure Vessel Code, and the augmented inspections of austenitic stainless steel piping required by the Comission, provide adequate assurance that the structbral integrity of components important to safety will be maintained. Any signific6nt degradation by an active mechanism would be discovered and the mechanical equipment or component restored to an acceptable conoition. Therefore, the age of the mechanical equipment or component should not be a consideration in the extension of the operating license for the Pilgrim Nuclear Power Station.
Electric _ Equipment Aging analysis has been performed for safety-related electrical equipment in accordance with 10 CFR 50.49, " Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants." Qualified lifetimes have been identified for the equipment as part of this analysis.
Since electrical
. equipment and components can be replaced, these lifetimes will be incorporated into the Pilgrira 5tction uuintenance and replacement proceoures to ensure that all _ safety-related equipn.ent remains qualified for the life of the plant regardless of the overall age of the plant.
We conclude from our evaluation that all issues associated with safety-related electrical equipment aging have been~ adequately addressed. Therefore, the age of electrical equipnent or components should not be a consideration in the extension of the operating license for the Pilgrim Nuclear Power Station.
Structures In evaluating the design of Category I structures for the Pilgrim Station, the staff considered the a) geology and nature of the. foundation, b) criteria for
' design loads, load combinations cha cesign stresses, and c) seismic design criteria anc nethod of analysis.
The general requirements' for the design of the Category I structures and equipment include provisions for resisting dead, live and operating combination loads within the allowable stress requirenents of local and state building codes, the Uniform Building Code, the ASME Boiler and Pressure Vessel Code, the U.S. Standards 831.1.0 Piping Code, the American Institute of Steel Construction Code and the American Concrete Institute Code.
Industrial experience with Category I structures constructed to the appropriate standeras confirm that a service life in excess of forty (40) years may be anticipateo.
i ;
The use of the indicated codes, standards, and specifications in the design, analysis, and construction, Appendix B of 10 CFR Part 50 for quality assurance, and the identified testing and inservic surveillance requirements provide reasonable assurance that the Category. structures will withstand continued service without loss of function for an extenced perico of 3 years, 9-1/2 months at the Pilgrim Nuclear Power Station.
Reactor Vessel The Final Safety Analysis Report (FSAR) states that the reactor vessel for the riigrim Nuclear Power Station was designed and fabricated for a service life 0140 years at 80% plant capacity.
The vessel was cesigt.ec, f abricated, inspected and tested in accordance with Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition, including Winter 1966 Addenda, and applicable requirements for the Class A pressure vessels at the time of purchase. Operation limitations on temper 6ture and pressure are establishec using Appencix G 07 Section 111 of the ASME Boiler and Pressure Vessel Code and Apperc1x G of 10 CFR Part 50. The inservice inspection program is periodically upgraded to comply with the recomendations of Section 50.55a(g) of 10 CFR Part 50, which incorporatesSection XI of the ASME Boiler and Pressure Vessel Code.
The integrity and perforwance capability of the ferritic materials in the reactor vessel for the Pilgrim Nuclear Power Station is assured because the fracture toughness is monitored with a surveillance program in confomance to i
the extent practical with 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 Boiler and Pressure Vessel Code and Appendix G,10 CFR Part 50, " Fracture Toughness Properties."
By letter dated February 28, 1986, the Boston Edison Company proposed that the pressure and temperature limits in the Technical Specifications for the Pilgrim Nuclear Power Station be changed to account for the insertion of a low-leakage core at the start of Fuel Cycle 5.
The operating pressure and temperature limit curves were developed using neutron flux values obtained from ausinietry measurements from the surveillance capsule taken from the reactor vessel after Fuel Cycle 4.
Rigorous radiation transport calculations by the licensee showed that the low-leakage core significantly reduced the neutron fluence at the wall of the reactor pressure vessel. Thus, the irradiation damage had been over-estimated in the current pressure temperattre limit curves.
The Comission approved the proposed change in Amendment No. 94 to Facility Operating License No. DPR-35 for the station by letter, dated May 28, 1986.
Generic Letter 88-11 "NRC Position on Radiation Embrittlement of Reactor Vessel L.tt;ri61s and Its impact on Plant Operations," dated July 12, 1988, informed 611 liter. sees that the methods described in Revision 2 to Regulatory Guide 1.99 should be used to predict the effect of neutron radiation on reactor vessel materials as required by paragr'aph V.A. of Appendix G to 10 CFR Part 50, unless the use of other methods can be justified.
. In its response to GL 88-11, dated December 2,1988, Boston Edison stated they had analyzed the regulatory guide for impact on the Pilgrim Nuclear Power Station and determined that compliance with its guidelines will require a license amendment to rcvise the reactor pressure vessel presure-temperature limits.
We conclude that there are no special considerations to indicate reactor vessel degradation for the Pilgrim Nuclear Power Station by increasing the duration of useful lif e for an additional three years, 9-1/2 months. The structural integrity of the rea: tor vessel is assured because it was originally designed and constructed for 32 EFPY usage at a minimum.
As of October 16, 1990, the reactor vessel has seen only 8.58 EFPY which represents approximately 27% of the minimum desigr usage.
The reactor vessel 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.
Summary and Findir,0s u
The NRC staff concluded in the Environmental Assessment that the' annual radiological effects during the additional years of operation that would be authorized by the proposed license amendments are not more than were previously estimated in the Final Environmental Statement, and are acceptable.
The staff concludes from its considerations of the design, operation, testing and monitoring of the mechanical equipment, electrical equipment, structures, and the reactor vessel that an extension of the operating license for the Pilgrim Nuclear Power Station to a 40-year. service life is consistent with the FStR NRC Safety Evaluations supporting amendments, and submittals made by the licensee. Therefore, there is reasonable assurance that the unit will be able to continue to operatt safely for the additional period authorized by this amendment. The plant is operated in compliance with the Commission's regulations, and issues associated with plant degradation and aging have been adequately addressed.
In sumary, we find that extension of the operating license for the Pilgrim Huclear Power Station to allow a 40-year service life is consistent with the Final Environmental Statement and Safety Analysis Peport for the Pilgrim fluclear Power Station and that the Commission's previous findings are not changed.
ENVIRONMENTAL CONSIDERATION A notice of Issuance of Environmental Asnssment and Finding of-No Significant
.1mpact -relating to the proposed extensi/n of the Facility-Operating License-termination date for the Pilgrim Nuclea: Power-Station was published in the Federal Register on November 27, 1990 (5 FR49351),
i
_ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _. _. _ _ _ _ _ _,, _. _. _ _ _ _. _ _ _ _ _ _ _ _ _ _.. ~
e.
7 CONClust0N The Connission made a proposed de terr.jhot's n it et the amendment involves no significant hazards consideration which was published in the Federal Reaister on April 23, 1986 (51 FR 15393).
No public connents were received,~ and the Cormonwealth of Massachusetts aid not have any comments.
We have concluoted, based on the considerations cinussea cbove, that (1) there is reasonable assurance that tt'e health ano scfety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be ccnducted in compliance with the Comission's regulations, and (3) the issuance of the amendment wi l not Le it.uonal to tLe contwn cefense and l
security or to the health ano safety of the public.
Principal Contributors:
D. Mcdonald F. Litton H. Li Dated:
January 29, 1991 4
't
_