ML20081K979
| ML20081K979 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/18/1991 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20081K976 | List: |
| References | |
| NUDOCS 9107020256 | |
| Download: ML20081K979 (11) | |
Text
. _ . - - . .
PROPOSED TECilNICAL SPECIFICATION CilANGES i
j 910702025i6' 910618 PDR ADOCK 05000368 P PDR
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1 3/4.4.9 PRESSURE / TEMPERATURE LIMITS RE/LG70R C00ldHI_ SYSTEM LIMITING _G_0 @1 TION FOR OPERAIE N 3.4.9.1 .The Reactor Coolant System (except the pressurizer) temperature
- and pressure shall be limited in accordance with the limit lines shown on
. Figures 3.4-2A, 3.4-2B and 3.4-2C during heatup, cooldown, criticality, and l '
inservice Icak and hydrostatic testing operations with: I A maximum heatup of 50 F, .60 F, 70 F or- 80'F in any one hour I
. a.
period in accordance with curves A, B, C or D, respectively, in- :
Figure 3.4-2A..
- b. .A maximum cooldown rate based on :
t-J RCS Temperature (Tel- Maximum Cooldown Rate T > 220'F- 100'F per hour (constant) or 50*F in any half hour period (step) 140"F 5 T" 5 220*F 60'F per hour (constant) or 30'F in any-half hour period (step)
T < 140'F 25'F per hour (constant) or 12.5 F
-in any half hour period (step)
- c. A maximum temperature change of s 10'F in any one hour period during inservice hydrostatic and Icak testing operations above the heatup and cooldown limit curves.
APPLICABILLII: At all times.
ACTION:
With any of the above.11mits exceeded, restore the temperature.and/or pressure to within the acceptable region of the applicabic curve within 30 l minutes; perform an engineering evaluation-to' determine the effects of the out-of-limit condition- on the. fracture toughness properties of the Reactor
-Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the naxt- 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200*F and less than 500 psia, respectively, witEin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l l
l ARKANSAS - UNIT 2 3/4 4-22 Amendment No.
)
ERYEILLAEE REQHREggs ,.
4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to ho within the limits at 1 cast once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Tabic 4.4-5. The results of these examinations shall be used to update l'igures 3.4-2A, 3.4-2B and 3.4-2C.
ARKANSAS - UNIT 2 3/4 4-22a amendment No. :
1 -.,
Figure 3.4-2A ARKANSAS NUCLEAR ONE UNIT 2 HEATUP CURVE - 21 EFPY REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS 2500 llllll 11 1 CURVE A - -+
ll1lll11 ! j
~~-~~ CURVE B g [ f;;
~~~
.{ CURVE D
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l 2000 TTTT~~~ i n.
ui EudEc _'1_ - ~
. _____ .! l g _____ __-__ __-___ .-_--- _. _
y .
L [
1500 I
', )) _ ACCEPTABLE g
W ._
f_ ..
REGION ---
g _ ___. ___.
3 LOWEST -
M SERVICE
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3
-~~~~~
1000
.i j j
@ TEMPERATURE l
170'F _.
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y _, ____ ___.. . _ __.__. _ ____
y ___ _____ ._ _
5- 500
--III 2 454 - --
15 ___ _
jy
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INIMUM BOLTUP TEMPERATURE (70'F) _ _ _
0 I l 0 100 200 300 400 500 600
! INDICATED REACTOR COOLANT TEMPERATURE T , F c
l l
l H'U LIMIT CURVE M'U RATE LIMIT i
A 50'F/HR B 60*F/HR C 70*F/HR D 80*F/HR (NCN-CRITICAL CORE)
E l 80*F/HR - (CRITICAL CORE)
ARKANSAS - UNIT 2 3/4 4-23 Amendment No.
l
Figure 3.4 2B j ARKANSAS NUCLEAR ONE UNR 2
-COOLDOWN CURVE- 21 EFPY REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS 2500 I
I 4 ._.__ _ ._. ._.
l y 2000 i CL l l
m _ ._. ___-
AC LE c'
4' _i g __ _..____ _.
f -- -_-__
E ._.--__ -_. . .__-_- -
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! U g ___ . _____
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COOLDOM ~~
1000 ' '' i
[* LOWEST g l l l n.
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i SERVICE --~T
! c TEMPERATURE-- L -_.
uJ 170'F J N l
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500 I 5 ,' '
l; z 454 - lll i'
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W l ~
T~. MINIMUM y .
BOLTUP TEMPERATURE -(70'F)
~~~{.- .--
0 100 200 300 400 500 600 INDICATED REACTOR COOLANTTEMPERATURE T , F c
l' RCS TEMP (Tc) C/D R ATE STEP
- T > 220*F 100*F/HR s 50*F IN ANY 1/2 HR PERIOD 140F s T s 220*F 60*F/HR s 30*F IN ANY 1/2 HR PERIOD T < 140*F 25'F/HR s 12.5'F IN ANY 1/2 HR PERIOD
- Not to exceed the specified instantaneous decrease in temperature with a subsequent thirty minute hold ARKANSAS - UNIT 2 3/4 4-23a Amendment No. j l
Figure 3.4-2C ARKANSAS NUCLEAR ONE UNIT 2 INSERVICE HYDROSTATIC TEST CURVE - 21 EFPY REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS 2500
. _ . . _ . _ _ _ _ _ ..! L .1.. , _ . _ _ _ _ _. .._ _ ..___
+
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u) _-__I I REGION N 1500 . h) ES T IIIIIII IIIIII!II k~j-1 SERVICEI lill l l C"
TEMPERATURE
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% 170 *F 1._.- .L_ . . _ . - . ____ _ _ _ ' -_ -~I - .- _ _ -.
E .-. , __ 5 INSERVICE --
3 l HYDROSTATIC I I t~~
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_.._._... _. ~
~] TEST TITHRT E
q.yLi]-1 li{_11 LUJJ Gr J_ .
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l __[ l 11 i UI I I I I TT D MINIMUM BOLTUP TEMPERATURE (70'F) . . .. _ .
0 llI i I Illl N Ibi I II O 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE T , F c
A maximum temperature change of < 10*F in any one hour period during inservice hydrostatic and leak testing operations ai>ove the heatup and cooldown limit curves. Otherwise, the heatup and cooldown limit curves apply.
ARKANSAS - UNIT 2 3/4 4-23b Amendment No.
4
-. .=
EEACTOR COOLANT SYSTEM i AASES .
steam generator tube rupture accident in conjunction with an' assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpm and a
- concur rent- loss of of fsite electrical power. The values _for the limits-on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative-in that specific site paramotors of the Arkansas Nuclear One site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.
The ACTION statement permitting p0WER OPERATION to continue for limited time periods with the primary coolant's specific activity ) 1.0 pC1/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure
, _3.4-1, accommodates possible lodine spiking phenomenon which may occur l following changes in TilERMAL POWER.
. Reducing T to ( 500 F prevents the release of activity should a steam get.erator ku6e rupture since the saturation pressure of the primary coolant is below the lif t pressure of the atmospheric steam relief valves. The surveillance requirements provide. adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on lodine spiking will be used to_ assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissiblo if justified by the data obtained.
! 3/4;4.9 PRESSURE / TEMPERATURE LIMITS All components in .the Reactor Coolant System are designed to withstand the ef fects of- cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.. The various categories of load cycles used for design. purposes are provided in Section 5.2.1.5 of the FSAR.
During_startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates do not exceed the design assumptions and satisfy the stress limits for cyclic p -operation.
Operation within the limits of the appropriate heatup and cooldown curves E assure the integrity of the reactor vessel against fracture induced by combined thermal and pressure stresses. As the vessel is subjected to L increasing fluence, the toughness of the limiting material continues to L decline, and even more restrictive pressure / temperature limits must bc
- observed. The current limits, Figures 3.4-2A, 3.4-2B and 3.42-C are for up l'
to and including 21 Effective Full Power Years (EFPY) of operation.
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l ARKANSAS - UNIT 2 B 3/4 4-5 Amendment No. 92,
REACTOR COOLANT SYSTEM EASES The reactor vessel materials have been tested to determine their initial RT I * * *"""
- I "'"* ""' "#* * "" '" ^
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NDT Reactor operation and resultant fast neutron (E > 1 Mov) Irradiation will cause an increase in the RT limit curves shown on Figure 3.4-2A, 3.4-2B anh.4-20 3. The heatup include and cooldown predicted r.djustments for this shift in RT at the end of the applicable service period, as well as N
adjustmentshT ar the location and for possible errors in the pressure and temperature sensing instruments. It should be noted that the location adjustment considered the operation of three RCPs from a RCS temperature of 70 F and above. The heatup, cooldown, and hydrostatic test limits are presented in tabular form in Table B 3/4.4-2.
The shift in the material fracture toughness, as represented by RT ,
is calculated using Regulatory Guide 1.99, Revision 2. For21EFpY,apD{he 1/4t position, the adjusted reference temperature (ART) value is 111 F. At the 3/4t position the ART value is 96 F. These values are conservatively based on a reactor vessel inner surface fluence of 3.74 x 10nyt. The fluence at the 1/4t point is 2.33 x 10nyt and the fluence of the 3/4t point is 9.06 x 10*"nyt. These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G to calculate heatup and cooldown limits in accordance with the requirements of 10 CFR Part 50, Appendix G.
To develop composite pressure / temperature limits for the heatup transient, the isothermal,1/4L heatup, and 3/4t heatup pressure / temperature limits are compared for a given thermal rate. Then the most restrictive pressure / temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event.
To develop composite pressure / temperature limit for the cooldown event,the isothermal pressure / temperature limits must be calculated.
The isothermal pressure / temperature limit is then compared to the pressure / temperature limit associatea with both the constant cooldown rate and the corresponding step change rate (an 1nstantaneous drop in temperature followed by a hold period). The more restrictive allowable pressure / temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline.
Both 10CFR part 50, Appendix G and ASME Code Section III, Appendix G, require the development of pressure / temperature limits which are applicable to inservice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the test pressure (1.1 times normal operating pressure) and locating the corresponding temperature. This curve is shown for 21 EFPY on Figure 3.4-2C.
Similarly, 10CFR Part 50 specifies that core critical limits be established based on material considerations. This limit is shown on the heatup curve, Figure 3.4-2B. Note that this ))mit does not consider the core reactivity safety analyses that actually control the temperature at which the core can be brought critical.
ARKANSAS - UNIT 2 B 3/4 4-6
~
TABLE B 3/4.4-2 ARKANSAS NUCLEAR ONE UNIT 2 21 EFPY - TECHNICAL SPECIFICATION PRESSURE-TEMPERATURE LIMITS BEATUP HYDROSTATIC COOLDOWN BELTLINE BELTLINE BELTLINE COMPOSITE CURVE COMPOSITE CURVE COMPOSITE CURVE P-ALLOWABLE P-ALLOWABLE P-ALLOWABLE (PSIA)
RCS 60 F/ 70 F/ 80 F/ (PSIA)
TEMPERATURE (PSIA) ISO 50 F/ HOUR THERMAL HOUR BOUR FOUR DEG. F 1 433.8 419.7 406.5 393.8 655.0 1 70 358.6 448.4 l
72.5 368.6 -
406.5 393.8 675.E I 464.0 433.8 419.7 80 -
)
378.6 82.5 65 388.6 -
406.5 393.8 699.8 482.0 433.8 419.7 90 -
95 408.6 -
418.6 97.5 419.7 406.5 393.8 727.5 100 - 502.8 433.8 428.6 102.5 - -
107.5 448.6 -
393.8 759.5 439.2 420.2 406.5 110 - 526.8 796.6 452.6 427.9 409.3 393.8 120 488.6 554.6 399.1 839.4 586.7 472.1 442.1 418.4 130 -
132.5 528.6 -
434.1 409.7 889.0 498.6 623.9 497.8 462.6 140 489.2 455.7 427.0 946.2 150 558.6 666.8 529.5 1012.4 567.7 522.4 483.5 449.7 160 628.6 716.4 479.5 1988.9 613.9 562.4 518.1 170 698.6 773.8 560.1 515.4 1177.4 788.6 840.2 668.0 610.1 180 666.3 609.8 559.6 1279.6 888.6 916.9 731.0 190 732.1 668.2 611.3 1397.8 998.6 1005.5 893.9 200 808.7 737.4 673.4 1528.2 1108.0 1108.0 888.2 210 897.8 817.9 744.9 1678.0 220 1226.5 1226.5 987.6 1851.3 1102.3 1001.2 911.1 830.0 230 1363.4 1363.4 927.3 2051.5 1521.8 1234.4 1121.1 1019.7 240 1521.8 1146.1 1042.4 2283.0 1704.8 1704.8 1386.7 1259.9 250 1420.5 1291.9 1173.7 2550.7 260 1916.5 1916.5 1562.1 1768.1 1606.4 1460.0 1328.4 -
270 2161.1 2161.1 1504.7 -
2443.9 2005.6 1821.3 1656.6 280 2443.9 1883.2 1712.2 -
2770.8 2770.8 2279.1 2070.0 290 2144.3 1948.5 -
- 2594.0 2357.5 300 -
- 2690.0 2447.0 2226.3 -
310 -
- 2798.3 2542.5 -
- +
320 AMENDMENT NO.
ARKANSAS - UNIT 2 B3/4 4-9
..- p.
l' l
ggS The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic test pressure l: (624 psia). This temperature is defined as equal to the most limiting
- l. RTNDT_f r a ance f e eactor. Coolant System component
'(conservatively estimated as 50'F) plus 100'F, per Article NB 2332 of Section III.of the ASME Boiler and Pressure Vessel Code. Temperature
- i. Instrument uncertainty _is conservatively estimated as 20*F.
l l.
The horizontal line between the minimum boltup temperature and the '
L Lowest Service Temperature is defined by the ASME Boiler and Pressure Vessel Code as 20% of the pre-operational hydrostatic test pressure.
u -The minimum boltup temperature is the minimum allowable temperature at l, pressures bel _ow 20% of the pre-operational system hydrostatic test pressure. The minimum is defined as the initial RT f r the material of NDT L the higher stressed. regic;n of the reactor vessel plus any ef fects for
! irradiation per Article G-2222 of Section III of the ASME Boller_and Pressure Vessel Code. The initial reference temperature of the reactor vessel and closure head flanges was determined using the certified material I l
l- test reports and Branch Technical Position MTEB 5-2. The maximum initial RT associated with the stressed region of the vessel flange is 30 F.
NDT The minimum boltup temperature including temperature instrument uncertainty is 30 F + 20 F = 50 F. liowever., for additional conservatism, a minimum -
boltup temperature of 70'F is utilized.
The number of reactor vessel irradiation surveillance specimens and the frequencies-for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix 11 to 10 CFR -
l- Part 50.
The limitations imposed on the pressurizer heatup and cooldown rates i are provided to assure that the pressurizer is operated within the design l' criteria-assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
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L ARVANSAS-- UNIT 2 B 3/4 4-10 Amendment No. 49, l
4 i ATTACHMENT 1 ABB-COMBUSTION ENGINEERING REPORT A-MPS-ER002
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