ML19317E854
| ML19317E854 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/04/1977 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317E852 | List: |
| References | |
| NUDOCS 8001031025 | |
| Download: ML19317E854 (19) | |
Text
C%,;
[~-
.f UNIT ED STATES
, CLEAR REGULATORY COMMisslON 4,
3 r7 f ' (f WASWNGTON, D. C. 20L55
- g. w.'.... f DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO. 1 A!!ENDMENT TO FACILITY OPERATING LICENSE Amendment No. 51 License No. DPR-38 l.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duke Power Company (the licensee) dated June 6,1977, 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 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.
8 0010310.Q,[
)
J 2
2.
Accordingly, tne license is amended by changes to the Technical Specifications as indicated in the attachment to this license amencment and paragraph 3.B of Facility License No. DPR-38 is herecy amended to read as follows:
3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.51, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
Thi s license amenemer.6 is effective within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY C0rtMISSION
'llWlW6&
).
A. Schwencer, Chief Operating Reactors Branch al Division of Operating Reactors
Attachment:
Cnanges to the Technical Specifications Date of Issuance: November 4,1977 i
W
4
)
UNITE 3 STATES f
4 NUCLEAR RE!ULATCRY COMMIS$10N
-'#2
// y CASHINGTON, C. C. 20585 j
DUKE POWER C0WANY OOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 51 License No. OPR-47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
The application for amendment by Duke Power Company (the licensee)
A.
dated June 6,1977, 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 confonnity 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 tne 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.
l l
i
(
,,.,)
)
.. 2.
Accordingly, the license is amended by changes to the Technica'r Specifications as indicated in the attachment to this license amendment and paragrapn 3.8 of Facility License No. DPR-47 is hereby amended to read as follows:
3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised througn Amendment No. 51, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amencment is effective within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/ /l
.8tt*6'W!%
s A. Schwencer, Chief Operating Reactors Branch 71 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 4,1977 a
e
((q -
'4 UNITE 3 STATES
- 4
, JCLEAR RE!ULATORY COMMISSION
}*
[
CASHINGTON, C. C. 20666 e
%,.m /
DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 48 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 June 6,1977, 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 witVthe 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 issaance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
i l
i l
L 1
p i
2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.5 of Facility License No. DPR-55 is hereby amendec to read as follows:
3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 48, are hereby incorporated in the license. The licensee shall operate tne facility in accorcance with the Technical Specifications.
3.
This license amendment is effective within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
((.d ld( h y
A. Schwencer, Chief Operating Reactors Branch al Division of Operating Reactors
Attachment:
Changes to the Technical Speci fic ations Date of Issuance: November 4,1977
-)
i ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 51 TO DPR-38 AMENDMENT NO. 51 TO DPR-47 AMENDMENT NO. 48 TO DPR-55 DOCKET NOS. 50-269, 50-270 and 50-287 Revise Appendix A as follows:
Remove the following pages and replace with identically numbered page2.
3.1-3 3.1-3a 3.1-4 3.1-5 3.1-6a 3.1-7a 3.1-7b 3.1-8 3.1-9 Add the following new pages:
3.1-6b 3.1-7c 3.1-7d
1 1
3.1.2 Pressurization, Hestup, 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 3.1.2-2B Unit 2 3.1.2-2C Unit 3.
3.1.2.2 Leak Testa Lerk 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 ASHI 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 iOOoF/hr.
The spray shall not be used if the temperature dif ference 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.
These revised limits shall be submitted l
to the NRC at least 90 days prior to exceeding four (Units 1 & 2) l effective full power years of operation or an integrated exposure 2 or DTT 144*F for Unit 3.
l of 1.7 x 10 M n/cm 3.1-3 Amendments Nos. 51, 51 & d8 1
Bases - Unica 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 lead 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, art given in BAW-1421(7) and BAW-1437(8).
l The ficures 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 l
full power years of operation.
These curves are adjusted by 25 psi and 100F for possible errors in the pressure and temperature sensing instruments.
The pressure li=1t 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.
l The pressure-temperature 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 for hydrostatic testing have been provided to assure ccmpliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice hydrostatic testing, The actual shift in RTgg; of the beltline region material will be established periodically during operation by re=oving and evaluating, reatter vessel material irradiatien surveillance spe:imens which are installed near the inside wall of this or a similar reactor vessel in the core areas, or in test reactors.
The limitation on steam generator pressure and temperature provide protection At against nonductile failure of the secondary side of the steam generator.
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 esti=ated 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 The limitations include margins of 25 psi and 100F for coolant temperature.
possible instrument error.
The spray temperature difference is impose.d to maintain the thermal stresses the pressurizer spray line nozzle below the design limit.
at 3.1-3a Amendments Nos. 51, 51 & 48
_Ba s e s l' nit 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 1C00F per hour satisfies stress limits for cyclic operation.
(2)
The 237 psig pressure limit for the secondary side 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 and a maximum NDTT value of 200F has been determined based on Charpy V-Notch The maximum NDTT value obtained for the steam generator shell material tests.
and welds was 40cF.
Fiaeres 3.1.2-1C 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.
Thesc 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
=aximum SDTT 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 sa=ples 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 MWt ratad power and an integrated exposure 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 exposure for 40 years operation at 80 percent load.
(4)
Figure 3.1.2-1C is based on l
the design value which is considerably higher than the calculated value.
The DTT value for Figure 3.1.2-lC 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 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 N!!! will be established periodically during plant operation by testing vessel =aterial samples which are irradiated cumulatively by securing them near the inside wall of this or a similar yessel in.the core area or in test reacters.
To compensate for the increases is the SDTT caused Joy-- =
irradiation, the linits on the pressure-temperature relationship are periodicelly changed to stay within the established stress limits during heatup and coolcowr..
3.1-4 Amendment Nos. 51, 51 & 48 i
The NDTT shift and the magnitudes of the thermal and pressure stresses are sensitive to integrated reactor power and not to instantaneous power level.
Figure 3.1.2-3C and 3.1.2-2C are applicable to reactor core thermal ratings l 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 23 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 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 3.1.2-1C.
During this two year period, a fracture toughness criterion applicable to Oconee Unic 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 Ocenee Units 2 and 3 reactor surveillance programs.
The spray temperature difference restriction is imposed to maintain the 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)
FSAE Section 4.1.2.4 (2)
ASMI 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.
(S)
Analysis of Capsule OC11-C from Duke Power Company Oconee Unit 2 Reaeter Vessel Materials Surveillance Program, BAR-1437, May,
.1977.
Amendment Nos. 51, 51 & 48 3.1-5
2400 i
i t
O OCONEE UNIT 11
~
2200 E
reing ragss.
itur.
f.,2000 A
487 70
$ 1800 e
645 158 C
649 283 0
2250 312 g
1600 E
O 283 w
p ses 2e3
(
a sus 323 APPLICABLE 80R NEATUP RATES
~
Y s
of < 100*flNR E 1400 n
2250 352 t,
( < 50*f la Any 1/2 Nous PERIOD)
Catiltalliv 3 1200 Laust
~
(epsNI O3
=
1000 3
u 800 n > 2 :o c-6
-oOm=
e c
O 'oy%QH 3
G00 F
D' er.
b.i..
e.d t. te,.,i,bi.. ih.
1h. eu.,iebi.,,...........rai.e.
c..bineii.n.
Rg gr gm m
3 -5 M
OJ O E
ll.it c.rv.(s).
Th. II.it c.rv..
Incl.d. th. pr....r.
diff.r.nti.I b.tw.en th.
400
.....,,...,,........e......ni ead ih.,r....,..a ti. r.ect.r.....
r..
.n g"'D mmg contr.Illag th. ii.it c.rve, th.y incl.de en additional.ergin.f.ef.ty for Z roo c
3f g yyp 6
C g-O --4 >
f 200
- lbl*
la**'***"t.rrer (25 pel and 10*f).
ih. c.ector...t not be.ed.
2*
N 2
critical.ntil the pre...r..te.,aret.re
- e. bla.tl.a. er. 1. th. right of th.
N.
o.-t en '
y
',_,,2 c ri tic.l lig II.i t.
l l
llE l
l g
- as
,o x -<
0
--4 t
40 80 120 160 200 240 280 320 360 C
-4 "9f g
Indicateo Reactor Coolant Temperature Tc, 'I d.E 0 o F
Z m-
- v1 3 m-O
-< -4 W
--4 s-O2 v1 3
i
POINT TEMP.
PRESS.
A 40 550 B
275 550 C
275 1400 2400 0
380 2275
\\
2200
-D 2000 m.
1800 3
UPPER PRESSUR12ATl0N a
LIMIT 1400
-C Z
E 1200 a
8
[
1000 3
o*
600 -
=
2
~
3 600 i
E A
~
4 00 MAXIMUM HEATUP RATE,*F/HR 200 5t 100
=
0 I
I I
I O
100 200 300 400 500 215 Indicated Reactor Coolant System T:mperature,'F b\\
bu rset s.
OCONEE NUCLEAR STATION Uni: 3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS
( A FPLICABLE UP TO AN INTEGRATED EXPOSURE OF 1 7 x 10 38 n cm2 OR DTT 144 'F)
F i gu r e 3.1. 2 -iC Snendment Nos. 51,51 &4E 3.1-6 b r
4 oo I
6 I
I i
i i
i e
i f
s e..
.z..
. e
_. ana
,s e
vi 6 -
o e-
=.,
- - o -
. - o e
_u a
,a.
s e~
e.s e G # # M f 48 E
88 S S W a e.. e ee
- =
ae e h b
s, O en e
e*
e e 8 g
m et
-d 5
3 9 - **
a=
3 O
a -
e O
E e D E C g
6 O -
c.
O e
a B 9 es, a.
g eg S U O t.e O
O b G
- 8 g
an
- 3 m -
g e e 3 4 w
g, A e e 4 te 9 e WO
.E 5
O m 9 O
O b
a b F
E e
e e -
ab e.e O
4 e R 9 9 6 a
es es.
e es a g e
. 6
.4
.e
=
B e 0
- a g g e.
O u
C
- G
- W N
3 >=
D u es.
d e*
N 9 U =
- e. a.s e W E 5 e g.
J
$ a e - S p
ie e.
O V0
- U U W 4
ed
==
mer A a es a.
B n.e a-a w
o a O
e A mP m.
3 a
E a=
W e S
- 8 g
- '8 6
se O.
E
- E e
==
m b
3 e.
O g
g E O g 3 e
3 3 8 h
=
gy e_m E
8'*
3D 3 e y n,
p,
'ds *=
=med 3,
i.,
=
e e.
t.
3 e.e m e
.e.
as O
g g
B em e B at *=
U
.g es e O
- '8 O 88 #
e g e - S i
a=
a S== m 4
d E E W ao 3 -
e*
l 9"8 E
%'8 5 3 g
eE et p.
- '8 O
a E B E m
,g 4.
5.e i
I e, e==
g
'O'8 e A e e
a a 1
e.
g
- 3 O
O -
E g,,,
- O 8
m >= m e -
=== =. u c
a = o "
a 2
O **
- o==
e e,
- z - =
m j
.e
. O e
g x *.
3
.u 3
- =.
.e u
E ~ " *
, e _
e.
==
.x
- w =-
=
-. =. =. -
o,
a e
=.
= a o
.-=
=
e..a..
o
- se
_e -
- s = =
e.
=
- o
=.x.
.z
=
a.
e~e, x
-=*
- y 3.2
=
- - - - e
=
c g 2, ;;
3_
o_.,
a e.nsE_
"'."85-
.=
x -. =
x =- o
8 8*
B
- W' D
. o 8
,8 a w a a vy-m
. =.- s
- -a a
u a,
. s o.., o.
e
=
~wo. >-
"[h
)
s E 5 - W.
3m = _
-g.
c
_ =
w g.
,c o
i s
.- - ---. s
. e
. O N G.
S.
e5.
3
.z x x-x xu Oa 1
1 1
1 I
I f
1 I
I f
e,
)
o e
o.
.o
.e e
o o
a o
e o
o o
=
o.
.o o.
o, o
o c.
e o
c.
e.
n n
e.
e.
(JartinssaJd utta dool) I!sd 'aJnssaJJ luttoo3 J013:33 palestoul i
1 I
1 UNIT 2 REACTOR COOLANT SYSTEM
)
NORMAL OPERATION C00LDOWN LIMITATIONS 1
APPLICABLE FOR FIRST 4.0 EFPY Ct.. met :
OCONEE NUCLEAR STATION Figure 3.1.2-2B 3.1-7a Amendment Nes. 51, 51 & 48
~
PolNT TEMP PRESS A
380 2275 8
275 1A00 C
275 550 0
250 550 E
250 A50 F
175 A50 2A00 -
G 175 200 g/
M 120 200 RC PuuP C0maiNAil0NS ALL0 SABLE 2200 -
AB0VE 185F Alt y 2000 -
BELOW 185F l A.1 8;0- A 2 B;1 1. 0 9;0 A.1 8 t1) SHEN DEC1Y HEAT REuGVAL SYSTEu (OH) IS
{ 1800 -
OPERATING tlTHOUT ANY RC PUNPS OPERATING.
INDICATED OH RETURN TEMP TO THE REACTOR O
[ 1600 -
VESSEL SHALL BE USED 8
(2) IN THE TEMPERATURE RANGE 260F TO 175F. A Milluuu STEP TEMPERATURE CHANGE OF 75F
~
CL iS ALLORABLE FOLLOWED BY A ONE HOUR 1200 MINiuuu HOLO ON TEMPERATURE. IF THE STEP CHANGE IS TAKEN BEL 0s 250F RC TEMPERATURE.
8 THE Milluuu ALLOWABLF STEP SHALL BE TH AT
$ 1000 WHICH YlELOS A FlW".L TEuPER ATURE OF 175F.
C THE STEP TEuPERATURE CH ANGE IS DEFINED AS B00 RC TEMPERATURE (BEFORE STOPPING ALL RC PuuPS) 0 ulNUS THE GH RETURN TEMPERATURE TO THE REACTOR f
600 VESSEL.
C l
-r 3
/
A00
/
[
E g
g i
p 200 UPPER PRESSURI Z ATION _
Lluli 0
24ttuuw C00LD0sN RATE
- F HR
,- (2) q l
1 100
-l SC I
l t
t t
i i f
f f f 4
1 260 i 175 120 i
$30 275 600 500 A00 300 200 t00 II' inci:stec Reactor C:otant System Temcerature *F REACTOR CC0LaNT SYSTEu C00LD01N Llulilil0NS
( AFPL t C ABL E UP TO OTT = 135'F1 Unit 3
\\
{ont roan, OCONEE NUCLEAR STATION Y
Figure 3.1.? - 2 c Artendment Nos. 51, 51 & 43 3.1-7 b
j1900
- 2 0
1 2500 b
IHL ACCEPIABLE PRES 5URE AND 1EMPERA100.0 COMulNATIONS ARE DELOW AtJD TO THE HlCHT Of THE LIMIT CudVE. MARGINS OF
'; 2300 25 PSIC AND 10f AftL INCLUDED FOR POSSIBLE INSTRUMENT a
ERROR.
TOP. COOLDOWN, NOILS I AND 2 ON FIGURE 3.1.2-2A AME APPLICABLE.
2l00 __
c
!1900 g
{1700 O
APPLICABLE FOR HEATUP OR COOLDOWN E 1500 RATES OF s. 500F/HR ( s 250F IN ANY g
i/2 HOUR PERIOD).
O i
a 1300 l u C
c
[
31100 n
a lo NI PNi$$ UNI IIMPlNAIURI j
900 A
s2s 70 a
s2s eso I
(
inac i90 I00
~
o 2soo zia o
A 8
500 E
o 1
1 l
l l
l l
l 1
1 1
1 1
l l
1 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 370 340 I$k Indicated Reactor Coolant System Tempesature ic.*F UN(Tl ljg REACTOR COOLANT SYSTEM HEATUP AND d5 C00LDOWN LIMITATIONS FOR INSERVICE f
HYDROSTATIC TESTS (NO FUEL ASSEMBLIES IN THE CORE), APPLICABLE FOR FIRST m
. ~. '
- l. EFPY OCONEE NUCLEAR STATION Figure 3.1.2-3
I I
I I
I I
I I
I P
o OCONEE UNIT 11 2400 -
C 2200 Zo
.arcticastt von ntatur satts or 2000 -
..,f,,
,e
( < so ** in any s/2 noun etnico) 5 1800 2
1600 entss igu, e Point u
o g
a as' so arrticants ton cootoons natts or <
~
g E
1400 e
eso saa c
ago 255 g
h o
2500 282
( < so*r in any 1/2 nova reasool 1200 E
isos as.
~
F 2500
.08 a.
g 1000 f
.00 e
e o
600 O > x c
=
O um2mz
'h' ***** ''*-*** " ' **""****"*
i
'O M 'A > *-*
.elow es. to the right of the limit curve (s),
the
~
)OZ [
hh iloit curve. i. cia.e the pre.sure.iffere. ties.etwee.
p' j
g N
the pol.t of system pres.ure seesurees.t
..d the pressure 3
Ct3 O n
'u~
Q.
g 0
200
.. the reactor vesnes resi.a c..tr.ili.e the simit curve. they i.ci..e.. e..
u..ei.er...,.e e,
e.,.....ie i..
r..e.t e,ror(25,.i e..
io,)
mm o
bw n.,, g c o
.- r-o r 9 5-
- r., co. i.e w....ie...
.. 2.. vi..r. 3.i.2-2. are e,,iice.i..
I I
m i
m :cox2 71 r >
40 80 120 160 200 240 200 320 360 400 440 2 N >
o
-i s
3 e w
- z z en o
CD (pg 30 0 -<
Indicated Reactor Coolant Temperature. Tc. 'I
-4 $ C r "-t m
2<m w
-4 A (O 3
- o m
O ovo 2
-t m m - -t 9*
-71 o >
"U 2 -1 S=
-< (n -
O
~
e 3.1.3 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 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 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 be maintained suberitical by at least 1%Ak/k until 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 4
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 5250F is prohibited except where necessary for icw power physics tests.
The potential reactivity insertion due to the moderator pressure coef icient(2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximate 1y'O.lak/k.
In addi-During physics tests, special operating precautions will be taken.
tion, tne strong negative Doppler coefficient (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 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 Amendment'Nos. 51, 51 & 48
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 17. 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 d.' ring 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 l
I f
1 I
Arendment Nos. 51, 51 & 48 3.1-9 c.
av, a.
-a