ML19312C517
| ML19312C517 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/08/1978 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C516 | List: |
| References | |
| NUDOCS 7912160075 | |
| Download: ML19312C517 (15) | |
Text
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UNITED STATES p '-
'i NUCLEAR REGULATORY COMMISSION
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p DUKE POWER COMPANY DOCKET N0. 50-269 OCONEE NUCLEAR STATION, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 56 License No. DPR-38 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendments by Duke Power Company,(the licensee) dated September 14, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission:
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 Ccmmission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by a change to the Techr.ical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. C.1-38 is hereby amended to read as follows:
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2 "3.S Tecnnical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 56, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Tecnnical Specifications."
3.
This license amendment is effective within 30 days after the date of issuance.
FOP THE NUCLEAR REGULATORY COMMISSION
,&:vuxw A. Schwencer, Chief Operating Reactors Branch al Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: February 8,1978 "i
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UNITED STATES NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D. C. 20555 f
- %......o' DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT N0. 2 AMENDMEl4T TO FACILITY OPERAT!NG LICENSE Amendment No. 56 License No. DDR-47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendments by Duke Power Company (the licensee) dated September 14, 1977, 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 I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Cornission:
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 connon 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 a change to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. DPR-47 is hereby amended to read as follows:
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i 2-1 "3.S Technical Scecifications I
The Technical Specifications contained in Appendices A and S, as revised through Amendment No.S6, are j
hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications."
3.
This license amendment is effective within 30 days after the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
ktM&WAL A. Schwencer, Chief Operating Reactors Branch *1 Division of Operating Reactors
Attachment:
Changes to the Technical i
Specifications Date of Issuance:
February 8,1978 l
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C) 4r UNITED STATES f*),
NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555 e
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D DUKE POWER COMPANY DOCKET NO. 50-287 OCD',EE NUCLEAP STATION, UN:
NC. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendmer.: No. 53 License No. OPR-55 1.
The Nuclear Regulatory Comission (the conmission) has found that:
A.
The application for amendments by Duke Power Company (the licensee) dated September 14, 1977, 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 I; B.
The f acility will operate in conferrity with the applicatiGn, the provisions of the Act, and the rules and regulations of the Comission:
C.
There is reasonaole assurance (i) tnat tne activities authorizec bv this amencrent can be corducted without endar,;ering the healtr and safety of the public, and (ii', that such activities will be condectec in coroliance witr. the Coimission's reg;lations; n.
The issoacce o' nis arerdrert will not te inimical to the c arr :r de's se anc securitj :r to the haaith and safety of tte publi:; and E.
The is bance of thi; a 0-d ert is ir ac;cedance,itn iC CF; Par.
51 of the Ccrrissien's c;-cfations 3r.c al' ap;l' nle c..;;1. y t; have oeer satis #iec.
2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license l
amendment and paragraph 3.8 of Facility License No. DPR-55 is hereby amended to read as follows:
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"3.B Technical Specifications The Technical Specifications contained in Accendices j
A and 5, as revised tnrough Amendment No. 53, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications."
3.
This license amendment is effective within 30 days after the date i
of issuance.
FOR THE NUCLEAR REGULATJRY COMMISSION i
't g 9.s c d S t --
A. Schwencer, Chief Operating Reactors Branch =l r
Division of Operating Reactnrs l
Attachment:
Changes to the Technical Specifications Date of Issuance: February - 8, 1978
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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT N0. 56 TO DPR-38 AMENDMENT NO. 56 TO DPR-47 AMENDMENT NO. 53 TO DPR-55 DOCKET NOS. 50-269, 50-276 AND 50-287 Revise Appendix A 65 follows:
Remove the following pages and insert pages 3.1-3 3.1-3a 3.1-4 3.1-5 3.1-6b 3.1-7b 3.1-8 Add page 3.1-7e
3.1.2 Prascurizntion, Hratup, and Cooldown Limitations Specification 3.1.2.1 The reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited as follows:
Heatup:
Heatup rates and allowable combinations of pressure and tempera-tures shall be limited in accordance with Figure 3.1.2-1A Unit 1 3.1.2-1B Unit 2 3.1.2-1C Unit 3.
Cooldown:
Cooldown rates and allowable combinations of pressure and tempera-ture shall be limited in accordance with Figure 3.1.2-2A Unit 1 3.1.2-23 Unit 2 3.1.2-2C Unit 3.
3.1.2.2 Leak Tests Leak test required by Specification 4.3 shall be conducted under the provisions of 3.1.2.1.
3.1.2.3 Hydro Tests For thermal steady state system hydro test the system may be pressurized to the limits set forth in Specification 2.2 when there are fuel assemblies in the core under the provisions of 3.1.2. 1 and to ASME Code Section III limits when no fuel assem-blies a.e present provided the reactor coolant system is to the right of and below the limit line in Figure 3.1.2-3A Unit 1 3.1.2-3B Unit 2 3.1.2-3C Unit 3.
l 3.1.2.4 The secondary side of the steam generator shall not be pressurized above 237 psig if the temperature of the vessel shell is below 1100F.
3.1.2.5 The pressurizar heatup and cooldown rates shall not exceed 1000F/hr.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 41007.
3.1.2.6 Pressurization heatup and cooldown limitations and hydro test limits shall be updated based on the results of the reactor vessel materials surveillance program.
These revised limits shall be subnitted to the
- RC at least 90 days prior to exceeding four effective full power years of operation.
I 3.1-3 Amendments 56, 56 t 5'
Banac - Units 1, 2 rnd 3 All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, startup and shutdown operations, and inservice leak and hydrostatic tests.
The various categories of load cycles used for design purposes are provided in Table 4.8 of the FSAR.
The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10CFR50.
Results of this analysis, including the actual pressure-temperature limitations of the reactor enolant pressure boundary, are given in BAW-1421(1), BAW-1437(2) and BAW-1438(3).
l The figures specified in 3.1.2.1, 3.1.2.2 and 3.1.2.3 present the pressure-temperature limit curves for normal heatup, normal cooldown and hydrostatic test, respectively.
The limit curves are applicable up to the indicated effective full power years of operation.
These curves are adjusted by 25 psi and 100F for possible errors in the pressure and temperature sensing instru-ments.
The pressure limit is also adjusted for the pressure dif ferential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.
The pressure-te=perature limit lines shown on the figure specified in 3.1.2.1 for reactor criticality and on the figure specified in 3.1.2.3 tor hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inser-vice hydrostatic testing.
The actual shift in RTNDT of the beltline region material will be established periodically during operation by removing the evaluating, in accordance with Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region.
The limitation on steam generator pressure and temperature provide protection against nonductile failure of the secondary side of the steam generator.
At metal temperatures lower than the RTNDT of +600F, the protection against non-ductile failure if achieved by limiting the secondary coolant pressure to 20 percent of the preoperational system hydrostatic test pressure.
The limitations of 1100F and 237 psig are based on the highest estimated RTNDT of +400F and the preoperational system hydrostatic test pressure of 1312 psig.
The average metal temperature is assumed to be equal to or greater than the coolant temperature.
The limitations include margins of 25 psi and 100F for possible instrument error.
The spray temperature difference is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.
3.1-3a Amendrents 36, 56 & 53
.s REFERENCES (1)
Analysis of Capsule OCl-F from Duke Power Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1421 Rev. 1, September 1975.
(2)
Analysis of Capsule OC2-1C from Duke Power Company Oconee Unit 2 Reactor Vessel Materials Surveillance Program, BAW-1437, April, 1977.
(3) Analysis of Capsule OCIII-A from Duke Power Company Oconee Unit 3 Reactor Vessel Materials Surveillance Program, BAW-1438, July, 1977.
1 Amendments 56, 56 & 53
C ENTIRE PAGE DELETED 1
l 3.1-5 Arendments 56, 56 & 53
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3.1.3 Minimum Conditiona for Criticality Specification 3.1.3.1 The reactor coolant temperature shall be above 525CF except for portions of low power physics testing when the requirements or Specification 3.1.9 shall apply.
3.1.3.2 Reactor coolant temperature shall be above the criticality limit of 3.1.2-1A (Unit 1) 3.1.2-13 (Unit 2) 3 1.2-1C (Unit 3) l 3.1.3.3 When the reactor coolant temperature is below the =inimum tem-pera :ure specified in 3.1.3.1 above, except for portions of low powe. physics testing when the requirements of Specification 3.1.9 shall apply, the reactor shall be subcritical by an amoent equa to or greater than the calculated reactivity insertion due to d. pressurization.
3.1.3.4 The eactor shall be maintained suberitical by at least 1%Ak/k unti a steam bubble is formed and a water level between 80 and 396 inches is established in the pressurizer.
3.1.3.5 Except for physics tests and as limited by 3.5.2.1, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality.
The regulating rods shall then be positioned within their position limits defined by Specification 3.5.2.5 prior to deboration.
Bases At the beginning of the initial fuel cycle, the moderator temperature coef fi-cient is expected to be slightly positive at operating temperatures with the operating configuration of control rods.(1) Calculations show that above 5250F, the consequences are acceptable.
Since the moderator temperature coefficient at lower te=peratures will be less negative e more positive than at operating temperature,(2) startup and operation of t a rasetor when reactor coolant temperature is less than 5250F is prohibited excep: where necessary for low power physics tests.
The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.lak/k.
During physics tests, special operating pro:autions will be taken.
In addi-tion, the strong negative Doppler coef ficient(1) and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduc-tion of moderator density.
The requirement that the reactor is not to be made critical below the limits of Specification 3.1.2.1 provides increased assurance that the proper rela-tionship between primary coolant pressure and temperature will be maintained relative to tihe NDTT of the primary coolant system.
Heatup to this tempera-ture will be accomplished by operating the reactor coolant pumps.
Arendments 56, 56 & 53 3.1-8