ML20058A250
| ML20058A250 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/01/1978 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058A248 | List: |
| References | |
| NUDOCS 7811170265 | |
| Download: ML20058A250 (5) | |
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o UNITED STATES o
8 ',,., r, ' ';T NUCLE AR REGULATORY COMMISSION
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SAFETY EVALUATION DY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDt ENT NOS. 32 AND 26 TO FACILITY LICENSE NOS. OPR-42 AND OPR-60 NOP.THER!i STATES POWER COMPANY PRAIRIE ISLAND tJUCLEAR GENERATING PLANT Utl!T t:05.1 AND 2 00CKET N05. 50-282 AND 50-306 I ntroduction By letters dated December 9,1977 and September 8,1978 Northern States Power Concany (the licensee) requested amendment of the Technical Specifications appended to Facility Operating License Nos. OPR-42 anc DPR-60 for the Prairie Island Nuclear Generating Plant, Unit Nos. I and 2 (PINGP). The proposed amendments would reduce the maximun pressurizer heatup rate from 200*F per hour to 100*F per hour, change the pressure-temperature operating limits and permit renoval of the part length rods.
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PRESSURIZER HEATUP RATE AND PRESSURE-TEMPERATURE LIMITS Backcround In August 1977, flitsubishi Heavy Industries, Ltd., of Japan noted an inconsistency in the pressurizer heatup rate stated in their Technical Specifications. Specification 3.4.9 required a heatup rate of 200*F/hr: Specification 5.7.1, however, required a heatup rate of 100*F/hr. This discrepancy was reported to the Westinghouse Electric Corporation (Westinghouse), who then reviewed tneir analysis of the pressurizer heatuo rate and determined that the correct heatup rate is 100*F/hr, and that the correct cooldown rate is 200*F/hr; the Technical Soecifications for the PINGP stated that pressuri:er heatup and ccol-down rates were 200*F/hr. Westinghouse then notified the Nuclear Reculstory Ccmnission (the Ccmmission) and the licensee of this proolen.
The requested ameadment would correct the error in the cressurizer heatup rate limit.
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Evaluation In designing the pressurizer, Westinghouse perfomed a thermal stress l
analysis which analyzed the fatigue resulting from a heatup rate of l
100*F/hr and a cooldown rate of 200'F/hr. This analysis meets the standards of the ASME Code,Section III, which requires that the analysis be based on a usage factor. The usage factor represents the fraction of the fatigue life (the total amount of stress that a particular component is designed to handle), with a usage factor of zero implying that no stress has been exerted on the component, and a usage factor of I
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one implying that the stress exerted on the component is equal to the amount of stress that the component is designed to handle. For any l
piece of equipment, certain ccmponents receive more stress than others.
For the pressurizer, this component is the surge nozzle, which has a usage factor of 0.9 for the design numbers listed above. This_ usage factor i
is such that if the heatup and cooldown rates used in the analysis were r
exceeded nore than a few times, the actual usage factor for the surge nozzle would exceed 1.0, which is not allot ale under the ASME Code.
Thus, we conclude that reducing the heatup rate limit from 200*F/hr to 100*F/hr is necessary to maintain themal stresses in the pressurizer i
to allowable levels. For the same reasons, we further conclude that the cooldown rate limit presently listed in the Technical Specifications i
is adequate.
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Because the currt:nt Technical Specifications provision authorized higher rates of pressurizer heatup than the correct limit, the question arose l
l as to whether the correct limit of 100*F/hr has been exceeded in the past. Discussions with Westinghouse indicate that this is unlikely.
This is because system capabilities and Technical Specification limits Y
on the rate of reactor coolant system heatup and pressurization effectively preclude pressurizer heatup rates in excess of 50*F to 75'F/hr.
i Furthermore, the licensee reviewed operating records for the PINGP and i
found that the 100*F limit had never been exceeded. System design precludes heatup rates in excess of 70*F/hr. Accordingly, we conclude that the only action required by the licensee is modification of the Technical-
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Specifications to reduce the limiting pressurizer heatup rate of i
200*F/hr to ICO*F/hr.
The Staff has corresponded with Westinghouse on this matter and has l
received confimation frcm the ccmpany that it is performing a review of the stress analyses for components of the reactor coolant pressure l
i boundary to assure that no similar inadvertent error appears in any other portion of the applicable Technical Scecifications. -This action will be confirmed by Westinghouse.
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. II. REACTOR PRESSURE VESSEL In its letter dated Decenber 9,1977, the licensee also submitted a request to change the Technical Specifications of PINGP regarding the reactor pressure vessel pressure-tenperature operating limits.
The proposed operating limit curves are based on the test results obtained fron material surveillance Capsule V that was removed from the Unit I reactor vessel.* These revised operating limits are proposed for operation through 10 effective full power years (EFPY).
We have reviewed the proposed operating limits and the test results from ul e '. The capsule received an average neutron fluence of 5.21 x Cap"n/cn. Weld metal and base metal showed increases in RT at this f
10fluence of 32 and 15'F respectively. Therewasnomeasurabibropin the upper shelf energy.
The reactor vessels for Unit I and Unit 2 are both made from A508 Class 3 forging naterial having very low copper content; below 0.1"..
As indicated by the test results from the surveillance program, this material has execlient resistance to radiation damage. The limiting material is the weld metal from Unit I having a copper content of 0.14",
and a phosphorus content of 0.016"..
Using the above material as the limiting material for both units, we made independent calculations for pressure-temperature limits through 10 EFPY. We found that the operating limits proposed by the licensee conform to those calculated by us and are in accordance with Appendix G, 10 CFR Part 50. Such compliance with Appendix G in establishing safety operating limitations will ensure adequate safety margins during operation, testing, naintenance and postulated accident conditions and constitutes an acceptable basis for satisfying the requirements of NRC General Design Criterion 31, Appendix A,10 CFR Part 50.
Based on the above, we conclude that the procosed operating limits are acceptable for c;:eration through 10 EFPY,
' Analysis of Capsule V fren.'lSP Prairie Island Unit No.1 Reactor Vessel Surveillarce Progran WCAP-8916, transnitted by NSP letter dated Sectenber 12, 1977
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111. PART LENGTH RODS t
Discussion and Evaluation p
Part length control rods were initially installed to suppress xenon induced power oscillations li. the axial direction, should such oscillations occur. They were also intended for use in axial offset calforation tests or low power physics tests.
The Technical Specifications, as now written, require that these part length rod cluster control assemblies (PLRCCAs) be withdrawn-and excluded from the core at all times during reactor operations.
The PLRCCAs 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 dropping into the core during operation. Such an event could cause abnormal flux distribution or reactor.tnutdown.
In addition, in order to preserve the current dynamic operating characteristics of the reactor (i.e., pressure drops, coolant flow I
rates, etc.) which could be affected if just removal of the PLRCCAs l
were to be performed, the licensee proposes to install thimble plug j
assemDlies in the spaces previously occupiec by PLRCCAs. The tnfmble plug assembly consists of a flat base plate with short rods suspended
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from the bottom surfa ? and a spring pack assembly. The twenty short rods, called thimble giugs, 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 l
t 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 l
l with the fuel assembly top nozzles by resting on the adapter plate.
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The spring pack is compressed by the upper core plate when the upper i
internals assembly is icwered into place. Each thimble plug is per-manently attached to the base plate by a nut which is locked to the l
threaded end of the plug by a pin welded to the nut.
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i 5-All components in the thimble plug assembly, except for the springs, are construced 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 PLRCCAs have been removed, just as they currently limit bypass flow in those assemblies which do not contain control rods, source rods, or burnable poison rods.
f Based on the considerations that (1) the PLRCCAs are not needed for reactor operatics, (2) that removal of these assemblies will remove l
the chance for an abnormal flux distribution reactor shutdown, and l
(3) that insertion of the thimble plug assemblies will preserve the current dynamic operating characteristics of the reactor, we conclude
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that this change is acceptable.
r Environmental Consideration r
We have determined that the amendments do not autnorize a change in l
1 effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made i
j this determination, we have further concluded that the amenaments involve an action which is insignificant from the standpoint of j
environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environ-t j
mental impact appraisal need not be prepared in connection with the
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issuance of these amendments, f
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Conclusion We have concluded, based on the considerations discussed above, I
that:
(1) because the amendments do not involve a significant
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increase in the probability or consequences of accidents previously i
considered and do not involve a significant decrease in a safety i
margin, the amencnent.; do not involve a significant hazards consideration, j
(2) there is reasonaole 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 cenpliance 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 7
l and safety of the public, j
i Da te; November 1, 1973 I
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