ML20117K199
| ML20117K199 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/04/1996 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20117K186 | List: |
| References | |
| NUDOCS 9606100104 | |
| Download: ML20117K199 (8) | |
Text
.
s l5.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
15.2.1 SAFETY LIMIT, REACTOR CORE Applicability:
i Applies to the limiting combinations of thermal power, reactor coolant system j
pressure, and coolant temperature during operation.
j j
l Objective:
To maintain the integrity of the fuel cladding.
Specification:
j 1.
The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 for Unit]1andFigure15.2.12forUr.it2j. The safety limit is exceeded if the point defined by the combination of reactor coolant system average temperature and power level is at any time above the appropriate pressure line.
Basis:
The restrictions of this safety limit prevent overheating of the fuel and pos-sible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fu'el cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excess cladding temperature because of the onset of departure from nucleate boiling (DNS) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore thermal power and Reactor Coolant temperature and pressure have been related to DNB.
$ 5 S ?kUIN S $ 5 W N $$5 N b N W b 5b?ENYN W NYNSCIS M8ZEEEVJIMDEiMEGGk1EEEE2 Unit 1 - Amendment No. 442 15.2.1-1 October 27, 1993 Unit 2 - Amendment No. 146 9606100104 960604 PDR ADOCK 05000266 P
l 4
FfgirTilBifl?!
P6fMT7tdE4tKfNUCLEARiPLANTiljN!Tsl17AND]
0 EN 670 l
660 l
650
[
2425 psia 640 m
g 2250 psia
_N M 630 o
"U v
M
$.620 2000 psia p
T 3 610 5
600 1775 psia 590 580 570 0
0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
1.1 1.2 Core Power (fraction of 1518.5 MWt) fEDfiiEHiEfiraMfeEMUREMf311TsiguzititiraTew jpg]istgirigtgg Urigg p32SiPyj orjlo201324jgjigurX15L2]E21appi j esitoitin 5
Unit 1 - Amendment No.
i l
Figure 15.2.1-( ~;-
REACTOR CORE SAFETY ~(LIMITS 1
POINT BEACH UNIT I 1
660<
650<
2400 PSIA 640<
2250 PSIA 650<
1 2000 PSIA t0 o620<
>e H
1775 PSIA GIO<
600-590<
l 580-9.
.1
.2
.5 4
.5
.6
.7
.8 9
1.
1.1 1.2 POWER treection or nominell f
LFg5T@lg_u. reg {gureTafi[ra.@_pp_li ent6) Un@i tfl_g@gijf((((J{Qy
[ jR 15!_2_31si_V j iiT D g
e Unit 2 - Amendment No. M9 Nov:dcr 17, 1005
Figure 15.2.1-2 REACTOR CORE SAFETY LIMITS POINT BEACH UNIT 2 670 660 -
650 2425 psia 640 2250 psia 630 C
3 620 f
2000 psia e
[ 610
'M h
l 600 1775 psia l
i 590 580 l
570 l
560 0
0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
1.1 1.2 Core Power (fraction of 1518.5 MWt)
Unit 2 - Amendment No. M9 k;; bcr 17, 1995 l
r
r l
(3)
Lcw pressurizer pressure -
21865 psig for operation at 2250 psia primary systea pressure 21790 psig for operation at 2000 psia primary system pressure (4)
Overtemperature 1
AT ( 1 +r,S )
sat, (K -K,(T( 1 +r,S )-T') ( 1 +r*S1+r,S, K,(P-P')-f(AI))
1
"+
3 where M1}Leslay]MllFiSTEffsgeraQ,oglatW6fjij[glef[ifalaYl@,5y j
fifi gn g s3 Er 7WEOid{Eifid}
indicated AT at rated power, F ATo
=
average temperature, *F T
T',
s 573{.9 F[ (Ur.it 1) e.,. n. n.o..r m.
a..
.n.,
r.
pressurizer pressure, psig P
2235 psig [2'j$]{s]E@@jgQ@yj P'
$ ? W 9 [1y8]!]jyli[G000IpljH 6psfB165[6d]y[]
K 4,40 ll2fi(2iS0lfs[QMi il@Klollyf t
3
$didMMdg@R$p([pl{Ifjp@{Qfnly]y{$
0.0200 y{pjjl[2210]pl(([ifefjQjnlo]ly!
K,
=
FNDIORHsMffilEHEf1EEEMill 0.000791 yJ00J3Z(2250ZpstE6fifiK5JF Til K,
=
Ef6+#sRE01ElfD00Tiill!~oEEEsEfisI6illiH 25 sec r3 3 sec j
r, 2 sec for Rosemont or equivalent RTD r,
O sec for Sostman or equivalent RTD 2 sec for Rosemont or equivalent RTD j
r,
=
O sec for Sostman or equivalent RTD and f(AI) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where q, and q, are the percent power in the top and bottom halves of the core respectively, and q, + q, is total core power in percent of rated power, such that:
(a) for q, - q, within -17, +5 percent, f(AI) - 0.
(b) for each percent that the magnitude of q, - q, exceeds +5 percent, the AT trip setpoint shall be automatically reduced by I
an equivalent of 2.0 percent of rated power fer Ur.it 1, er by l
l an equivalent of 3.1 percent of rated power for Unit 2 FTThise yalues apply to U6W'2"f61lisiini~02R22Td'~iT'UEiUFf6113~KE'g'01R24]
~
^
'^"~~^ Prior to'U1R24, the' values for Unit I are:' T' s 573'.9*F,^^P- 2235 p~sigf ^ '
K 's 1.30,';Kj1 02,00' ~and12 0.000791.'['~ ^~ ~
~ ~ ' ~ ' ~
0 t
Unit 1 - Amendment No, M6 15.2.3-2 Octcber 23, 100?
Unit 2 - Amendment No. MO
(c) for each percent that the magnitude of q, - q, exceeds -17 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power.
l (5)
Overpower AT ( 1 +r,S )
r,S 1
1
$4T,[K.-K,( r,S+1 ) ( 1 +r.S ) T -K. (T( 1 +r.S ) - T'))
I i
whore @NFifi~ppfinSMitoYoperatWn7sE65EinfR03{iGikTd}ff$iffgfg indicated AT at rated power,' F AT, average temperature, *F T
T' s
573(.9 F] (Ur.it 1)
T' 570. 0 F
'%' t 2' s
K.
s 4r489 {@ of rated power $
0.0262 for increasing T K,
i 0.0 for decreasing T
=
0.00123 for T :a: T' K.
0.0 for T < T'
=
10 sec r,
2 sec for Rosemont or equivalent RTD r,
0 sec for Sostman or equivalent RTD 2 sec for Rosemont or equivalent RTD l
r.
0 sec for Sostman or equivalent RTD (6)
Undervoltage - 275 percent of norm'al voltage (7)
Indicated reactor coolant flow per loop -
l 290 percent of normal indicated loop flow l
(8)
Reactor coolant pump motor breaker open (a)
Low frequency set point 255.0 HZ (b)
Low voltage set point 275 percent of normal voltage.
UM2fMdIif6I 6TIWiifyIMin2M Ell}2f311"ofwyst rlugjg1 gggTggi7;p{lii@ry_ligentoss Iibinia EFSThWe]gJ.
g i
Unit 1 - Amendment No. 443 15.2.3-3 Cet:bcr 27, 1903 l
Unit 2 - Amendment No. 446 i
With normal axial power distribution, the reactor trip limit, with allowance for errors"), is always below the core safety limit as shown on Figurej 15.2.1-1 kr Unit--I and Figur: 15.2.1-2 fer Unit 2.
If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear 1
detectors, the reactor trip limit is automatically reduced")").
The overpower, overtemperature and pressurizer pressure system setpoints include the effect of reduced system pressure operation (including the effects of fuel densification). The setpoints will not exceed the core safety limits as shown in FigureY 15.2.1-1 for Unit I and Figer: 15.2.1-2 fer Unit 2.
l The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur.
The reactor is prevented from reaching the overpower limit condition by action of the nuclear overpower and overpower AT trips.
The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such that the reactor is tripped before the safety valves actuate. The low pressurizer pres-sure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident").
The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or both reactor coolant pumps. The setpoint specified is consistent with the value used in the accident analysis"). The low loop flow signal is caused by a condi-tion of less than 90 percent flow as measured by the loop flow instrumentation.
The loss of power signal is caused by the reactor coolant pump breaker opening Unit 1 - Amendment No. MB 15.2.3-6 Octob:r 27, 1993 Unit 2 - Amendment No. 446
G.
OPERATIONAL LIMITATIONS
\\
The following DNB related parameters shall be maintained within the limits i
shown during Rated Power operation:
1 l
l 1.
T,., shall be maintained bel:w 578% @5fTIagig5RitfQ 2.
Reactor Coolant System (RCS) pressurizer pressure shall be maintained:
Unit 1:
2:2205 psig during operation at 2250 psia, or a1955 psig during operation at 2000 psia.
,.. t i.t...i..4 mi. nee
.....i..
e.nn.n.
u.
y...
u.
a
<e.
y<.y
..v sy
.y ar 3.
Reactor Coolant System raw measured Total Flow Rate 'rS S::i:' Mi~alf i_
P m b.ww.w.-s. wen h w e kn-ww.FmaTEmneTiElW8061Eed.
i t. 4..
1 01..O nn.
f.f =. 4..
1 1 s
,.n_
3 y...
i u,.
I.t 4.
- 17. J, n.n. a, - _ t.t _ J.
4 9e I
iss.
e.,
s T
. y y s..
..s.
Basis:
The reactor coolant system total flow rate for Unit 1 of 181,800 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm). The re::ter :: lant Oy:te: tet:1 flew r:t: for Unit 2 :t r:ted p:wcr i: 174,000 gp=.
Thi: 1: b:,cd en :: :::umed ::::ure: nt unecrt:inty f 2.1 percent ver : the ::1 design flew of 170,400 gpm.
However, Unit 2 i: :n:ly:cd t: :upport operation with a re::ter :::lant y:t:: tot:1 flew rat: limit Of
- i. e n, e n n. _, _._...
- v..u.., i.. u.-...a
.........__ _.-.. ___..____......,.,, 4 r
... ~....................................
y Over : the ::1 de:ign flew of 155,000 gpm.
If the Unit 2 "CS r:w ::::ured tot:1 flew r:t: i: les: th:.n 174,000 gp: but gre:ter th:n Or equal to 150,500 gp=,
Operatier, i: limited te le:: th:n er equal t: CS" r:ted power :: de:cribed in the not: to Speciffection 15.2.1.0.2.b.
The raw measured flow is based upon the use of normalized elbow tap differential pressure which is calibrated against a precision flow calorimetric at the beginning of each cycle.
_MPw5Tfiss- _ithTE_ip_siyw_Ef_tM_f51_TWTRF.U_2k22iisaT_E_Us_itIIIf6TTWEFUlms W
IWU
~
m I
hh l
l For Unit 2:
If the " ::ter C clant Sy:te r:w ::::ured total flew rat i:
i.i. e..... e u......,, u, 4. _. 4..,a...--n..o.,.
, i s,,,, n.n.n ___ u.... mien..e.nn
.. m
....r....
..m.
... 3 r...,
,y...
Unit 1 - Amendment No. MS 15.3.1-19 November 17, 1005 Unit 2 - Amendment No. M9