ML20055G955

Safety Evaluation Supporting Amends 131 & 106 to Licenses DPR-51 & NPF-6

ML20055G955
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Site:	Arkansas Nuclear
Issue date:	07/16/1990
From:	Office of Nuclear Reactor Regulation
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ML20055G952	List: ML20055G951 ML20055G954 ML20055G955
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NUDOCS 9007240401
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4 SAFETY EVALUATION BY-THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N05.131 AND 106TO.

a FACILITY OPERATING LICENSE NOS. DPR-51 AND NPF-6 l

ENTERGY OPERATIONS. INC.

i ARKANSAS NUCLEAR ONE. UNIT NOS 1 AND 2 DOCKET NOS. 50-313 AND 50-368.

l'. 0 INTRODUCTION" l

By letters -dated October.30,1987 as supplemented on September 27, 1989 L

for Units l'and 2-and' January 29,1990 for Unit 1 only, Arkansas Power. and L

Light Company (AP&L) requested amendments to the Facility Operating l

License Nos. DPR-51 and NPF-6 for Arkansas Nuclear One, Units 1 and 2 (ANO-182).. The proposed amendments would change the expiration date

.for.ANO-1 from December 6, 2008 to May 20, 2014, and for ANO-2 from December 6, 2012 to July 17, 2018.-

2.0. DISCUS $10N~

l Section 103.c of the Atomic Energy Act of 1954 provides that a license is n

.to be issued for. a specified period not exceeding 40 years. L The Code of Federal Regulations.in 10 CFR'50.51 specifies that each license will be issued for a fixed period of time not'to exceed 40' years from date of L

issuance. Also, 10 CFR'50.56 and 10.CFR 50.57 allow the issuance-of an.

operating license pursuant to 10 CFR 50,51 after the construction of the.

facility has'been substantially completed, in. conformity with the; con-struction permit and when other provisions specified in 10 CFR 50.57 are L

met. The currently licensed terms for Arkansas Nuclear One, Units 1 and 2 are 40 years comencing with the issuance of the construction permits-(December 6,~1950andDecember6,1972). Accounting for'the. time that was required for plant' construction, this represents an effective operating i

L license term of-341 years for each unit. Consistent with Section 103.c of the Atomic Energy Act and Sections 50.51, 50.56 and 50.57 of the Comission's -

' regulations,. AP&L, by its application of October 30, 1987, seeks extensions.

b of the operating. license terms for ANO-182 from the date of operating L

license issuance.

3.0 EVALUATION i

AP&L's request for extension of the operating licenses is b'ased on the i

fact that a 40-year service life was considered during the design and construction.of the plant. Although this does not mean that some components will not wear out during the plant lifetime, design features were incorporated which maximize the inspectability of structures, systems and equipment. Surveillance and maintenance practices which were

. implemented in accordance with the ASME code and the facility Technical Specifications provide assurance that any unexpected degradation in plant E

eautpment will be identified and corrected.

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2 l3.1 Mechanica1' Equipment

-TheorliginalSafetyAnalysisReports-forANO-182asapprovedbyNRC's'

-Safety. Evaluation _ Reports, have evaluated the adequacy of safety-related

' mechanical-systems, equipment, and components for 40 years of plant operation.

It is clear that the design of the plant considered a 40-year service life. Where a specific design lifetime is specified in the-Safety Analysis Report, it is at least 40 years (i.e., 32 EFPY at.80% capacity j

factor). Examples include the description of the reactor pressure vessel,;

6 reactor coolant system components, and control element drive mechanisms.

In ether cases, performance requirements govern the design and no specific-design lifetime is stated.

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Although some mechanical equipment and components night wear out or need i

replacerent curing the plant operating lifetime, existing surveillance and maintenar.ce prograns.are sufficient to raintain or determine the need for replacement of safety-related components.

Periodic. inservice inspection i

and testing requirements have been incorporated into procedures to. provide the added assurarce that any unanticipated degradation in systems or-equiprent will be identified end corrected in a tirely manner. Therefore, tht stoff concludes that sefety-related mechanical systems, equipment, and con.ponents considered will nct be impacted by a 40-year operating' lifetime.

3.2. Electrical Equipment

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The staff has also evaluated the safety implications of extending the-Ah0-182 operating license on safety-related electrical systems.and u

equiprent.

This evaluation considered AP&L's review of extended servict l

life irpacts on equipnent, integrated dose qualifications and environmental qualifications in response to 10 CFR 50.49. For safety-related electrical equiprent within the scope of 10 CFR 50.49, aging reviewsihave been; conducted by'AP&L so as'to establish a qualified life for the equipeent.

For this equipment, the staff believes that'the licensee has controis-in place to ensure that required surveillance and maintenance are performed.

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These are described in the Environmental Qualification Program Manual and 1

ANO procedures. The current AP&L Equipment Qualification (EQ) program is in compliance with 10 CFR 50.49. There are currently no known open EQ issues which are affected by the extension of the operating license.

Based on this evaluation, the staff concludes that electrical systems E

design, electrical equipment selection and application, and environmental qualification cf electrical equipment either considered the effects of a 40-year operational lifetime or will not be affected by a 40-year operational lifetime.

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3.3 Reactor Yessel Integrity The' ANO-1' reactor vessel was designed considering(the effects' of 40 years of operation with a plant capacity factor of 80% 32EFPY). The'B&W Owners-Group'(BWOG)_IntegratedReactorVesselMaterialSurveillanceProgramand' 1

the. Cavity Dosimetry Program provide the means to continuously sionitor the-

~ cumulative effects of, neutron exposure on reactor vessel materials through 32 EFPY. The analyses of four ANO-1 plant-specific surveillance capsules, which were-irradiated at Davis Besse, are documented in the following reports: BAW-1440 (April 1977); BAW-1698 (November 1981) BAW-1836 (July' 1984),andBAW-2075(May1989).

Specificallyforthemogrecepttest,thecapsulereceivedanaverage

. fast fluence of 1.46X10 n/cm (Egreaterthan1.0MeV).

Based on the calculated fast flux at the reactor vessel wall, an 801 load factor, and the planned fuel management, the projected fast fluence that the ANO-1 reactorvesselinsicygsurfage will receive in 40 calendar years of operation is 9.75X10 n/cm (Egreaterthan1.0MeV).

These reports conclude that the current analytical techniques used for predicting the change in both the increase in RT and the decrease in upper-shelftoughnesspropertiesare'conservativkT In addition, the results indicate that the reactor vessel a64terials exhibited normal changes in tensile strength from exposure to neutron fluence. These analyses show that the expected cumulative neutron fluence on the ANO-1 reactor vessel will not be a-limiting consideration for 32 EFPY of plant operation. Completion of the BWOG Integrated Reactor Vessel Material Surveillance Program will ensure compliance with 10 CFR Part 50, 1,ppendices G and H~,-through 32 EFPY.

-The method of Regulatory Guide 1.99', Rev. 2, predicts a decrease in the Charpy upper-shelf energy to:below 50 ft-lbs for the' controlling weld metc1 at the vessel inside wall. However, using surveillance data and the prediction techniques presented in BAW-1803, AP&L calculated that none-of the reactur. vessel material (including the most limiting weld metal) uppct-shelf energies will decrease to below 50 ft-lbs during the vessel design-life. The uncertainties of the procedures used to evaluate the materials' upper-shelf energies necessitated that a. fracture analysis be performed on the most' limiting weld metal. This low upper-shelf elastic-plastic toughness analysis of the. controlling weld in the ANO-1 reactor vessel is provided in Section 9 of BAW-2075. The analysis _used the methodology docur.ented in B&W Topical _ Report BAW-10046A, Rev. 2.

The NRC approved the B&K0G analysis procedures in 1986. The low upper-shelf fracture analysis demonstrated that the most limiting weld. metal has adequate irradiated toughness properties to assure safe operation to 32 EFPY.

The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure. The Charpy impact data results exhibited the characteristic increase in transition temperature and decrease in upper-shelf energy. These results demonstrated that the current techniques used for predicting the change in both RTNDT ""d

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' upper-shelf properties due;to irradiation are conservative. The B&W i

= recomended operating period was extended to 32 EFPY as a result of the fourth capsule evaluation; These new operating limitations are in accordance with the requirements of Appendix G of 10 CFR Part 50. AP&L plans to submit new pressurization, heatup and cooldown limit curves-in 1990 based on the results of the fourth capsule examination.

AP&L.also plans-to install neutron dosimetry in the ANO-li cavity. in 1990 m

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~to meet Appendix H monitoring requirements. This dosimetry will be j

consistent with the B&W Owners Group Dosimetry Program, as described in e

j BAW-1875.

-As required by 10 CFR 50.61, AP&L has submitted to the NRC the projected-

' values of RT for ANO-1 reactor vessel materials through the end of the current liceElh and through 32 EFPY.

This submittal included results

' derived from the B&W Owners Group Report BAW-1895, " Pressurized Thermal

- Shock Evaluations:in Accordance With 10 CFR 50.61 for BWOG' Reactor Pressure Yessels," dated January 1986. Table 4-6 of the referenced report shows that the RT NRCscretniEhcforallANO-1reactorvessal-materialsiswellwithinthe riteria through both the current license term,-and 32 EFPY.

The most limiting material is the middle circumferential weldL L

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WF-112, which will have an RT of 251'F upon expiration of.the current of 264'F N the end of 32 EFPY. These values are p

-license and an RT

-belowthe10'CFRNS61 screening criteria of 300'F for circumferential-

' welds. As required by the NRC Safety Evaluation for the ANO-1 PTS evalua-tion, AP&L is required to submit a revaluation of RT withthepredictedvaiuewithfuturepressure-temper $lbr.andcomparison e submittals-as per-10 CFR Part 50, Appendix G.

The ANO-2 reactor vessel was also designed considering the effects of-40 years.of; operation at a plant capacity factor of 80% (32 EFPY),.The reactor Vessel Material' Surveillance Program for_ANO-2 contains six

.in-reactor surveillance capsules that are used to monitor cumulative-i effects of power operation on-reactor vessel materials. This program-L ensures that the AN0-2 reactor ve u el will meet the requirements of~10 CFR Fart 50,: Appendices G anc H through 32 EFPY.

All the reactor vessel materials are predicted to have a low' susceptibility to neutron radiation damage because of their high unirradiated charpy V-notch upper shelf energy, and their low copper, phosphorous and nickel c

content. The results of the first capsule analysis, which were documented -

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in:a Battelle Columbus Laboratory Report dated May 1, 1984 supported the predictionthatcumulativeneutronfluencewillnotbeaIImitingconsideration L

forreactorvesseloperationthroughfulldesignlif{g Thegeresults.

L indicated that at a fast neutron fluence of 3.5 X 10 n/cm base metal L

longitudinal specimens had the largest reduction in upper shelf energy.

E This reduction was from 155 ft-lb to 142 ft-lb, which is still substantially l'

higher than'the 50 ft-lb criteria of 10 CFR Part 50, Appendix G.

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surveillance capsules are scheduled to be removed and analyzed to further characterize irradiation-induced property changes for the ANO-2 reactor vessel over the 32 EFPY.

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As required.by 10 CFR 50.61, AP&L has determined projected values of RT for-the*ANO-2 reactor vessel materials. The results of-this analysis suSMttedtothe-NRConJanuary 22, 1986, showed that the RT for each materialintheANO-2reactorvesselwasbelowtheNRCscree$Ikgcriteria through the! current license ano through 32-EFPY. The most limiting material was intermediate shell plate, heat ~ number C8161-3, which will of 173' upon expiration of the current license and.a RT have an RT of 179.6'F N the end of 32 EFPY, Thesevaluesarewellwithinthe16Tf,FR P

50.61 screening criteria of 270*F for plate materials. The NRC review of the AP&L-10 CFR 50.61 submittal found the material properties of' reactor vessel beltline materials, the projected fluence at the inner surface of the reactor vessel for the end of-life of the plant and the calculated.

RT for the end of life of the plant to be acceptable. The NRC' Safety Ev$I$ationdatedJuly 20, 1987, alsofoundthattheRT{gichis.beyond'the value of 179.6*F-for the limiting plate material at the end of 32 EFPY,P current expiration date of the license to be acceptable.. As required by the NRC Safety Evaluation for the ANO-2 PTS evaluation, AP&L will submit a.

and comparison with the predicted value with future reevaluationofRT@submittelsasper10CFRPart50,AppendixG.

pressure-temperatu We conclude that there are no special considerations to indicate reactor vessel degradation for Arkansas Nuclear One, Units 1 and 2 due to the proposed operating lifetime extensions. The structural integrity of the-reactor vessels is assured because each vessel was originally designed-for 32 EFPY-usage (40 years at 80f plant capacity); each is monitored, inspected, and tested to detect degradation processes at an early stage of development; and each unit is operated with procedures to assure.that design conditions are not exceeded.

3.4 Structures-For ANO-1 and ANO-2, the auxiliary and turbine buildings and intake structures are constructed of reinforced concrete and structural steel.

The: reactor-building (ANO-1) is a steel-lined, post-tensioned and reinforced concrete structure. Thecontainment(ANO-2)isalsoasteel-lined,

. post-tensioned and reinforced concrete structure.

Industrial experience with such materials establishes that a service life well in excess of forty (40)' years can be anticipated.

Surveillance, inspection, and testing programs are in place to monitor the condition of the reactor building / containment structures so that any degradation can be identified and corrected.

In particular the ANO-1 reactor building and the ANO-2 containment integrated leak rate test (ILRT) performed at least three times every 10 years, verifies the leak. tightness of these structures throughout their service life. Surveillances of the gost-tensioning system are also provided which further verify reactor suilding/ containment integrity.

From the results of tendon surveillances so far performed for the reactor building / containment, there is the possibility that some of the tendon forces may go below the Ininimum required values if the operating license terms are extended from the date of operating license issuance. This condition is to be identified and rectified as required by the facility technical specification on tendon surveillance.

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1 3.5 Elevated Reactor Building Temperatures (ANO-1)

The reactor building air temperature has been higher than expected throughout the operating history of ANO-1. Since 1974, this has been, and continues to _be, the' subject of investigative and corrective actions. The Justification forContinuedOperation(JCO),submittedtotheNRConAugust 27, 1987 1

described the results of a comprehensive evaluation of safety _ implications' of the elevated reactor building temperature. The evaluation included a detailed review of the plant's design basis, accident analysis, structural performance, system and equipment performance, and equipment qualification..

-The JC0 concluded.that operation with elevated reactor building temperatures-has no_significant adverse effects on plant components, ;ystems, and structures, nor on overall safety or the plant's response to postu' lated h

accidents and transients.

Furthermore, AP&LLhas connitted to develop a long term action plan. Actions will be oriented toward two areas:

further evaluation of elevated tempera-

.ture effects;.and reduction of reactor building bulk average temperature.

-Long term inspections, preventative maintenance, and replacement intervals for structures and systems adversely affected by past and future reactor-

.e building temperatures will be modified as necessary to provide assurance that their performance during normal and transient conditions is as required to ensure safe operation of ANO-1.

l' During 16te 1988, AP&L informed the NRC of a self-imposed operating limitation upon ANO-1, based on AP&L's discovery of reduced service water flow, and other discrepancies which could have reduced the post LOCA long-term containment cooldown and therefore call into question the qualification.of safety-related equipment ~ located in the containment.

These deficiencies were identified during the ANO-1 eighth refueling outage -_(IRS). and reported per 10 CFR = 50.73.on March 30, 1989. - Analysis was performed during-the IR8 outage which supported full power operation for Lake Dardanell. temperatures up to 70'F. Operation of ANO-1 with lake temperature in excess of 70'F was precluded at that time pending further evaluation of the discrepancies.

AP&L's analysis of the effects of the reduced service water flow and the other identified discrepancies was completed in early April 1989. AP&L evaluated the effects of the ideatified deficiencies with respect to the

' post LOCA containment temperatw e response and had determined that the lant could be operated at full power with lake temperature up to 95'F p

p(maximum anticipated lake temperat ee) without invaiidating the environmental qualification of the equipment contan.M in the Reactor Building.

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- The staff has had followup discussions with AP&L concerning the analysis and AP&L provided an additional submittal dated July 19, 1989. The staff has completed its review of AP&L's analysis and justification for full power operation with lake water temperatures up to 95'F and has found them to be acceptable. A safety evaluation is being prepared for issuance to reflect this.

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, fI-The staff believes that-the extension of the operating license dees not affect the results of the JC0 evaluations. -Plant operation is based on

an ongoing process of monitoring, inspection, evaluations, and maintenance' to assure acceptable operability, and safety. Programs such as Technical.

Specification surveillance, inservice inspection, and equipment qualification provide this assurance regardless of the duration of the operating license.

3.6-Summary of Findings k

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 Statekeents, and are acceptable.

The staff concludes from its considerations of the design, operation, testing ano monitoring of the mechanical equipment,-structures, and the x

-reactor vessels that an extension of the operating licenses for Arkansas Nuclear. One, Units 1 and 2 to a 40-year service life is consistent with the FSARs, SERs, and submittels made by the AP&L, and that there is reasonable assurance that the units will be able to continue to operate safely for the additional periods authorized by the amendments. The plants are cperated in compliance with the Commission's regulations, and issues associated with plant degradation have been adequately addressed.

.In sumary, we find that extension of the operating licenses for ANO-1&2-to allow 40-year. service lives is consistent with the Final Environmental Statements and Safety Evaluation Reports for the units and that the omission's previous findings are not changed.

'4.0 Environmental Consideration A Notice of Issuance of Environmental Assessment and Finding of No Significant Impact relating to the proposed extension of the Facility Operating License termination dates for Arkansas Nuclear One was published in the Federal Register on July 13, 1990 (55FR28850).

5.0 Conclusion The= staff has concluded, based on considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of-the public will'not be endangered by operation in the proposed manner and (2)suchactivitieswillbeconductedincompliancewiththeCommIssion's regulations,-and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Date:

July 16, 1990 Principal Contributors: Chester Poslusny, PDIV-1 Tom Alexion, PDIV-1 i

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