ML20214T642
| ML20214T642 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/24/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214T635 | List: |
| References | |
| NUDOCS 8612080651 | |
| Download: ML20214T642 (9) | |
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1 UNITED STATES
'l' NUCLEAR REGULATORY COMMISSION o
i-WASHINGTON, D. C. 20666
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j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT K0.105TO FACILITY OPERATING LICENSE NO. DPR-51
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ARKANSAS POWFP AFD LIGHT COMPANY l
ARKANSAS NUCLEAR ONE, UNIT NO. 1 4
DOCKET NO. 50-313 1
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1.0 INTRODUCTION
By letter dated September 10, 1986 (Ref. 1), with supportino data provided by letter dated September 19, 1986 (Ref. 14), and a revision provided by let ter dated November 7,1986 (Pef.15), Arkansas Power and Light Company (AP&L or the licensee) requested amendment to the TechnicalSpecificationsappendedtoFacilityOperating). License No. DPR-51 for Arkansas Nuclear One, Unit No. 1 (ANO-1 The i
proposed amendment would modify the Technical Specifications to permit operation for an eighth cycle (Cycle 8). The safety analyses performed and the resulting modifications for ANO-1 are described in the Cycle 8 j
ReloadPeport(Ref.2).
The safety analysis for the previous seventh cycle of operation at ANO-1 I
is being used by the licensee as the reference cycle for the proposed eighth cycle of operation. Cycle 7 operated with no anomalies that would 4
adversely affect Cycle 8.
Where conditions are identical or limitin I
theseventhcyclesafetyanalysis,ourpreviousevaluation(Ref.3) gin
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continues to apply.
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Our evaluation of the safety analysis for the ANO-1 Cycle 8 reload follows.
2.0 EVALUATION
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2.1 Description of the Cycle 8 Core The ANO-1 core cor.sists of 177 fuel assemblies, each of which is a 15x15 array containing 208 fuel rods,16 control ' rod guide tubes, 4
and one incore instrument guide tube. The fuel management scheme is t
basically a low-leakage design with loading pattern and enrichments chosen to provide a Cycle 8 length of 420 effective full power days (EFPDs). The loading pattern consists of one batch 7 (redesignated l
batch 70) lead test assembly (LTA) located at the center of the core; 44 batch 8 (redesignated batch 88) assemblies will be shuffled to new locations, with 12 on the core periphery; 60 of the batch 9 assemblies will be shuffled to locations at or 'near the core periphery, with 8 batch 9 assemblies surrounding the center i
location; 64 fresh batch 10 assemblies will be leaded in a symmetric j
checkerboard pattern throughout the core. The batch 8B' assemblies 8612080651 861124 ADOCK0500g3 DR i
-2 are characterized as being twice-burned assemblies while the batch 9 assemblies are once-burned. The batch 70 assembly is a thrice-burned assembly. The fuel enrichments for batches 70, 88, 9, and 10 are 2.95, 3.21, 3.30, and 3.35 weight percent uranium-235, respectively.
Reactivity control for Cycle 8 will be provided by 60 full-length silver-indium-cadmium control rods, 64 burnable poison rod assemblies (BPRA) centaining varying amounts of 8 C admixed with Al 0, and soluble boron in the primary coolant., Cycle 8 will not 2 3 contain a centrally located control rod. The core will contain eight axial power shaping rods (APSRs) for additional control of the axial power distribution.
Except for the five centrally located fuel assemblies and those fuel assemblies located on the core periphery, each fuel assembly will contain either a control rod or a BPRA.
Cycle 8 will operate at full power such that only regulating control rod Bank 7 is partially inserted and such that the Bank 8 APSRs are within the range of 9.5 to 33.3 percent withdrawn for most of Cycle 8.
After 380 EFPDs, the APSRs will be completely withdrawn from the reactor core.
The licensed core full power level is 2568 MW. The safety analysis providedinthereloadreport(Ref.2)demonsfratesthesafe operation of ANO-1 throughout Cycle 8 at full power. The following sections describe our evaluation of the safety analysis.
2.2 Evaluation of the Fuel System Design 2.2.1 Fuel Assembly Mechanical Design The 64 Babcock and Wilcox (B&W) Mark B4 15x15 fuel assemblies to be loaded as batch 10 fuel for Cycle 8 operation are mechanically interchangeable with batches 70, BB, and 9 fuel assemblies previously loaded at ANO-1. The batt5 10 fuel assemblies incorporate the design features of anti-straddle lower end fittings and annealed guide tubes. The anti-straddle lower end fitting prevents mispositioning a fuel assembly in the lower grid during fuel assembly movemert. The annealed guide tubes reduce incore irradiation fuel assembly growth which pennits higher burnup capability. The Park MK-BEB fuel assembly loaded as batch 70 differs from the Mark B4 assemblies in that some fuel rods can be easily removed and windows are cut in the upper grid skirt to permit observation of fuel rod growth.
2.2.2 Fuel Rod Design Batches 88, 9 and 10 in the ANO-1 Cycle 8 core utilize the same BAW Mark B4 fuel design. The Batch 10 fuel parameters are identical to the previously loaded batches 8B and 9 except for enrichment, which has been increased to 3.35 weight percent uranium-235.
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- There has been a change in the fuel rod pre-pressurization in that the batch 10 fuel rods have a decrease in the fuel rod pre-pressure of 50 psi. The licensee states that this change will improve fuel performance and has been included in all mechanical and thermal analyses.
The one fuel assembly in batch 70 is an extended burnup LTA, which is scheduled for its fourth cycle of burnup in Cycle 8.
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This assembly, which is described in Reference 4. is similar in design to the Mark B4 assemblies except for changes to the fuel rod and fuel assembly structure to extend its burnup capability.
Wepreviouslyconcluded(Ref.5)thattheirradiationof the LTA in ANO-1 is acceptable.
i The cladding stress, strain and collapse analyses are bounded by conditions previously analyzed for ANO-1 or were analyzed specifically for Cycle 8 using methods and limits previously i
reviewed'and approved by the NRC.
2.2.3 Fuel Rod internal Pressure Section 4.2 of the Standard Review Plan (Ref. 6) addresses a number of acceptance criteria used to establish the design j
bases to evaluate the fuel system. Among those which may I
affect the operation of a fuel rod is the internal pressure limit. The NRC staff's current criterion is that fuel rod internal gas pressure should remain below nominal system pressure during normal operation unless otherwise justified.
AP8L states that fuel rod internal pressure will not exceed nominal system pressure during nomal operation of Cycle 8.
This is based on analyses perfomed with the approved B&W I
TACO 2 code (Ref. 7). We conclude that the rod internal l
pressure limit has been acceptably considered for Cycle 8 cperation.
2.2 4 Fuel Thermal Design There are no major changes between the thermal design of the new batch 10 fuel and previous batches that will be reinserted in i
the Cycle 8 core. The licensee presented results of the thermal design evaluation of the Cycle 8 core. These are based on analyses performed with the approved TACO 2 code (Ref. 7).
The Cycle 8 core protection limits are based on a linear heat generation rate (LHGR) to centerline fuel melt of 20.5 kV/f t, which is applicable to fuel batches BB, 9 and 10..The LHGR a
limit for the one batch 70 fuel assembly is greater than 70.5 kW/ft. The results of the thermal design evtluation show no difference between the new batch 10 fuel and the previous batches 8 and 9 fuel. We have reviewed the fuel thermal j
design parameters for normal operation and find them acceptable, i
s 2.2.4.1 Loss of Coolant Accident (LOCA) Initial Conditions l
In addition to the steady-state conditions, the average fuel temperature as a function of LHGR and lifetire fuel pin pressure data used in the '.0CA analysis (see Section 7.2 of Ref. 2) are also calculated with the TACO 2 code (Ref. 7). The reload report (Ref. 2) states that the fuel temperature and pin pressure data used in the generic LOCA analysis (Ref. 8) are conservative compared to those calculated for ANO-1 Cycle 8.
The bounding values of the allowable LOCA LHGRs (see Table 7.3 of Ref. 2) include the effects of NUREG-0630 regarding fuel cladding swelling and rupture behavior during LOCA.
2.2.5 Conclusion on Cycle 8 Fuel System Design We have reviewed the fuel system design and analysis for ANO-1 Cycle 8 operation and find it acceptable, as discussed above.
2.3 Evaluation of the Nuclear Design To support Cycle 8 operation of ANO-1, the licensee has provided analyses using analytical methods and design bases established in licensing topical reports that have been approved by the NRC. The licensee has provided a comparison of the core physics parameters for Cycles 7 and 8 as calculated with these approved methods. The parameters for Cycle 7 were generated using P0007 (Ref. 9) while the parameters for Cycle 8 were generated using the NOODLE code (Ref.10). The two codes give comparable results when compared to measured data. There are slight differences in the parameters compared between Cycles 7 and 8.
These differences can be attributed to differences in new fuel assembly enrichment, BPRA loading, and shuffle pattern. All of the accidents analyzed in the Final Safety Analysis Report (FSAR) were reviewed for Cycle 8 operation. The Cycle 8 parameters were conservative when compared to analyses accepted for previous cycles and no new accident analyses are included in the reload report (Ref. 2).
We conclude that the licensee's predicted nuclear parameters are acceptable because they were obtained using approved methods, the validity of which has been reinforced through a number of cycles of predictions, including startup tests, for this and other reactors.
As a result of this review of the nuclear parameters compared to previous cycles, we concur with the licensee's conclusions regarding Cycle 8 accident analysis. The licensee plans to withdraw the APSRs near the end of Cycle 8 at 380 EFPDs. The calculated stability index is -0.022 per hour at 384 EFP0s, which ensures the axial stability of the core to axial xenon transients.
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4 j The licensee has made a number of changes in the nuclear design of Cycle 8.
These changes are (1) the center control rod has been removed, (2) the lumped burnable poison (LBP) has a 4.5 inch longer poison stack than was used for Cycle 7 that is, 121.5 versus 117 inches of B C-A10. (3) the N0ODLE code was used to calculate the physics para, meters for Cycle 8, and (4) the power level hold 3
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requirements of Technical Specifications 3.5.2.4 and 3.5.2.5 have been removed. The removal of the center control rori has been taken into account in the nuclear design and, according to the licensee, i
had a negligible effect on the Cycle 8 nuclear parameters. The LBP design alters the axial power shape and increases operating l
flexibility at the beginning of the cycle. The N000LE code has been reviewed and approved by the staff (Ref. 11). An extensive analysis has been perfomed by B&W for the licensee (Ref.13) to justify removal of the power level cut-off requirements. This power level i
cutoff had been utilized to accommodate transient xenon effects on j
power peaking factors before ascending to 100% power. The analysis.
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showed that the 5 percent total xenon factor applied in the
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computation of LOCA margin provides conservative operating limits.
The 2.5 percent radial xenon factor applied in the evaluation of initial condition departure from nucleate boiling (DNB) margin was also shown to be conservative. We conclude that these changes in the Cycle 8 nuclear design are acceptable since the nuclear design and i
resulting Technical Specifications for Cycle 8 include the effects l
of the changes calculated with approved methods.
I 2.4 Evaluation of the Thermal-Hydraulic Design The thermal-hydraulic design of Cycle 8 is identical to that of Cycle 7 as shown in the comparison of maximum design conditions in i
Table 6-1 of Reference 2.
The same methods and models approved for J
use in Cycle 7 are used for Cycle 8.
The fresh batch 10 fuel j
assemblies are hydraulically and geometrically similar to irradiated batches 88 and 9 fuel assemblies. The modified lower end fitting of the batch 10 fuel has, according to the licensee, negligible impact on the themal-hydraulic design. The one batch 70 LTA is never the i
limiting assembly during Cycle 8 operation. No departure from i
i nuclear boiling ratio (DNBR) penalty is required since the approved rod bow topical report (Ref. 12) shows that the reductinn in power production capability more than offsets any rod bow effects as burnup increases. Based on the similarities of Cycle 8 with Cycle 7 and the use of approved methods and models, we conclude that the thermal-hydraulic design of Cycle 8 is acceptable.
2.5 Evaluation of the Accident and Transient Analyses The licensee has examined each FSAR transient and accident analysis j
with respect to changes in the Cycle 8 parameters to ensure that the
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calculated consequences still meet applicable criteria. The key
. parameters having the greatest effect on the outcome of a transient i
or accident are the core thermal parameters, the thermal-hydraulic parameters, and the physics static and kinetic parameters. Fuel thermal analysis values are listed in Table 4-2 of Reference 2 for i
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' all fuel batches in Cycle 8.
Table 6-1 of P.eference 2 compares the thermal-hydraulic parameters for Cycles 7 and 8.
These parameters are the same for both cycles. The physics parameters are provided in Table 5-1 of Reference 2.
A comparison of key kinetic parameters from the FSAR and for Cycle 8 is provided in Table 7-2 of Reference
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These changes indicate no significant changes or changes in the conservative direction for all parameters except for the hot-7ero power all rod group worth. The value for Cycle 8 is somewhat less than the value in the FSAR analysis. However, the licensee has demonstrated ample shutdown margin for Cycle 8.
The effects of fuel densification on the FSAR accident analyses have also been evaluated.
A generic LOCA analysis for the B&W 177-fuel assembly, lowered loop plant design has been performed using the Final Acceptance Criteria (FAC) emergency core cooling system (ECCS) evaluation model (Pef. 8).
That analysis used the Ifmiting values of the key parameters for all plants in the 177-FA lowered-loop category and is, therefore, bounding for the ANO-1 Cycle 8 operation.
The radiological dose consequences of the accidents presented in the FSAR have been reevaluated for Cycle 8.
The reason for the reevaluation is the increased amount of energy produced by fissioning plutonium caused by the extended cycle fuel management strategy. The bases used in the radiological dose evaluation are the same as in the FSAR except for three factors: (1) the fission yield and half-lives used in the Cycle 8 evaluation are based on current data, (2) whole body gama dose conversion factors are based on updated (lowered) factors, and (3) the steam generator tube rupture accident (SGTR) evaluation considers the increased amount of steam released to the environment because of a post-TPI modification. All radio-4 logical doses are bounded by the values presented in the FSAR or are a small fraction (10%) of the 10 CFR Part 100 limits except for the maximum hypothetical accident (PHA) which meets 10 CFR Part 100 limits.
We conclude from the examination of Cycle 8 core thermal and kinetic parameters, with respect to previous cycle values and with respect to the FSAR values, that this core reload will not adversely affect the ANO-1 plant's ability to operate safety during Cycle 8.
2.6 Technical Specifications s
8 As indicated in our evaluation of the nuclear design, provided in Section 7.3, the operating characteristics of Cycle 8 were calculated with well-established, approved methods. The proposed Technical Specifications are the result of the cycle-specific analyses for power peaking, control rod worths, and quadrant tilt allowance. The removal of the power level cut-off to accommodate transient xenon effects was discussed in Section 2.3.
We conclude that the Technical Specification changes proposed by the licensee in Reference I and repeated in Section 8 of'the Cycle 8 i
Reload Report (Ref. 2) are acceptable. The proposed Technical Specification changes are as follows:
1.
A new Figure 3.2-1 will be provided giving the boric acid addition tank volume and concentration as a function of reactor coolant system temperature to accommodate Cycle 8 shutdown margin requirements.
2.
TS 3.5.2.4.1 Quadrant Tilt The power level cutoff requirement has been deleted.
3.
TS 3.5.2.5.3 Control Rod Positions 4
New control rod insertion limits are provided for 4, 3 and 2 pump operation, as well as a function of burnup interval.
APSR limits are also provided as a function of burnup interval.
4.
TS 3.5.2.5.4 Contrni Rod Positions The power level cutt rf eequirements have been deleted.
5.
TS 3.5.2.6 Reactor Power Imbalance Power-imbalance curves for Cycle 8 as a function of burnup interval are added.
6.
TS 3.5 i
The basis has been modified to include the Cycle 8 power-imbalance curve and the maximum allowed linear i
heat rate as a function of burnup interval that meets the FAC on ECCS.
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TS 4.7.1 4
This TS has been changed to reflect the removal of a rod bow penalty on DNBR margin. However, since the rod bow penalty is no longer required based on an approved B&W rod bow topical report (Ref. 12), the original Cycle I control rod insertion time is proposed, by the licensee. We conclude that this change is acceptable for the reason cited above.
2.7 Startup Testing We have reviewed the startup physics testing program for ANO-1 Cycle 8 presented in Reference 2.
We conclude that this program is acceptable since it will provide confirmation that measurements for the as-loaded core conform to the Cycle 8 nuclear desian and since the data required by the Technical Specifications will be satisfied.
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4 2.8 Conclusions We have reviewed the fuel system design, nuclear design,
. thermal-hidraulic design, and the transient and accident analysis information presented in the ANO-1 Cycle 8 Reload Report. We conclude that the proposed reload and associated modified Technical Specifications are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
We have concluded, based on the 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) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: November 24, 1986 Principal Contributors:
D. Fieno, G. Vissing 1
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REFERENCES
,j 1.
T. Gene Campbell (AP&L) letter to J. F. Stolz (NRC) on " Cycle 8 Reload Report and Proposed Technical Specification Change Request,"
dated September 10, 1986.
2.
" Arkansas Nuclear One, Unit 1 - Cycle 8 Reload Report," BAW-1918, Babcock & Wilcox Company Report August 1986 (transmitted with Reference 1 above).
3.
Guy S. Vissing (NRC) letter to John M. Griffin (AP&L) transmitting Amendment No. 92 to Facility Operating License No. DPR-51, December 20, 1984.
4.
D. C. Trimble (AP&L) letter to R. W. Reid (NRC) dated November 6,1980, transmitting BAW-1626.
5.
Robert W. Reid (NRC) letter to William Cavanaugh III (AP&L) on Amendment No. 52, Cycle 5, dated March 9,1981.
6.
Standard Review Plan, Section 4.2, Rev. 1. " Fuel System Desion,"
U. S. Nuclear Regulatory Commission Report, NUREG-0800, July 1981.
7.
" TAC 02 - Fuel Pin Performance Analysis," Y. Hsii, et al., BAW-10141P-A, Rev. 1, June 1983.
8.
"ECCS Analysis of B&W's 177-FA Lowered-Loop NSSS, W. L. Bloomfield, et al...BAW-10103, Revision 1, September 1975.
9.
" Babcock & Wilcox Version of PDQ-User's Manual," Hassan,.H. H.,
et al.,
BAW-10117P-A, Babcock & Wilcox, January 1977.
10.
"N0ODLE - A Multi-Dimensional Two-Group Reactor Simulator," Mays, C. W.,
et al., BAW-10152, Babcock & Wilcox, September 1984 11.
NRC memorandum from L. S. Rubenstein to D. M. Crutchfield on review of the N000LE code, April 10, 1985.
12.
" Fuel Rod Bowing in Babcock & Wilcox Fuel Designs," BAW-10147P-A, Rev. 1, Babcock & Wilcox Company, May 1983.
13.
"ANO-1 Power Level Cutoff Removal Analysis," P. L. Holman and B. J. Delano Babcock & Wilcox Company report, May 1986 (transmitted by letter from J. Ted Enos (AP&L) to John F. Stolz (NRC), September 19,1986).
14 J. Ted Enos (AP&L) letter to J. F. Stolz (NRC) on "ANO-1 Cycle 8 Reload Report," dated September 19, 1986.
15.
J. Ted Enos (AP8L) letter to J. F. Stolz (NRC) on "ANO-1 Cycle 8 Reload Report (BAW-1918)," dated November 7, 1986.
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