ML19312C388
| ML19312C388 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/29/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C379 | List: |
| References | |
| NUDOCS 7912130915 | |
| Download: ML19312C388 (11) | |
Text
UNITED STATES i
NUCLEAR REGULATORY COMMISSION WASH?NGTON, D. C. 20666 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 45 TO LICENSE NO. OPR-38 AMENDMENT NO. 45 TO LICENSE NO. DPR-47 AND AMENDMENT NO. 42 TO LICENSE NO. DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION UNITS NOS. 1, 2 AND 3 DOCXETS NOS. 50-269, 50-270 AND 50-287 Introduction By letter dated May 6, 1977(I)and as supplemented June 21, 1977(2) and July 11,1977, Duke Power Company (the licensee) requested changes to the Technical Specifications appended to the Facility Operating License for the Oconee Nuclear Station, Unit 2 (Oconee 2)
The proposed changes would
. By letter dated permit Oconee({21Qperation as reloaded for Cyc 10}1, March 1, 1977 as supplemented May 5, 1977 the licensee also requested a change to Technical Specifications appended to the Facility Operating Licenses which would revise the reactor internal vent valves testing program for all three Oconee Units.
Evaluation The Oconee 2 reactor core consists of 177 fuel assemblies. The reload for Cycle 3 will involve the removal of all 61 Batch 2 fuel assentlies and 12 of the Batch 3 fuel assemblies, and the relocation of the residual Batch 3 and Batch 4 fuel assemblies.
The removed fuel will be replaced by 5 Batch 1 fuel assentlies from Oconee 2,12 Batch 1 fuel asse211es from Oconee 3, and 56 new Batch 5 fuel assent 11es. These assenclies will occupy primarily the periphery of the core and four locations in the interior regions of the core.
All but four of the Cycle 3 core fuel assentlies have a 15x15 array of fuel rods.
Of these four, two of the Batch 4 and two of the new Batch 5 fuel assentlies are demonstration Mark C & CR assentlies, respectively.
Each of these demonstration asse211es consist of a 17x17 array of fuel rods. A description of the progrgg to irradiate the assentlies was provided by letter dated January 28, 1976Wi.
In addition, a Babcock & Wilcox (B&W) report on the rradiation of 17x17 demonstration assemblies in Oconee 2 of January 1976(j1, was provided which describes the mechanical, nuclear and thermal-hydraulic characteristics of these demonstration assecblies.
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. The licensee's reload analyses and Technical Specification changes submitted by letter dated May 6,1977 were based on an originally planned 296 effective full power days (EFPD) of Oconee 2 Cycle 2 opp,rgtion. The licensee, however, advised us by letter dated June 21, 1977td/ that Cycle 2 operation was being terminated early at 277 EFPD.
As a result, the burnup distribution in the Batch 3 and 4 fuel assemblies, which are to remain in the core for Cycle 3 operation, will be different from that assumed in the original reload analysis. This June 21, 1977 submittal also included changes to the reload configuration (Removal of the 12 Batch 3 fuel assemblies and insertion of the 12 Batch 1 fuel asseablies from Oconee 3).
Based on a reanalysis of the new burnup distribution of the Batch 3 and 4 fuel assemblies, and the change in the fuel reload configuration, the licensee submitted revisions to the reload report and Technical Specifications.
Fuel Mechanical Design Table 4-1 of Reference 5 summarizes the reload core fuel assembly parameters.
The Batch 5,15x15, (Mark B-4) fuel (ggsembly design has been reviewed and accepted by us for use in Oconee 2.
/ This type of assembly is currently operating in Oconee 2.
The 56 new Batch 5 fuel assemblies, therefore do not represent any unreviewed change in mechanical design from the reference cycle.
Five of the new Cycle 3 replacement fuel assemblies are once-burned Batch I fuel assemblies.
These assemblies were removed from the Oconee 2 reactor following Cycle 1 operation and have been stored in the spent fuel pool during Cycle 2.
These assemblies are of the Mark B-2 type and have been previously reviewed and approved by us for operation in Oconee 2.
They do not significantly change the mechanical design of the Cycle 3 core.
Twelve of the Cycle 3 replacement fuel assemblies are once-burned Batch 1X fuel assemblies removed from Oconee 3 (Batch IX).
These assemblies are of the Mark B-3 design similar to the twice-burned, Batch 3, Oconee 2 assemblies. This mechanical design which has been previously reviewed and approved for use in Oconee 2, does not significantly change the mechanical design of the Cycle 3 core.
As stated earlier, there are four demonstration fuel assemblies proposed for Cycle 3 operation in Oconee 2.
Two of these are Mark C fuel assemblies and two are Mark CR fuel assemblies. The Mark CR assemblies were placed in the Cycle 2 core, and will continue irradiation in the Cycle 3 core.
These assemblies have a 17x17 fuel rod configuration. There are two different length fuel pellets used in these 17x17 assemblies. The fuel rod outside and inside diameters have been decreased in the Mark C assemblies. The Mark C assemblies are mechanically compatible and l
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. interchangeable with Mark B type assemblies with the exception of the reviewed and found acceptable for operation in econee 2.lggn previously control rod component interface. These assedlies have h
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The two demonstration Mark CR fuel asse211es#) and the Mark C fuel assemblies are identical except that the Mark CR assent 11es have re-constitutable lower end fittings.
The reconstitutable feature is provided by positioning the lower end fitting to the lower grid by flange sleeves on the guide tubes rather than by welding the lower grid to the lower end fitting.
Also, the lower end fitting is fastened to the guide tubes by torque nuts as in the Mark C demonstration assemblies; however, the nuts are prevented from rotating by swaged locking cups rather than by welding.
The cups are based on a retainer plate that is restrained flush against the lower end fitting by a guide tube nut.
The retainer plate cup brazement captures all 24 nuts by means of deformed metal tangs, so all nuts have to be untorqued before the brazement, including nuts, can be removed.
The Mark CR design has been subjected to a 300 hour0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> test at simulated reactor full power conditions with no deterioration or wear of any of the parts of the retainer system (brazement nuts, lower end fittings, flange sleeves).
Cold water tests were perfonned and the Mark CR design was found to have the same resonant frequency and amplitude as the Mark C design.
Bench tests have been perfonned on the retainer system to assure the locking cups are securely brazed to the nut plate and the swaged cups will prevent loosening of the nuts which are captured within the nut plate brazement. The Mark CR demonstration assentlies have been designed to maintain their structural integrity through three cycles of operation and to successfully withstand seismic and loss-of-coolant loads. This reconstitutable mechan; cal locking configuration has been used in a similar function (new B&W tube specimen holder design); tested under simulated conditions, and completely analyzed and is being used in other B&W reactors. Therefore, the use of the reconstitutable lower end fitting is acceptable for the demonstration Mark CR assemblies.
These mechanical design variations have been taken into account in the various mechanical analyses. The Batch 3 fuel is generally limiting, because of it's low initial fuel pellet density, low initial pre-pressure, and previous incore exposure. The results of these analyses have shown that the mechanical design differences in the Oconee 2 Cycle 3 fuel assemblies are of negligible affect and are acceptable.
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. Fuel rod cladding creep collapse analyses were performed for the fuel batches which will be present in the Cycle 3 core.
The calculational method $ qssumptions, and data have been previously reviewed and approved by us.tl31 The CROV computer code was used to calculate the time to fuel rod cladding creep collapse. The most restrictive power profiles, to which the once-burned and new fuel assemblies may be exposed, were used in the Batch 4 and Batch 5 analysis. The actual reactor operating history along with the most restrictive power histories for the forthcoming cycle were used in the analyses of the Batch 3 fuel.
The fuel cladding material properties are the same as those used in the CROY code. The analysis assumed no fission gas production (maximum differential pressure),
lower tolerance limit on cladding thickness, and upper tolerance limit on cladding ovality.
Based on the analyses performed, the fuel rod design has been shown to meet the required design life limits for fuel cladding creep collapse and is, therefore, acceptable.
Frem the viewpoint of cladding stress and strain Cycle 3 operation is acceptable. The cladding stress (creep stress due to differential pressure, thermal stress due to temperature gradient and bending stress due to axial loads and restraints) will not exceed the yield stress or ultimate strength of the cladding material. The Batch 3 fuel is most limiting with respect to stress, because of its low prepressurization and density.
The cladding strain for Cycle 3 operation is less than the generally used 1% plastic strain acceptance criteria. The strain analysis assumed maximum specification value for the fuel pellet diameter, density, and burnup, and minimum specification tolerance on fuel cladding inside diameter. These assunptions conservatively represent the cladding strain.
The Batch 3 fuel will again be limiting in the Cycle 3 core based on the cladding strain. Again this is because of its irradiation history, lower prepressurization, and lower fuel pellet density.
The Batch 5 fuel assentlies are not new in concept and do not use different component materials.
The fuel assentlies for Cycle 3 operation will not exceed their design life limits.
In addition, it.has been shown that the presence of the demonstration arsentlies in the Cycle 3 core will have an insignificant effect on operation. We conclude therefore that the fuel mechanical design for Cycle 3 operation is acceptable.
Fuel Thermal Design Thefgs thermal design analysis was conducted using the TAFY-3 computer code.
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This analysis established heat flux limits to fuel centerline melt. The analysis cppidered the effect of a power spike from fuel pellet densification.lol Modifications consisting of change to the void probability F, and size distribution. F, have been previous reviewed g
k and approved by us for Oconee 2 fuel thermal design analysis.1 This '
analysis is based on the lower tolerance limit on fuel density and ass mes isotropic diametral densification shrinkage and anisotropic axial snrinkage densification. The calculated gap conductance was reduced by 25% in accordance with ou nterim evaluation of TAFY. These assumptions havebeenapprovedbyus.{9
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. During Cycle 3 operation, the highest rele.cive asse21y power levels occur in Batch 1 and 3 fuel asse2 11es. The fuel temperature analysis for Cycle 3 is based on limiting beginning-cF-cycle (BOC) conditions (zero burnup) and conservative peaking factors. The analysis is performed to establish linear heat generation rates to preclude central fuel melting and stored energy limits for LOCA analyses.
Although Batch 4 and 5 fuel assed11es have a reduced active fuel length and greater linear heat generation rate, the maximum predicted centerline temperature of this fuel is lower than that of Batch 1 and 3 fuel assedlies. The maximum predicted centerline temperature for the Batch 1X fuel assed11es was also lower than that predicted for the Batch 1 and 3 fuel asse211es. This is due to the higher initial density of the Batches IX, 4 and 5 fuel asseelies.
Therefore, the therTnal design analysis for the Batch 1 and 3 fuel assedlies thennal design analysis is bounding, and we conclude that the fuel thermal design for Oconee 2 Cycle 3 core is acceptable.
I Nuclear Analysis The reactor core physics parameters for Oconee 2 Cycle 3 operation were I
calculated using a PDQ07 computer code. Since the cere has not yet reached an equilibrium cycle, there were minor differences in the physics parameters between the Cycle 2 and Cycle 3 cores.
For example, EOC Doppler and moderator coefficient changes by less than 1% from Cycle 2 to Cycle 3.
These changes are to be expected and are not significant.
The effects of the four demonstration fuel assemblies on the Cycle 3 nuclear design have been reviewed and found to be negligible.
In view of the above and the fact that startup tests (to be conducted prior to power operation) will verify that the significant aspects of the core performance are within the assumptions of the safety analysis, we find the licensee's nuclear analysis for Cycle 3 to be acceptable.
Thermal-Hydraulic Analysis l
The major acceptance criteria which we used for the thermal-hydraulic design are specified in Standard Review Plan (SRP) 4.4.
These criteria establish t
acceptable limits on departure from nucleate boiling. The thermal-hydraulic analysis for Oconee 2 Cycle 3 reload were made using previously approved models and methods.
Certain aspects of the thennal-hydraulic des.ign are new for the Cycle 3 core and are discussed below.
Reactor Coolant System Flow Rate The reactor coolant flow rate was accurately measured during Cycle 1 I
operation and determined to be 111.5% of the system design flow. The licensee has proposed to take credit in the Cycle thermal-hydraulic analysis (as was done in Cycle 2) for this higher flow. The licensee will
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. by measurement verify the reactor coolant system flow rate for Oconee 2 Cycle 3.
The licensee will include conservatisms for uncertainties in the mes urement of the flow in the thermal-hydraulic analysis. The licensee has used a flow-rate of 106.5% in the Oconee 2 Cycle 3 analysis with these conservatisms to be consistent with the flow rate used in Oconee 1 themal-hydraulic analysis.
There are differences in the flow resistance between the Mark B-3 fuel asse@ lies and the-Mark B-4 fuel asse211es. The flow resistance for the Mark B-4 fuel asse211es is less than that measured for the Mark B-3 assed11es.
Also, the Mark C and CR assedlies have a greater flow resistance than either of the other two fuel assembly types.
These differences have been analyzed. The Cycle 3 core has flow resistance characteristics that are similar to the Cycle 2 core. The possible introduction of core cross flow due to the different flow resistances has been considered. This phenomenon was shown to be of negligible effect from analyses for the previous cycle as discussed in reference 6.
Fuel Rod Bow In the submittal dated May 6,1977, the licensee sumarized the method and results of the rod bow analysis. This rod bow analysis was perfomed with an as yet unapproved model.
Therefore, the licensee was requested to provide analyses with the NRC approved rod bow model or to show sufficient compensatory margin.
The licensee chose to show sufficient core flow margin in order to offset the difference between models. The approved rod bow model requires a DNBR penalty of approximately 12% as compared to the unapproved rod bow model which has about a 6% DNBR penalty. The 6%
difference in DNBR penalty will be accomodated by approximately a 3%
margin in reactor coolant system flow rate, which is flow margin above the design valve of 106.5%. The required flow rate is specified in the Technica1 > Speci fications. The licensee will verify by measurement the reactor coolant system flow rate.
Free previous experience, we are assured that reactor coolant flow will remain essentially constant throughout the cycle. This approach will assure adequate thermal hydraulic design margin.
Mark C and CR Demonstration Assemblies l
The themal hydraulic design analysis has been based on a core configuration consisting of 177 Mark B (15x15) fuel assemblies.
Comparative analyses have been perfonned to show that the insertion of the Mark C and CR (17x17) demonstration assemblies would have decreased the core thermal hydraulic margin if they were located in the high power region of the core. Therefore, the demonstration assemblies have been placed in low power producing core locations to ensure that these assemblies will not be limiting and to provide minimum impact on the hot assembly performance.
Thus, the two Mark C and the two Mark CR demonstration assemblies are not limiting and their presence in Cycle 3 will not significantly affect the themal-hydraulic characteristics of the reactor.
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s In summary, the licensee has proposed that a reactor coolant flow rate based on actual measured flow with uncertainties taken into account, be used in the Oconee 2 Cycle 3 thermal hydraulic analysis.
The licensee has also assured us that there will be sufficient RCS flow to compensate for the difference between the approved and the unapproved rod bow models.
The licensee has considered the impact of the Mark C and CR demonstration assemblies and has ensured the impact to be negligible.
Based on our review, we find that the licensee has included appropriate conservatisms in its analysis and that the proposed Technical Specifications provide assurance that the criteria of SRP 4.4 will be met.
Therefore, we conclude that the thernal hydraulic analysis as previously approved and discussed are acceptable.
Accident and Transient Analysis The accident and transient analyses provided by the licensee demonstrates that the Oconee FSAR analyses conservatively bounds the predicted conditions of the Oconee 2 Cycle 3 core and are therefore acceptable.
Each FSAR accident analysis has been examined, with respect to changes in Cycle 3 parameters, to determine the effects of the Cycle 3 reload and to ensure that thermal performance is not dagraded during hypothetical transients.
The core thermal parameters used in the FSAR accident analyses were design operating values based on calculated values plus uncertainties.
Cycle 1 values (FSAR values) of core thernal parameters were compared with those used in the Cycle 3 analysis.
For each accident of the FSAR, a discussion and the key parameters were provided. A comparison of the key parameters from the FSAR and Cycle 3 was provided with the accident discussion to show that the initial conditions of the transient are bounded by the FSAR analysis. The effects of fuel densification on the FSAR accident results have been evaluated and are reported in the Oconee 2 fuel densification report. Since Cycle 3 reload fuel assemblies contain fuel rods with theoretical density higher than those considered in this report, the conclusions derived in that report are still valid.
Calculational techniques and methods for Cycle 3 analyses remain consistent with those used for the FSAR. No new dose calculations were performed i
for the reload report. The dose considerations in the FSAR were based on maximum peaking and burnup for all core cycles; therefore, the dose considerations are independent of the reload batch.
Startup Program A startup program will be conducted to verify that the core performance is within the assumptions of the safety analyses and provide the necessary data for continued plant operation.
The startup test is similar to that previously approved for Cycle 2 operation.t6) program Additionally, th,e program was discussed with the licensee for
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8-clarafication of control rod worth and power distribution measurements and comparison to predicted values. These measurements and comparisons will be perfonned by the licensee.
The licensee also will provide a sumary within 90 days following completion of physic testing.
This s'artup test program is acceptable.
ECCS-U-Baffle Pressure Droo By letter dated June ll,1977, the licensee referenced a reanalysis of the ECCS perfomance with revised reactor coolant system pressure drop characteristics (14) using the same ECCS model previously approved for Oconee 2.
This reanalysis was performed because of an identified error in the input value to the reactor vessel inlet nozzle U-baffle pressure loss characteristics.
The reanalysis shows that lower peak cladding temperatures (PCT) would be obtained for the worst break analysis during a postulated LOCA.
The trends of the break spectrum, sensith1ty, and LOCA limits studies for the pruiously approved analysis for Oconee 2 remain valid.
Therefore, only the limiting size break needed reanalysis.
The licensee has confirmed that the reanalysis is appropriate to all three Oconee plants.
The reduction in PCT as compared to that for the generically approved ECCS analysis (B&W-10103) was due to the enhanced core flow during blowdown (more cooling), lower metal-water reaction rates (because of lower temperatures, less heat generation due to exothermic reaction), and improved reflooding of the core (cooling attained sooner). These benefits are based on an improved system pressum distribution, i.e., the reanalyzed RCS pressure drops are less than that assumed from B&W-10103.
The revision to the RCS pressure drops is based on both experimental and analytical verification techniques.
Pressure drop measurements were made during the Oconee 1 hot functional testing. The other two Oconee plants are identical to Oconee 1, so that from this data the Once Through Steam Generator (OTSG) and reactor vessel pressure drops were established for all the Oconee plants.
The pressure drop characteristics within the reactor vessel and the OTSG were then analytically estab}: ghed to match this data.
Additionally, there were vessel model flow testst 31 which further substantiate the decrease in pressure drop observed in the hot functional test data and established by analysis.
The reactor vessel inlet nozzle was originally assumed to be a long leg U-baffle, however, it is not.
As shown by tests, the change in pressure drop for this component l
between originally assumed and-as-built conditions is substantial. All the.qbanges to RCS pressure drops have been verified experimentally and analytically.
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. We have reviewed the RCS pressure drops and their inoact on the ECCS performance analysis. We agree with the licensee that the ECCS calculations for the current Oconee 1, 2 & 3 fuel loading and those submitted by Units 1 and 2 are in compliance with the criteria of 10 CFR 50 Section 50.46 and Appendix K.
Although the reanalysis has lower PCT than those of B&W-10103, the allowable Linear Heat Generation Rate (LHGR) limits for all Oconee plants will be maintained at the same values as previously approved.
l We find this analysis acceptable.
Surveillance Testino of Reactor Internal Vent Valves The licensee proposed a change to the subject testing, such that, the required opening differential pressure for the reactor internals vent valves would be equivalent to 1.0 psid.
The licensee has shown that this change has no significant effect on the peak claqiding temperature (PCT) during the limiting LOCA, i.e., <30F.(10) This is not a significant increase and does not cause the limiting LOCA PCT to exceed any of the 10 CFR 50.46 criteria, nor does this change affect which LOCA break is limiting.
The licensee has supplied this information in reference 10.
Based on this information and the continued surveillance requirements on the reactor internals vent valves we find this change to be acceptable.
The previously discussed analyses, which were presented as justification for operation, were conducted in compliance with NRC's regulations and approved methods and, furthermore, are conservative relative to NRC regulations. The proposed changes to the Technical Specifications are acceptable on the bases that the health and safety of the public will not be endangered by operation in the proposed manner.
Environmental Consideration We have determined that these 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 these amendments involve an action which is insignificant from the standpoint of environmental impact a nd pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
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. References 1.
Letter from W. O. Parker, Jr., (Duke Power Company) to Director NRR (NRC) dated May 6, 1977.
2.
La".er from W. O. Parker, Jr., (Duke Power Company) to Director NRR, dated June 21, 1977.
3.
Letter from W. O. Parker, Jr., (Duke Power Company) to B. C.
Rusche, dated January 28, 1976.
4.
" Irradiation of Two 17x17 Assemblies in Oconee 2, Cycle 2,"
BAW-1424, January 1976.
5.
"0conee Unit 2, Cycle 3-Reload Report", BAW-1452, April, 1977.
6.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments Nos. 27, 27 and 23 to Facility License Nos. DPR-38, DPR-41, and DPR-55 Duke Power Company, Oconee Nuclear Station 2, June 30,1976.
7.
"TAFY-Fuel Pin Temperature and Gas Pressure Analysis," BAW-10044, May 1972.
8.
" Fuel Densification Report," BAW-10055, Revision 1, June 1973.
9.
" Technical Report on Densification of Babcock and Wilcox Reactor Fuels", ONRR, July 6, 1973.
10.
Letter from W. O. Parker, Jr., (Duke Power Company) to Director NRR, dated May 5, 1977.
11.
Letter from W. O. Parker, Jr., (Duke Power Company) to N. C.
Moseley (NRC) dated March 10, 1977.
12.
Letter from W. O. Parker, Jr., (Duke Power Company) to Benard C. Rusche dated March 1, 1977, 13.
Letter from A. Schwencer (NRC) to J. F. Mallary (B&W) dated January 29, 1975.
14 J. H. Taylor (Babcock & Wilcox) letter to R. L. Baer (NRC) dated July 8,1977 15.
Reactor Vessel Model Flow Tests, B&W-10012, October 1969 16.
Letter from W. O. Parker, Jr., (Duke Power Company) to E. G. Case i
dated July 27, 1977.
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3 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKETS NOS. 50-269, 50-270 AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U. S. Nuclear Regulatory Comission (the Comission) has issued Amendments Nos. 45, 45, and 42 to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55, respectively, issued to Duke Power Company (the licensee), which revised Technical Specifications for operation of the Oconee Nuclear Station Units Nos.1, 2 and 3, (the facilities) located in Oconee County, South Carolina.
The amendments are effective as of the date cf issuance.
The amendments revise the Technical Specifications (1) to establish operating limits for Unit 2 Cycle 3 operation and (2) to establish requirements for testing reactor core internal vent valves.
The applications for the amendments comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations.
The Comission has made appropriate findings as required by the Act and the Comission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments.
Prior public notice of these amendments was not required since the amendments do not involve a significant hazards consideration.
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