ML16012A349: Difference between revisions
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
||
| Line 16: | Line 16: | ||
=Text= | =Text= | ||
{{#Wiki_filter:Appendix D Scenario Outline Form ES-D-1 FCS 2015 NRC Simulator Scenario ES | {{#Wiki_filter:Appendix D Scenario Outline Form ES-D-1 FCS 2015 NRC Simulator Scenario ES | ||
-D-1 Outline Final As Run Facility: | -D-1 Outline Final As Run Facility: Fort Calhoun Station Scenario No.: | ||
Fort Calhoun Station Scenario No.: | |||
4 Op Test No.: | 4 Op Test No.: | ||
Dec 2015 NRC Examiners: | Dec 2015 NRC Examiners: | ||
| Line 30: | Line 29: | ||
Critical Tasks: | Critical Tasks: | ||
Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature > 110°F. (Event 2). | Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature > 110°F. (Event 2). | ||
OR Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions | OR Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions. (Event 5) Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level. (Event 7) Event No. Malf. No. Event Type* | ||
. (Event 5) Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level | |||
. (Event 7) | |||
Event No. | |||
Malf. No. | |||
Event Type* | |||
Event Description 1 +20 min R (ATCO) N (BOPO, CRS) | Event Description 1 +20 min R (ATCO) N (BOPO, CRS) | ||
Raise Power Using Control Rods to 7% per OP | Raise Power Using Control Rods to 7% per OP | ||
-2A, Plant Startup. Place Steam Dump and Bypass Valves in AUTO per OI | -2A, Plant Startup. Place Steam Dump and Bypass Valves in AUTO per OI | ||
-MS-1A. 2 +30 min C (ATCO, CRS) TS (CRS) Raw Water Pump Discharge Line Leak Upstream of HCV | -MS-1A. 2 +30 min C (ATCO , CRS) TS (CRS) Raw Water Pump Discharge Line Leak Upstream of HCV | ||
-2879A in the Auxiliary Building. | -2879A in the Auxiliary Building. 3 +45 min I (BOPO , CRS) TS (CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On Steam Generator RC | ||
3 +45 min I (BOPO, CRS) TS (CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On Steam Generator RC | |||
-2A. 4 +60 min C (ATCO, CRS) TS (CRS) Loss of Instrument Bus AI | -2A. 4 +60 min C (ATCO, CRS) TS (CRS) Loss of Instrument Bus AI | ||
-40A. Loss of Letdown and Pressurizer Level Control. | -40A. Loss of Letdown and Pressurizer Level Control. | ||
(Alternate Path Event 8) 5 +60 min M (ATCO, BOPO, CRS) Reactor Coolant Pump RC | (Alternate Path Event 8) 5 +60 min M (ATCO, BOPO, CRS) Reactor Coolant Pump RC | ||
-3A Trip. | -3A Trip. Automatic Reactor Trip Failure, Manual Reactor Trip Required. | ||
Automatic Reactor Trip Failure, Manual Reactor Trip Required. | |||
6 +65 min C (BOPO) Instrument Air Compressor CA | 6 +65 min C (BOPO) Instrument Air Compressor CA | ||
-1B and CA | -1B and CA-1C Trip. Bearing Cooling Water Pump AC | ||
-1C Trip. Bearing Cooling Water Pump AC | |||
-9B Trip. 7 +70 min M (ATCO, BOPO, CRS) Steam Line Break inside Containment on RC | -9B Trip. 7 +70 min M (ATCO, BOPO, CRS) Steam Line Break inside Containment on RC | ||
-2A @ 0.65% Severity on 5 Minute Ramp. | -2A @ 0.65% Severity on 5 Minute Ramp. | ||
| Line 60: | Line 51: | ||
Appendix D Scenario Outline Form ES-D-1 FCS 2015 NRC Simulator Scenario ES | Appendix D Scenario Outline Form ES-D-1 FCS 2015 NRC Simulator Scenario ES | ||
-D-1 Outline Final As Run Facility: | -D-1 Outline Final As Run Facility: Fort Calhoun Station Scenario No.: | ||
Fort Calhoun Station Scenario No.: | |||
4 Op Test No.: | 4 Op Test No.: | ||
Dec 2015 NRC Examiners: | Dec 2015 NRC Examiners: | ||
| Line 74: | Line 64: | ||
Critical Tasks: | Critical Tasks: | ||
Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature > 110°F. (Event 2). | Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature > 110°F. (Event 2). | ||
OR Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions | OR Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions. (Event 5) Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level. (Event 7) Event No. Malf. No. Event Type* | ||
. (Event 5) Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level | |||
. (Event 7) | |||
Event No. | |||
Malf. No. | |||
Event Type* | |||
Event Description 1 +20 min R (ATCO) N (BOPO, CRS) | Event Description 1 +20 min R (ATCO) N (BOPO, CRS) | ||
Raise Power Using Control Rods to 7% per OP | Raise Power Using Control Rods to 7% per OP | ||
-2A, Plant Startup. Place Steam Dump and Bypass Valves in AUTO per OI | -2A, Plant Startup. Place Steam Dump and Bypass Valves in AUTO per OI | ||
-MS-1A. 2 +30 min C (ATCO, CRS) TS (CRS) Raw Water Pump Discharge Line Leak Upstream of HCV | -MS-1A. 2 +30 min C (ATCO , CRS) TS (CRS) Raw Water Pump Discharge Line Leak Upstream of HCV | ||
-2879A in the Auxiliary Building. | -2879A in the Auxiliary Building. 3 +45 min I (BOPO , CRS) TS (CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On Steam Generator RC | ||
3 +45 min I (BOPO, CRS) TS (CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On Steam Generator RC | |||
-2A. 4 +60 min C (ATCO, CRS) TS (CRS) Loss of Instrument Bus AI | -2A. 4 +60 min C (ATCO, CRS) TS (CRS) Loss of Instrument Bus AI | ||
-40A. Loss of Letdown and Pressurizer Level Control. | -40A. Loss of Letdown and Pressurizer Level Control. | ||
(Alternate Path Event 8) 5 +60 min M (ATCO, BOPO, CRS) Reactor Coolant Pump RC | (Alternate Path Event 8) 5 +60 min M (ATCO, BOPO, CRS) Reactor Coolant Pump RC | ||
-3A Trip. | -3A Trip. Automatic Reactor Trip Failure, Manual Reactor Trip Required. | ||
Automatic Reactor Trip Failure, Manual Reactor Trip Required. | |||
6 +65 min C (BOPO) Instrument Air Compressor CA | 6 +65 min C (BOPO) Instrument Air Compressor CA | ||
-1B and CA | -1B and CA-1C Trip. Bearing Cooling Water Pump AC | ||
-1C Trip. Bearing Cooling Water Pump AC | |||
-9B Trip. 7 +70 min M (ATCO, BOPO, CRS) Steam Line Break inside Containment on RC | -9B Trip. 7 +70 min M (ATCO, BOPO, CRS) Steam Line Break inside Containment on RC | ||
-2A @ 0.65% Severity on 5 Minute Ramp. | -2A @ 0.65% Severity on 5 Minute Ramp. | ||
Revision as of 16:42, 8 July 2018
| ML16012A349 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/14/2015 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | |
| References | |
| Download: ML16012A349 (35) | |
Text
Appendix D Scenario Outline Form ES-D-1 FCS 2015 NRC Simulator Scenario ES
-D-1 Outline Final As Run Facility: Fort Calhoun Station Scenario No.:
4 Op Test No.:
Dec 2015 NRC Examiners:
Operators:
Initial Conditions:
MODE 2 at ~1% power
- RCS Boron is 959 ppm (by sample).
Turnover:
Continue in OP
-2A, Plant Startup and OI
-RR-1, Reactor Regulating System Operation to raise Reactor power to 7% power. When MODE 1 is entered, place Steam Dump and Bypass Valves in AUTO per OI-MS-1A, Main Steam System Operation.
Critical Tasks:
Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature > 110°F. (Event 2).
OR Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions. (Event 5) Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level. (Event 7) Event No. Malf. No. Event Type*
Event Description 1 +20 min R (ATCO) N (BOPO, CRS)
Raise Power Using Control Rods to 7% per OP
-2A, Plant Startup. Place Steam Dump and Bypass Valves in AUTO per OI
-MS-1A. 2 +30 min C (ATCO , CRS) TS (CRS) Raw Water Pump Discharge Line Leak Upstream of HCV
-2879A in the Auxiliary Building. 3 +45 min I (BOPO , CRS) TS (CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On Steam Generator RC
-2A. 4 +60 min C (ATCO, CRS) TS (CRS) Loss of Instrument Bus AI
-40A. Loss of Letdown and Pressurizer Level Control.
(Alternate Path Event 8) 5 +60 min M (ATCO, BOPO, CRS) Reactor Coolant Pump RC
-3A Trip. Automatic Reactor Trip Failure, Manual Reactor Trip Required.
6 +65 min C (BOPO) Instrument Air Compressor CA
-1B and CA-1C Trip. Bearing Cooling Water Pump AC
-9B Trip. 7 +70 min M (ATCO, BOPO, CRS) Steam Line Break inside Containment on RC
-2A @ 0.65% Severity on 5 Minute Ramp.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 3 Abnormal events (2
-4) 2 Major transients (1
-2) 1 EOPs entered/requiring substantive actions (1
-2) 0 EOP contingencies requiring substantive actions (0
-2) 2 Critical tasks (2
-3)
Appendix D Scenario Outline Form ES-D-1 FCS 2015 NRC Simulator Scenario ES
-D-1 Outline Final As Run Facility: Fort Calhoun Station Scenario No.:
4 Op Test No.:
Dec 2015 NRC Examiners:
Operators:
Initial Conditions:
MODE 2 at ~1% power
- RCS Boron is 959 ppm (by sample).
Turnover:
Continue in OP
-2A, Plant Startup and OI
-RR-1, Reactor Regulating System Operation to raise Reactor power to 7% power. When MODE 1 is entered, place Steam Dump and Bypass Valves in AUTO per OI-MS-1A, Main Steam System Operation.
Critical Tasks:
Upon Loss of Cooling to Reactor Coolant Pumps (RCPs), Trip the Reactor and Trip Associated RCP(s) within 5 minutes of Total Loss of CCW flow or Motor Bearing Temperatures > 203°F but prior to exceeding 210°F or within 5 minutes of CCW temperature > 110°F. (Event 2).
OR Manually Trip Reactor to meet Core Design Criteria of Lowering Reactor Power and Negative Startup Rate to Verify Reactivity Control Established During ATWS Event Prior to Exiting EOP-00, Standard Post Trip Actions. (Event 5) Isolate the Affected Steam Generator to Prevent Excess Plant Cooldown and Reactivity Additions Prior to Steam Generator Level = 0% Wide Range Level. (Event 7) Event No. Malf. No. Event Type*
Event Description 1 +20 min R (ATCO) N (BOPO, CRS)
Raise Power Using Control Rods to 7% per OP
-2A, Plant Startup. Place Steam Dump and Bypass Valves in AUTO per OI
-MS-1A. 2 +30 min C (ATCO , CRS) TS (CRS) Raw Water Pump Discharge Line Leak Upstream of HCV
-2879A in the Auxiliary Building. 3 +45 min I (BOPO , CRS) TS (CRS) Inadvertent Channel B Auxiliary Feedwater Actuation Signal On Steam Generator RC
-2A. 4 +60 min C (ATCO, CRS) TS (CRS) Loss of Instrument Bus AI
-40A. Loss of Letdown and Pressurizer Level Control.
(Alternate Path Event 8) 5 +60 min M (ATCO, BOPO, CRS) Reactor Coolant Pump RC
-3A Trip. Automatic Reactor Trip Failure, Manual Reactor Trip Required.
6 +65 min C (BOPO) Instrument Air Compressor CA
-1B and CA-1C Trip. Bearing Cooling Water Pump AC
-9B Trip. 7 +70 min M (ATCO, BOPO, CRS) Steam Line Break inside Containment on RC
-2A @ 0.65% Severity on 5 Minute Ramp.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 1 Malfunctions after EOP entry (1-2) 3 Abnormal events (2
-4) 2 Major transients (1
-2) 1 EOPs entered/requiring substantive actions (1
-2) 0 EOP contingencies requiring substantive actions (0
-2) 2 Critical tasks (2
-3)