ML19312C509
| ML19312C509 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/23/1977 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C507 | List: |
| References | |
| NUDOCS 7912160069 | |
| Download: ML19312C509 (18) | |
Text
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UNITED STATES
>1 NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 s*..,
DUKE POWER COMPANY DOCKET NO. 50-269 OCOMEE NUCLEAR STATION, UNIT N0. 1 AMEllDf:E!JT TO FACILITY OPERATING LICENSE Amendment No. 38 License No. DPR-38 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duke Power Company (the licensee) dated October 1,1975, complies with the standards and requirecents 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; 8.
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.
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2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility License No. DPR-38 is hereby amended to read as follows:
(2) Technical Specifications
/
The Technical Specifications contained in Appendices A and B, as revised thraugh Amendment No. 38, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY CCf1 MISSION 6
A. Schwencer, Chief Operating Reactors Branch $1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 23, 1977 f
_)
I ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT N0. 38 TO DPR-38 l
AMENDMENT NO. 38 TO DPR-47 1
i AMENDMENT N0. 35 TO DPR-55 1
00CKETS tl05. 50-259, 50-270 AND 50-287 j
Revise Appendix A as follows:
1.
Remove pages 3.1-3 through 3.1-9 and insert revised identically numbered pages.
2.
Add pages 3.1-3a, 3.1-6a, 3.1-7a and 3.1-7b.
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3,g,2 Pressurization, Heatup, and Cooldown Limitations Sgecification 3.1.2.1 The reactor coolant pressure and the system heatup and cocidown 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 Units 2&3.
%vsidown:
Cooldown rates and allowable combinations of pressure and tempera-ture shall be limited in accordance with Figure 3.1.2-2A Unit 1 3.1.2-2B Units 2&3.
3.1.2.2 Leak Tests Leak tests 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 are present provided the reactor coolant system is to the right of and below the limit line in Figure 3.1.2-3.
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 pressurizer 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 4100F.
3.1.2.6 Pressurization, heatup and cooldown limitations shall be updated based on the results of the reactor vessel materials surveillance Program described in Specification 4.2.8.
3.1-3 Amendments Nos, 38, 38 & 35
n.
Bases - Unit l_
An 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 bo.ndary have been analyaed in accordance with Appendix G to 10CFR50.
Results of this analysis, including the actual pressure-te=perature limitations of the reactor coolant pressure boundary, are given in BAW-1421(7),
Figures 3.1.2-1A, 3.1.2-2A, 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 fifth effective full power year of operation. These curves are adjusted by 25 psi and 100F for possible errors in the pressure *and temperature sensing instruments.
The pressure limit is also adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.
The pressure-temperature limit lines shown on Figure 3.1.2-1A for reactor criticality and on Figure 3.1.2-3 for hydrostatic testing have been provided to assure compliar:e with the minimum temperature require =ents of Appendix C to 10CFR50 for reactor criticality and for inservice hydrostatic testing.
The actual shif tI in RTET of the beltline region material will be established periodically during operation by removing and evaluating, in accordance with Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of the 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 RTET of +600F, the protection against nonductile 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 esti=ated RTET 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 Amendmenu *ics, 38, 38 & 35
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Bases Units 2 and 3_
l All reactor coolant system components are designed to withstand the effects (1)
These of cyclic loads due to system temperature and pressure changes.
cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations.
The number of ther=al and loading cycles used for design putposes are shown in Table 4-8 of the FSAR.
The maximum unit heatup and cooldown rate of 1000F per hour satisfies stress limits for cyclic operation.
(2) The 237 psig pressure limit for the secondary side of the steam generator at a ':emperature less than 11GoF satisfies stress levels for temperatures below the DTT.
(3)
The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maximum NDTT value of 200F has been determined based on Charpy V-Notch tests. The maximum NDTT value obtained for the steam generator shell material and welds was 40cy.
Figures 3.1.2-13 and 3.1.2-2B contain the limiting reactor coolant system l
pressure-temperature relationship for operation at DTT(4) and below to assure that stress levels are low enough to preclude brittle fracture.
These stress levels and their bases are defined in Section 4.3.3 of the FSAR.
As a result of fast neutron irradiation in the region of the core, there will be an increase in the NDTT with accumulated nuclear operation. The predicted maximum NDTT increase for the 40-year exposure is shown on Figure 4.10.(4)
The actual shift in NDTT will be determined periodically during plant opera-tion by testing of irradiated vessel material samples located in this reactor vessel.(5)
The results of the irradiated sample testing will be evaluated and compared to the design curve (Figure 4-11 of FSAR) being used to predict the increase in transition temperature.
The design value for fast neutron (E > 1 MeV) exposure of the reactor vessel is 3.0 x 1010 n/cm2 -- s at 2,568 We rated power and an integrated exposure l
of 3.0 x 1019 n/cm2 for 40 years operation.
(6)
The calculated maximum values are 2.2 x 1010 n/cm2 -- s and 2.2 x 1019 n/cm2 integrated exposare for 40 years operation at 80 percent load.
(4)
Figure 3.1.2-1B is based on the design value which is considerably higher than the calculated value.
The DTT value for Figure 3.1.2-1B is based on the projected NDTT at the end of the first two years of operation.
During these two years, the energy output has been conservatively estimated to be 1.7 x 106 thermal megawatt days, which is equivalent to 655 days at 2,568 W t core power. The projected 18 fast neutron exposure of the reactor vessel for the two years is 1.7 x 10 n/cm2 which is based on the 1.7 x 106 thermal megawatt days and the design value for fast neutron exposure.
The actual shift in NDTT will be established periodical'ly during plant operation by testing vessel m.~.terial samples which are irradiated cumulatively by securing them near the inside wall of the vessel in the core area.
To compensate for the increases in the NDTT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heacup and cooldown.
3.1-4 Amendments Nos. 38, 38 & 35 i
The NDrr shift and the magnitudes of the thermal and pressure stresses are sensitive to integrated reactor power and not to instantaneous power level.
Figure 3.1.2-1B and 3.1.2-2B are applicable to reactor core thermal ratings up to 2,568 MWt.
The pressure limit line on Figure 3.1.3-1B has been selected such that the l
reactor vessel stress resulting from internal pressure will not exceed 15 percent yield strength considering the following:
1.
A 25 psi error is measured pressure.
4 2.
System pressure is measured in either loop.
3.
Maximum differential pressure between the point of system pressure measurement and reactor vessel inlet for all operating pump combinations.
For adequate conservatism in fracture toughness including size (thickness) affect, a maximum pressure of 550 psig below 2750F with a maximum heatup and cooldown rate of 500F/hr has been imposed for the initial two year period as shown on Figure 3.1.2-13.
During this two year period, a fracture toughness criterion applicable to Oconee Units 2 and 3 beyond this period will be developed by the AEC.
It will be based on the evaluation of the fracture toughness properties of heavy section (thickness) steels, both irradiated and unirradiated, for the AEC-HSST program and the PVRC program, and with considerations of test results of the Oconee Units 2 and 3 reactor. l surveillance programs.
The spray temperature difference restriction is imposed to maintain the I
thermal stresses at the pressurizer spray line nozzle below the design limit.
Temperature requirements for the steam generator correspond with the measured NDTT for the shell.
REFERENCES (1)
FSAR Section 4.1.2.4.
(2) ASME Boiler and Pressure Code,Section III, N-415.
1 (3)
FSAR Section 4.3.10.5.
(4) FSAR Section 4.3.3.
(5)
FSAR Section 4.4.6.
(6)
FSAR Sections 4.1.2.8 and 4.3.3.
l (7) Analysis of Capsule OCl-F from Duke Power Company Oconee Unit 1 l
Reactor Vessel Materials Surveillance Program, SAW-1421 Rev. 1 September 1975.
3.1-5 Amendments Nos. 38, 38 & 35 l
2400 "a
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E 2000 INE ACCEPT &glE PRESSURE AND TEMPERATURE COM8snAT10n3 ARE BELow
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AND TO THE RIGNT OF inE LIMIT CURVE.
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RIGNT Of TME CRITICAttiv LIMai C; lave. HARGINS or 25,PSIG AND 80F c
ARE INCluGED FOR POS$18LE INSTRUMENT EPROR.
1600 m
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1200 w
APPLICABLE FOR HEATUP RATES OF < 100*F/HR T
$ 1000 (i $0"F IN ANY 1/2 HOUR PERIOD)-
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20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 IndeCated Reactor C00lant System Temperatufe, it,*F M
Unit i Only REACTOR' COOLANT SYSTEM HEATUP LlHITATIONS, APPLICABLE FOR FIRST 4 EFPY mi remR OCONEE NUCLEAR STATION Figure 3.1.2-l A
7 PolNT TEMP.
PRESS.
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1800 af 1600 UPPER PRESSURIZATION LIMIT a
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O 100 200 300 400 500 275 Indicated Reactor Coolant System Temperature. F e.aireein OCONEE NUCLEAR STATION Unita 2 & 3 REACTOR COOLANT SYSTEM NEATUP LIMITATIONS
'(APPLICABLE UP TO AN INTEGRATE 0 EIPOSURE 0F l.7 x 10 18 n/ c m2 OR OTT - 144 F)
F i gu r e 3.1.2 1 B 3.1-6 a Arendments Nc. 38, 38 & 35
2600 3 2400 in[ ACCEPIABLE PRES $uRE AnD TEMPERATURE COMBlnAT10ns ARE BELOW AND 16
{ 2200 Int R Gni 0F inE LeMil CuRvt. MARGen$ OF 25 PSiG AnD 10f ARE anClu0E0 FOR POSSIBLE In5TRUMENT ERROR.
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OPERAltnG. Tut In0ICATED DnB SYSTEM RETURN TEMPERAluRE 10 INE REAC10R VESSEL SMALL BE USED.
E 2.
A MAIIMUM STEP TEMPERATURE CHANGE OF 75*f 15 ALLOWABLE Wnta NEMOVING 2 1600 ALL RC PUMP 5 tROM OPERATION wiin inE One 5YSTEM OPERAlinG.
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3 STEP TEMPERATURE CHANGE 15 DEFINED AS inE RC IEMP (PRIOR To i
STOPPING ALL RC PUMPS) Minus Tnt Ona REluda IEMP (AtIER E,
1400 STOPrinG ALL RC PUMPS) inE 100*f/nt RAMP DECREASE 15 0
ALLOWAILE soin sEf0RE AND AFTER inE STEP TEMP CnAnGE.
APPLICABLE FOR COOLDOWN RATES Of s 100'F/HR (s 50'I IN ANY
^ 1200 w
mi Note: Applicable craldown rates -
532 to 4320F, s 1000F/HR; 1/2 H0llR PEH100) (2)
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Unit i Only REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS, APPLICABLE FOR FIRST 4 EFPY OCONEE NUCLEAR STATION Figure 3.1.2-2A
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120 200 RC PUMP COMBINATIONS ALLOWABLE.
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.* 2000 (I) WHEN DECAY HlAT REMOVAL SYSTEM (OH) IS g 1800 OPERATING Elf HOUT ANY RC PUNPS OPERaithG.
O IN0lCATED OH RETURN TEMP TO THE REACTOR
[ 1600 VESSEL SHALL BE USED.
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(2) IN THE TEMPERATURE RANGE 260F TO 175F. A 1400 MAllMUM STEP TEMPERATURE CHANGE OF 75F 15 ALL0 SABLE FOLLOWED BY A ONE HOUR 3}1200-MINIMUM HOLO CN TEMPERATURE. IF THE STEP CHANGE IS TAKEN BELOW 250F RC TEMPERATURE.
1000 THE MAXIMUM ALL0 SABLE STEP SHALL BE H AT WHICH YlELOS A FINAL TEMPERATURE OF 175F.
3 800 THE STEP TEMPERATURE CHANGE is DEFINED AS RC TEMPERATURE (BEFORE STOPPING ALL RC PUNPS) g MINUS THE OH RETURN TEMPERATURE TO THE REACTOR
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,,,30 0 200 100 Indicated Reactor Coolant System Tesoerature.'F(II RE ACTOR COOL ANT SYSTEM COOLDOWN Llul T ATIONS
( APPLICA8L E UP TO OTT = 185'F)
Units 2 & 3 M' OCONEE NUCLEAR STATI If0*";
Figure 3.1.2 28 3.1-7 a Amendments Nos. 38, 38 & 35
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D 2500 THE ACCEPTABLE PRESSURE AND TEMPERATURE COM3INATIONS ARE E
8ELOW AND TO THE RIGHT OF THE LIMIT. CURVE. MARGINS OF 2300 a
25 PSIG AND 10F ARE INCLUDED FOR POSSIBLE INSTRUMENT-5 ERROR. FOR C00LDOWN, NOTES I AND 2 ON FIGURE 3.1.2-2A ARE APPLICABLE.
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$1900
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0 APPLICABLE FOR HEATUP OR C00LDOWN 2 1500 RATES OF s 500F/HR (1250F IN ANY O
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20 40 60 80 100 120 140 150 180 200 220 240 260 280 300 320 340 co Indicated Reactor Coolant System Temperature. Tc.*F REACTOR COOLANT SYSTEM HEATUP AND C00LD0VN LIMITATIONS FOR INSERVICE O
HYDROSTATIC TESTS (NO FUEL ASSEMBLIES IN THE CORE), APPLICABL2 FOR FIRST 4 EFPY OCONEE NUCLEAR STATION Figure 3.1.2-3
m 1.1.3 Minimum conditions for criticality Specification _
3,1.3.1 The reactor coolant temperature shall be above 5250F except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply.
3.1.3.2 Reactor coolant temperature shal1 be above the criticality limit of 3.1.2-1A (Unit 1) or above DTT + 100F (Units 2 and 3).
3.1.3.3 When the reactor coolant temperature is below the minimum te=pera-ture specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply, the reactor shall be suberitical by an amount equal to or greater than the calculated reactivity insertion due to depressuri-zation.
3.1.3.4 The reactor shall he maintained subcritical by at least lLik/k until a steam bubble is for=ed and a water level between 80 and 396 inches is established in the press.urizer.
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 =argin 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 coeffi-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 temperatures will be less negative or more positive than at operating temperature,(2) startup and operation of the reactor when reactor coolant temperature is less than 3250F is prohibited except 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 precautions will be taken.
In addi-tion, the strong negative Doppler coefficient (1) and the small integrated Ak/k would limit the magnitude of a power excursion resulting from a redcc-l l
tion of moderator density.
The requirement that the reactor is not to be made critical below the limits i
of Specification 3.1.2-1 provides increased assurance that the proper relatien-l ship between primary coolant pressure and temperature will be maintained rela-tive to the NDTT of the primary coolant system.
Heatup to this temperature will be accomplirhad by operating the reactor coolant pumps.
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3.1-8 Amendments Nos. 38, 38 & 35
If the shutdown margin required by Specification 3.5.2 is maintained, there is no Possibility of an accidental criticality as a result of a decrease of coolant pressure.
The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1% suberitical will assure that the reactor coolant system cannot become solid in the event of a red withdrawal accident or a startup accident.(3)
The requirement that the safety rod groups be fully withdrawn before criti-cality ensures shutdown capability during startup.
This does not prohibit rod latch confirmation, i.e., withdrawal by group to a caximum of 3 inches withdrawn of all seven groups prior to safety rod withdrawal.
The requirement for regulating rods being within their rod position limits ensures that the shutdown margin and ejected rod criteria at hot zero power are not violated.
REFERENCES (1) FSAR, Section 3 (2)
FSAR, Section 3.2.2.1.4 (3) FSAR, Supplement 3 Answer 14.4.1
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3.1-9 Amendments Nos. 38, 38 & 35
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UNITEO STATES
/k NUCLEAR REGULATORY COMMISSION f fgT j
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J l WASHINGTON, D. C. 20566 i.C3 ' /
, d e]
p DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATIOil, UtlIT NO. 2 AfiENDMENT TO FACILITY OPERATING LICENSE Amendment No. 38 License No. DPR-47 1.
The Nuclear Regulatory Commission (the Ccamission) has found that:
A.
The application for anendment by Duke Power Company (the licensee) dated October 1,1975, ccmolies with the standards and requirenents 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; 8.
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 anendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with tne Commission's regulations; i
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 amendnent is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, the license
< cnended by changes to the Technical Speci fication: as indicatku in the attachment to this license amendment and par agraph 3.8 or Facility License No. OPR-47 is hereby amended to re.* 's follows:
(2) Technical Specifications j
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.38, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 1
A. Scnwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: February 23, 1977 l
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UNITED STATES y 'i
, - 5, NUCLEAR REGULATORY COMMISSION j ;..I.
k WASHINGTON. D. C. 20555
i
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- ....f DUKE POWER COMPANY DOCKET N0. 50-287 OCONEE NUCLEAR STATION, UNIT N0. 3 AMENDMENT TC FACILITY OPERATING LICENSE Amendment No. 35 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 October 1,1975, complies with the standards and requirenents of the Atonic 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 Commission; C.
There is reasonable assurance (1) 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.B of Facility License No. DPR-55 is hereby amended to read as follows:
(2) Technical Specifications
/
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 35, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0f1 MISSION 64 2-A. Sc'hwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical
. Specifications Date of Issuance: February 23, 1977 O
e