ML20205F018
| ML20205F018 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/13/1986 |
| From: | Miosi A COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 1979K, NUDOCS 8608190067 | |
| Download: ML20205F018 (39) | |
Text
,
/
\\ Commonwealth Edison
(
One First National Plaza Chicago,12nois
\\
Address Reply 12 Post Office Box 767 Chicago, Illinois 60690 0767 August 13, 1986 Mr. Harold R. Denton U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC. 20555
Subject:
Braidwood Station Units 1 & 2 Comments on Proof & Review Tech Specs NRC Docket 50-456 & 50-457
Reference:
June 10, 1986 J.A.
Stevens letter to D.L.
Farrar
Dear Mr. Denton:
Enclosed are typographical errors discovered during the review of the Proof and Review copy of the Braidwood Technical Specifications.
These errors are delineated in the Attachment 1.
Please include these in the Final Draft of the Technical Specifications.
Also included are the revised heatup and cooldown limitation curves included in Attachment 2 for Braidwood Unit 1 and Unit 2.
Please incorporate these in the Final Draft of the Technical Specifications.
Should you have any questions concerning tP.is matter please contact this office.
One signed original and fifteen copies of this letter and attachments are being provided for your review.
Very truly yours, l2 A.
D. Miosi Nuclear Licensing Administrator
/klj cc:
J. Stevens 1979K 8608190067 860813
[
PDR ADOCK 05000456 g
A PDR t
l
(.-.
ATTACHMENT 1 Typographical Errors Page #
2-3 2-7 B2-2 B2-7 3/4 3-30 3/4 3-29 3/4 5-3 3/4 6-26 3/4 7-16 3/4 7-17 3/4 8-40 3/4 8-41 3/4 8-42 3/4 8-43 3/4 8-44 3/4 8-45 3/4 8-46 3/4 8-47 3/4 9-2 3/4 9-5 3/4 9-14 3/4 11-3 3/4 11-16 3/4 12-12 B3/4 2-5 B3/4 2-6 B3/4 3-3 B3/4 3-5 B3/4 3-6 B3/4 4-9 B3/4 4-10 B3/4 11-1
'I I
--e-
_,-.e.,
--n_,-,- - - -, _,---,~,-
n.,-,--
,.,.,-m_..,- - - -, -.,,
n
PROOF AND REVIEW COPY SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent within the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY:
As shown for each channel in Table 3.3-1.
ACTION:
With a Reactor Trip System Instrumentation or Interlock Setpoint less a.
conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint
- value, b.
With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:
1.
Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> M quation 2.2-1 was satisfied for the affected
- m ipr G.'oIe.
2.
Declare the channel i rableandapp1MheappicableACTION statement requiremen of S;;cific:t4e# 3.3-1 u il the channel is restored to OPER LE status with its Setp 'nt adjusted I
consistent with the tpoint value.
Equation 2.2-1 Z + RE + SE 1 TA Where:
Z=
The value for Column Z of Table 2.2-1 for the affected channel, RE = The "as measured" value (in percent span) of rack error for the affected channel, j
SE = Either the "as measured" value (in percent span) of the sensor error, or the value for Column SE (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value for Column TA (Total Allowance) of Table 2.2-1 for l
the affected channel.
l f
BRAIDWOOD - UNITS 1 & 2 2-3 I
\\
e g
TABLE 2.2-1 (Continued) y TABLE NOTATIONS I
8 O
e NOTE 1: OVERTEMPER URE AT fl
[T (3 f g) - T'] + K (P - P') - f (al)}
I r*
g{
l(1, gag) $ AT,(K
-K AT 3
i g
4/
g Measured AT by RTD Manifold Instrumentation, Where:
AT
=
lead-lag compensator on measured AT,
=
3 Time constants utilized in lead-lag compensator for AT, r
=8s,
=
Ts. T2 12=3s, yf,3 Lag compensator on measured AT,
=
Time constants utilized in the lag compensator for AT, 13=0s,
=
T3 AT, Indicated AT at RATED THERMAL POWER,
=
1.164, K
=
i ;
0.0265/*F, K
=
2 N
I + **5 The function generated by the lead-lag compensator for T,yg E
=
3,
,3 dynamic compensation, g
i T4. Is Time constants utilized in the lead-lag compensator for T,yg, 14 = 33 s, E
=
15-4s, K3N Average temperature,
'F, T
=
1 1 + tsS Lag c spensator on measured T,yg, g
=
4 n
PROOF AND E1EW COPY SAFETY LIMITS I
ASES n
\\
~
j Y
\\
C' REACTOR CORE (Continued)
NY jj Fh=1.49[1+0.3(1-P)]
LL Where P is the fraction of RATED THERMAL POWER.
f 49
}7 lThese limiting heat flux conditions are higher than those calculated for
- w the range of all control rods fully withdrawn to,the maximum allowable control 3
rod i'nsertion assuming the axial power imbalance is within the limits of the vM
( I) function of the Overtemperature trip. When the axial power imbalance l
( = [ perature AT trips will reduce the Setpoints to provide protection c is not within the tolerance, the axial power imbalance effect on the Overtem-with core Safety Limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety limit protects the integrity of the eactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reacter coolant from reaching the containment atmosphere.
The reactor vessel, pressurizer, and the RCS piping, valves, and fittings are. designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.
The entire RCS is hydrotested at 3110 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
l i
1 BRAIDWOCD - UNITS 1 & 2 B 2-2
LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%
of full power equivalent), an automatic Reactor trip will occur if the flow in more than one/
rap helow 90% of nominal full loop flow.
Above P-8 (a power leve approximateTX 30% of RATED THERMAL POWER) an automatic Reactor trip will cur if the flow in any single loop drops below 90% of nominal full loop floi.
Conversely 3 on dec'reasing power between P-8 and P-7 an automatic Reactor Srio will occur on low reactor coolant flow in more than one loop and below P ' thetripfunctionipautomaticallyblocked.
/
Steam Generat'ot Water Leve'l The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater.
The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System.
Undervoltace and Underfrequency - Reactor Cnolant Pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant l
flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached.
For undervoltage, the delay is set so j
that the time required for a signal to cause a reactor trip after the Under-l voltage Trip Setpoint is reached shall not exceed 1.5 seconds.
For under-frequency, the delay is set so that the time required for a signal to cause a reactor trip after the Underfrequency Trip Setpoint is reached shall not exceed 0.6 second. On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of
)
approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.
BRAIDWOOD - UNITS 1 & 2 8 2-7
TABLE 3.3-4 (Continued)
TABLE NOTATIONS
" Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are
> 50 seconds and i< 5 seconds.
CHANNEL CALIBRATION shall ensure that ese time constant' are adjusted to these. values.
J
- The time constant utilized in the rate-lag controller for Steam Line Pressure -
Negative Rate - High is greater than or equal to 50 seconds.
CHANNEL CALIBRA-TION shall ensure that this time constant is adjusted to this value.
pu nu,nbacs de dd nek
(((l W.2 s d _Co e cdscee 0 4
BRAIDWOOD - UNITS 1 & 2 3/4 3-29
PROOF AND REVH COPY TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.
Manual Initiation a.
Safety Injection (ECCS)
N.A.
b.
Containment Spray N.A.
c.
Phase "A" Isolat. ion N.A.
d.
Phase "B" Isolation N.A.
e.
Containment Vent Isolation N.A.
f.
Steam Line Isolation N.A.
g.
Feedwater Isolation N.A.
h.
Auxiliary Feedwater N.A.
i.
Essential Service Water N.A.
1 j
Containment Cooling Fans N.A.
k.
Start Diesel Generator N.A.
1.
Control Room Isolation N.A.
m.
Turbine Trip N.A.
2.
Containment Pressure-High-1 1 7.(1)/12(5) 2 a.
Safety Injection (ECCS) 1)
Reactor Trip 52
$ (3) 7 2)
Feedwater Isolation 3)
Phase "A" Isolation 12(6) 4)
Containment Vent Isolation 17 5)
Auxiliary Feedwater 160 6)
Essential Service Water 142(1) 7)
Containment Cooling Fans 140(1) 8)
Start Diesel Generator 112 9)
Control Room Isolation N.A.
10)
TurbineTrip[
N.A.
3.
Pressurizer Pressure-Low a.
Safety Injection (ECCS) 127(1)/12(5) l 1)
Reactor Trip 12 1 (3) 7 2)
Feedwater Isolation 3)
Phase "A" Isolation 12(6)
A)
Containment Vent Isolation 17 BRAIDWOOD - UNITS 1 & 2 3/4 3-30
?
.---:~T~*
' ' r. L T -
r
EMERGENCY CORE COOLING SYSTEMS PRDOF AND REVIEW COPY 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two. independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
a.
One OPERABLE centrifugal charging pump, b.
One OPERABLE Safety Injection pump, c.
One OPERABLE RHR heat exchanger, d.
An OPERABLE f, low path
- capable o.f taking suction from the refueling water storage tank on a Safety Injection signal and automatic opening of the containment sump suction valves.
{
APPLICABILITY:
MODES 1, 2, and 3.
ACTION:
a.
With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.
The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
"During! MODE 3, tne di h
e paths of both Safety Injection pumps may be isolated by closing _I 8835 for a period of up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform surveil-lancet'astingasrequireo(bySpecification4.4.6.2.2.
I
\\
4 "' 's %
cC d HD l
)
N (gn BRAIDWC00 - UNITS'1s& 2 3/4 5-3
CONTAINMENT SYSTEMS PROOF AND CEEW COPY ELECTRIC HYOROGEN RECOMBINERS LIMITING CONDITIC1 FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE.
APPLICABILITY: MODES 1 and 2,
ACTION:
i With one Hydrog,en Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:
e a.
At least once per 6 months by verifying, during a Recombiner System functional test that the minimum heater sheath temperature increases to greater than or equal to 1200*F within 90 minutes.
Upon reaching 1200*F, increase the temperature controller to maximum setting for 2 minutes and verify that the power is greater than or equal to 38 kW, and b.
At least once per 18 months by:
1)
Performing a CHANNEL CALIBRATION of all recombiner instrumen-tation and control circuits, 2)
Verifying through a visual examination that there is no evidence 1
of abnormal conditions within the recombiners enclosure (i.e.,
loose wiring or structural connections, deposits of foreign materials,etc.),and 3)
Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test.
The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.
BRAIDWC00 - UNITS 1 & 2 3/4 6-26
PROOF AND REN copy PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) f.
After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the Emergency Makeup System at a flow rate of 6000 cfm + 10%; and g.
After each complete or partial replacement of a charcoal adsorber bank in the Emergency Makeup System by verifying that the cleanup system satisfies the in place penetration testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow rate of 6000 cfm 10%.
h.
At least once per 18 months or (1) after any structural maintenance on the charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the recirculation charcoal adsorber by:
(1) Verifying that the recirculation charcoal adsorber satisfies the in place penetration testing acceptance criteria of less than 2%
total bypass and uses the test procedure guidance in Regulatory Positions C.S.a and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 49,500 cfm i 10% for the recirculation charcoal adsorber; (2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample from the recirculation charcoal adsorber obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1% when tested at a temperature of 30*C and a relative humidity of 70%; and (3) Verifying a system flow rate of 49,500 cfm i 10% for the Recircula-tion Charcoal Adsorber when tested in accordance with ANSI N510-1980.
1.
After each complete or partial replacement of a charcoal adsorber bank in the Recirculation Charcoal Adsorber System by verifying that the cleanup system satisfies the in place penetration testing acceptance criteria of less than 0.1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating at a system flowrate of 49,500 cfm i 10%.
j.
After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of Recirculation Charcoal Adsorber operation by l
verifying within 31 days after removal, that a laboratory analysis of I
a representative carbon sample obtained in accordance with Regulatory i
Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 meets the laboratory testing, criteria of Regulatory Guide 1.52, Revision 2, March 1978 for a methylQodide penetration of less than 1% when tested l
at a temperature of 30*C}(3A) leMek is and a relative humidity of 70%.
b wer-AS) i i
BRAIDWOOD - UNITS 1 & 2 3/4 7-16 i
o
(
m._...
PROOF AND G1EW COPY PLANT SYSTEMS 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Three independent non-accessible area exhaust filter plenums (50%
capacity each) shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one non-accessible area exhaust filter plenum inoperable, restore the inoperable plenum to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.7 Each non-accessible area exhaust filter plenum shall be demonstrated OPERABLE:
a.
At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that operation occurs for at least 15 minutes; b.
At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the exhaust filter plenum by:
1)
Verifying that the exhaust filter plenum satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% when using the test procedure guidance in Regulatory Positions C.S.a. C.5.c and C.S.d of Regulatory Guide 1.52, Revi-sion 2, March 1978, and the flow rate is 66,900 cfm i 10% for the train and 22,300 cfm 210% per bank; 2)
Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample from each bank of adsorbers of the train obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52. Revision 2, March 1978, for methyl ddedid. penetration of less than 1% when tested at the temperature of 30'C and a relative humidity of 70%;
jedid4 BRAIDWOOD - UNITS 1 & 2 3/4 7-17
PR00f AND REYlEW COPY TABLE 3.8-2a MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES UNIT 1 VALVE NUMBER
~
FUNCTION 00G059 Unit 1 Suct Isol V1v H Recomb 2
00G060 Unit 1 Dischange.Isol Viv H2 Recombinee.-
00G061 Unit Dischaage.Xtie for H2 Recombinee-00G062 Unit Xtie on Dischaage of H2 Recombinee-00G063 Unit Suction Xtie for H Recombh r -
2 00G064 Unit Suction Xtie for H Recombiaece-2 00G065 08 H Analyzer Inlet Isol Viv 2
00G066 08 H Recomb Disch Isol Viv 2
.10G057A OA. H Recomb. Disch/ Isolf V:h:
'/11 10G079 H Recomb Disch/ dnet/ Isol Valvelf 2
10G080 H Recomb Suct/ Cnet,s Isol Velve '
2 10G081 H2 Recomb Suction. Cnsty I o[l
- 10G082 OA H2 Recomb Disch Cnat Iso Viv 10G083 OA H Recomb Disch Cnst Isol Viv 2
10G084 OA H Recomb Cnat Outlet Isol Viv 2
10G085 H Recomb Cnst Outlet Isol.Viv 2
1AF006A 1A AF Pp SX Suct Isol Viv 1AF0068 18 AF Pp SX Suct Dwst Isol Viv E
kc.k AFMtrDryhDisch[HdrDwstIsolViv 1AF013A 1AF013B AF Mtr Drv Pop eseW Hdr Dwst Isol Viv 1AF013C AF Ntr Drv Pp Disch Hdr Dwst Isol V1v 1AF0130 AF Ntr Dry Pp% Hdr Dwst l~sil Viv Disch Hdr Dwst Isol V1v -
AF Ds1 Dry -Pe
~
9 1AF013E 1AF013F AF Ds1 Drv Pp 4ssA,Hdr Dwst Isol Viv 1AF013G AF Ds1 Drv Pp DeeA.Hdr Dwst Isol V1v 1AF013H AF Ds1 Dry Pp M Hdr Dwst Isol Viv w :,a 1AF017A 1A AF Pp SX Suct Upst Isol Viv 1AF0178 18 AF Pp SX Suct Upst Isol Viv ICC685 RCP Thermal Bare +ee Outlet Hdr Cnmt Isol Viv ICC9412A CC to RH HX 1A Isol Viv 1CC9412B CC to RH HX 18 Isol Viv Cnd ICC9413A RCP CC Supply Dwst Isol Vb/
1CC94138 RCPs CC Supply Upst E Isol Vlv 1CC9414 CC Water from RCPs Isolf Velve.'/iv ICC9415 Unit 1 Serv / Loop Isol Viv E lfs 1CC9416 CC Wtr from 4GFFIsol/ Veh d Viv ICC9438 CC Wtr from OC ^... Ahermal Bar Isol/ Vehe VW ICC9473A Disch Hdr % Isol Viv ICC94738 Disch Hdr-X4$4(Isol Viv
'/.&
BRAIDWOCD - UNITS 1 & 2 3/4 8-40
PR00F AND REVIEW CO TABLE 3.8-2a (Continued)
MOTOR-OPERATED VALVES THERMAL OVERLOAD
~
PROTECTION DEVICES UNIT 1 (Continued)
VALVE NUMBER FUNCTION IC5001A 1A CS Pp suct from RWST 1B CS Pp Suct h from RWST IC50018 1A CS Pp % Disch Line Dwst Isol VyD*2"3 1CS007A 1C5007B 18 CS Pp '19. Disch Line fownstrear Isol V1v IC5009A 1 4 d 5 "Pr l' M g Suction from 1A Recirc Sump L.: b L3 1CS0098 1B CS Cent "-circ-5_- B-Svet-bel Vi v tc M Pp 1C5019A CS Eductor 1A Suct h Conn Isol Viv h e.
s...., f 1CS019B CS Eductor 1B Suct h Conn Isol Viv MOVVCT-OutletUpst[IsolVCTViv
- 1CV1128 1CV112C MOV VCT Outlet Dwstp Isol VCT Viv 1CV1120 MOV RWST to Chg Pp Suct Hdr ICV 112E MOV RWST to Chg Pp Suct Hdr ICV 8100 MOV RCP Seal Leakoff Hdr Isol 1CV8104 MOV Emerg Boration Viv s'tisch Hdr Isol Viv u
MOV Chrg-Pp/ Disch Hdr Isol V1v_'Q.
1CV8105 csD MOV % Pp Dd5*
1CV8106 1CV8110 MOV A & B Chgf.pprRecire Dewast Isol-gg 1CV8111 MOV A & B Chg Pp Recirc Wpetream lsol r
ICV 8112 RC Pump Seal Water Return Isol.'Velve-V b cnmt ICV 8355A MOVRCP1ASealInjInletto<ontehmen?Isol 1CV8355B MOV RCP 18 Seal Inj Inlet Isol 1CV8355C MOV RCP 1C Seal Inj Isol ICV 83550 MOV RCP 10 Seal Inj Isol Ci h,
s MOV RH/-Sys X-Tit. Vlv to Chrgmy Pump Sucth Hdr A/Bf 1CV8804A g
c.
IRH610 14 RH M 1RHG1PA Recire,' Line Isol," '/b 1RH611 13 RH PP' 1RHelPB Recire,* Line Isolf '/2,j ;.,
1RH8701A RC Loop 1A to RHJr" %,Isol,'Vrhea IRH8702A RC Loop IC to RHJr.,%Isol,'Velwf RC Loop 1A to RHR,%Isol/'Velve4 RC Loop IC to RHi[ Peop Qsol, Valve [ P 1RH8701B P
1RH87028 1RH8716A RH HX 1RH02AA DwnstenhIsoT Viv 1RH8716B RH HX 1RH02AB Dwntt Isol Velve.'ll1 77 wA 1RY8000A
/ReliefIsolfVe4*g_1A yl9 1RY80008 ftM/ Relief Isol/ Vah 18 7
ISI8801A SI ChuginpPump_Disch Isol Viv _ c,i,,I p
1518801B SI Chargia; n Disch Isol Viv o
ISI8802A SI PA 1A Disch Line Dwst Cont Isol Viv 15188028 SI M 1B Disch Line Owst Isol Viv i
$y BRAIDWOOD - UNITS 1 & 2 3/4 8-41
PRGDF AND REVIEW COPY TABLE 3.8-2a (Continued)
MOTOR-OPERATED VALVES THERMAL OVERLOAD
~
PROTECTION DEVICES l
UNIT 1 (Continued)
VALVE NUMBER
' FUNCTION 7
X6e 7
ISI88048 SI 18SuctM-tMfromR[HX Mrst f
15I8806 SI
";:t 01->Sucttorr Isol V W ISI8807A SI to Chg\\
Suct h Crosstie Isol Viv 15I88078 SI to Chg 'PP' Suetiee Crosstie Isol Viv ISI8808A Accump1ADisch/Isol,tVe%:
15I8808B Accum,t 18 Disch/ Isole V h Vtv ISI8808C Accum./ IC Disch/ Isol'>
15I88080 Accuse 10 Disch/ Isol/ M ISI8809A g g HX 1A esrA Line Dwst Isol Viv coe.k 15I88098 SI AW HX 18 eschiLine Dwst Isol Viv ISI8811A SI Cnet Sump A Outlet Isol Viv 15188118 SI Cnet Sump B Outlet Isol Viv ISI8812A W fr %SISI Awn to RH Pp 1A Outlet Isol Viv 15188128 to RH Pp 18 Outlet Isol Viv 1518813 SI Pumps IA-18 Recire Line Dwst Isol Hv 15I8814 SI Pump 1A Recire Line Isol Viv 15I8835 SI Pumps -Ft4(51sch 1 sol VIC
<--T so l 15I8840 SI RTHX Disch Line UpstroTo~nben 4*M V1v ISI8821A pp __Sl F 1A Disch Line F Isol Viv 15I88218 18 Disch Line Isol Viv
-re m
15I8920 SI 18 Recire Line Isol Viv ISI8923A SI & 1A Suct h Isol Viv ISI89238 SI hop 1B Suct Isol Velve! ;# Nd 9
15I8924 SI Pump 1A Suct h, %n Ownsted Isol Viv
$g ISX0168 RCFC B&D Sr Supply MOV ISX016A RCFC A&C SX Supply MOV,
1SX027A ACFC A&C Sturr.- C/, kht-m ISX0278 RCFC 8&D SX Return MOV OSX007 CC HX Outlet V1v e b N
~
OSX063A SX to Cogefrig Cdsr 0A OSX0638 SX to Cow Rm Refrig Cdsr 08 OSX146 CC Hx "0" Outlet 4
OSX147 CC Hx "0" Outlet I
BRAIDWOOD - UNITS 1 & 2 3/4 8-42
v PROOF AND REVIEW 00p TABLE 3.8-2a (Continued)
~
MOTOR-OPERATED VALVES TNERMAL OVERLOAD PROTECTION DEVICES UNIT 1 (Continued)
VALVE NUMBER
. FUNCTION l
OSX165A SX Train A Return Valve to Pond OSX16SB SX Train B Return Valve to Pond ISX001A 1A SX Pp Suct Viv MOV ISX001B 18 SX Pp Suct Viv MOV 1SX004 U-1 SX Supply to U-1 CCW HX MOV 1SX005 18 SX Pp Supply to O CCW HX MOV iSX007 CC HX Outlet Viv ISX010 U-1TrnAketurnV1vA8 X41 ISX011 Trn A Trn B Unit 1 Return-X-th V1v AB ISXO33 1A SX Po Disch-X-tit tiO L y( 4. ',,
1SX034 18 SX Pg Disch-X-th MOV ISX136 Unit 1 Trn 8 (eturn Viv A8 w fe Q Chilled wt.d ee..Is, 1WOOO6A He"1A & IC Supply Isol vi 1W0006B Chilled wte s*Me? 1B & ID Supply Iso 1 wiv-
'lI
1 WOO 20A Chilled wtr coMs 1A & IC Return Isol #1 IWOO20B Chilled wt e d 1B & 10 Return Isol 1 WOOS 6A Chilled Wete Cnat/ Isolf Y:he y Viv IWOOS6B Chilled Wet Cnst.' Isolf Valve-Wk BRAIDWOOD - UNITS 1 & 2 3/4 8-43
PROOF AND liEYlEW COPY TABLE 3.8-2b MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES UNIT 2 I
VALVE NUMBER
' FUNCTION 00G059 Unit 1 Suct Isol V1v H Recomb 2
00G060 Unit 1 Dischaage Isol Viv H2 Recombiner-00G061 Unit Dischaege Xtie for H2 Recombiaer.
00G062 Unit Xtie on Dischuge of H2 Recombines.
00G063 Unit Suction Xtie for H Recombiae*
2 00G064 Unit Suction Xtie for H Recombinees.
2 00G065 08 H2 Analyzer Inlet Isol Viv 00G066 08 H Recomb Disch Isol Viv 2
20G057A OB Hs Recomb Disch/-Isole V%
20G079 H
Recomb Disch.Cnet,s.Isol,eYr Viv
/
2 20G080 H Recomb Suct,* Cnat,s Isol e Whc.
2 20G081 H Recomb Suction Cnet,eIsol, eval 2
20G082 08 H Recomb Disch Cnet Isol Viv 2
20G083 08 H Recomb Disch Cnet Isol Viv 2
20G084 08 H Recomb Cnet Outlet Isol Viv 2
20G085 H Recomb Cnet Outlet Isol Viv 2
2AF006A 2A AF Pp SX Suct Isol Viv 2AF0068 28 AF Pp SX Suct Dwst Isol Viv Pr bis: ~
2AF013A AF Mtr Drv M DischjHdr Dwst Isol Viv 2AF0138 AF Mtr Dry PapescVHdr Dwst Isol Viv 2AF013C AF Mtr Dry Pp Disch Hdr Dwst Isol V1v 2AF0130 AF Mtr Dry Pp Disch Hde Dwst Isol Vh 2AF013E AF Ds1 Dry-WBsch Hdr 6wst Isol VTv~~ 9 2AF013F AF Ds1 Drv Pp DecS Hdr Dwst Isol V11 2AF013G AF Ds1 Drv Pp GschtHdr Dwst Isol Viv 2AF013H AF Ds1 Drv Pp OschjHdr Dwst Isol Viv t'i. _4 2AF017A 2A AF Pp SX Suct Upst Isol Viv 2AF0178 28 AF Pp SX Suct Upst Isol Viv 2CC685 RCP Thermal Bar 4ee!0utlet Hdr Cnmt Isol Viv 2CC9412A CC to RH HX 2A Isol Viv 2CC94128 CC to RH HX 28 Isol V1v de 2CC9413A RCP CC Supply Dwst sol Viv 2CC94138 RCPs CC Supply Upst GNW Isolm 2CC9414 CC Water from RCPs Isold \\ h ' lie 2CC9415 Unit 2 Serv A Loop Isol _V_1_
-- WW 3
2CC9416 CC Wtr from RGM(Tsolj Ve+ve.'/W 2CC9438 CC Wtr from P.C P d i? Thermal Bar Isol/ Valve 2CC9473A Disch Hdr X-tts Isol V1v 2CC9473B Disch Hdr X-tiMIsol Vlv
'b**
I BRAIDWOOD - UNITS 1 & 2 3/4 8-44 i
e
PROOF AND EIEW COPY TABLE 3.8-2b (Continued)
MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES UNIT 2 (Continued)
VALVE NUMBER
' FUNCTION I
2C5001A 2A CS Pp Suct from RWST 2C50018 28 CS Pp Suetio# from RWST 2CS007A 2A CS Pp 1!4 Disch Line Dwst Isol V12 h st 2C5007B 1B CS Pp 28 Disch Line 9ewstreenfIsol Viv 2A C5 P -2A-Puse Suctioe from-INRecire Sump ~ 2 A 2C5009A r
bd +"
.E, 2C5009B 28 CS Ge+Reekc Sep " Sect Isel V1# to-GS-77b"i7i 2C5019A CS Eductor 2A Suctiev Conn Isol V1v 2CS0198 CS Eductor 28 Suctte# Conn Isol Viv
['IsolVCTViv MOV VCT Outlet Upstp/ Isol VCT Viv 2CV1128 MOV VCT Qutlet Owst 2CV112C 2CV1120 MOV RWST to Chg Pp Suct Hdr 2CV112E MOV RWST to Chg Pp Suct Hdr 2CV8100 MOV RCP Seal Leakoff Hdr Isol 2CV8104 MOV Emerg Boration V1v s'Disch Hdr Isol Viv 2CV8105 Ch MGhrg Pp/Olsch Hdr Isol Viv h MOV Cheg-Pp bws-5 2CV8106 2CV8110 MOV A & 8 Chg/ Pp Recirc Dewatr Isol 2CV8111 MOV A & B Chg Pp Racire W u d.
r 2CV8112 RC Pump Seal Water Return Isolf Velve-Viv MOVRCP2ASealInjInlettocoed.s..[Cw+.
Isol 2CV8355A 2CV8355B MOV RCP 28 Seal Inj Inlet Isol 2CV8355C MOV RCP 2C Seal Inj Isol 2CV83550 MOV RCP 2D Seal Inj Isol 7
MOV RHg Sys X-f..i,t, Viv to CHOP 8mp Suctie# Hdr A/8/-
2CV8804A y
a 2RH610 RH Pf"i""01"A Recire, Line Isol/ 'li>
2RH611 RH M1" ;01?ir Recirc/ Line Isol, 'l3 '
2RH8701A RC Loop 2A to RHJt P m Isol./ Valve-Isol/ Valve-- 'llv RC Loop 2C to RHR %omp Isol,eVelve{
2RH8702A RC Loop 2A to RHR*P 2RH8701B 2RH8702B RC Loop 2C to RHR' Pump Is_ol/ Va,1_d p',
2RH8716A RH HX 2RH02AA Gwnet m Isol Viv RH HX 2RH02A8 Dweysg1 Mve.' liv 2RH87168 2RY8000A Tv-
,2 Relief 1s01) Vehe 2A Esa/ Relief Isolj Ve4ve g"' y[f r
2RY80008 2SI8801A SI therg % Fw,_Disch Isol V1v
~?7 25I88018 SI Charging-Pumgr Disen Isol Viv 2518802A SI @ 2A Disch Line Dwst Cont Isol V1v 25188028 SI -PP 28 Disch Line Dwst Isol Viv b
BRAIDWOOD - UNITS 1 & 2 3/4 8-45
PROOF AND TEEW COPY TABLE 3.8-2b (Continued)
MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES UNIT 2 (Continued) g i
VALVE NUMBER
. FUNCTION
{
'F yJ,e e
2SI88048 SI hoh28 Suct MQ from RH)b'X 25I8806 SI P e u SI to Chg}pstwam'Suctied Is01 'Ise
)
2SI8807A PP'Suctiew Crosstie Isoi '. v 2SI88078 SI to Chg PP' Suctiew Crosstie Isol Viv 2SI8808A Accum/ 2A Disctu Iso 1A Yr%
25188088 Accus/ 28 Dische Iso 1/ Wh: se 2SI8808C Accum/ 2C Disch/ Isolj Velve-1 25188080 Accum/ 20 Disch/-Isol/ Vew 2SI8809A g ~~SI At HX 2A Os A Line Dwst Isol Viv SA 25I88098 SI'RN HX 28 Osch Line Owst Isol Viv 2SI8811A SI Cnet $ ump A Outlet Isol V1v 25I88118 SI Cnet Sump 8 Outlet Isol Viv 2SI8812A SI N 5 p A to RH Pp 2A Outlet Isol Viv 2SI88128 SI Rws4-to RH Pp 28 Outlet Isol Viv 25I8813 SI Pumps 2A-20 Recire Line Dwst Iso 1Viv 2518814 SI Pump 2A Recire Line Isol Viv SI Pumps M-44e'Ufsch Isol V W M I
2SI8835 r
SI RHrHX Disch Line Upstre' Cont Pen 4s[Le 25I8840 V1v 2SI8821A SLfB 2A Disch Line Mlsol Viv 7? SI Pums 28 Disch Line X44ellsol V1v 25188218 fr n
25I8920 SI Pump 28 Recire Line Isol V1v 2SI8923A SI PP 2A Suction.Isol Viv 2 SIS 9238 SI Pump 28 Suct Isol Ve+ve'l',
, b w :.7 25I8924 SI Pump 2A SuctievX-tia Dwnstnr Isol V1v x %~
2SX0168 RCFC B&D SX Supply MOV 2SX016A RCFC A&C SX Supply MOV 2SX027A RCFC A&C SX Return NOV 2SX0278 RCFC B&D SX Return MOV OSX146 CC HX "0" Outlet VelveJ~
OSX147 CC HX "0" Outlet Velve-OSX165A SX Train A Return Vei to Pond OSX1658 SX Train 8 Return Ye to Pond 8RAIDWC00 - UNITS 1 & 2 3/4 8-46
- - - ~ - -
PROOF AND GEYlBV COPY TABLE 3.8-2b (Continued)
MOTOR-OPERATED VALVES THERMAL OVERLOAD
~
PROTECTION DEVICES UNIT 2 (Continued)
VALVE NUMBER FUNCTION 2SX001A 2A SX Pp Suct Viv MOV 2SX0018 28 SX Pp Suct Viv MOV 2SX004 U-2 SX Supply to U-2 CCW HX MOV 25X005 28 SX Pp Supply to O CCW HX MOV 2SX007 CC HX Outlet Viv 2SX010 U-2 Trn A Eeturn Viv AB M
2SX011 Trn A Trn B Unit 2 Teturn -X-tN Viv AB 2SX033 2A SX Pp Disch X-tie MOV 28 SX Pp Disch M-Me MOV 'p,'e 2SX034 2SX136 Unit 2TrnBEeturnVivAB
'.& d.d5 cotitI2A'& 2C Supply Isol %
\\/IV 2W0006A Chilled ce Ht/28 & 20 Supply Isol h 2W00068 Chilled 2 WOO 20A Chilled wtrs cotW 2A & 2C Return Isol vW 2 WOO 208 Chilled wtp ceHe 28 & 20 Return Isol v2W 2 WOOS 6A Chilled Wateg Cnet/Isol,t Valve, ytv 2 WOOS 6B Chilled, y Cnet/ Isoif V+1vd OSX0Q7 CC HX Outlet Viv_ A b l OSXO63A SX to Rm Refrig Cdsr 0A OSX0638 SX to Rm Refrig Cdsr OB BRAIDWOOO - UNITS 1 & 2 3/4 B-47 1
-. ~.
~
N00F AI,'D An'is ggpy7 REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION
~
LIMITING CONDITION FOR OPERATION 3.9.2 Asaminimum,twoSourceRangeNeutronFluxMonitorsshallbeOPERyBLEi each with continuous visual indication in the control room and one with aadible indication in the containment and control room.
APPLICABILITY: MODE 6.
ACTION:
a.
With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
b.
With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:
a.
A CHANNEL CHECX at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.
An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initial start of CORE ALTERATIONS, and c.
An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
BRAIDWJ00 - UNITS 1 & 2 3/4 9-2
P!!DOF MD liEVlEW Copy REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 4.9.4.2 Verifying that the fuel handling building exhaust filter plenums maintain the fuel building at a negative pressure of greater than or equal
.to.1/4. inch water gauge relative to the outside atmosphere with the equipment hatch removed, Prior to CORE ALTERATIONS or movement of irradiated fuel)and a.
b.
At least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel.
Verificatfon of negative pressure will be performed with systems in the normal REFUELING MODE.
BRAIDWOOD - UNITS 1 & 2 3/4 9-5
PRDOF A 3 00PY
~
REFUELING OPERATIO!<S t
1 3/4.9.12 FUEL' HANDLING BUILDING EXHAUST FILTER PLENUMS
~
LIMITING CONDITION FOR OPERATION
~
3.9.12 Two independent Fuel Handling Building Exhaust Filter Plenums shall be OPERABLE.
APPLICABILITY: Whenever irradiated fuel is in the storage po 1.
ACTION:
With one Fuel Handling Building Exhaust Filter Plenum inoperable, a.
fuel movement within the storage pool, or crane operation with loads over the storage pool, may proceed provided the OPERABLE Fuel Handling Building Exhaust Filter Plenum is capable of being powered from an OPERABLE emergency power source and is in operation and taking suction from at least one train of HEPA filters and charcoal adsorbers.
b.
With no Fuel Handling Building Exhaust Filter Plenums OPERABLE, suspend all operations involving movement of fuel within the storage pool, or crane operation with loads over the storage pool, until at least one Fuel Handling Building Exhaust Filter Plenum is restored to OPERABLE status.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
c.
SURVEILLANCE REQUIREMENTS
- 4. 9. 2 The above required Fuel Handling Building Exhaust Filter Plenums shall be demonstrated OPERABLE:
At least once per 31 days on a STAGGERED TEST BASIS by initiating, a.
from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes; b.
At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system, by:
BRAIDWOOD - UNITS 1 & 2 3/4 9-14
PROOF APID REVIEW COPY TABLE 4.11-1 (Continued)
RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(1)
TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml)
I7)
C7)
- 3. Continuous W
W Principal Gama 5x10 7 Release (5)
Grab Sample Emitters (3)
Essential Service Water Reactor C. Aim JAmrhm4 Fan Cooler (RCFC) i Outlet Line I-131 1x10.s Oissolved and Entrained Gases (Gama Emitters) 1x10.s H-3 lx10 5 t
BRAIDWOOD - UNITS 1 & 2 3/4 11-3
RADI0 ACTIVE EFFLUENTS PROOF AND REV!BV CDPY EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to ZE by volume whenever the hydrogen concentration exceeds 4% by volume.
APPLICA8ILITY:
At all times.
ACTION:
a.
With the concentration of oxygen in the WASTE GAS HOLOUP SYSTEM graatar than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
b.
With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 4% by volume and the hydrogen concentration greater than 4% by volume, inmediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 4% by volume, then take ACTION a. above.
c.
The prosistons of Spe:ifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentratlons of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM shall be determined to be within the above limits by continuously monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.,1r.
10 o
BRAIDWOOD - UNITS 1 & 2 3/4 11-16
TABLE 4.12-1 (Continued)
PROOF AND REVH3f COPY TABLE NOTATIONS (Continued)
It should be recognized that the LLO is defined as a before th'e fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement.
Analyses shall be performed in such a manner that stated LLDs will be achieved under i
routine conditions.
Occasiona ekground fluctuations, unavoidable small sample sizes, tIhe presen interfering nuclides, or other uncon-i trollable circumstances may ren these LLDs unachievable.
In such cases, the contributing factors shall be identified and described in the-Annual Radiological Environmental Operating Report pursuant to Specifica-tion 6.9.1.6.
(4) LLD for drinking water samples.
If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.
9 5
O BRAIDWC00 - UNITS 1 & 2 3/4 12 12
POWER DISTRIBUTION LIMITS N00F ED EN COPY BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continueo)
The' limit of 155 for Fh does not assume any specific uncertainty on the measured value of F An appropriate uncertainty of 4% (nominal) or greater g.
saddedtothemeasuredvalueofFhbeforeitiscomparedwiththerequirement.
i When an F measurement is taken, an allowance for both experimental error q
and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
The Radial Peaking Factor, F,y(Z) is measured periodically to provide assurance that the Hot Channel F (Z) remains within its limit. The F limit RTP) q for RATED THERMAL POWER (F as provided in Specification 3.2.2 wa determined from expected power control maneuvers over the full range of burnup conditions in the core.
I The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect flow degradation which could lead to operation outside the acceptable limit.
3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power dis-tribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts.
A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater i
than 1.02 but less than 1.09 is provided to allow ideritification and correc-tion of a dropped or misaligned control rod.
In the event such action doe not correct the tilt, the margin for uncertainty on F is reinstated by r uc e.
q ing the maximum allowed power by 3% for each percent of tilt in excess of For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incere l
flux map or two sets of four symmetric thimbles.
The two sets of four sym-metric thimbles is a unique set of eight detector locations.
These locations are C 8, E-5. E-11, H-3, H 13, L-5, L 11, N-8.
BRAIDWOOD - UNITS 1 & 2 B 3/4 2 5 i
- ~. - -,,.-.,-._ _ _
c.
1 t
l POWER DISTRIBUTION LIMITS PR00f AND REV1EMI CD.'Y 8ASES I
3/4.2.5 DN8 PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are saintained within the. normal steady-state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design DNBR throughout each analyzed transient.
The calculated tele::
l DNS-relatedparameterswillbeanaverageoftheindicatedvaluesfor/(ofthe he OPERABLE channels.
(,g 3
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
BRAIDWOOD - UNITS 1 & 2 8 3/4 2 6
PROOF AND REYlEW COPY _,
INSTRUMENTATION BASES Engineered Safety Features Actuation System Interlocks The Engineered Safety Features Actuation System interlocks perform the following functions:
Reactor tripped'- Actuates Turbine trip, closes main feedwater P-4 valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped.
Reactor not tripped -
events manual block of Safety Injection.
P-11 On increasing pressul 11 automatically reinstates Safety Injection actuation on low pre:
zer pressure and low steamline pressure and P
automatically blocks amlin isolation on negative steamline pressure
( V, ;#gy/
M e.
Un decr 6 sing pressur P-11 allows the manual block of Safety Injection low pressurizer pr ure and low steamline pressure and allows steamline isolation on negative steamline. pressure rate to er become active upon manual block of low steamline pressure SI.
P-12 On increasing reactor coolant loop temperature, P-12 automatically provides an arming signal to the Steam Dump System.
On decreasing reactor coolant loop temperature, P-12 automatically removes the arming signal from the Steam Dump System.
P-14 An increasing steam generator water level, P-14 automatically trips all feedwater isolation valves and inhibits feedwater control valve modulation.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MCHITORI:.G FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level sonitored by each channel reaches its Setpoint and (2) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance.
The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether i
or not predetermined limits are being exceeded.
If they are, the system sends actuation signals to initiate alarms and automatic actuation of Emergency Exhaust or Ventilation Systems.
The radiation monitor Setpoints given in the requirements are assumed to be values established above normal background radiation levels for the particular area.
Radia* ion monitors ORE-AR055 and 56 serve a dual purpose for plant operations as criticality and fuel handling accident sensors.
Although these monitors are designed primarily to detect fuel handling accident releases, they are capable of detecting an inadvertent criticality incident.
The Setpoint given in the requirement is estabitshed for the fuel handling building isolation function but is also adequate for an inadvertent criticality.
BRAIDWCC3 - UNITS 1 & 2 B 3/4 3-3
INSTRUMENTATION FR00F MD RM CON BASES SEISMIC INSTRUMENTATION (Continued)
The response spectrum analyzer computes the response spectrum of the event for two sensor locations, compares it to the design response spectra of the 1
plant, and indicates whether the event exceeded the operating basis earthquake criteria or the safe shutdown earthquake criteria.
This instrument'ation is consistent with the recommendations of Regulatory Guide 1.12. " Instrumentation for Earthquake," April 1974.
3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological. instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERA 81LITY of the remote shutdown instrumentation ens 0res that sufficient capability is available to permit shutdown and maintenance of HOT STANOBY of the facility from locations outside of the control room.
This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.
3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION t
The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.
This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 3. " Instrumentation
{
for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," May 1983 and NUREG 0737, " Clarification of TMI Action Plan Requirements," November 1980.
/
3/4.3.3 7 CHLORINE DETECTION SYSTEMS (g,# Q k ww.) L*$d a
g4 The OPERABILITY of the Chlorine Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release.
This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, Revision 1. " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," January 1977.
BRAIDWOOD - UNITS 1 & 2 B 3/4 3-5
00f Ul0 Eyg gag INSTRUMENTATION BASES m
3/4.3.,(XFIREDETECTIONINSTRUMENTATION l
7
\\
,j 0ERAB)bilityisavailableforthepromptdetectionoffiresandtha LITY of the detection instrumentation ensures that both adequate warning a
Fire. Suppression.Sys,tems,'that are actuated by fire det.ectors, will discharge extinguishing agents in a timely manner.
Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.
The zone designations given in Table 3.3-11 are from electrical schematic diagrams.
Fire detectors that are used to actuate Fire Suppression Systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.
Consequently, the minimum number of operable fire detectors must be greater.
The loss of detection capability for Fire Suppression Systems, actuated by fire detectors represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the M#eted areas is required to provide detection capability until the a
inoperanle \\nstrumentation is restored to OPERABILITY.
U8
\\
l 3/4.3.3 5 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capabilfty/is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components.
The allowable out-of-service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary system of Light-Water-Cooled Reactors," May 1981.
/ 1 l
3/4.3.$.10'RADI0ACTIVELIOUIDEFFLUENTMONITORINGINSTRUMENTATION
\\
/
Th y adioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid 4
effluents during actual or potential releases of liquid effluents.
The Alarm /
Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILI nd use of this instrumentation is consistent with the requirements of Gen 10 ign Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
/0 i
3/4.3..M A010 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION j
\\'
/
The_ radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.
The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted BRAIDWOOD - UNITS 1 & 2 B 3/4 3-6
REACTOR COOLANT SYSTEM N00F MD mn m BASES PRESSURE / TEMPERATURE LIMITS (Continued)
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.
In the calculation procedures a semi-elliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.
The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.
Therefore, the reactor opera-tion limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure.
To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNOT, is used and this includes the radiation-induced shift, ARTNOT, c rresponding to the end of the period for which heatup and cooldown curves are generated.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup g
or cooldown cannot be greater than the reference stress intensity factor, Kgg.
for the metal temperature at that time.
K is obtained from the reference ig fracture toughness curve, defined in Appendix G to the ASME Code. The K 7g curve is given by the equation:
Kgg = 26.78 + 1.223 exp (0.0145(T-RTNDT + 160)]
(1)
Where:
K is the reference stress intensity factor as a function of the metal IR temperature T and the metal nil-ductility reference temperature RT
- Thus, HDT.
the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
T CKgg + K K
(2)
IR Where:
Kgg = the stress intensity factor caused by membrane (pressure)
- stress, T
K
= the stress intensity factor caused by the thermal gradients, g
4 BRAIDWOOD - UNITS 1 & 2 8 3/4 4-9
l PROOF AND PRIEW COPY REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) l Kgg = constant provided by the code as a function of temperature relative to the RTNOT. f the material, C=
2.0 for level A and B service limits, and C=
1.5 for inservice hydrostatic and leak test cperations.
At any time during the heatup or cooldown transient, K is determined by yg the metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference fracture toughness curve.
The thermal stresses resulting from temperature graoients through the vessel wall are calculated j
and then the corresponding thermal stress intensity factor, KIT, for the reference flaw is computed.
From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
C00LDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall.
Duiing cooldown, the controlling location of the flaw is
)
always at the inside of the wall because the thermal gradients produce tensile J
stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.
From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.
During cooldown, the 1/4T vessel location is at a higher temperature than the fluid edjacent tc the vessel ID.
This condition, of course, is not true for the steady-state situation.
It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T location for finite IR cooldown rates than for steady-state operation Furthermore, if conditions 7
exist such that the increase in K exceeds K the calculated allowable IR pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct cortrol on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp.
The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-10
PROCF AND REVU COPY 3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION' This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2.
This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within:
(1) the Section II.A design objectives of Aooendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population.
The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
This specification applies to the release of radioactive materials in liquid effluents from all units at the site.
The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).
Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J.
K., " Detection Limits for Radicanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sec-tions II.A, III.A and IV.A of Appendix I, 10 CFR Part 50.
The Limiting Condi-tion for Operation implements the guides set forth in Section II.A of Appen-dix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculation methodology and parameters in the ODCM implement the require-7 ments in Section III.A of Appendix I that conformance with the g" Ma t b b. M <
BRAIDWOOD - UNITS 1 & 2 8 3/4 11-1
i ATTACHMENT 2 Heatup/Cooldown Curves Page #
3/4 4-33 3/4 4-34 3/4 4-35 3/4 4-36
MATERIAL PROPERTY BASTS Controlling Matirid : Wald Metal Copper Content
- Conservatively assumed to be 0.08 WIS (actual content = 0.04 WTS)
Phosphorus Content
- 0.R15Wr5 Initial RT
- 40 7 g
RT After 32 EFFY
- 1/4T, 138 L
- 3/4T, 84 7, Curve applicable for heatup rates up to 100*F/hg for the service period up to 32 EFFY and contains margins of 107 and 60 psig for possible instrtanent errors I I I I I I I f i 1 1 I r
i jI iiiI I I I I I I
(
i LEAK TEST LIMIT
/
j j
x 2250
'l l
l
}
l l
\\
l i
I l
1 2000
/
/
l l
HEATUP RATES UP
/
/
- 973, TO 1007/HR N
/
/
2.)
l
~
p 1500
/
/
i E
tis 5CCEPTABLE t,
OPERATION
[
[
w 1250 l
l k
/
/
ACCEPTABLE E
OPERATION
/
1 1000
/
/
O 750 J
' CRITICALITY LIMIT j
f BASED ON INSERVICE
~'
HYDROSTATIC TEST
)
f Soo TEMPERATURE (283 F)
FOR THE SERVICE PERIOD 3
UP TO 32 EFPY l
250 l
l l
0 30 100 150 2o0 250 3o0 350 doc 450 SoO tunicarco itwtnatung (oco.r)
FIGURE 1./JBraidwood Unit 1 Reactor Coolant System Heatup Limitations Applicable up to 32 EFPY l
3h 4-0
t MATERIAL PROPERTY BASIS 3 Controlling Material : Weld Metal g
Copper Content
- Conservatively asstaned to be 0.08 W5 1
(actual content = 0.04 WI5) l 0.01(WI5 Phosphorus Content Initial RT 40 t g
- 1/47, 109 Eyr
- 3/4T, 71
]
Curve applicable for heatup rates up to 100*F/ht or the service f
period up to 16 EFFY and contains margins of 107 and 60 psig for possible instrument errors 2500 7 r
g,gg i
2250 l
f f
l l
i i
i 2000 r
r HEATUP RgTES UP
/
/
j 1750
,TO 100 F/HR N i,
i x
I i
I I
7 1500 UNACCEPTABLE
/
/
E OPERATION r
1
)
J w 1250 ACCEPTABLE l
/
/
/
/
OPERATION
- w t 1000 l
l l
S O
750
/
CRITICALITY LIMIT BASED ON INSERVICE
/
HYDROSTATIC TEST TEMPERATURE (254 F)
FOR THE SERVICE PERIOD
)
250 00 SO 100 150 200 250 300 350 400 450 500 inoicarso itwcnatung (oco.r)
FIGUREJ.4-14Braidwood Unit 2 Reactor Coolant System Heatup Limitations Applicable up to 16 EFPY M 1-n
19 MATEDTAL PROPERTY BASTS 3
?
Controlling Material : Weld Metal Conservatively asstaned to be 0.08 WT5 Copper tentent (actual content = 0.04 WTS)
Phosphorus content
- 0.Q15WT5 e
Initial RT
- 40 7 g
- 3/4T, 84 7 Curves applicable for cooldown rates up to 100 (/hr for the service period up to 32 EFPY and contains margins of 107 and 60 psig for possible instrtanent errors 2500 l
r 1250 j
i
?
2000
)
~
1 f
1750 NACCEPTABLE OPERATION
/
g gg
~
r E
/
w 1250 B
C
/
{tooo
/
ACCEPTABLE OPERATION s
8 A
$ 750 @/HR' N' RATES S
C
-7 ffj/
9 in'ii sm/
S00 0.
', $[Y
_, - 20
- Qp', '
- ". 4 0 -
60 C 250
- 100 '
l s
O SO 100 150 2o0 250 300 350 400 450 300 llelCATED 1EtrER ATURC (DCO.F)
Braidwood Unit 1 Reactor Coolant System Cooldown 1. imitations FIGURE M.(Applicable up to 32 EFPY
}/Y 4-35
MATEMAL PROPERTY BASIS Controlling Noterial : Weld Metal Copper Content
- Conservatively asstaned to be 0.08 WIS (actual content = 0.04 VI5)_.
0.01gWIS
~
Phosphonas Centent Initial RT 40 F g
1/4T,109%
RT After 16 EFFT g
- . 3/4T, 717 Curves applicable for cooldown rates up to 100I/hr for the service period up to 16 EFFT and contains nargins of 107 and 60 psig for possible instnment errors 2500 l
I 2250 I
l 2M i
i I
1730 UNACCEPTABLE j
OPERATION, l
@ 1500 7
h
/
w r
w 1250 8
l C
I
' 1000
/
ACCEPTABLE S
e OPERATION O
7hS @r/HR C
N RATES
+
ti jf 4
m g
- Om A
Soo - 20 2 f,97
~ 40 a
/ ~ V' 250 100 -
-~
~
i' 0O SC 100 150 200 250 300 350 400 450 500
.lMDICATED TEwtRATURE (DES.7)
FIGURE M-!fBraidwood Unit 2 Reactor Coolant System Cooldown Limitations Applicable up to 16 EFPY 3/y y-yg