ML20055A298
| ML20055A298 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 07/08/1982 |
| From: | Dixon O SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8207160103 | |
| Download: ML20055A298 (10) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _
e o
f south CAROUNA EtccTmc & GAS COMPANY m.,o,.c
.o. >..
Coi.uumia. south Camouna 2921 O O W. Demon, Ja.
vect Pne s. Den t Nvc e.r a n oer na t.ou.
July 8, 1982 Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Virgil C.
Summer Nuclear Station Docket No. 50/395 Reactor Coolant System Narrow Range RTD's
Dear Mr. Denton:
Because of problems encountered in cross calibration of the reactor coolant system temperature instrumentation during previous hot functional tests, South Carolina Electric and Gas (SCE&G) is changing the reactor coolant system narrow range RTD's from Rosemount Model 176KF to RdF Model 21204.
As a result of this change, minor revisions must be made to the Technical Specifications, the FSAR and Table 3-4 of the Reactor Protection System / Engineered Safety Features Actuation System Setpoint Methodology.
These revisions result from the RdF RTD's having a slightly slower response time along with a small increase in sensor drift in comparison to the Rosemount RTD's.
Since the RdF RTD's have a slower (2 second) time response in comparison to the Rosemount RTD's (0.5 second), the time constants used in the lag compensators (filters) for A T and Tavg are reset to zero.
This is reflected in Note #1 of Table 2.2-1 in Technical Specifications.
As a result of qualification testing of the RdF RTD's, a sensor drift of + 1.0*F has been included in the setpoint and margin analysis for the overtemperature A T, overpower A T and Tavg-Low-Low functions.
When combined with the other uncertainties using the Reactor Protection System / Engineered Safety Features Actuation System Setpoint Methodology, the margin is reduced by approximately 0.2% but has no impact on safety analysis assumptions.
Technical Specification Table 2.2-1 (Item 7 and 8 and p\\
8207160103 820700 PDR ADOCK 05000395 A
Mr. Harold R. Denton July 8, 1982 Page #2 Notes 2 and 4) and Table 3.3-4 (Itemse4d and 9b) are marked with the effects of this change.
A revision to Table 3-4 of the setpoint and margin study will be forwarded as it becomes available.
Also enclosed are marked up FSAR Tables 3.10-2 and 3.11-0 reflecting changes that will be included in the next FSAR amendment.
If you have any questions please contact us.
Very truly yours,
/
A o.
W.
- Dixon, r.
AW:OWD/fjc cc:
V.
C.
Summer G.
H.
Fishcher H.
N.
Cyrus T.
C.
Nichols, Jr.
O.
W.
Dixon, Jr.
M.
B.
Whitaker, JR.
J.
P.
O'Reilly H.
T.
Babb D.
A.
Nauman C.
L. Ligon (NSRC)
W.
A.
Williams, Jr.
R.
B.
Clary O.
S.
Bradham A.
R.
Koon M.
N.
Browne G.
J. Eraddick J.
L.'Skolds J.
B.
Knotts, Jr.
B.
A.
Bursey F.
Mangan NPCF File
-g TABLE 2.2-1 9
i
~
~.
REAC10R TRIP SYSTEM Ills 111UMEllTATION TRIP SETPOIllTS 3
Total Functional Unit Allowance QA),
Z S
Trip Setpoint Allowable Value 1.
Manual Reactor Trip tiot Applicabie NA flA fiA
,NA
$111.2% of RTP 2.
Power Range, Heutron Flux 7.5 4.56 0
$109% of RTP liigh Setpoint d
Low Setpoint 0.3 4.5G 0
$25% of RTP
$27.2% of RTP 3.
Power Range, Neutron Flux ~
1.6 0.5 0
<5% of RTP with
<G.3% of RTP with liigh Positive Rate 3 time constant a time constant 12 seconds 12 seconds 4.
Power Range, Neutron Flux 1.G 0.5 0
$5% of RTP with
$6.3% of RTP with liigh Negative Rate a time constant a time constant x,
j, 12 seconds
>2 seconds 5.
Intermediate Range, 17.0 0.4 0
$25% of RTP
$31% of RTP Neutron Flux G.
Source Range, Neutron Flux..
17.0 dil 10.0 0
$10s cps 31.4 x 10 cps 5
/.6 l
7.
Overtemperature AT 7.1
- 7. 94 Or6-See note 1 See note 2
- 1. 2 8.
Overpower AT 4.5 1.4 4)r2-See note 3
.See note 4 l
9.
Pressurizer Pressure-l.ow
' 3.1 0.71 1.5 11870 psig 11859 psig 10.
Pressurizer Pressure-liigh
- 3.1 0.71 1.5
$2300 psig 12391 psig 11.
Pressurizer Water Level-liigh 5.0 2.10 1.5 192% of instrument
$93.0% of instrumeni, if,.
span span 12.
Loss of Flow
- 7. 5 1.0 1.5 190% of loop 209.2% of loop design flow" design flowa Loop design flow = 90,000 gpm
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TABL '2.2-1 (Continued) i i
REACTOR TRIP SYSTEM lilSTRUMEtiTATI0ft TRIP SETPOIIITS J'
fl0TATI0tl (Continued)
~
HOTE 1:
(Continued)
T' 507.4"F Reference T at RATED TilERMAL POWER avg si h
.0006720 K
=
3 C
l'rcssuriter pressure, psig P
=
2235 psig, Nominal RCS operating pressure.
P'
=
Laplace transform operator, sec 1
~S
=
and f (AI) is a function of the indiceited dif ference between top and bottom detectors of the t
T power-range nuclear lon chambers; with gains to be selected based on measured instrument response during plant startup t.csts such that:
(i) for qt 4
between - 34 percent'and + 0 percent f (AI) = 0 where qt and q are percent t
,b
~
h is RATED TilERMAL POWER in the top and bottom halves of the core respectively, and q( + qb total TilERMAL POWER in percent oh RATED TilERMAL POWER.
(ii) for each percent that the magnittide of qt g
exceeds -34 percent, the AT trip setpoint b
shall be automatically reduced bf 1.67 percent of its value at RATED TilERMAL PdWER.
(iii) for cach percent that the magnitude of q q
exceeds +8 percent, the AT trip setpoint b
shall be automatically reduced by 1.11 percent of its value at RATED TilERHAL POWER.
HOTE 2:
The channel's maximum trip setpoint shall not exceed its computed tr,ip point by more than j
' M. percent AT span.
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g TABLE 3.3-4 (Continued) 5g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Total E
Functional Unit Allowance (TA)
Z S
Trip Setpoint Ailowable Value w
4.
STEAM LINE ISOLATION a.
Manual NA NA NA NA NA b.
Automatic Actuation, Logic NA NA NA NA NA and Actuation Relays c.
Reactor Building Pressure-3.0 0.71
- 1. 5 16.35 16.61 High 2 d.
Steam Flow in Two Steamlines-20.0 13.16 1.5/
< a function
< a function defined High, Coincident with
- 1. 5 IIefined as as follows:
A op follows:
A AP corresponding to 44%
{
corresponding of full steam flow to 40% of full between 0% and 20%
w steam flow load and then a op O
between 0% and increasing linearly 20% load and to a op corre-then a Ap sponding to J14.0%
increasing of' full steam linearly to a flow at full load.
Ap correspond-ing to 110% of
~
full steam flow at full load I,2 553.6 Tavg - Low-Low 4.0 1.-12 0 2553 F 155 M F e.
Steamline Pressure - Low 20.0 10.71
- 1. 5 1675 psig 1635 psig(1) mM (1)
Time constants utilized in lead lag controller for steamline pressure low are as follows:
T1 = 50 secs.
T2 = 5 secs.
e N
.m
?
]
- /',
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY. FEATURE ACTUATIO!I SYSTEM lil5TitVMENTATI0tl TRIP SETPOINTS 9
Total c
Z S
Trip Setpoint Allowable Value
.} functional Unit Allowance (TA),
t" 9.
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM IllTEllLOCKS i$
INTERLOCKS
)
i PressuriterPressurb,P-11 3.1
.71 1.5 1985 psig 11974 psig &
a.
<1996 psig
- '2
~rro.s SSC.+
4.0 1.12
-Ore-553*F 1650-5'F & <555.5%
Tavg Low-Low, P.12 -
b.
c.
Reactor Trip, P-4 NA NA tiA NA NA l
t n
te.
PJ a
U D
O
'V N.I -
me
' a,
~
TABLE 3.10-2 I
IDENTIFICATION OF NUCLEAR STEAM SUPPLY SYSTEM SEISMIC CATECORY I INSTRUMENTATION, ELECTRICAL EQUIPMENT AND SUPPORTS
(
Item Method 6
1.
Pressur,e Transmitters ** and Differen-Bi-axial, multifrequency*
14 s
Pressure Transmitters **
2.
Process Control Equipment Cabinets **
Single axis sine beat, bi-axial multifrequency 3.
NSSS Solid State Protection System Single axis sine beat Cabinets 4.
Nuclear Instrumentation System Single axis sine beat, bi-axial 6
Cabinets multifrequency 5.
Safeguards Test Racks Single axis sine beat bisay Mults*meu scagle fesyvwey 33 6.
Resistance Temperature Detectors **
g*ngk ani: : c _ ;d; 1 8
l 7.
Instrument Supply Inverters **
Single axis sine beat, bi-axial
(
sine beat 8.
Reactor Trip Switchgear Multi-axis, multifrequency*
14 t
9.
Power Range Neutron Detectors Single axis sinusoidial
- 10. Post Accident Monitoring Equipment Multi-axis, multifrequency*
l g
(Indicaiors** and Recordere) 8 i
i
- 11. Post Accident Electric Hydrogen Single axis sine beat for Recombiners recombiners, bi-axial sine but for control panel
- Not yet completed.
- Required for safe shutdown (assuming norr.al operation and not post accident 14 conditions).
AMENDMENT M U 3.10-15 gern, 4944 Tu ty,111;
I TABLE 3.11-0 i
NUCLEAR STEAM SUPPLY SYSTEM CLASS 1E EQUIPMENT IN CONTAliMENT Safety function Equipment Tag or II 4
Descrip t f ort Manufacturer Category FSAR Chapter Location Time g Location Number Instrumentation RdF 33 e
Narrow Range Resistance "x.o..
a2, c1 Section '. 5, y RBc, by pass W Response TE-412 A.B.C.D l Jf" Temperature Detectors Response tt 031.47 loop to 031.47 TE-422 A.B.C.D i
TE-432 A,B,C,0
=.
=...m R===-
-- m
,. = u =m m...,,... =,
isu l
L.y:
t "-- n-ter *:n ce R e pc m - te 031.'?
O 0". '7 423, 430, G Reactor Coolant Flow Barton cl, c2 Section 7.2 RBc (2)
FT-414, 415, 416 l 27 DP Transmitters RBa FT-424, 425, 426 l 27 RBb FT-434, 435, 436
! 27 Pressurizer Pressure Barton al, a2 Section 7.2, W RBd W Response PTOISS, 456, 457 l 27 P Transmitterr Response to 031.47 to 011.47 III Section 7.2, 7.3, 7.5 RBd W Response LT-459, 460, 461 l 27 Pressurizer Level Barton al III DP Transmitters a2 W Response to 031.47 to 031.47 III Steam Generator Level Barton al Section 7.2, 7.3, 7.5 RBe W Rt.oonse LT-474, 475, 476 l 27 bI Narrow Range DP a2 W Response to 031.47 to 031.47 LT-484, 485, 486 Transmitters LT-494, 495, 496
!!h III
[] h Reactor Coolant System Barton al Section 7.5 W RBc W Response PT-402, 403 l 27 III
'k Pressure Wide Range a2 Response to 031.47 to 031.47
-Y f 4w P Transmitter 4w tw 18
.