ML12285A221
| ML12285A221 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/26/2012 |
| From: | Nelson J, David Reeser Operations Branch III |
| To: | Exelon Generation Co |
| Shared Package | |
| ML11354A119 | List: |
| References | |
| Download: ML12285A221 (12) | |
Text
DEOP 100 RPV CONTROL Dresden Nuc ear Power Station Uni s 2(3)
EMERGENCY OPERA NG PROCEDURE RPV CONTROL DEOP 100 1 0 T t e Number Rev CATEGORY 1 Scram.
A l rods in to at least 0
?
No / Unknown Yes EN RY CONDI IONS FAILURE O SCRAM:
Exit this procedure En er DEOP 00-5
£ Yes Wi l the reactor s ay shutdown under all cond tions without boron
?
No / Unknown En er DGP 2-3 wh le continuing here.
£ EN ER SCRAM PROCEDURE:
LEVEL Verify needed auto act ons:
- Isola ions
- ECCS s art
Ex t this procedu e Enter DEOP 00-1 at h
£ Cannot stay be ow Fig D, Primary Containment P essure Limit Stop inject on from ou s de the primary containment not needed for core cooling.
CAU ION: Exceed ng NPSH/Vortex Limits (Figs V Y) may cause system damage Control RPV wa er evel between 8 in. and 8 in. using any of the systems isted below:
Check Deta l A o ind out wh ch inst umen s to use.
IF HEN Cannot estore level above 8 in.
and hold it there Hold level above -1 3 in. (TAF).
Use Alterna e Injec ion Sys ems f needed De a l E, DEOP 500 3).
Cannot ho d evel above -59 in.
Inhib t ADS.
To 2
- Condensate/Feedwater
- HPCI CAU ION: Exceeding 140°F suc ion tempera ure may cause system damage Use CST suct on f you can. OK to de eat high torus level t ansfer (DEOP 500-2).
OK to defeat high area temperature isolation (DEOP 500-2).
IF HEN RPV water level ca be restored and held above -1 3 in. (TAF)
RPV water level unknown Exit this procedure Enter DEOP 00 1 at h.
£ Cannot s ay below F g D, Primary Containment Pressure Lim t Go to 1.
FLOOD HE RPV:
Stop injection from outside the primary containment not needed for core coo ing.
To 1 In tiate IC.
Preferred Inject on Sys ems Inhibit ADS.
2 or more subsystems (Deta l F) ava lable
?
No Yes Start lining up all A ternate Injection Systems (De ail E, DEOP 500-3) wh le continuing here.
£ WAI until evel drops to
-1 3 in.
(TAF)
Any subsystem (Deta l F) ined up with a pump running
?
No Yes Maximize inject on with A terna e Injec ion Sys ems (Deta l E, DEOP 500-3).
BEFORE RPV water evel d ops to
-16 in.
(Blow down) inject on source lined up w th a pump running
?
No Yes S EAM COOL:
Ex t th s procedure En er DEOP 00-3
£ BLOW DOWN:
Enter DEOP 00-2 whi e continuing here.
£ Injection Subsystems F
- Condensate
- Core Spray Pump A
- Core Spray Pump B
- LPCI IF HEN D ywell pressure above 2.0 psig Befo e RPV pressure drops to 350 psig, prevent Core Spray and LPCI inject on not needed for core coo ing.
Ant cipa e RPV Blowdown DEOP 00-2)
Depressur ze RPV rap dly using IC and main turbine bypass valves.
OK to exceed 100°F/hr cooldown.
Reactor power above IRM Range 7 AND steady or increasing FA LURE O SCRAM:
Ex t this procedu e Enter DEOP 00-5
£ PRESSURE Stabi ize RPV pressure be ow 1060 ps g using main turbine BPVs.
Use A te nate P essure Control Sys ems ( isted be ow) if needed.
C
- ADSVsonly if orus water level s above 6 ft.
Use prefe red sequence if you can: A, C, E, D, B f D ywe l Pneumat cs lost, place the Ta get-Rock swi ch (203-3A) in AUTO.
- HPCI CAU ON: Exceeding NPSH/Vortex Limits (Figs V Y) or 140°F suction emperature may cause system damage Use CST suction if you can.
- Max Recycle Reboi er
- G and Seal Steam
- Of gas preheater C vent valves
- Main steam ine d ains
- RWCU, recirculation mode Bypass demins.
OK to de eat so at ons, including SBLC (DEOP 500-2).
- RWCU, b owdown mode A ternate Pressure Control Systems Depressurize the RPV using main turbine BPVs.
Hold coo down rate be ow 100°F/hr.
Use A ternate Pressure Control Systems
( isted above) f needed.
If Drywell Pneumat cs ost, minim ze use of the Target-Rock ADSV (203-3A).
IF HEN Cool down to cold shutdown using Shutdown Coo ing.
Hold cooldown ra e below 100°F/hr.
Shutdown Coo ing does not work Ho d recirculation suct on temperature below 350°F using A te na e P essure Control Sys ems.
WAI until Shu down Cooling high temperature interlock clears (reci cula ion suction temperature below 350°F).
Cannot estore level above
-1 3 in. (TAF) and hold it there Go to 2 Any ADSV cycling
?
No Yes Init ate IC and open ADSVs to lower RPV p essure o 9 5 psig.
CAU ION: Exceeding NPSH Vortex L mits (F gs V Y) may cause system damage Control RPV water level above -1 3 in. (TAF) using any of the Pre erred Injection Systems ( isted below).
Use A ternate Inject on Systems if needed (Detail E, DEOP 500-3).
- Condensate Feedwater
- HPCI CAU ION: Exceeding 140°F suction temperature may cause system damage Use CST suction if you can. OK o defeat high o us evel transfer DEOP 500-2).
OK to defeat h gh area empe ature solat on DEOP 500-2).
IF HEN Cannot restore level above -16 in.
and hold it there AND Neither Core Spray loop flow is at or above 750 gpm
- 1. Maximize Injection using Preferred a d Alternate Injection Systems (De a l E, DEOP 500-3).
you st ll cannot estore level above -16 in. and hold it there, TSC/EOF has command, FLOOD CON A NMEN :
- 2. IF.......
AND....
THEN..
Exit all DEOP F owcharts Enter all SAMGs
£ Cannot restore level at or above
-190 in. and hold it there CAU ION: Exceeding NPSH/Vortex Limits (Figs V Y) may cause system damage Control RPV water level above -1 3 in. (TAF) using any of the Preferred Injection Systems (l s ed above).
Use Al ernate Injection Systems f needed De a l E, DEOP 500 3).
Alternate Injection Systems E
- Standby Coo ant Supply
- SBLC bo on tank
- SBLC test tank
- ECCS Keep Fi l
- CRD crosst e
- Reactor head cooling
- Service un t back lush
- Condensate Transfer
- Fire Sys em
- LPCI crosst e
- Core Spray or LPCI pump with CST suc ion D
Primary Containment Pressure Limit 0
10 20 30 0
50 60 70 80 90 100 93 Primary Containment Water Level ( t)
Torus Bo tom Pressure (ps g) 70 60 50 0
30 20 10 0
0 5000 10,000 15 000 20 000 25,000 30,000 35,000 Total ECCS Flow (gpm)
W de Range Torus Water Level ( t) 11 10.5 10 9.5 9
V ECCS Vortex Limit C Minimum Usable Indicating Levels Fuel Zone
( 60 o -3 0 )
-297
-297
-298
-299
-300
-301 32 to 100 101 to 200 201 to 300 301 o
00 01 to 500 501 o
558 Drywe l temperature (°F)
TR 2(3)-13 0 1, Point 9 or 10 Medium/Nar ow Range A l Rx B dg Temps
( 60 o -60 )
Medium/Nar ow Range Rx Bldg Temps 181°F or less
( 60 o -60 )
Wide Range
( 330 to -70 )
-38
-39
- 1
- 3
- 3
- 3
-60
-60
-60
-60
-60 60
-68
-51
-21 17 68 107 Drywell Temperature (°F)
TR 2 3)-13 0-1, Point 9 or 10 OR Reactor Bu lding Temperatu e (°F) 600 500 400 300 200 0
100 200 300 400 500 600 700 800 900 1000 1100 RPV Pressure (ps g)
B RPV Saturation emperature A
RPV Water Level Instruments CAU ON: RPV water evel nstruments may be unre iable due to boi ing in he ns rument runs when drywell or reactor bu lding temperatures near the instrument runs are above Fig B RPV Saturation emperature.
An RPV water level instrument may not be used if the instrument reads at or below its Minimum Usable Level (Detail C).
W LPCI / Core Spray NPSH Limit ECCS Flow Up o 10 750 gpm 0
1000 2000 3000 000 5000 6000 LPCI / Core Spray Pump Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3 5 ps g 5 ps g 10 ps g 15 psig Torus Bo tom P essu e (ps g)
X LPCI / Core Spray NPSH Limit ECCS Flow Up o 25 500 gpm 0
1000 2000 3000 000 5000 6000 LPCI / Core Spray Pump F ow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3 5 ps g 5 ps g 10 ps g 15 psig Torus Bottom Pressure (ps g) 0 1000 2000 3000 000 5000 6000 7000 HPCI Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3 5 ps g 5 psig 10 psig 15 ps g Torus Bot om Pressure (psig)
ENTRY CONDITIONS D
PRIMARY CONTAINMENT PRESSURE IF THEN Cannot hold pressure below 2.0 psig Go to t.
BEFORE Drywell pressure reaches 9 psig
£ CAUTION: Exceeding LPCI NPSH/Vortex Limits (Figs V-X) may cause system damage.
- 1. Trip all recirc pumps.
- 2. Trip all drywell cooling fans.
- 3. Start drywell sprays.
Do not use pumps needed for core cooling.
OK to use external spray sources if you can restore and hold torus bottom pressure and primary containment water level inside Fig D, Primary Containment Pressure Limit.
Reducing primary containment pressure affects margin to NPSH limits. Check Figs W-Y, NPSH Limits.
IF Torus sprays running THEN Before torus pressure drops to 0 psig, stop torus sprays.
t (Torus spray)
Drywell sprays running Before drywell pressure drops to 0 psig, stop drywell sprays.
Torus water level
?
At or above 27.5 ft Below 27.5 ft.
CAUTION: Exceeding LPCI NPSH/Vortex Limits (Figs V-X) may cause system damage.
Start torus sprays.
Do not use pumps needed for core cooling.
OK to use external spray sources.
Reducing primary containment pressure affects margin to NPSH limits. Check Figs W-Y, NPSH Limits.
WAIT until drywell pressure is above 9 psig Below Fig K, Drywell Spray Initiation Limit
?
No Yes IF THEN Keep trying to lower drywell and torus pressures below 9 psig.
Cannot stay inside Fig L, Pressure Suppression Pressure Go to y.
BEFORE Torus bottom pressure reaches Fig D, Primary Containment Pressure Limit (Vent)
Vent to stay below Fig D, Primary Containment Pressure Limit (DEOP 500-4).
OK to exceed release rate limits.
y DRYWELL TEMPERATURE IF THEN Hold drywell temperature below 160°F using drywell cooling.
Drywell temperature affects RPV water level indication.
Check Detail A.
Cannot hold drywell temperature below 160°F Go to i.
BEFORE Drywell temperature reaches 281°F
- 1. Scram.
- 2. ENTER RPV CONTROL:
Enter DEOP 100 while continuing here
- 3. BLOW DOWN:
Enter DEOP 400-2 while continuing here i
£
£ (Drywell spray)
TORUS TEMPERATURE IF THEN Hold torus bulk temperature below 95°F using torus cooling.
Cannot hold torus bulk temperature below 95°F Go to o.
Start all available torus cooling
Do not use pumps needed for core cooling.
BEFORE Torus bulk temperature reaches 110°F IF THEN Cannot hold torus bulk temperature below Fig M, Heat Capacity Limit o
(Scram, Enter DEOP 100)
- 1. Scram.
- 2. ENTER RPV CONTROL:
Enter DEOP 100 while continuing here.
£ Hold torus bulk temperature below Fig M, Heat Capacity Limit.
- 1. IF.........
THEN..
you are not in RPV Flooding (DEOP 400-1) or Steam Cooling (DEOP 400-3),
lower RPV pressure to stay below Fig M, Heat Capacity Limit.
OK to exceed 100°F/hr cooldown rate.
- 2. IF.........
THEN..
you still cannot stay below Fig M, Heat Capacity Limit, BLOW DOWN:
£ Enter DEOP 400-2 while continuing here.
CAUTION: Exceeding LPCI NPSH/Vortex Limits (Figs V-X) may cause system damage.
TORUS WATER LEVEL IF THEN Hold torus water level between -4.5 in. and -1.5 in. narrow range.
Sample the torus before discharging water.
Low Level: Cannot hold level above -4.5 in.
Go to s.
High Level: Cannot hold level below -1.5 in.
Go to a.
a s
High Level (Above -1.5 in.)
HYDROGEN IF THEN Monitor hydrogen and oxygen concentrations in drywell and torus (DOP 2400-1).
Hydrogen or oxygen monitor is unavailable Sample the drywell and torus for hydrogen and oxygen.
Hydrogen (at or above 1%) or oxygen (at or above 5%) is detected in the drywell or torus
£ Hold drywell and torus pressures below 2.0 psig using SBGT and drywell purge (DOP 1600-1).
BLOW DOWN:
Enter DEOP 400-2 while continuing here.
IF Drywell sprays running THEN Before drywell pressure drops to 0 psig, stop drywell sprays.
Below Fig K, Drywell Spray Initiation Limit
?
No Yes IF THEN Keep trying to lower drywell temperature below 160°F.
Drywell temperature affects RPV water level indication.
Check Detail A.
Cannot restore drywell temperature below 281°F and hold it there Low Level (Below -4.5 in.)
ENTER HYDROGEN CONTROL:
Enter DEOP 200-2 while continuing in other sections of this procedure.
Hydrogen or oxygen concentration in the drywell or torus is unknown
£ ENTER HYDROGEN CONTROL:
Enter DEOP 200-2 while continuing in other sections of this procedure.
Start all available drywell cooling
OK to defeat drywell cooling isolations (DEOP 500-2).
IF THEN Hold torus water level above 12 ft. (HPCI exhaust).
Cannot hold level above 11 ft.
(downcomers)
- 1. Scram.
- 2. ENTER RPV CONTROL:
Enter DEOP 100 while continuing here
- 3. BLOW DOWN:
Enter DEOP 400-2 while continuing here Trip HPCI.
Even if core cooling will be lost.
Cannot hold level above 12 ft.
(HPCI exhaust)
IF THEN Hold torus water level below 18.5 ft (ring header).
Cannot hold level below 18.5 ft (ring header)
- 1. Scram.
- 2. ENTER RPV CONTROL:
Enter DEOP 100 while continuing here
- 3. Stop injection from outside the primary containment not needed for core cooling or to shut down the reactor.
- 4. IF.........
THEN..
you cannot restore torus water level below 18.5 ft and hold it there, BLOW DOWN:
Enter DEOP 400-2 while continuing here.
£
£ DEOP 200-1 PRIMARY CONTAINMENT CONTROL Dresden Nuclear Power Station Units 2(3)
EMERGENCY OPERATING PROCEDURE PRIMARY CONTAINMENT CONTROL DEOP 200-1 Title Number 1 0 Rev CATEGORY 1 D
Primary Containment Pressure Limit 0
10 20 30 40 50 60 70 80 90 100 93 Primary Containment Water Level (ft)
Torus Bottom Pressure (psig) 70 60 50 40 30 20 10 0
£
£ 0
5000 10,000 15,000 20,000 25,000 30,000 35,000 Total ECCS Flow (gpm)
Wide Range Torus Water Level (ft) 11 10.5 10 9.5 9
V ECCS Vortex Limit CAUTION: Exceeding LPCI NPSH/Vortex Limits (Figs V-X) may cause system damage.
- 1. Trip all recirc pumps.
- 2. Trip all drywell cooling fans.
- 3. Start drywell sprays.
Do not use pumps needed for core cooling.
OK to use external spray sources if you can restore and hold torus bottom pressure and primary containment water level inside Fig D, Primary Containment Pressure Limit.
Reducing primary containment pressure affects margin to NPSH limits. Check Figs W-Y, NPSH Limits.
K Drywell Spray Initiation Limit Drywell Pressure (psig) 20 18 16 14 12 10 8
6 4
2 0
Drywell Temperature (°F)
TR 2(3)-1340-1 Points 5 and 6 600 500 400 300 200 100 10 11 12 13 14 15 16 17 18 19 20 Wide Range Torus Water Level (ft)
L Pressure Suppression Pressure 10.9 Torus Bottom Pressure (psig) 40 30 25 20 15 10 5
0 35 18.6 M
Heat Capacity Limit 66 psig Use only when torus level is at or below 17 ft.
See TSGs for torus levels above 17 ft.
0 100 200 300 400 500 600 700 800 900 1000 1100 RPV Pressure (psig)
Torus Bulk Temperature (°F) 230 190 180 170 160 150 140 130 200 210 220 C Minimum Usable Indicating Levels Fuel Zone
(+60" to -340")
-297
-297
-298
-299
-300
-301 32 to 100 101 to 200 201 to 300 301 to 400 401 to 500 501 to 558 Drywell temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 Medium/Narrow Range All Rx Bldg Temps
(+60" to -60")
Medium/Narrow Range Rx Bldg Temps 181°F or less
(+60" to -60")
Wide Range
(+330" to -70")
-38
-39
-41
-43
-43
-43
-60
-60
-60
-60
-60
-60
-68
-51
-21 17 68 107 Drywell Temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 OR Reactor Building Temperature (°F) 600 500 400 300 200 0
100 200 300 400 500 600 700 800 900 1000 1100 RPV Pressure (psig)
B RPV Saturation Temperature A
RPV Water Level Instruments CAUTION: RPV water level instruments may be unreliable due to boiling in the instrument runs when drywell or reactor building temperatures near the instrument runs are above Fig B, RPV Saturation Temperature.
An RPV water level instrument may not be used if the instrument reads at or below its Minimum Usable Level (Detail C).
W LPCI / Core Spray NPSH Limit ECCS Flow Up To 10,750 gpm 0
1000 2000 3000 4000 5000 6000 LPCI / Core Spray Pump Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3.5 psig 5 psig 10 psig 15 psig Torus Bottom Pressure (psig)
X LPCI / Core Spray NPSH Limit ECCS Flow Up To 25,500 gpm 0
1000 2000 3000 4000 5000 6000 LPCI / Core Spray Pump Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3.5 psig 5 psig 10 psig 15 psig Torus Bottom Pressure (psig) 0 1000 2000 3000 4000 5000 6000 7000 HPCI Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3.5 psig 5 psig 10 psig 15 psig Torus Bottom Pressure (psig)
ENTRY CONDITIONS Any of the following in the reactor building:
CAUTION:
High temperature or low water level in the spent fuel pool can affect radiological conditions and water levels in the reactor building and adjacent areas. Reactor building access may also be restricted by steam.
IF THEN Reactor building exhaust radiation above 4 mr/hr Verify:
- Reactor Building Ventilation isolation
- SBGT start Reactor Building Ventilation isolates AND Reactor building exhaust radiation below 4 mr/hr Start Reactor Building Vent:
OK to defeat high drywell pressure and low RPV water level interlocks (DEOP 500-2).
TEMPERATURE / RADIATION Isolate all discharges into affected areas except systems needed for:
- Fire fighting
- Other DEOP actions IF THEN Primary system discharging into reactor building AND Discharge cannot be isolated Go to k Shut down the reactor.
BEFORE Any area temperature, radiation, or water level reaches max safe value (Details S, T, U)
- 1. Scram.
- 2. ENTER RPV CONTROL:
Enter DEOP 100 while continuing here.
WAIT until 2 or more areas above max safe values of same parameter (Details S, T, U)
BLOW DOWN:
Enter DEOP 400-2 while continuing here.
WATER LEVELS IF THEN Operate sump pumps to hold floor drain sump level below the hi-hi alarm setpoint and remove water from reactor building areas.
Floor drain sump level cannot be restored below the hi-hi alarm setpoint OR Water cannot be removed from an area Go to j j
Operate area coolers and Reactor Building Ventilation.
Reactor building temperature affects RPV water level indication. Check Detail A.
IF THEN Any area temperature or radiation above max normal (Details S, T)
Go to j k
(Scram, Enter DEOP 100)
WAIT until 2 or more areas above max safe values of same parameter (Details S, T, U)
ENTRY CONDITIONS DEOP 300-1 SECONDARY CONTAINMENT CONTROL DEOP 300-2 RADIOACTIVITY RELEASE CONTROL
- Operate Turbine Building Ventilation.
- Isolate all primary system discharges outside primary and secondary containments except systems needed for other DEOP actions.
IF THEN Primary system discharging outside primary and secondary containments AND Discharge cannot be isolated Before off-site release rate reaches the General Emergency level (Dresden EALs):
- 1. Scram.
- 2. ENTER RPV CONTROL:
Enter DEOP 100 while continuing here.
- 3. BLOW DOWN:
Enter DEOP 400-2 while continuing here.
Dresden Nuclear Power Station Units 2(3)
EMERGENCY OPERATING PROCEDURE Radioactivity Release Control DEOP 300-2 Title Number Rev CATEGORY 1 0 2 HPCI Room Shutdown Cooling Pump Room Shutdown Cooling Ht X Room Clean Up Demin Room Clean Up Pump & Ht X Area Isolation Condenser Area Area Max Normal Temperature (°F) 150 150 150 150 150 150 210 180 180 210 210 180 Max Safe Temperature (°F)
S Reactor Building Area Temperatures HPCI Cubicle, Unit 2(3)
East LPCI Pump Area West LPCI Pump Area East CRD Module Area West CRD Module Area Vessel Instrument Rack Area Area Max Normal Radiation (mr/hr) 150 (100) 12 9
30 50 30 2500 2500
- 2500
- 2500
- 2500
- 2500
- Max Safe Radiation (mr/hr)
Clean Up System Area 30 2500
- Isolation Condenser Area, Unit 2(3) 10 (2500 * )
- Measured by local survey.
T Reactor Building Area Radiation Levels 2500 East Corner Room Floor West Corner Room Floor Area Max Safe Water Level (in.)
8 8
U Reactor Building Area Water Levels Dresden Nuclear Power Station Units 2(3)
EMERGENCY OPERATING PROCEDURE Secondary Containment Control DEOP 300-1 Title Number Rev 09 CATEGORY 1 C Minimum Usable Indicating Levels Fuel Zone
(+60" to -340")
-297
-297
-298
-299
-300
-301 32 to 100 101 to 200 201 to 300 301 to 400 401 to 500 501 to 558 Drywell temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 Medium/Narrow Range All Rx Bldg Temps
(+60" to -60")
Medium/Narrow Range Rx Bldg Temps 181°F or less
(+60" to -60")
Wide Range
(+330" to -70")
-38
-39
-41
-43
-43
-43
-60
-60
-60
-60
-60
-60
-68
-51
-21 17 68 107 Drywell Temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 OR Reactor Building Temperature (°F) 600 500 400 300 200 0
100 200 300 400 500 600 700 800 900 1000 1100 RPV Pressure (psig)
B RPV Saturation Temperature A
RPV Water Level Instruments CAUTION: RPV water level instruments may be unreliable due to boiling in the instrument runs when drywell or reactor building temperatures near the instrument runs are above Fig B, RPV Saturation Temperature.
An RPV water level instrument may not be used if the instrument reads at or below its Minimum Usable Level (Detail C).
IF THEN All rods in to at least 04 OR Reactor will stay shutdown under all conditions without boron RPV water level indication is available Cannot stay inside Fig D, Primary Containment Pressure Limit Core damage is occurring AND TSC/EOF has command Go to h RETURN TO FAILURE TO SCRAM:
Exit this procedure Enter DEOP 400-5 at 6 (Level)
Enter DEOP 400-2
Stop injection from outside the primary containment not needed for RPV flooding or to shut down the reactor.
FLOOD CONTAINMENT:
Exit all DEOP Flowcharts Enter all SAMGs
START From DEOP 400-5 G
Terminate and prevent all RPV injection except boron and CRD.
- Condensate/Feedwater
OK to defeat high RPV water level trips (DEOP 500-2).
CAUTION:
Exceeding NPSH/Vortex Limits (Figs V-X) may cause system damage.
Use HXs as soon as you can.
CAUTION:
Injecting too fast may damage the core.
Using only the systems listed below, slowly raise injection to establish and hold the following conditions:
- At least 2 ADSVs open, AND
- RPV pressure above the value in Table J, but as low as possible.
IF THEN Cannot raise RPV pressure above the value in Table J and hold it there with at least 2 ADSVs open Slowly raise injection using Alternate Flooding Systems (Detail Q, DEOP 500-3).
IF THEN RPV water level indication is available Cannot stay inside Fig D, Primary Containment Pressure Limit RETURN TO RPV CONTROL:
Exit this procedure Enter DEOP 100 at 1 (Level)
Enter DEOP 400-2
Stop injection from outside the primary containment not needed for RPV flooding.
START From DEOP 100, 400-3 H
Above 6 ft At or below 6 ft 2 or more None or 1 Open all 5 ADSVs.
Leave switches in MAN.
OK to exceed 100°F/hr cooldown.
- Condensate/Feedwater
OK to defeat high RPV water level trips (DEOP 500-2).
CAUTION: Exceeding NPSH/Vortex Limits (Figs V-X) may cause system damage.
Use HXs as soon as you can.
CAUTION: Exceeding NPSH/Vortex Limits (Figs V-X) may cause system damage.
- Alternate Injection Systems (Detail E, DEOP 500-3)
Flood the RPV to the main steam lines using any of the systems listed below.
Isolate the following steam lines:
- Main steam line drains
?
No Yes Using any of the Emergency Depressurization Systems (Detail O3), depressurize the RPV until RPV pressure is less than 66 psi above drywell pressure.
OK to exceed 100°F/hr cooldown.
OK to exceed release rate limits.
OK to defeat interlocks (DEOP 500-2).
Yes No Is RPV pressure 66 psi or more above drywell pressure
?
WAIT until the RPV is flooded to the main steam lines Isolate the following steam lines:
- Main steam line drains
- HPCI Keep the RPV flooded to the main steam lines. Control injection flow as low as possible.
How many ADSVs open
?
2 or more None or 1 WAIT until RPV pressure is below value in Table J.
RPV Pressures J
Number of Open ADSVs RPV Pressure (psig) 5 330 4
420 3
560 2
850 Isolate the following steam lines:
- Main steam line drains
Can all 5 ADSVs be opened
?
Yes No Is RPV pressure 66 psi or more above drywell pressure
?
Yes No Using any of the Emergency Depressurization Systems (Detail O2), depressurize the RPV until:
- Available injection sources can control RPV pressure above the value in Table J.
OK to exceed 100°F/hr cooldown.
OK to exceed release rate limits.
OK to defeat interlocks (DEOP 500-2).
Yes No Can you control RPV pressure above the value in Table J with available injection sources
?
WAIT until the RPV is flooded to the main steam lines Above 6 ft At or below 6 ft Open all 5 ADSVs.
Leave switches in MAN.
OK to exceed 100°F/hr cooldown.
Torus water level
?
Isolate the following steam lines:
- Main steam line drains
- HPCI Keep the RPV flooded to the main steam lines. Control injection flow as low as possible.
Core damage is occurring AND TSC/EOF has command How many ADSVs open
?
FLOOD CONTAINMENT:
Exit all DEOP Flowcharts Enter all SAMGs
Emergency Depressurization Systems (ATWS RPV Flooding)
O2
- ADSVsonly if torus water level is above 6 ft.
- Max Recycle Reboiler
- Gland Seal steam
- Offgas preheater
- Main steam line drains
CAUTION: Exceeding NPSH/Vortex Limits (Figs V, Y) or 140°F suction temperature may cause system damage.
- Head vent
- IC vent valves Alternate Flooding Systems Q
CAUTION: Exceeding NPSH/Vortex Limits (Figs V, Y) or 140°F suction temperature may cause system damage.
OK to defeat high RPV water level trip (DEOP 500-2).
Use CST suction if you can. OK to defeat high torus level transfer (DEOP 500-2).
OK to defeat high area temperature isolation (DEOP 500-2).
CAUTION: Exceeding NPSH/Vortex Limits (Figs V-X) may cause system damage.
- Standby Coolant Supply
- SBLC boron tank
- SBLC test tank
- ECCS Keep Fill
- CRD crosstie
- Reactor head cooling
- Service unit backflush
- Condensate Transfer
- Fire System
- LPCI crosstie
- Core Spray or LPCI pump with CST suction Alternate Injection Systems E
- Standby Coolant Supply
- SBLC boron tank
- SBLC test tank
- ECCS Keep Fill
- CRD crosstie
- Reactor head cooling
- Service unit backflush
- Condensate Transfer
- Fire System
- LPCI crosstie
- Core Spray or LPCI pump with CST suction Torus water level
?
Emergency Depressurization Systems (non-ATWS RPV Flooding)
O3
- ADSVsonly if torus water level is above 6 ft.
- Max Recycle Reboiler
- Gland Seal steam
- Offgas preheater
- Main steam line drains
CAUTION: Exceeding NPSH/Vortex Limits (Figs V, Y) or 140°F suction temperature may cause system damage.
- Head vent
- IC vent valves
- RWCU, recirculation mode
Bypass demins.
- RWCU, blowdown mode
- Shutdown Cooling DEOP 400-1 RPV FLOODING Dresden Nuclear Power Station Units 2(3)
EMERGENCY OPERATING PROCEDURE RPV FLOODING DEOP 400-1 Title Number Rev 0 7 CATEGORY 1 D
Primary Containment Pressure Limit 0
10 20 30 40 50 60 70 80 90 100 93 Primary Containment Water Level (ft)
Torus Bottom Pressure (psig) 70 60 50 40 30 20 10 0
0 5000 10,000 15,000 20,000 25,000 30,000 35,000 Total ECCS Flow (gpm)
Wide Range Torus Water Level (ft) 11 10.5 10 9.5 9
V ECCS Vortex Limit W
LPCI / Core Spray NPSH Limit ECCS Flow Up To 10,750 gpm 0
1000 2000 3000 4000 5000 6000 LPCI / Core Spray Pump Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3.5 psig 5 psig 10 psig 15 psig Torus Bottom Pressure (psig)
X LPCI / Core Spray NPSH Limit ECCS Flow Up To 25,500 gpm 0
1000 2000 3000 4000 5000 6000 LPCI / Core Spray Pump Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3.5 psig 5 psig 10 psig 15 psig Torus Bottom Pressure (psig) 0 1000 2000 3000 4000 5000 6000 7000 HPCI Flow (gpm)
Torus Bulk Temperature (°F) 250 200 150 100 50 3.5 psig 5 psig 10 psig 15 psig Torus Bottom Pressure (psig)
DEOP 200 Figures A, B, and C, (RPV Water Level Instruments)
CATEGORY 1 UNIT 2(3)
DOP 6400-08 REVISION 29 13 of 15 FIGURE 5 DRESDEN UNIT 3 GENERATOR CAPABILITY CURVES AND UNDER EXCITATION LIMITER SETTINGS - REGULATOR IN SERVICE
CATEGORY 1 UNIT 2(3)
DOP 6400-08 REVISION 29 14 of 15 FIGURE 6 DRESDEN UNIT 3 GENERATOR CAPABILITY CURVES AND UNDER EXCITATION LIMITER SETTINGS - REGULATOR OUT OF SERVICE
CATEGORY 1 UNIT 2(3)
DOP 6400-08 REVISION 29 15 of 15 FIGURE 7 DRESDEN UNIT 3 GENERATOR CAPABILITY CURVES (EXPANDED VIEW)
REGULATOR IN SERVICE
Dresden Annex Exelon Nuclear June 2012 DR 3-10 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluent RG1 Offsite dose resulting from an 1 2 3 4 5 D-actual or IMMINENT release of gaseous radioactivity greater than 1000 mRem TEDE or 5000 mRem Thyroid CDE for the actual or projected duration of the release using actual meteorology.
RS1 Offsite dose resulting from an 1 2 3 4 5 D-actual or IMMINENT release of gaseous radioactivity greater than 100 mRem TEDE or 500 mRem Thyroid CDE for the actual or projected duration of the release using actual meteorology.
RA1 Any release of gaseous or 1 2 3 4 5 D-liquid radioactivity to the environment greater than 200 times the ODCM for 15 minutes or longer.
RU1 Any release of gaseous or 1 2 3 4 5 D-liquid radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Radiological Effluents Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.
- 1.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 7.90 E+09 uCi/sec for 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 2.
Dose assessment using actual meteorology indicates doses at or beyond the Site Boundary of EITHER:
- 3.
Field survey results at or beyond Site Boundary indicate EITHER:
- a. Gamma (closed window) dose rates
> 1000 mR/hr are expected to continue for
> 60 minutes.
- b. Analyses of field survey samples indicate
> 5000 mRem CDE Thyroid for 60 minutes of inhalation.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.
- 1.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 7.90 E+08 uCi/sec for 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate.)
- 2.
Dose assessment using actual meteorology indicates doses at or beyond the Site Boundary of EITHER:
- a.
- b.
- 3.
Field survey results at or beyond Site Boundary indicate EITHER:
- a.
Gamma (closed window) dose rates
> 100 mR/hr are expected to continue for
> 60 minutes.
- b.
Analyses of field survey samples indicate
> 500 mRem CDE Thyroid for 60 minutes of inhalation.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.
- 1.
VALID reading on any of the following effluent monitors > 200 times alarm setpoint established by a current radioactivity discharge permit for
> 15 minutes.
Radwaste Effluent Monitor 2/3-2001-948 OR Discharge Permit specified monitor OR
- 2.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 9.02 E+07 uCi/sec for 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 3.
Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates
> 200 times ODCM Limit with a release duration of
> 15 minutes.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.
- 1.
VALID reading on any of the following effluent monitors > 2 times alarm setpoint established by a current radioactivity discharge permit for
> 60 minutes.
Radwaste Effluent Monitor 2/3-2001-948 OR Discharge Permit specified monitor OR
- 2.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 9.02 E+05 uCi/sec for 60 minutes. (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 3.
Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates
> 2 times ODCM Limit with a release duration of
> 60 minutes.
Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-11 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluent RA2 Damage to irradiated fuel or loss 1 2 3 4 5 D-of water level that has resulted or will result in the uncovering of irradiated fuel outside the reactor vessel.
RU2 UNPLANNED rise in plant 1 2 3 4 5 D-radiation levels.
EAL Threshold Values:
EAL Threshold Values:
Table R1 - Refuel Floor ARMs Refuel Floor High Range ARM Station #2(4)
Fuel Pool Radiation Monitor
- 1.
VALID reading > 1000 mR/hr on any Table R1 Radiation Monitor due to EITHER:
Damaged irradiated fuel.
OR Water level drop.
- 2.
Water level drop in the Reactor Refueling Cavity, Spent Fuel Pool or Fuel Transfer Canal that will result in irradiated fuel becoming uncovered.
- 1.
- a. UNPLANNED water level drop in the reactor Refueling Cavity, Spent Fuel Pool or Fuel Transfer Canal as indicated by:
Refueling Cavity water level < 466 in. (Refuel Outage Reactor Vessel and Cavity Level Instrument LI 2(3)-263-114).
OR Spent Fuel Pool water level < 19 ft. above the fuel (33 ft. 9 in. indicated level).
OR Report of visual observation of a drop in water level in the Refueling Cavity, Spent Fuel Pool or Fuel Transfer Canal.
AND
- b. VALID Area Radiation Monitor reading rise on one or more radiation monitor in Table R1.
- 2.
UNPLANNED VALID Area Radiation Monitor readings or survey results indicate rise by a factor of 1000 over NORMAL LEVELS.
RA3 Rise in radiation levels within 1 2 3 4 5 D-the facility that impedes operation of systems required to maintain plant safety functions.
RU3 Fuel Clad Degradation.
1 2 3 EAL Threshold Values:
EAL Threshold Values:
Abnormal Rad Levels Table R2 Areas Requiring Continuous Occupancy Main Control Room Central Alarm Station - (by survey)
Dose rate > 15 mR/hr in ANY area requiring continuous occupancy (Table R2) to maintain plant safety functions.
Fuel clad degradation resulting in EITHER:
Offgas system radiation monitor HI-HI alarm.
OR Specific coolant activity > 4.0 uCi/gm Dose Equivalent I-131.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-12 EP-AA-1004 (Revision 29)
Fission Product Barrier Matrix Hot Matrix GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT FG1 Loss of ANY two barriers AND Loss or 1 2 3-Potential Loss of third barrier.
FS1 Loss or Potential Loss of ANY two barriers.
1 2 3-FA1 ANY Loss or ANY Potential Loss of either 1 2 3-Fuel Clad or RCS.
FU1 ANY Loss or ANY Potential Loss of 1 2 3-Containment.
FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss
- 1. Pri Coolant Activity Coolant activity > 300 uCi/gm Dose Equivalent I-131 None None None None None 2.RPV Water Level
- 1.
RPV level cannot be restored and maintained > - 164 inches.
- 2.
RPV level cannot be restored and maintained > - 143 inches. (TAF).
- 3.
RPV level cannot be determined.
- 1.
RPV level cannot be restored and maintained > -143 inches. (TAF).
- 2.
RPV level cannot be determined.
None None Plant conditions indicate that Primary Containment Flooding is required.
- 3. Primary Cont Conditions None None
- 1. Drywell pressure > 2.0 psig.
AND
- 2. Drywell pressure rise due to RCS leakage.
None
- 1.
Rapid unexplained drop in Drywell Pressure following initial pressure rise.
- 2.
Drywell pressure response not consistent with LOCA conditions.
- 3.
Drywell pressure 62 psig and rising.
- 4.
- a.
Drywell or torus hydrogen concentration 6%.
AND
- b.
Drywell or torus oxygen concentration 5%.
- 5.
Heat Capacity Limit (DEOP 200-1 Fig. M) exceeded.
- 1.
UNISOLABLE Main Steam Line (MSL), Isolation Condenser, HPCI, Feedwater, or RWCU line break.
OR 2
Emergency RPV Depressurization is required.
- 3. RCS leakage > 50 gpm inside the drywell.
- 4. UNISOLABLE primary system leakage outside drywell resulting in EITHER:
Secondary Containment area temperatures > DEOP 300-1, Maximum Normal operating levels.
OR Secondary Containment radiation levels > DEOP 300-1, Maximum Normal operating levels.
- 5. Pri Cont Rad Monitoring Drywell radiation monitor reading
> 6.70E+02 R/hr.
None Drywell Radiation monitor reading
> 100 R/hr.
None None Drywell radiation monitor reading
> 1.60 E+03 R/hr.
- 6. Primary Cont Isolation Failure or Bypass None None None None
- 1. a. Failure of isolation valves in any one line to close.
AND
- b. Direct downstream pathway to the environment exists after a primary containment isolation signal.
- 2.
Intentional venting/purging of Primary Containment per EOPs or SAMGs due to accident conditions.
- 3.
UNISOLABLE primary system leakage outside drywell resulting in EITHER:
Secondary Containment area temperatures > DEOP 300-1, Maximum Safe operating levels.
OR Secondary Containment radiation levels > DEOP 300-1, Maximum Safe operating levels.
None
- 7. ED Judgment.
- 1. Any condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.
- 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.
- 1. Any condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.
- 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.
- 1. Any condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier.
- 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.
Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled
Dresden Annex Exelon Nuclear June 2012 DR 3-13 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT System Malfunction MG1 Prolonged loss of all offsite power and 1 2 3 all On-Site AC power to emergency busses.
MS1 Loss of all Off-site and all On-Site 1 2 3 AC power to emergency busses for 15 minutes or longer.
MA1 AC power capability to emergency 1 2 3 busses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in station blackout.
MU1 Loss of all Off-site AC power to 1 2 3 busses for 15 minutes or longer.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Loss of AC Power Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
Loss of all off-site AC power to unit ECCS busses.
AND
- 2.
Failure of DG 2(3), shared DG 2/3 and SBO DG 2(3) emergency diesel generators to supply power to unit ECCS busses.
AND
- 3.
- a. Restoration of at least one unit ECCS bus in
< 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.
- b. RPV level cannot be determined to be
> - 143 inches (TAF).
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
Loss of all off-site AC power to unit ECCS busses.
AND
- 2.
Failure of DG 2(3), shared DG 2/3 and SBO DG 2(3) emergency diesel generators to supply power to unit ECCS busses.
AND
- 3.
Failure to restore power to at least one ECCS bus in
< 15 minutes from the time of loss of both offsite and onsite AC power.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
AC power capability to unit ECCS busses reduced to only one of the following power sources for
> 15 minutes:
Reserve auxiliary transformer TR-22(TR-32)
Unit auxiliary transformer TR-21(TR-31)
Unit Emergency Diesel Generator DG 2(3)
Shared Emergency Diesel Generator DG 2/3 Station Blackout Diesel Generator SBO DG 2(3)
Unit crosstie breakers AND
- 2.
Any additional single power source failure will result in a unit blackout.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
Loss of all off-site AC power to unit ECCS busses for
> 15 minutes.
MG2 Automatic Scram and all manual actions fail 1 2-to shutdown the reactor and indication of an extreme challenge to the ability to cool the core exists.
MS2 Automatic Scram fails to shutdown the 1 2-reactor and manual actions taken from the reactor control console are not successful in shutting down the reactor.
MA2 Automatic Scram fails to shutdown the 1 2 reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor.
MU2 Inadvertent criticality.
3 EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
RPS Failure / Inadvertent Criticality
- 1.
Automatic scram was not successful as indicated by Reactor Power > 6%.
AND
- 2.
Manual scram/ARI actions were not successful as indicated by Reactor Power > 6%.
AND
- 3.
EITHER of the following exists:
RPV level cannot be restored and maintained
> - 164 inches.
OR Heat Capacity Limit (DEOP 200-1 Fig. M) exceeded.
- 1.
Automatic scram was not successful as indicated by Reactor Power > 6%.
AND
- 2. Manual scram/ARI actions were not successful from the Reactor Console as indicated by Reactor Power
> 6%.
- 1.
Automatic scram was not successful as indicated by Reactor Power > 6%.
AND
- 2.
Manual scram/ARI actions were successful from Reactor Console as indicated by Reactor Power < 6%.
UNPLANNED sustained positive period observed on nuclear instrumentation.
Modes:
1 - Power Operation, 2 - Startup, 3 -Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-14 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT System Malfunction MS3 Loss of all vital DC power for 1 2 3 15 minutes or longer.
EAL Threshold Values:
DC Power Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
Loss of all vital DC power based on < 105 VDC on 125 VDC battery busses #2 and #3 for > 15 minutes.
MS4 Inability to monitor a SIGNIFICANT 1 2 3 TRANSIENT in progress.
MA4 UNPLANNED Loss of safety system 1 2 3 annunciation or indication in the control room with EITHER (1) a SIGNIFICANT TRANSIENT in progress, or (2) compensatory indicators unavailable.
MU4 UNPLANNED loss of safety system 1 2 3 annunciation or indication in the control room for 15 minutes or longer.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Annunciators Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
Loss of greater than approximately 75% of the following for > 15 minutes:
Safety System annunciators (Table M1)
OR Safety System indications (Table M1)
AND
- 2.
SIGNIFICANT TRANSIENT in progress (Table M2).
AND
- 3.
Compensatory indications (computer points) are unavailable.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1. UNPLANNED loss of greater than approximately 75% of the following for > 15 minutes:
Safety System annunciators (Table M1)
OR Safety System indications (Table M1)
AND
- 2.
- a. SIGNIFICANT TRANSIENT in progress (Table M2).
- b. Compensatory indications (computer points) are unavailable.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
UNPLANNED loss of greater than approximately 75% of the following for > 15 minutes:
Safety System annunciators (Table M1)
OR Safety System indications (Table M1)
Table M1 - Safety Systems Table M2 - Significant Transients ECCS Containment Isolation Reactor Scram Process/Area Radiation Monitoring Turbine trip Reactor scram ECCS actuation Recirc Runback >25% power change Thermal power oscillations > 10 % Reactor Power change Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-15 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT System Malfunction MU5 RCS leakage.
1 2 3 EAL Threshold Values:
RCS Leak
- 1.
Unidentified or pressure boundary leakage into the Drywell > 10 gpm.
- 2.
Identified leakage into the Drywell > 25 gpm.
MU6 Loss of all On-site 1 2 3 or Off-site communications capabilities.
EAL Threshold Values:
Communications Table M3 - Communications Capability System Onsite Offsite Plant Radio System X
Plant Paging System X
Sound Power Phones X
In-Plant Telephones X
All telephone lines (commercial and microwave)
X ENS X
Satellite Phones X
HPN X
Cellular Phones X
- 1.
Loss of all Table M3 Onsite communications capability affecting the ability to perform routine operations.
- 2.
Loss of all Table M3 Offsite communications capability affecting the ability to perform offsite notifications.
MU7 Inability to reach required shutdown 1 2 3 within Technical Specification limits.
EAL Threshold Values:
T. S. Time Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-16 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other Conditions Affecting Plant Safety HG1 HOSTILE ACTION resulting in loss 1 2 3 4 5 D-of physical control of the facility.
HS1 HOSTILE ACTION within the 1 2 3 4 5 D-PROTECTED AREA.
HA1 HOSTILE ACTION within the 1 2 3 4 5 D-OWNER CONTROLLED AREA or airborne attack threat.
HU1 Confirmed SECURITY CONDITION 1 2 3 4 5 D-or threat, which indicates a potential degradation in the level of safety of the plant.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Security
- 1.
A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain ANY safety function (Table H1).
- 2.
A HOSTILE ACTION has:
Caused failure of Spent Fuel Pool Cooling Systems AND IMMINENT fuel damage is likely for freshly offloaded reactor fuel in the pool (e.g., within 120 days).
A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.
- 1.
A validated notification from NRC of an airliner attack threat < 30 minutes from the site.
- 2.
Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA.
- 1.
A credible site-specific security threat notification as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.
- 2. A validated notification from NRC providing information of an aircraft threat.
- 3.
Notification by the Security Force of a SECURITY CONDITION that does not involve a HOSTILE ACTION.
HS2 Control Room evacuation has 1 2 3 4 5 D-been initiated and plant control cannot be established.
HA2 Control Room evacuation has 1 2 3 4 5 D-been initiated.
EAL Threshold Values:
EAL Threshold Values:
C. R. Evacuation Table H1 - Safety Functions Reactivity Control (ability to shut down the reactor and keep it shutdown)
RCS Inventory (ability to cool the core)
Decay Heat Removal (ability to maintain heat sink)
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
Control Room evacuation has been initiated.
AND
- 2.
Control of the plant cannot be established per DSSP 0100-CR in < 30 minutes.
Entry into DSSP 0100-CR for Control Room evacuation.
HA3 FIRE or EXPLOSION affecting 1 2 3 4 5 D-the operability of plant safety systems required to establish or maintain safe shutdown.
HU3 FIRE within the PROTECTED 1 2 3 4 5 D-AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA.
EAL Threshold Values:
EAL Threshold Values:
Fire / Explosion Table H2 - Vital Areas Reactor Building Aux Electric Room Control Room Diesel Generator Rooms 4-KV ECCS Switchgear Area Battery Rooms CRD & CCSW Pump Rooms Turbine Building Cable Tunnel Turbine Building Safe Shutdown Areas Crib House FIRE or EXPLOSION resulting in any of the following:
VISIBLE DAMAGE to a Table H2 permanent structure.
OR VISIBLE DAMAGE to safety system equipment contained within a Table H2 area.
OR Control Room indication of degraded safety system equipment performance contained within a Table H2 area.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
FIRE in any Table H2 area not extinguished in
< 15 minutes of Control Room notification or verification of a Control Room FIRE alarm.
- 2.
EXPLOSION within PROTECTED AREA boundary affecting a Table H2 area.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-17 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other Conditions Affecting Plant Safety HA4 Natural and destructive 1 2 3 4 5 D-phenomena affecting VITAL AREAS HU4 Natural and destructive 1 2 3 4 5 D-phenomena affecting the PROTECTED AREA.
EAL Threshold Values:
EAL Threshold Values:
Natural / Destructive Phenomena Table H2 - Vital Areas Reactor Building Aux Electric Room Control Room Diesel Generator Rooms 4-KV ECCS Switchgear Area Battery Rooms CRD & CCSW Pump Rooms Turbine Building Cable Tunnel Turbine Building Safe Shutdown Areas Crib House Table H3 - Internal Flooding Areas Condenser Pits Condensate pump Rooms Containment Cooling Service Water Vaults Crib House East Corner Room West Corner Room
- 1. a.
Seismic event > Operating Basis Earthquake (OBE) as indicated by seismic instrumentation:
> 0.10 g (Channel 1 or 3)
> 0.067 g (Channel 2)
AND
- b.
Confirmed by EITHER:
Earthquake felt in plant.
OR National Earthquake Center.
OR Control Room indication of degraded performance of systems required for the safe shutdown of the plant.
- 2.
ANY of the following resulting in VISIBLE DAMAGE to any Table H2 structure OR Control Room indication of degraded performance of a safety system in any Table H2 area:
Tornado strike OR High winds > 100 mph OR Vehicle crash OR Turbine failure-generated PROJECTILES OR
- 3.
Flooding in any Table H3 area that results in ANY of the following:
Degraded safety system performance as indicated in the Control Room.
OR Industrial safety hazards (e.g., electric shock) that preclude access necessary to operate or monitor safety equipment.
OR Water level > DEOP 300-1, Maximum Safe.
- 1. Seismic event identified by any TWO of the following:
Earthquake felt in plant.
OR Seismic event confirmed by station seismic monitor procedure.
OR National Earthquake Center.
- 2. EITHER of the following occurring within the PROTECTED AREA boundary:
Tornado strike OR Sustained (> 15 minutes) high winds > 100 mph OR
- 3. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
- 4.
Flooding in any Table H3 area that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-18 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other Conditions Affecting Plant Safety HA5 Access to a VITAL AREA 1 2 3 4 5 D-is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of operable equipment required to maintain safe operations or safely shutdown the reactor.
HU5 Release of toxic, corrosive, 1 2 3 4 5 D-asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS.
EAL Threshold Values:
EAL Threshold Values:
Toxic / Flammable Gas Table H2 - Vital Areas Reactor Building Aux Electric Room Control Room Diesel Generator Rooms 4-KV ECCS Switchgear Area Battery Rooms CRD & CCSW Pump Rooms Turbine Building Cable Tunnel Turbine Building Safe Shutdown Areas Crib House Note: If the equipment in the stated area was already inoperable, or out of service, before the event occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.
Access to a Table H2 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases, which jeopardize operation of systems required to maintain safe operations or safely shutdown the reactor.
- 1.
Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS.
- 2.
Report by Local, County or State Officials for evacuation or sheltering of site personnel based on an offsite event.
HG6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY.
HS6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY.
HA6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of an ALERT.
HU6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Judgment Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to equipment needed for the protection of the public.
Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-19 EP-AA-1004 (Revision 29)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI Malfunction E-HU1 Damage to a loaded cask 1 2 3 4 5 D-CONFINEMENT BOUNDARY EAL Threshold Values:
ISFSI Damage to a loaded cask CONFINEMENT BOUNDARY resulting in radiation readings > 10 times normal..
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled HOT MATRIX HOT MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-20 EP-AA-1004 (Revision 29)
This Page Intentionally Left Blank
Dresden Annex Exelon Nuclear June 2012 DR 3-21 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluent RG1 Offsite dose resulting from an 1 2 3 4 5 D-actual or IMMINENT release of gaseous radioactivity greater than 1000 mRem TEDE or 5000 mRem Thyroid CDE for the actual or projected duration of the release using actual meteorology.
RS1 Offsite dose resulting from an 1 2 3 4 5 D-actual or IMMINENT release of gaseous radioactivity greater than 100 mRem TEDE or 500 mRem Thyroid CDE for the actual or projected duration of the release using actual meteorology.
RA1 Any release of gaseous or 1 2 3 4 5 D-liquid radioactivity to the environment greater than 200 times the ODCM for 15 minutes or longer.
RU1 Any release of gaseous or 1 2 3 4 5 D-liquid radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Radiological Effluents Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.
- 1.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 7.90 E+09 uCi/sec for 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 2.
Dose assessment using actual meteorology indicates doses at or beyond the Site Boundary of EITHER:
- 3.
Field survey results at or beyond Site Boundary indicate EITHER:
- a. Gamma (closed window) dose rates
> 1000 mR/hr are expected to continue for
> 60 minutes.
- b. Analyses of field survey samples indicate
> 5000 mRem CDE Thyroid for 60 minutes of inhalation.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results.
- 1.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 7.90 E+08 uCi/sec for 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate.)
- 2.
Dose assessment using actual meteorology indicates doses at or beyond the Site Boundary of EITHER:
- a.
- b.
- 3.
Field survey results at or beyond Site Boundary indicate EITHER:
- a.
Gamma (closed window) dose rates
> 100 mR/hr are expected to continue for
> 60 minutes.
- b.
Analyses of field survey samples indicate
> 500 mRem CDE Thyroid for 60 minutes of inhalation.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.
- 1.
VALID reading on any of the following effluent monitors
> 200 times alarm setpoint established by a current radioactivity discharge permit for > 15 minutes.
Radwaste Effluent Monitor 2/3-2001-948 OR Discharge Permit specified monitor OR
- 2.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 9.02 E+07 uCi/sec for 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 3.
Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates
> 200 times ODCM Limit with a release duration of
> 15 minutes.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.
- 1.
VALID reading on any of the following effluent monitors > 2 times alarm setpoint established by a current radioactivity discharge permit for
> 60 minutes.
Radwaste Effluent Monitor 2/3-2001-948 OR Discharge Permit specified monitor OR
- 2.
The sum of VALID readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 9.02 E+05 uCi/sec for 60 minutes. (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 3.
Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates
> 2 times ODCM Limit with a release duration of
> 60 minutes.
Modes: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-22 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluent RA2 Damage to irradiated fuel or loss 1 2 3 4 5 D-of water level that has resulted or will result in the uncovering of irradiated fuel outside the reactor vessel.
RU2 UNPLANNED rise in plant 1 2 3 4 5 D-radiation levels.
EAL Threshold Values:
EAL Threshold Values:
Table R1 - Refuel Floor ARMs Refuel Floor High Range ARM Station #2(4)
Fuel Pool Radiation Monitor
- 1.
VALID reading > 1000 mR/hr on any Table R1 Radiation Monitor due to EITHER:
Damaged irradiated fuel.
OR Water level drop.
- 2.
Water level drop in the Reactor Refueling Cavity, Spent Fuel Pool or Fuel Transfer Canal that will result in irradiated fuel becoming uncovered.
- 1.
- a. UNPLANNED water level drop in the reactor Refueling Cavity, Spent Fuel Pool or Fuel Transfer Canal as indicated by:
Refueling Cavity water level < 466 in. (Refuel Outage Reactor Vessel and Cavity Level Instrument LI 2(3)-263-114).
OR Spent Fuel Pool water level < 19 ft. above the fuel (33 ft. 9 in. indicated level).
OR Report of visual observation of a drop in water level in the Refueling Cavity, Spent Fuel Pool or Fuel Transfer Canal.
AND
- b. VALID Area Radiation Monitor reading rise on one or more radiation monitor in Table R1.
- 2.
UNPLANNED VALID Area Radiation Monitor readings or survey results indicate rise by a factor of 1000 over NORMAL LEVELS.
RA3 Rise in radiation levels within 1 2 3 4 5 D-the facility that impedes operation of systems required to maintain plant safety functions.
EAL Threshold Values:
Abnormal Rad Levels Table R2 Areas Requiring Continuous Occupancy Main Control Room Central Alarm Station - (by survey)
Dose rate > 15 mR/hr in ANY area requiring continuous occupancy (Table R2) to maintain plant safety functions.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-23 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU1 AC power capibilty to emergency 4 5 busses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in station blackout.
EAL Threshold Values:
Loss of AC Power CA1 Loss of all Off-site and all On-Site AC 4 5 D-power to emergency busses for 15 minutes or longer.
EAL Threshold Values:
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
Loss of all off-site AC power to unit ECCS busses.
AND
- 2.
Failure of DG 2(3), shared DG 2/3 and SBO DG 2(3) emergency diesel generators to supply power to unit ECCS busses.
AND
- 3.
Failure to restore power to at least one ECCS bus in
< 15 minutes from the time of loss of both offsite and onsite AC power.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
AC power capability to unit ECCS busses reduced to only one of the following power sources for
> 15 minutes:
Reserve auxiliary transformer TR-22(TR-32)
Unit auxiliary transformer TR-21(TR-31)
Unit Emergency Diesel Generator DG 2(3)
Shared Emergency Diesel Generator DG 2/3 Station Blackout Diesel Generator SBO DG 2(3)
Unit crosstie breakers AND
- 2. Any additional single power source failure will result in a unit blackout.
CU2 Inadvertent criticality.
4 5 EAL Threshold Values:
Inadvertent Criticality UNPLANNED sustained positive period observed on nuclear instrumentation.
CU3 Loss of required DC power for 15 minutes 4 5-or longer.
EAL Threshold Values:
DC Power Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
Loss of required DC power based on < 105 VDC on unit 125 VDC battery buses #2 and #3 for > 15 minutes.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-24 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Loss of all On-site 4 5 D-or Off-site communications capabilities.
EAL Threshold Values:
Communications Table C1 - Communications Capability System Onsite Offsite Plant Radio System X
Plant Paging System X
Sound Power Phones X
In-Plant Telephones X
All telephone lines (commercial and microwave)
X ENS X
Satellite Phones X
HPN X
Cellular Phones X
- 1.
Loss of all Table C1 Onsite communications capability affecting the ability to perform routine operations.
- 2.
Loss of all Table C1 Offsite communications capability affecting the ability to perform offsite notifications.
CA5 Inability to maintain plant in cold shutdown.
4 5-CU5 UNPLANNED loss of decay heat removal 4 5-capability with irradiated fuel in the RPV.
EAL Threshold Values:
EAL Threshold Values:
Heat Sink Table C2 - RCS Reheat Duration Thresholds RCS Containment Closure Duration Intact N/A 60 minutes*
Established 20 minutes*
Not Intact Not Established 0 minutes
- If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, then EAL Threshold #1 is not applicable.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
UNPLANNED loss of decay heat removal capability results in RCS temperature > 212°F for > Table C2 duration.
- 2.
UNPLANNED RPV pressure rise > 10 psig as a result of temperature rise due to loss of decay heat removal.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
UNPLANNED loss of decay heat removal capability results in RCS temperature > 212F.
- 2.
Loss of the following for > 15 minutes:
All RCS temperature indications AND All RPV level indications Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-25 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Loss of RPV inventory affecting 4 5-fuel clad integrity with containment challenged.
CS6 Loss of RPV inventory affecting 4 5-core decay heat removal capability.
CA6 Loss of RPV inventory.
4 5-CU6 RCS leakage.
4-EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
AND
- b. Any Containment Challenge Indication (Table C4).
- 2.
- a. RPV level unknown for > 30 minutes.
AND
- b. Loss of RPV inventory as indicated by any of the following:
Table C3 indication.
OR Erratic Source Range Monitor indication.
OR Refuel Floor Hi Range ARM > 3000 mR/hr.
AND
- c. Any Containment Challenge Indication (Table C4)
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1. With CONTAINMENT CLOSURE not established, RPV level < - 60 inches.
- 3.
- a. RPV level unknown for > 30 minutes.
AND
- b. Loss of RPV inventory as indicated by any of the following:
Table C3 indications.
OR Erratic Source Range Neutron Monitor indication.
OR Refuel Floor Hi Range ARM > 3000 mR/hr.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
RPV level < - 54 inches.
- 2.
- a.
RPV level unknown for > 15 minutes.
AND
- b.
Loss of RPV inventory per Table C3 indications.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
RCS leakage results in the inability to restore and maintain RPV level > 0 inches for > 15 minutes.
RCS Leakage / Inventory Table C4 - Containment Challenge Indications Primary Containment Hydrogen concentration > 6%
Oxygen > 5%.
UNPLANNED rise in containment pressure CONTAINMENT CLOSURE not established.
Any Secondary Containment radiation monitors reading > DEOP 300-1 Maximum Safe operating level.
Table C3 - Indications of RCS Leakage UNPLANNED floor or equipment sump level rise UNPLANNED Torus level rise UNPLANNED vessel make-up rise Observation of leakage or Inventory loss CU7 UNPLANNED loss of RPV inventory.
5-EAL Threshold Values:
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1. UNPLANNED Reactor Cavity or Vessel level drop that meets EITHER:
When controlling level above the RPV flange, Reactor Cavity level drop below the RPV flange for > 15 minutes.
OR When controlling level below the RPV flange, vessel level drop below the procedurally established limit > 15 minutes.
- 2.
- a. RPV level unknown.
AND
- b. Loss of RPV inventory per Table C3 indications.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-26 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other Conditions Affecting Plant Safety HG1 HOSTILE ACTION resulting in loss 1 2 3 4 5 D-of physical control of the facility.
HS1 HOSTILE ACTION within the 1 2 3 4 5 D-PROTECTED AREA.
HA1 HOSTILE ACTION within the 1 2 3 4 5 D-OWNER CONTROLLED AREA or airborne attack threat.
HU1 Confirmed SECURITY CONDITION 1 2 3 4 5 D-or threat, which indicates a potential degradation in the level of safety of the plant.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Security
- 1.
A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain ANY safety function (Table H1).
- 2.
A HOSTILE ACTION has:
Caused failure of Spent Fuel Pool Cooling Systems AND IMMINENT fuel damage is likely for freshly offloaded reactor fuel in the pool (e.g., within 120 days).
A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.
- 1.
A validated notification from NRC of an airliner attack threat < 30 minutes from the site.
- 2.
Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA.
- 1.
A credible site-specific security threat notification as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.
- 2. A validated notification from NRC providing information of an aircraft threat.
- 3.
Notification by the Security Force of a SECURITY CONDITION that does not involve a HOSTILE ACTION.
HS2 Control Room evacuation has 1 2 3 4 5 D-been initiated and plant control cannot be established.
HA2 Control Room evacuation has 1 2 3 4 5 D-been initiated.
EAL Threshold Values:
EAL Threshold Values:
C. R. Evacuation Table H1 - Safety Functions Reactivity Control (ability to shut down the reactor and keep it shutdown)
RCS Inventory (ability to cool the core)
Decay Heat Removal (ability to maintain heat sink)
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
Control Room evacuation has been initiated.
AND
- 2.
Control of the plant cannot be established per DSSP 0100-CR in < 30 minutes.
Entry into DSSP 0100-CR for Control Room evacuation.
HA3 FIRE or EXPLOSION affecting 1 2 3 4 5 D-the operability of plant safety systems required to establish or maintain safe shutdown.
HU3 FIRE within the PROTECTED 1 2 3 4 5 D-AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA.
EAL Threshold Values:
EAL Threshold Values:
Fire / Explosion Table H2 - Vital Areas Reactor Building Aux Electric Room Control Room Diesel Generator Rooms 4-KV ECCS Switchgear Area Battery Rooms CRD & CCSW Pump Rooms Turbine Building Cable Tunnel Turbine Building Safe Shutdown Areas Crib House FIRE or EXPLOSION resulting in any of the following:
VISIBLE DAMAGE to a Table H2 permanent structure.
OR VISIBLE DAMAGE to safety system equipment contained within a Table H2 area.
OR Control Room indication of degraded safety system equipment performance contained within a Table H2 area.
Note: The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.
- 1.
FIRE in any Table H2 area not extinguished in
< 15 minutes of Control Room notification or verification of a Control Room FIRE alarm.
- 2.
EXPLOSION within PROTECTED AREA boundary affecting a Table H2 area.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-27 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other Conditions Affecting Plant Safety HA4 Natural and destructive 1 2 3 4 5 D-phenomena affecting VITAL AREAS HU4 Natural and destructive 1 2 3 4 5 D-phenomena affecting the PROTECTED AREA.
EAL Threshold Values:
EAL Threshold Values:
Natural / Destructive Phenomena Table H2 - Vital Areas Reactor Building Aux Electric Room Control Room Diesel Generator Rooms 4-KV ECCS Switchgear Area Battery Rooms CRD & CCSW Pump Rooms Turbine Building Cable Tunnel Turbine Building Safe Shutdown Areas Crib House Table H3 - Internal Flooding Areas Condenser Pits Condensate pump Rooms Containment Cooling Service Water Vaults Crib House East Corner Room West Corner Room
- 1. a.
Seismic event > Operating Basis Earthquake (OBE) as indicated by seismic instrumentation:
> 0.10 g (Channel 1 or 3)
> 0.067 g (Channel 2)
AND
- b.
Confirmed by EITHER:
Earthquake felt in plant.
OR National Earthquake Center.
OR Control Room indication of degraded performance of systems required for the safe shutdown of the plant.
- 2.
ANY of the following resulting in VISIBLE DAMAGE to any Table H2 structure OR Control Room indication of degraded performance of a safety system in any Table H2 area:
Tornado strike OR High winds > 100 mph OR Vehicle crash OR Turbine failure-generated PROJECTILES OR
- 3.
Flooding in any Table H3 area that results in ANY of the following:
Degraded safety system performance as indicated in the Control Room.
OR Industrial safety hazards (e.g., electric shock) that preclude access necessary to operate or monitor safety equipment.
OR Water level > DEOP 300-1, Maximum Safe.
- 1. Seismic event identified by any TWO of the following:
Earthquake felt in plant.
OR Seismic event confirmed by station seismic monitor procedure.
OR National Earthquake Center.
- 2. EITHER of the following occurring within the PROTECTED AREA boundary:
Tornado strike OR Sustained (> 15 minutes) high winds > 100 mph OR
- 3. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
- 4.
Flooding in any Table H3 area that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-28 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other Conditions Affecting Plant Safety HA5 Access to a VITAL AREA 1 2 3 4 5 D-is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of operable equipment required to maintain safe operations or safely shutdown the reactor.
HU5 Release of toxic, corrosive, 1 2 3 4 5 D-asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS.
EAL Threshold Values:
EAL Threshold Values:
Toxic / Flammable Gas Table H2 - Vital Areas Reactor Building Aux Electric Room Control Room Diesel Generator Rooms 4-KV ECCS Switchgear Area Battery Rooms CRD & CCSW Pump Rooms Turbine Building Cable Tunnel Turbine Building Safe Shutdown Areas Crib House Note: If the equipment in the stated area was already inoperable, or out of service, before the event occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.
Access to a Table H2 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases, which jeopardize operation of systems required to maintain safe operations or safely shutdown the reactor.
- 1.
Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS.
- 2.
Report by Local, County or State Officials for evacuation or sheltering of site personnel based on an offsite event.
HG6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY.
HS6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY.
HA6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of an ALERT.
HU6 Other conditions existing which in 1 2 3 4 5 D-the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
EAL Threshold Values:
Judgment Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to equipment needed for the protection of the public.
Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
Dresden Annex Exelon Nuclear June 2012 DR 3-29 EP-AA-1004 (Revision 29)
COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI Malfunction E-HU1 Damage to a loaded cask 1 2 3 4 5 D-CONFINEMENT BOUNDARY EAL Threshold Values:
ISFSI Damage to a loaded cask CONFINEMENT BOUNDARY resulting in radiation readings > 10 times normal..
Modes:
1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled COLD SHUTDOWN / REFUELING MATRIX COLD SHUTDOWN / REFUELING MATRIX
OP-DR-104-1001 Revision 08 Page 11 of 25 CHOOSE CURVE BASED ON INITIAL WATER TEMPERATURE Attachment A Time-to-Boil with Head On 0
5 10 15 1
10 100 1000 10000 Time Since Shutdown (hr)
Time-to-Boil (hr) 65 °F 100 °F 120 °F 180 °F
OP-DR-104-1001 Revision 08 Page 13 of 25 CHOOSE CURVE BASED ON INITIAL WATER TEMPERATURE Attachment C Time-to-Boil while Flooded to Flange Dryer/Separator In 0
5 10 15 20 1
10 100 1000 10000 Time Since Shutdown (hr)
Time-to-Boil (hr) 65 °F 100 °F 120 °F 180 °F
OP-DR-104-1001 Revision 08 Page 15 of 25 CHOOSE CURVE BASED ON INITIAL WATER TEMPERATURE Attachment E Time-to-Boil while Flooded to Flange Dryer/Separator Removed 0
5 10 15 20 1
10 100 1000 10000 Time Since Shutdown (hr)
Time-to-Boil (hr) 65 °F 100 °F 120 °F 180 °F
OP-DR-104-1001 Revision 08 Page 17 of 25 CHOOSE CURVE BASED ON INITIAL WATER TEMPERATURE Attachment G Time-to-Boil while Flooded Up Gates In; Dryer/Separator In (Dryer/Separator Pit not included) 0 10 20 30 40 50 1
10 100 1000 10000 Time Since Shutdown (hr)
Time-to-Boil (hr) 65 °F 100 °F 120 °F 180 °F
OP-DR-104-1001 Revision 08 Page 19 of 25 CHOOSE CURVE BASED ON INITIAL WATER TEMPERATURE Attachment I Time-to-Boil while Flooded Up Reactor and Dryer/Separator Pit 0
5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 1
10 100 1000 10000 Time Since Shutdown (hr)
Time-to-Boil (hr) 65 °F 100 °F 120 °F 180 °F
OP-DR-104-1001 Revision 08 Page 21 of 25 CHOOSE CURVE BASED ON INITIAL WATER TEMPERATURE Attachment K Time-to-Boil with Gates Out Cavity, Dry/Separator Pit, & Spent Fuel Pool 0
10 20 30 40 50 60 70 80 1
10 100 1000 10000 Time Since Shutdown (hr)
Time-to-Boil (hr) 65 °F 100 °F 120 °F 180 °F
ECCSOperating 3.5.1 Dresden 2 and 3 3.5.1-1 Amendment No.233/226 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND ISOLATION CONDENSER (IC) SYSTEM 3.5.1 ECCSOperating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.
APPLICABILITY:
MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.
ACTIONS
NOTE------------------------------------
LCO 3.0.4.b is not applicable to HPCI.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One Low Pressure Coolant Injection (LPCI) pump inoperable.
A.1 Restore LPCI pump to OPERABLE status.
30 days B.
One LPCI subsystem inoperable for reasons other than Condition A.
OR One Core Spray subsystem inoperable.
B.1 Restore low pressure ECCS injection/spray subsystem to OPERABLE status.
7 days C.
One LPCI pump in each subsystem inoperable.
C.1 Restore one LPCI pump to OPERABLE status.
7 days (continued)
ECCSOperating 3.5.1 Dresden 2 and 3 3.5.1-2 Amendment No.233/226 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D.
Required Action and associated Completion Time of Condition A, B, or C not met.
NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 3.
D.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E.
Two LPCI subsystems inoperable for reasons other than Condition C.
E.1 Restore one LPCI subsystem to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> F.
Required Action and associated Completion Time of Condition E not met.
F.1 Be in MODE 3.
AND F.2 Be in MODE 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours G.
HPCI System inoperable.
G.1 Verify by administrative means IC System is OPERABLE.
AND G.2 Restore HPCI System to OPERABLE status.
Immediately 14 days H.
One ADS valve inoperable.
H.1 Restore ADS valve to OPERABLE status.
14 days I.
Required Action and associated Completion Time of Condition G or H not met.
NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 3.
I.1 Be in Mode 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (continued)
ECCSOperating 3.5.1 Dresden 2 and 3 3.5.1-3 Amendment No.233/226 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME J.
Two or more ADS valves inoperable.
J.1 Be in Mode 3.
AND J.2 Reduce reactor steam dome pressure to
< 150 psig.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours K.
Two or more low pressure ECCS injection/spray subsystems inoperable for reasons other than Condition C or E.
OR HPCI System and one or more ADS valves inoperable.
OR One or more low pressure ECCS injection/spray subsystems inoperable and one or more ADS valves inoperable.
OR HPCI System inoperable and either one low pressure ECS injection/spray subsystem is inoperable or Condition C entered.
K.1 Enter LCO 3.0.3 Immediately
ECCSOperating 3.5.1 Dresden 2 and 3 3.5.1-4 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify each ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.3 Verify correct breaker alignment to the LPCI swing bus.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.4 Verify each recirculation pump discharge valve cycles through one complete cycle of full travel or is de-energized in the closed position.
In accordance with the Inservice Testing Program SR 3.5.1.5 Verify the following ECCS pumps develop the specified flow rate against a test line pressure corresponding to the specified reactor pressure.
TEST LINE PRESSURE NO.
CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray 4500 gpm 1
90 psig LPCI 9000 gpm 2
20 psig In accordance with the Inservice Testing Program (continued)
ECCSOperating 3.5.1 Dresden 2 and 3 3.5.1-5 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.6
NOTE--------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with reactor pressure 1005 and 920 psig, the HPCI pump can develop a flow rate 5000 gpm against a system head corresponding to reactor pressure.
In accordance with the Inservice Testing Program SR 3.5.1.7
NOTE--------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with reactor pressure 180 psig, the HPCI pump can develop a flow rate 5000 gpm against a system head corresponding to reactor pressure.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.8
NOTE--------------------
Vessel injection/spray may be excluded.
Verify each ECCS injection/spray subsystem actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program (continued)
ECCSOperating 3.5.1 Dresden 2 and 3 3.5.1-6 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.9
NOTE--------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.10 Verify each ADS valve actuator strokes when manually actuated.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.11 Verify automatic transfer capability of the LPCI swing bus power supply from the normal source to the backup source.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.12 Verify ADS pneumatic supply header pressure is > 80 psig.
In accordance with the Surveillance Frequency Control Program
SGT System 3.6.4.3 Dresden 2 and 3 3.6.4.3-1 Amendment No.233/226 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One SGT subsystem inoperable.
A.1 Restore SGT subsystem to OPERABLE status.
7 days B.
Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.
NOTE-------------
LCO 3.0.4.a is not applicable when entering MODE 3.
B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C.
Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.
NOTE------------
LCO 3.0.3 is not applicable.
C.1 Place OPERABLE SGT subsystem in operation.
OR Immediately (continued)
SGT System 3.6.4.3 Dresden 2 and 3 3.6.4.3-2 Amendment No.233/226 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.
(continued)
C.2.1 Suspend movement of recently irradiated fuel assemblies in secondary containment.
AND C.2.2 Initiate action to suspend OPDRVs.
Immediately Immediately D.
Two SGT subsystems inoperable in MODE 1, 2, or 3.
D.1 Restore one SGT subsystem to OPERABLE status.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> E.
Required Action and associated Completion Time of Condition D not met.
NOTE-------------
LCO 3.0.4.a is not applicable when entering MODE 3.
E.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> F.
Two SGT subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs.
F.1
NOTE--------
LCO 3.0.3 is not applicable.
Suspend movement of recently irradiated fuel assemblies in secondary containment.
AND F.2 Initiate action to suspend OPDRVs.
Immediately Immediately
SGT System 3.6.4.3 Dresden 2 and 3 3.6.4.3-3 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for 10 continuous hours with heaters operating.
In accordance with the Surveillance Frequency Control Program SR 3.6.4.3.2 Perform required SGT filter testing in accordance with the Ventilation Filter Testing Program (VFTP).
In accordance with the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or simulated initiation signal.
In accordance with the Surveillance Frequency Control Program