ML19345F587
| ML19345F587 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 02/10/1981 |
| From: | Delgeorge L COMMONWEALTH EDISON CO. |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| LOD-81-40-10, NUDOCS 8102180467 | |
| Download: ML19345F587 (28) | |
Text
'
Commonwealth Edison One First Nabonal Ptata. CNcago. Ilknois Address Reply to: Post Office Box 767 Chicago, lilinois 60690 February 10, 1981 Mr. B.J. Youngblood, Chief Licensing Branch 1 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
LaSalle County Station Units 1 r, 2, Instrumentation and Control Systems Resolution of SER Open issues, NRC Docket Nos. 50-373/374 LOD 81-40-10
Dear Mr. Youngblood:
Attached for your review are supplemental materials submitted in response to NRC Staff questions related to the following Instrumentation and Control Systems issues:
1.
Test Techniques - see Enclosure 1 2.
Setpoints - see Enclosure 2 3
Separation Criteria - see Enclosure 3 4.
Testing of low-Low Setpoint Logic - see Enclosure'4 5
RG 1.97 and Safe Shutdown Indication - see Enclosure 5 6.
RBM System - see Enclosure 6 7.
Use of Non-Safety Grade Equipment Each of these issues is discussed in the enclosure to this letter as referenced above.
It is judged that these supplemental traterials provide the technical basis for resolving these issues. This judgement is based on our review of the subject areas and detailed discussions conducted with the NRC Staff iuring the week of February 2,1981.
Item 7, although discussed at sor s igth, is not addressed in the enclosures inasmuch as the Staff has nc decided whether additional information in addition to technical spec
'lons already committed is required.
In the event yt o ve any further questions on these matters, please direct them to this office.
Very truly yours, L.O. DelGeorge
~
Nuclear Licnesing Administrator Enclosures 1-6 cc:
NRC Resident Inspector-LSCS I
~
81 oniso 9m
/
Mr. B.J. Youngblood, Chief February 10, 1981 ENCLOSURE 1 TEST TECHNIQUES in order to address concerns raised by the NRC Staff relative to the technique used in the performance of periodic instrumentation survelliances on LaSalle County Unit 1, a detailed delintation of all applicable test procedures was provided in response to Q 031.284.
In addition, the testing frequencies associated with these tests as defined in the LaSalle County Technical Specifications was reviewed with the Staff.
At the conclusion of that review, the Staff requested a clarification of the input to q 031.284 Section E, " Logic System Functional Tests Covered By Technical Specifications".
This question was reviewed in greater depth by Commonwealth Edison and the following information provided to the Staff:
- 1. Logic System Functional Tests, as described in Section 4.3-of the Technical Specifications are performed on an 18 month interval during refueling outage.
- 2. Item Q 031.284 E.1, "RPS Instrumentation" does not require circuit disturbance of the types discussed in Q 031.284 to accomplish the required periodic surveillance.
- 3. Item Q 031.284 E.2, " Isolation Instrumentation" was reassessed and it was determined that fuse pulling was required only for item 1.b (Hi Drywell Pressure) in Table 4.3.2-1 and that an annunciator alarm which is received on removal of the fuse which must be cleared as part of the return of the system to normal.
- 4. Item Q 031.284 E.3, "ECCS Actuation Instrumentation" and E.4, "RCIC Actuation Instrumentation" were reassessed and it was determined that fuse pulling was required for items a.1.f, a.1.1, a.2.c, a.2.g, b.1.d and b.1.g.
These tests require check of time delay relays and manual push buttons, An annunciator alarm is received on removal of the fuse which must be cleared as part of the return of the system to normal.
This information was discussed with Mr. D. Thatcher of ICSB on February 6,1981 and is reported here for documentation. Question Q 031.284 will be modified to document this information in a future l
amendment to the FSAR
ENCLOSURE 2 SETPOINTS At the review meeting between Commonwealth Edison and the NRC staff on December 17, 1981, the applicant was requested to provide an amendment to Tables 7.1-2, 7.3-1, 7.3-2 and 7.6-1 of the FSAR which would identify the device range for instruments delineated therein.
These tables have been reviewed, and changes made as requested.
A copy of the prosed amendment to those tables is provided as an attachment to this enclosure.
Formal documentation of these amended tables will be made in a future amendment to the FSAR.
P
_ ~
~ --
,N v
h h it L. E 75f:C& (91f eck TABLE 7.3-1 ECCS ACTUATION INSTRUMENTATION SETPOINTS [d ANALYTIC OR DESIGN BASIS DESIG! -BASIS bh FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE LIMIT ACCURACY CALIBRATION ALLOUASLE gggfje _
!o+ tis,v (1) Reactor Water Level
-50'in.
-57 in.
-70 in.
G.0 in.
2.0 in.
7.0 in.
- Low, Level #2 (2). Drywell Pressure - High 1.69 psig 1.89 psig 2.0 psig 0.04 psi 0.04 psi 0.2 psi o.2.- 4 %
(3) Condensato Storage Tank 7.6[n7"
[dhM.7
{l%[f,,,q DD 0.5 in.
0.5 in.
1.0 in.
Lcvol - Low n
va
'! -/8 (4). Suppression Pool Water
'700 fc. / in.
700 ft. 2.in.
DD 0.5 in.
0.5 in.
1.0 in.
T lli
- 8N Lovel - 11igh
- J ADS (5). Drywell Pressure, !!igh 1.69.psig 1.89 psig 2.0 psig 0.04 psi 0.04 psi 0.2 psi O.2-6 [.
(G)
Reactor Water Lovel
-129 in.
-13G in.
-149 in.
G.0 in.
2.0 in.
7.0 in.
.m g,.,
- Low, Level #1
.(7)
ADS Timer.
105 seconds 117 seconds 120 seconds 2.0 sec.
2.0 sec.
12.0 sec g g f.g, x a
au
- 8) ;LPCS Pump Discharge 14G psig 135 psig 125 psig 5.0 psi 2.0 psi 10.0 psi
{e g.f o.J.fo u
Pressure
. Permissive-gy U
l w55 W'
~$%
(eePicure M. Ta for-e k..-,,,
',, + c,rt, t) 'Se.e.hcGd S c'/j<e A,v a & ya/ceLal "Ph
~
ce=3 xp-b.
,A.
, n, i
4 t
J TABLE.7.3-1 (Cont'd)-
(LS ANALYTIC OR DESIGN DASIS DESIGN-BASIS I MC"'
FUSCTIONAL UNIT TRIP SETPOINT ALLOWADLE VALUE LIMIT ACCURACY CALIBRATION ALLO"ABLE
/24N C
-(9)
RIIR - (LPCI Mode) Pump
,,,,].?A" psig 106 psig 100 psig 2.0 psi 2.0 psi 8.0 psi l /O-2fo Discharge Pressure -
jfy Permissive LPCS 4
(10)
Reactor Vessel Water:
-129 in.
-136 in.
in.
G.0 in.
2.0 in.
7.0 in.
l~/f0 j fp l./
Lovel - Low, Levol il a
.( 1 11. Injection valve AP - High 729 pcid 709 psid 700 psid 3.0 psi 3.0 psi 20.0 psi po.8co g n
(12). Drywell Pressure - 1Iigh 1.69 psig 1.89 psig 2.0 psig 0.04 psi 0.04 psi 0.2 psi T o.2.-C. h
[
~
Y Rl!R (LPCI Mode) g (13)
Drywell Pressuro - liigh 1.69 psig.
1.89 psig 2.0 psig 0.04 pei 0.04 pci 0.2 psi c.2.- 4 M (14)
Reactor vessel *1ater Level
.-129 in.
-136 in.
-149 in.
G.0 in.
2.0 in.
7.0 in.
l
-/5'D[o[#o4,
- Low, Lovel $1.
(15)- Injection valve AP - High 729 psid 709 paid 700 psid 3.'O psi 3.0 psi 20,0 psi 0 -IJO M o p*
M 2d LD C3 xg M
U '3 i
=yeA O
1 m
A E
I 1
a 2:53 b
r%
TABLE 7.3-2 PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION ACTUATION IMSTRUMENT SETPOINTS ANALYTIC OR b
DESIGN-DASIS DESIGN-BASIS FUNCTIONAL UNIT TRIP SETPOINT ALLONABLE VALVE LIMIT ACCURACY CALIDRATION ALLO!.'IsNCE M4N(t' Reactor Core Isolation Cooling L9'S
-h/3 3,o 4 (1)
RCI C Steamlino Plow 290%
M 300%
1 in.
1 in.
6 in.
- High L1-? 5'in. )
M &7"in.)
(191 in.)
t78 t 85' (2)
RCIC Steam Supply Pressurc 57 psig 53 psig DB 1.0 psi 1.0 psi 2.0 psi l f-/#C h$
- Low 8
(3)
RCIC Pipo Routing Arca
- 280
- 2 # 2 *b 2.0* F 1.0' F G.0* F g_3fo F a
Temperature - Ifigh r.7
- /0Y
./360f ho-/roF 3.0' F 0.5' F 3.0* r (4)
RCIC Pipo Routing Arca A Temperature - Iligh
Diaphragm Prcscuro - High 2.0* P 1.0* F G.0* F
$b - 370 [
(6)
P"'
2quipment Room
- M 200
- /d s 8-2 W' *I
.:+.rature - Ilich n
r -
)[
~
~
D ou Shutdown Cooling Isolation g*Ej Nm
~gC f0,o gg)
(8)
Reactor Vessel Water 12.5 in.
11.0 in.
7.5 in.
1.5 in.
0.5 in.
1.5 in.
8 Lovel - Lou, Level #3 off M
curing preoperational testing, the trip setpoints will *>c cet at 40' P above space ambient temperature, o
g Final setpoints will be established based upon operatio tal data af ter calibration of detectors and modules.
N
?
g EED
<m TABLE 7.3-2 (Cont'd)
ANALYTIC OR DESIGN-3 ASIS DESIGN-BASIS d cyt c c FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALVE LII-IIT ACCURACY CALIBRATION ALLO'.iANC E ggg e (9)
Reactor steam-De-e
/Oo/A 135 psig 145 psig 1G1 psig 5.0 psi 2.0 psi 10 psi
/d-210 pf Pressuro - High
/pof) S f37pSsy fff
/s i fm fo 2.0 h fo g o - f#6 Reactor Water Cleanup System (10) a Flow - High 70 gpm 87.5 gpm 104.5 gpm 11.5 gpm 13 gym 17.5 gpm 0- / 00 fD (11) Arca Temperature - !!igh
- /ST*F
- l'l/ */'
2.0* F 1.0* F G.0* F T0- 370 'F (12) Arca Ventilation Temp.
5 AT-I!igh
- 77*/
- /0/ V 3.0* F 0.5' F 3.0* F p o-/9 *F f
(13)
Reactor vessel Water Lovel
-50 in.
-57 in.
-70 in.
6.0 in.
2.0 in.
7.0 in.
2/
go
- Low, Level #2
?j ja Rc';idual !! cat Removal (14)
RI!R Flow - High 123 in.
128 in.
DD 6 in.
3 in.
5 in.
l O-3fD,*v (15)
Rt!R Equipment Arca
- SO 'F
/wF e
2' F 1r G' r so.3ro r Tempera ture - !!igh g 'lG)
R!!R Equipment Arca
- b
- N 3' F 0.5* F 3' F 0-/70 [
g3 a Temperature - Iligh oa L19 -
,3 g m
M 17)
Reactor Vessel Water Level 12.5 in.
11.0 in.
7.5 in.
1.5 in.
0.5 in.
1.5 in.
~;i 0- b O '^#
- Low, Level #3 ga Q
?
o Reactor Vesuol Water Level
-50 in.
-57 in.
-70 in.
G.0 in.
2.0 in.
7.0 in.
M ]l8)
--/ 58 o @
- Low, Level v2
,.y Sc y
W W
b
m.
j%
w TABLE 7.3-2 (Cont'd)
ANALYTIC OR DESICN-DASIS DESIGN-DASIS fMcC" TUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALVE LIMIT ACCURACY CALIBRATION ALLO *.fANCE A 4 W e
(19) Drywell Pressure - Ifigh 1.G9 psig 1.89 poig 2.0 psig 0.04 psi 0.04 psi 0.2 psi o.2. d[
(20) ' Main'Steamline Radiation 3.0 x full 3.6 x full' DD 202 21 10%
ll-/O
- Irigh power power background background mg (21) Main Steamline Pressure 854 psig 834 psig 825 psig 3.0 poi 3.0 psi 20.0 poi 20,- nnpd
- Low
~(22)' '!!ain Steamlino Flow - Iligh 111 psid 116 psid 123 psid 0.5 psid 0.5 psid 3.0 poi
- M
.a L.
(23). Temperature -'!!igh f3d.
/.,'
Main Steamlino Tunnel
- / p #[
- /k, *f 2* r l' F G* F p 3ro*F
. (24). ' !!ain Steamlino
- 20
- M 3* P 0.5* F 3* P O /J3 *8 A Temperature - !!igh
- 2 f fs' //y
- 23 /d h (25). Condensor vacuum - Low 0.3 in.
0.3 in.
0.6 in.
o,7-27.7.
I!g
!!g lig
/**
=(2G)
Reactor nullding Exhaunt mq.
y1 A L.
O*6/-/6h/h
- 107, 2t 121 3
Rad - !!igh d
M
.(27)
Drywoll Precouro - !!igh 1.69 psig 1.09 psig 2.0 psig 0.04 pai 0.04 psi 0.2 psi
- d. W rJgp c
(28) P.cactor Vecsol Water Lovel.
12.5 in.
11.0 in.
7.5 in.
1.5 in.
0.5 in.
1.5 in. 0 - SO Y. =
- CQ
- Low, Levol 43 kcc=: 3
$====
-=
i%
c 5
, m
yi~o, TABLE 7.3-4 W
REACTOR PROTECTION SYSTEM INSTRUMENT SETPOINTS ANALYTIC OR
'DESIGM-3 ASIS DESIGN-BASIS M.*
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUC LIMIT ACCURACY CALIBRATION ALLONAMCE P
g g-(1) Intermediate Range Monitor 120 Divisions 122 Divisions DD 0.1%
0.5%
2.01 2 ?. f.
Neutron. Flux - Upscalc 64 y, 4 t
(2) Average Power Range Monitor 15%
201 25%
0.1%
0.5%
2.0%
Neutron Flux - Upecalc fy' (Not Run Mode)
(3)
Averago Power Range Monitor, (.GGw + 51%)
(.66w + 545) 117%
0.1%
0.5%
2.0%
(
Simulated Thermal. Power (113.5% max.)
(115.5% max.)
m 7
- Upscale 4
(4) AveragePowerRanbcMonitor, 118%
1201 121%
0.1%
0.5%
2.0%
h
- Upecale (Run Mo c)
(5) ' Reactor Vessel h Domo
.1043 psig 10G3 psig 1071 psig 3.0 psi
-3.0 psi 20.0 psi 2co-st**
Pressure - iligh P
'(65 Reactor Vessel Water Level 12.5 in, 11.0 in.
7.5 in.
1.5 in.
0.5 in.
1.5 in.
o /.,o,o
--Low, Level #3*
(7) Main Steamlino Isolation
.94% open 93% open 90% open 1.01 1.01 1.01 Valve - Closurc gn (8)
/ X/*&
To i
- M
. Main Steamline Radiation 3.0 x full 3.6 x full DD 20t 2t 10%
ff y
- Ifigh power power Mh y y
background background e
- p
$[
- Refers to instrument zero whcihe is 527.5 inches abovo vessel r.cro.
25E3 s
_ - _ _ _ _ _ _ _ _. - ~ - - - - - - - - - - - - - - - - - - - -
~
.-.s A...
4 TABLE 7.3-4 (Cont'd)
ANALYTIC OR DESIGN-EASIS DESIGN-BASIS M FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE LIMIT ACCURACY CALIDPATION ALLONANCE _ (qu (9) Drywell Preccurc - Itigh 1.69 psig 1.89 psig 2.0 psig 0.04 psi 0.04 psi 0.2 psi h.2- @
(10). ' Scram Discharge Volumo Water *
- 0. 5 in.
0.5 in.
1.0 in.
Lovel - !!igh (11) Turbine Stop Valve 95% open 93% open 90% open 1.01 1.0%
2.0%
- Closurc g
(12) Turbino Control valvo 500 psig 414 psig 400 psig 5.0 psi 5.0 psi 30.0 psi YO
- Fast Closure, y
Trip Oil Pressure - Low 5
'U Q
h 5
A 8%
@ Note:
dem All reactor water levcis are referenced to instrument,zero at 527.5 inches above the Vercel acmes M I5
@ [3
- t ::
Q To be determined from preoperaticnal testing.
"E U"
g=
81%
c===
w p==
i=
a N.
j~ y
?%
t TABLE 7.3-5 Y
CONTROL ROD BLOCK INSTRUMENTATION SCTPOINTS ANALYTIC OR DESIGN kfjfc e!"
DESIGN BASIS DRIFT FUNCTIONAL UNIT TRIP SETPOINT ALLOWADLE VALUE LIMIT ACCURACY CALIDRATION ALLOMANCE bWI Average Power Range Monitor (1). Neutron Flur. - Upscale
-(0.66w +.421)
(0.G6w + 451)
D3 1.0%
1.04 3.0%
(ficw reforonecd)
(2) Noutron Flux - Downscalc.
51 31 D3 0.1%
0.5%
2.0%
(3) Noutron' Flux Upscalo 121 14%
D3 0.1%
0.5%
2.0%
n
.(Not Run Modo)
Rod Block Monitor (4) Upscale (0.66w + 40%)
(0.6Gw + 4 3%)
111'.8%
1.01 1.01 3.0%
gp k'
-(5)
Downscalc St 31 DD 0.1%
0.5%
2.0%
to f$
' Source Rango Monitors
[.
5 5
0' (G) - Upscale 2 x 10 cps 5 x 10 DB 0.lt*
0.55*
2.0t*
/o
/0 (7).Downscale' 3'eps 2 cps DD 0.1t*
0.5t*
2.0t*
D Intermediate Rango Monitor k.D (0) Upscalo 108/125 110/125 DD 2.0%
C.St 2.01 2,'/, d (9) Downcealo 5/125-3/125 DD 2.01 0.5%
2.01 E84 O
22) d.4 8
74 -
u o
no
$s
?
-h.
J 5
.Uguivalent Lincar Full Scale.
c.
ENCLOSURE 3 SEPARATION CRITERIA The criteria used in the separation and physical independence design for LaSalle County Station Units 1 & 2 is described in Section 8.3.1 of the FSAR.
In response to inquiries from the NRC Staff a reassessment of these criteria has been performed with the intention of establishing the degree of conformance to Regulatory Guide 1.75 " Physical Independence of Electric Systems" which was published (Feb. 3, 1974) after the LaSalle County Construction Permit was issued (Sept. 1973) and revised (Jan., 1975) after the LaSalle County final design basis was implemented.
The discussion of RG 1.75 contained in Appendix B of the FSAR is being revised to clarify the bases upon which it is judged that the LaSalle County design provides a technically acceptable alternative to the regulatory position of RG 1.75.
This revision was discussed with the NRC Staff on February 6, 1981 and judged to be tentatively acceptable pending adequate documentation.
A copy of the proposed revision to Appendix B is attached and will be' formally documented in a future amendment to the FSAR.
I f
,;e -.,-
PkaPosED)
LSCS-FSAR AMENDMENT. W FEBRUARY 1980 REGULATORY GUIDE 1.75 g
A Initial Issue:
Revision 0, February 1974 Current Issue:
Revision 1, January 1975 La Salle C.P.
Issued:
September 10, 1973 PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS Regulatory Guide 1.75.describas a method acceptable to the NRC for assuring physical independence of the circuits and electric equipment comprising or associated with Class lE power systems, protection systems, systems actuated or con-trolled by the protection system, and auxiliary or support-ing systems that must be operable for the protection systems to perform their safety-related functions.
Regulatory Guide 1.75 was issued after the La Salle County Station construction permit.
However, the LSCS design meets Mo sT the inuat of the regulatory positions, as described in two Subsection 8.3.1.
Aside from thektr.rc_ exceptions noted e-5'7 below (Positions 1,)*Iand 10) the LSCS design complies with this guide.37o s,7, o u a ss Nor APPLac Ast r.
'The LSCS design differs with Position 1 in that the basis for precluding the use of a fault-current-actuated inter-l rupting device as an isolation device within the context l
of Regulatory Guide 1.75 is questionable.
This conclusion seems to preclude the use of a circuit breaker as an iso-l lating device and thus prevents achieving a practical design of an electrical auxiliary system.
A qualified (Class lE) l circuit breaker could otherwise fulfill the literal inter-l pretation of Section 3.8 by preventing (by its successful interruption of fault current) a malfunction in one section l
of a circuit from causing unacceptable influences (i.e.,
failures) in other sections of the circuit (or other cir-
)
cuits).
To preclude Class lE circuit breakers from these applications casts doubt upon the safety-related portions of the auxiliary power systems in all nuclear plants.
Such systems invariably utilize Class 1E metal-clad switchgear and molded case circuit breakers as circuit protection l
devices.
SEE T 'E E2C2 'cci;r full; crmplicc fith Tcciticnc 2, 2,
4, 3,
e g -
- -e e n citi--, -mec-r t-Fectier 4.e.2 cf :::
22 127:
c c^.ccr-i.;
cr-ciccc '" i-etrr-^-trtier
--S ccntrol circuit:
)
net Scing rc7 ire?'t^ "2 rer2-'*^6
'rc-cccccicted circuitc.
T"2 '?^? de-igr ir i-c;rc^-^-t 'itb ?ccticr 5.2 ci ::::
b 4
/
20 '
12~'
ir tP:t -- -ciccc 1 r i-c t r um e n t c hic r cr.3 c o n t rc l
)
l circuitr cre -et rc~uire? tr 'c capcrcted frcr cc.cccictcd s,
i B.1-94 1
INSERT The La Salle County Station design complies with Positions 3 thru 9.
With respect to the separation of associated circuits, the La Salle County Station design complies with Section 4.5 o f IEEE-384-1974.
Although separation of associated circuits is, in some cases, slightly different that that dictated in Sections 5.13 and 5.62 of IEEE-384, Class lE circuits are not degraded below an acceptable level for the following reasons:
a.
Cables associated with one safety-related division are never routed in cable trays or conduit containing cables of a redundant safety-related division.
This is true for all general plant areas.
Lesser separation than that dictated by IEEE-384 occurs only within control panels located in the control room and auxiliary equipment room.
- b. All cables used to interconnect associated circuits are the same high quality as that utilized in Class lE circuits, i.e., all associated cables comply with the requirements of IEEE 383-1974.
Therefore, this cable has been proven to be highly fire retardant by testing.
l
- c. A lesser separation than that dictated by Sections 5.1.3 l
and 5.62 of IEEE-384 is limited to control and l
instrumentation circuits which, by their very naturc, l
are low energy circuits.
Control circuits are geners':ly l
120V a.c. or 125V d.c.,
whereas, the insulation rating of the cable utilized at La Salle County is 600V.
- d. There are not power cables in contact with the control and instrumentation cables in the cable spreading area or in the control room and auxiliary equipment room.
Also, there are no high energy sources located within f
control panels installed in these areas.
I e.
Fire stops are installed in the bottom entrances of all control panels.
L
With respect to the separation of non-Class lE from Class 1E control and instrumentation circuits, La Salle County Station complies with Section 4.6.2 of IEEE-384.
Although the separation of non-Class lE from Class IE control and instrumentation circuits is, in some cases, less than that dictated by Sections 5.1.3 and 5.6.2 of IEEE-384, these l
circuits have been analyzed to show that Class lE circuits are not degraded below an acceptable level for the following reasons:
- a. Non-Class 1E cables are routed in separate cable trays from Class lE and associated cables in general plant areas.
- b. Non-Class lE cables which come in close proximity to Class 1E and associated cables at one end do not come in contact with redundant Class 1E or associated circuits at their other end.
(This has been confirmed by a study of the La Salle County Station installation.)
- c. All cables useo to interconnect associated circuits are the same high quality as that utilized in non-Class lE
- circuits, i.e.,
all associated cables comply with the requirements of IEEE 383-1974.
Therefore, this cable has been proven to be highly fire retardant by testing.
- d. A lesser separation than that dictated by Sections 5.1.3 and 5.6.2 of IEEE-384 is limited to control and instrumentation circuits which by their very nature are low energy circuits.
Control circuits are generally 120V a.c. or 125V d.c.,
whereas, the insulation rating of-the cable utilized at La Salle County is 600V.
- e. There are no power cables in contact with the control and instrumentation cables in the cable spreading area or in the control room and auxiliary equipment room.
Also, there are no high energy sources located within c ontrol panels installed in these areas.
- f. F ire stops are installed in the bottom entrances of all c ontrol panels.
PRCP3 SED LSCS-FSAR AMENDMENT g FEBRUARY 1980 cir'_l'_
~1
-_ L F -n :c ar; lu.
- n rg; c;r;ulu and ;11 cah1_.3 t[
g
(
s z
-2 H tb - ' air'_it:
_r-f " '_1 ;
7 211ric;I to the rup;rc: :nt; cf I :: 20] 1 74.
7 7eqe m,i 7 c.,11,-,~,f ;;
. i g, n..; i t i. ;
- 2. -I 0.
Posi-h tion 10 refers to Section 5.1.2 of IEEE 384-1974 concerning cable and raceway identification.
The LSCS design utilizes cable trays with permanent colored identification markers at each routing point which are assigned an alphanumeric code per Table 8.3-6 of the LSCS-FSAR.
Each cable is assigned a number and segregation code.
This information is placed on a colored tag, of permanent design, which is affixed to each end of the cable.
A similar tag is also affixed to the cable where it enters and exits a penetration.
The LSCS design complies with Positions 11, 12, 13, 14, 15, and 16.
/
i l
l l
l
\\ e
\\
B.1-95
Mr. B.J. Youngblood, Chief Februa ry 10, 1981 ENCLOSURE 4 TESTING OF LOW-LOW SETPOINT SYSTEM LOGIC The low-low setpoint system logic which is being Installed on LaSalle County Units 1 & 2 as a product improvement is intended to reduce the number of relief valve repetative actuations thereby enhancing S/RV load margins. This system is being functionally tested as a part of the LaSalle County initial Test program to assure proper system ope ra t ion. At the request of the NRC Staff, a copy of the system test program was provided on February 6, 1981.
It is our understanding that the NRC will review this information and determine whether periodic surveillance of this system, which is not required to satisfy the plant design basis,need be considered.
In the event periodic surveillance is required, it is our understanding that it will be limited to a logic systen functional test on an 18 month (refueling outage) cycle.
l
Mr. B.J. Youngblood, Chief Februzry 10, 1981 ENCLOSURE 5 POST-ACCIDENT MONITORING AND SAFE-SHUTDOWN INDICATION The LaSalle County Station Unit 1 & 2 design has been upgraded to address the issue of edequate. monitoring during and following an accident. Significant improvements have already been implemented to improve menitoring of Engineered Safety Features (FSAR Section 7.8),
Post-Accident Tracking (FSAR Section 7.5.1.2) and other instrumentation systems required for safety which are related to the subject of monitoring during and following an accident (FSAR Section 7.6).
In addition significant improvements have also been made in response to NUREG-0737 Item II.F.1 related to Post-Accident Honi toring (FSAR Appendix L.29).
Activities are currently underway to assess the need to provide additional instrumentation to satisfy th'e requirements of NUREG-0737 Items II.D.2 and ll.F.2(1) for which criterion applicable to BWR facilities has not been determined as yet by the NRC.
In addition, related areas addressed in NUREG-0737 I tems I l l. A.1.2 and Il l. A.2 on Emergency Preparedness are also being reviewed and will be resolved on a schedule consistent with that ultimately prescribed by the NRC, conditioned only on unforeseen difficulties in procuring and installing additions or modifications to in place instrumentations.
in this regard, Commonwealth Edison will take guidance from th(
position defined in Regulatory Guide 1.97, " Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident".
It is expected that the review of the adequacy of the LaSalle County instrumentation design will be conducted in the context of NUREG-0737 Items 11.D.2, ll.F.2 (1), Ill.A.1.2 and Ill.A.1 and will be resolved on the schedule established for conformance with that document or future revisions to it.
In any event, resolution of these issues prior to LaSalle County Unit 1 Fuel loading is required only insofaras the quidance presently defined in NUREG-0737 require.
One additional area in which the LaSalle County design has been modified to provide additional information related to safe-shutdown i
l Indication is the Rod Position Indication System (RPIS). With the agreement of the NRC Staff a modification to this system will be made as l
l soon as possible (current schedules indicate that this modification can be completed prior to Unit 1 fuel loading) to provide continuous unterruptible-power to the RPIS.
This modificatton and proposed amendments to the FSAR are presented in greater detail in the attachment to this enclosure.
l l
l t
i
(Pksvosav)
LSCS-FSAR AMENDMENT 5/
JANUARY 1981 radiatior. monitor is displayed on an eight-decade meter and on separate stripchart recorders located in the control room.
Redundancy The subsystem utilizes a redundant instrumentation channel so that a single failure cannot prevent subsystem operation.
Separation Each of the redundant pairs of gamma-sensitive instrumentationsis physically separated.from the other and is powered from a separate power bus.
Inspection and Testing A built-in source of current is provided with each radiation monitor for test purposes to provide a point reading equivalent to 105 R/hr.
In addition, the operability of each monitoring channel can be routinely verified by comparing the outputs of the channels at any time.
Environmental Considerations The gross gamma monitoring equipment read-outs in the
(
control room.
See Appendix H for a description of_the reactor building and control room environment.
Operational Considerations The gross gamma subsystem is operational at all times during normal and accident conditions except when taken out of service i
for calibration.
This subsystem covers the range of 100 to 10s R/hr, which is greater than the dose rates in the H2 and O2 sample lines following the loss-of-coolant accident.
l 7.5.1.3 Shutdown, Isolation, and Core Coolina Indication The information furnished to the control room operator permits him to a'ssess reactor shutdown, isolation, and availability of l
emergency core cooling following the postulated accident.
f a.
Operator verification that reactoe shutdown has l
occurred may be made by observing one or more of the following indications:
SEE INSERT 1.
C ntr:1 red rt:t:0 1 rPC indic; tins :a:h - J '
l FO R !. AND2.
52117
!"c^"I^^
SE22 EiTII1 5 1.) Ei i
l
." O ' '9 r COur"O IE CE2 t'. ;
^#
g 7.5-7
INSERT FOR SUBSECTION 7.5.1.3 a 1.
Control rod status lamps indicating each rod fully inserted.
See Figure 7.5-1.
The power source for these lights is +ksmes safety-rGtated S.wetshown on Figures lE-1-4206AE and lE-1-4206AF (in Section 1.7).
2.
Computer printout of rod position information indicating actual rod position.
The power source for both the computer and the rod position indication logic panel (lH13-P615) is the uninterruptible power supply.
The power source to the RPI logic panel (lH13-P615) is shown on Figure lE-1-4213AC (in Section 1.7).
t
---m-
IRO Po3EDj LSCS-FSAR AMENDMENT )MI
, JANUARY 1981 g
iratruz:nt ; ; bu 3[.
Control rod scram pilot valve status indicating open valves._ See Figura 7.5-1.
The power sources are RPS MG sets.
See Figure lE-1-4215AJ (Section 1.7).
Hors ' Ris M q SET 5 ARE PowietD *aam moron sournau
N reRs unsen Ana rewanse reem rose EmaRqancy 06GsEt G en sa.4rces ou toss or oss Po w e e.
Y,.
Neutron monitoring power range ch,75annels and J
recorders downscale.
See Figure 7.5-1.
The l
power sources are RPS MG setse WHita ARE VLTI'.1 ATELV I
rowe nen av r;[sneassucy masac c susanTen on L oss oc e
sre ssre Pow 5 /.
Annunciators for reactor protection system variables and trip logic in the tripped state.
See Figure 7.5-1.
The power source is de from a station battery.
b.
The operator may verify reactor isolation by observing one or more of the following indications:
1.
Isolation valve position lamps indicating valve closure.
See Figure 7.5-2.
The power source is the same as for the associated motor operator.
2.
Main steamline flow indication downscale.
See Figure 7.5-1.
The power source is instrument ac from one of the standby a-c buses.,
9' Operation of'the emergency core cooling and the RCIC c.
custem following an accident may be verified by eving the following indications:
1.
Flow and pressure indications for each emergency core cooling system.
See Figure 7.5-2.
The power sources are independent and from the same standby buses as the. driven equipment.,
2.
RCIC isolation valve position indicating open valves.
See Figure 7.5-2.
The power source is from the same bus as the valve motive power.
i l
3.
Injection valve position lights indicating either open or closed valves.
See Figure 7.5-2.
The power source is the same as the valve motor.
4.
Relief valve position status by open or closed indicator lamps.
See Figure 7.5-2.
The power source is the same as for the pilot solenoid, 120 l
Vac from standby a-c systems.
l 7.5-8 l
)
i LSCS-FSAR AMENDMENT 48
' FEBRUARY 1980 i
detect pressurization upward on the drywell floor due to the suppression pool bypass phenomena, in order to insure that design limits of the drywell floor are not exceeded.
These transmitters feed redundant divisional indicators in the control room to provide instantaneous pressure readings to the operator.
In addition, for the non-accident case, an alarm is provided in the control room to call the operator's attention to these indicators if there is an abnormal pressurd reading.
The transmitters are qualified to IEEE 323-1971 standards, and-each loop in the redundant pair is powered from emergency a-c power, which is capable of being powered by the diesel generators.
4.
Emeroency Core Coolino Performance of emergency core cooling systems following an accident may be verified by observing redundant and independent indications as described in Subsection 7.5.1.2 item c and fully satisfies the need for operator verification of operation of the system.
5.
postaccident Tracking The various indications described in Subsection 7.5.1.2 provide adequate information regarding the status of the reactor vessel level and pressure to allow operators to make proper decisions regarding core and containment cooling operations; they also fully satisfy the need for postaccident surveillance of these variables.
c.
Safe Shutdown Display I
The applir2ble inctre r-tatier ::::: fr:r th: h1;t in Subcertier '.5.1.2 2nd includ:: th
- ntrcl,md
-ct2tuc 12 pe, scra pilot v217c ct:tuc 1:=ps, and 70utrer Tonitoring inctrument: tier Th::: displays are cepected te re=?ir Operabl^ fer 2 1:ng enough IE6 JN S E RT 2 d tim fell: Ming 3"
accident te indic2te the c::urren :
cf caf chutd :7 medundancy ir previded by di/idir; the displays uma thrce ceperete syst-rre red pc=1tice
,-d
, cut:on monit ring Outpute are recorded, th former by the precere cc putar Tha ryst^ r cited are cither e
-'nuelly er aute==tically cornectable to th: ctandbj 2-c pe'cr.
7.5-11
INSERT 2 INSERT FOR SAFE SHUTDOWN DISPLAY - P.7.5-ll PARA.7.5.2.2.C 7.5.2.2.C SAFE SHUTDOWN DISPLAY The applicable instrumentation comes from the list in subsection 7.5.1.2 and includes the computer printout of rod position information, scram pilot valve status lamps and neutron monitoring instrumentation.
These displays are expected to remain operable for a long enough time following an accident or loss of off-site power to indicate the occurance of safe shutdown.
Redundancy is provided by dividing the displays into three separate systems.
The rod position and neutron monitoring outputs are recorded, the former by process computer.
Although the rod position indication on panel P6.3 is lost due to a loss of off-site power, the RPIS cabinet will provide indication of rod position to the computer. (RPIS is powered from the uninterrupted power supply.)
LSCS-FSAR AMENDMENT 48
' FEBRUARY 1980 Comoliance with IEEE 279-1971 Neither the rod status circuitry nor the scram pilot valve status circuitry alone meets the requirements of IEEE 279-1971.
Jointly, however, they do meet the requirements applicable to display instrumentation except for seismic qualification.
The neutron monitoring system is designed to meet all the requirements of IEEE 279-1971 as a part of the reactor protection system.
However, its RPS function is a " fail-safe" function, while safe shutdown display is not.
Further, its RPS function terminates with the generation and maintenance of a shutdown signal.
- n thic regard, pert-D?? crvircr.cnt condition may cauce malfunctier but not until the PPS functier Of ccr:r generctier ic concluded.
Thic,pggg7g 2kec it imperrible te claim centinucuc indicating capability for refe rhutde"a dirplay by the neutrer eenitoring ryste=
In addition, Icer of peeer ic ecceptable to the OPS perform 2nce, but not to the erfe chutdemn display performance 2lthough cuch 2 j
conditier is highly unlikely - Redundancy, power switching capabilities, RPS capabilities, and expected time to failure under DBA environment conditions allow the neutron monitoring system to meet the functional requirements of IEEE 279-1971 as applicable to display instrumentation.
()
7.5-12 7
QO31.281 (Response)
(1) Section 7.5 has been revised to show that all post-accident instrumentation and read-outs are powered from safety-grade power sources, and the read-out devices themselves, as well as the instrumentation which provides the signals are seismically qualified, in accordance with IEEE 344-1975.
(2) Section 7.5.1.2 has been revised to show that suppression pool temperature is recorded on redundant, seismically-qualified recorders, and that all post-accident tracking parameters are redundant, with at least one channel recorded, and indication on all channels.
(3) Section 7.5.1.2 a) has been revised to sh'cw that the rod position information system, which feeds shutdown position information to the process computer is fed from the uninterruptible power supply, and that the control rod scram pilot valve status lamps, and the neutron monitoring power range channels are ultimately fed from safety grade power supplies.
(
g e.-,
m -.
e--e-e
Mr. B.J. Youngblood, Chief February 10, 1981 ENCLOSURE 6 ROD BLOCK MONITOR SYSTEM The NRC Staff has conducted a thorough review of the LaSalle County Station Reactor Manual Control System (RMCS) and specifically the Rod Block Monitor (RBM) Subsystem. This system, the logic and design of which is identical to the Zimmer Station system reviewed in NUREG-0528, represents a design improvement over that previously IIcensed on the Hatch-2 docket.
In that regard, the LaSalle County system design is the same as Hatch-2 except for the new electronic multiplexing feature.
A GE/NRC generic meeting was held on January 16, 1981 to discuss the RMCS and to specifically discuss the propriety of utilizing the RBM in the analysis of certain operational transients. The new electronic RMCS utilized at LaSalle County was described in detail with emphasis on the reliability redundancy and self-testing features of the system.
At that time the NRC Staff indicated tentative approval of the design and transient analysis with the possible addition of periodic Technical Specification surveillance to better assure proper system operation. The LaSalle County technical specification will be amended to include this additional functional testing of the multiplexing feature of the RMCS.
Proposed additions to the appropriate technical specification are included for Information as an attachment to this enclosure
)
TABLE 4.3.6-1 S
CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREl(ENTS U
CHANNEL OPERATIONAL CHANNEL FUNCTI0f:AL CHANNEL C0liDITIONS FOR WtilCH l
CALIBRATION (")
SURVEILLANCE REQUIRED TRIP FUNCTION CHECK TEST 4 R00 BLOCK MONITOR N 1.
a.
Upscale NA S/U Q
la S/U,b ) ',
HA b.
Inoperative NA L
Q la S/U c.
Downscale NA 2.
APRM a.
Flow Biased Simulated Thermal S/UfD,H Q
1 4%
Power - Upscale HA' HA 1, 2, 5 b.
Inoperative NA S/U(b),H f
c.
Downscale NA S/U(b),H Q
l O
d.
Neutron Flux - Upscale, Startup NA S/U
,H Q
2, 5 j
4[
3.
SOURCE RANGE MONITORS S/U(b) (c) 2, 5 im a.
Detector not full in HA y
b.
Upscale NA S/U Q
2, 5 NA 2, 5 S/U(b)' (c) g c.
Inoperative itA d.
Dounscale NA S/U Q
2, 5 4.
INTERMEDIATE RANGE HONITORS g
S/U(b), IC)
ID)
NA 2, 5 J
a.
Detector not full in NA b.
Upscale NA S/U Q
2, 5 I
(c) c.
Inoperative NA S/U NA 2, 5 k'
d.
Downscale NA S/U Q
2, 5 5.
Water Level-High itA Q
R 1, 2, 5""
b.
Scram Trip Bypassed NA H
NA 1, 2, 5"a 6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW II a.
Upscale NA S/UI ),H Q
l
!!A I
b.
Inoperative 11A S/U(b),H c.
(Cczparator) (Downscale)
NA S/U
,g 9
Q l S W *WU l " TrY g
l,gh
\\lJ' TABLE 4.3.6-1 (Continued)
CONTROL RCD WITH0RAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS l
I
.l l
1 NOTES:
a.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
i l
b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the 1
previous 7 days.
l c.
When making an unscheduled change from OPERATIONAL CONDITION 1 to OPERATIONAL CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering OPERATIONAL CONDITION 2.
With THERMAL POWER > (20)% of RATED THERMAL POWER.
f l
With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
i l
i l
8 I
b i
e l
GE-sTs 3/4 3-55 1
- - - - - - -..