ML19319D921
| ML19319D921 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/02/1978 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19319D920 | List: |
| References | |
| NUDOCS 8003270594 | |
| Download: ML19319D921 (8) | |
Text
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RANCHO SECO UNIT 1 TECHNICAL SPEClflCATIONS Limiting Condi tions for Operation 3.1.2 PRESSURIZATION, HEATUP, AND C00LDOWN LIMITATIONS Specification
-3 1.2.1 Hydro Tests:
For thermal steady state system hydro tests the system may be pressurized to the limits set forth in Specification 2.2 when there are fuel assemblies in the core and to ASME Code Section ill limits when no fuel assemblies are present provided that:
a.
During the first five effective full power (EFP) years of operation the reactor coolant system temperature / pressure limits in accordance wi th figure 3.1.2-4.
3.1.2.2 Inservice Leak Tests Pressure temperature limits -for the fi rst five EFP years of inservice leak and
. hydrostatic tests are given in Figure 3.1.2-4.
Heatup and cooldown rates shall be restricted to 100"F/hr.
3.1.2.3 For the fi rst five EFP years of power operation the. reactor coolant pressure and the system-heatup and cooldown rates (with the exception of the pres-surizer) shall be limited in accordance with figure 3.1.2-2 and figure 3.1.2-3 respectively. Heatup and cooldown rates shall not exceed 100*F/hr.
3 1.2.4 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 130*F.
Proposed Amendment No. 59 O
8003 270 fff 3.3
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS i
Limiting Conditions for Operation 3.1.2 5 The pressurizer heatup and cooldown ratet shall'not exceed 100*F in any l-hour period.
s 3.1.2.6 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 410*F.
Bases The pressure-temperature limits of the reactor coolant pressure boundary are es tablished in accordance wi th the requi rements of Appendix G to 10 CFR 50 and with the thermal and loading cycles used for design purposes.
The limitations prevent non-ductile failure during normal operation, including anticipated operational occurrences and system hydrostatic tests.
The limits also prevent exceeding stress limits during cyclic operation.
The loading conditions of interest include:
1.
Normal heatup 59 2.
Normal cooldown 3
Inservice leak and hydrostatic test The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50. The closure head region, reactor vessel outlet nozzles and the beltline region have been identified to be the only regions of the reactor vessel, and consequently of the reactor coolant pressure boundary, that determine the pressure-tempera-ture limitations concerning non-ductile failure.
The closure. head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting f rom bolt pre-load). This region largely controls 'the pressure-temperature limitations of the first several service periods. The outlet nozzles of the reactor vessel also a f fect the pressure-temperature limit curves of the first several periods.
This is due to the high local stresses at the -inside corner of the nozzle which can be two to three times the' membrance stresses of the shell. After the first several years of neutron irradiation exposure, the RT tempera-ture of the beltline region materials will be high enough so thak the beltline region of the - reactor. vessel will start to control the' pressure-temperature Proposed Amendment No. 59 3-3a O
9
r m
RANCHO SECO UNIT 1 TECHNICAL SPECIEICATIONS Limiting Conditions for Operation limitations of the reactor coolant pressure boundary.
For the service period
- for which the limit cursas are established, the maximum allowable pressure as a function of fluid temperature is cbtained through a point-by point comparison of the limits imposed by the closure head region, outlet nozzles, and beltline region. The. maximum allowable pressure is taken to be the lowest pressure of the three calculated pressures. The pressure limit is adjusted for'the pressure differential between the point of system pressure measurement and the limiting component for all reactor coolant pump combinations. The limit curves were prepared based upon-the most limiting adjusted reference tempera-ture of all the beltline region materials at
' full power year.
the end of the fif th effective 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 10. CFR 50, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Because the neutron energy spectra.at the specimen location and at the vessel inner wall location are essentially the same, the measured transition shift for a sample can-be applied with confidence to the adjacent section of the reactor vessel. The limit curves must be reca.lculated when the ART 59 NDT determined from the surveillance capsule is different from the calculated ART for the equivalent capsule radiation exposure.
NDT The unirradiated impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available.
The unirradiated impact properties and residual elements of the beltline region materials are listed in Table 3.1.2-1.
The adjusted reference temperatures are calculated by adding the radiation-induced ARTNDT and the unirradiated RTNOT.
The predicted ART NDT are calculated using the respective neutron fluence and copper and phosphorus contents in accordance with Reg. Guide 1.99 Figure 3.1.2-1 illustrates the design curves for predicting the radiation-induced ARTNDT as a function of the material's copper and phosphorus content and neutron fluence.
The adjusted RTNDT's of the baseline region materials at the end of the fif th' full power year are listed in Table 3.1.2-1.
The adjusted RT
's are given for. the 1/4 'T and 3/4 T (T is' wall thickness) vessel wall NDT locations..The assumed RTNDT of the closure head region is 60*F and the outlet nozzle' steel forgings is 60'F.
The 1 Imitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is
~
! operated wi thin. the design criteria assumed for the fatigue analysis pe'rformed in ~ accordance wi th the ASME code requi rements.
Proposed Amendment No. 59 3-4 a
j
.~
i MAXIMUM CURVE 300 FOR 0.15', Cu
- 167 MAX. AND PH0SPHATES a[
WlTHIN ASTM E
250 REQUIREMENTS 139 MAXINUM CURVE FOR 33 UNCONTROLLEO OR MAXIMUM CURVE UNDETERMINED Cu 200 FOR 0.10 Cu MAX.
111 g
=
SATISFIED BY g
E
^
B&W SPECIAL
~
n CHEMICAL ROMT'S.
=m y
150 83 5
M E
100 56 50 28 r
e lx10 I8 2
4 6 8 lx10 39 2
4 6 8 lx1020 6
8 lx102I 2
l Neutron Fluence n/cm2 witn E > 1 Mey EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RTNOT FOR REACTOR VESSEL STEELS EXPOSED TO 550 F TD'PERATURE Figure 3.1.2-1 i
i g '
BANCl!O SECO UNIT ONE i
TECilNICAL SPECIFICATIONS 2600 l
oo THE ACCEPrABLE PRESSURE AND TEMPERATURE COMBINATIONS ARE BELOW R 2400 AND 10 THE RIGHT OF THE LlHIT CURVE.
THE REACTOR HUST NOT BE h
MADE CRITICAL UNTIL Tile PRESSURE-TEMPERATURE COMBIN ATIONS ARED G
c TO THE RIGHT OF THE CRITICALITY LlHIT CURVE.
MARGINS OF 25 PSIG 3
AND 10F ARE ikCLULiD FOR POSSIBLE INSTRUMENT ERROR.
g
[ 2000 E
CRIIICAllTY g 1800 7
LIMIT (EFG) e
-)
e 1600 8
[
1400 P
T a
A 381 97 f,1200 200 m
C 525 283
/*
E APPLICABLE FOR HEATUP RATES OF < 100*F 1000 0 2250 312 l}
<50 F step change in any 1/2 hr period E
O 323 F 525 323 830 g
G 2250 352
}
600 B
C p
4G0 A
o" 200 -
i e
i I
E GD 100 140 180 220 260 300 340 380 indicated Reactor CO0lant System Temperature, Tc, *F REACTOR COOLANT SYSTEM. NORMAL OPERATION-HEATUP LIMITATIONS, APPLICABLE FOR FIRST 5.0 EFFECTIVE FULL POWER YEARS
RANCII0 SECO UNIT ONE TEClINICAL SPECIFICATIONS J'
2600 ft THE ACCEP T ABLE PRESSURE AND [EMPERATURE COMBINATIONS ARE BELOW D
AND TO THE RIGHT OF ruf LlHIT CURVE.
HARGlMS OF 25 PSIG AND 10F 00 ARE iMCtuotD FOR P0ssisLE i N s T RU:4ENT ERROR.
z
++
1.
. WHEN THE DECAT MEAT REH0 VAL (DNR) SYSTEN 85 OPfRATING WITH NO AC PUMPS I
2200 OPERATING, THE INDICATED DHR SYSTEM RETURN TEMPERATURE To THE REACTOR c
VESSEL $ HALL BE USED, e-E 2000 2.
A HAxlHUN $TEP TEMPERATURE CHANGE Of 75*F IS ALLOWASLE WHEN o"
REH0VING ALL RC PUHPS FROM OPERATION WITN THE DHR SYSTEM b
1800 OPERATING.
THE STEP TEMPERATURE CHANGE IS DEFINED AS THE RC IEMP (PRIOR TO STOPPING ALL AC PUNPS) HINUS THE DHR RETURN I
,y TEMP (AFTER STOPPING ALL RC PUMPS).
THE 100*F/ HR R AMP APPLICABLE FOR C00LD01N b
1600 DECREASE is ALLOWABLE BOTN BEFORE AND AFTER THE STEP E
TEH{ CHANGE..
RATES OF < 100*F.1ii
<50"F st'ep change in s
1400 l.
.n...../2 hour period ay1 a
E 1200 P
T 3
A 152 60 O
1000 B
500 183 0
C 500 215 000 0
860 221
[
E 1250 261 600 f 2250 377 6
C
.5 8
400 200 -
A 1
I I
f f
I i
60 100 140 180 220 2G0 300 340 380 InasCatea Reactor Coolant System Temperature, TC,*f REACTOR COOLANT SYSTEM, NORMAL OPERATION-COOLOOWN LlulTATIONS, APPLICABLE FCR flRST 5.0 EFFECTIVE FULL P0 lier YEARS g g,,
,,y
f.
l RANClio SECO UNIT ONE TECilNICAL SPECIFICATIONS j'~
i 2000 THE AECEPT ABLE PRESSURE AND TEMPER ATURE C0H8lNAT10h3 ARE DELOW E APPLICABLE FOR HEATUP cs Ag0 TO THE RICHI Of IHE LlHIT CURVE.
MARGINS OF 25 PSIG AND 10F RATES OF <100*F/HR O
~
ARE INCLUBE0 f0R POSSIBLE INSTRUMENT ERROR.
FOR C00LDOWN THE y
FOLLOWING NOTES ARE APPLICASLE.
<50 F step cha e in a
2200 any 1/2 hou g
I.
WHEN Tite DECAY HEAT REMOVAL (DHR) SYSTEM is OPERATiMG Wilit NO period RC PUMPS OPERATING.
THE INDICAIED DnR SYSTEM RETURN TEHPERATURE a-2000 TO TuE REACiOR VESSEL SuaLt SE uSED.
5 D
2 A HAXIMUM STEP TEMPERATURE CHANGE OF 75'F l$ ALLOWABLE WHEN
~
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REMOVING ALL RC PUMPS FROH OPERATION WITH THE DHR SYSTEM OPERATlhG.
lite STEP TEMPERATURE CHANGE IS DEFINED AS APPLICABLE FOR C00LD0nN RATES OF { 100*F/Hft 1600 Tuf RC itMP (PRIOR To STOPPlNG ALL RC PUMPS) MINuS THE DMR RETURN TEMP
<500F step change in any h
(AFTER STOPPING ALL RC PUMPS). THE 100*F/HR RAMP DECREASE IS ALLOWABLE 80TH BEFORE AND 1/2 hr period S
1400 AtTER inE STEP TEMP CHANGE.
m S
O 1200 P
T k
A 230 80 1000 -
j B
525 ISB C
525 255 800 -
y E
2500 323
~
600 -
o F
2500 W
8 C
1400 267
.S 400 -
C
_E A
200 e
i i
60 100 140 180 220 2G0 300 340 3G0
)
i
(
Indicated Reactor Coolant System Temperature. Tc. *F REACTOR COOL ANT SYSTEM INSERVICE LEAK AND HYORDSTATIC TE i
HEATUP AND COOLDOWN LIMITATIONS APPLICABLE FOR FIRST 5.0 (EFFECIIVE FULL POWER YEARS)
g.
RANCIIO SECO BELTLINE REGION MATERIALS i.
'}
Table 3.1.2-1 H7;ct Unirrad'd' Neutron Fluence @
Radiation-Induced ART Adjusted RT
. Numb *r RT~ T.
%Cu.
%P End of 5 EFPY-End of 5 EFPY*
Endof5EFFVT NDT ND 1/4 T 3/4 T 1/4 T.
3/4 T 1/4 T 3/4T
-f ZV 4281 10
.1'5
.009 2.40x1018 5.76x1017 56 28 56 38 C5062-1 4
. 12
.013 2.40x1018 5.76x1017 52 26 56 30 C5062-2
-10
.12
.013 2.40x1018 5.76x1017 52 26 42 16 C5070-l' 0'
.10
.G10 2.40x1018 5.76x1017 34 17 -
34 17
,C5070-2
-10
.10
.010 2.40x1018 5.76x1017 34 17 34 7
WF-233
+20
.22
.015 2.40x1018 5.76x1017 105 51 125 71 WF-29f*
+20
.16
.017 2.05x1018 4.98'x1037 74 37 94 57 WF-154
.+20
.20
.015 2.40x1018 5.76x1017 96 46 116 66
.WF-29**
+20
.16
.017 1.83x1018 4.42x1017 71 35 91 55 WF-70
+2'O
.27
.014 1.83x1018 4.42x1017 111 54 131 74' WF-29**
+20
.16
.017 1.83x1018 4.42x1017
'71 35 91 55 i
- Per Reg. Guide 1.99 i
- WF-29 in 3 different locations Proposed Technical Specification No. 59
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