ML16138A707
| ML16138A707 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/10/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A708 | List: |
| References | |
| NUDOCS 9005230139 | |
| Download: ML16138A707 (4) | |
Text
UNITED STATES' NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 184TO FACILITY OPERATING LICENSE DPR-38 AMENDMENT NO.
184TO FACILITY OPERATING LICENSE DPR-47 AMENDMENT NO. 181TO FACILITY OPERATING LICENSE DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1, Z AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287
1.0 INTRODUCTION
By letter dated January 22, 1988 (Ref. 1),
Duke Power Company, licensee for the Oconee Nuclear Station (ONS),
Units 1, 2, and 3, proposed a revision to Technical Specification (TS)
Figure 3.5.2-16 for the loss of coolant accident (LOCA)-Limited Maxinium Allowable Linear Heat Rate (LHR).
The results of the analysis in BAW-2001P (Ref. 2) were provided in support of a peak power limit increase from 14.0 to 14.5 kW/ft for the 0-1000 megawatt days per metric ton uranium (MWd/mtU) burnup parameter.
A modification was submitted in a letter dated October 9, 1989 (Ref. 3) for the elimination of the 1000-2600 MWd/mtU burnup parameter.
The NRC staff review of the BAW-2001P report was limited to its applicability to the Oconee plant for the increase in LOCA-Limited Maximum Allowable LHR.
The BAW-2001P report presents the results of a series of analyses concerning low pre-pressure fuel rods.
These fuel rods are the standard Mark-B designs in which the pre-pressure (fill gas pressure) has been reduced by 50 psi.
The purpose of this reduction is to provide for extended burnup capabilities and for improved LOCA margins.
The overall burnup limit was extended by 4,000 MWd/mtU.
The increased burnup limits were obtained by trading off margin between the creep collapse and the pin pressure burnup limits.
Reducing the pre-pressure caused the pin pressure burnup limit to increase and the creep collapse limit to decrease.
The amount of pre-pressure reduction was chosen so that both resulting burnup limits were nearly equivalent.
This resulted in an increase in the overall fuel and rod burnup limit.
The analysis effort is described in this report.
2.0 EVALUATION The licensee used the most restrictive LOCA for the analysis, which was a 8.55 ft 2 double ended rupture at the reactor coolant pump discharge, with the peak power at the 2-foot core elevation (core inlet) during BOL conditions.
This analysis had resulted in an allowable LHR of 14.0 kW/ft, with a peak cladding temperature of 1865 0 F, which is in conformance with the acceptance criteria of 9005230139 900510 PDR ADOCK 05000269 FDC
-2 10 CFR 50.46. The method of analysis is described in B&W Topical Report BAW-10104, Revision 5 (Ref. 4), and the recent report (Ref. 5), "Bounding Analytical Assessment of NUREG-0630 Models on LOCA kW/ft Limits With Use of FLECSET."
For the reanalysis, the licensee used current LOCA analysis codes which include:
CRAFT2 (Ref. 6). for loss of primary coolant, REFLOD3 (Ref. 7) for refill and reflood of the reactor vessel, FLECSET (Ref. 5) to determine fuel pin heat transfer coefficients during reflood, and the THETA1-B (Ref. 9) for calculation of the fuel pin cladding temperature response. All of these codes have been approved by the NRC. The analysis included the TACO2 (Ref 10) fuel model, the NUREG-0630 cladding swell and rupture models and the B&W modified FLECHT-SEASET heat transfer correlation. The LOCA limits analysis to determine operating limits for the TSs was done at the 2-foot core elevation. For B&W operating plants, the 2-foot peak power core elevation is of most interest because it generally has the most restrictive LHR.
The design change made by the licensee was a reduction by 50 psi in the fuel rod pre-pressure, resulting in a reduction of the internal pin pressure during plant operation. The reduced internal pin pressure (1) allows for longer burnup periods (fuel cycles), and (2) delays the time to clad rupture during a LOCA.
An evaluation by the licensee indicated that the pin pre-pressure reduction would delay rupture by a time increment equal to a rise in the LHR of.5 kW/ft.
The LOCA analysis, therefore, utilized an increased LHR of 14.5 kW/ft, as compared to the current 14.0 kW/ft limit at the 2-foot core elevation for the 0-1000 MWd/mtU burnup parameter.
The licensee presented the results of the reduced pin pre-pressure LOCA analysis in Table 2 of Reference 3 which also included the previous 2-foot core elevation results for comparison purposes. The 50 psi pin pre-pressure reduction resulted in a 150 psi reduction in the operational internal pin pressure. This pressure reduction corresponds to an allowable LHR of 14.5 kW/ft at the 2-foot core elevation which was used in the analysis. In the analysis, the peak cladding temperature was 2028 0 F. This is in conformance with the acceptance criteria of 10 CFR 50.46 which requires that the calculated peak cladding temperature not exceed 2200 0F.
In a letter dated October 9, 1989 (Ref. 3), the licensee modified their original proposal to eliminate the 1000-2600 MWd/mtU burnup parameter and to have the "after 2600 MWd/mtU" burnup parameter relabled "after 1000 MWd/mtU." This change is in agreement with the tabulated results in Table 3-2 of the BAW-10104 report (Ref. 5), which was approved by NRC in a letter dated October 12, 1987 (Ref. 10), and is therefore acceptable.
3.0 FINDINGS On the basis of the LOCA analysis, which used acceptable codes and in which the results were in conformance with approved limits and criteria, the NRC staff concludes that:
-3 (1) The reduced pin pressure results in delayed rupture and, consequently, lower peak cladding temperatures for the same LHR.
(2) The reduced pin pressure maintains lower internal pin pressures for the entire fuel cycle.
(3) Based on the reanalysis of the 2-foot core elevation LHR, as reported herein, the 2-foot peak power location LOCA LHR limit as shown in TS Figure 3.5.2-16, can be raised from 14.0 kW/ft to 14.5 kW/ft for the 0-1000 MWd/mtU burnup parameter for the Oconee Nuclear Station, Units 1, 2 and 3. Also, the 1000-2600 burnup parameter can be removed because of the results provided in the approved report BAW-1915.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.32, the Commission has determined that the issuance of these amendments will have no significant impact on. the environment (55 FR 18991).
5.0 CONCLUSION
The Commission issued Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Opportunity for Hearing, which were published in the Federal Register (53 FR 20196) on June 2, 1988, and (55 FR 2720) on January76, 1990.
No requests for hearing were received.
The NRC staff has 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 issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
6.0 REFERENCES
- 1. Letter from Hal B. Tucker, Duke Power Company, to USNRC, dated January 22, 1988.
- 2. BAW-2001P, "Low Pre-Pressure Fuel Rod Program Burnup Extension LOCA Analysis," the B&W Owners Group Core Performance Committee, June 1987.
- 3. Letter from Hal B. Tucker, Duke Power Company, to USNRC, dated October 9, 1989.
- 5. BAW-1915, "Bounding Analytical Assessment of NUREG-0630 Models on LOCA kW/ft Limits With Use of FLECSET," April 1986.
-4
- 6.
BAW-10092P, Rev. 3, "CRAFT2, A Fortran Program for Digital Simulation of a Multinode Reactor Plant During. Loss of Coolant," October 1982 (Proprietary).
- 7.
BAW-10148, "REFLOD3
- Model for Multinode Code Reflooding Analysis,"
May 1981.
- 8.
BAW-10094, Rev. 3, "THETAl-B, A Computer Code for Nuclear Reactor Core Thermal Analysis (IN-1445)," February 1981.
- 9.
BAW-10141P, Y. H. Hsii, et al., "TACO2 Fuel Pin Performance Analysis,"
January 1979.
- 10. Letter from A. C. Thadani,
Principal Contributor:
H. Balukjian, SRXB L. Wiens Dated:
May 10, 1990