ML20195E010
| ML20195E010 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/09/1998 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20195D999 | List: |
| References | |
| NUDOCS 9811180217 | |
| Download: ML20195E010 (43) | |
Text
{{#Wiki_filter:,_. . _ . .. _ _ _ . . _ _ _ _ _ . . _ _ _ _ . _ _ _ _ . . . - . . . . . _ _ . _ . . _ , . . . . . . _ . . _ _ . - - . . . . . . _. I TABLE 3.3.2-1 , ISOLATION ACTUATION INSTRUIENTAT10ll ' VALVE GROUPS MINIMUM OPERABLE APPLICA8tE OPERATED BY- CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL . TRIP SYSTEM (b) C0191T1011 ACTION
- A. ADIOMATIC INITIATION .
- 1. PRIMARY CONTAll0ENT ISOLAT10ll '
- a. Reactor Vessel Water Level fil- Low, Level 3 7 2 1, 2, 3 I f2'i Level 2 2, 3 20 '
L3) Low Low Low Low, Low, Level 1 1, 10 2 1, 2, 3 20 2 1, 2, 3 20
- b. Drywell Pressure - High 2, 7, 10 i 2 1, 2, 3 20
/
- c. -Main Steam Line i
11 DELETED 2'I Pressure - Low 1 2 1 l 3? Flow - High 23 1 2/11ne'* 1, 2, 3 21
- d. DELETED
- e. Main Steam Line Tunnel i ATemperature - High 1 2 ,
l' ' ',',3 ' ,2" " 3 ' ,
. 3 ,g
~
21 g Condenser Vacuum - Low kO f. 1 2 1, 2*, 3' 21
- 2. SEC0 LEARY CONTA110ENT ISOLATI0ll a.
O O. O 3 b Reactor Bulldt Ven i PlenumRadiaNontExhaust.4'*"*' High 2 1, 2, 3 and ,,
- b. Drywell Pressure - High 24
/ cT-4 '*"*8 2 1, 2, 3 24
- c. Reactor Vessel Water Level - Low Low, Level 2 4'*"** 8 Cf yw 1, 2, 3, and
- d. Fuel Pool Vent Exhaust Radiation - High 4 ""*'
.2 2 1, 2, 3, and ,,
24 24 L XD
-Q y 3
' o-X ' 5, - A LA SALLE - UlllT 2 3/4 3-11 Amendment No.100 L ! 9811180217 981109 PDR
- d P
ADOCK 05000373' PDR a
.. - =. . ._ . . .- . ~ . . ... - . - .
d TABLE 3.3.2-1 (Continued) ISOLATION ACTUATION INSTRIMENTATION VALVE GROUPS 'NININUN OPERA 8LE 'APPLICA8tE ; OPERATED BY CHANNELS PER OPERATIONAL ' TRIP FUNCTION SIGNAL TRIP SYSTEN (b) 00lWITION I ACTION _ .
- 3. REACTOR WATER CLEANUP SYSTEN ISOLATION
- a. A Flow - Wigh 5
# 1 1,2,3 22 ,
t
- b. Heat Exchanger Area 5 1/ heat 1, 2, 3 22 N Temperature - High y exchanger :
T c. 5- .1/ heat 1, 2, 3 22 f LtJ HeatExchaErArea Ventilat AT - Hloh
~
exchanger
! l d. SLCS Initiation 5"' NA 1, 2, 3 22 d e. Reactor Vessel Water Level - Low Low, Level 2 5 2 1, 2, 3 22
- 4. REACTOR CORE ISOLATION COOLING SYSTEN ISOLATION
- a. RCIC Steam Line Flow - High 8 1 1, 2, 3 22
- b. RCIC Steam Supply Pressure - Low 8, 9 2 1, 2, 3 22 '
- c. RCIC Turbine Exhaust .
Olaphrage Pressure - High 8 2 1, 2, 3 ,. 22 !
- d. RCIC Equipment Room . i I Temperature - High 8 1 1, 2, 3 22 t' e. RCIC Steam Line Tunnel Temperature - High 8 1 1, 2, 3 22 ;
- f. RCIC Steam Line Tunnel !
A Temperature - High 8 1 1, 2, 3 22 i
- g. Drywe11 Pressure - High 9 2 1, 2, 3 22
- h. RCIC Equipment Room ;
A Temperature - High 8 1 1, 2, 3 22 - l t LA SALLE - UNIT 2 3/43-12 Amendment No. 87
ATTACHMENT B MARKED-UP PAGES FOR PROPOSED CHANGES + INSERT A* VALVE GROUPS M1NLMUM APPLICABLE OPERATED BY- OPERABIE OPERATIONAL TRIP FUNCITON SIGNAL CHANNEllPER CONDITION ACTION TRIP SYTIEM M E Pump and Valve Area Temperature - High 5 1/ area 1,2,3 22
- g. Pump and Valve Area Ventilation AT - High 5 1/ area 1, 2, 3 22 It Holdup Pipe Area Temperature - High 5 1 1,2,3 22
- i. Holdup Pipe Area Ventilation AT - High 5 1 1,2,3 22
- j. Filter /Demineralizer Valve Room Area Temperature - High 5 1 1,2,3 22.
- k. Filter /Demineralizer Valve Room Area Ventilation AT - High 5 1 1,2,3 22
- 1. Pump Suction Flow - High 5 1 1,2,3 22
- Headings in Italics are provided for information only.
B-2
TA8tE 3.3.2-1 (Continued) ISOLATION ACTUATION INSTNISENTATIoll El yED vAtvE sRouPS OPERATED BY nlNinun OPERA 8tE CHAlelELS PER APPtiCA8tE OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b) C0151T1011 ACTION
- 5. FRHR SYSTIM STEAM CONDENSING MODE ISOLAT10ll x
- a. RHR Equipment Ares
& Temperature - High 8 1/lHtarea 1, 2, 3 22
- b. RHR Area Temperature -
High 8 1/RHR area 1, 2, 3 22
- c. RHR Heat Exchanger Steam .
Supply Flow - High 8 1 1, 2, 3 22 ]
- 6. RHR SYSTEH SHUTDOWN COOLING MODE ISOLATION i a. Reactor Vessel Water
/ Level - Low, Level 3 6 2 1, 2, 3 25
. b. Reactor Vessel (RHR Cut-in Permissive)
Pressure - High 6 1 1, 2, 3 25
- c. RHR Pump suctlen Flow - High 6 1 1, 2, 3 25 I
/.d RHR Area Temperature -
High 6 1/ftHR area 1, 2, 3 25 (e. RHR Equipment Area AT - High 6 I/IIHR area I, 2, 3
- 8. MAlluAL INITIATION
- 1. Inboard Valves I,2,5,6,7 1 group 1, 3 26
- 2. Outboard Valves 1 5, 6 1 group 1 3 26 1
- 3. Inboard Valves 9 4",(' 1 group 1, 3 and.**,f 26
- 4. Outboard Valves 4 "' I group 1, 3 and **,# 26
- 5. Inboard Valves 3,8,9 1 valve 1 ,3 26
- 6. Outboard Valves 3 ,8, 9 1 valve 1 ,3 26
- 7. Outboard Valve 83 1 group 1 . 3 26 LA SALLE - UNIT 2 3/4 3-13 Amendment No. 105
0 OfS) WOVIQ
' Q "ABtr 3 3 H 1 (Continued)' bCOMfh4 0 ISOLA" ION ACTUA" ION INSTRUMENTATI Y ACTION STATEMENTS ACTION 20 -
Be in at least NOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. ACTION 21 - Se in at least STARTUP with the associated isolation valves closed I within 6 hours or be in at least HDT SHUTDOWN within 12 hours <and l in COLD SHUTDOWN within the next 24 hours. 1 ACTION 22 - Close the affected system isolation valves within I hour and declare the affected system inoperable. ACTION 23 - Be in at least STARTUP within 6 hours. i ACTION 24 - Establish SECONDARY CONTAlletENT INTEGRITY with the standby gas < treatment system operating within I hour. I
\
ACTION 25 - Locx the affected system isolation valves closed within I hour ,and l declare the affected system inoperable. ! ACTION 26 - Provided that the manual initiation function is OPERABLE for each other group valve, inboard or outboard, as applicable, in each line, restore the manual initiation function to OPERABLE status within 24 hours; otherwise, restore the manual initiation function to OPERA 8LE status within 8 hours; otherwise:
- a. Be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, or
- b. Close the affected system isolation valves within the next hour and declare the affected system inoperable.
TABLE NOTATIONS May be bypassed with all turbine stop valves not full open. ! When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
# During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
(a) Deleted. (b) A channel may be placed in an inoperable status for up to 6 hours for I required surveillance without placing the channel in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter. In addition for those trip systems with a design providing only one channel per trip system, the channel may be placed in an inoperable status for up to 8 hours for required surveillance testing without placing the channel in the tripped condition provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is operable and all required actuation instrumentation for that redundant valve is OPERABLE, or place the trip system in the tripped condition. (c) Also actuates the standby gas treatment system. i (d) A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE. i (e) Also actuates secondary containment ventilation isolation dampers per (f) ses 1 bsysteminletoutboardvalve. i ( LA SALLE - UNIT 2 3/4 3-14 Amendment No. 109 l l ym - - w 7 -cy , y. -- --- m
0 0n003 fPOV$ TABLE 3.3.2-1 (Continued) rbD U) @ l HQI.El (Continued) (g) Requires RCIC steam supply pressure-low coincident with drywell pressure-high. (h) Manual initiation isolates 2E51-F008 only,and only with a coindident reactor vessel water level-low, level 2, signal. (1) Both channels of each trip system may be placed in an inoperable status for up to 4 hours for required reactor building ventilation system corrective maintenance, filter changes, damper cycling and surveillance tests, other than Surveillance Requirement 4.6.5.1.c, without placing the trip system in the tripped condition. (j) Both channels of each trip system may be placed in an inoperable status for up to 12 hours due to loss of reactor building ventilation or for . performance of Surveillance Requirsiment 4.6.5.1.c without placing the trip system in the tripped condition. I
~
LA SALLE - UNIT 2 3/4 3 343' Amendment No. 96
JMit 3.3.2-2 i ISOLATION AC10 Ail 0N INSTRUMENTATION SETP0lNTS 1 RIP FUNCTION ALLOWABLE-IRIP SEIPOINT VALUE ' A. AUTOMATIC INITIATION
- 1. I PRIMARY CONTAll8 TENT ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, level 3 2 12.5 inches *
- 2) Low Low, Level 2 1 11.0 inches
- 2 -50 inches
- 1 -57 inches *
- 3) Low low Low, Level 1 2 -129 inches
- i'
- b. Drywell Pressure - High 2 -136 inches
- s 1.69 psig s 1.89 psig
- c. Main Steam Line 1
- 1) DELETED
- 2) Pressure - Low 2 854 psig .!
- 3) Flow - High 1 834 psig s 111 psid s 116 psid
- d. DELETED i
- e. Main Steam Line Tunnel a Temperature - High 5 65'F
- f. Condenser Vacuum - Low $ 70*F
> 7 inches Hg vacuum > 5.5 inches Ng vacuum
- 2. SEC0lWARY CONTAllflENT ISOLATION
- a. Reactor Sullding Vent Exhaust !
Plenum Radiatten - High s 10 er/h
- b. Drywell Pressure - High s 1.69 psig s 15 mr/h. t
- c. Reactor Vessel Water s 1.89 psig Level - Low Low, Level 2 2 -50 inches *
- d. Fuel Pool Vent Exhaust 2 -57 inches
- Radiation - High 5 to ar/h s 15 mr/h
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION .
- a. AFlow - High i s 70 gpa s 87.5 gym
- b. Heat Exchanger Area Temperature
- High r 'A.
M37 s 81* Ifii[8F
- c. Heat Exchanger Area Ventilation 3(87*I) -
AT - High T *
- d. SLCS Initiation s 5 457 ' 3h*F
- e. Reactor Vessel Water Level -
Low Low, level 2 M.A. N.A. 3f ; T " 2 -50 inches
- 2 -57 inches
- LA SALLE - UNIT 2 I J_NSERTB 3/4 3-15 Amendment No. too i !
v ATTACHMENT B MARKED-UP PAGES FOR PROPOSED CHANGES INSERT B* AllDiVABLE TRIP FUNCTION TRIP SETPOINT VALUE e f. Pump and Valve Area Temperature - High s 201*F 5 209*F -
- g. Pump and Valve Area Ventilation AT- High s 86*F s 92.5*F
- h. Holdup Pipe Area Temperature - High s 201*F s 209'F
- i. Holdup Pipe Area Ventilation AT - IIigh 5 86*F s 92.5*F
- j. Filter /Demineralizer Valve Room Area Temperature - High s 201*F s 209*F -
- k. Filter /Demineralizer Valve Room Area Ventilation AT - High 5 86 F s925*F L Pump Suction Flow - High s 560 gpm 5 610 gpm
- Headings in Italics are provided for information only.
B-3
- v. .
5 - TABLE 3.3.2-2 (Continued) ISOLATION ACTUATION IllSTRISENTATION SETPOINTS e
- ALLOWASLE TRIP FnsecTI0le , TRIP SETPOINT VALUE
, '4. REACTOR CORE ISOLATION C00LIIIB SYSTEM IN_4TIGII '
- a. KIC Steam Line Fleet P Wish { 29 5 of rated flew, 178" N e
- b. ly Pressure - Low 2 < 295K of rated flow, 185" H O KIC Steam .-> 57 psig i 53 psig 2
- c. ScIC Tidine whamat Rfaphrase -
Pressure - Niet - $ 10.0 psig 5 20.8 psis - d, K IC Egulpment Asse Temperature - High < 200T e.
-< 206T KIC Steam Line Tunnel Temperature - Nigh < 200*F g -
-< 206*F
- f. KIC Steam Line Tunnel '
- A Temperature - Nigh . -
1 117'F < 123*F y g. 11 Pressure - High 1 1.69 psig $ 1.89 psig g h. C Equipment Reen
& Temperature - Nfsh 5 120*F i 126*F 5 [AHR $YSTEM STEAN CafEIEllSING IEIDE ISOLA ,
- s. M Equipment Area "
A Tamperature - Nigh 5 50*F 1 56*F
- b. M Area Cooler Temperature -
N1gh 5 200*F $ 206*F
' c. M Neat Exchanger steam Supply Flow - liigh 1 123" N20 < 128" H 2O
- ~
'.. DELETED '
9
l
. 9, t
O W . 1 .1 5 s ER t 's e - s Al vi vi vi v i, I f I g .
- 1s i
}l
?,g 3 -{
lc -
~
s a gac t* >- j = 2
.s h Al vi vi vt VI N $ .
s g . m j E s . I l
< ~ e s
'S _ ."
1s t .
.... . t, ,
?
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"rIj'Ifg j 13 1
.[ [.[g
!!j 13.1 !j='2j=i =
- n. 1 LI3:s .
a gi _ ,u a3 hs ! w.i :sse ! ' ae e s as
'LA SALLE - UlGT 2 3/4 3-17
. - - . ~ - - . . . . - . - . - . - . . - . - . . . - . - . - . . - . . . - . _ - . - . . . . . . ~ . . . - - . . - TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME-TRIP FUNCTION RESPONSE TIME (Seconds M A. AUTOMATIC INITIATION
- 1. PRIMARY CONTAIPMENT ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, Level 3 N/A
- 2) Low Low, Level 2 N/A
- 3) Low Low Low, Level 1 g 1. 0** " ,
- b. Drywell Pressure - High N/A
- c. Main Steam Line -
l
- 1) DELETED I
- 2) Pressure - Low $ 2.0** "
- 3) Flow - High 5 0. 5** " *
- d. DELETED
- e. Condenser Vacuum - Low N/A
- f. Main Steam Line Tunnel ATemperature - High N/A
- 2. SECONDARY CONTAINMDfT ISOLATION N/A
- a. Reactor Building Vent Exhaust Plenum Radiation - High
- b. Drywell Pressure - High
- c. Reactor Vessel Water Level - Low, Level 2
- d. Fuel Pool Vent Exhaust Radiation - High
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION N/A
- a. AFlow - High
- b. Heat Exchanger Area Temperature - High c.
d. Heat Exchanger Area Ventilation AT-High SLCS Initiation INSERTC
- e. Reactor Vessel Water Level - Low Low, Level 2
- 4. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION N/A
- a. RCIC Steam Line Flow - High
- b. RCIC Steam Supply Pressure - Low
- c. RCIC Turbine Exhaust Diaphragm Pressure - High
- d. RCIC Equipment Room Temperature - High
- e. RCIC Steam Line Tunnel Temperature - High
- f. RCIC Steam Line Tunnel ATemperature - High
- g. Drywell Pressure - High
- h. RCIC Equipment Room ATemperature - High
- 5. RHR SYSTEM STEAM CONDENSING MODE ISOLATION N/A
- a. RHR Equipment Area ATemperature - High
- b. RHR Area Cooler Temperature - High Q. RHR Heat Exchanger Steam Supply Flow High LA SALLE - UNIT 2 DELETED 3/4 3-18 Amendment No. 100
1 ATTACHMENT B MARKED-UP PAGES FOR PROPOSED CHANGES I 1 INSERT C
- f. . Pump and Valve Area Temperature - High
- g. Pump and Valve Area Ventilation AT - IIigh
- h. Holdup Pipe Area Temperature - Iligh
- i. IIoldup Pipe Area Ventilation AT - High _
- j. Filter /Demineralizer Valve Room Area Temperature - High j
- k. Filter /Demineralizer Valve Room Area Ventilation AT - High
- 1. Pump Suction Row - Iligh t
IM
TABLE 3.3.2-3 (Continued) ISOLATION SYSTEM INSTRtMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Second gf
- 6. RNR SYSTEM SHUTfM C00LINc M00E ISOLATION N/A
- a. Reactor Vessel Water Level - Low, Level 3
- b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High
- c. RiA P - Suction F1_ow - High -
~. A R Area Cooler Temperaf,Ure Hig RNA Equipment Area AT High
- 5. MANUAL INITIATION N/A ,
I. Inboard Valves
.1. Outboard Valves
- 3. Inboard Valves
- 4. Outboard Valves
- 5. Inboard Valves
- 6. Outboard Valves
- 7. Outboard Valve TABLE NOTATIONS Isolation system instrumentation response time for MSIVs only. No diesel generator delays assumed.
Isolation system instrumentation response time specified for the Trip Function actuating the MSIVs shall be added to MSIV isolation time to obtain ISOLATION SYSTEM RtSPONSE TIME for each valve. Sensor is eliminated from response time testing for the MSIV actuation logic circuits. Response time testir.g and conformance to the administrative limits
-for the remaining channel including trip unit and relay logic are required.
N/A Not Applicable. LA SALLE - UNIT 2 3/4 3-19 Amendment No.100
- ..m_.._ . _ _ - . __ _ _ - _ . . - . _ _ . - . - . _ . _ . -_._ .__ _. -
=
1ABLE 4.3.2.1-1 ISOLATION AC1UATION INSTRUNENTATION SURVEILLANCE REOUIREE NTS CHANNEL OPERATIOML CHAllNEL FUNCTIONAL CHAlWlEL CONDITIONS FOR WHICH
-TRIP FUNCT[0!i CHECK TEST A. AUTOMATIC INITIATI0lt CAL 1811AT10ll SURVEILLANCE REGUIRED a
- 1. PRIMARY CONTAllWlENT ISOLATI0lt
- a. Reactor Vessel Water Leve,1
- 1) Low, Level 3 5 Low Low, Level 2 4 R I, 2, 3
- 2) flA '
Q R I, 2, 3
- 3) Low Low Low, Level 1 5
- b. Drywell Pressure - High NA Q R 1, 2, 3
- c. Main Steam Line Q Q 1, 2, 3
- 1) DELETED
- 2) Pressure - Low NA
- 3) Flow - High MA Q Q l
- d. DELETED Q R 1, 2, 3
- e. Condenser Vacuum - Low NA
- f. Main Steam Line Tunnel Q Q 1, 2*, 3* i
& Temperature - High NA Q R 1, 2, 3
- 2. SEColeARY CONTAllBIENT ISOLATI0li
- a. Reactor Building Vent Exhaust Plenum Radiatfor. - High S
- b. Drywell Pressure - High NA Q R I, 2, 3 and ** i
- c. Q Q I, 2, 3 Reactor Vessel Water -
Level - Low Low, Level 2 NA
- d. Fuel Pool Vent Exhaust Q R 1, 2, 3, and f '
Radiation - High S Q R 1, 2, 3 and **
- 3. REACT 0lLWATER.CLEAlluP SYSTDI ISOLATION
- a. A Flow - High 5
- b. Heat Exchanger Area Q R 1, 2, 3 Temperature - High IIA !
- c. Heat Exchanger Area Q Q 1, 2, 3 Ventilation AT - High itA
- d. SLCS Initiation MA Q
R Q' 1, 2, 3
- e. Reactor Vessel Water NA 1, 2, 3 Level - Low Low, Level 2 11A
% Q R 1, 2, 3 -
LA SALLE - UNIT 2 [h.bbkTb 3M 3-20 h ndment 900. 100 I 1
I ATTACHMENT B MARKED-UP PAGES FOR PROPOSED CHANGES INSERT D* OfANNEL- OPERA 110NAL OfANNEL FUNCT'ONAL CHANNEL CONDH10NS FOR WHICH TRIP FUNCTION CHECK TEST CAI1BRATIO_N SURVRITI ANQi REQUIRRD
- f. Pump and Valve Area Temperature - High NA Q Q 1,2,3
- g. Pump and Valve Area Ventilation AT - High NA Q Q 1,2,3
- h. Holdup Pipe Area Temperature - High NA Q Q 1,2,3
- i. Holdup Pipe Area Ventilation AT - High NA Q Q 1,2,3
- j. Filter /Demineralizer Valve Room Area Temperature - High NA Q Q 1,2,3
- k. Filter /Demmeralizer Valve Room Area Ventilation AT - High NA Q Q 1,2,3 L Pump Suction Flow - High S Q R 1,2,3
- Headmgs in Italics are provided for information only.
. B-5
TABLE 4,3.2,1-1 (continued) ISOLATION ACTUATION INSTRtetENTATION SURVEf t. "E REQUImmegyg CHANNEL OPERATIONAL
, CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR NHICH TRIP FUNCTION CIIECK TEST CALIBRATION SURVEILLANCE REQUIRED
- 4. REACTOR CORE ISOLATION COOLING SYSTEN ISOLATION
- a. RCIC Steam Line Flow - High NA Q Q 1, . 2, 3
- b. RCIC Steam Supply Pressure -
Low NA Q Q 1, 2, 3
- c. RCIC Turbine Exhaust Diaphragm Pressure - High NA Q Q 1, 2, 3
- d. RCIC Equipment Room Temperature - High NA Q Q 1, 2, 3
-e. RCIC Steam Line Tunnel Temperature - High NA Q Q 1, 2, 3,
- f. RCIC Steam Line Tunnel A Temperature - High NA Q Q 1,2,3
- g. Drywell Pressure - High NA Q Q 1, 2, 3
- h. RCIC Equipment Room A Temperature - High NA Q Q 1,2,3 5.
.kHRSYSTEMSTEAMCONDEMSINGMODEISOLATIOM -
- a. RHR Equipment Area A
, Temperature - High NA Q Q 1, 2, 3 *
- b. RHR Area Cooler Temperature -
High NA Q Q 1, 2, 3
- c. RHR Heat Exchanger Steam Supply Flow - High NA Q Q 1, 2, t
JELETED
^
i l t LA SALLE - UNIT 2 3/4 3-21 Amendment No. 90
b 3 TABLt . 3.2.1-1 (Continued) ISOLATION ACTETION INSTRISENTATION SEVEILLima.E Namlimma,3 CHAIRIEL CMMEL FUNCTIONAL OPEMileML ' TRIP FUNCTION CMMEL CO MITIONS FOR M ICH CHECK TEST CALIBRATIOll SARTEILLANCE REBUIRED 6. RHR SYSTEM S Minnuu emuINE MODE ISOLATidll
- a. Reactor Vessel Water Level -
Low, Level 3
- b. 5 Q-Reactor Vessel R 1. 2, 3.
i (IHL Cut-in Permissive) ' Pressure-High fl4 1
- c. lHt Pumo Section Flow-Hlah Q Q 1, 2, 3 (i. M 0 0 Area Temperature-High ' NR 1. 2. 3-Q s (e. IIHR Equipment Area AT-High M Q 1, 2, 3 *
.R.
_ Q Q 1, 1. }}_ MANAL INITIATI0li
- 1. Inboard Valves
- 2. Outboard Valves M R M I, 2, 3
- 3. MA R Inbeard Valves MA IIA 1, 2, 3
- 4. Outboard Valves IIA R M I. 2, 3 and ** #
- 5. Inboard Valves R M 1, 2, 3 and **,,8
- 6. HA R Outboard Valves flA HA 1, 2, 3
- 7. Outboard Valve R IIA 1, 2, 3 llA R IIA 1, 2, 3
** Not required when all turbine stop valves are not full open.
11 hen hand 11ag irradiated the reactorfuel in the secondary containment and during ColtE ALTERATIONS and op potential for draining vessel. During CORE ALTERAllolls and operations with a potential for draining the reactor vessel. . LA SALLE - UlllT 2 3/4 3-22
- AmendmentNo.(g
l ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edi:m (Comed) has evaluated this proposed amendment and determined that it involves no significant hazards consideration. According to 10 CFR 50.92(c), a proposed amendmer'.c to an operatiry license involves nc significant hazards consideration if operation of facility in accordance with the proposed amendment will not' Involve a significant increase in the probability or consequences of an accident previously evaluated; Create the possibility of a new or different kind of accident from any previously analped; of Involve 2 significant reduction in a rnargin of safety. Comed proposes .o amend Appendix A, Technical Specifications, of Facility Operating License NPF-1s, LaSaile County Station Unit 2. The proposed amendment requests a change to the Technical Specifications which: adds automatic primary containment isolation on Ambient and Differential Temperature (AT). Iligh for the Reactor Water Cleanup System (RWCU) Pump, Pump Valve, Holdup Pipe, and Filter /Demineralizer (F/D) Valve Rooms; l adds automatic primary containment isolation on RWCU Pump Suction Flow - Iligh; l revises the ambient temperature and AT isolation serpoints in the RWCU Heat Exchanger Rooms; eliminates the Residual IIeat Removal (RHR) System steam condensing mode isolation actuation instrumentation; and a eliminates the ambient temperature and AT alarm and isolation functions for the RHR System shutdown cooling mode. Currently the RWCU pump rooms have no temperature monitoring because they contain
" cold" piping. A design modification that restores " hot" suction to the RWCU pumps l determined that the RWCU pump rooms and associated new pump valve room require leak l detection isolation instrumentation. The serpoints for the heat exchanger rooms are being L changed as a result of new design basis calculations. The new ambient temperature and AT leak l detection for the RWCU holdup pipe area and the F/D valve room and the RWCU pump suction high flow switch are being added to minimize the impact ofline breaks in these areas.
- The steam condensing mode of the RIIR system is no longer utilized. Area temperature monitoring of the RHR shutdown cooling mode lines is being deleted, because the system mode has been recognized as a moderate energy line; and, because area temperature monitoring is not C-1 i
i
~. . . ._
I i l ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION effective, since the energy in these lines is not sufficient to increase the area temperatures to ' detectable levels. %ese lines are in service in Cold Shutdown or at low reactor pressures in Hot < Shutdown. I The proposed leak detection changes for LaSalle Unit 2 are same, except for minor differences, as the LaSalle Unit 1 changes approved by Amendment 129 to LaSalle County Station Unit 1 Facility Operating License NPF-11, issued July 6,1998. Unless otherwise specified, all references 4 to Technical Specifications are to LaSalle Unit 2 Technical Specifications. The determination that the criteria' set forth in 10 CFR 50.92 are met for this amendment request is indicated below:
- 1) Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
a) %ere is no effect on accident initiators so there is no change in probability of an accident. A line break in the subject areas, would consist of an instantaneous circumferential break downstream of the outermost isolation valve of one of these systems. The leak detection isolation is only a precursor of a break, and thus does not affect the probability of a break. b) %ere is minimal effect on the consequences of analyzed accidents due to changing the leak detection ambient temperature or AT serpoint and allowable values to detect 25 gpm equivalent leakage. The addition of more ambient temperature and AT leak detection monitoring, along with the addition of the high flow break detection will actually decrease the consequences of the associated accidents. The worst case accident outside the pehacy containment boundary is a main steam line break which bounds the dose con,equences of all line breaks and therefore bounds any size ofleak. The deletion of the RHR steam condensing mode isolation a:tuation instmmentation trip functions from the LaSalle Technical Specifications does not increase the probability or consequences of an accident previously evaluated, because this mode of operation of the RHR system has been deleted from the LaSalle design basis and the lines that were previously high energy lines are isolated during unit operation, including Operational Condition 1 (Run mode), Operational Condition 2 (Startup mode), and Operational Condition 3 (Hot Shutdown) The deletion of the RHR shutdown cooling mode leak detection T and AT isolation L actuation instmmentation trip functions from the LaSalle Techrscal Specifications does not l increase the probability or consequences of an accident previously evaluated, because the leak detection is only a precursor of a break, and thus does not affect the probability of a C-2 l i
I
~
ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION 1 l break. Also, there are two other methods of detecting abnormal leakage and isolating the I system in Technical Specification trip functions A.6.2, Reactor Vessel Water Level - Low, Level 3 and A.6.c, RHR Pump Suction Flow - High. In addition, other means to detect leakage from the RHR system, such as sump monitoring and area radiation monitoring, are also available. In accordance with Technical Specification Administrative Requirement 6.2 F.1, LaSalle has a leakage reduction program to reduce leakage from those portions of systems outside primary containment that contain radioactive fluids. RHR, including piping and components asociated with the shutdown cooling mode, is part of this program, which includes periodic visual inspection of the system for leakage. He sump monitoring, radiation monitoring and periodic inspections for system leakage makes the probability of a leak of 5 gpm going undetected for more than e day very low. Also, due to the low reactor pressures (less than 135 psig) at which RHR shutdown cooling l mode is able to operate, reactor coolant makeup and outflow is very low compared to i normal plant operation. A change in flow balance due to a leak is thus more readily detectable with reactor coolant water level changes and makeup flow rate, and thus precludes a significant leak going undetected before break detection instrumentation would cause automatic isolation. Therefore, this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
@ Create the possibility of a new or different kind of accident from any accident previously evaluated because:
The purpose of the leak detection system, as it applies to the RWCU and RHR system areas,is to provide the capability for leak detection and automatic isolation of the system as necessary in the event ofleakage in these areas. His change maintains this capability with at least two different methods of detection of abnormal leakage for protection from the flooding concems of a significant leak or line break when the RHR system is operating in the shutdown cooling mode, so that redundant systems will not be affected. His change also maintains or adds primary containment isolation logic for the leak detection isolation based on temperature monitoring in RWCU areas and break detection based on RWCU pump suction flow - high. The Jitimal instrumentation and the associated isolation logic is the same or similar to existing instrumentation and logic for containment isolation actuation instrumentation, so no new failure modes are created in this l way. herefore, these proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. i C-3
ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION
- 3) Involve a significant reduction in the margin of safety because:
%e change to the automatic isolation setpoint for high AT leak detection in the heat exchanger rooms is based on current configuration calculated / analyzed response to a small leak compared to a circumferential break. The increased leakage rate in the RWCU heat exchanger rooms that is necessary to actuate isolation on ambient temperature during winter conditions, does not adversely affect the margin of safety. This increased leakage rate is below the critical crack leakage rate as represented in UFSAR Figure 5.2-11. Additionally, differential temperature leak detection is conservative under these same conditions, and will actuate isolation at a leakage rate less than the established limit. The leak detection isolation logic is unchanged and thus remains single failuce proof.
The addition of automatic primary containment isolation on ambient temperature and AT-Iligh for the Reactor Water Cleanup System (RWCU) Pump, Pump Valve, IIoldup Pipe, and Filter /Demineralizer (F/D) Valve Rooms and the addition of the RWCU Pump Suction Flow IIigh line break isolation add to the margin of safety with respect to leak detection and line breaks in the RWCU system, because the system isolation diversity is increased and the amount of system piping monitored for leakage is increased. The setpoints for the ambient temperature and AT leak detection isolations being changed or added and the RWCU pump suction flow - high are set sufficiently high enough so as not to increase the possibility of spurious actuation. In the event that a spurious actuation does occur,little safety significance is presented since the RWCU system performs no safety function. The setpoints and allowable values for the proposed changes also assure sufficient margin to the analyti:al values and are high enough to prevent spurious actuations based on calculations consistent with Regulatory Guide 1.105. The deletion of the RIIR steam condensing mode isolation actuation instmmentation does not effect the margin of safety, because this mode is no longer utilized by LaSalle in Operational Conditions 1,2, or 3 (Run mode, Startup mode, or IIot Shutdown). He elimination of the temperature based trip functions for the RilR shutdown cooiing rnode area is based on the determination that temperature is not the appropriate parameter for leak detection as it does not provide meaningful indication and will not provide serpoints that would be sufficiently above the normal range of ambient conditions to avoid spurious isolations. ! here are two other methods of detecting abnormal leakage and isolating the system in Technical Specification trip function A.6, which are A.6.a, Reactor Vessel Water Level - Im, Level 3 and A.6.c, RIIR Pump Suction Flow - Iligh. In addition, other means to detect leakage from the RIIR system, such as sump monitoring and area radiation monitoring, are also available. Also, in accordance with Technical Specification C-4
_ _ _ _ _ = . _. . . - i-l l
. ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Administrative Requirement 6.2.F.!, LaSalle has a leakage reduction program to reduce leakage from those portions of systems outside primary containment that contain radioactive fluids. RIIR, including piping and components associated with the shutdown cooling mode, is part of this program, which includes periodic visual inspection of the system for leakage.
The previous evaluation of diversity ofisolation parameters, as presented in Table 5.2 8 of the UFSAR remains unchanged. Adequate diversity ofisolation parameters is maintained because there are at least two different methods available to detect and allow isolation of the system for a line break, as necessary. Therefore, these changes do not involve a significant reduction in the margin of safety. Therefore, based upon the above evaluation, Comed has concluded that these changes involve no significant hazuds consideration. I t l i C-5 t I
ATTACHMENT D ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW Comed has evaluated this proposed operating license amendment request against the criteria for identification oflicensing and regulatory acticns requiring environmental assessment in accordance with 10 CFR 51.21. Comed has determined that the proposed license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9) and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b). His determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that change an inspect;on or a surveillance requirement, and the amendment meets the following specific criteria: (i) the amendment involves no significant hazards consideration. As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards consideration. (n) there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite. As documented in Attachment C, there will be no change in the types or significant increase in the amounts of any effluents released offsite. (iii) there is no significant increase in ind;vidual or cumulative occupational radiation exposure. The proposed changes will not result in changes in the operation or configuration of the facility. %ere will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change. j 1 l I I I l D-1 j
b I J i 7 a 1 ATTACHMENT E
SUMMARY
OF SARGENT AND LUNDY (S&L) CALCULATIONS FOR REVISED HEAT LOADS FOR THE RWCU AREAS AFFECTED BY THE MODIFICATION THAT CHANGES THE RWCU SYSTEM FROM A COLD SUCTION TO A HOT SUCTION f E-1
i Summary of Sargent and Lundy (S&L) Calculations for Revised Heat Loads for the RWCU Asene Affected by the Modification that Changes the RWCU System imm a Cold Suction to a Hot Suction Cooling Load Requirements and Supply Air Flowrates for the RWCU Pump Valve Room and the North and South Pump Rooms; Cooling Load Requirements and Supply Air Flowrates for the RWCU Holdup Pipe Room; and Cooling Load Requirements and Supply Air Flowrates for the RWCU Heat Exchanger and Valve Rooms Summary. Sargent & Lundy (S&L) Calculations L-001325, L-001332, and L-001374 were revised to address both LaSalle Units 1 & 2. These calculations determine the revised heat loads for the areas affected by the modification that changes the RWCU system from cold suction to hot suction. Air flowrates necessary to maintain the room design temperature are calculated based on the new heat loads. %ese heat loads and air flowrates are used as input to the RWCU area ambient temperature and differential temperature (r and AT) leak detection calculation. It should be noted that the RWCU Filter /Demineralizer Valve Room was not affected by the change to hot suction for the RWCU pumps, so the original calculations for heat balance are not affected and were performed similarly to the discussion below. For piping heat loads, values for the heat loss per foot of pipe length are obtained through the use of S&L Program HEATLS Version 1.0 (Program No. 03.7.290-1.0). He input data for the program includes the pipe outer diameter, insulation thickness, ambient room temperature, pipe emissivity and the temperature of the fluid inside the pipe. The heat gain due to the heat loss from the piping, loads from equipment (such as the pump motor for the pump rooms and the heat exchangers for the heat exchanger rooms) and lighting are then calculated. Finally, based on the heat load and room design temperature, the required supply air flowrate to each of the rooms is calculated. Note that the " Mezzanine Area" as referred to in the calculations is the same as the RWCU IIoldup Pipe Room. The calculations were originally prepared for the piping, equipment, and room configurations for the Unit 1 areas. The revisions were prepared to incorporate the heat load and supply air flowrates for the corresponding Unit 2 areas. The piping, equipment, and room configurations for Unit 2 were compared to those of Unit 1 and no differences were noted that affect the calculated heat loads or air flowrates. Accordingly, the heat loads and supply air flowrates originally calculated for Unit 1 are confirmed to be applicable for Unit 2. l l E-2
r Summary of Results: $ t Tnp Function Top Function (Bru/hr) . CF) (SCFM) Ref. [ areas or rooms subareas or rooms ' Heat Load With Design Airflow ! Calc. No.' '[ Equipment in Temperature. . M _ RWCU Holdup Ppe 17,602 122 889 Ie001332 i (Mezzarune) ~ , RWCU Pump and Pump Valve Room 17,928 122 905 Pump Valve Room North Pump Room 33,840 122 1709 1,001325 South Pump Room 33,840 122 1709 RWCU Heat South Hz 34,142 122 1724 Exchanger Room North Hx 37,953 122 1917 'I 001374 South Valve 13,119 122 663 North Valve 13,189 122 666 E-3
e%a mkm- 4 L-4-. k%---e4-
- 4 I
ATTACHMENT F
SUMMARY
OF CALCULATION NO. I-001281 RWCU AREAS TEMPERATURE RESPONSE DUE TO HIGH ENERGY RWCU FLUID LEAKAGE l F-1 l
. ._ . . ._ .__- - __ =-- -
I l l Summary of Calculation No. Ir001281 l RWCU Aseas Temperature Response Due to High Energy RWCU Fluid Leakage f 1 Calculation 1 001281 was revised to address both LaSalle Unit 1 and Unit 2. It determines the l steady-state room temperatures and room differential temperatures for the Reactor Water Cleanup l (RWCU) system aren as a result of high energy fluid leakage. Based on these temperature responses, analytical limits for establishing leak detection instmment setpoints are determined. The Leak Detection system consists of ambient temperature as well as differential temperature sensors located in the IIVAC supply and retum air rystem for the RWCU system areas. In the event of a pipe failure in one of the rooms, a percentage of the high energy fluid released from l the pipe would fluh to steam. As a result, the ambient room temperature would quickly rise, setting off high temperature alarms in the control room. If the temperature continues to rise, the RWCU system is isolated by actuation of the inboard and outboard isolation valves, (G33-F001 and G33-F004) preventing additional flow into the system. l The calculation of the room temperature response for a given fluid leakage rate is based on solving steady-state equations for mass and energy balances for the room in question. The mass
- and energy of the leaking fluid is added to the inlet air stream and the thermodynamic properties l
of moist air are used to determine the exit air conditions. If the leaking fluid provides sufficient moisture to produce 100% relative humidity in the room, the properties of saturated air are used to define the exiting air temperature. The analytical limits are based on conservative analysis of l temperature responses for various operating scenarios (e.g., summer, winter, equipment opera'ing l or not operating, etc.). In addition, the potential for false alarms is minimized. The room air flowrates are based on Calculations 1 001325,1,001332, and 1 001374 which were revised to address both LaSalle Unit 1 and Unit 2. The results of this calculation are used as input into the leak detection temperature serpoint calculation (I,001420), 1 l 1he results of this calculation for Unit 2 are the same as the Unit 1 results. Table 1 on the i following page summarizes the results for the Normal (Zero Leak), Alarm (5 gpm) Analytical l Limit, and Isolation (25 gpm) Analytical Limit Cases. l l l l l-r i 4 F-2 i I
Summary of Results:
'Ihe analytical limits for alarm are determined for temperatures corresponding to i gpm leakage. 'Ihe analytical limits for isolation are determined for temperatures corresponding to a 25 gpm leak. Table 1 contains the results of Calculation 1 001281. These results are valid for both LaSalle IJnits 1 and 2.
Table 1 RWCU Leak Detection System Analytical Limita RWCU Leakage Analytical Limit Analytical IJmit Room (gpm) AT (*F) Thigh ( F) RWCU Pump Normal 0 18.0 122.0 Rooms Alarm 5 32.4 136.4 Isolation 25 94.0 212.0 RWCU Pump Normal 0 17.6 121.6 Valve Room Alarm 5 41.0 145.0 Isolation 25 94.0 212.0 RWCU Heat Normal 0 17.6 121.6 Exchanger Room Alarm 5 24.6 128.6 Isolation 25 41.8 159.8 RWCU Mezzanine Normal 0 17.3 121.3 Room Alarm 5 40.8 144.8 Isolation 25 94.0 212.0 RWCU F/D Normal 0 19.6 123.6 Valve Room Alarm 5 44.5 148.5 Isolation 25 94.0 212.0 l l 9 o F-3 i
au- _. +.au m,,-r. Aa,M - .,,4 _a . mu,a .4-- u- dE- 4 .A - - 4--A6A, ;A.- 4.- 4-+4-- -* e -
- A- Me+4-e.A s - J. .A aae & a -
ATTACHMENT G
SUMMARY
OF CALCUIATION NO. L-001324 AREA AMBIENT AND DIFFERENTIAL TEMPERATURE DESIGN BASIS CALCUIATIONS FOR REACTOR COOLANT LEAK DETECTION 4 l l i G-1
f Summary of Calculation No. I-001324, Area Ambient and Differential Temperature Design Basis Calculation for Reactor Coolant Leak Detection LaSalle County Station Units 1 and 2 have a Leak Detection (LD) system for pipelines containing reactor coolant that are outside of the primary containment. %e purpose of safety-related calculation 1,001324 is to define and document the methodology and calculations used in determining inputs for the LD temperature sensors. 'lhese inputs (analytical setpoints) consist of a determination of the theoretical room area ambient temperature (T) and differential room temperature rise (AT) due to pipe leaks of various sizes. The calculation contained an unverified assumption conceming the applicability of the results to Unit 2. Therefore, the calculation was revised in order to verify that the results were also applicable for Unit 2. %e verification process required the preparation of a new scoping document (AttachmentJ of the calculation), which identified the Unit 2 rooms that should contain a temperature based LD system. The conclusion of this review is that temperature based leak detection is appropriate for the same Unit 2 rooms as previously identified for Unit 1. In addition to summarizing the results of LD calculations contained in documents I 001281, RWCU Areas Temperature Response Due to High Energy RWCU Fluid Leakage; 3C7-1184-001, ECCS Room Leak Detection Serpoints; and BSA-I 95-05, LaSalle MS Tunnel Temperature Response Due to Steam Leakage with Ver.tilation System in Operation, analyses are included to determine LD analytical setpoints for the RCIC Pipe Chase p10*-0") and RCIC Equipment Room (673'-4"). Also, a check is included for the RWCU Pipe Tunnel 696'-0") to determine the pipe leak rate which would " shut off" the inlet air flow normally induced into the tunnel by the VR system exhaust fans. All calculations are performed using an enhanced version of an Excel spreadsheet initially developed for use in Calculation L-001281. The enhanced version includes additional analytical capability to' allow for the analysis of rooms for which the inlet air flow is held constant (e.g., the RCIC Pipe Chase), and rooms where the high pressure steam leaking from the pipe expands to superheated steam in the room (e.g., the RCIC Pipe Chase and RCIC Equipment Room). The results of the RCIC Pipe Chase and RCIC Equipment Room calculations are applicable to Unit 1 and Unit 2 and demonstrate that a 25 gpm leak will produce analytical ambient temperatures and differential temperatures which are higher than the current allowable isolation values in the Technical Specification. The calculations also show that a 5 gpm leak will produce temperatures greater than the current detection alarm setpoints. G-2
25 gym leak - 25 gym leak - 5 gym leak - 5 gym leak -
" Isolation", " Isolation", " Detection," " Detection,"
Ambient T . Diferential T Ambient T Difesential T Calc. Allow- Calc. Allow- Calc. Allow- Calc. Allow-able able able able RCIC Pye , 270 206 170 123 244 180 154 110 Chue RCIC 299 206 195 126 144 120 45 300 Equipment Room
%e RWCU Pipe Tunnel inlet and exhaust ventilation air flow rate is high,7650 SCFM. The existing temperature LD sensors are not located at the room ventilation discharge. Therefore, when a pipe leak occurs in the tunnel the sensors will not become effective until the pipe leak
. rate reaches a level which shuts off the induced ventilation air flow, and steam completely fills the tunnel. Le calculation is applicable to Unit 1 and Unit 2 and determines that a pipe leak rate of 92 gpm will cause the induced ventilation air flow to be ahut off. (This calculated leak rate is lower than the previously calculated value of f 66 gpm due to the enhanced calculation methodology.) Until the pipe leak rate reaches 92 gprn, the LD system in the RWCU Pipe Tunnel is believed ta be adequate for the following reasons:
he steam will be exhausted by the room ventilation system out of the ventilation stack.
%erefore, it is not released to any general plant areas. Some piping in the RWCU Tunnel contains fluid that has just been drawn from the reactor vessel with high radiation activity levels. he radiation monitors in the ventilation stack would identify a leak below 92 gpm in this piping in the RWCU Pipe Tunnel.
Redundant LD is required in " equipment areas." There is no safety related equipment in the RWCU Pipe Tunnel. This area only contains piping. Therefore, a pipe leak will have no significant effect on plant operation or safe shutdown. Some piping in the RWCU Tunnel contains water from downstream of the demineralizers which is being retumed to the feedwater system. Therefore, any radiation released due to a leak in this piping will be minor due to the reduced radiation activity level of the fluid at this location in the RWCU system, and will not affect off site doses.
%e room contains a sump monitoring LD system which will detect leak rates below 92 gpm.
G-3 U
,j_.h. _ -m. A A e , sLas.. .m u5" b *+d"- h *-M 4 ^--'O ^ * * ^ ' "* ' * ^ ' # * " * ~ ' b i ATTACHMENT H
SUMMARY
OF CALCULATION NO. 3C7-1184-001 ECCS ROOM LEAK DETECTION SETPOINTS 11-1
Summary of Calculation No. 3C7-1184-001, ECCS Room Leak Detection Setpoints Calculation 3C71184-001 determines the room temperature in the ECCS equipment rooms (RilR Pump Room A, RIIR Pump Room 13 &C, and RCIC Pump Room) for both LaSalle Units 1 & 2 due to high energy fluid leakage within the room. The calculations were originally performed for i the Steam Condensing Mode of operation. No chr.nge was made to this calculation as it already applies to both LaSalle Units 1 & 2. The steam condensing rnode of the RilR while in operating conditions 1,2, or 3, was deleted as reported in the Comed October 20,1992 letter. The UFSAR was updated accordingly at the time, however no Technical Specification change was submitted. The need to have consistent documentation results in this Technical Specification change request to delete the associated isolation actuation trip functions. Area temperature monitoring of the RIIR shutdown cooling mode lines is intended to provide assurance that important variables are monitored with a sufficient precision so that the isolation control system shall respond correctly to the sensed variables. Ilowever, calculations for temperature response show that the RIIR shutdown cooling area temperature and AT sensors are not affected significantly as a result of the leak because the shutdown cooling lines are moderate energy lines. Therefore, these variables are not the important variables that can be monitored with precision to assure correct response of the isolation control system. Ilecause the steam condensing mode of operation has been removed from the UFSAR there is no need to postulate high energ, une breaks during this mode. Therefore, the requirement for a leak detection system acociated with the steam condensing mode is being removed from the UFSAR, and the associated isolation trip functions should be eliminated from the LaSalle Technical Specifications. Also, the RIIR shutdown cooling lines are moderate energy lines. Leakage from moderate energy lines is not detectable by a T and AT leak detection system. Therefore, there is no technical basis to support the requirement in the LaSalle Technical Specification for a T and AT leak detection system in the'RIIR Rooms. As such, these isolation trip functions should be eliminated from the LaSalle Technical Specifications. l l 11-2
l l \ l l l 1 1 I 1 i ATTACHMENT I i
SUMMARY
OF CALCULATION NO. L-001384, REACTOR BUILDING ENVIRONMENTAL TRANSIENT CONDITIONS FOLLOWING RWCU AND RCIC HELBs CALCULATION t I-1
i
;Q I Su==ary of Calculation No. I 001384, Reactor Building Environmental Tra nsient Conditions following RWCU and RCIC HELBs Calculation Calculation I 001384, " Reactor Building Environmental Transient Conditions Followir. ; RWCU and RCIC HELBs," wu revised to address both LaSalle Unit 1 and Unit 2. his cal alation determines the short-term environmental conditions, specifically temperature and rein.ve humidity,in the general floor areas of the Reactor Building that result from postulao i RCIC and RWCU HELBs, and the short-term RCIC Turbine Room transient temperatures tis result from postulated RCIC HELBs, In addition, Calculation I 001384 determines the uppei .>ound of the analytical limit for the high RWCU flow instrumentation based on the system volumetric flow ;
rate which would result from a postulated guillotine pipe break in the RWCU system during i normal RWCU operatim. He calculation results indicated no impact to the existing RCIC Technical Specification setpoints. Therefore, this calculation summary only discusses the RWCU portion of the calculation. A postulated HELB for the purpose of Calculation L.001.284 is considered to be guillotine rupture of a pipe with the ends offset and assumed break opening time of 1 msec. Per FSAR l Appendix C, Section C.2, pipe breaks are postulated for lines of 4-inch diameter or larger. Mass l and energy releases from the postulated pipe breaks are determined based on the methodology ! l outlined in ANSI /ANS-58.2-1980. Mass and energy releases from selected pipe break locations we calculated by considering blowdown flow from both ends of the guillotine pipe break. The . l blowdown of the fluid downstream of the pipe break is determined based on removal of all fluid inventory between the pipe break and the first check valve downstream of the pipe break at a choked flow rate. %e initial blowdown of the fluid inventory upstream of the pipe break is determined based on the removal of all fluid inventory between the Reactor Pressure Vessel and the pipe break at a choked flow rate. Following the removal of the inventory upstream of the
- pipe break, a steady-state blowdown is calculated until the pipe break is isolated from the Reactor Pressure Vessel.
RWCU pipe breaks are postulated in each area of the Reactor Building through which the RWCU piping is routed. RWCU pipe breaks are considered to be detected by the high RWCU flow instrumentation. The time required to isolate the RWCU break from the Reactor Pressure Vessel is determined based upon the time required to accelerate the flow above the high RWCU flow setpoint limit, the RWCU flow instrument response time including a time delay incorporated to limit spurious isolations resulting from operational transients, and the time required to stroke the RWCU isolation valves closed. Bounding RWCU mass and energy releases based upon high RWCU flow detection and isolation are considered in each respective area of the Reactor Building containing RWCU pipe. A model of the Reactor Building is generated by dividing the Reactor Building into nodes, which represent rooms. The boundaries of rooms are based on features such as walls, doorways, blowout panels, hallways, and grating. The rooms are connected via a system of" room junctions" that provide paths for mass and energy to travel to surrounding rooms. The " room i junctions" consist of stairways, equipment hatches, doorways, hallways, open areas above walls 9 that do not extend to the top of the ceiling (wall-top slots), blowout paths, and otner open areas l providing a means of communication between rooms. Blowout paths are junctions such as doors I2
-, , , - ,- -- -- m --- gmn r
-. .- - - - - . - - . . - . - ~ - . . . --- -- .- . - - . -- .. .
). and blowout panels that are initially closed, yet are assumed fully open when the differential pressure across the junction is greater than its specified blowout linut. Flow between rooms via
. the room junctions are a function of the cross-sectional area available in each room for flow to travel (usually occurring at the actual junction between rooms), the effective inertia for each flow ,
path, and the hydraulic loss coefficients associated with flow from one room to the next. %e l heat loads associated with normal plant operations and heat transmission between Reactor Building rooms at different temperatures are included in the model. %e effects of steam
;~ condensing on the concrete and steel surface area are included. When the surface temperature of a heat sink is equal to or greater than the room temperature, a natural convection heat transfer is ;
4 considered. Finally, heat transfer via natural convection through various equipment hatches, stairway openings, hallways, and tunnels has been incorporated into the model during the post-4 IIELB period. i i Using the Reactor Building model developed and the mass and energy releases calculated for the l postulated RWCU and RCIC HELBs, transient analyses are performed using the COMPARE l . computer program. %is code was developed by los Alamos National Laboratory under contract l
- for the US NRC to perform subcompartment analysis. In addition, COMPARE is cited in Section 6.2.1.2 of the Standard Review Plan "Subcompartment Analysis
- " as the applicable tool used by the NRC for review of subcompartment analysis. COMPARE /MODT-PC supplements !
the original COMPARE / MOD 1 code to permit compliance with requirements of Appendix B of I J NUREG 0588 for environmental qualification calculations. He COMPARE model is used to j determine the room temperature and junctions flow rates. The COMPARE /MODT-It modelis run, and the results include the transient room temperatures and relative humidities for each 4 room in the model. From the results, the maximum room temperature and relative humidities . are determined for the various line breaks considered. A The results of this calculation indicate the following peak room temperatures: [ Room Peak Temperature (T) Break Location Main Room g61'0") 144 RWCU Pump Valve Room L (General Access Area) l Main Room G86'6")(General 143 South RWCU Heat 5 Access Area) Exchanger Valve Room Main Room - East Side 114 South RWCU Heat , (820'6")(General Access Exchanger Valve Room j Area) t Calculation 1-001384 also determines the upper bound high flow analytical limit for a guillotine I
- break of 4" or larger RWCU pipe. The high flow analytical limit is the minimum flow rate from !
a full guillotine pipe break during normal RWCU operation. This value is determined for normal ! J (Mode A) RWCU operation. This value is based on a process temperature of 534 'F, and a i reactor dome pressure of 1020 psia, his analytical limit is design input for determination of the high flor setpoint. He analytical limit is determined to be 650 gpm and includes a reduction of i l 5.13% to account for errors in the measured RWCU flow rate. 4 i- t I-3 1 l 4 1
- 5. ~ , ~ .-, , . . - . .., - - . , . . - - , ,.
l l
%e upper bound of the analytical limit for the RWCU high flow instrumentation is determined based on the system volumetric flow rate which would result from a postulated guillotine pipe break of the smallest size pipe postulated in the FSAR for the RWCU system during normal l RWCU operation. Larger breaks would result in higher flows which would be detected and l isolated. l I
Note that 650 GPM is approximately 85% greater than the normal operating flow rate of 352 GPM. His high flow analytical limit is used as input to the RWCU high flow setpoint calculation (1-001443). Calculation 1,001384 was initially prepued for Unit 1 and Unit 2, but contained unverified assumptions for Unit 2. %c revised calculation verifies that the results are valid and applicable for both LaSalle Units 1 & 2. 1 l 7 l l-4 L l I
l I l ATTACHMENT J
SUMMARY
OF CALCULATION NO. I 001420 RWCU AREA AMBIENT AND DIFFERENTIAL TEMPERATURE SETPOINT CALCULATION 3-1
i l Summary of Calculation No. Ie001420, RWCU Area Ambient and Differential ; Temperature Setpoint Calculation Calculation 1001420, "RWCU Room Serpoint Maigin Analysis," determines the serpoints and allowable values for the RWCU Room Differential Temperature and Ambient Temperature
- isolation functions. It also determines the calibration setpoints for the high RWCU Room Differential and Ambient Temperature alarms. This instrumentation detects a RWCU pipe leak i in the room based on a rise in the ambient or the differential temperatures in the room. Each room has differential and ambient temperature loops that will alarm when a 5 gpm leak occurs and isolate the RTCU system when a 25 gpm leak occurs. The room temperatures during !
normal operation,5 gpm leak, and 25 gpm leak conditions for both summer and winter, were determined in Calculation I 001281 "RWCU Areas Temperature Response Due to High Energy
)
RWCU Fluid 12akage," and are used in this calculation. This calculation revises the previous RWCU Room Serpoint Margin Analysis and Loop Accuracy , calculation submitted for Unit 1. The revision employs the same methodology but incorporates l the Unit 2 specific instrumentation. There are no differences between the units that impact diis l calculation. The results of Calculation 1 001420,"RWCU Room Setpoint Margin Analysis" conclude that the following alarm setpoints, isolation setpoints, and allowable values provide assurance that the Analytical Limit will not be exceeded for the Unit 2 RWCU Leak Detection System. i l J-2
AT 8 Aminent RWCU inst. No. Allowable Isolation Alarm Inst. No. Allowable Isolation Alarm Rooms -(2E31-) Value (*F) Setpoint Setpoint (2E31-) Value (F) Setpoint Serpoint CO CO CO CO Pump N600A,B,C, N601A,B, Rooms D 92.5 86 C,D 209 201 (A,B) R611B 28 R608B 131 Pump N600E,F 92.5 86 N601E,F 209 201-Valve Room R611B 37 R608B 140 HX N600G,H, N601G,H, Rooms J,K 40.3 33 J,K 156.8 149 (A,B) R611B 20 R608A 124 Holdup N621A,B 92.5 86 N620A,B 209 201 Pipe Area R611A 37 R608A 140 [ F/D No23A,B 92.5 86 N622A,B 209 201 Valve Room R611A 40 R608A 143 J-3
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i I l l ATTACHMENT K
SUMMARY
OF CALCULATION NO. L-001443 REACTOR WATER CLEANUP HIGH FLOW ISOLATION SETPOINT CALCULATION l l l l K-1
l Su===y of Calculation No. IA01420, RWCU Ases Ambient and Differential l Temperature Setpoint Calculation Calculation 1-001443," Reactor Water Cleanup High Row Isolation Setpoint Error Analysis," calculates the calibration setpoint and allowable value for the RWCU High Flow Switch. The I setpoint is selected based on detecting a postulated guillotine pipe break in the RWCU piping of l 4-inch or larger diameter during normal (Mode A) RWCU operation. Le RWCU Flow ba;ed I on a postulated guillotine pipe break in the RWCU piping was calculated by Calculation l I 001384," Reactor Building Environmental Transient Conditions following RWCU and RCIC HELB's" and is used by this calculation. This calculation (performed for Unit 1) was revised to incorporate the Unit 2 instrumentation and i employs the same methodology for both units. A difference exists between units for the flow element output. %is difference is accounted for in the scaling of the differential pressure transmitter and does not impact the flow setpoint determination. here are not other differences j
- between units for the RWCU High Flow Instmmentation.
l l The results of Calculation 1 001443, " Reactor Water Cleanup High Flow Isolation Setpoint Error Analysis," conclude that a RWCU High Flow Nominal Trip Setpoint of 560 GPM and Allowable Value of 610 GPM provides assurance that the Analytical Limit of 650 GPM will not be exceed for Unit 2. i I I f 4 i K-2 I}}