ML19312B894
| ML19312B894 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/06/1977 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 7911250082 | |
| Download: ML19312B894 (14) | |
Text
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. I 3.1.2 Pressurization, Heatup, 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 limitad 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. s Cooldown:
Cooldown rates and allowable combinations of pressure and tempera-ture shall be limited in accordance with Figure 3.1.2-2A Unit 1 3.1.2-2B Unit 2 3.1.2-2C Unit 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 3.1.2-3A Unit 1 3.1.2-3B Unit 2.
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 and hydro test limits shall be updated based on the results of the reactor vessel materials surveillance program described in Specification 4.2.9. These revised limits shall be submitted to the NRC at least 90 days prior to ex-ceeding four (Unit 1) effective full powar years of operation or an six (Unit 2) 8 integrated exposure of 1.7 x 10 n/cm or DTT 144 F for Unit 3 . ,
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Bases - Unit l and 2 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 coolant ;
pressure boundary, are given in BAW-1421(7) and BAW-1437(8). I 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 j full power yearsof 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 differentf61 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 the figure specified in 3.1.2.1 f6r reactor criticality and on the figure specified in 3.1.2.3 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice hydrostatic testing.
The actual shift in RTNDT 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 this or a similar l 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 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 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.
1 The spray temperature difference is imposed to maintain the thermal stresses l at the pressurizer spray line nozzle below the design limit.
l 3.1-3a
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Bases Unit 3 All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. (1) These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes 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 l of the steam generator at a temperature less than 1100F 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 j 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 400F.
Figures 3.1.2-lc and 3.1.2-2C 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) l The actual shif t in NDTT will be determined periodically during plant opera-4 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 l 1s 3.0 x 1010 n/cm2 -- s at 2,568 MWt rated power and an integrated exposure l of 3.0 x 1019 n/cm2 for 40 years operation. (6) The calculated maximum l values are 2.2 x 1010 n/cm2 -- s and 2.2 x 1019 n/cm2 integrated exposure !
for 40 years operation at 80 percent load. (4) Figure 3.1.2-1c is based on l l the design value which is considerably higher than the calculated value. J The DTT value for Figure 3.1.2-1C is based on the projected NDTT at the end l ,
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 MWt core power. The projected fast neutron exposure of the reactor vessel for the two years is 1.7 x 1018 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 NDIT will be established periodically during plant l operation by testing vessel material 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 NDIT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay
-within the established stress limits during heatup and cooldown.
3.1-4
The NDTT shift and the magnitude of the thermal cud pressure stresses are sensitive to integrated reactor power and not to instantaneous power level.
Figure 3.1.2-]C and 3.1.2-2C are applicable to reactor core thermal ratings {
up to 2,568 MWt.
The pressure limit line on Figure 3.1.3-1C 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.
- 2. System pressure is measured in either loop.
- 3. Maximum differential pressure between the point of system pressure measurement and reactor ves.lel 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-lC. During this two year period, a fracture toughness criterion applicable to Oconee Unit. 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 surveillance programs.
The spray temperature difference restriction is imposed to maintain the j 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.
(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.
(7) Analysis of Capsule OCl-F from Duke Power Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1421 Rev. 1, September 1975.
(8) Analysis of Capsule OC11-C from Duke Power Company Oconee Unit 2
! Reactor Vessel Materials Surveillance Program, BAW-1437, June, 1977.
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InE AccEPIABLE PRESSURE AND IEMPE R A TURE C0136mA110NS ANE SELOW G ano 10 Int RIGHI 0F THE LIMil CURVE. THE REACIOR HUST NOI UE MADE
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Indicated Reactor Coalant System Temperature,'F lb ont rono; OCONEE NUCLEAR STATION Unit 3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (APPL l CABLE UP TO AN INTEGRATE 0 EXPOSURE OF 1.7 x 10 18 n/ c m2 OR OTT - 144 'F)
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f2000 _ l. WHEN INE DECAf HEAT REMOVAL (DHR) SYSTEM iS OPERAllNG WifM NO RC PUMPS OPERAllMG. THE INDICA 1ED OMR SYSIEH REIURN TEMPERATURE 10 THE REACIOk U
' VESSEL SHALL BE USED.
c 1800 _
E 2. A maximum STEP i[HPERATURE CHANGE OF 75*f IS ALLOWABLF WHIM R E MOV ING THE o" 1600 _
ALL RC PUMPS fROM OPERAll0N WilH THE CNR SYSTEM OPERAI4NG.
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Note: Applicable cooldown rates -
S32 to 4320F, s1000F/HR; i .'2 H0tlR PEll100) (2 )
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REACTOR COOLANT SYSTEM NORMAL CPERATION C00LDOWN LIMITATIONS i APPLICABLE FOR FIRST 6.0 EFPY l OCONEE NUCLEAR STATION
- 3. I-h Ou Figure 3.1.2-28
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PolNT VEh. PRESS A 380 8275 8 275 1400 C 275 550 0 250 550 E 250 450 175 450 2400 - F
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H 120 200 2200 - RC PUMP COMBINATIONS ALL0 SABLE:
ABOVE 185F ALL 2000 - BEL 0s 185F 1 A,1-B;0 A.2 8;1 4.0 8;0 4.1 B
.=a (1) SHEN DECAY HEAT REMOVAL SYSTEM (OH) IS OPERATING slTHOUT ANY RC PUMPS OPERATING.
f1800
- lNDICATED DH RETURN TEMP. TO THE REACTOR 2 1800 VESSEL SHALL BE USED. -
- 1400 -
MAIIMUM STEP TEMPtHATURE CHANGE OF 75F
$ l$ ALL0 SABLE FOLL0sEO BY A ONE HOUR j 1200 - MINIMUM HOLO ON TEMPERATURE. IF THE STEP 8 CHANGE IS TAKEN BEL 0s 250F RC TEMPERATURE,
- 1000 - THE MAIiMUM ALL0 SABLE STEP SH ALL BE TH AT 3 eHICH YlELOS A FINAL TEMPERATURE OF 175F.
THE STEP TEMPERATURE CHANGE IS DEFINED AS 3 800 -
e RC TEMPERATURE (BEFORE STOPPING ALL RC PUMPS)
WINUS THE OH RETURN TEMPFRATURE TO THE REACTOR 600 - /
- VESSEL.
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UPPER PRES $URI Z Ail 0N _.
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MAllMUM C00LD0sN RATE. *F.HR
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530 i i i i 260 i 175 120 1 275 400 300 200 100 600 500 indicatea Reactor Coolant System Temaerature.*F II)
REACTOR COOL ANT SYSTEM C00L00sN LIMITAil0NS
( APPLICA8LE UP TO OTT = IB5*F)
Unit 3 pufonte, OCONEE NUCLEAR STATION M Figure 3.1.2 2 e 3,1-7 b
'[2700 i 0
- 2500 -
THE ACCEPTABLE PRESSURE AND TEMPERATURE COMBINATIONS ARE B BELOW AND TO THE RIGHT OF THE LIMIT CURVE. MARGINS OF 2300 -
25 PSIG AND 10F ARE INCLUDED FOR POS$1BLE INSTRUMENT ERROR. FOP,C00LDOWN, NOTES I AND 2 ON FIGURE 3.1.2-2A ARE APPLICABLE.
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3 APPLICABLE FOR llEATUP OR C00LDOWN E 1500 -
RATES OF s 500F/HR (s 250F IN ANY J B 1/2 HOUR PERIOD).
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> 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 QQ uw indicated Reactor Coolant System Temperature, ic.*f UN(Tl "R REACTOR COOLANT SYSTEM HEATUP AND d C00LDOWN LIMITATIONS FOR INSERVICE HYDROSTATIC TESTS (NO FUEL ASSEMBLIES IN THE CORE), APPLICABLE FOR FIRST 4 EFPY OCONEE NUCLEAR STATION Figure 3.1.2-3
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800 600 5
T m 4A 3= H The acceptable pressure-temperature combinations are Q below and to the right of the limit curve (s).
12 hhoN E N ~
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$ hhhg limit curves include the pressure differential between the point of system pressure measurement and the pressure
] 7 pD -g on the reactor vessel reelen controlline the limit curve, ther mn o 200 -
[C "I
O Or rl h b include an additional margin of safety for possible instrument error (25 psi and 10*f) f or cooldown, notes I and 2 on figure 3.1.2-28 are applicable.
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m Indicated Reactor CO0lant Temperature, Tc, *F
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1 3.1.3 Minimus Conditions for Criticrlity 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.
{
j 3.1.3.2 Reactor coolant temperature shall be above the criticality limit of 3.1.2-1A (Unit 1) or above DTT + 100F (Unit 3).
. 3.1.2-1B (Unit 2) 3.1.3.3 When the reactor coolant temperature is below the minimum tempera-ture specified in 3.1.3.1 above, except for portions of low power j physics testing when the requirements of Specification 3.1.9 shall j 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 be maintained subcritical by at least 1%Ak/k until a steam bubble is formed and a water level between 80 and i 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 j the approach to criticality. The regulating rods shall then be l 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 525oF, 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 i op.1 ration of the reactor when reactor coolant temperature is less than 5250F is prohibited except where necessary for low power physics tests.
i 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 coeff.;cisar(1) and the small integrated dk/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 relation-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 accomplished by operating the reactor coolant pumps.
3.1-8 l
- - l
. ~'
l i
- j If the shutdown margin required by Specification 3.5.2 is maintained, th3re ,
is no po6sibility 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 rod 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 maximum 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
- ~
A 38/38/35 2/23/77
, _ _ _