ML19261C342
| ML19261C342 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/02/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19261C341 | List: |
| References | |
| NUDOCS 7903220298 | |
| Download: ML19261C342 (6) | |
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NUCLEAR F,s2AATOFY cOMM:53:ON j,1 ;i WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.10 TO FACILITY OPERATING LICENSE NO. DPR-66 ALABAMA POWER COMPANY FARLEY NUCLEAR PLANT, UNIT N0. 1 DOCKET N0. 50-348 INTRODUCTION By letter dated November 15, 1978, supplemented by letters dated December 21, 1978 and January 12, 1979, Alabama Power Company (APC) proposed changes to the Appendix A Technical Specifications for Farley Nuclear Plant (FNP), Unit No.1.
The proposals included the following:
1.
Modifications to Specification 4.2.2.2.e.2 and Bases 3/4.2.2 and 3/4.2.3 relating to the radial peaking factor limits as a function of core height.
2.
Changes to Specification Table 3.3-1, Reactor Trip System Inter-locks, and associated Bases relating to the setpoint for a reactor trip following a turbine trip.
3.
Deletion of references to part length control rods which will be removed from the reactor during the first refueling outage.
4.
Addition of a Maintenance Superintendent as a member of the Plant Operations Review Committee in Specification 6.0, Administrative Controls.
DISCUSSION AND EVALUATION 1.
Radial Peaking Factor (Specification 4.2.2.2.e.2 and Bases 3/4.2.2 and 3.4.2.3)
By letter dated November 15, 1978, APC proposed a Technical Speci-fication change to increase the radial peaking factor for RATED THERMAL POWER, FxyRTP, limits for the remainder of Cycle 1 and subsequent cycles of operation.
Specifically, APC proposed changing Section 4.2.2.2.e.2 to increase FxyRTP from <1.55 for unrodde.d core planes to the following:
~~l q 03pOM 8 F RTP <1.68 up to core elevations of 2.4 feet F
TP <1.75 for core elevation from 2.4 feet to 7.2 feet xy F RTP <l.61 for core elevation above 7.2 feet Thus, the radial peaking factor for unrodded core planes would become core elevation dependent. Such changes have been previously considered by the NRC staff and found acceptable.M Fxy is used during periodic surveillance calculations to show that the heat flux hot channel factor, FQ, is within the current limit of 2.32.
Calculations perfomed by Westinghouse Electric Corporation for the Farley Unit No.1 cores were provided by APC letters of December 21, 1978 and January 12, 1979. Our review shows that the Farley cores would now and in the future exhibit flatter than the initial axial shape, with a corresponding lower axis' peaking, Fz. Thus, the proposed increase in radial peaking factor, Fxy, lt would be offset by the decreased axial peaking Fz and would resu in the predicted peak to core average linear heat rate ratio, Fg, remaining within current Technical Specification limits. The FQ limits would not be altered and the F XP_/ limit envelope would 2
Q remain unchanged with a maximum value of 2.32.
Similar changes in the calculational methods for FQ have been previously supported genericallyFfor Westinghouse plants where eighteen load follow power histories were analyzed to ensure that the FgXP limit was not exceeded. Westinghouse has described a subset of three of these eighteen cases.f!/ To account for the increased uncertainty inherent in analyzing a smaller number of cases, an uncertainty multiplier, denoted F S was applied. The NRC staff is currently reviewing this new method. No problems have been identified which would prohibit its use for Farley.
In fact, as noted above (reference to the Trojan Plant Amendment No. 30) the NRC staff has approved the use of a subset of three cases rather than the eighteen cases.
E etter from A. Schwencer, USNRC, to C. Goodwin, Jr., Portland l
General Electric Company, June 22, 1978 transmitting Amendment No. 30 for Trojan Nuclear Plant.
UF XP is the product of the total peaking factor and the fractional Q
core power.
E etter from C. Eicheldinger, Westinghouse Electric Corporation to l
D. B. Vassallo, USNRC, July 16, 1975.
O etter from C. Eicheldinger, Westinghouse Electric Corporation to lJohn F. Stolz, USNRC, April 6,1978.
By letters dated December 21, 1978 and January 12, 1979, APC has shown, and we agree, that the three subset method used to calCJI!!e bounding values of F XP predict that, even with the proposec Q
increase in Fxy, the FQ limit of 2.32 will not be exceeded for the remainder of Cycle 1 and all of Cycle 2.
There is no change in the core local peak power liait because FQXP has not increasec.
Thus, the present total peaking FQ limits and peak local power limits are not increased. The change does not change the margin to safety (local core power limit). Therefore, the proposec charge in Fxy is acceptable for operation through Cycle 2.
The FQ limi; envelope should be verified by the licensee for operation beyond Cycle 2.
2.
Setpoint for Reactor Trio Following a Turbine Trip By letter dated November 15, 1978, APC propa3ed to change the se acint for the reactor trip following a turbine trip. The proposal would modify the turbine trip function shown in Specification Table 3.~-1, Reactor Trip System Interlocks. APC proposed to delete the turbine trip from the permissive interlock P-7 (neutron flux > 11%) and 3 add a new permissive interlock, P-9 (neutron flux > ST%).
Tne P-3 interlock would allow a reactor trip following a tiirbine trip whe, reactor power is at or above 50% power. APC states that this change would significa.1tly reduce the down time without compromising adequate safety margins.
APC has performed an analysis confirming that deleting the reactc-trip following a turbine trip from 50% power or less has no adverse affect on plant safety. We have reviewed the APC submittal as well as the Farley Plant Final Safety Analysis Report with specific attention to Chapter 15, Accident Analyses.
The previously reviewed and NRC staff approved analysis (Section 15.2.7) for a loss of External Electrical Load and/or Turbine Trip takes no credit for the direct reactor trip following a turbine trip from 102% power.
Therefore, changing the interlock setpoint from 11% to 51% power would fall well within the bounds of an accident previously analyzed and found acceptable.
We note that the proposed Reactor Protection System design changes for both the P-7 and P-9 interlocks should be made using the same design techniques as for the existing protection logic relays.
-ne APC staff has agreed that such was its intention.
APC ras c-c: ort:
accomplishing the modifications during the upcoming reft.elir; ru 1;s.
Thus, this par.t of the Technical Specification revisicn acui: :e:: e effective prior to startup for Cycle 2 Operation.
' Based on our review of APC's submittal and the discussion noted above, the proposed design changes and Technical Specification changes are acceptable.
3.
References to Part Lenath Rod Cluster Control Assemblies (PLRCCA's)
By letter dated November 15, 1978 APC proposed to remove the PLRCCA's during the first refueling outage.
Use of part length rods is currently prohibited during power operation. APC states that removal of these rods will reduce outage time and radiation exposure associated with surveillance testing.
Part length control rods were initially installed to suppress xenon induced power oscillations in the axial direction, should such os-cillations occur. They were also intended for use in axial offset calibration tests or low power physics tests.
The Technical Specifications, as now written, require that these PLRCCA's be withdrawn and excluded from the core at all times during reactor operations. The PLRCCA's are not needed, used or assumed to be available to achieve required reactor shutdown conditions.
The proposed removal, therefore', will not cause any change in required reactivity characteristics, or safety margins at full power, low power or shutdown. To the contrary, removal will eliminate the potential for part length rods remaining in the core during operation. Such an event would cause an abnormal core flux distribution.
In addition, in order to preserve tne current dynamic operating characteristics of the reactor (i.e., pressure drops, coolant flow rates, etc.) which could be affected if just removal of the PLRCCA's were to be performed, APC proposes to install thimble plug assemblies in the spaces previously occupied by PLRCCA's. The thimble plug assembly consists of a flat base plate with short rods suspended from the bottom surface and a spring pack assembly. The twenty short rods, called thimble plugs, project into the upper ends of the guide thimbles to reduce the bypass flow area. Fuel assemblies without control rods, burnable poison rods, or source rods use identical devices. Similar short rods are also used on the source assemblies and fuel assembly guide thimbles. At installation in the core, the thimble plug assemblies interface with both the upper core plate and with the fuel assembly top nozzles by resting on the adapter plate.
The spring pack is compressed by the upper core plate when the upper internals assembly is lowered into place.
Each thimble plug
. is permanently attached to the base plate by a nut which is locked to the threaded end of the plug by a pin welded t.o the nut.
Al'l components in the thimble plug assembly, except cor the springs, are constructed from type 304 stainless steel. The springs are wound from Inconel x-750 for corrosion resistance and high strength.
These thimble plugs will effectively limit bypass flow through the '
rod cluster control guide thimbles in the fuel assemblies from which the PLRCCA's have been removed, just as they currently limit bypass flow in those assembl.ies which do not contain control rods, source rods, or burnable poison ro,ds.
Based on the considerations that (1) the PLRCCA's are not needed for reactor operation, (2) that removal of these assemblies will remove the chance for an abnormal flux distribution reactor shutdown, and (3) that insertion of the thimble plug assemblies will preserve the current dynamic operating characteristics of the reactor, this change and the associated Technical Specification changes are acceptable. Technical Specification changes become effective prior to startup for Cycle 2 operation.
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4.
Addition of Maintenance Superintendent to the Plant Operatiens Review Comittee PORC By letter dated November 15, 1978 APC proposed Administrative Control Changes to Technical Specifications 6.5.1.2, 6.5.3.1 and Figure 6.2-2.
The changes would reflect recent changes to the facility organization. The Maintenance Superintendent has assumed responsibility for Maintenance and Instrumentation and Control activities. The Operations Superintendent's duties become strictly operations oriented.
The prt.viously approved organization included one Operations Superintendent having line responsibility for three sections (Operations, I&C and Maintenance). The proposed organization places all maintenance activities under a new Maintenance Superin-tendent. APC states that the new organization increases the effectiveness of the review and approval process for procedures relating to nuclear safety.
The change also would provide added expertise to the PORC. The Maintenance Superintendent would become a member of PORC.
The change in Operations Superintendent's duties would allow more personalized operations control. at the same time.
. Since the changes proposed are improvements to the previously reviewed and approved organization, the Technical Specification changes are acceptable.
ENVIRONMENTAL CONSIDERATION We have detennined that the amendment does 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 concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement, or negativa declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
CONCLUSION We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangsred by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Connission'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: March 2, 1979