ML23012A169
| ML23012A169 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 10/25/2022 |
| From: | Nuclear Management Co, Xcel Energy |
| To: | NRC/RGN-III/DORS/OB |
| Iskierka-Boggs T | |
| Shared Package | |
| ML21188A296 | List: |
| References | |
| Download: ML23012A169 (1) | |
Text
RG1 RS1 RA1 RU1 Note:
Note:
Note:
RG1.1 1
2 3
4 5
DEF RS1.1 1
2 3
4 5
DEF RA1.1 1
2 3
4 5
DEF RU1.1 1
2 3
4 5
2 3
4 5
DEF RS1.2 1
2 3
4 5
2 3
4 5
DEF RS1.3 1
2 3
4 5
DEF RA1.2 1
2 3
4 5
DEF RU1.2 1
2 3
4 5
DEF RA1.3 1
2 3
4 5
DEF RU1.3 1
2 3
4 5
DEF RA1.4 1
2 3
4 5
- 1.
RG2 RS2 RA2 RU2 2.
Note:
RA2.1 1
2 3
4 5
DEF RU2.1 1
2 3
4 5
- a.
RG2.1 1
2 3
4 5
DEF RS2.1 1
2 3
4 5
DEF RA2.2 1
2 3
4 5
- 1.
2.
AND
- b.
RA2.3 1
2 3
4 5
DEF RA3 Note:
RA3.1 1
2 3
4 5
DEF RA3.2 2
3 4
HG1 HS1 HA1 HU1 HG1.1 1
2 3
4 5
DEF HS1.1 1
2 3
4 5
DEF HA1.1 1
2 3
4 5
DEF HU1.1 1
2 3
4 5
- a.
a.
a.
b.
AND b.
b.
HA1.2 1
2 3
4 5
DEF HU1.2 1
2 3
4 5
- 1.
HU1.3 1
2 3
4 5
HU2 HU2.1 1
2 3
4 5
- a.
HU3 b.
Note:
HU3.1 1
2 3
4 5
DEF HU3.2 1
2 3
4 5
DEF a.
HU3.3 1
2 3
4 5
DEF HU3.4 1
2 3
4 5
DEF b.
HU3.5 1
2 3
4 5
DEF HU3.6 1
2 3
4 5
DEF HU4 Note:
HU4.1 1
2 3
4 5
DEF a.
AND b.
HU4.2 1
2 3
4 5
DEF a.
AND b.
AND c.
HU4.3 1
2 3
4 5
DEF HU4.4 1
2 3
4 5
DEF HA5 Note:
a.
HA5.1 2
3 4
b.
a.
a.
a.
AND c.
b.
b.
b.
HS6 HA6 Note:
HS6.1 1
2 3
4 5
HA6.1 1
2 3
4 5
DEF a.
AND b.
a.
HG7 HS7 HA7 HU7 HG7.1 1
2 3
4 5
DEF HS7.1 1
2 3
4 5
DEF HA7.1 1
2 3
4 5
DEF HU7.1 1
2 3
4 5
DEF EU1
- b. 1.
EU1.1 1
2 3
4 5
DEF 2.
Approval: PCR# 602000016746 Note:
- If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
- If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
System Malfunct.
Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode.
AND ATWS Communica-tions / Hazard Events Heat Capacity Limit (HCL) exceeded.
Control Room Indications Reactor Coolant System The occurrence of ANY of the following hazardous events:
Seismic event (Earthquake)
Internal or external flooding event High winds or tornado strike FIRE EXPLOSION Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode.
Damage to a loaded cask CONFINEMENT BOUNDARY.
Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16.
Loss of all but one AC power source to essential buses for 15 minutes or longer.
SS5.1
- 12 Emergency Diesel Generator 1R Reserve Transformer Reactivity control (Modes 1 and 2 only)
RPV water level RCS heat removal Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.
A tornado strike within the Plant PROTECTED AREA.
A FIRE within the Plant PROTECTED AREA or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
A FIRE within the Plant PROTECTED AREA or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
The Emergency Director should declare the Site Area Emergency promptly upon determining that 10 minutes has been exceeded, or will likely be exceeded.
An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel.
The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
FIRE potentially degrading the level of safety of the plant.
River level less than 902.4 ft el.
Receipt of multiple (more than 1) fire alarms or indications Control of ANY of the following key safety functions is not reestablished within 10 minutes.
Control Room evacuation resulting in transfer of plant control to alternate locations.
Inability to control a key safety function from outside the Control Room.
An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel.
Note:
The FIRE is located within ANY of the Table H2 plant rooms or areas The existence of a FIRE is not verified within 30-minutes of alarm receipt.
A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported by the Security Shift Supervisor.
AND AND Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Automatic or manual runback greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load ANY of the following transient events in progress.
Loss of all Vital DC power for 15 minutes or longer.
Suppression Pool Temperature Note:
Table S1 1AR Reserve Transformer Hazardous Event.
A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.
Thermal power oscillations greater than 10%
Indicated voltage is less than 110 VDC on ALL 125 VDC Vital DC buses for 15 minutes or longer.
EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
2R Auxiliary Transformer Movement of personnel within the Plant PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
1 2
SG8.1 1
SS8.1 ECCS actuation Note:
SA2 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
RPV Water Level Reactor scram Reactor Power Seismic event greater than Operating Basis Earthquake (OBE) as indicated by Annunciator OPERATIONAL BASIS EARTHQUAKE (6-C-13) received.
Prolonged loss of all offsite and all onsite AC power to essential buses.
The Emergency Director should declare the General Emergency promptly upon determining that 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> has been exceeded, or will likely be exceeded.
3 SA1.1 Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.
Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16 for 15 minutes or longer.
ANY condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier.
1AR Reserve Transformer AND AC power capability to essential buses 15 and 16 is reduced to a single power source (Table S1) for 15 minutes or longer.
SU1.1 1
2 1
ALERT SU1 Loss of all offsite AC power capability to essential buses for 15 minutes or longer.
Note:
3 Table S2 1R Reserve Transformer Loss of ALL offsite AC power capability (Table S2) to essential buses 15 and 16 for 15 minutes or longer.
A.
ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.
A.
Seismic event greater than OBE levels.
RM-17-453A or B Fuel Pool Monitoring System 2
3 Spent Fuel Pool low water level alarm Visual observation of an uncontrolled water level drop below a fuel pool skimmer surge tank inlet Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Loss of Power 100 mR/hr A-3 1027 RB W Stairway FPB Degradation Liquid Effluent Monitors RB Vent Effluent Monitor (Ch A or B) 6 E+05 µCi/sec UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
26 mR/hr Observation of water leakage into the drywell or the reactor building from piping penetrations surrounding the drywell.
A-1 1027 RB NE Low A-2 1027 RB N High UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
50 mR/hr 50 mR/hr SG8
- 2. RPV Water Level Primary Containment Pressure Suppression Pool Level 1
2 3
RPV Pressure An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
Reactor Power RPV Water Level An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
Indicated voltage is less than 110 VDC on ALL 125 VDC Vital DC buses for 15 minutes or longer.
RPV Pressure 3
Primary Containment Pressure Suppression Pool Level Note:
SU2 UNPLANNED loss of Control Room indications for 15 minutes or longer.
SA2.1 Offgas Pretreatment Radiation Monitor (RM-17-150A or RM-17-150B) high radiation alarm (4-A-12) received.
2 3
ANY of the following is successful in reducing reactor power to less than 4%:
A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Note:
RCS leakage for 15 minutes or longer.
Coolant sample activity greater than 0.2 µCi/gm dose equivalent I-131.
1 A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
RCS identified leakage greater than 25 gpm for 15 minutes or longer.
SU4.3 1
2 3
3 SU3 Reactor coolant activity greater than Technical Specification allowable limits.
SU3.1 1
2 Suppression Pool Temperature AND SS5 All manual actions to shutdown the reactor are not successful in reducing reactor power to less than 4%.
Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.
EITHER of the following conditions exist:
Reactor vessel water level cannot be restored and maintained above -150 (Minimum Steam Cooling RPV Water Level).
AND AND An automatic or manual scram did not reduce reactor power to less than 4%.
1 River level greater than 919 ft el.
SA5 Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the main control boards are not successful in shutting down the reactor.
An automatic or manual scram did not reduce reactor power to less than 4%.
Manual actions taken at the main control boards are not successful in reducing reactor power to less than 4%.
2 3
1 A.
A.
Coolant activity is greater than 300
µCi/gm dose equivalent I-131.
RPV water level cannot be restored and maintained above -126 in. or cannot be determined.
RPV water level cannot be restored and maintained above -126 in. or cannot be determined.
A.
Entry into the room or area is prohibited or impeded.
If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Dose rate greater than 15 mR/hr in ANY of the following areas:
Secondary Alarm Station (by survey)
An UNPLANNED event results in radiation levels that prohibit or impede access to any of the Table H1 plant rooms or areas.
A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.
Notification of a credible security threat directed at MNGP.
A validated notification from the NRC providing information of an aircraft threat.
If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
River level greater than 918 ft el.
The FIRE is located within ANY of the Table H2 plant rooms or areas.
A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:
Release of a toxic, corrosive, asphyxiant or flammable gas into any of the Table H1 plant rooms or areas.
Report from the field (i.e., visual observation)
Control Room (A-20 Control Room Low Range)
Central Alarm Station (by survey)
UNPLANNED rapid drop in primary containment pressure following primary containment pressure rise Primary containment pressure response not consistent with LOCA conditions.
A.
OR B.
OR C.
- 2. RPV Water Level Primary containment pressure greater than 1.84 psig due to RCS leakage.
Greater than or equal to 6% hydrogen and greater than or equal to 5% oxygen in Drywell or Torus Heat Capacity Limit (HCL) exceeded
SAMG entry is required.
A.
A.
SG1 SS1 Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer.
SA1 Note:
Note:
The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
2 ANY condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.
- 3. Primary Containment Isolation Failure A.
A.
Containment High Range Rad (Drywell Radiation) monitor reading greater than 6.2 E+01 R/hr.
- 3. RCS Leak Rate
- 4. Primary Containment Radiation Containment High Range Rad (Drywell Radiation) monitor reading greater than 1.5 E+03 R/hr.
- 5. Emergency Director Judgement
- 5. Emergency Director Judgement Max Normal Operating Area Radiation Level A.
A.
SAMG entry is required.
ALERT UNUSUAL EVENT Lowering of spent fuel pool level to 15.25'.
Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.
NOTE
- The Emergency Director should declare the Emergency Action Level promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
- If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Stack Effluent Monitor (Ch A or B)
TBNWS 9 E+04 cpm Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
- The pre-calculated effluent monitor values presented in EAL RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
- The pre-calculated effluent monitor values presented in EAL RS1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.
Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.
8 E+05 µCi/sec
- The pre-calculated effluent monitor values presented in EAL RA1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Power Operation Startup Refueling Defueled Modes:
1 2
5 DEF 4
3 Cold Shutdown Hot Shutdown Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NUE.
Receipt of a single fire alarm (i.e., no other indications of a FIRE).
SS8 Field verification of a single fire alarm SS1.1 1
2 2
AND EITHER of the following:
Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16 for 15 minutes or longer.
3 Loss of all AC and Vital DC power sources for 15 minutes or longer.
The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
3 Reactor vessel water level cannot be restored and maintained above -150" (Minimum Steam Cooling RPV Water Level).
4 8
Table E1 (Applicable to HSMs 1A/B-5A/B)
Total Dose Rate (Neutron + Gamma mR/hr) 800 200 40 Location of Dose Rate 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.
Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading greater than any of the values listed in Table E1, E2, or E3 as applicable to the HSM.
On-Contact HSM-H Door Centerline On-Contact End shield wall exterior UNISOLABLE break in ANY of the following: MSL; HPCI; RWCU; RCIC as indicated by high flow/temperature isolation setpoints Emergency RPV Depressurization A.
OR OR B.
GENERAL EMERGENCY ALERT FG1 1
2 3
FS1 1
2 3
FA1 1
2 3
GENERAL EMERGENCY SITE AREA EMERGENCY SITE AREA EMERGENCY Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
Loss of any two barriers and Loss or Potential Loss of the third barrier.
Loss or Potential Loss of any two barriers.
FISSION PRODUCT BARRIER REFERENCE Fuel Cladding Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
- 1. RCS Activity
- 1. Primary Containment Pressure A.
Primary containment pressure greater than 56 psig
- 1. Primary Containment Conditions SU2.1 1
2 3
The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Max Normal Operating Temperature UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.
- 3. N/A
- 4. Primary Containment Radiation SU5 Automatic or manual scram fails to shutdown the reactor.
Note:
SU4 SU3.2 2
3 SU4.1 1
2 3
SU6.3 1
2 An initial automatic or manual scram did not reduce reactor power to less than 4%.
AND AND SU5.1 Other events with similar hazard characteristics as determined by the Shift Manager SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
SU6 Loss of all onsite or offsite communications capabilities.
1 1
2 3
1 Manual scram pushbuttons Mode switch to shutdown Alternate Rod Insertion (ARI)
Subsequent automatic scram SA5.1 2R Auxiliary Transformer HOT CONDITIONS River level less than 900.5 ft el.
Commercial Telephones Direct Dedicated Telephones Federal Telecommunications System (FTS)
Commercial Telephones 3
SA9.1 Loss of ALL of the following NRC communications methods:
EITHER of the following:
Radio/Receiver Transmitter 1
Loss of ALL of the following onsite communication methods:
Commercial Telephones Plant Telephones Portable radios Plant PA System Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode.
SU6.2 RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
SU4.2 1
Loss of ALL of the following Offsite Response Organization (ORO) communications methods:
SU6.1 Leakage from the RCS to a location outside primary containment greater than 25 gpm for 15 minutes or longer.
2 3
A.
B.
Intentional primary containment venting per EOPs C.
Max Safe Operating Temperature UNUSUAL EVENT The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Note:
OR A.
- 5. Emergency Director Judgement A.
ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.
UNISOLABLE primary system leakage that results in exceeding Control Room indication of EITHER of the following:
Max Safe Operating Area Radiation Level UNISOLABLE primary system leakage that results in exceeding Control Room indication of EITHER of the following:
OR OR
- 4. Primary Containment Radiation Containment High Range Rad (Drywell Radiation) monitor reading greater than 3.3 E+04 R/hr.
Reading on the following radiation monitor greater than the reading shown for 15 minutes or longer:
Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Reading on ANY of the following effluent radiation monitors greater than the listed values for 60 minutes or longer:
Discharge Canal 2000 cps Service Water 700 cps Gaseous Eflluent Monitors Stack Effluent Monitor (Ch A or B) 4 E+05 µCi/sec RB Vent Effluent Monitor (Ch A or B) 3 E+04 µCi/sec Liquid Effluent Monitors Discharge Canal 900 cps Gaseous Effluent Monitors Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY.
Stack Effluent Monitor (Ch A or B) 8 E+06 µCi/sec Service Water 300 cps TBNWS 4 E+04 cpm Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY.
Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors:
RM-17-452A RB Vent Exhaust Plenum Monitor Ch A 26 mR/hr Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
The Emergency Director should declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of exposure.
Monitor Alarm/Trip A-1 1027 RB NE Low 20 mR/hr A-2 1027 RB N High 200 mR/hr Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
A-3 1027 RB W Stairway Spent fuel pool level at 15.25'.
Significant lowering of water level above, or damage to, irradiated fuel.
UNPLANNED loss of water level above irradiated fuel.
HOSTILE ACTION within the Plant PROTECTED AREA.
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.
Confirmed SECURITY CONDITION or threat.
RM-17-452B RB Vent Exhaust Plenum Monitor Ch B A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.
RM-17-453A Fuel Pool Radiation Monitor Ch A Lowering of spent fuel pool level to 24.75'.
Uncovery of irradiated fuel in the REFUELING PATHWAY.
RM-17-453B Fuel Pool Radiation Monitor Ch B GENERAL EMERGENCY SITE AREA EMERGENCY A.
ANY condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.
A.
A.
3 feet from the HSM Surface On-Contact Outside HSM Door Centerline On-Contact End shield wall exterior Restoration of at least one AC essential bus in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.
SG1.1 1
HOSTILE ACTION resulting in loss of physical control of the facility.
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 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 warrant declaration of an Alert.
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
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.
Table E3 (Applicable to HSMs 11A/B-15A/B)
Location of Dose Rate On-Contact HSM-H Front Bird Screen Total Dose Rate (Neutron + Gamma mR/hr) 1300 Abnormal Rad Release Rad Effluent Hazards Stack Effluent Monitor (Ch A or B) 8 E+07 µCi/sec Damage to spent fuel has occurred or is IMMINENT.
Reactivity Control RPV water level RCS heat removal Effluents / Release Rates Natural & Destructive Phenomenon Fire or Explosion Toxic, Corrosive, Asphyxiant or Flammable Gases Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported by the Security Shift Supervisor.
Spent Fuel Pool Level /
In Plant Rad Spent fuel pool level cannot be restored to at least 15.25' for 60 minutes or longer.
Reading on the following radiation monitor greater than the reading shown for 15 minutes or longer:
Control Room Evacuation Security EITHER of the following has occurred:
ANY of the following safety functions cannot be controlled or maintained.
Spent fuel pool level cannot be restored to at least 15.25' for 60 minutes or longer.
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY.
ISFSI Events Cask Confine.
Boundary Shift Manager /
Emergency Director Judgment Table E2 (Applicable to HSMs 6A/B-10A/B)
Location of Dose Rate On-Contact HSM or HSM-H Front Surface On-Contact HSM or HSM-H Door Centerline On-Contact End shield wall exterior Total Dose Rate (Neutron + Gamma mR/hr) 1400 200 40 Table H2 All All EFT Building All All Control and Administration Building Diesel Generator Building Diesel Fuel Oil Transfer House Intake Structure Control Room, Cable Spreading Room, and Battery Rooms All Reactor Building HPCI Building Turbine Building Room(s)/Area(s) with Safety Equipment Building Name All All Modes 1, 2, 3 F:\\EALM 5790-101-02\\5790-101-02 MONTICELLO NUCLEAR GENERATING PLANT 5790-101-02 TITLE:
EMERGENCY ACTION LEVEL MATRIX Revision 14 Page 1 of 2 (Reactor Building 896' SW and SE)
MCC-133 Room (Turbine Building 911' East)
MVP Room Rooms BOTH RHR Rooms (Turbine Building 908' West)
Applicable Modes 3 / 4 2 / 3 Table H1
RG1 RS1 RA1 RU1 CG1 CS1 CA1 CU1 Note:
Note:
Note:
Note:
CG1.1 4
5 CS1.1 4
5 CA1.1 4
5 CU1.1 4
5 a.
- a.
AND AND Note:
Note:
Note:
b.
- b.
RG1.1 1
2 3
4 5
DEF RS1.1 1
2 3
4 5
DEF RA1.1 1
2 3
4 5
DEF RU1.1 1
2 3
4 5
DEF CG1.2 4
5 CS1.2 4
5 CA1.2 4
5 CU1.2 4
5
- a.
- a.
- a.
- a.
AND AND AND AND b.
b.
- b.
- b.
CS1.3 4
5 RG1.2 1
2 3
4 5
DEF RS1.2 1
2 3
4 5
DEF AND a.
- c.
AND
- b.
RG1.3 1
2 3
4 5
DEF RS1.3 1
2 3
4 5
DEF RA1.2 1
2 3
4 5
DEF RU1.2 1
2 3
4 5
DEF RA1.3 1
2 3
4 5
DEF RU1.3 1
2 3
4 5
DEF RA1.4 1
2 3
4 5
DEF CA2 CU2 Note:
Note:
CA2.1 4
5 DEF CU2.1 4
5 DEF
- a.
RG2 RS2 RA2 RU2 AND Note:
RA2.1 1
2 3
4 5
DEF RU2.1 1
2 3
4 5
- b.
- a.
RG2.1 1
2 3
4 5
DEF RS2.1 1
2 3
4 5
DEF RA2.2 1
2 3
4 5
DEF AND
- b.
CU4 Note:
CU4.1 4
5 RA2.3 1
2 3
4 5
DEF CA3 CU3 RA3 Note:
Note:
Note:
CA3.1 4
5 CU3.1 4
5 RA3.1 1
2 3
4 5
DEF CA3.2 4
5 CU3.2 4
5 RA3.2 2
3 4
HG1 HS1 HA1 HU1 CU5 HG1.1 1
2 3
4 5
DEF HS1.1 1
2 3
4 5
DEF HA1.1 1
2 3
4 5
DEF HU1.1 1
2 3
4 5
DEF CU5.1 4
5 DEF a.
AND b.
HA1.2 1
2 3
4 5
DEF HU1.2 1
2 3
4 5
- 1.
HU1.3 1
2 3
4 5
DEF CU5.2 4
HU2 HU2.1 1
2 3
4 5
DEF CU5.3 4
5 DEF HU3 Note:
CA6 HU3.1 1
2 3
4 5
DEF Note:
HU3.2 1
2 3
4 5
DEF CA6.1 4
5 HU3.3 1
2 3
4 5
- a.
HU3.4 1
2 3
4 5
DEF HU3.5 1
2 3
4 5
DEF AND HU3.6 1
2 3
4 5
DEF b.
- 1.
HU4 Note:
2.
HU4.1 1
2 3
4 5
DEF a.
AND b.
HU4.2 1
2 3
4 5
DEF a.
AND b.
AND c.
HU4.3 1
2 3
4 5
DEF HU4.4 1
2 3
4 5
DEF HA5 Note:
HA5.1 2
3 4
a.
AND b.
HS6 HA6 Note:
HS6.1 1
2 3
4 5
HA6.1 1
2 3
4 5
DEF a.
AND b.
HG7 HS7 HA7 HU7 HG7.1 1
2 3
4 5
DEF HS7.1 1
2 3
4 5
DEF HA7.1 1
2 3
4 5
DEF HU7.1 1
2 3
4 5
DEF EU1 EU1.1 1
2 3
4 5
DEF Approval: PCR# 602000016746 EITHER of the following:
Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode.
Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode.
Hazardous Events Cold SD/
Refuel System Malfunct.
Comm-unications COLD CONDITIONS
- If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
- If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode.
AND SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.
SECONDARY CONTAINMENT: SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System, the Offgas Dilution Fans, and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e.,
this may include violent acts between individuals in the owner controlled area).
VISIBLE DAMAGE: Damage to a SAFETY SYSTEM that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.
SITE BOUNDARY: For Dose Assessment and Protective Action Recommendation purposes the SITE BOUNDARY is the closest distance at which members of the public would be exposed to a radioactive release. The SITE BOUNDARY for liquid releases of radioactive material is defined in ODCM-02.01 (LIQUID EFFLUENTS). The SITE BOUNDARY for gaseous releases of radioactive material is defined in ODCM-03.01 (GASEOUS EFFLUENTS).
INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool, or fuel transfer canal PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. In areas where two fences are present, the inner fence is designated as the Protected Area barrier.
ISFSI PROTECTED AREA: The area surrounding the Independent Spent Fuel Storage Installation encompassed by the double chain link fence surrounding the ISFSI as defined in the Security Plan; the ISFSI Protected Area is excluded from the Plant Protected Area.
FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
OWNER CONTROLLED AREA: The OCA boundaries consist of the plant property enclosed by a three strand barbed wire fence and a posted boundary on the Wright County side of the river.
Fire or Explosion FIRE potentially degrading the level of safety of the plant.
The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:
Report from the field (i.e., visual observation)
Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm The FIRE is located within ANY of the Table H2 plant rooms or areas Receipt of a single fire alarm (i.e., no other indications of a FIRE).
The FIRE is located within ANY of the Table H2 plant rooms or areas.
The existence of a FIRE is not verified within 30-minutes of alarm receipt.
A FIRE within the Plant PROTECTED AREA or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
A FIRE within the Plant PROTECTED AREA or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
DEFINITIONS CONFINEMENT BOUNDARY: The barrier(s) between areas containing spent fuel and the environment once the spent fuel is processed for dry storage.
IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
Seismic event (Earthquake)
Internal or external flooding event High winds or tornado strike FIRE EXPLOSION Inability to control a key safety function from outside the Control Room.
Control Room evacuation resulting in transfer of plant control to alternate locations.
River level greater than 919 ft el.
River level less than 900.5 ft el.
Other events with similar hazard characteristics as determined by the Shift Manager EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.
Movement of personnel within the Plant PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
River level greater than 918 ft el.
River level less than 902.4 ft el.
Commercial Telephones Direct Dedicated Telephones UNPLANNED rise in RCS temperature to greater than 212 °F for greater than the duration specified in Table C2.
UNPLANNED RCS pressure rise greater than 10 psig.
ISFSI Events Cask Confine.
Boundary Damage to a loaded cask CONFINEMENT BOUNDARY.
Temp. Control Shift Manager /
Emergency Director Judgment Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.
Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert.
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NUE.
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 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.
Toxic, Corrosive, Asphyxiant or Flammable Gases Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Entry into the room or area is prohibited or impeded.
The Emergency Director should declare the Site Area Emergency promptly upon determining that 10 minutes has been exceeded, or will likely be exceeded.
An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel.
The occurrence of ANY of the following hazardous events:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Radio/Receiver Transmitter Federal Telecommunications System (FTS)
Commercial Telephones Loss of ALL of the following NRC communications methods:
Loss of ALL of the following Offsite Response Organization (ORO) communications methods:
Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Dose rate greater than 15 mR/hr in ANY of the following areas:
Secondary Alarm Station (by survey)
An UNPLANNED event results in radiation levels that prohibit or impede access to any of the Table H1 plant rooms or areas.
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
Control Room (A-20 Control Room Low Range)
Central Alarm Station (by survey)
A-2 1027 RB N High 200 mR/hr A-3 1027 RB W Stairway Plant PA System Loss of Power Refueling Floor radiation monitor reading greater than 3 R/hr.
UNPLANNED rise in drywell floor or equipment drain sump levels of sufficient magnitude to indicate core uncovery.
ANY indication from the Containment Challenge Table C1.
Liquid Effluent Monitors 4 E+05 µCi/sec Service Water 300 cps TBNWS RB Vent Effluent Monitor (Ch A or B) 3 E+04 µCi/sec Inventory Control NOTE
- The Emergency Director should declare the Emergency Action Level promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
- If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Gaseous Eflluent Monitors Reading on ANY of the following effluent radiation monitors greater than the listed values for 60 minutes or longer:
Stack Effluent Monitor (Ch A or B) 8 E+06 µCi/sec Stack Effluent Monitor (Ch A or B) 8 E+05 µCi/sec 100 mR/hr RM-17-452B RB Vent Exhaust Plenum Monitor Ch B 26 mR/hr Commercial Telephones Plant Telephones Portable radios Loss of ALL of the following onsite communication methods:
Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer.
Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer.
The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
UNPLANNED rise in RCS temperature.
Inability to maintain the plant in cold shutdown.
The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Loss of Vital DC power for 15 minutes or longer.
1R Reserve Transformer 1AR Reserve Transformer 2R Auxiliary Transformer
- 11 Emergency Diesel Generator Loss of all onsite or offsite communications capabilities.
UNPLANNED rise in RCS temperature to greater than 212 °F.
Indicated voltage is less than 110 VDC on required 125 VDC Vital DC buses for 15 minutes or longer.
- 12 Emergency Diesel Generator Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.
Table S1 Loss of ALL offsite and ALL onsite AC Power to essential buses 15 and 16 for 15 minutes or longer.
SECONDARY CONTAINMENT established.
RPV level less than -126 in. (TAF).
RPV level cannot be monitored for 30 minutes or longer.
Loss of RPV inventory affecting fuel clad integrity with containment challenged.
The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Loss of all but one AC power source to essential buses for 15 minutes or longer.
The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Core uncovery is indicated by EITHER of the following:
RPV level cannot be monitored for 15 minutes or longer.
UNPLANNED rise in drywell floor or equipment drain sump levels due to a loss of RPV inventory.
Core uncovery is indicated by EITHER of the following:
RPV level cannot be monitored for 30 minutes or longer.
Refueling Floor radiation monitor reading greater than 3 R/hr.
ANY indication from the Containment Challenge Table C1.
RPV level less than -47 in.
RPV level cannot be monitored.
UNPLANNED rise in drywell floor or equipment drain sump levels of sufficient magnitude to indicate core uncovery.
GENERAL EMERGENCY UNPLANNED rise in drywell floor or equipment drain sump levels.
AC power capability to Essential buses 15 and 16 is reduced to a single power source (Table S1) for 15 minutes or longer.
Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
UNUSUAL EVENT GENERAL EMERGENCY ALERT SITE AREA EMERGENCY UNPLANNED loss of reactor coolant results in RPV level less than a procedurally required lower limit for 15 minutes or longer.
The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
Loss of RPV inventory affecting core decay heat removal capability.
The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
UNPLANNED loss of RPV inventory for 15 minutes or longer.
The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
RPV level less than -126 in. (TAF) for 30 minutes or longer.
SECONDARY CONTAINMENT not established.
Loss of RPV inventory.
Loss of RPV inventory as indicated by level less than -47 in..
ALERT UNUSUAL EVENT SITE AREA EMERGENCY Stack Effluent Monitor (Ch A or B) 8 E+07 µCi/sec Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.
Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.
Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.
Discharge Canal 900 cps Liquid Effluent Monitors RB Vent Effluent Monitor (Ch A or B) 6 E+05 µCi/sec Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY.
Gaseous Effluent Monitors TBNWS 9 E+04 cpm Stack Effluent Monitor (Ch A or B)
- The pre-calculated effluent monitor values presented in EAL RA1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Reading on the following radiation monitor greater than the reading shown for 15 minutes or longer:
Reading on the following radiation monitor greater than the reading shown for 15 minutes or longer:
Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Discharge Canal 2000 cps Power Operation Modes:
Release of a toxic, corrosive, asphyxiant or flammable gas into any of the Table H1 plant rooms or areas.
Startup Hot Shutdown Cold Shutdown Refueling Defueled RM-17-453A or B Fuel Pool Monitoring System 2
DEF RCS heat removal An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel.
Control of ANY of the following key safety functions is not reestablished within 10 minutes.
Reactivity control (Modes 1 and 2 only)
RPV water level A tornado strike within the Plant PROTECTED AREA.
Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.
3 4
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.
A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported by the Security Shift Supervisor.
A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported by the Security Shift Supervisor.
A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.
5 4 E+04 cpm Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY.
1 Confirmed SECURITY CONDITION or threat.
A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.
EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
Notification of a credible security threat directed at MNGP.
A validated notification from the NRC providing information of an aircraft threat.
Seismic event greater than OBE levels.
Seismic event greater than Operating Basis Earthquake (OBE) as indicated by Annunciator OPERATIONAL BASIS EARTHQUAKE (6-C-13) received.
Hazardous Event.
UNPLANNED loss of water level above irradiated fuel.
20 mR/hr Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of exposure.
Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors:
50 mR/hr Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY.
Uncovery of irradiated fuel in the REFUELING PATHWAY.
Spent fuel pool level cannot be restored to at least 15.25' for 60 minutes or longer.
Service Water 700 cps A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.
UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
Spent Fuel Pool low water level alarm Visual observation of an uncontrolled water level drop below a fuel pool skimmer surge tank inlet Observation of water leakage into the drywell or the reactor building from piping penetrations surrounding the drywell.
UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
Monitor Alarm/Trip A-1 1027 RB NE Low A-3 1027 RB W Stairway A-1 1027 RB NE Low A-2 1027 RB N High RM-17-453A Fuel Pool Radiation Monitor Ch A Hazards Security Natural &
Destructive Phenomenon Damage to spent fuel has occurred or is IMMINENT.
Reactivity Control RPV water level HOSTILE ACTION resulting in loss of physical control of the facility.
HOSTILE ACTION within the Plant PROTECTED AREA.
Abnormal Rad Release Rad Effluent
- The pre-calculated effluent monitor values presented in EAL RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
- The pre-calculated effluent monitor values presented in EAL RS1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Control Room Evacuation Effluents / Release Rates Spent Fuel Pool Level / In Plant Rad Spent fuel pool level cannot be restored to at least 15.25' for 60 minutes or longer.
Spent fuel pool level at 15.25'.
EITHER of the following has occurred:
ANY of the following safety functions cannot be controlled or maintained.
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
RCS heat removal Significant lowering of water level above, or damage to, irradiated fuel.
The Emergency Director should declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
Lowering of spent fuel pool level to 15.25'.
Lowering of spent fuel pool level to 24.75'.
Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
RM-17-453B Fuel Pool Radiation Monitor Ch B 50 mR/hr RM-17-452A RB Vent Exhaust Plenum Monitor Ch A 26 mR/hr Table E1 (Applicable to HSMs 1A/B-5A/B)
Table E2 (Applicable to HSMs 6A/B-10A/B)
Table E3 (Applicable to HSMs 11A/B-15A/B)
Location of Dose Rate Total Dose Rate (Neutron + Gamma mR/hr)
Location of Dose Rate Total Dose Rate (Neutron + Gamma mR/hr)
Total Dose Rate (Neutron + Gamma mR/hr)
Location of Dose Rate On-Contact Outside HSM Door Centerline On-Contact End shield wall exterior On-Contact HSM or HSM-H Front Surface 1400 On-Contact HSM-H Front Bird Screen 1300 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading greater than any of the values listed in Table E1, E2, or E3 as applicable to the HSM.
On-Contact HSM or HSM-H Door Centerline 200 On-Contact HSM-H Door Centerline 4
On-Contact End shield wall exterior 40 On-Contact End shield wall exterior 8
800 40 200 3 feet from the HSM Surface Table H2 All All EFT Building All All Control and Administration Building Diesel Generator Building Diesel Fuel Oil Transfer House Intake Structure Control Room, Cable Spreading Room, and Battery Rooms All Reactor Building HPCI Building Turbine Building Room(s)/Area(s) with Safety Equipment Building Name All All Intact N/A 60 minutes*
- If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
Table C2: RCS Heat-Up Duration Thresholds RCS SECONDARY CONTAINMENT Heat-up Duration Not Established 0 minutes Established 20 minutes*
Not intact Modes 4, 5, DEF F:\\EALM 5790-101-02\\5790-101-02 MONTICELLO NUCLEAR GENERATING PLANT 5790-101-02 TITLE:
EMERGENCY ACTION LEVEL MATRIX Revision 14 Page 2 of 2 (Reactor Building 896' SW and SE)
MCC-133 Room (Turbine Building 911' East)
MVP Room Rooms BOTH RHR Rooms (Turbine Building 908' West)
Applicable Modes 3 / 4 2 / 3 Table H1 Containment Challenge Table C1
- SECONDARY CONTAINMENT not established*
- Greater than or equal to 6% hydrogen and greater than or equal to 5% oxygen in primary containment
- If SECONDARY CONTAINMENT is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
- Two or more Reactor Building areas exceed Max Safe Radiation Levels (C.5-CSD-2 Table R (Mode 4) or C.5-RF Table R (Mode 5))
- UNPLANNED rise in containment pressure resulting in Drywell Pressure greater than 1.84 psig
ENTRY CONDITIONS (MODE 1, 2, 3)DW or torus hydrogenabove 2%Torus temperatureabove 90°FTorus levelbelow -inchesORabove +3.0 inchesDW temperatureabove 135°FDW pressureabove 1.84 psig HYDROGEN TORUS TEMPERATURE TT-1
! Exceeding NPSH or Vortex Limit may damage RHR (Figs N2, V; SPDS 74, 77)
Maintain torus temperature below 90°F using torus cooling.
TT-2 IF THEN Cannot maintain torus temperature below 90°F Go to r TT-3 Start all available torus cooling.
p Do not use RHR pumps you need for core cooling.
(Scram, enter C.5-1100)
Torus temperature reaches 110°F BEFORE TT-4 Maintain torus temperature below Fig M, Heat Capacity Limit (SPDS 78).
TT-6 IF THEN Cannot maintain torus temperature below Fig M, Heat Capacity Limit (SPDS 78)
- 1. IF......... you are not in C.5-2001 or C.5-2006, AND.....permitted by C.5-1100 / 2007 Pressure branch, THEN...lower RPV pressure to stay below limit.
p Minimum pressure for adequate HPCI/RCIC flow is 150 psig.
p OK to exceed 100°F/hr cooldown.
- 2. IF......... still cannot stay below Fig M, Heat Capacity Limit (SPDS 78),
THEN...RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
14 TORUS LEVEL Maintain torus level -4.0 to +3.0 inches.
TL-1 IF THEN Low Level:
Cannot maintain level above -4.0 inches Go to y High Level:
Cannot maintain level below +3.0 inches Go to t Maintain DW temperature below 135°F using DW coolers.................................................... B.08.16-05 p DW temperature affects RPV water level indication.
Check Detail A.
DWT-1 IF THEN Cannot maintain DW temperature below 135°F Go to u DWT-2 Start all available DW cooling.
p OK to defeat trips.........C.5-3503 17 DW TEMPERATURE (Scram; enter C.5-1100)
DW temperature reaches 281°F BEFORE DWT-3 BEFORE THEN DWT-5 DW pressure drops below 0 psig Stop DW sprays................................C.5-3502
! Exceeding NPSH or Vortex Limit may damage RHR (Figs N2, V; SPDS 74, 77).
IF......... DW temperature below Fig K, DW Spray Limit (SPDS 79),
AND.....torus level below 4.2 ft, THEN...1. Verify recirc pumps tripped,
- 2. Place DW cooling FAN CONTROL switches in OFF.
- 3. Start DW sprays................................................................C.5-3502 p Do not use pumps you need for core cooling.
p Reducing primary containment pressure affects margin to NPSH limits.
p DW temperature affects RPV water level indication. Check Detail A.
DWT-6 IF THEN Cannot restore and maintain DW temperature below 338°F RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
DW/TORUS PRESSURE Maintain DW and torus pressure below 1.84 psig using SBGT............................................................B.04.01-05 PCP-1 IF THEN Cannot maintain pressure below 1.84 psig Go to i 18 BEFORE THEN PCP-2 DW pressure drops below 0 psig Stop DW sprays...................................C.5-3502 Torus pressure drops below 0 psig Stop torus sprays.................................C.5-3502 Primary containment pressure reduction is required to:
- Restore and maintain adequate core cooling OR
- Reduce the total offsite dose Vent the primary containment..............C.5-3505 p OK to exceed release rate limits.
p OK to defeat isolations.
p Reducing primary containment pressure affects margin to NPSH limits.
Cannot maintain DW or torus pressure above -2.0 psig Restore and maintain DW and torus pressures above -2.0 psig by adding nitrogen (B.04.01-05).
p OK to exceed release rate limits.
p OK to defeat isolations.
C.5-1200 PRIMARY CONTAINMENT CONTROL (MODE 1, 2, 3)
Enter C.5-1100 while continuing here.
SCRAM:
+
DWT-4 Enter C.5-1100 while continuing here.
SCRAM:
+
TT-5 Monitor hydrogen and oxygen in DW and torus...........................C.5-3501 H-1 IF THEN H2/O2 analyzers do not work Have DW and torus sampled for hydrogen and oxygen.......... A.2-415 DW or torus hydrogen reaches 4%
Leave all EOPs
=
Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
EAL FG1 EAL FPB SS5.1 EAL FPB EAL FPB SU4.1 SU4.2 (Torus spray)
DW pressure reaches 12 psig BEFORE PCP-3 PCP-4
! Exceeding NPSH or Vortex Limit may damage RHR (Figs N2, V; SPDS 74, 77)
! Drywell pressure less than 8.6 psig may not provide adequate NPSH to prevent ECCS pump cavitation.
IF......... torus level below 14.1 ft, THEN...start torus sprays................................................................... C.5-3502 p Do not use pumps you need for core cooling.
p Reducing primary containment pressure affects margin to NPSH limits.
PCP-5 IF THEN Cannot maintain DW pressure below 12 psig Go to o IF......... DW temperature below Fig K, DW Spray Limit (SPDS 79),
AND.....torus level below 4.2 ft, THEN...1. Verify recirc pumps tripped,
- 2. Place DW cooling FAN CONTROL switches in OFF.
- 3. Start DW sprays................................................................C.5-3502 p Do not use pumps you need for core cooling.
p Reducing primary containment pressure affects margin to NPSH limits.
PCP-6 IF THEN Cannot maintain DW pressure within Fig O, Pressure Suppression Pressure (SPDS 80)
RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
Cannot maintain DW pressure below Fig D, DW Pressure Limit (SPDS 81)
Vent to control DW pressure below Fig D, DW Pressure Limit (SPDS 81)....................C.5-3505 p OK to exceed release rate limits.
p OK to defeat isolations.
p Reducing primary containment pressure affects margin to NPSH limits.
19 EAL FPB EAL FPB 16 Low Level (below -4.0 inches)
TL-6 Restore and maintain torus level above -4.0 inches............................ C.5-3401 p Torus level affects margin to Heat Capacity Limit and ECCS Vortex Limit.
Check Fig M (SPDS 78).
Check Fig V (SPDS 77).
Maintain torus level above -3.3 feet (downcomers).............. C.5-3401 TL-9 IF THEN Cannot maintain torus level above -3.3 feet (downcomers)
RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
Cannot maintain torus level above
-3.7 feet (HPCI exhaust)
AND HPCI is not needed for adequate core cooling
! Operating HPCI with torus level below -3.7 ft may damage the primary containment Trip HPCI.
(Scram, enter C.5-1100)
Torus level drops to -3.3 feet (downcomers)
BEFORE TL-7 Enter C.5-1100 while continuing here.
SCRAM:
+
TL-8 EAL FPB 15 (Scram, enter C.5-1100)
Torus level rises to +3.7 feet (vent header)
BEFORE TL-3 TL-4
- 1. SCRAM:
Enter C.5-1100 while continuing here.
+
- 2. Stop RPV injection from outside primary containment not needed for core cooling or to shut down the reactor.
Raising torus level above 15 ft will result in loss of the hard pipe vent.
Discharging water from the torus may increase radiation levels near the pipe runs.
Maintain torus level below +3.7 feet (vent header).............. C.5-3402 p Do not use pumps needed for RPV injection.
p OK to defeat isolations.
p OK to exceed release rate limits.
TL-5 IF THEN Cannot restore and maintain torus level below +3.7 feet RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
Cannot maintain torus level below +4.2 feet (vacuum breakers)
- 1. Stop DW sprays........ C.5-3502
- 2. Open MO-2373.........C.5-3304 Steps 1,2,3 High Level (above +3.0 inches)
TL-2 Restore and maintain torus level below +3.0 inches.............................C.5-3402 EAL FPB C.5-1200 23 PRIMARY CONTAINMENT CONTROL (Mode 1, 2, 3)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY OPERATING PROCEDURE Rev Number Title 602000019453 PCR C.5-1200 23 PRIMARY CONTAINMENT CONTROL (Mode 1, 2, 3)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY OPERATING PROCEDURE Rev Number Title 602000019453 PCR DW Pressure P (psid)
(PT7251A - PT7251B) 0 DW Pressure (psig)
D 0
5 10 15 10 20 30 40 50 60 70 80 0
5 10 15PT 7251A PT 7251B 62 13.6 Torus Level (ft)
DW Pressure P (psid)
(PT7251A - PT7251B) 0 DW Pressure (psig)
D 0
5 10 15 10 20 30 40 50 60 70 80 0
5 10 15PT 7251A PT 7251B 62 13.6 Torus Level (ft) 15 psig 7 psig 0 psig Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 0
1000 2000 3000 4000 5000 6000 7000 8000 9000 Torus Temperature (°F)
RHR Loop Flow (gpm) 2 pumps 1 pump 1 pump 2 pumps N2 RHR NPSH Limit (SPDS 74) 15 psig 7 psig 0 psig Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 0
1000 2000 3000 4000 5000 6000 7000 8000 9000 Torus Temperature (°F)
RHR Loop Flow (gpm) 2 pumps 1 pump 1 pump 2 pumps N2 RHR NPSH Limit (SPDS 74) 0 Total Flow From Torus Ring Header (gpm x 1000) *
-8 Torus Water Level (ft)
V
- Total Flow = RHR + Core Spray + HPCI + RCIC 5
10 15 20 25 30
-7
-6
-5
-4
-6.7 ft 0
Total Flow From Torus Ring Header (gpm x 1000) *
-8 Torus Water Level (ft)
V
- Total Flow = RHR + Core Spray + HPCI + RCIC 5
10 15 20 25 30
-7
-6
-5
-4
-6.7 ft 0
DW Pressure (psig) 0 DW Temperature (°F)
K 2
4 6
8 10 12 14 60 100 200 300 400 500 600 700 Spray OK Do Not Spray 3
EAL FPB 0
DW Pressure (psig) 0 DW Temperature (°F)
K 2
4 6
8 10 12 14 60 100 200 300 400 500 600 700 Spray OK Do Not Spray 3
EAL FPB
-3.3
-5 Torus Level (ft) 0 Drywell Pressure (psig)
Pressure Suppression Pressure (SPDS 80)
O
-4
-3
-2
-1 0
1 2
3 4
5 5
10 15 20 25 30 35 40 3.7
-3.3
-5 Torus Level (ft) 0 Drywell Pressure (psig)
Pressure Suppression Pressure (SPDS 80)
O
-4
-3
-2
-1 0
1 2
3 4
5 5
10 15 20 25 30 35 40 3.7 A
RPV Water Level Instruments An RPV water level instrument may not be reliable if:
- Temperatures near runs for that instrument are at or above Fig B, RPV Saturation Temperature (SPDS 71).... C.5-3206 OR
- Level is below Fig C, Minimum Usable Levels 100 200 300 400 500 Above 500 0
Highest DW Temperature Near Reference Leg Vertical Run (°F) (C.5-3206)
-325 -279
-48
-36
-36
-24
-4 21 54 78 Feedwater Safeguards (RB Temp <200°F)
Vessel Flood Safeguards (RB Temp >200°F)
Fuel Zone (RB Temp <200°F)
Fuel Zone (RB Temp >200°F)
C Minimum Usable Levels (in.)
1 0
RPV Pressure (psig) 200 Temperature Near Instrument Runs (°F) (C.5-3206)
RPV Saturation Temperature (SPDS 71)
B 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 A
RPV Water Level Instruments An RPV water level instrument may not be reliable if:
- Temperatures near runs for that instrument are at or above Fig B, RPV Saturation Temperature (SPDS 71).... C.5-3206 OR
- Level is below Fig C, Minimum Usable Levels 100 200 300 400 500 Above 500 0
Highest DW Temperature Near Reference Leg Vertical Run (°F) (C.5-3206)
-325 -279
-48
-36
-36
-24
-4 21 54 78 Feedwater Safeguards (RB Temp <200°F)
Vessel Flood Safeguards (RB Temp >200°F)
Fuel Zone (RB Temp <200°F)
Fuel Zone (RB Temp >200°F)
C Minimum Usable Levels (in.)
1 0
RPV Pressure (psig) 200 Temperature Near Instrument Runs (°F) (C.5-3206)
RPV Saturation Temperature (SPDS 71)
B 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 ENTRY CONDITIONS (MODE 1, 2, 3)DW or torus hydrogenabove 2%Torus temperatureabove 90°FTorus levelbelow -inchesORabove +3.0 inchesDW temperatureabove 135°FDW pressureabove 1.84 psig HYDROGEN TORUS TEMPERATURE TT-1
! Exceeding NPSH or Vortex Limit may damage RHR (Figs N2, V; SPDS 74, 77)
Maintain torus temperature below 90°F using torus cooling.
TT-2 IF THEN Cannot maintain torus temperature below 90°F Go to r TT-3 Start all available torus cooling.
p Do not use RHR pumps you need for core cooling.
(Scram, enter C.5-1100)
Torus temperature reaches 110°F BEFORE TT-4 Maintain torus temperature below Fig M, Heat Capacity Limit (SPDS 78).
TT-6 IF THEN Cannot maintain torus temperature below Fig M, Heat Capacity Limit (SPDS 78)
- 1. IF......... you are not in C.5-2001 or C.5-2006, AND.....permitted by C.5-1100 / 2007 Pressure branch, THEN...lower RPV pressure to stay below limit.
p Minimum pressure for adequate HPCI/RCIC flow is 150 psig.
p OK to exceed 100°F/hr cooldown.
- 2. IF......... still cannot stay below Fig M, Heat Capacity Limit (SPDS 78),
THEN...RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
14 TORUS LEVEL Maintain torus level -4.0 to +3.0 inches.
TL-1 IF THEN Low Level:
Cannot maintain level above -4.0 inches Go to y High Level:
Cannot maintain level below +3.0 inches Go to t Maintain DW temperature below 135°F using DW coolers.................................................... B.08.16-05 p DW temperature affects RPV water level indication.
Check Detail A.
DWT-1 IF THEN Cannot maintain DW temperature below 135°F Go to u DWT-2 Start all available DW cooling.
p OK to defeat trips.........C.5-3503 17 DW TEMPERATURE (Scram; enter C.5-1100)
DW temperature reaches 281°F BEFORE DWT-3 BEFORE THEN DWT-5 DW pressure drops below 0 psig Stop DW sprays................................C.5-3502
! Exceeding NPSH or Vortex Limit may damage RHR (Figs N2, V; SPDS 74, 77).
IF......... DW temperature below Fig K, DW Spray Limit (SPDS 79),
AND.....torus level below 4.2 ft, THEN...1. Verify recirc pumps tripped,
- 2. Place DW cooling FAN CONTROL switches in OFF.
- 3. Start DW sprays................................................................C.5-3502 p Do not use pumps you need for core cooling.
p Reducing primary containment pressure affects margin to NPSH limits.
p DW temperature affects RPV water level indication. Check Detail A.
DWT-6 IF THEN Cannot restore and maintain DW temperature below 338°F RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
DW/TORUS PRESSURE Maintain DW and torus pressure below 1.84 psig using SBGT............................................................B.04.01-05 PCP-1 IF THEN Cannot maintain pressure below 1.84 psig Go to i 18 BEFORE THEN PCP-2 DW pressure drops below 0 psig Stop DW sprays...................................C.5-3502 Torus pressure drops below 0 psig Stop torus sprays.................................C.5-3502 Primary containment pressure reduction is required to:
- Restore and maintain adequate core cooling OR
- Reduce the total offsite dose Vent the primary containment..............C.5-3505 p OK to exceed release rate limits.
p OK to defeat isolations.
p Reducing primary containment pressure affects margin to NPSH limits.
Cannot maintain DW or torus pressure above -2.0 psig Restore and maintain DW and torus pressures above -2.0 psig by adding nitrogen (B.04.01-05).
p OK to exceed release rate limits.
p OK to defeat isolations.
C.5-1200 PRIMARY CONTAINMENT CONTROL (MODE 1, 2, 3)
Enter C.5-1100 while continuing here.
SCRAM:
+
DWT-4 Enter C.5-1100 while continuing here.
SCRAM:
+
TT-5 Monitor hydrogen and oxygen in DW and torus...........................C.5-3501 H-1 IF THEN H2/O2 analyzers do not work Have DW and torus sampled for hydrogen and oxygen.......... A.2-415 DW or torus hydrogen reaches 4%
Leave all EOPs
=
Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
EAL FG1 EAL FPB SS5.1 EAL FPB EAL FPB SU4.1 SU4.2 (Torus spray)
DW pressure reaches 12 psig BEFORE PCP-3 PCP-4
! Exceeding NPSH or Vortex Limit may damage RHR (Figs N2, V; SPDS 74, 77)
! Drywell pressure less than 8.6 psig may not provide adequate NPSH to prevent ECCS pump cavitation.
IF......... torus level below 14.1 ft, THEN...start torus sprays................................................................... C.5-3502 p Do not use pumps you need for core cooling.
p Reducing primary containment pressure affects margin to NPSH limits.
PCP-5 IF THEN Cannot maintain DW pressure below 12 psig Go to o IF......... DW temperature below Fig K, DW Spray Limit (SPDS 79),
AND.....torus level below 4.2 ft, THEN...1. Verify recirc pumps tripped,
- 2. Place DW cooling FAN CONTROL switches in OFF.
- 3. Start DW sprays................................................................C.5-3502 p Do not use pumps you need for core cooling.
p Reducing primary containment pressure affects margin to NPSH limits.
PCP-6 IF THEN Cannot maintain DW pressure within Fig O, Pressure Suppression Pressure (SPDS 80)
RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
Cannot maintain DW pressure below Fig D, DW Pressure Limit (SPDS 81)
Vent to control DW pressure below Fig D, DW Pressure Limit (SPDS 81)....................C.5-3505 p OK to exceed release rate limits.
p OK to defeat isolations.
p Reducing primary containment pressure affects margin to NPSH limits.
19 EAL FPB EAL FPB 16 Low Level (below -4.0 inches)
TL-6 Restore and maintain torus level above -4.0 inches............................ C.5-3401 p Torus level affects margin to Heat Capacity Limit and ECCS Vortex Limit.
Check Fig M (SPDS 78).
Check Fig V (SPDS 77).
Maintain torus level above -3.3 feet (downcomers).............. C.5-3401 TL-9 IF THEN Cannot maintain torus level above -3.3 feet (downcomers)
RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
Cannot maintain torus level above
-3.7 feet (HPCI exhaust)
AND HPCI is not needed for adequate core cooling
! Operating HPCI with torus level below -3.7 ft may damage the primary containment Trip HPCI.
(Scram, enter C.5-1100)
Torus level drops to -3.3 feet (downcomers)
BEFORE TL-7 Enter C.5-1100 while continuing here.
SCRAM:
+
TL-8 EAL FPB 15 (Scram, enter C.5-1100)
Torus level rises to +3.7 feet (vent header)
BEFORE TL-3 TL-4
- 1. SCRAM:
Enter C.5-1100 while continuing here.
+
- 2. Stop RPV injection from outside primary containment not needed for core cooling or to shut down the reactor.
Raising torus level above 15 ft will result in loss of the hard pipe vent.
Discharging water from the torus may increase radiation levels near the pipe runs.
Maintain torus level below +3.7 feet (vent header).............. C.5-3402 p Do not use pumps needed for RPV injection.
p OK to defeat isolations.
p OK to exceed release rate limits.
TL-5 IF THEN Cannot restore and maintain torus level below +3.7 feet RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
Cannot maintain torus level below +4.2 feet (vacuum breakers)
- 1. Stop DW sprays........ C.5-3502
- 2. Open MO-2373.........C.5-3304 Steps 1,2,3 High Level (above +3.0 inches)
TL-2 Restore and maintain torus level below +3.0 inches.............................C.5-3402 EAL FPB C.5-1200 23 PRIMARY CONTAINMENT CONTROL (Mode 1, 2, 3)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY OPERATING PROCEDURE Rev Number Title 602000019453 PCR DW Pressure P (psid)
(PT7251A - PT7251B) 0 DW Pressure (psig)
D 0
5 10 15 10 20 30 40 50 60 70 80 0
5 10 15PT 7251A PT 7251B 62 13.6 Torus Level (ft) 15 psig 7 psig 0 psig Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 0
1000 2000 3000 4000 5000 6000 7000 8000 9000 Torus Temperature (°F)
RHR Loop Flow (gpm) 2 pumps 1 pump 1 pump 2 pumps N2 RHR NPSH Limit (SPDS 74) 0 Total Flow From Torus Ring Header (gpm x 1000) *
-8 Torus Water Level (ft)
V
- Total Flow = RHR + Core Spray + HPCI + RCIC 5
10 15 20 25 30
-7
-6
-5
-4
-6.7 ft 0
DW Pressure (psig) 0 DW Temperature (°F)
K 2
4 6
8 10 12 14 60 100 200 300 400 500 600 700 Spray OK Do Not Spray 3
EAL FPB
-3.3
-5 Torus Level (ft) 0 Drywell Pressure (psig)
Pressure Suppression Pressure (SPDS 80)
O
-4
-3
-2
-1 0
1 2
3 4
5 5
10 15 20 25 30 35 40 3.7 A
RPV Water Level Instruments An RPV water level instrument may not be reliable if:
- Temperatures near runs for that instrument are at or above Fig B, RPV Saturation Temperature (SPDS 71).... C.5-3206 OR
- Level is below Fig C, Minimum Usable Levels 100 200 300 400 500 Above 500 0
Highest DW Temperature Near Reference Leg Vertical Run (°F) (C.5-3206)
-325 -279
-48
-36
-36
-24
-4 21 54 78 Feedwater Safeguards (RB Temp <200°F)
Vessel Flood Safeguards (RB Temp >200°F)
Fuel Zone (RB Temp <200°F)
Fuel Zone (RB Temp >200°F)
C Minimum Usable Levels (in.)
1 0
RPV Pressure (psig) 200 Temperature Near Instrument Runs (°F) (C.5-3206)
RPV Saturation Temperature (SPDS 71)
B 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 0
RPV Pressure (psig) 140 Torus Temperature (°F)
Heat Capacity Limit (SPDS 78)
M 100 200 300 400 500 600 700 800 900 1000 1100 150 160 170 180 190 200 210 220 230 240 250 260-ft to +3.7 ft-in. to +3.0 in.
127 psig EAL FPB 0
RPV Pressure (psig) 140 Torus Temperature (°F)
Heat Capacity Limit (SPDS 78)
M 100 200 300 400 500 600 700 800 900 1000 1100 150 160 170 180 190 200 210 220 230 240 250 260-ft to +3.7 ft-in. to +3.0 in.
127 psig EAL FPB-in. to +2.0 ft 0
RPV Pressure (psig) 140 Torus Temperature (°F)
Heat Capacity Limit (SPDS 78)
M 100 200 300 400 500 600 700 800 900 1000 1100 150 160 170 180 190 200 210 220 230 240 250 260-ft to +3.7 ft-in. to +3.0 in.
127 psig EAL FPB-in. to +2.0 ft
ENTRY CONDITIONS (MODE 1, 2, 3)
Any of the following in the reactor building:
TEMPERATURE/RADIATION WATER LEVELS Operate available area coolers and Secondary Containment HVAC.
p Reactor building temperature may affect RPV water level indication. Check Detail A.
SC-2 IF THEN An area temperature or radiation level is above max normal (Tables T, R)
Go to h Restore and maintain floor drain sump levels below max normal (Table W1).
SC-3 IF THEN Cannot restore and maintain floor drain sump level below max normal (Table W1)
OR Any area water level is above 0 in. (Table W2)
Go to h SC-4 Isolate all discharges into area except systems needed for:
- Reactor shutdown
- Core cooling
- Damage control
- Primary containment venting 26 IF THEN SC-5 Primary system is discharging into reactor building AND Discharge cannot be isolated Go to j SC-6 WAIT until 2 or more areas above max safe values of same parameter (Tables T, R, W2)
SC-7 Shut down the reactor.
(Scram, enter C.5-1100)
Any area temperature, radiation, or water level reaches max safe value (Tables T, R, W2)
BEFORE SC-8 Enter C.5-1100 while continuing here.
SCRAM:
+
SC-9 SC-10 WAIT until 2 or more areas above max safe values of same parameter (Tables T, R, W2) 27 ENTRY CONDITIONS (MODE 1, 2, 3)
IF THEN RR-1 Turbine Building Ventilation is shutdown AND Operation of Turbine Building Ventilation is allowed by C.4-B.4.1.B Operate Turbine Building Ventilation.
Isolate all radioactivity releases outside primary and secondary containments except systems needed for:
- Reactor shutdown
- Core Cooling
- Primary containment venting RR-2 IF THEN A primary system is discharging outside primary and secondary containments Before offsite release rate reaches the General Emergency level:
- 1. SCRAM:
Enter C.5-1100 while
+
continuing here.
- 2. RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
W2 Area Water Level Limits Location Max Safe Torus area 32 in.
RCIC 6 in.
RHR A 15 in.
RHR B 53 in.
HPCI 6 in.
Steam Chase 36 in.
Tank Room NA C.5-1300 SECONDARY CONTAINMENT CONTROL (MODE 1, 2, 3)
C.5-1400 RADIOACTIVITY RELEASE CONTROL (MODE 1, 2, 3)
FUEL POOL TEMPERATURE FUEL POOL LEVEL Extended loss of fuel pool cooling may result in fuel pool overflow, reactor building flooding, and loss of habitability.
FPL-1 IF THEN Cannot maintain fuel pool level above 24.75 ft Start lining up fuel pool sprays....................... A.8-02.01 p Do not use systems needed for RPV injection.
Anticipate thermal expansion from prolonged loss of fuel pool cooling Restore and maintain fuel pool level between 24.75 ft and 36.47 ft R
Area Radiation Limits (SPDS 110)
Elev Area Instrument Max Normal Max Safe 1 R/hr 1 R/hr 1 R/hr 1 R/hr 1 R/hr 1 R/hr 10 R/hr 1 R/hr 1 R/hr 1 R/hr 1 R/hr 1 R/hr 10 R/hr 1 R/hr 1 R/hr 1 R/hr 1 R/hr 1 R/hr 110 mr/hr Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint 1000 mr/hr Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Alarm Setpoint Local survey A-16 A-17 A-18 A-14 A-15 NUMACs A-10 A-11 A-12 A-9 A-8 Local survey A-7 A-6 A-5 A-4 A-1, A-2, A-3 Torus Room RCIC Equipment Area East CS & RHR Area West CS & RHR Area HPCI Turbine Area Rx Bldg Drain Tank Steam Chase East CRD HCU West CRD HCU Rx Bldg TIP Drive Rx Bldg East Outside RWCU RWCU Pump Room Chemistry Sample Decon Area Fuel Pool Tank Source Storage Refuel Floor 985' 896' 931' 935' 962' 1001' 1027' W1 Floor Drain Sump Level Limits Location Max Normal Annunciator 42.5 in.
38 in.
6 ft 3 in.
Reactor Building Floor Drain Sump Reactor Building Equipment Drain Sump Reactor Building Floor Drain Tank 84A-A-9 84A-A-4 84A-A-11 While in this procedure:
IF THEN SC-1 Reactor building exhaust plenum radiation above 26 mr/hr
- 1. Verify:
- Secondary Containment HVAC isolation
- SBGT start
- 2. IF.........SBGT cannot restore and maintain Reactor Building DP below 0 in.,
AND.....temperatures, radiation levels, or radioactivity may interfere with operation of systems needed for damage control or EOP actions, THEN...restart Secondary Containment HVAC.
p OK to defeat all Secondary Containment HVAC isolations............C.5-3601 p OK to exceed release rate limits.
Secondary Containment HVAC isolates AND Reactor building exhaust plenum radiation below 26 mr/hr AND SBGT not available Restart Secondary Containment HVAC.
p OK to defeat high DW pressure and low-low RPV water level isolations...........C.5-3601 EAL RA2.2 EAL RA1.1 EAL FPB EALs RU2 RA2 RA3 SU4.3 EAL HU3.2 EAL RU2.1 EAL RA2.3 EAL FPB EAL FPB EAL FPB EAL FPB EAL FPB SC-11 RPV BLOWDOWN REQUIRED.
p See overrides in C.5-1100, C.5-2001, C.5-2006, C.5-2007.
FP1 Alternate Fuel Pool Makeup Sources
- RHR crossties:.................................B.02.01-05
- Torus/CST through RHR
- Fire Water through RHR
- Hose connections on refuel floor:.... B.02.01-05
- Condensate Service Water hose connection
- Fire Water hose connection
- Demin Water hose connection
- Portable Diesel Pumps........................ C.5-4301 FPT-1 Extended loss of fuel pool cooling may result in fuel pool overflow, reactor building flooding, and loss of habitability.
68° to 120°F using FPCC..................................
C.5-1300, C.5-1400 19 SECONDARY CONTAINMENT CONTROL RADIOACTIVITY RELEASE CONTROL (Mode 1, 2, 3)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY OPERATING PROCEDURE Rev Number Title 602000019454 PCR (Maximize cooling)
Fuel pool temperature reaches 140°F BEFORE FPT-2 Operate all available fuel pool cooling.
p Do not use systems needed for RPV injection.
FPT-3 IF THEN Secondary Containment HVAC and SBGT unavailable Before fuel pool temperature reaches 212°F, establish airflow through the secondary containment.............. C.5-4501 p OK to defeat isolations and interlocks.
p OK to exceed release rate limits.
(Spray)
Fuel pool level drops to 15.25 ft (instrument downscale)
BEFORE FPL-2 EAL RS2.1 EAL RG2.1 Operate fuel pool sprays.............................................A.8-02.01 p Do not use sources needed for RPV injection.
FPL-3 IF THEN Fuel pool level drops below 15.25 ft (instrument downscale)
Cool spent fuel bundles using either or both of the following methods:
p Even if RPV injection will be reduced.
- Fuel pool sprays......A.8-02.01
- Restore and maintain fuel pool level above 15.25 ft (instrument downscale).
p Use any available Alternate Fuel Pool Makeup Sources (Detail FP1).
T Area Temperature Limits Elev Area Instrument Max Normal Max Safe 985' 896' 931' 935' 962' 1001' 1027' Torus Room RCIC Equipment Area East CS & RHR Area West CS & RHR Area HPCI Turbine Area Rx Bldg Drain Tank Steam Chase East CRD HCU West CRD HCU Rx Bldg TIP Drive Rx Bldg East Outside RWCU RWCU Pump Room Chemistry Sample Decon Area Fuel Pool Tank Source Storage Refuel Floor TR-4926 Pt. 8,11,21,25 TR-4926 Pt. 22,23,24 TR-4926 Pt. 5,6,7 TR-4926 Pt. 9,10,12 TR-4926 Pt. 26,27,28 Local survey TR-4926 Pt. 1,2,3,4 Local survey Local survey Local survey Local survey Local survey TR-4926 Pt. 13 Local survey Local survey Local survey Local survey Local survey 130°F 130°F 130°F 130°F 130°F 130°F 165°F 120°F 120°F 120°F 120°F 120°F 145°F 120°F 120°F 120°F 120°F 120°F 212°F 212°F 212°F 212°F 212°F 212°F 212°F 130°F 130°F 212°F 130°F 212°F 212°F 158°F 128°F 128°F 128°F 212°F 140°F 212°F SFP Levels Minimum Safe Operating Level 24.75' Top of Fuel Racks 14.75' Low Level Alarm 36.58' Normal Level 37.75' Operating Floor Level 38.75' Top of Curb 39.01' Instrument Downscale 15.25 A
RPV Water Level Instruments An RPV water level instrument may not be reliable if:
- Temperatures near runs for that instrument are at or above Fig B, RPV Saturation Temperature (SPDS 71).... C.5-3206 OR
- Level is below Fig C, Minimum Usable Levels 100 200 300 400 500 Above 500 0
Highest DW Temperature Near Reference Leg Vertical Run (°F) (C.5-3206)
-325 -279
-48
-36
-36
-24
-4 21 54 78 Feedwater Safeguards (RB Temp <200°F)
Vessel Flood Safeguards (RB Temp >200°F)
Fuel Zone (RB Temp <200°F)
Fuel Zone (RB Temp >200°F)
C Minimum Usable Levels (in.)
1 0
RPV Pressure (psig) 200 Temperature Near Instrument Runs (°F) (C.5-3206)
RPV Saturation Temperature (SPDS 71)
B 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600
ENTRY CONDITIONS (MODE 1, 2, 3)RPV water levelbelow 9 in.RPV pressureabove 1056 psigDW pressureabove 1.84 psigReactor powerabove 4.0% or unknownwhen scram required 1
No/Unknown Yes 2
All rods in to at least 02
?
No/Unknown Yes 3
Will reactor stay shutdown under all conditions without boron
?
FAILURE TO SCRAM:
Leave this procedure Enter C.5-2007
=
4 IF THEN L-1 Cannot stay within Fig D, DW Pressure Limit (SPDS 81)
Stop RPV injection from outside the primary containment not needed for core cooling.
Raw water injected OR High level is needed for use of RCIC, HPCI, Alternate Injection Systems (Detail L2),
or Shutdown Cooling Restore and maintain RPV water level -126 in. (TAF) to 100 in.
p OK to defeat high RPV water level trips.
Restore and maintain RPV water level 9 to 48 in. using Preferred Injection Systems (Detail L1).
p Use Alternate Injection Systems if needed (Detail L2).
p Check Detail A for instrument limits.
L-2 IF THEN Cannot restore and maintain level 9 to 48 in.
Go to 2 PRESSURE C.5-1100 RPV CONTROL (MODE 1, 2, 3)
EAL FPB P-2 Stabilize RPV pressure below 1056 psig using main turbine bypass valves.............................................................................................. C.5-3302 p Use Alternate Pressure Control Systems (Detail P1) if needed.......... C.5-3302 p If SRV pneumatics lost:
Place SRV D switch in AUTO and SRV G switch in CLOSE.
Place DIV II SRV LO-LO SET LOGIC switch in BYPASS.
P-4 WAIT until:
Shutdown Cooling interlock clears (75 psig)
AND further cooldown required Cool down to cold shutdown using Shutdown Cooling......................B.03.04-05 p Do not use RHR pumps you need for maintaining RPV water level above 9 in.
p Maintain cooldown rate below 100°F/hr.
P-5 IF THEN Cannot use Shutdown Cooling Maintain RPV pressure below 75 psig (shutdown cooling interlock) using Alternate Pressure Control Systems (Detail P1)............................C.5-3302 EALs SU5.1 SA5.1 EALs SU2.1 SA2.1 FPB Verify needed auto actions while continuing here:
- Isolations
- ECCS start
- Diesel generator start
+
5 While in this procedure:
IF THEN 6
Anticipate that available Injection Subsystems (Detail L3) will not provide adequate core cooling Leave this procedure
=
Enter C.5-2001 USE ALTERNATE LEVEL/PRESSURE CONTROL:
RPV water level unknown Leave this procedure
=
Enter C.5-2006 at n FLOOD THE RPV:
Ongoing and progressive core damage is occurring..... A.2-208 Leave all EOPs =
Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
P-3 Depressurize the RPV using main turbine bypass valves................... C.5-3302 p Maintain cooldown rate below 100°F/hr.
p Use Alternate Pressure Control Systems (Detail P1) if needed......... C.5-3302 p If SRV pneumatics lost, minimize number of SRV D and SRV G cycles.
IF THEN P-1 DW pressure above 1.84 psig Before RPV pressure drops to 320 psig, prevent Core Spray and LPCI injection not needed for core cooling................. C.5-3205 Anticipate RPV Blowdown requirement Depressurize RPV rapidly using main turbine bypass valves.
p OK to exceed 100°F/hr cooldown.
RPV Blowdown Required by any EOP Leave Pressure
=
Go to 5 BLOW DOWN:
Cannot maintain torus temperature below Fig M, Heat Capacity Limit (SPDS 78)
Lower RPV pressure to stay below limit.
p OK to exceed 100°F/hr cooldown.
EALs FPB SS5.1 LEVEL IF THEN L-3 RPV water level can be restored and maintained above 9 in.
Return to 1 p OK to execute the RPV Level Recovery Strategy............................... C.5-3208 Cannot maintain level above -47 in. (ADS initiation)
Inhibit ADS.
1 RPV Blowdown 5
No Yes P-8 Torus level above SRV discharges
(-5.9 ft)
?
P-9 Open all 3 ADS valves.
p Leave switches in OPEN.
p OK to exceed 100°F/hr cooldown.
No Yes P-10 All 3 ADS valves verified open
?
P-11 Open other SRVs until a total of 3 are verified open.
p Leave switches in OPEN.
p OK to exceed 100°F/hr cooldown.
No Yes P-6 DW pressure at or above 1.84 psig
?
P-7 Prevent Core Spray and LPCI injection not needed for core cooling........C.5-3205 EAL FPB EAL FPB P-13 WAIT until:
Shutdown Cooling interlock clears (75 psig)
AND further cooldown required Allow the RPV to depressurize.
p OK to exceed 100°F/hr cooldown.
p Reducing RPV pressure affects margin to RPV Saturation Temperature. Check Detail A.
P-12 IF THEN Less than 3 SRVs can be opened AND RPV-to-DW P at or above 50 psid Rapidly lower RPV-to-DW P less than 50 psid using Alternate Depressurization Systems (Detail P2).
p OK to exceed 100°F/hr cooldown.
p OK to exceed release rate limits.
p OK to defeat trips and isolations.
Cool down to cold shutdown using Shutdown Cooling.
p Do not use RHR pumps you need for maintaining RPV water level above 9 in.
P-14 IF THEN Cannot use Shutdown Cooling Cool down to cold shutdown using Alternate Depressurization Systems (Detail P2).
p OK to defeat trips and isolations.
2 EAL FPB Restore and maintain RPV water level above -126 in. (TAF) using Preferred Injection Systems (Detail L1).
p Use Alternate Injection Systems if needed (Detail L2).
L-4 IF THEN Cannot restore and maintain level above -126 in. (TAF)
Go to 3 Less than 2 Injection Subsystems (Detail L3) available Start lining up all Alternate Injection Systems (Detail L2) while continuing here.
+
(Blow down)
RPV water level drops to
-150 in.
BEFORE L-5 See override in Pressure.
RPV BLOWDOWN REQUIRED:
+
L-6 3
Yes No L-7 1 Core Spray and at least one other ECCS pump available
?
Inject as needed using Preferred and Alternate Injection Systems (Details L1 and L2).
p Check Detail A for RPV water level instrument limits.
L-8 IF THEN Less than 2 ECCS pumps available Go to 4 No Core Spray pump injecting at least 3150 gpm Cannot restore and maintain level above -174 in.
Leave all EOPs = Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
Maximize injection using Preferred and Alternate Injection Systems (Details L1 and L2).
p Check Detail A for RPV water level instrument limits.
L-9 IF THEN Cannot restore and maintain level above -150 in.
Leave all EOPs = Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
4 EAL FG1 EAL FG1 To 5 P2 Alternate Depressurization Systems
! Note Main steam line drains 150,000 lb/hr HPCI Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N4, V; SPDS 76, 77).
p OK to defeat:........................................................C.5-3202 High RPV water level trip High temperature trip 100,000 lb/hr RCIC Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N3, V; SPDS 75, 77).
p OK to defeat:........................,,,..............................C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 15,000 lb/hr Deareating Steam 95,000 lb/hr Gland Steam 14,000 lb/hr SJAEs 6,200 lb/hr Recombiners 300 lb/hr RWCU, normal mode p Bypass regen HXs and filter/demins as soon as you can.
p OK to defeat isolations, including SBLC.
xxxx RWCU, dump mode p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
xxxx SRVs p Only if torus level is above -5.9 ft.
800,000 lb/hr Main turbine bypass valves 972,000 lb/hr Shutdown Cooling p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
p Do not use RHR pumps you need for LPCI injection.
xxxx Head vent 700 lb/hr System / Details (C.5-3303)
Capacity P2 Alternate Depressurization Systems
! Note Main steam line drains 150,000 lb/hr HPCI Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N4, V; SPDS 76, 77).
p OK to defeat:........................................................C.5-3202 High RPV water level trip High temperature trip 100,000 lb/hr RCIC Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N3, V; SPDS 75, 77).
p OK to defeat:........................,,,..............................C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 15,000 lb/hr Deareating Steam 95,000 lb/hr Gland Steam 14,000 lb/hr SJAEs 6,200 lb/hr Recombiners 300 lb/hr RWCU, normal mode p Bypass regen HXs and filter/demins as soon as you can.
p OK to defeat isolations, including SBLC.
xxxx RWCU, dump mode p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
xxxx SRVs p Only if torus level is above -5.9 ft.
800,000 lb/hr Main turbine bypass valves 972,000 lb/hr Shutdown Cooling p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
p Do not use RHR pumps you need for LPCI injection.
xxxx Head vent 700 lb/hr System / Details (C.5-3303)
Capacity 50 100 150 200 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 Torus Temperature (°F)
Core Spray Pump Flow (gpm)
Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 15 psig 7 psig 0 psig N1 Core Spray NPSH Limit (SPDS 73) 50 100 150 200 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 Torus Temperature (°F)
Core Spray Pump Flow (gpm)
Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 15 psig 7 psig 0 psig N1 Core Spray NPSH Limit (SPDS 73) 15 psig 7 psig 0 psig Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 0
1000 2000 3000 4000 5000 6000 7000 8000 9000 Torus Temperature (°F)
RHR Loop Flow (gpm) 2 pumps 1 pump 1 pump 2 pumps N2 RHR NPSH Limit (SPDS 74) 15 psig 7 psig 0 psig Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 0
1000 2000 3000 4000 5000 6000 7000 8000 9000 Torus Temperature (°F)
RHR Loop Flow (gpm) 2 pumps 1 pump 1 pump 2 pumps N2 RHR NPSH Limit (SPDS 74)
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5)
Overpressure
- 15 psig 7 psig 0 psig 50 100 150 200 250 300 0
100 200 300 400 500 Torus Temperature (°F)
RCIC Pump Flow From Torus (gpm)
Limit applies only when suction is from torus.
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5)
Overpressure
- 15 psig 7 psig 0 psig 50 100 150 200 250 300 0
100 200 300 400 500 Torus Temperature (°F)
RCIC Pump Flow From Torus (gpm)
Limit applies only when suction is from torus.
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5)
Overpressure
- 15 psig 7 psig 0 psig 100 150 200 250 300 0
1000 2000 3000 4000 Torus Temperature (°F)
HPCI Pump Flow From Torus (gpm)
Limit applies only when suction is from torus.
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5)
Overpressure
- 15 psig 7 psig 0 psig 100 150 200 250 300 0
1000 2000 3000 4000 Torus Temperature (°F)
HPCI Pump Flow From Torus (gpm)
Limit applies only when suction is from torus.
N4 HPCI NPSH Limit (SPDS 76) 0 Total Flow From Torus Ring Header (gpm x 1000) *
-8 Torus Water Level (ft)
V
- Total Flow = RHR + Core Spray + HPCI + RCIC 5
10 15 20 25 30
-7
-6
-5
-4
-6.7 ft 0
Total Flow From Torus Ring Header (gpm x 1000) *
-8 Torus Water Level (ft)
V
- Total Flow = RHR + Core Spray + HPCI + RCIC 5
10 15 20 25 30
-7
-6
-5
-4
-6.7 ft DW Pressure P (psid)
(PT7251A - PT7251B) 0 DW Pressure (psig)
D 0
5 10 15 10 20 30 40 50 60 70 80 0
5 10 15PT 7251A PT 7251B 62 13.6 Torus Level (ft)
DW Pressure P (psid)
(PT7251A - PT7251B) 0 DW Pressure (psig)
D 0
5 10 15 10 20 30 40 50 60 70 80 0
5 10 15PT 7251A PT 7251B 62 13.6 Torus Level (ft)
A RPV Water Level Instruments An RPV water level instrument may not be reliable if:
- Temperatures near runs for that instrument are at or above Fig B, RPV Saturation Temperature (SPDS 71).... C.5-3206 OR
- Level is below Fig C, Minimum Usable Levels 100 200 300 400 500 Above 500 0
Highest DW Temperature Near Reference Leg Vertical Run (°F) (C.5-3206)
-325 -279
-48
-36
-36
-24
-4 21 54 78 Feedwater Safeguards (RB Temp <200°F)
Vessel Flood Safeguards (RB Temp >200°F)
Fuel Zone (RB Temp <200°F)
Fuel Zone (RB Temp >200°F)
C Minimum Usable Levels (in.)
1 0
RPV Pressure (psig) 200 Temperature Near Instrument Runs (°F) (C.5-3206)
RPV Saturation Temperature (SPDS 71)
B 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 A
RPV Water Level Instruments An RPV water level instrument may not be reliable if:
- Temperatures near runs for that instrument are at or above Fig B, RPV Saturation Temperature (SPDS 71).... C.5-3206 OR
- Level is below Fig C, Minimum Usable Levels 100 200 300 400 500 Above 500 0
Highest DW Temperature Near Reference Leg Vertical Run (°F) (C.5-3206)
-325 -279
-48
-36
-36
-24
-4 21 54 78 Feedwater Safeguards (RB Temp <200°F)
Vessel Flood Safeguards (RB Temp >200°F)
Fuel Zone (RB Temp <200°F)
Fuel Zone (RB Temp >200°F)
C Minimum Usable Levels (in.)
1 0
RPV Pressure (psig) 200 Temperature Near Instrument Runs (°F) (C.5-3206)
RPV Saturation Temperature (SPDS 71)
B 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 L1 Preferred Injection Systems
! Exceeding NPSH or Vortex Limits may damage systems (Figs N1-N4, Fig V; SPDS 73-77).
- Condensate/Feedwater p OK to defeat high RPV water level trip.
9000 gpm
- HPCI Exceeding 180°F suction temperature may damage system.
p Use CST suction if available.
p OK to defeat:...................................................... C.5-3202 High RPV water level trip High temperature trip High torus level suction transfer 3000 gpm
! Exceeding 180°F suction temperature may damage system.
p Use CST suction if available.
p OK to defeat:...................................................... C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 400 gpm
- CRD.......................................................................... C.5-3204 74 gpm
- Core Spray 3150 gpm
4000 gpm System / Details Capacity L1 Preferred Injection Systems
! Exceeding NPSH or Vortex Limits may damage systems (Figs N1-N4, Fig V; SPDS 73-77).
- Condensate/Feedwater p OK to defeat high RPV water level trip.
9000 gpm
- HPCI Exceeding 180°F suction temperature may damage system.
p Use CST suction if available.
p OK to defeat:...................................................... C.5-3202 High RPV water level trip High temperature trip High torus level suction transfer 3000 gpm
! Exceeding 180°F suction temperature may damage system.
p Use CST suction if available.
p OK to defeat:...................................................... C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 400 gpm
- CRD.......................................................................... C.5-3204 74 gpm
- Core Spray 3150 gpm
4000 gpm System / Details Capacity L2 Alternate Injection Systems Note
- SBLC.......................................................................................C.5-3203 28 gpm
- Condensate Service to pressurization station.........................C.5-3203 800 gpm
- Fire Water crosstie to LPCI..................................................... C.5-3203 1500 gpm
- Service Water Crosstie to Condensate/Feedwater................. C.5-3203 8000 gpm
- RCIC manual operation.......................................................... A.8-05.01 400 gpm
- Portable Diesel Pump Injection to RPV................................... C.5-3203 350 gpm System Capacity L2 Alternate Injection Systems Note
- SBLC.......................................................................................C.5-3203 28 gpm
- Condensate Service to pressurization station.........................C.5-3203 800 gpm
- Fire Water crosstie to LPCI..................................................... C.5-3203 1500 gpm
- Service Water Crosstie to Condensate/Feedwater................. C.5-3203 8000 gpm
- RCIC manual operation.......................................................... A.8-05.01 400 gpm
- Portable Diesel Pump Injection to RPV................................... C.5-3203 350 gpm System Capacity L3 Injection Subsystems Note Condensate 9000 gpm Core Spray A 3150 gpm Core Spray B 3150 gpm LPCI-A 4000 gpm LPCI-B 4000 gpm Subsystem Capacity L3 Injection Subsystems Note Condensate 9000 gpm Core Spray A 3150 gpm Core Spray B 3150 gpm LPCI-A 4000 gpm LPCI-B 4000 gpm Subsystem Capacity P1 Alternate Pressure Control Systems
! Note Main steam line drains 150,000 lb/hr HPCI Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N3, V; SPDS 76, 77).
p Use CST suction if available.
p OK to defeat:........................................................C.5-3202 High RPV water level trip High temperature trip High torus level suction transfer 100,000 lb/hr RCIC Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N4, V; SPDS 75, 77).
p Use CST suction if available.
p OK to defeat:........................,,,..............................C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 15,000 lb/hr SRVs p Only if torus level is above -5.9 ft.
p Use LO-LO SET or preferred sequence if you can.
800,000 lb/hr Deareating Steam 95,000 lb/hr Gland Steam 14,000 lb/hr SJAEs 6,200 lb/hr Recombiners 300 lb/hr RWCU, normal mode p Bypass regen HXs and filter/demins as soon as you can.
p OK to defeat isolations, including SBLC.
xxxx RWCU, dump mode p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
xxxx System / Details (C.5-3302)
Capacity P1 Alternate Pressure Control Systems
! Note Main steam line drains 150,000 lb/hr HPCI Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N3, V; SPDS 76, 77).
p Use CST suction if available.
p OK to defeat:........................................................C.5-3202 High RPV water level trip High temperature trip High torus level suction transfer 100,000 lb/hr RCIC Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N4, V; SPDS 75, 77).
p Use CST suction if available.
p OK to defeat:........................,,,..............................C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 15,000 lb/hr SRVs p Only if torus level is above -5.9 ft.
p Use LO-LO SET or preferred sequence if you can.
800,000 lb/hr Deareating Steam 95,000 lb/hr Gland Steam 14,000 lb/hr SJAEs 6,200 lb/hr Recombiners 300 lb/hr RWCU, normal mode p Bypass regen HXs and filter/demins as soon as you can.
p OK to defeat isolations, including SBLC.
xxxx RWCU, dump mode p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
xxxx System / Details (C.5-3302)
Capacity C.5-1100 23 RPV CONTROL (Mode 1, 2, 3)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY OPERATING PROCEDURE Rev Number Title 602000019451 PCR C.5-1100 23 RPV CONTROL (Mode 1, 2, 3)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY OPERATING PROCEDURE Rev Number Title 602000019451 PCR ENTRY CONDITIONS (MODE 1, 2, 3)RPV water levelbelow 9 in.RPV pressureabove 1056 psigDW pressureabove 1.84 psigReactor powerabove 4.0% or unknownwhen scram required 1
No/Unknown Yes 2
All rods in to at least 02
?
No/Unknown Yes 3
Will reactor stay shutdown under all conditions without boron
?
FAILURE TO SCRAM:
Leave this procedure Enter C.5-2007
=
4 IF THEN L-1 Cannot stay within Fig D, DW Pressure Limit (SPDS 81)
Stop RPV injection from outside the primary containment not needed for core cooling.
Raw water injected OR High level is needed for use of RCIC, HPCI, Alternate Injection Systems (Detail L2),
or Shutdown Cooling Restore and maintain RPV water level -126 in. (TAF) to 100 in.
p OK to defeat high RPV water level trips.
Restore and maintain RPV water level 9 to 48 in. using Preferred Injection Systems (Detail L1).
p Use Alternate Injection Systems if needed (Detail L2).
p Check Detail A for instrument limits.
L-2 IF THEN Cannot restore and maintain level 9 to 48 in.
Go to 2 PRESSURE C.5-1100 RPV CONTROL (MODE 1, 2, 3)
EAL FPB P-2 Stabilize RPV pressure below 1056 psig using main turbine bypass valves.............................................................................................. C.5-3302 p Use Alternate Pressure Control Systems (Detail P1) if needed.......... C.5-3302 p If SRV pneumatics lost:
Place SRV D switch in AUTO and SRV G switch in CLOSE.
Place DIV II SRV LO-LO SET LOGIC switch in BYPASS.
P-4 WAIT until:
Shutdown Cooling interlock clears (75 psig)
AND further cooldown required Cool down to cold shutdown using Shutdown Cooling......................B.03.04-05 p Do not use RHR pumps you need for maintaining RPV water level above 9 in.
p Maintain cooldown rate below 100°F/hr.
P-5 IF THEN Cannot use Shutdown Cooling Maintain RPV pressure below 75 psig (shutdown cooling interlock) using Alternate Pressure Control Systems (Detail P1)............................C.5-3302 EALs SU5.1 SA5.1 EALs SU2.1 SA2.1 FPB Verify needed auto actions while continuing here:
- Isolations
- ECCS start
- Diesel generator start
+
5 While in this procedure:
IF THEN 6
Anticipate that available Injection Subsystems (Detail L3) will not provide adequate core cooling Leave this procedure
=
Enter C.5-2001 USE ALTERNATE LEVEL/PRESSURE CONTROL:
RPV water level unknown Leave this procedure
=
Enter C.5-2006 at n FLOOD THE RPV:
Ongoing and progressive core damage is occurring..... A.2-208 Leave all EOPs =
Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
P-3 Depressurize the RPV using main turbine bypass valves................... C.5-3302 p Maintain cooldown rate below 100°F/hr.
p Use Alternate Pressure Control Systems (Detail P1) if needed......... C.5-3302 p If SRV pneumatics lost, minimize number of SRV D and SRV G cycles.
IF THEN P-1 DW pressure above 1.84 psig Before RPV pressure drops to 320 psig, prevent Core Spray and LPCI injection not needed for core cooling................. C.5-3205 Anticipate RPV Blowdown requirement Depressurize RPV rapidly using main turbine bypass valves.
p OK to exceed 100°F/hr cooldown.
RPV Blowdown Required by any EOP Leave Pressure
=
Go to 5 BLOW DOWN:
Cannot maintain torus temperature below Fig M, Heat Capacity Limit (SPDS 78)
Lower RPV pressure to stay below limit.
p OK to exceed 100°F/hr cooldown.
EALs FPB SS5.1 LEVEL IF THEN L-3 RPV water level can be restored and maintained above 9 in.
Return to 1 p OK to execute the RPV Level Recovery Strategy............................... C.5-3208 Cannot maintain level above -47 in. (ADS initiation)
Inhibit ADS.
1 RPV Blowdown 5
No Yes P-8 Torus level above SRV discharges
(-5.9 ft)
?
P-9 Open all 3 ADS valves.
p Leave switches in OPEN.
p OK to exceed 100°F/hr cooldown.
No Yes P-10 All 3 ADS valves verified open
?
P-11 Open other SRVs until a total of 3 are verified open.
p Leave switches in OPEN.
p OK to exceed 100°F/hr cooldown.
No Yes P-6 DW pressure at or above 1.84 psig
?
P-7 Prevent Core Spray and LPCI injection not needed for core cooling........C.5-3205 EAL FPB EAL FPB P-13 WAIT until:
Shutdown Cooling interlock clears (75 psig)
AND further cooldown required Allow the RPV to depressurize.
p OK to exceed 100°F/hr cooldown.
p Reducing RPV pressure affects margin to RPV Saturation Temperature. Check Detail A.
P-12 IF THEN Less than 3 SRVs can be opened AND RPV-to-DW P at or above 50 psid Rapidly lower RPV-to-DW P less than 50 psid using Alternate Depressurization Systems (Detail P2).
p OK to exceed 100°F/hr cooldown.
p OK to exceed release rate limits.
p OK to defeat trips and isolations.
Cool down to cold shutdown using Shutdown Cooling.
p Do not use RHR pumps you need for maintaining RPV water level above 9 in.
P-14 IF THEN Cannot use Shutdown Cooling Cool down to cold shutdown using Alternate Depressurization Systems (Detail P2).
p OK to defeat trips and isolations.
2 EAL FPB Restore and maintain RPV water level above -126 in. (TAF) using Preferred Injection Systems (Detail L1).
p Use Alternate Injection Systems if needed (Detail L2).
L-4 IF THEN Cannot restore and maintain level above -126 in. (TAF)
Go to 3 Less than 2 Injection Subsystems (Detail L3) available Start lining up all Alternate Injection Systems (Detail L2) while continuing here.
+
(Blow down)
RPV water level drops to
-150 in.
BEFORE L-5 See override in Pressure.
RPV BLOWDOWN REQUIRED:
+
L-6 3
Yes No L-7 1 Core Spray and at least one other ECCS pump available
?
Inject as needed using Preferred and Alternate Injection Systems (Details L1 and L2).
p Check Detail A for RPV water level instrument limits.
L-8 IF THEN Less than 2 ECCS pumps available Go to 4 No Core Spray pump injecting at least 3150 gpm Cannot restore and maintain level above -174 in.
Leave all EOPs = Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
Maximize injection using Preferred and Alternate Injection Systems (Details L1 and L2).
p Check Detail A for RPV water level instrument limits.
L-9 IF THEN Cannot restore and maintain level above -150 in.
Leave all EOPs = Enter A.7-SAMG-01 and A.7-SAMG-02 ENTER SAMGs:
4 EAL FG1 EAL FG1 To 5 P2 Alternate Depressurization Systems
! Note Main steam line drains 150,000 lb/hr HPCI Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N4, V; SPDS 76, 77).
p OK to defeat:........................................................C.5-3202 High RPV water level trip High temperature trip 100,000 lb/hr RCIC Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N3, V; SPDS 75, 77).
p OK to defeat:........................,,,..............................C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 15,000 lb/hr Deareating Steam 95,000 lb/hr Gland Steam 14,000 lb/hr SJAEs 6,200 lb/hr Recombiners 300 lb/hr RWCU, normal mode p Bypass regen HXs and filter/demins as soon as you can.
p OK to defeat isolations, including SBLC.
xxxx RWCU, dump mode p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
xxxx SRVs p Only if torus level is above -5.9 ft.
800,000 lb/hr Main turbine bypass valves 972,000 lb/hr Shutdown Cooling p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
p Do not use RHR pumps you need for LPCI injection.
xxxx Head vent 700 lb/hr System / Details (C.5-3303)
Capacity 50 100 150 200 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 Torus Temperature (°F)
Core Spray Pump Flow (gpm)
Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 15 psig 7 psig 0 psig N1 Core Spray NPSH Limit (SPDS 73) 15 psig 7 psig 0 psig Overpressure *
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5) 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 0
500 1000 1500 2000 2500 3000 3500 4000 4500 0
1000 2000 3000 4000 5000 6000 7000 8000 9000 Torus Temperature (°F)
RHR Loop Flow (gpm) 2 pumps 1 pump 1 pump 2 pumps N2 RHR NPSH Limit (SPDS 74)
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5)
Overpressure
- 15 psig 7 psig 0 psig 50 100 150 200 250 300 0
100 200 300 400 500 Torus Temperature (°F)
RCIC Pump Flow From Torus (gpm)
Limit applies only when suction is from torus.
- Overpressure (psig) = Drywell Pressure + 0.4 x (Torus Level (ft) + 4.5)
Overpressure
- 15 psig 7 psig 0 psig 100 150 200 250 300 0
1000 2000 3000 4000 Torus Temperature (°F)
HPCI Pump Flow From Torus (gpm)
Limit applies only when suction is from torus.
N4 HPCI NPSH Limit (SPDS 76) 0 Total Flow From Torus Ring Header (gpm x 1000) *
-8 Torus Water Level (ft)
V
- Total Flow = RHR + Core Spray + HPCI + RCIC 5
10 15 20 25 30
-7
-6
-5
-4
-6.7 ft DW Pressure P (psid)
(PT7251A - PT7251B) 0 DW Pressure (psig)
D 0
5 10 15 10 20 30 40 50 60 70 80 0
5 10 15PT 7251A PT 7251B 62 13.6 Torus Level (ft)
A RPV Water Level Instruments An RPV water level instrument may not be reliable if:
- Temperatures near runs for that instrument are at or above Fig B, RPV Saturation Temperature (SPDS 71).... C.5-3206 OR
- Level is below Fig C, Minimum Usable Levels 100 200 300 400 500 Above 500 0
Highest DW Temperature Near Reference Leg Vertical Run (°F) (C.5-3206)
-325 -279
-48
-36
-36
-24
-4 21 54 78 Feedwater Safeguards (RB Temp <200°F)
Vessel Flood Safeguards (RB Temp >200°F)
Fuel Zone (RB Temp <200°F)
Fuel Zone (RB Temp >200°F)
C Minimum Usable Levels (in.)
1 0
RPV Pressure (psig) 200 Temperature Near Instrument Runs (°F) (C.5-3206)
RPV Saturation Temperature (SPDS 71)
B 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 L1 Preferred Injection Systems
! Exceeding NPSH or Vortex Limits may damage systems (Figs N1-N4, Fig V; SPDS 73-77).
- Condensate/Feedwater p OK to defeat high RPV water level trip.
9000 gpm
- HPCI Exceeding 180°F suction temperature may damage system.
p Use CST suction if available.
p OK to defeat:...................................................... C.5-3202 High RPV water level trip High temperature trip High torus level suction transfer 3000 gpm
! Exceeding 180°F suction temperature may damage system.
p Use CST suction if available.
p OK to defeat:...................................................... C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 400 gpm
- CRD.......................................................................... C.5-3204 74 gpm
- Core Spray 3150 gpm
4000 gpm System / Details Capacity L2 Alternate Injection Systems Note
- SBLC.......................................................................................C.5-3203 28 gpm
- Condensate Service to pressurization station.........................C.5-3203 800 gpm
- Fire Water crosstie to LPCI..................................................... C.5-3203 1500 gpm
- Service Water Crosstie to Condensate/Feedwater................. C.5-3203 8000 gpm
- RCIC manual operation.......................................................... A.8-05.01 400 gpm
- Portable Diesel Pump Injection to RPV................................... C.5-3203 350 gpm System Capacity L3 Injection Subsystems Note Condensate 9000 gpm Core Spray A 3150 gpm Core Spray B 3150 gpm LPCI-A 4000 gpm LPCI-B 4000 gpm Subsystem Capacity P1 Alternate Pressure Control Systems
! Note Main steam line drains 150,000 lb/hr HPCI Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N3, V; SPDS 76, 77).
p Use CST suction if available.
p OK to defeat:........................................................C.5-3202 High RPV water level trip High temperature trip High torus level suction transfer 100,000 lb/hr RCIC Exceeding NPSH Limit, Vortex Limit, or 180°F suction temperature may damage system (Figs N4, V; SPDS 75, 77).
p Use CST suction if available.
p OK to defeat:........................,,,..............................C.5-3201 Low RPV pressure trip High RPV water level trip High exhaust pressure trip High temperature trip 15,000 lb/hr SRVs p Only if torus level is above -5.9 ft.
p Use LO-LO SET or preferred sequence if you can.
800,000 lb/hr Deareating Steam 95,000 lb/hr Gland Steam 14,000 lb/hr SJAEs 6,200 lb/hr Recombiners 300 lb/hr RWCU, normal mode p Bypass regen HXs and filter/demins as soon as you can.
p OK to defeat isolations, including SBLC.
xxxx RWCU, dump mode p Only if NO boron injected or reactor will remain shutdown under all conditions without boron.
xxxx System / Details (C.5-3302)
Capacity C.5-1100 23 RPV CONTROL (Mode 1, 2, 3)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY OPERATING PROCEDURE Rev Number Title 602000019451 PCR EAL FG1 0
RPV Pressure (psig) 140 Torus Temperature (°F)
Heat Capacity Limit (SPDS 78)
M 100 200 300 400 500 600 700 800 900 1000 1100 150 160 170 180 190 200 210 220 230 240 250 260-ft to +3.7 ft-in. to +3.0 in.
127 psig EAL FPB 0
RPV Pressure (psig) 140 Torus Temperature (°F)
Heat Capacity Limit (SPDS 78)
M 100 200 300 400 500 600 700 800 900 1000 1100 150 160 170 180 190 200 210 220 230 240 250 260-ft to +3.7 ft-in. to +3.0 in.
127 psig EAL FPB-in. to +2.0 ft 0
RPV Pressure (psig) 140 Torus Temperature (°F)
Heat Capacity Limit (SPDS 78)
M 100 200 300 400 500 600 700 800 900 1000 1100 150 160 170 180 190 200 210 220 230 240 250 260-ft to +3.7 ft-in. to +3.0 in.
127 psig EAL FPB-in. to +2.0 ft I
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