ML15112B055
| ML15112B055 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/08/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112B054 | List: |
| References | |
| NUDOCS 8205100047 | |
| Download: ML15112B055 (5) | |
Text
0 UNITEDSTATES STTE NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. l1lTO FACILITY'OPERATING LICENSE NO. DPR-38 AMENDMENT NO. I1lTO FACILITY OPERATING LICENSE NO.
DPR-47 AMENDMENT NO. 108TO FACILITY OPERATING LICENSE NO.
DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 1.0 Introduction By letter dated November 13, 1981, Duke Power Company (Duke or the licensee) submitted an application (the reload application) to amend the common Oconee Nuclear Station (0NS) Technical Specifications (TSs) to support full power operation of Unit 2 during fuel Cycle 6. The reload application also provided, as an enclosure, Babcock and Wilcox Report BAW-1691, "Oconee Unit 2, Cycle 6 Reload Report,"
August 1981, in support-of the proposals. This report incl.udes-a summary of the operating parameters and contains the safety analyses supporting Unit 2 operation during Cycle 6.
By letter dated March 24, 1982, Duke provided information on the degraded condition discovered on a Mark BZ demonstration fuel assembly which was to be reinserted in the core for Cycle 6 operation.
As a result of the movement detected in zircaloy spacer grids on this assembly, a decision was made by Duke not to reinsert it for cycle 6 but ijnstead to insert an assembly with similar reactivity from the spent fuel pool with Inconel spacer grids. Analyses were performed which showed that the conclusions of the reload application remain valid.
2.0 Discussion and Evaluation 2.1 Evaluation of Fuel System Design The reload application described the core loading to be used in Cycle 6.
Seventy-two fresh assemblies having an initial enrichment of 3.17 weight percent U-235 will be loaded. Cycle 6 is to have an extended length of approximately 400 effective full power days.
For this reason burnable poison assemblies are used to limit the -required beginning of cycle soluble boron concentration.
The seventy-two Babcock and Wilcox (B&W) Mark B4 15x15 fuel assemblies loaded as Batch 8 at the end of Cycle 5 (EOC 5) are mechanically interchangeable with Batches 6B and 7 fuel assemblies previously loaded at Oconee 2. The.Cycle 6 core will also contain four previously irradiated burnable poison rod assemblies (BPRAs)2_
8205100047 820408 PDR ADOCK 05000269 p
Although.all batches in the Oconee 2 Cycle 6 core will utilize the same Mark B fuel design, the Batch 8 assemblies incorporate-a slightly different active fuel length. The change, based on undensified fuel length, is a consequence of a minor modification.in the fuel fabrication
,process. The stabilities (densification resistance) of all fuel types are almost identical.
As a consequence, the densified fuel stack height is nearly the same for all Cycle 6 assemblies.
In addition to the permanent reactivity control system (soluble boron and controlrods), 52 previously-irradiated BPRAs will be discharged and 64 fresh BPRAs will be added to control reactivity changes due to fuel burnup and fission product buildup. The irradiated BPRAs are normally removed from the reactor at the end of each cycle and fresh BPRAs are inserted for the subsequent cycle of operation, particularly where extended cycle operation is anticipated. Four previously irradiated BPRAs will remain in the Cycle 6 core for a second cycle to gather burnup data on these assemblies.
Tihe licensee has considered the impact of these four assemblies on the operation of Oconee 2 and has determined that they will not adversely affect Cycle 6 operation.
The reload application states that the cladding collapse, stress and strain analyses are bounded by conditions previously analyzed and approved by the NRC. We agree with these conclusions.
The reload application also states that fuel rod internal pressure will not exceed normal system pressure during normal operation for Cycle 6.
The analysis is based.on the use of the B&W TAFY code rather than a newer B&W code called TACO-1.
Although both of these codes have been approved for use in safety analysis, we believe that the newer TACO code is capable of more correctly calculating fission gas release (and therefore rod pressure) at very high burnups. B&W has stated that the internal fuel rod pressure predicted by TACO is lower than that predicted by TAFY for fuel rod exposures of up to 42,000 MWd/MtU. Although we have not examined the comparison, we note that the maximum expected exposure (37,046 MWd/MtU), in Oconee 2 at EOC 6, for all assemblies is lower than this predicted value. We, therefore, conclude that the rod internal pressure limits have been adequately considered for Cycle 6 operation.
The average fuel temperature as a function of linear heat rate and lifetime pin pressure data used in the Loss of Coolant Accident (LOCA) analysis (Section 7.2 of the reload application) are also calculated with the TAFY code. Duke has stated that the fuel temperature and pin pressure data used in the generic LOCA analysis are conservative compared with those calculated for Cycle 6 at Oconee 2.
The chemical and material compatibility of possible fuel, cladding and coolant interactions is unchanged from the previous cycle of operation.
The impact of material compatibility on theoperational safety of Oconee 2 need not be reconsidered for Cycle 6 operation.
-3 The licensee has calculated a fuel rod bowing penalty with 'a method similar to that previously approved. The rod bowing magnitude correlation used in that method is approved, and we.conclude that.it adequately accounts for gap closure as a function of burnup ih the Mark B fuel design.
We have reviewed those sections of the reload application for Oconee 2, Cycle 6, dealing with the fuel system design. We find those portions of the application acceptable.
2.2 Evaluation of Nuclear Design The nuclear characteristics of the core have been computed'by methods previously used and approved for B&W reactors. Comparisons are made between the physics parameters for Cycles 5 and 6. The differences that exist between the parameters are due to the increased cycle length which tends to increase values of critical boron concentrations. Changes in the radial flux and burnup distributions.between cycles also accounts for the differences in control rod worths, including ejected and stuck rod worths. All sdfety criteria are still met. Shutdown margin values at beginning and end of cycle are 3.74 and 2.40 percent Ak/k respectively compared to the required 1.0 percent.
Beginning of cycle radial power distributions show acceptable margins to limits. Based on our review, we conclude that approved methods have been used, that the nuclear design parameters meet applicable criteria and that the nuclear design of Cycle 6 is acceptable.
The key kinetics parameters for Cycle 6 have been compared to the values used in the Final Safety Analysis Report (FSAR) and densification report.
It is shown that in all cases Cycle 6 values are bounded by those previously used. We conclude that the FSAR transient and accident analysie~s are valid.
We have :reviewed the proposed TSs for Cycle 6. The limiting safety systems settings and the limiting conditions for operation have been established by previously used and approved methods. The rod withdrawal limits for the various pump combinations and times in life are presented. On the basis that previously approved methods were used to obtain the limits, we find them acceptable.
The effects of the recently discovered under-estimate of the errors in certain modules of the reactor protection system have been included. The nuclear overpower trip setpoint was (rjd.ceTfrom 105.5 to 104.9 percent full power, and the high reactor coolant temperature trip was reduced from 619 to 618 degrees '
Fahrenheit. On the basis that these setpoints were established by previously accepted methods, we conclude that the revised limits are acceptable.
2.3 Evaluation of Thermal-Hydraulic Design The incoming Batch 8 fuel is hydraulically and geometrically similar to the fuel remaining in the core from the previous cycles.
For Cycle 6 reload!
£
-- 8 BPRAs will be inserted, 12 more than the previous cycle. The fewer number
-5 An additional change related to Oconee 1 involves the clarification of the Bases on page 2.1-2 related to the DNBR margin. The bases indicate that a 10% margin results when all penalties are taken into account for Cycle 7-----!
operation. However, as a result of the under-estimate of module errors discussed above, this is no longer the case. Therefore, we have removed this statement to eliminate any confusion. Since this change is only a clarification of a Bases, we consider it to be an acceptable, administrative change.
2.5 Summary We have reviewed the physics, fuels, thermal-hydraulic and transient and accident information presented in the Oconee 2 Cycle 6 reload application and find the proposed reload and the associated modified TSs to be acceptable.
We have also reviewed the additional changes to the TSs and find them to be acceptable.
3.0 Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and,.pursuant to 10 CFR i5l.5(d)(4),
that an environmental impact statement, or negative declaration and environ mental impact appraisal need not be prepared in connection with the issuance of these amendments.
4.0 Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a signi ficant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission s regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated: April 8, 1982 The following NRC staff personnel have contributed to this Safety Evaluation:
P.,C. Wagner, L. Kopp, S. Sun, J. Vogelwede.
-4 gf unplugd gide tubes in il oad resultsin decrese of maximum bypass tfowlto~776% f6m 8.T for Cycle5.
The decreased-bypass flow and consequent7 in6rease in core flow indicates that with other core parameters unchanged, the safety margin for Cycle 6 is.at least comparable to that of Cycle 5. The reinserted BPRAs have been designed to'ensure that the impact on thermal-hydraulic analysis is insignificant.
The rod bow Departure from Nucleate Boiling Ratio (DNBR) compensation applicable to Cycle 6 was calculated using the interim rod bow compensation evaluation procedure similar to that previously approved. The burnup used to calculate the rod bow compensatio-was the highest assembly burnup in Batch 8, 19,1)00 MWd/MtU, which contains the limiting (maximum radial peak factor) fuel assembly. The resultant hot rod bow compensation factor after inclusion of the one percent flow area reduction factor credit is 0.4 percent reduction in DNBR. To demonstrate that the rod bow compensation for Batch 8 fuel is the most limiting, the licensee performed a series of thermal-hydraulic analyses. The analytical results.
for the limiting assemblies in fuel Batches 6B and 7, based on steady state power distributions, demonstrate that the increase in DNBR associated with the lower peaking of these assemblies relative to the limiting Batch 8 assembly offsets the increased rod bow DNBR compensation that would be calculated on the basis of maximum assembly burnup valaes for these batches.
We, therefore, conclude that the available margin for Cycle 6 more than offsets the 0.4 percent DNBR rod bow compensation and that the thermal-hydraulic design is, therefore, acceptable.
To support the operation of Oconee 2 at full power during Cycle 6, the licensee proposed modifications to core protection safety limits of TS 2.1 (Figures 2.1-2B, 2.1-3B and 2.3-2B).
The changes reflect a revised flux/flow setpoint of 1.08 which remains within the safety limit DNBR criterion of 1.30 and is, therefore, acceptable.
The pertinent thermal-hydraulic parameters are summarized in Table 6.1 of the reload application and are identical for Cycles 5 and 6. The fuel in Cycle 6 is geometrically similar to Cycle 5 fel'which has been previously approved for Oconee 2, and the thermal-hydraulic models and methodology used for Cycle 6 have been previously approved. We conclude that this core reload will not adversely affect the capability to operate Oconee 2 safely during Cycle 6.
2.4 5AdditionalChaiges As mentioned in Section 2.2 of this Safety Evalution, recentlyfd scovered l
under-estimate of the errors in certain modules of the reactor protection system require changes to some setpoints in the TSs. By license amendments dated November 2,1981, the NRC approved similar changes for the Oconee 1 TSs. One item,.Maximum.Nuclear Power, was not. revised'on Page 2.3-11 although it was changed in all other areas. This inconsistency was recently discovered, and we have included a revised page indicating the correct value of 1.07 vs. 1.08. Since this change:only corrects a previous omission, we consider it to be an acceptable, administrative change.