ML15112A900
| ML15112A900 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/04/1980 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112A898 | List: |
| References | |
| NUDOCS 8001180132 | |
| Download: ML15112A900 (23) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
75 License No. DPR-55
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee),
dated December 13, 1979, as supplemented December 28, 1979, and January 2, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter 7; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-55 is hereby amended to read as follows:
3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 75, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
8 001180
S
-2
- 3. This litcen'se amendment becomes; effectitve wi t-in 5 days, after Its, issuance.
FOR' THE NUCLEAR REGULATORY COMMISSION Robert W. Rkeid, Chief Operating Reactors Branch #4 D4itiions of Operatilng; Reactors
Attachment:
Changes. toi the Techndica.l Specifi caitiions Dte: of lissqance:
January 4), 1980
ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO. 78 TO DPR-38 AMENDMENT NO. 78 TO DPR-47 AMENDMENT NO. 75 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:
Remove the following pages and insert the revised identically numbered pages.
viii ix 2.1-8 2.3-9 3.5-16a 3.5-16b 3.5-19a 3.5-19b 3.5-22a
- 3.
5-22b 3.5-25a 3.5-25b Changes on the revised pages are identified by marginal lines.
LIST OF FIGURES (cONT'D)
Figure 3.1.2-3A Reactor Coolant System Inservice Leak and Hydrostatic Test Heatup and Cooldown Limitation - Unit 1 2.1-7c 3.1.2-3B Reactor Coolant System Inservice Leak and Hydrostatic Test Heatup and Cooldown Limitation -
Unit 2
-. 1-7d 3.1.2-3C Reactor Coolant System Inservice Leak and Hydrostatic Test Heatup and Cooldown Limitation - Unit 3 3.1-7e 3.1.10-1 Limiting Pressure vs. Temperature Curve for 100 STD cc/Liter H20 3.1-22 3.5.2-1A1 Rod Position Limits for Four Pump Operation - Unit 1 3.5.2-1A2 Rod Position Limits for Four Pump Operation Unit 1 3.5.2-181 Rod Position Limits for Four Pump Operation - Unit 2 3.5.2-182 Rod Position Limits for Four Pump Operation - Unit 2 3.5-lea 3.5.2-183 Rod Position Limits for Four Pump Operation -
Unit 2 3.5-16b 3.5.2-101 Rod Position Limits for Four Pump Operation - Unit 3 3.E-17 3.5.2-lC2 Rod Position Limits for Four Pump Operation - Unit 3 3.5-17a 3.5.2-1C3 Rod Position Limits for Four Pump Operation - Unit 3 3.5-17b 3.5.2-2A1 Rod Position Limits for 2 and 3 Pump Operation - Unit 1 3.5-15 3.5.2-2A2 Rod Position Limits for 2 and 3 Pump Operation - Unit 1 2.5-15a 3.5.2-2B1 Rod Position Limits for 2 and 3 Pump Operation - Unit 2 3.E-1 3.5.2-2B2 Rod Position Limits for 2 and 3 Pump Operation - Unit 2 3.5-19a 3.5.2-2B3 Rod Position Limits for 2 and 3 Pump Operation
- Unit 2 3.5-19b 3.5.2-2C1 Rod Position Limits for 2 and 3 Pump Operation - Unit 3 3.5-20 3.5.2-2C2 Rod Position Limits for 2 and 3 Pump Operation - Unit 3 3.5-20a 3.5.2-2C3 Rod Position Limits for 2 and 3 Pump Operation - Unit 3 3.E-2Cb 3.5.2-3A1 Operational Power Imbalance Envelope - Unit 1 3.E-2" 3.5.2-3A2 Operational Power Imbalance Envelope - Unit 1 3.5-2a 3.5.2-381 Operational Power Imbalance Envelope - Unit 2 3.5-22 3.5.2-382 Operational Power Imbalance Envelope - Unit 2 3.5-2a Amendments Nos. 78, 78, & 75 viii
LIST OF FIGURES (CONT"D)
Figure Page 3.5.2-3B3 Operational Power Imbalance Envelope
- Unit 2 3.5-22bi 3.5.2-3C1 Operational Power Imbalance Envelope - Unit 3 3.5-23 3.5.2-3C2 Operational Power Imbalance Envelope - Unit 3 3.5-23a 3.5.2-3C3 Operational Power Imbalance Envelope - Unit 3 3.5-235 3.5.2-4A1 APSR Position Limits - Unit 1 1.5-24 3.5.2-4A2 APSR Position Limits - Unit 1 3.6.24a 3.5.2-4B1 APSR Position Limits - Unit 2 3.5-25 3.5.2-482 APSR Position Limits -
Unit 2 3.5-25a 3.5.2-4B3 APSR Position Limits -
Unit 2 3.5-25b[
3.5.2-4C1 APSR Position Limits -
Unit 3 3.5-26 3.5.2-4C2 APSR Position Limits -
Unit 3 3.5-26a 3.5.2-4C3 APSR Position Limits -
Unit 3 3.5-26b 3.5.2-5 LOCA - Limited Maximum Allowable Linear Heat 3.5-27 3.5.4-1 Incore Instrumentation Specification Axial Imbalance Indication 3.5-31 3.5.4-2 Incore Instrumentation Specification Radial Flux Tilt Indication 3.5-32 3.5-4-3 Incore Instrumentation Specification 1-5-33 4.5-1-1 High Pressure Injection Pump Characteristics 4.5-4 4.5-1-2 Low Pressure Injection Pump Characteristics 4.5-5 4.5.2-1 Acceptance Curve for Reactor Building Spray Pumps
-. 5-9 6.1-1 Station Organization Chart 5.1-7 6.1-2 Management Organization Chart 6.1-8 Amendments Nos.
78, 78 & 75 ix
Rated Thermal Power, %
ONBR Limit
--120 kW./ft Limit kV/ft Limit k~l ft Li it(-33.6,112) 112 (34. 94,112)
M 2 = 5.55
)090 ACCEPTABLE 60 4 PUMP 4'2 OPERATION
(-43,10oo) 100 90 C.3O (D ACCEPTABLE 80 3&4 PUMP
(-43.73*4OPERATIOP 3
83(38,68.3) 0
-20
.0 ACCEPTABLE 2,3i4 PUMP OPERATION EA50
(-43, (384 1 2
-- 40 REACTOR C OOLANT CURVE FLOW, GPM 30 I374,880 2
280,035 20 3
183,690
.. 10
-5O
-50
-40
-30
- 20
-10 0
10 20 30 40 50 G0 Axial Power Imnalance, CORE ?ROTECTION SAFETY LIMITS uUtEowM OCONEE NUCLEAR STATION Amendments Nos. 78, 78 & 75
Thermal Power Level, %
UNACCEPTABLE OPERATION
- 110
(-23.10,105.5) 105.5 (23.714, 105.5)
+1.56 ACCEPTABLE 100.86 4 PUMP OPERATION
(-33,9o)
(26.90)
- 80 78.81.
ACCEPTABLE 364 PUMP OPERATION 70
(-33,53.4)
(26, 63.3) 51070 ACCEPT46LE "50 2,3 & 4 PUMP OPERATION
-- 40
(-33,36. 3)
(6 6 2
-30 20 10
-60
-50
-40
-30
-20
-10 0
10 20 30 40 50 SO Power Imbalance, %
PROTECTIVE SYS:EM U!17: 2
- unrow, OCONEE NUCLEAR STATION Amendments Nos. 78, 78 & 75 2.3-9
110 OPERATION IN i'NIS (133102) 100 REGION IS NOT ALLOWED 2
.253 7.92) 90 80 (235. 80)
POWER LEVEL SHUTDOWN MARGIN CUTOFF 70 LIMIT RESTRICTED 100%
REGION 6M 60 0
((P ERMISSIBLE OPERATING o
o (52. 50)
(168 511)
REGION 40 a
30 S(0,27) 20 10 0
I I1 0
20 40 60 80 100 120 140 180 lEO 200 220 240 260 280 300 Roa Index, '4 M0 0
25 50 75 100 0
25
- 50.
75 100 Group 5 Group 7 0
25 50 100 Group 6 ROD POSITION LIMITS FOR FOUR-PUM' OPERATION From 250 10EFPDTo 307 CONEE 2 EFPD' I m'KEWER; OCONEE NUCLEAR STATION Figure 3.5.2-1B2 Amendments Nos. 78, 78, & 75 3.5-16a
110 OPERATION IN THIS REGION (153,102)
(285,102) 100 IS NOT ALLOWED 90 (2ss.2) 90
-8
,9 SHUTDOWN 80 MARGIN (270, 80)
LIMIT 70 RESTRICTED REGION (66. 50) 60 u
50 r
(168,
- 50) 40 30 PERMISSIBLE 30 OPERATING REGION 20 (12,
I I
I 0
25 50 75 100 0
25 50 75 100 Group 5 Group 7 0
25 50 75 100 Group6 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 297 TO 363 EFPD UNIT 2 OCONEE NUCLEAR STATION Amendments Nos.78, 78, & 75 Figure 3.5.2-1B3 3.5-16b
0 110 OPERATION IN THIS (133. 102 (180 102)
(235. 102) 3:0 102!
100
.REGION IS NOT
.(
0 1 2 ALLOWED WITH 2 OR 3 PUMPS i16E 92) 90 80 SHUTDOWN MARGIN 0
LIMI o+
168.63) 6 0-
'o 50 (52.50) 40 PERMISSIBLE 30 OPERATING REGION (0, 27) 20 10 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Roo Index, % WO 0
25 50 75 100 0
25 50 75 100 Group 5 Group 7 I
1 11 I
0 25 50 75 100 Group 6 ROD POSITION LIMITS FOR TWO AND THREE PWT OPERATION From 250 +
10 EF?D to 307 EFPD OCONEE 2 ouKw
~WE OCONEE NUCLEAR STATION Figure 3.5.2-2B2 Amendments Nos. 78, 78, & 75 3.5-19a
110 (153, 102)
(185 102)
(270, 102)
(300, 102) 100 OPERATION IN THIS REGION IS NOT ALLOWED WITH 2 OR 3 PUMPS (168,92) 90 80 70 60 SHUTDOWN (168,63)
LIMIT 50 (66,50) 40 PERMISSIBLE 30 OPERATING REGION 20 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, W 0 0
25 50 75 100 0
25 50 75 100 Group 5 Group 7 0
25 50 75 100 Group 6 ROD POSITION L:MITS FOR 7o AND THREE PU>!? OPERATION FROM 297 to 363 EFPD UNIT 2 19UKEPOWER OCONEE NUCLEAR STATION Figure 3.5.2-2B3 Amendments Nos. 78, 78, & 75 3.5-1wo9
Power c
ot 2568 MWt 110 RESTRICTED REGION
(-22.
. 102)
\\
-100
(-20.7.92)
( 7.
- 92)
-- 80 PERMISSIBLE OPERATING REGION 70 60 50 40 30 20 10 0
-30
-20
-10 0
10 20 30 ImoaIance, %
OPERATIONAL POWER L'2ALANCE ENVELOPE FOR OPERATION FROM 250 10 EFPDto OCONEE 2 307 EFPD UKEPOWE OCONEE NUCLEAR STATION Figure 3.5.2-3B2 Amendments Nos.
78, 78, & 75 3.5-22a
Power, %of 2568 mwt Ila 100 RESTR ICTED R EG 10 N
(-16. 5,92) 90 (8. 5192)
(-17.9, 80) 80 70 60
(-20. 7, 50) 50 PERMISSIBLE_ 40 OPERATINMG REGION 30 20 10
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power Imbalance, S OPERAT'iONAL POWER 24BAILNCE E~NVELOPE FOR OPEFA.T:ION F-,Om 297 TO 363 EFPD (eoi UNIT
~KPWJOCONEE NUCLEAR STATION Amendments Nos. 78, 78, & 75 35 -22b Figure 3.5..2-3B3
RESTRICTED REGION 110 61 102) 32 1:2) 100 3 2 92 90 (2 3 92) 80 (0 80)
'33 80 70 REST:
C'ED REGIl>
80 PERMISSIBLE S 60 l OPERATING 4-REGION 30 40 (100
- 40) 30 20 10 0
10 20 30 40 50 60 70 B0 90 APSR %a WO APSR POSITION LIMITS FOR OPERATION FROM 250 + 10 EFPD to 307 OCONEE 2 EFPD
'EPQiEwi OCONEE NUCLEAR STATION IDJA POAVFigure 3.5.2-4B2 Amendments Nos. 78,378, & 75 3.5-25a
(6. 1, 102)
(17.5, 102)
RESTRICTED (18.5,92)
REGION 90 (2.3,92)
(30,80) 80 (0,80) 70
~60 C0 (64.4,50) 0 40 PERMISSIBLE OPERATING..
REGION 20 10 0
I 1
1I I
0 10 20 30 40 50 60 70 80 90 100 APSR, WO APSR POSION LIMITS FOR OPERATION FROM 297 TO 363 EFPTD UNIT 2 4 UKEPOWE OCONEE NUCLEAR STATION Amendments Nos.
78, 78, & 75 3.3-25b Figure 3.5.2-4B3
- k~ REGO4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 78 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 78 TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO. 75 TO FACILITY OPERATING LICENSE NO. DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 Introduction By letter dated December 13, 1979, Duke Power Company (DPC) requested amendment of the common Technical Specifications (TS) appended to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55 for the Oconee Nuclear Station, Units Nos. 1, 2 and 3 (ONS-1, 2 & 3).
The requested change would modify the ONS-2 figures relating to core protection safety limits, protective system setpoints, control rod position, power imbalance, and axial power shaping rod (APSR) position consistent with extending the duration of the ONS-2 present operating cycle (Cycle 4) from 297 + 10 effective full power days (EFPD) to 363 EFPD.
Background
By letter dated September 18, 1978, DPC requested amendment of the ONS common TS to provide operating limits consistent with the fuel loading to be used during ONS-2 Cycle 4. The safety analysis supporting this request was contained in the Babcock
& Wilcox report "Oconee Unit 2, Cycle 4 Reload Report", BAW 1491, August 1978, which was included in DPC's September 18, 1978 submittal.
On December 15, 1978, the Commission issued Amendments Nos. 66, 66 and 63 for the ONS-1, 2 & 3, and a supporting Safety Evaluation which revised the TS to support the operation of ONS-2 during Cycle 4. On June 22, 1979, the Commission issued Amendments Nos. 73, 73 and 70 for the ONS-1, 2 and 3, and a supporting Safety Evaluation which revised the TS in regard to power level cut-off.
Initial criticality for Cycle 4 was achieved on December 27, 1978, The 100% power level of 2568 MW(t) was reached on January 9, 1979. It is presently estimated that the current fuel loading will achieve its design cycle length of 297 + 10 EFPD on January 5, 1980. This would normally be the point at which ONS-2 would shut down and begin refueling for operation in Cycle 5.
Due to an extended refueling and modification outage of ONS-1 and the associated difficulties of managing concurrent outages and load considerations, DPC has requested by letter dated December 13, 1979, an extension of their 0NS-2 Cycle 4 by about 66 EFPD.
Based upon an ongoing Commission review on the subject of control rod guide tube wear at nuclear facilities, it was determined that the review of the ONS-2 Cycle 4 extension would include an evaluation on this subject. On December 20, 1979, DPC
-2 presented to the Commission preliminary results of their control rod guide tube inspection program. Additional information was provided by letter dated December 28, 1979.
EVALUATION A comprehensive evaluation of Cycle 4 for its nominal design cycle length is given in our Safety Evaluation of December 15, 1978. This evaluation addresses only those issues which are pertinent to the extension of Cycle 4 from 297 + 10 to 363 EFPD. The main differences are discussed below.
- 1) An increase in the estimated residence time for all fuel batches in the current Cycle 4 fuel load by about 1584 Effective Full Power Hours (EFPH)
(66 EFPD) and its effect on cladding creep collapse:
Babcock & Wilcox (B&W) generic analyses, which have been approved by the Commission, show that the time to rod cladding collapse will be in excess of 30,000 EFPH. At the conclusion of the extended ONS-2 Cycle 4 operation no fuel rod will have accumulated 30,000 EFPH. Therefore, the original Cycle 4 analysis is bounding.
- 2) An increase in the core burnup in MWD/MTU and its effects on rod bow, cladding strain, and fission gas release and associated internal rod pressures:
- a. DPC has applied a rod bow departure from nucleate boiling ratio (DNBR) penalty of 11.2% to all analyses that define plant operating limits and to design transients, up to a maximum burnup of 33,000 MWD/MTU (Ref. 2).
However, subsequent rod bow penalty analysis done by B&W, and accepted (after being revised by B&W) by the Commission (Ref. 3) indicated that the 11.2% penalty is highly conservative. Therefore, the extended Cycle 4 operation is well protected against the effects of rod bow.
- b. In Reference 4 the anticipated cladding strain for Mark B-2 fuel (most limiting for Cycle 4 operation because of its low initial density) was shown to be less than 1% of the plastic strain limit for burnup to 55,000 MWD/NTU. The maximum anticipated burnup at the end of the extended Cycle 4 is well below 55,000 MWD/MTU.
C. B&W calculations using the TACO Code which we have approved for use in the range of this analysis (Ref. 5), have shown that changes in internal fuel rod pressures and average temperatures due to fission gas release are acceptable up to 42,000 MWD/MTU burnup. The maximum anticipated burnup at the end of the extended Cycle 4 is well below 42,000 MHD/MTU.
- 3) A decrease in calculated shutdown margin at the end of cycle (EOC):
The shutdown margin was calculated to be 1.45 Ak/k at the EOC 4. However, that calculation was carried out fora cycle life of 250 EFPD. At 250 EFPD in Cycle 4, the transient bank 7 is nearly fully inserted.
After 250 EFPD this bank starts to be withdrawn, thus increasing the shutdown margin over the calculated value of 1.45 Ak/k. Therefore, the requested extensioni in Cycle 4 will not decrease the EOC shutdown margin as calculated in
-3 the Cycle 4 Reload Report (Ref. 2).
- 4) An increase in the negative moderator temperature coefficient (MTC) and its effect on dropped rod transient and the steam line break (SLB) accident:
At the end of the proposed extension in Cycle 4, the 5TC will be -2.68 E-4 LW/k-0F as opposed to an EOC 4 (Ref. 2) value of -2.58 E-4 Ak/k-0F. However, in the dropped rod transient and the SLB accident analyses, an EO MTC value of -3.0 E-4 Ak/k-,F was used, thus bounding the new MTC value at the end of the extended Cycle 4.
- 5) An increase in the Negative Doppler Coefficient and its effect on the rod ejection accident:
The EGC 4 (Ref. 2) value of the Doppjer Coefficient is -1.59 E-5 Ak/k-oF.
This negative value will increase during the additional 66 EFPD of the Cycle 4 extension period due to the U-238 accumulation. A conserva tive value of -1.33 E-5 Ak/k-oF was used in the rod ejection accident analysis included in DPC's Final Safety Analysis Report (FSAR). There fore, the Doppler Coefficient change during the extended Cycle 4 oper ation will result in an increase in the safety margin above that assumed in the FSAR.
The TS would be revised to establish new reactor protection system limits and set points to ensure operation of the core within the prescribed DNB and accident limits such as peak clad temperature. The TS curves (which are acceptable) were derived following the methodology of B&W Topical Report BAI!-10121 entitled RPS Limits and Set Points, June 1978.
These curves are the part of the input used in the evaluation of transients and accidents.
Based upon our review of the fuel mechanical design, nuclear design, thermal hydraulic analysis, and accident and transient analysis which is summarized in the five point evaluation above, we conclude that the requested Cycle 4 extension for ONS-2 and the associated TS changes do not increase the proba bility or consequences of accidents or malfunctions previously considered nor involve a significant decrease in a safety margin.
Control Rod Guide Tube Wear By letter dated November 23, 1979, the Commission requested DPC to provide detailed information on the wear characteristics of the control rods on the guide tubes in fuel assemblies at the ONS. In response, DPC engaged B&W to perform confirmatory inspections on selected control rod guide tubes. The purpose of the inspections was to provide assurance that the fuel could sustain the proposed extended Cycle 4 operation at ONS-2 without experiencing through-the-wall wear in the guide tubes, in addition to providing generic information on B&W control rod guide tube wear.
The results of the preliminary inspection program were presented to the Commission at a December 20, 1979 meeting. Additional information was provided by DPC's letter dated December 28, 1979.
The inspections were performed by Eddy Current Test (ECT) techniques on fuel assemblies from the ONS-1 and 3 spent fuel pools. The EFPD of operation exper ienced by the inspected fuel ranged from'264 EFPD to 793 EFPD.
The ECT measurements that were performed'using an encircling coil technique and calibrated with machined standards are considered preliminary inspections by the Commission. Results of the inspections indicated minimal loss of wall thickness due to wear of the Zircaloy guide tubes by fretting action of the stainless steel clad control rods in the parked position.
No through-wall wear was observed in any of the tubes examined, and the maximum degradation reported was no greater than 24% through-wall. To provide additional assurance that guide tube wear indicated by the test results would not affect th.e structural integrity of the fuel, B&W reviewed the strength aspects of degradation.
The review considered uniform circumferential wear, one-sided wear, and two sided wear. Preliminary results indicate that the allowable wear in the limiting wear scenario would be in excess of 50% through-wall.
Based on the guide tube wear potential indicated by the ECT examinations, and the stress analysis results which included wear degradation, B&W and DPC con cluded that control rod guide tube wear does not appear to be a significant problem in B&W fuel and the extension of ONS-2, Cycle 4 will not compromise the structural integrity of the fuel.
We agree with the preliminary conclusions reached by DPC and B&W to the extent that the ECT measurements appear to indicate that there is a sufficient margin between the actual wear observed or expected during the extension of Cycle 4 and design limitations. Additional confirmatory evidence on the conservatism of the ECT measurements using the encircling coil technique is expected to be available in February 1980, based upon the results from surveillance programs currently being formulated by B&W and licensees of other B&W plants.
Our approval for extending the current fuel cycle for ONS-2, based on the pre liminary ECT inspections, has considered the following:
- 1) guide tube wear is a time-dependent process,
- 2) available evidence indicates that a sufficient margin exists between guide tube wear observed to date in B&W plants and design limits, and
- 3) confirmatory inspections are planned for February 1980.
ENVIRONMENTAL CONSIDERATIONS 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 951.5(d)(4),
that an environmental impact statment, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
-5 CONCLUSION We have concluded, based on the considerations discussed above, that: (1) because the amendments do not involve a significant increase in the probability or con sequences of accidents previously considered.and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable 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 compliance 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 and safety of the public.
Dated: January 4, 1980
-6 References
- 1. Letter, W. 0. Parker (Duke Power Company), to H. R, Denton (NRC), December 13, 1979.
- 2. Oconee-2, Cycle 4 Reload Report, BAW-1491, August 1978.
- 3. Letter, J. H. Taylor (B&W), to S, A. Varga (NRC), June 22, 1979.
- 4. Oconee-2, Cycle 3 Reload Report, BAW-1452, April 1977.
- 5. TACO -
Fuel Pin Performance Analysis, BAW-10087.
- 6. Oconee-2, Cycle 5 Reload Report, BAW-1565, October 1979.
- 7. Letter, W. 0. Parker (Duke Power Company), to H. R. Denton (NRC), December 28, 1979.
- 81.
Letter. R. W. Reid (NRC), to W. 0. Parker (Duke Power Company), December 15, 1978
- 9. Letter, R. W. Reid (NRC), to W. 0. Parker (Duke Power Company), November 23, 1 979.
10'.
Letter, W. 0. Parker (Duke Power Company), to H. R. Denton (NRC), January 2, 1980.
- 11.
RPS Limits and Setpoints, BAW-10121, June 1978.
7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKETS NOS. 50-296, 50-270 AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendments Nos. 78 78 and 75to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55, respectively, issued to Duke Power Company, (the licensee),
which revised the Station's common Technical Specifications for operation of the Oconee Nuclear Station, Units Nos. 1, 2 and 3, located in Oconee County, South Carolina.
The amendments become effective within five days after the date of issuance.
These amendments revise the core protection safety limits, protective system setpoints, control rod position, power imbalance, and axial power shaping rod position for Unit 2 consistent with extending the duration of the Unit's present operating cycle (Cycle 4) from 297 + 10 to 363 effective full power days.,
The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission'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.
The Commission has determined that the issuance of these amendments will not result in any significant environmental impact and that pursuant to 8001180
0 7590-01 2
10 CFR §51.5(d)(4) an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of these amendments.
For further details with respect to this action, see '(I) the application for amendments dated December 13, 1979, as supplemented December 28, 1979 and January 2., 1980, (2) Amendments Nos. 78, 78, and 75, to Licenses Nos.
DPR-38, DPR-47 and DPR-55, respectively, and (3) the Commission's related Safety Evaluation.
All of these items are available for public inspection at the Com mission's Public Document 'Room, 1717 H Street, N.W.,
Washington, D. C.,
and at the Oconee County Library, 201 South Spring Street, Walhalla,, South Carolina.
A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, Attention:
Director, Division of Operating Reactors.
Dated at Bethesda, Maryland, this 4th day of January 1980.
FOR THE NUCLEAR REGULATORY COMMISSION Robert W.. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors