ML15112A996
| ML15112A996 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/10/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112A995 | List: |
| References | |
| NUDOCS 8103030914 | |
| Download: ML15112A996 (9) | |
Text
o AGUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 93 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 93 TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO. 90 TO 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 August 25, 1980(1), as supplemented December 22, 1980(2) and January 22, 1981(8), Duke Power Company (DPC or the licensee) re quested amendments to the Appendix A Technical Specifications (TSs) of the Oconee Nuclear Station, Units 1, 2 and 3, Licenses Nos. DPR-38, DPR-47, and DPR-55. One request was to support the operation of Oconee Unit No.
3 at full rated power during Cycle 6. There were two other requested changes: one to add a new TS 3.1.11, Shutdown Margin, and the second to add a new Section 3.5.2.9 to TS 3.5.2, Control Rod Group and Power Dis tribution Limits; both apply to Oconee Units 1, 2 and 3.
2.0 Evaluation 2.1 Fuel Assembly Mechanical Design The sixty-eight Babcock and Wilcox (B&W) Mark B-4 15x15 fuel assemblies loaded as Batch 8 at the end of Cycle 5 (EOC 5) are mechanically inter changeable with Batches SB, 6 and 7 fuel assemblies previously loaded at Oconee 3. Fuel assemblies of the Mark B-4 design have been used in four previous refuelings of Oconee 3. The design was most recently approved(3) for the previous cycle of operation (4) and is used in other B&W nuclear steam supply systems. Two assemblies will contain regenerative neutron sources, and retainers will be used to contain the sources. Justification for the design and use of the neutron source retainer is described in the "Burnable Poison Rod Assembly Retainer Design Report"(5). A discussion of the burnable poison rods themselves is presented in Section 2.1.1 of this evaluation.
2.1.1 Reactivity Control System In addition to the permanent reactivity control system (soluble boron and control rods), 60 burnable poison rod assemblies (BPRAs) are being added to control reactivity changes due to fuel burnup and fission product buildup. The BPRAs are normally removed from the reactor at the end of first cycle and reinserted only for extended cycle operation, such as that proposed for Cycle 6. In April 1978, two BPRAs were accidentally ejected from the core of another B&W-designed reactor at Crystal River (6.
The sto o 40& 4
-2 ejected BPRAs were carried out of the reactor vessel by the coolant flow to the steam generator, where damage to the steam generator tube ends resulted. B&W determined that the ejection of the BPRAs from the core resulted from fretting wear in the holddown latching mechanism. In order to avoid similar problems at other plants, B&W redesigned and replaced the BPRA holddown mechanism on all operating B&W cores. The NRC staff has generically approved (7).the new design. We therefore conclude that changes to the core reactivity control system have been adequately con sidered for Cycle 6 operation.
2.1.2 Fuel Rod Design Although all batches in Oconee 3 Cycle 6 utilize the same Mark B-4 fuel, the Batch 8 assemblies incorporate a slightly higher initial fuel density.
The change, from 94 to 95 percent of theoretical density, is a consequence of using a modified fuel fabrication process. The stability (densification resistance) of both fuel types is similar. As a consequence, the densi fied fuel stack height is virtually unchanged for the Batch 8 assemblies.
Densification in Oconee 3 Cycle 6 fuel is discussed further in Section 2.3 of this report.
2.2.1 Cladding Collapse The licensee has stated(8) that the cladding collapse analysis in the Cycle 6 Reload Report(l) is bounded by conditions previously analyzed in the Oconee Unit 3 Final Safety Analysis Report (FSAR) or analyzed specifi cally for Cycle 6 conditions using methods and limits previously reviewed and approved by the NRC. We conclude that additional NRC staff review of the cladding collapse analysis is unnecessary for Cycle 6 operation due to the similarity of Cycle 6 fuel to previous fuel.
2.2.2 Cladding Stress The licensee has stated(8) that the cladding stress analysis described in the Cycle 6 Reload Report(l) is bounded by conditions previously analyzed in theOconee 3 FSAR or analyzed specifically for Cycle 6 conditions using methods and limits previously reviewed and approved by the NRC. We con clude that additional NRC staff review of the cladding stress analysis is unnecessary for Cycle 6 operation.
2.2.3 Cladding Strain The licensee has stated(8) that the cladding strain analysis described in the Cycle 6 Reload Report(l) is bounded by conditions previously analyzed in the Oconee 3 FSAR or analyzed specifically for Cycle 6 conditions using methods and limits previously reviewed and approved by the NRC. We conclude that additional NRC staff review of the cladding strain analysis is unneces sary for Cycle 6 operation.
-3 2.2.4 Rod Internal Pressure Section 4.2 of the Standard Review Plan (SRP)(9) addresses a number of acceptance criteria used to establish the design bases and evaluation of the fuel system. Among those which may affect the operation of the fuel rod is the internal pressure limit. The acceptance criterion (SRP 4.2,Section II.A.1(f)) is that fuel rod internal gas pressure should remain below normal system pressure during normal operation unless otherwise justified.
DPC has stated(l) that fuel rod internal pressure will not exceed nominal system pressure during normal operation for Cycle 6. This analysis is based on the use of the B&W TAFY code(1O) rather than a newer B&W code called TACO(11).
Although both of these codes have been ppproved for use in safety analyses, we believe(12). that only the newer TACO code is capable of correctly calculating fission gas release (and therefore rod pressure) at very high burnups. B&W has responded(13) to this concern with an analytical comparison between both codes. In this response, they have 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 analyses exceed the expected exposure (37,000 MWd/MTu) in Oconee 3 at EOC 6. Therefore, we conclude that the rod internal pressure limits have been adequately considered.
2.3 Fuel Thermal Design 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 submittal) are also calculated with the TAFY code(10).
B&W has stated(l) that the fuel temperature and pin.pressure data used in the generic LOCA analysis(14) are conservative compared with those calculated for Cycle 6 at Oconee 3.
As previously mentioned in Section 2.2.4 of this evaluation, B&W currently has two fuel performance codes, TAFY(10) and TACO(11), which could be used to calculate,the LOCA initial conditions.
The older code, TAFY, has been used for the Cycle 6 LOCA analysis.
Recent information(15) indicates that the TAFY code predictions do not produce higher peak cladding tempera tures than TACO for all Cycle 6 conditions as suggested in Ref. 13.
The issue involves calculated fuel rod internal gas pressures that are too low at beginning of life. The rod internal pressures are used to determine swelling and rupture behavior during LOCA. B&W has proposed(16)a method of resolving this issue which we accepted(17).
The method involves the use of reduced LOCA kW/ft limits at low core elevations during the first 50 effective full power days (EFPD) of operation.
The licensee has incor porated(2) these changes into the Oconee Nuclear Station TSs to support the operation of Oconee 3 at full rated power during Cycle 6. We have reviewed (18)these changes and find them acceptable. We conclude that the initial thermal conditions for LOCA analysis have been appropriately con sidered for Cycle,6 operation.
-4 2.4 Material Compatibility The chemical and material compatibility of possible fuel, cladding and coolant interactions is unchanged from the previous cycle of operation.
The impact of this issue on the operational safety of Oconee 3 need not be reconsidered for Cycle 6 operation.
2.5 Operating Experience B&W has accumulated operating experience with the Mark B 15xl5 fuel assembly at all of the eight operating B&W 177-fuel assembly plants. A summary of this operating experience as of April 30, 1980, is given on page 4-3 of Ref. 1.
2.6 Fuel Rod Bowing The licensee has stated that a fuel rod bowing penalty has been calculated according to the procedure that was approved in Ref. 19.
The burnup used in that calculation was the maximum fuel assembly burnup of the batch that contains the limiting fuel assembly. For Cycle 6, this burnup is 23,411 MWd/MTu in a Batch 7 assembly. The resultant rod bowing penalty was found to be a 2.1% reduction in Departure from Nucleate Boiling Ratio (DNBR).
To offset the 2.1% penalty, the licensee has drawn upon both generic and plant-specific margins. The generic margin employed was a thermal credit equivalent to 1% DNBR. This credit is a result of the standard flow-area reduction factor included in all B&W hot-channel thermal-hydraulic analyses.
The plant-specific margin employed was a 10% DNBR credit available because plant operating limits were set at conservative values that correspond to the original method (20)of calculating rod bowing penalties rather than the new procedure.
During our review of this reload application, we audited the Cycle 6 DNBR penalty due to fuel rod bowing. We were unable to reproduce the penalty, and therefore requested additional information from the licensee.
Based on information supplied in the response(8) from DPC, we were able to duplicate the DNBR penalty as previously specified in Section 6 of the reload report(1).
We conclude that the DNBR reduction due to fuel rod bowing has been con servatively calculated for Cycle 6 operation.
In order to provide for a proper accounting of margins used to offset the DNBR penalty, we required-as on other operating reactors--that.the bases for the TSs for Oconee Unit 3 be amended to identify each generic or plant-specific'margin that was used.
The licensee provided such an amendment that identified the generic margin, and we provided the plant-specific margin in the TS bases.
2.7 Nuclear Design We have reviewed the effect on the rod insertion limit and axi&1 imbalance limiting conditions of operation caused by the reduction in allowable heat generation rate at the bottom of the core due to the TAFY-TACO conversion.
In order to meet the reduced limit on the power in the lower half of the core during the first 50 EFPD of the cycle, the allowable negative imbalance has been reduced, the amount of control rod insertion allowed at full power has.
been decreased, and the amount of permitted withdrawal of the axial power
-5 shaping rods has been reduced. All of these actions are in a direction to reduce the power at the bottom of the core. The techniques used to obtain the revised limiting conditions of operation are the same as have been previously used to obtain limiting operating conditions. On the basis of our review, which is discussed above, we conclude that the revised TSs are acceptable.
A further TS (3.5.2.9) specifies that the curves shown in the various Specifications shall be valid only to the end of the nominal cycle length (in spite of the open ended nature, e.g., Figure 3.5.2-1C3 which is designated for use after 200 + 10 EFPD However, use of these curves would be permitted after the end of the nominal cycle if analyses are performed which confirm their suitability.
Such use would not, therefore, involve a TS change. If analyses failed to confirm the suitability of the curves, a TS change would have to be obtained to continue operation beyond the nominal cycle length. We find this approach to be acceptable.
2.8 Shutdown Margin The licensee proposed adding a new TS 3.1.11, Shutdown Margin. The current TSs included a shutdown margin only for the operating and refueling conditions.
TS 3.1.11 provides for a shutdown margin greater than 1% Ak/k with the highest worth control rod withdrawn for all modes of operation and ensures.
the reactor can remain subcritical during various shutdown conditions. We conclude that TS 3.1.11 provides conservative shutdown margins for all modes of reactor operation and is thus acceptable.
2.9 Thermal and Hydraulic Design The thermal and hydraulic design of the reload core was reviewed to confirm that it uses acceptable analytical methods, is equivalent to or is a justi fied extrapolation from previously approved core designs, and provides an acceptable margin of safety from conditions which would lead to fuel damage during normal reactor operation and anticipated operational transients.
Oconee 3 Cycle 6 consists of 68 new Mark B-4 Batch 8 fuel assemblies. There are 60 BPRAs inserted for Cycle 6 operation. Retainers are used on these assemblies as described in Ref. 5. Two assemblies contain regenerative neutron sources. The number of open assemblies is 46. The Cycle 5 and 6 maximum design conditions are provided in Table 6-1 of Ref. 1. The burnup used to calculate the rod bow penalty is the highest Batch 7 burnup of 23,411 MWd/MTu.
2.9.1 Evaluation of Thermal-Hydraulic Design The incoming Batch 8 fuel is hydraulically and geometrically similar to the fuel remaining from the previous cycles.
The thermal-hydraulic models and methodologies used to support Cycle 6 operation aredescribed in Ref. 21, 22 and 23.
The main differences between Cycle 6 and the Reference Cycle 5 are discussed below.
-6 Core Bypass Flow The maximum core bypass flow in Cycle 5 was 10.4%. For Cycle 6 operation, 60 BPRAs will be inserted, leaving 46 open assemblies, resulting in a decrease in calculated maximum core bypass flow to 8.1% (i.e., net increase in core flow).
BPRA Retainers The retainers added to provide positive hold-down of BPRAs introduce a small DNBR penalty discussed in Ref. 5. However, the increase in core flow due to the BPRA insertion more than compensates for the decrease in DNBR due to the BPRA retainers.
Rod Bow DNBR Penalty The rod bow DNBR penalty applicable to Cycle 6, according to the licensee, was calculated using the interim rod bow penalty evaluation procedure approved in Ref. 19.
The burnup used to calculate the penalty was the highest Batch 7 burnup, 23,411 MWd/MTu. The calculated rod bow penalty using this procedure is 2.1%. Utilizing the 1% DNB credit for the flow area reduction hot channel factor, the actual penalty is 1.1%. However, according to the licensee, all plantooperating limits based on DNBR criteria include a minimum of 10% DNBR margin available due to the plant operating limits being set at conservative values that correspond to the original method (20) of calculating rod bow penalties rather than the new procedure given in Ref. 19. The licensee wants to do this for their convenience of establishing the set points once for all the future reloads. Therefore, we find the licensee's minimum DNBR limit value of 1.43 to be conservative and acceptable.
3.0 Evaluation of Transients and Accidents The licensee has examined each FSAR (21) accident analysis with respect to the changes in Cycle 6 parameters to determine their effect on the plant performance during the analyzed transients. The parameters having an effect on the outcome of a transient are the core thermal parameters, thermal-hydraulic parameters, and the physics and kinetics parameters. The kinetics parameters, including reactivity feedback coefficients and control rod worths, have the greatest effect on the outcome of a transient. The licensee, in Table 7-1 of Ref. 1, compared the Cycle 6 input parameters to the FSAR values. Our review of these input parameters indicate that Cycle 6 is bounded by the FSAR values.
Fuel thermal analysis values are listed in Table 4-2 of Ref. 1 for all fuel batches in Cycle 6. Table 6-1 of Ref. 1 compares the thermal-hydraulic para meters for Cycles 5 and 6. These parameters are the same for both cycles with the exception of the higher value of design maximum DNBR for Cycle 6 (2.05 as compared to 1.98 for Cycle 5).
According to the FSAR (Ref. 21), loss of flow (2 pump coast down) is the worst transient and the minimum DNBR is 1.4326, which is within the licensee's acceptable limit of 1.43.
We conclude from our review of the Cycle 6 core accident-related parameters, with respect to acceptable previous cycle values and with respect to the FSAR values, that this core reload.design will enable safe operation of Oconee 3 during Cycle 6.
-7 4.0 TS Changes The proposed modifications to the Core Protection Safety Limits of Specifi cation 2.1 (Figure 2.1-3C, Page 2.1-12 of Ref. 1) have been reviewed for the Oconee 3 Cycle 6 operation, and we find the revised TSs acceptable, The TS changes related to cycle length and shutdown margin have been reviewed in Sections 2.7 and 2.8 of this Safety Evaluation.
5.0 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 CER k5l.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.
6.0 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 thepublic will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission4 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: February 10, 1981
-8 REFERENCES i*W. 0. Parker, Jr. (Duke) letter dated August 25., 1980vto R. W..Reid (NRC) transmitting Oconee Unit 3, Cycle 6 Reloadeport (BAW-1634) dated August 1980.
- 2. W. 0. Parker, Jr. (Duke) letter to H. R. Denton (NRC) dated December 22, 1980.
- 3. R. W. Reid (NRC) letter to W. 0. Parker, Jr. (Duke) dated June 22, 1979.
- 4. Oconee Unit 3 Cycle 5 Reload Report,Babcock & Wilcox Company Report BAW-T5_22, March 1979.
- 5. BPRA Retainer Design Report, Babcock & Wilcox Company Report BAW
- 1496, a-y 198.
W. P. Stewart (Florida Power Corporation) letter to C. Nelson (NRC) on "Crystal River Unit Three Status Report -
May 1, 1978," dated May 4, 1978.
- 7. T. M. Novak (NRC) memorandum to E. L. Jordon (NRC) dated December 22, 1980.
- 8. W. 0. Parker, Jr. (Duke) letter to H. R. Denton (NRC) dated January 22, 1981.
- 9. Standard Review
- Plan, Section 4.2 (Rev.
1),
"Fuel System Design,'
Lt.
S. NcerRgatry CormTli55Oneport NUREG-75/087.
- 10. C. D. Morgan and H. S. Kao, "TAFY-Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.
"TACO-Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10087P-A, Rev.2, August 1977.
- 12. D. F. Ross, Jr. (NRC) letter to J. H. Taylor (B&W) dated January 18, 1978.
- 13.
J. H. Taylor (B&W) letter to P. S. Check (NRC), dated July 18, 1978.
-9
Babcock and Wilcox Company Report BAW-10105, June 1975.
- 15. R. 0. Meyer (NRC) memorandum to L. S. Rubenstein (NRC) on "TAFY/TACO Fuel Performance Models in B&W Safety Analyses," dated June 10, 1980.
- 16. J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated September 5, 1980.
- 17. L. S. Rubenstein (NRC) letter to J. H. Taylor (B&W) dated October 28, 1980.
- 18. W. V. Johnston (NRC) memorandum for R. W. Reid (NRC) on "Oconee Unit 3 Reload for Cycle 6" dated January 13, 1981.
- 19.
L. S. Rubenstein (NRC) letter to J. H. Taylor (B&W) on "Evaluation of Interim Procedure for Calculating DNBR Reduction Due to Rod Bow," dated October 18, 1979.
- 20. D. F. Ross and D. G. Eisenhut (NRC) memorandum to D. B. Vassallo and K. R.
Goller (NRC) on "Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors," dated December 8, 1976.
- 21.
Oconee Nuclear Station, Units 1, 2 and 3 -
Final Safety Analysis Reports, Dockets Nos. 50-269, 50-270 and 50-287, Duke Power Company.
- 22. Oconee Unit 2, Cycle 4 Reload Report, BAW-1491, Babcock and Wilcox, Lynchburg, Virginia, August 1978.
- 23. Oconee 2, Fuel Densification Report, BAW-1395, Babcock and Wilcox, Lynchburg, Virginia, June 1973.