ML19319D251
| ML19319D251 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/26/1977 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML19319D247 | List: |
| References | |
| NUDOCS 8003130879 | |
| Download: ML19319D251 (11) | |
Text
4
~
,7 r
i s
' ENCLOSURE 3 l.0.
Single Failure Evaluation _ of Com.r.on Instronat Sensino Lines for the
.g R115_0vergwer Irip thned upon Reactor Cool. int Flow and Axial Power -
Imbalance 1.1 Reactor Coolant Flow Instrumentation in each primary loop, reactor coolant flow is detected by measuring t..e aP developed across a flow tube that is an integral' part of the outlet piping of the loop. As illustrated in Figure 1, each flow tube has a high pressure (HP) tap and a low pressure (LP) tap. As illustrated in Figure 2, connections to the taps are made with 1-inch linas. The 1-inch lines =are terminated at root valves located inside the secondary shield wall.
Frcm the root valves,1/2-inch tubing runs through the secondary shield wall to HP and LP headers. Four (4) AP trans-mitters are connected between the-tno headers.
These transmitters provide flow information to:
the reactor protective system, the operator with flow indication and alarms at the control console, and the ICS.
f Each _of the four (4) reactor protective system channels receives
-a t.P signal from a diffsrent;one of the four AP transmitters.
In other words, one transmitter is exclusively assigned to one (1) protective channel. The identical arrangement and assignment of transmitters is used for each of the two (2) primary reactor coolant loops.
008130 l
<u
m n
Within _each reactor protective system channel, the square roots of the aP' signals' from each loop are extracted to obtain loop flow signals.
The loop flow signals are suated to obtain a total reactor coolant flow signal. The four (4) flow signals are displayed within the channel's cabinets and monitored by the plant computer.
The' reactor operator can read the individual loop flows and total flow at.the control console, within each reactor protective channel cabinet.- and from the plant computer.
1.2 Failures Considered The folicwing failures are considered:
(a) Break in one of the 1-inch instrument lines.
(b) Break in one of the 1/2-inch instrument lines.
(c) A leak in one of the instrument lines.
(d) plugging of instrument line between the flow annulus and flow transmitter.
l '. 2.1_
Break in 1-inch Instrument Line
- A' break of a 1-inch instru:::ent line will resul t in a reactor trip due to low RC pressure.
If the break occurs in a HP line, the reactor will trip due to a.high power / flow ratio if the power / flow limit 1is exceeded.
m.._
_._-..w
< s e
--+-- - -~-
- -+
t
7-
.~.
r' m
The operator will receive at least the following alarms and indications:
Al o rt:is (1) Break in 1" HP Instrument Line (a) Low RC loop flow (b) Letdown storage low level (c) Pressurizer low level (d ) Low reactor coolant prassure (e ) Plant computer alarm and printout for low RC pressure (2) Break in 1"LP Instrument Line Identical alar:as as listed for HP line break except RC flow is not alarmed on.high value.
Ind ica tion (1) Break in a 1" HP Instrument Line (a) LControl room indication of the reactor building atmosphere particulate and gas radioactivities-increases.
..... ~ - -
. ~...
...... ~
n p
-(b)) Loop l flow indication on console falls to zero.
(c) Loop flow indication.in each RPS channel falls E
'to zero.
?
(d) ~ Total flow indication on ' console falls to approximately 50%.
(e) ' To'tal-flow-indication in each RPS cnannel falls approxima tely 50%.
. (f) Makeup flow goes higher than normal.
~
(g) liC pressure falls on console indicators and within etich RPS cliannel.
'(h) Reactor; building ressure ana temperature indication -
p
~~-
l rises.
12.-
Break in a '1".LP Instrument Line i
Identical ~ indication ^as listed for HP line break except all
~
Lloop flow indication 'goes full _ scale, t).. flow indication increases'above normal.
}
f',
'u
+
r 3 4 -
..-u
m i
_.m
]
O
~
~
~
1.2.2
. Break in a 1/2-inch Instrument Line A break of a 1/2-inch-instrument line.will result in a reactor trip due to low RC pressure.
If the break occurs in a HP line, the reactor will trip due to.a high power / flow ratio if the power flow limit is exceeded.
-The operator will receive the same alarms and indications
' as. described for the 1-inch instrument line break.
1.2.3 Leak in One of the Instrument Lines If a leak occurs in either the HP or LP lires, the full RC pressure will rapidly increase the release cate to that of the break discussed above.
In addition, if in the unlikely event that the' leak -rate did not progress to a break condition, the reactor b'uilding radiation monitors will. readily detect activity
' levels resulting from leakage in excess of I gpm and result in
' leak evaluation,.and subsequent action as required by Technical
. Spec i fica tions.
In.the unlikely event the radiation monitors idid not detect this leak,- the op-?rator would observe a change in
'iridicdted flow 'when he performs a required flow indication
. comparison between expected and actual indication as required by Surveillance Precedure SP-300.
Depending 'on the-size of the leak, alarms and indication described in-Paragraph 1.2.1 may occur.
[15 4 e
l l'
~j
g7,
^'
n
, /;
r
- 1. 2. 4
' Plugging ofLInstrument Line SetUeen Flot Annulus and Flow-Transmitters
. The probability of any mechanism which: could completely block one lof these lines is at'least several orders of cugnitude less than that. cf a break or leak.
The _ reactor coolant system is a very cleanLsystem and is continuously filtered to assure that no significant particulate matter is circulated. The boric acid in the coolant is in concentrations about a factor or two belcw the; boric ' acid solubility limit at 700 F.
Therefore, no boron
- precipitation would occur within the reactor coolant system.
The
-entire fica monitoring system is essentially stagnant because it is a' pressure - sensitive device, and, as. such, there is no free exchange s
of-fluid between the reactor coolant syste., and the sensing lines oto induce material into these, lines to cause plugging.
If the assumption is cude that a sensin:; line became blocked, the tol'in.sinij indications, alarms, and survtillsace procedures uculd enable the operator to detect such a failure in the flow instru. antation.
.Indic.ition and Alarais The reactor coolan't' loop flow indication with a~ plugged sensing r
line are sensitive to reactor coolant systea pressure fluctuations.
The' pressurizer. heater control loop has a design control band which
~
- produces a re.ictor coolant sys ter.; pres sui a cycle. of about 35 psi ui th-af period of approkicately 10. tic.tes during s teady s tate -
c
. opera t i'on.
This pressure cyc1c.has.no effec: on the reactor coolant m
- flou indications 1when. imposed ~ehuall !u?:n.the high and'lo.. pressure sidesL of 'the. flow transr.itters.
Mc.teser,:ic the etent that'the highfor'iow pressure sensing line be:c e3 ;!u?;9d, the penstre
- l. -~\\
p.
4 s
.+ hrs 14-
- [
L u.
i
~
r
..S t
i
' cycle (35 psi) is imposed lon only one side of the'jg p' flow trans-mitters and the resulting flow indication will shift approximately 1.4 x 106 lbs/hr for.every lb.' of reactor coolant system pressure change.
This shif t in flow indication will provide the operator with. identical' flow related indication and alarms as that listed for the case of a line break (see Section 1.2.1).
Plant Surveillance.
(1) In accordance with Plant Technical Specifications & Surveillance Procedure SP-300, the Operating Daily Surveillance Log must be completed each shift and daily while the unit is operating in x
Modes 1, 2, 3 or 4.
As part of the surveillance required by the daily log, the operator must record the indicated reactor coolant system ficw and compare it against the required flow corresponding to the numbe. of P..C. pumps in operation and also ccmpare it againc'. readings take, at earlier shif ts to determine devi tinns.
(2) Folloding each calibration of the RPS Reactor Coolant Flow sensors during refueling, the response time of the reactor
-coolant flow protection channel must be verified. This-response
. time testing of the reactor coolant flow system would provide additional indication of-pluggage within sensing lines, as pluggage would alter the response time measured.
e<
-4, I
n-4
(
']3 1.3 conclusion The conclusion of this analysis is that the operator has adequate indica' tion and alarm facilities to quickly recognize a common mode failure in the flow instrumentation for the reactor protection system.
Corrective action would therefore be positive and prompt, in order to preclude reactor operation outside of acceptable limits described in the Plant Technical Specifications.
For these reasons, modifica-tions to the pressure sensing lines to the reactor coolant system flow differential pressure transmitters is not required at this time.
l
,M-
f
~.
/
-- w!_N.
M'.. s d
s
\\
[
J_
= _ _.
J I
,Q Q
N xm 1
W
.e
,9 sp.
s,.
jbk
[ ki\\
d'S N ..,..
' 9 -
.u l
e
.a
,,,,... er. 2 _..
- - 4.-
g s
si,'
wa g
\\i s' L
[ [ t',' 17 t' 1
D**]D T))W
@l 6 alu skis ]lhl A
~
es, g
..a-i
\\
i f.s,.e,
,:.,v.
p.,
\\,'
f<I
(
s.
l
'\\
I
/
r t
,:/,
I p
.c. ;-
l eo I 5
/
' bh.'.YE I
b
<t
/)
lI a'
('
/
.A>j
'{
,/
s
{
i
~
,,,e Ql:/
.,s i
r
./
/
t
'[
I.
//
' x, f,,,
/,I
.., ~ ~ ~ ~
l' s,
,.:. 't
'i !
.m c
/'
/
' -l iz) f.-
i y'
j
'i
[
l 7
/ G 2 '-
lh':3 #
.f f\\
, /.'
)
i, l'
f i
i.
3
'N.
I e,
l e
- f.Y,,
s, t
l t
l
\\
e i
s
)
f
'3fijf.([
/
f j
k l
\\
i' I
'g l
\\
i
\\
1
(
l s
s 1
I x
N\\
\\
j
/
l" 6,, *.
- -9
- . L, s
i i
i
/
/
,/
Fi ure 2.
W G
l%e 1 o f 2' 4n
\\'
s r
s N,
p
,s l
. *
- s t l S.
g s -
z.,.
==we s
-N Pl.EL. / / 9 ' 0" s
I?..
~x. N _~
- s...
~~'N_
N.N
<. 'v,
N p' L {
et
' ' '\\,'\\
y 'I !,D,1
,,Tx. 'iIh $ L.P-
'S-
!-s
- f.. lX.. 'N,l xx y
's v.3,N +r.
,,s, 1 : / ; p r,-
a f
f AA
_{ c.s
'. r rll P-L 't
'%ff' s-f&,
J s e y
s(s a
v ?,,
e %.
(v
/
i f.[Q'$*i-0s.
l
~.
/
M Figure ?
P-1 * ; ' 2 O f
c,,.,, { l, *. * ; 7,, e s e,-
.3 w.