ML20206C471
| ML20206C471 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/08/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20206C469 | List: |
| References | |
| NUDOCS 8811160293 | |
| Download: ML20206C471 (5) | |
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UNITr0 D? TES NUCLEAR RPGUtM00 COMMISSION c
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- g SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.113 TO
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FACILITY OPERATING LICENSE NO. OPR-51 j
ARKANSAS POWER AND LIGHT CMPANY I
ARKANSAS NUCLEAR ONE,, UNIT,N0,1 2
DOCKET NO. 50-313 1.0,lp,T RODUCTION 4
l In a 'etter dated July 20, 1988, Arkansas Power and Light Corpany made applica-ti:n to modify the Technical Specifications for Arkansas Nuclear One, Unit 1
( ANO-1) to permit operation for Cycle 9.
An associated additional application was submitted by letter dated August 21, 1988 to modify the variable low pressure trip setpoint. The safety analyses performed are described in the Cycle 9 reload report. The reference cycle for this reload is Cycle 8.
All accidents analyzed in the Final Safety Analysis Report (FSAR) have been reviewed for Cycle 9 operation, i
1.1 Description, of, the, Cycle,9, Core The ANO-1 Cycle 9 core consists of 177 fuel assemblies (FAs), each of which is i
r a 15x15 array containing 208 fuel rods, 16 control rod guide tubes and one i
incere instrument guide tube. Reactivity is controlled by 60 full-lenghth Ag-In-Cd control rods, 52 burnable poison rod asserblies (BPRAs) and soluble boron shim. Eight Inconel-6CO axial power shaping rods (gray APSRs) are r
provided for additinnal control of the axial power distribution.
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1.2 Significant Areas of Review for this Reload For the most part, Cycle 9 of ANO-1 will be identical in operation to Cycle 8 and most Technical Specification changes such as reactor core safety limit i
trip setpoints, rod insertion limits and irtalance limits are the result of changes resulting from insertion of new fuel, cycle lif etime and time of l
l APSR(s)withdrawalwhichareofti made on B&W reactors. Significant changes l
are replacing the Ag-In-Cd with Inconel for neutron absorption in the t
APSRs, using a low leakage fuel cycle design and using a mixed core fuel assembly with Inconel and Zircaloy spacer grids. These changes have been made f
previously on B&W reactors and are evaluated in the fuel design and nuclear l
performance section of this Safety Evaluation, j
I 2.0 EVALUATION OF THE FUEL SYSTEM DESIGN 2.1 FuelAsserplyMechanical.0isin t
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The feed batch, batch 11, consists of 60 assemblies of the MK-B6 type with t
uranium enrichrent of 3.451. Cycle 9 will consist of I batch 60, 52 Batch 98, l
l 8811160293 estgog l
l hDR ADOCX 05000313 I
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e 64 batch 10 and 60 batch 11 assemblies. The differences between MK-86 and other f1K-B types are the zircaloy spacer grids, the fact that it is reconstitutable, and the method us6d to retain fixed control components during reactor operation. ic.is type of fuel has been previously used on other B&W reactors. All fuel assemblies are identical in concept and are j
mechanically interchangeable.
2.2. Fuel Rod Design The cladding stress, strain and collapse analyses methods used for the Cycle 9 fuel rod design are the same ones used for previous cycles. We find that no i
further review in these areas is necessary, 2.3 Fuel Thermal Design All fuel asseeblies in the Cycle 9 core are thermally similar.
The design of l
the batch 11 ftarkC6 assutlies is such that the thereal performance of this fuel is equivalent to the fuel design used in the rernainder of the cure.
l Cycle 9 core protection limits are based on a linear heat rate (LFR) to centerline fuel rett limit of 20.5 kw/ft. Maximum fuel asser.bly burnup at
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EOC9 is predicted to be less than 42,800 VWd/MtU. The fuel rod internal pressures have been evaluated for the highest burnup fuel rods and predicted to be less than the nemir.a1 reactor pressure of 2200 psia. The fuel thermal i
design analyses was performed using TACO 2 which was previously reviewed and q
approved by the staff (Ref. 1).
i 2.4 Gray APSR_D s,ign i
The gray APSR design was analyzed for cladding stress due to pressure, i
terperature and ovality.
It was fcund that the gray APSR had sufficient cladding and weld stress margins. The gray APSR was also analyzed for cladding strein due to thercal and irradiation swelling.
The results of B&W analysis shewed that no cladding strain is er duced due to thernel expansior, er irradiation swelling of the inconel absorbers.
I' The staff has reviewed the mechanical design of the gray APSR's and has previously found it acceptable (Ref. 2). This design has been used in several
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B&W reactors.
l 3.0 EVALUATIONOFTENUC).JECJS,1GN t
i The core design changes for Cycle 9 are the use of gray APSRs and the l
replac'erent of the Inconel intermediate spacer grids with Zircaloy spacer grids. The gray APSRs are longer and use a weaker absorber. Calculations with the standard three dirensional model verified that these APSRs provide i
adequate axial power distribution control. The substitution of Zircaloy spacer grids reduces the parasite absorption of reutrons.
The calculational rethods used to obtain the irportant nuclear design parareters for this cycle I
were the same ones as used for the reference cycle.
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O 3-4.0 EyAlpA,T10N,0F,THER$A}.-HJDPAp).1C DES!GN The design basis chosen for Cycle 9 thermal-hydraulic analyses was a full core of Zircaloy grid assemblies, containing 40 BPRAs, for which the core bypass flow is 8.8%.
The actual Cycle 9 core contains 52 BPRAs with core bypass flow of 8.3%.
The pressure temperature safety limits have been recalculated using the BWR CHF correlation in the LYNXT crossflow analysis.
Based on the similarity with Cycle 8 and the use of approv6d models and methods, we conclude that the thertal-hydraulic design of Cycle 9 is acceptable.
5.0 EVALUATION OF TRANSIENT AND ACCIDENT ANALYS15 Each FSAR accident analysis has been examined with respect to changes in Cycle 9 paraceters to determine the effect of the Cycle 9 reloed snd to ensure that thern.a1 performance during hypothetical transients is not degraded.
The radiological dose consequences of the accidents presented in Chapter 14 of the updated FSAR were reevaluated, except for waste gas tank rupture, which is not cycle dependent. All of the calculated Cycle 9 accident doses were below the dose acceptance criteria specified in the NRC Standard Review Plan.
The key parareters that have the greatest effect on determining the outcore of a transient are core therral properties, therral-hydraulic parareters and kinetics paraceters. Corparisons of these parameters with those fron previous cycles and the FSAR values showed that the Cycle 9 parameters are within the bounds of those used for previous analyses. Thus, Cycle 9 is beunded by the previous analyses.
6.0 TECHNICAL SPECIFICATION CHANGES The Technical Specifications have been revised for Cycle 9 operation for changes in core reactivity, power peaking and control rod worths. The Cycle 9 core also includes very low leakage fuel cycle design, a mixed Park E4/ Park E6 fuel assertly core, gray APSRs, gray AFSP withdrawal flexibility and crossf1cw analysis. These contribute to the nueber of Techncial Specification changes needed.
Figures 2.1-1, 2.1-2, 2.1 3 and the bases to TS 2.1 Figures 2.3-1, 2.3-2, Table 2.3-1 and the bases to TS 2.3 These sections are modified to reflect the mixed core of fuel asserblies with Inconel and Zircaloy spacer grids, and to use cycle specific credits. Also included is a revised variable low pressure trip setroint.
Section 3.5.2.4, 3.5.2.5.3, 3.5.2.5.4 Figures 3.5.2 1, 3.5.2 2 and 3.S.2-3, 3.5.2 4 These sections are redified to reflect new quadrant power tilt limits, changes to control rod insection limits, operational limits on the gray axial pcwer shapingrods(AP!Rs),andnewoperaticnalpowerirbalancelimit.
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-4 Section 5.3.1 This section has been modified to reflect the reconstitutable fuel asserbly design and the gray axial power shaping rods.
7.0 ENVIRONMENTAL CONSIDERATION
Pi,rsuant to 10 CFR 51.21, $1.22 and 51.35, an environmental assessrent and finding of no signifiant irpact was published in the Federal Register on Nove?ber 4,1988 (53 FR 44684}'
Accordingly, based upon the environmental assessrent, the Corrnission has determined that issuance of this amendrent will not have a significant effect on the quality of the human envirorrent, 8.0 $UtHARY The staff has reviewed the fuel system design, nuclear design, therral-hydraulic design and the transient and accident analyses inferration presented in the Arkansas i:uclear One Unit 1 Cycle 9 reload submittals.
Eased on these const-derations, the staff has concluded that:
(1) There is reasonable assurance that the health and safety of the public will not be endangered by operatier.
in the proposed canner, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of this atendrent will rot be inirical to the corron defense and security or to the health and safety of the public.
Dated: November 8,1988 Principal Contributor:
M. Chatterton l
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REFERENCES 1.
Y. Hsii, et, al., "TACO 2-Fuel Pen Performance Analyses " Babcock & Wilcox Company Report BAW-10141P-A, Rev.1 dated June 1983.
. S. Rubenstein (NT.0) to G. C. Lainas (NRC) Safety Evaluation of the 2.
rystal River Unit 3 Cycle 6 Reload, dned June 18, 1985.
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