ML19317E860
| ML19317E860 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/22/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317E858 | List: |
| References | |
| NUDOCS 8001031035 | |
| Download: ML19317E860 (12) | |
Text
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SAFETY EVAllRTION BY THE OFFICE OF NUCLEA
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SUPPORTING AENDENT NO. 34 TO FACILITY LICE AENDMENT NO. 34 TO FACILITY LICENSE NO. DPR-47
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AENDMENT N0. 31 TO FACILITY LICENSE NO. DPR-55
_ DUKE POWER COPPANY
_0CONEE NUCLEAR STATION, INITS NOS.1, 2, AND 3
_ DOCKETS NOS. 50-269, 50-270, AND 50-287 Introduction By letter dated July 21,:1976, October 1 % October 2 0, and Octoberas supplemented August 20, October 7 20,1976, Specifications appended to Facility Operating Licen DPR-47, and DPR-55 for Units Nos.1, 2, and 3.
. DPR-38 which apply only to Unit 3, would permit operation of Unit No. 3 asThe reloaded for Cycle 2 operation.
performed are the Final Acceptance Criteria (FAC) for Emerg Cooling Systems, as required by the Commission's Order for Modif of License dated December 27, 1974.
single failure criterion. failure criterion was done concurrently with to single failure, the evaluationSince the two plants are identical in regard we made for Unit 2 dated June equally applies to Unit 3.
Technical Specifications and design hardware identified evaluation for Unit 2 for Unit 3 also.
with a 15x15 array of fuel rods.The Oconee Unit No. 3 reactor co i
, each removal of all of the Batch 1 fuel (56 assenblies) and the reloca of the Batch 2 and Batch 3 fuel.
The fresh Batch 4 fuel will occupy primarily the periphery of the core and ei ht locations in its interfor E
- . cycle of Oconee Unit 3 at the rated core power of 2 performed take into account the postulated effects of fuel densificatio The analyses and the Final Acceptance Criteria for Emergen We have.
concieded that Oconee thiit 3 cy Core Cooling Systems.
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'.: Cycle 2 at the rated power level of 2568 Mit.can be operated sa r
presented in this safety evaluation.
Details of our review are g3p
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- Evaluation 1.
Fuel Mechanical Design All of the Cycle 2 fuel asse211es are identical in concept and am mechanically interchangeable. The asse211es are described in the licensee's reload submittal of July 21,1976 as supplemented October.20, 1976. The fresh fuel does have minor modifications to the end fittings to reduce assedly pressure drop and increase the holddown margin. The only effect of these modifications is a slight re-distribution of core flow which is discussed under themal-hydraulic design in Paragraph 4 below. Also, four of the asse211es have a slightly higher enrichment and pellet stack length. These four assed11es were substituted for four of the original assed11es after two of the original assemblies were damaged during handling.
These four assemblies are described in the licensee's October 20, 1976 letter.
Fuel rod cladding creep collapse analyses were perfomed for the three fuel batches for the Cycle 2 core. The calculational methods, assumptions, and data have been previously reviewed and approved by the staff. The CROY computer code (BAW-10084 PA) was used to calculate the time to fuel rod cladding collapse. The most restrictive power profiles the new fuel asse211es may be exposed to were used in the analyses.
Conservative values were used for the cladding thickness and ovality and no credit was taken for fission gas release which yields conservative net differential pressures. Also, batches 2 and 3 cladding temperatures were calculated using outlet temperature which is also conservative. Based on the analyses perfomed, the fuel rod design has been shown to meet the required design life limits for fuel cladding creep collapse and is therefore acceptable.
From the viewpoint of cladding stress, Batches 2, 3, and 4 are identical.
The Batch 4 fuel asse211es are not new in concept and previously approved methods of analysis were used to analyze the mechanical perfonnance of the fuel. Also, this design was used in Oconee 2 Cycle 2 which we approved on June 30,1976. Based on our review.
un conclude that the fuel desigt is aWable.
2.
Thermal Desian The fuel thermal design analysis was performed using tSe TAFY-3 computer code, as described la "TAFY - Fuel Pia Temperatura and Gas Pressure Analysis." BAW-10044, My 1972.
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. As part of our interim evaluation of the TAfrY code, the following modifications to the code were approved for use in " Technical R on Densification of Babcock & Wficox Reactor Fuels", July 6.1973 (1) a code option for no nstrQcturing of the fuel.
(2)' calculated gap conductance was reduced by 25%.
Using the TAFY code, the damage threshold o any value expected during normal operation, anticipated operat transients, or a LOCA.
Based on our review, we conclude that the fuel thermal design for Cycle 2 is acceptable.
3.
Nuclear Design analysis The reactor core physics parameters for Cycle 2 operation were s
calculated using the PDQ07 computer code which has been previou Since the com has not yet reached an equilibrium cycle, the minor differences in the physics parameter approved by us for use.
which exist between the Cycle 1 and Cycle 2 cores are to be e and are not significant.
In view of the above and the fact that sta we find the licensee's nuclear design analys acceptable.
Thermal-Hydraulic Analysis 4.
The Mark B4 (Batch 4) assenbly differs from the Mark C3 (B This produces assembly primarily in the design of the end fitting.
Introducing a slightly smaller flow resistance for the B4 assenblies.
B4 assenblies into the core causes a slight change in the core f To obtain distribution, which we conclude to be a negligible effect.
the Cycle 2 core flow distribution, the themal-hydrau utilized the actual the hottest core locations.
The Raab coolant flow was measund durfag Cycle 1 operation.For the C measured flow was 1101 of the design flow. hydra 7
design dich is consistent with Units 1 and 2.
'.as it includes adequate conservatisms re l
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hydraulic calculations was accompanied by a correspon a
, the cors inlet temperature from 554 to 555.9F.
and inlet temperature are changes in calculational param p
do not represent changes in operation of the plant.
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indicates that the
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,. margin to DNB is greater for Cycle 2 than had been predicted for Cycle 1 operation.
The DNBR analysis for Cycle 2 operation considered maximum design conditions, as-built fuel assedly geometry, and hot operating conditions. This analysis resulted in the hot channel (Batch 3 fuel) minimum DNBR of 1.98 of 112% power. for undensified fuel. The
' DNBR calculations for undensified fuel are based on a 144-inch active length.
J The shortened stack length used in a second analysis for densified fuel was 141.12 inches. Although this is longer than the densified stack length of the Batch 3 fuel (140.30 inches) the gap size and power spike magnitude were large enough to give conservative results.
The densification effect results in a 5.93% reduction in the minimum DNBR. The minimum DNBR for Cycle 2 considering this effect, is still greater than for Cycle 1.
f Rod Bow i
An analysis was perfomed with the COBRA III-C code to determine l
the effect of a fuel rod bowing into the hot channel and reducing l
l its flow area. The results indicate that rod bow of the magnitude j
predicted is adequately compensated for by the flow area reduction factor. Rod bow away from the hot channel was also analyzed.
In this analysis the effect of a power spike was added to the hot rod in the area of the minimum DNBR. This analysis indicates that Cycle 2 DNBR nsults account for the effects of fuel rod bowing.
Core Vent Valve In the past, a 4.6% reactor coolant flow penalty had been assumed in the themal-hydraulic design analysis for the Oconee units. This penalty was assessed to allow for the potential of a core vent valve being stuck open during nomal operation. The core vent valves are incorporated into the design of the reactor internals to preclude the possibility ~of a vapor lock developing in the core following a postulated cold-leg break. By letter dated January 30, 1976, we advised the licensee that we had concluded that sufficient evidence had been provided by SIM to assure that the core vent valves would vennin closed during normal operation and that it could. therefom.
stadt an application for a license amendment to eliminate the vent valve flow penalty. In addition. the sendttal should include
. appspriate surveillance requinments to demonstrate, each refueling
~ utage. that the vent valves est not stuck open and that they operate o
.. freely.. By letter dated June 11. 1976, the licensee proposed
.' surveillance requirements.
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-S-Our letter dated June 30, 1976, issued the license amendments applying these surveillance requirements to all units. By letter dated August 20, 1976, the licensee requested that the requirement for a flow penalty be removed for Unit 3 Since the June 30, 1976 amendments provided for the necessary surveillance, we find the licensee's request to remove this flow penalty to be acceptable.
Critical Heat Flux cornlation (CHF)
The W-3 CHF correlation was used for the Unit 3 Cycle 1 core. The BAW-2 correlation has been reviewed and approved for use with the Mark B fuel assembly design.
In the application to.the Oconee 3, Cycle 2 core, two modifications, which have also been applied to the Oconee 1. Cycle 3, and Oconee 2. Cycle 2 cores, have been instituted.
1.
The pressure range applicable to the correlation has been extended downward from 2000 to 1750 psia.
2.
The limiting design DNBR of 1.30 was used. This corresponds to a 95% probability at a 95% confidence level that DNB will not occur.
Item 1. above, was based on a review of rod bundle CHF data taken at pressures below 2000 psia which indicate that the BAW-2 correlation conservatively predicts the data in this range.
Item 2. above is consistent with the standard review plan and industry practice.
We have previously reviewed the modifications identified above to the BAW-2 correlation and have concluded that they are acceptable for use in the Unit No. 3 analysis.
In addition, we recently completed a reevaluation of the BAW-2 CHF correlation to verify its continued suitability in relation to available rod bundle data. We determined that the BAW-2 correlation continues to be an acceptable correlation over the pressure, quality, massflux, rod diameter and rod spacing range of its original data base.
1 In sunnary the licensee has proposed a reactor coolant flow rate consistent with Units 1 and 2 for the Unit 3. Cycle 2 thermal-hydraulic analysis. The licensee has also requested elimination of a 4.6% went valve flow penalty. Based on our review. we fiave concluded it' the licensee has included.cy.ey. iate tenservatisms in its analysis rad that existing Technical Specifications provide added assurance that the reactor coolant flow is properly monitored. Based on the above we find that the thermal-hydraulic analysis is acceptable and that the Technical Specificatio..s related to the Cycle 2 thermal-hydraulic analysis, as proposed in the July 21. 1976 submittal are also acceptable.
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. 5.. Accident and Transient Analysis Each FSAR accident and transient analysis was reviewed. In all cases the importent parameters are bounded by PSAR assumed parameters or the results are conservative with respect to the PSAR and reference cycle analyses. Therefore, we conclude that the accident and transient enalyses are adequate.
6.
Startup Program The startup program tests will verify that the core perfomance is within the assumption of the safety analysis and will provide the necessary data for continued plant operation. The licensee has agreed by letter dated October 20, 1976, to provide certain confimatory infor-mation from the startup program. We find this to be acceptable.
7.
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On Decenber 27, 1974, the Atomic Energy Commission issued an Order for Modification of License implementing the requirements of 10 CFR 50.46, " Acceptance Criteria and Emergency Core Cooling Systems for Light Water Nuclear Power Reactors." One of the requirements of the Order was that the licensee shall submit a re-evaluation of ECCS l
cooling performance calculated in accordance with an acceptable evaluation model which conforms with the provisions of 10 CFR 50.46.
The Order also requimd that the evaluation shall be accompanied by such proposed changes in Technical Specifications of other license amendments as may be necessary to implement the evaluation msults. As required by the Order, the licensee, by letter dated Jhly 9,1975 as supplemented August 1,1975, submitted an ECCS reevaluation and nlated Technical Specifications.
In the reload application of July 21, 1976, the licensee has submitted the related Technical Specifications using the B&W ECCS evaluation model as described in BAW-10104 of May 1975.
The background ef-our review of the B&W ECCS evaluation model and its
.. application to Oconee is described la nur-Safety Evaluation Itaport for this facility dated December 27,1974, issued in connetetion with the.
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' Order for Modification of License. The bases for acceptance of the r, principal portions of the evaluation model are set forth in our
' Status Report of October 1974 and the Supplement to.the Status Report
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'of November 1974 which are referenced in the Decenter 27,1974 SER.
That SER describes the various changes required in the earlier version of the B&W sedel. Together, that SER, the Status Report and its Supplement describe an acceptable ECCS evaluation rodel ad the basis
for our acceptance of the model.
The Oconee 3 ECCS evaluation which is covend by this safety evaluation report properly confoms to the accepted model. The licensee's July 9,1975 submittal contains documentation by reference to B&W Topical Reports of the revised ECCS model (with the modifications described in our December 27.1974 SER) and a generic break spectrum ap(Revised April 1976), respectively.
propriate to Oconee 3; BAW-10104, May 1975 and BAW-10103 June 1975
'The generic annlysis in BAW-10103 identified the worst break size as the 8455 ft4 double-ended cold leg break at the pump discharge with a CD = 1.0.
The table below sumarizes the results of the LOCA limit analyses which determine the allowable linear heat rate limits as a function of elevation in the com for Oconee Unit 3:
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Elevation LOCA Peak Cladding Max. Local Time of 1
(ft)
Limit Temperature (OF)
Oxidation Rupture (kw/ft)
Ruptured Unruptured
(%)
(sec)
Node
, Node Oconee 3 2
15.5 2002 1978 3.92 12.25 4
16.6 2136 2072 4.59 13.01 6
18.0 2066 2146 5.46 14.55 8
17.0 1742 2110 5.19 14.01 10*
16.0 1642 1931 2.93 39.20
- See discussion below.
The maximum com-wide metal-water reaction for Oconee 3 was calculated to be 0.557 percent, a value which is below the allowable limit of 1 percent.
As shown in the tabulation, the calculated values for the peak clad temperatum and local metal-water reaction were below the allowable limits specified in 10 CPR 50.46 of 22000F and 17 percent, mspectively.
t BMW-19153 has also shown that the core gennetry remains==nahle-to l
cooling and that long-tem com cooling can be establi,shed.
We noted during our review of BAW-10103 that the LOCA limit
. calculation at the 10-foot elevation in the core showed aflood rates below 1 isch /second, 251 seconds fato the accident (Section 7.3.5).
Appendix K to 10 CFR 50.46 requins that when reflood rates am less than 1 inch /second. heat transfer calculations shall be based on the assumption that cooling is only by steam, and shall take into account
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any flow blockage calculated to occur as a result of cladding swelling or rupture as such blockage might affect both local steam flow and heat in the Status Report of October 1974 transfer. As indicated by us and supplement of Novenber 1974, a steam cooling model for reflood rates The less tian 1 inch /second was not submitted by B&W for our review.
steam cooling model submitted by B&W in BAW-10103 is therefore considered review and ACRS to be a proposed model change requiring our furtherAccordingly, consideration.
steam cooling model is reviewed, the heat transfer calculation at the 10-foot elevation during the period of steam cooling specified en BAW-10103 must In lieu of using their proposed steam cooling model, be further justified.
B&W has submitted the results of calculations at the 10-foot elevation using adiabatic heatup during the steamecooling period, where this period is defined by B&W as the time when the reflood rate first goes below 1 inch /
second to the time that REFLOOD predicts the 10-foot elevation is covered The new calculated peak cladding temperature, local by solid water.
metal-water reaction and core-wide metal-water reaction at the 10-foot
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elevation are 19460F, 3.021, and.647% respectively. These values remain us.
below the allowable limits of 10 CFR 50.46 and are acceptable to Until a steam cooling model has been accepted by us, these values will serve as the LOCA results for oconee 2 at the 10-foot elevation.
We have reviewed the Technical Specifications proposed by the licensee in the July 9,1975 submittal, to assure that operation of Oconee Unit 3 will be within the limits imposed by the Final Acceptance Criteria (FAC)
These criteria pemit an increase in the for ECCS system performance.
allowable heat generation rate from 15 to 16 kw/(ft at the 10 foot e comoared to the Interim Acceptance Criteria IAC). For Unit 3, the LOCA-related heat generation limits am bounded by the generic limit of as 18.0 kw/ft as contained in BAW-10103. We have concluded that the proposed Technical Specifications, as submitted for Unit 3, Cycle 1 operation meet Since Oconee Unit 3 is currently the necessary FAC and are acceptable.
undergoing refueling for Cycle 2 operation, we have also reviewed the proposed Technical Specifications for Cycle 2 operation to assure that We have determined that the LOCA related heat they also meet the FAC.
generation limits used in the BAW-10103 LOCA limits analysis are con-Based on th servative compared to those calculated for this reload.
above, we find that the proposed Technical Specifications for Cycle 2 cperation also meet the FAC-of EMS performance and are therefom i
a & dele.
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Our myfew of other plant-specific assugtfons discuss'ed in the fo11cring
- paragraphs regarding Oconee 3 analyses addressed the areas of single t
-7 failure criterion long-term baron conwir.iion, Wilal semerged n
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..,c equipumat, partial loop operation. emergency electrical power and the cont i
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sent pressun calculation.
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Single Failure Criterion _
Appendix K to 10 CFR 50 of the Comission's regulations r combination of ECCS subsystems to be assum occurred.
j Our review of the Unit 3 ECCS single failure criterion was done concurrently with the review of the Unit 2 single failure criterion.
Since the two plants are identical in regard to single failure, thee 30,1976, evaluation we made for Unit 2, dated June to Unit 3.
One of our requiremer.ts in the Unit 2 safety-evaluation was that valves LP-21 and LP-22 would be left in the open position d operation to minimize the potential for a water hamer due to th By letter dated August 20,1976, discharge of ECC water into a dry line.the licensee comitte Based on our review of the single failure criterion *, we conclud
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the criterion has been met and is therefore acceptable.
Emergency Electric Power _
The design of the power distribution system for the Ocon consists of two 87.5 MVA hydroelectric power generators at K that serve as onsite emergency power sources.
Oconee units is capable of supplying all the essential loads of all the There are two diverse methods of feeding em Units.
of the thrre Oconee Units. Dam through the 230KV site switch fomers whenever offsite power is unava down transfonner, redundant feeder breakers (SK1 and SK buses.
In addition to the two Keowee hydro units t
Station via an independent overhead 100KV transmission system l
hr evaluation of the Unit 2 amargency alactrihpower sy We have concluded that the design I 1976, applies to the. Unit-3 as mell.of the electric pow i l CCCS of either Units 2 or 3
,,f electric i.+,,u.t would not preclude the Dur. conclusion ans_ based in part.
from perforwing its function.
. seismic qualification of the Keowee O'verheadthe M the
.15g earthquate referred to in theJconee FSAR Unit 3.
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. t ted that although the s possible, the complexity, The licensee, by letter dated October 7.1976, s a l d d completion of the l
analyses are being completed as expedittous y a diversity, and vintage of the equipment has prec u e The licensee has spired.
tasks in the short period of time which has tranof the i
provided a schedule which shows complet on is forthcoming on a 1977.
i We conclude that since the confirmatory informat oneven l
reasonable schedule and a seismicthat it is acceptable for U probability,irmatory information.
GT this conf Submerged Electrical Equioment.
l to that performed for Unit 2.
lies to Unit 3 also, The Unit 3 review and evaluation are identica 30, 1976, app Our safety Evaluation issued on June
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and is acceptable.
Single Failure Conclusion indicated changes to Technic find that there is On the basis of our review, including the i
mponent level has sufficient assurance that the ECCS will rem
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occurred.
Containment Pressure.
licable to Unit 3 also. Th 30,1976, is app Class plants were Our Safety Evaluation dated June ECCS containment pressure calculations f this type as described in d
performed generically by B&W for reactors o d of November 1974.
BAW-10103 of June 1975.in the Status Report of Oc used for the ECCS i
We have concluded that the plant-dependent inform vative and, therefore.
containment pressure analysis for Oconee 3 is conser the calculated containment pressure are in accor 10 CFR 50 of the Comission's regulations.
l i m Ters Borce E m etration steur designed for tJe have reviewed the proposed procedures an reactor vessel during l
h preventing excessive boric acid buildups in t e f procedures for Unit-the long-term coolfag period after a LOCA.
i 1975, the licensee committed to the implanentation 3 which wculd allow adequate boron dilutionle failure crite h..J i
..uhich w'.11 comply with the s ng A
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As indicated in our June 30, 1976 Safety Evaluation and our letter dated October 4,1976, we concluded that the proposed procedures and modifi-cations are acceptable for preventing long-term boron concentration provided that some type of flow indication is provided on the hot leg drain. lines. We indicated that the nextrefueling cycle would be acceptable for installation on Unit 3 since we required testing of the hot leg drain'^ system prior to cycle 2 startup. The licensee.has
... comitted to this by letter dated October 19, 1976. We find this to be acceptable.
Partial Loop Analysis Our Safety Evaluation dated June 30, 1976, evaluated the operating mode of one idle reactor coolant pump and showed that this mode is supported by a LOCA analysis performed in accordance with Appendix K of 10 CFR 50.
An analysis of ECCS cooling perfomance with one idle reactor coolant l
pump in each loop was not submitted and power operation in this configuration was limited by Technical Specifications to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
t The June 30, 1976 evaluation is applicable to Unit 3 and we conclude that this mode of operating-is acceptable as indicated above.
a We have completed the review of the Oconee 3 ECCS perfomance re-analysis and have concluded:
(a) The proposed Technical Specifications are based on a LOCA analysis performed in accordance with Appendix K to 10 CFR 50.
(b) The ECCS minimum containment pressure calculations were perfomed in accordance with Appendix K to 10 CFR 50.
(c) The single failure criterion will be satisfied.
(d) The proposed procedures for long-tem cooling after a LOCA are acceptable. The implementation of these procedures during the Cycle 3 refueling outage is required to provide assurance that the ECCS can be operated in a manner which would prevent excessive boric acid concentration from occurring. A comitment by the
'.11cansas to install the positive indication.to show that the hat
' leg drain natuork is working during post-LOCA conditions is i
required and has been received by letter dated October 19, 1976.
1 (e)~ The proposed mode of reactor operation with one idle reactor
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l coolant pump is ww.ited by a LOCA analysis performed in accordance with Appendix K to 10 CFR 50. Operation with one idle
- pep in each loop is restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Requests for single l
loop operation will be reviewed on a case-by-case basis. - - -
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l We have completed our evaluation of the licensee's Unit 3 Cycle 2 reload application and conclude that the licensee has perfomed the required analyses and has shown that operation of the Cycle 3 core will be within applicable fuel design and perfomance criteria.
In addition, we conclude that the licensee's proposed Technical Specification changes meet the Final Acceptance Criteria based on an acceptable ECCS model conforming to the requirements of 10 CFR 50.46 and that the restrictions imposed on the facility by the Comission's December 27, 1974 Order for Modification of License should be terminated and replaced by the limitations established in accordance with 10 CFR 50.46.
We have determined that the 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 the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 151.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
. 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 Comissia.'s regulations and the issuance of these amendments will not be inimical to the comon defense and security or to the health and safety of the public.
Date:
October 22, 1976 m.ae w M
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.. - UNITED STATES NUCLEAR REGULATORY CDPM I)OCKETS N05. 50-269. 50-270. AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMEN 0MENTS TO FACILITY OPERATING LICENSES The U.S. Nuclear Regulatory Commission (the Commission) has is to Facility Operating Licenses Nos.
Amendments Nos. 34, 34 and 31 DPR-38, DPR-47 and DPR-55, respectively, issued to Duke Power Comp which revised the licenses for operation of the Oconee Nuclear Statio The Units Nos.1, 2 and 3, located in deonee County, South Carolina.
amenttnents are effective as of the date of issuance.
These amendments (1) revise the Technical Specifications to for Unit 3 Cycle 2 operation based upon establish operating limits an acceptable Emergency Core Cooling System evaluation model
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to the requirements of 10 CFR Section 50.46 and (2) terminate the operating restrictions imposed on Unit 3 by the Comission's D s
1974 Order for Modification of License.
The application for the amendments complies with the standa
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requirements 'of the Atomic Energy Act of 1954, as amended the Consission's rules and regulations. The Commission has made
.yy.e date findings as required by the Act and the Commissio r
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'and regulations la 10 tR thaptar 1. J1dt are sat forth-in the license
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amansbants.' Natice of Proposed Issuance of hht to Facility Op License No. DPR-55 in connection with this action was publishe Es request for a FEDERAL REGISTER on Septsaber 15,1976 (41 R 39848).
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f 2-hearing or petition for leave to intervene was filed following notice of the proposed action.
The Commission has determined that the issuance of these amendments will not result in any significant environmental impact and that pursuant to 10 CFR 551.5(d)(4) an environmental impact statement or negative declaration and environmental impact a,,praisal need not be prepared in connection with the issuance of these amendments.
For further details with respect to this action, see (1) the
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application for amendments dated July 21,1976, as supplemertted August 20. October 7. October 19, October 20, and October 20, 1976, (2) Amendments Nos. 34,34 and 31 to Licenses Nos. DPR-38, DPR-47 and DPR-55, respectively and (3) the Comission's related Safety Evaluation. All of these iten's are available for public inspection at the Comission's Public Document Room,1717 H Street, NW., Washington, D.C. and at the Oconee County Library, 201 South Spring Street, Walhalla, South Carolina 29691. A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Comission, Washington, D.C.
20555, Attention: Director,. Divisior: of Operating Reactors.
Dated at Bethesda, Maryland, this 22nd day of October 1976.
FOR THE LEAR REGULATORY C0ftUSSION
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A. Schwencer Chief '
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Operating Reactors Branch #1 Division of Operating Reactors i
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