ML20128K127
| ML20128K127 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 02/11/1993 |
| From: | ILLINOIS POWER CO. |
| To: | |
| Shared Package | |
| ML20128K097 | List: |
| References | |
| NUDOCS 9302180076 | |
| Download: ML20128K127 (6) | |
Text
-
TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES E
a f
RESPONSE TIME (Seconds)
Qc FUNCTIONAL UNIT
% O.;
3*
1.
NA
> tn
- a.
Heutron Flux - High NA 88 b.
Inoperative nM 2.
Average Power Range Monitor":
NA oc a.
Neutron Flux - High, Setdown
< 0.09**
88 b.
' Flow Blased Simulated Thermal Power - High c.
Neutron Flux - High j 0.09 ~
yg$
NA ca-d.
Inoperative
\\zw
< 0.33 i
3.
Reactor Vessel Steam Dome Pressure - High
~
7 1.03 4.
Reactor Vessel Water l'evel - Lcw, Level 3 7 1. 03 2
5.
Reactor Vessel Vater Level - liigh, Level 8 7*0 04 w
6.
Main Steam Line Isolation Valve - Closure RA w4 7.
Main Steam Line Radiation - High NA 8.
Drywell Pressure - High 9.
Scram Discharge Volume Water Level - High NA a.
Level Transmitter NA b.
Float Switches
(
< 0.04
)
10.
Turbine Stop Valve - Closure l
11.
Turbine Control valve Fast closure, Valve Trip System g
< 0.05 Oil Pressure - Low RA 12.
Reactor Mode Switch Shutdown Position NA I
13.
I
- Neutron detectors are exempt from response time testing. Response time shall be measured from the detector
{
output or from the input of the first electronic component in the channel.
- Hnt including a simulated thernal power time -constant ef C 0.0 ::: nds.hedNG IE fob
- Measured from start of turbire control vhlve fast closure.
TR*"
- S * ? E a~$i 8,8a t
,5" O
s u
TABLE 4.3.1.1-1 n
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS S
CHANNEL OPERATIONAL s:
CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH Z
FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)
SURVEILLANCE REQUIRED Q
+
1, h 1.
$0f) a.
Neutron Flux -'High S/U,5,(b)
S/U(c),W R
2 LDI l
81-15 @
S W
R 3, 4, 5 b.
Inoperative NA W
NA 2,3,4,5 h" "&
Average Power Range Monitor:(f)
[gl 2.
a.
Neutron Flux - High, S/U,5,(b)
S/U(c),W SA 2
1 O T.'
Setdown S
W SA 3, 4, 5 B7-15 [,B5 5
F Cl w
b.
Flow-Biased.Siculated E
Thermal Power - High S
S/U(c),y g(d)(e) SA,R(I) 1
' b(
c.
Neutron Flux - High S
S/U(c),y g(d)(e) SA 1
K..,
{2.'%
d.
Inoperative.
NA W-NA 1,2,3,4,5 3.
Reactor Vessel Steam Dome
['M Pressure - High S
M R(9) 1,2(I)
( : _$, l 4
Low, Level 3 5
M R(9' 1, 2 L O' 3
I 3 r
5.
/[
High, Level 8 S
M R(g) 1 6.
Main Steam Line Isolation k
Valve - Closure NA M
R 1
gy g
.m n
3*?%
E 7.
Main Steam Line Radiation -
1,2(3)
.E"$i
~
High S
M R
?
ME 8.
Drywell Pressure - High 5
M R(9) 1, 2(I) e3"w
.m
r e m
wE-
- Ua OQ*
m 3
hc.s sb&9m1Qc c
(! h D
IER ICf l u iQME R
5 5
N IE L
C 4
4 ASN
)
)
N!! A k
k OOL
(
(
3 3
IIL 5
5 TTI A!E 2
2 R0V 2
2 EHR S
PCU T
OCS 1
1 1
1 1
1 H
EH ER I
UQ E
)
R a
(
E H
C 0
N 1
A T
L LA L
ER I
la!
)
)
)
D dEV AL 9
l e
R A
I I
(
A A
l l
u U
CC A
R R
R lf n
S i
t N
n O
o I
L C
T A
(
A T
LO N
EI 1-E NT 1
M
! CT
)
)
t Al S l
(
(
1 R
ll uE S
R H
T CfT H
Q H H
3 N
4 I
E H
L E
B 1
L A
s E
T Y
lHK S
l C AE C
CC S
H I
A A
A N
l l A
A l l H
H I
1 i
1 l
C O
E T
r er m
0 e
u t e t
P a
s st W
o as l
fy RO e
r C
S T
m e
e C
u t
vp hn A
l t
s ll E
o i
e e
ar co R
V m
h v
VI ti w
it s
c l
e n
t a
l eo wi gh a
i V
ovL Ss rg r
w rl o
ai T
S p
t a-eP m
hi o
nV d
a 1
c l l
t t
o e
on r
1 s -
e a
S Cer Mw c
11 l
v o
ru o
S 0
Ol e
i e
eus rd e
L F
n nss ot l
i i oe t u a
L mv lA ae b
bl r ch u
l rL r
rCP aS n
u u
e a
0 c
T T
R M
I S
a h
T C
li 0
1 2
3 i
1 1
1 1
f 9
d#
4 n e.-
c
- 4
-Attachment 3 to U-602085 8
~
TABLE 4.3.1.1-1 (Continued]
004 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
,TABLC NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM and SRM channels shall be determined to overlap'for at least 1/2 de-cade during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1 decade dur-ing each controlled shutdown, if not performed within the previous 7 days.
(c) Within 24-hours prior to startup, if not performed within the previous 7 days.
(d) This calibration shall consist of the adjustment of the APRM channel-to conform to the power' values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER.
Adjust the APRM channel if the absolute diTference is greater than 2% of RATED THERMAL POWER.
(e) This cal-ibration shall consist-of a setpoint verification of the Neutron Flux-High and the' Flow Biased Simulated Thermal Power-High trip functio.ns.
The Flow Biased Simulated Thermal-High trip function is verified using a calibrated flow signal.
(f) The LPRMs shall be calibrated at least once per 1000 offective full power hours (EFPH).using the TIP' system.
(g) Calibrate the analog trip module at least once per 31 days.
(h) Deleted.
(i) This calibration shall consist _of verifyin C2^.C' n s;.d'simuIated thermal power. time constantkM4k)phs IW6 yelili(inW (4 tith (j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(k) With any control rod withdrawn.
Not 'applicab'le to cont'rol rods remov$d-
~
per Specification 3.9.10.1 or 3.9.10.2.
(1) This function is not required to be OPERAB'LE when DRYWELLLINTEGRITY is-not required to be OPERABLE per Special Test Exception 3.10.1.
(m)' The CHANNEL FUNCTIGNAL TEST and CHANNEL CALIBRATION shall include the turbine first: stage pressure instruments.
1 CLINTON - UNIT 1
.3/4 3-10
. Amendment No. 30
__________m.-_m_m_.________-
to U-602085 b shall consist of a statrix of airca?oy clad fuel rods with en initial composition of nstural or sl:e,htly enriched uraniws
[ Page 5 of 6
, dioxide as fuel material, and water ro.Ns).
Limited substitutiens LS-92-004 of zircon 1un alloy or stairdens steel fitler rods for fuel rods.
in eccordance with NRc-approvud applications of fuel.od configurations. may be used.
Fuel assemblies shall be limited to
[,
i l
those fuel designs that have been analyzed with applicable NRc y
staf f approved codes and methods, and shown by tests or analyses I to comply with all fuel safety design bases. A limited nua.ber of 1ead-veeassembliesthathavenotcompletedrepresentativetesting}
n may be placed in non limiting core regiors d
DFJIGNFEATURc5 j '- ^~
"~
~
SiCONDARY CONTAINMENT
- 5. 2. 3 The secondary containment consists of the fuel 'uilding, the ECCS puam rooms and the containment gas control boundary, includi g extension, and has a minimum free volume of 1,710,000 cubic feet.
l 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 624 fuel assemblic i+h-each-4e+1 = rWy-sent.Mnia%2 f:P r & 2nd U: ic vedt,- t444-with,7 i rca. :r 2.
E a c 4-f,al--eod-shatl-hew a. e. inal eetive-fueHength-of-lW~irscha The-inttieb eere--leedirig-sh*M-have-a-core-average ceic4enwr L70C uc!9ht-pereent-t4-GE. Echad
'; :.1 hall bc shHet -in-physieel-design-to 'the~icitiet-core-loadinge-CONTROL ROD ASSEMatIES con'f-' -----.s k.
h-tca;ist-5.3.2 Thet actor core shall contain 145,, control rod assemblicsz hi-;f - cruciform-ereey*f-eteinlest-+ tee 4--tubes-eoataining--143.-70 '=heH4-heon-c,wM4c, B,C, ;miee-eeecewoded-4.y-weoe4few-+haped-steieden +6ee4 _ _J-asifHal 6 hall be beron& ~n U.. % ea,b hie.
ea4/dc 5.4 REACTOR COOLANT SYSTEM 4
DNt*~ $ %. D c.nd er ha n;ms melal.J-i DE51CH PRESSURE AND TEMPERMURE
--m g
>_m S.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the cou requirements specified in Sectior 5.2 of' the FSAR, with allowance for normal degradation pursuaat to the applicable Surveillance Requirements, b.
For a pressure of-1.
1250 psig on the suction side of the recirculation pump.
2.
1650 psig from the recirculation punp discharge to the outlet side of the discharge shutoff val u.
3.
1550 psig from the discharge shutoff valve to the jet pumps.
c.
For a temperature of 575 F.
V01.UME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 16,000 cubic feet at a nominal steap dome saturation tetroerature of 549'F.
1
~..u,...
....n r '
5-5 Amendment No. 53
to U-602085 Page 6 of 6 LS-92-004 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETP0lRf5 (Continued)
P Average Power Range Monitor (Continued]
4 i
The APRM trip system is calibrated using heat balance data taken during steady-state conditions.
fission chambers provide the basic input to the system and therefore, the monitors respond directly and quickly to char,ges due to transient operation for the case of the Neutron Fiex-High setpoint; i.e; for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the n:cutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow-Biased Simulated Thermal Power-High setpoint, a time con-l stant of 5 0.0 ::cende'is introcuced into the flow-biased APRM in order to simula e the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.
In thes flow biased equations, the variable W is the loop recirculation flow as a per entage of the total loop recirculation flow which produces a rated
[~ 4. d ( Q 'i,. X W L g-care flos cf 84.5 million 1bs/hr.
The APRM setpoints were selected to provide adequate margin for the' Safety Limits ano yet allow operating cargin that reduces the poscibili.ty of unneces-sary shutdown.
3.
Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fissioa products. A pressure i
increase during operation will also tend to increase the power of the reactor by compressing voids, tnus adding reactivity. The trip will quickly reduce the -
neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of.the pressure measurement com-pared to the highest pressure that occurs in the system during a transient.
This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure and turLine control valve fast closure trips are bypassed.
For a turbine trip or load rejection under these conditions, the transient I-analysis indicated an adequate margin to the thermal hydraulic limit.
4.
Reactor Vessel Water level-Low The reactor vessel water level trip setpoint has been used in transient analy-ses dealing with coolant inventory decrease. The scram setting was chosen far enough telow the normal operating level to avoid spurious trips but high enough above the fuel La assure that there is adequate protection for the fuel and pressure limits.
CLINTON - UNIT 1 8 2-7 Amendment No. 18
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