ML15113A180
| ML15113A180 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/19/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15113A179 | List: |
| References | |
| NUDOCS 8510090088 | |
| Download: ML15113A180 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 142 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 142 TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO. 139 TO FACILITY OPERATING LICENSE NO. DPP-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 INTPODUCTION 8y letter dated May 31, 1985 (Ref. 1), Duke Power Com pany (the licensee) proposed changes to the Technical Specifications (TSs) of Facility Operatina Licenses Nos. DPR-38; DPR-47 and DPR-55 for the Oconee Nuclear Station,
'nits 1, 2 and 3.
These amendments would consist of changes to the Station's common TSs.
These amendments would authorize proposed changes to the Oconee Nuclear Station TSs which are required to support the operation of Oconee Unit 3 at full rated power during the upcoming Cycle 9. The proposed amendments would change the following areas:
- 1) Core Protection Safety Limits (TS 2.1); 2) Protective System Maximum Allowable Setpoints (TS 2.3); 3) Rod Position Limits (TS 3.5.2);
and 4) Power Imbalance Limits (TS 3.5.2).
To support the license amendment application, the licensee submitted a Duke Power Company report, DPC-RD-2005 (Ref.
2), "Oconee Unit 3, Cycle 9 Reload",
as an attachment to Reference 1. A summary of the Cycle 9 operating parameters is included in the report, along with safety analyses. The fuel system design, the nuclear design, the thermal-hydraulic design, and the accident and transient analysis of this reload are presented in the Reference 2 report. An evaluation of this analysis and the proposed TS changes follows.
The Oconee Unit 3, Cycle 9 reload consists of 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes and one incore instrument guide tube.
The fuel consists of dished-end cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4. The average nominal fuel loading is 463.6 kg of uranium. The undensified nominal active fuel length is 141.8 inches; the initial mean density is 95% of theoretical.
The fuel pellet outside diameter is.3686 inches and the initial enrichment is 3.22 w/o U-235. The new loading will contain 68 new fuel assemblies designated as Mark 15-Z, Batch 11.
Cycle 9 will operate in a rods out, feed-and-bleed mode. Reactivity contro' is supplied by soluble boron, full-length Ag-In-Cd control rods, burnable poison rod assemblies and Inconel axial power shaping rods.
The design of Cycle 9 was compared to the design of Cycle 8.
3510090088 850919 PDR ADOCK 05000269 P
EVALUATION 1.0 Evaluation of the Fuel System Design 1.1 Fuel Assembly Mechanical Design Batch 11 uses intermediate spacer grids made of Zircaloy-4. All of the 68 new fuel assemblies are mechanically interchangeable into any core position.
The Cycle 9 reload will include two regenerative neutron sources (built into the poison rod assemblies). The analysis methodology is that of Reference 3 which has been approved by the NRC staff.
1.2 Cladding Stress, Strain and Collapse Fuel Batch 98 has been shown to be the most limiting for cladding creep collapse for Cycle 9 due to its longer previous incore exposure. The Batch 9R assembly power histories were analyzed and the most limiting assembly was used to perform the creep collapse analysis using the CROV code and the procedures described in Reference 4. The TACO2 (Ref. 5) code was used to calculate internal pin pressure and clad temperature used as input to CROV.
The collapse time for the most limiting assembly was estimated to be 31,400 effective full power hours which is greater than the estimated residence time of 30,460 effective full power hours for Cycle 9.
The cladding stress was estimated in a conservative and generic manner as described in Reference 3 and in compliance with the provisions of Section Ill of the ASME Boiler and Pressure Vessel Code. Exception in the methodology of Reference 3 is the static stress analysis which complies with the requirements of ASME Code Article 111-2000 for the static stress analysis.
For the stress calculation, conservative cladding dimensions were assumed, combined with hiqh external pressure (110% of design), low internal pressure and the maximum possible radial temperature gradient through the clad.
The strain was estimated using the TACO2 code (Ref. 5), and it demonstrated that the uniform circumferential strain of the cladding is within the limit of 1.0".
Based on the above results for the cladding stress, strain and collapse, it was found that the cladding design is acceptable.
1.3 Fuel Thermal Design The Cycle 9 fuel analysis was performed using the approved TACO2 code (Re'. 5).
The design of the Batch 11 fuel, which is the new fuel in Cycle 9, is such as to be equivalent to the other batches present in Cycle 9 in the remainder of the core. Conservative parameters were used to determine for each fuel batch in the core the fuel melt limits.
The maximum average assembly burnup was estimated to be 39,758 MWD/MTU and the maximum fuel rod burnup to be 40,91?
MWD/MTJ. The fuel rod internal pressure was evaluated using TACO2 and was found to be less than the nominal reactor coolant system pressure of 2,?00 psi.
The results of the fuel thermal design are acceptable.
id
-3 2.0 Evaluation of the Nuclear Design Cycle 9 differs from Cycle 8 in that there are now 68 new Mark 85-7 assemblies, the use of gray axial power shaping rods and the use of new control rod group patterns.
The nuclear design calculations were carried out using the approved methods of Reference 3. The burnable poison rod assemblies are now being loaded in a different pattern as a result of the fuel assembly shuffle pattern; Cycle 9 is a transition cycle toward low leakage loadings. This affects also the power distribution and control rod worths. Analysis of the shutdown margin indicates that the minimum value is 2.74T A k/k compared to the required shutdown margin of 1.0%& k/k.
The results of the Cycle 9 physics analysis were found to be acceptable.
3.0 Evaluation of the Thermal-Hydraulic Design The methods described in the Oconee Station Final Safety Analysis Report, the Oconee reload methodology (Ref. 3), the Unit 3 Cycle 8 reload report (Ref. 6) and the Oconee Fuel Densification Report (Ref. 7) were utilized in the thermal hydraulic analysis of Unit 3 Cycle 9. Of the 68 new fuel assemblies in the Cycle 9 core, six have open guide tubes. Counting their contribution, the total core bypass flow is estimated to be 7.9% which, however, is less than the 8.2% assumed in the generic analysis.
The Mark 8Z fuel assembly has a slightly higher pressure drop than the Mark 8 which constitutes the remainder of the core (109 assemblies). Therefore, the limiting hot channel of the Mark 8 assemblies will receive more coolant than a full Mark 8 core.
The generic analyses based on the B&W - 2 critical heat flux correlation are bounding and applicable to the Cycle 9 core. The Mark BZ assemblies minimum departure from nucleate boiling ratio (DNBR) for the transition core is greater than 1.18 which is the BWC critical heat flux correlation limit (Ref. 8).
No fuel rod bow penalty was included in the DNBR limit used in the generic analysis.
This was justified and approved in Reference 9.
The methods used in the analysis and evaluation of the Oconee 3 Cycle 9 loading have been previously approved. Based on the results, we find the thermal-hydraulic design of Cycle 9 acceptable.
4.0 Evaluation of the Accident and Transient Analysis A generic loss of coolant 4acident (LOCA) analysis for the B&W 177 fuel assembly reactors has been performed using the final acceptance criteria in the emergency core cooling system (ECCS) evaluation model (Ref. 10).
In this analysis, the limiting parameter values for all plants were used. The values of the fuel temperature (as a function of the linear heat rate) and the pin pressure calculated for the Oconee 3 Cycle 9 are conservative compared to the corresponding values of the generic analyses (Ref. 10).
Therefore, the analysis and the LOCA limits reported:in Reference 10 provide conservative
-4 results for Cycle 9. The lower pre-pressurization of the Batch 11 assemblies has a negligible effect.on the LOCA analysis (Ref. 11).
The theoretical density of the fuel in Batch 11 is higher than that considered in the densification report (Ref. 7).
Finally, there was no need to recalculate doses because the estimates of Oconee 1 Cycle 9 are applicable to Oconee 3 Cycle 9 (Ref. 12).
From the review of the accident analyses for Oconee 3 Cycle 9 and on the bases of the parameters used with methods which have been previously approved, we conclude that the transient and accident analyses have been treated properly and are acceptable.
5.0 Evaluation of Technical Specification Changes The changes discussed above necessitated Technical Specification changes to account for the differences in power peaking and control,od worths. The changes are such that neither the thermal design criteria nor the ECCS acceptance criteria are violated.
The following Specifications have been affected:
2.1 Safety Limits, Reactor Core The modifications pertain to the value of the design flow (the Mark RZ assemblies have slightly higher hydraulic resistance), and the new radial and axial peaking values and the linear heat generation rates due to the use of the gray axial power shape rods. The revised Figure 2.1-2C specifies the acceptable limits of the thermal power level versus the reactor power imbalance. The analyses of the Mark BZ assemblies, the power distribution and the thermal-hydraulics have been performed with methods which have been approved and with a range of parameters which are accept able.
2.3 Limit Safety System Settings, Protective Instrumentation The power imbalance boundaries are established to prevent reactor thermal limits from being exceeded. The power imbalance affects the power level trip established by the power to flow ratio. The revised power level versus power imbalance limits are specified in Figure 2.3-2C.
The power level and flow rates have been estimated using approved methods and acceptable parameter ranges.
Therefore, we find the specified power level versus power imbalance limits acceptable.
-5 3.2 High Pressure Injection and Chemical Addition System Due to the change in the control rod configuration, the boron solution con centration.in the boric acid storage tank was changed to 11,000 parts per million (ppm). The nuclear analysis and estimation of this value was performed with approved codes and used an acceptable range of parameters.
Therefore, we find that the proposed change will be adequate to bring the reactor to a cold shut-down condition and is acceptable.
3.5.2 Control Rod Group and Power Distribution Limits The introduction.of the gray axial power shaping control rods affected the control rod position limits for two, three, or four pump operation versus burnup. The rod position limits for the axial power shaping rods are no longer needed. The proposed control rod position limits are shcwn in Figures 3.5.2-3, -6. -9. The operational power imbalance envelope versus burnup is shown in Figure 3.5.2-12. The nuclear characteristics of Cycle 9 have been estimated with approved codes using acceptable ranges of parameters. Therefore, we find the control rod position limits and the operating power imbalance envelope acceptable.
6.0 Evaluation Findings
We have reviewed the fuels, physics, thermal-hydraulic and accident analysis information presented in the Oconee Unit 3, Cycle 9 reload report as stated above. We find the proposed reload and the associated modified Technical Specifications acceptable.
ENVIRONMENTAL CONSIDERATION These amendments involve 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 amendments involve 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 these amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accordingly, these amendments meet theleligibility 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 these amendments.
-6 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 these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
September 19, 1985 Princioal Contributor: L. Lois
REFERENCES
- 1.
- Letter H. 8. Tucker, Duke Power Company, to H. R. Denton, Director, NRR, dated May 31, 1985.
- 2. DPC-RO-2005, "Oconee Unit 3, Cycle 9, Reload Report", Duke Power Company, May 1985.
- 3. NFS-1001A, "Oconee Nuclear Station Reload Design Methodology",
Technical Report, Duke Power Company, April 1984.
- 4. BAW-10084A, Rev.2, "Program to Determine In-Reactor Performance of B&W Fuels-Cladding Creep Collapse" Babcock and Wilcox Company, October 1978.
- 5. BAW-10141A, Rev. 1, "TACO?, Fuel Performance Analysis", Babcock and Wilcox, June 1983.
- 6. DPC-RD-2003, "Oconee Unit 3, Cycle 8. Reload Report", Duke Power Company, February 1984.
- 7. BAW-1399, "Oconee 3 Fuel Densification Report", Babcock and Wilcox, November 1973.
- 8. BAW-10143, Part 2. "Correlation of 15x15 Geometry Zircaloy Grid Rod Bundle CHF Data with the BWC Correlation", Babcock and Wilcox, March 1980.
- 9. BAW-10147PA, Rev. 1, "Fuel Rod Bowing in Babcock and Wilcox Fuel Designs", Babcock and Wilcox, May 1983.
- 10.
BAW-10103, Rev. 3, "ECCS Analysis of B&W's 177-FA Lowered-Loop NSS", Babcock and Wilcox, July 1977.
- 11.
Letter R. J. Walker, 8&W, to K. S. Canady, Duke Power Company, dated February 18, 1985.
- 12.
BAW-1841, "Oconee Unit 1. Cycle 9 Reload Report", Babcock and Wilcox, August 1984.