This section captures the fundamental reactor identification and baseline operational parameters immediately preceding the thermal anomaly detection. Accurate baseline data is critical for root cause analysis and delta calculations.
Reactor Unit Identification Code
SMR Model Designation
Facility Location & Grid Node
Baseline Telemetry Timestamp (Pre-Incident)
Shift Supervisor on Duty
Licensed Operators on Duty at Detection
Senior Reactor Operator (SRO)
Reactor Operator (RO)
Balance of Plant Operator (BOP)
Instrumentation & Control Technician
Maintenance Technician
Health Physics Technician
Pre-Incident Operational Mode
Power Operation (≥90% rated)
Reduced Power Operation (50-90%)
Startup Mode
Hot Standby
Hot Shutdown
Cold Shutdown
Refueling Outage
Testing Mode
Pre-Incident Reactor Power Level (% of rated thermal output)
Baseline Core Thermal-Hydraulic Parameters
Measurement Location | Inlet Temp (°C) | Outlet Temp (°C) | Flow Rate (kg/s) | Pressure (MPa) | |
|---|---|---|---|---|---|
Core Zone A (Upper) | 280 | 320 | 450 | 15.5 | |
Core Zone B (Middle) | 282 | 318 | 445 | 15.5 | |
Core Zone C (Lower) | 285 | 315 | 440 | 15.5 | |
Primary Coolant System Pressure (MPa)
Recent Maintenance or Testing Activities (Last 72 Hours)
Were any pre-existing alarms or equipment degradations active at baseline?
Environmental & Ambient Conditions at Baseline
Parameter | Value | Unit | |
|---|---|---|---|
Ambient Temperature | 22 | °C | |
Atmospheric Pressure | 101.3 | kPa | |
Relative Humidity | 45 | % | |
Seismic Activity Level | 0.01 | g | |
Provide a precise chronological account of the thermal anomaly detection and progression. Include all measured deviations from baseline with timestamps and affected systems.
Anomaly Detection Method
Automatic Digital Control System Alarm
Operator Routine Surveillance
Periodic Data Review
Maintenance Activity Discovery
External Monitoring System
Predictive Analytics Alert
Initial Anomaly Detection Timestamp
Temperature Deviation Measurements
Core Zone | Baseline Temp (°C) | Anomaly Temp (°C) | Delta T (°C) | Rate of Change (°C/min) | |
|---|---|---|---|---|---|
Core Zone A (Upper) | 320 | 342 | 22 | 2.2 | |
Core Zone B (Middle) | 318 | 325 | 7 | 0.8 | |
Core Zone C (Lower) | 315 | 318 | 3 | 0.5 | |
0 | |||||
0 | |||||
0 | |||||
0 | |||||
0 | |||||
0 | |||||
0 |
Maximum Temperature Observed Anywhere in Core (°C)
Design Limit Safety Margin Remaining (°C)
Affected Core Assemblies or Zones
Upper Core Region (0-30% height)
Middle Core Region (30-70% height)
Lower Core Region (70-100% height)
Central Fuel Assemblies
Peripheral Fuel Assemblies
Control Rod Guide Thimbles
Instrumentation Lances
Duration of Thermal Anomaly Before Stabilization (hh:mm)
Temperature Trend During Event
Rapidly Increasing (>5°C/min)
Gradually Increasing (1-5°C/min)
Oscillating/Fluctuating
Stable but Elevated
Decreasing After Peak
Unstable/Erratic
Did temperature exceed any Technical Specification limit?
Did automatic reactor protection actuation occur (e.g., SCRAM, power reduction)?
Upload SCADA/DCS Trend Data Files (CSV, Excel, or proprietary format)
Upload Thermal Imaging or IR Camera Photos if available
Detailed Chronological Narrative of Event (Include all key timestamps and observations)
Document the condition and performance of the secondary cooling loop and containment systems during and after the thermal anomaly. This includes steam generators, feedwater systems, and pressure boundary integrity.
Affected Secondary Loop Identification
Steam Generator Water Level Deviation from Setpoint (%)
Feedwater Flow Rate Variance from Normal (kg/s)
Was a containment isolation signal generated?
Was visual inspection of secondary side components required?
Secondary Loop Parameter Deviations
Parameter | Baseline Value | During Event | Deviation | Unit | |
|---|---|---|---|---|---|
Steam Pressure | 6.5 | 6.8 | 0.3 | MPa | |
Feedwater Temp | 220 | 215 | -5 | °C | |
Condenser Vacuum | 95 | 92 | -3 | kPa | |
Valve Position Anomalies Observed
Main Steam Isolation Valve (MSIV)
Feedwater Control Valve
Bypass Valve
Relief Valve
Safety Valve
Check Valve
No anomalies observed
Coolant Chemistry Analysis Results (if sampled)
Parameter | Measured Value | Normal Range | Within Limits? | |
|---|---|---|---|---|
pH | 9.2 | 9.0-9.6 | Yes | |
Conductivity | 12 | <20 µS/cm | Yes | |
Dissolved Oxygen | 0.005 | <0.01 ppm | Yes | |
Was any primary-to-secondary leakage indicated?
Radiation Monitoring in Secondary Containment
Monitor Location | Reading (µSv/h) | Alarm Setpoint | Alarm Actuated? | |
|---|---|---|---|---|
Steam Line Monitor | 15 | 50 µSv/h | ||
Feedwater Line Monitor | 8 | 30 µSv/h | ||
Turbine Building Area | 12 | 25 µSv/h | ||
Secondary Loop Structural Integrity Assessment
No Concern
Minor Concern - Monitor
Moderate Concern - Inspect
Significant Concern - Restrict Operation
Severe Concern - Shutdown Required
Upload Secondary Side Inspection Photos or Videos
Comprehensively log all automatic protective actions initiated by reactor protection systems and subsequent manual operator interventions to mitigate the thermal anomaly and restore stable conditions.
Did automatic reactor SCRAM occur?
Were control rods manually inserted or adjusted?
Primary Coolant Pump Adjustments
Pump ID | Pre-Event Speed (%) | Adjusted Speed (%) | Adjustment Time | Reason | |
|---|---|---|---|---|---|
RCP-1A | 100 | 85 | 2:30 PM | Reduce core flow to limit cooldown | |
RCP-1B | 100 | 85 | 2:30 PM | Match loop flow | |
Bypass Valve Operations
Valve ID | Opened? | Opening (%) | Operation Time | Purpose | |
|---|---|---|---|---|---|
RC-459 | Yes | 25 | 2:35 PM | Increase core inlet temp | |
RC-460 | 0 | Not required | |||
Manual Operator Actions Taken (Describe each action in sequence with rationale)
Time Stable Conditions Were Re-established
Effectiveness of Interventions in Controlling Temperature
Post-Stabilization Core Parameters
Parameter | Value | Unit | Within Normal? | |
|---|---|---|---|---|
Reactor Power | 65 | % rated | Yes | |
Core Outlet Temp | 310 | °C | Yes | |
Primary Pressure | 15.2 | MPa | Yes | |
Safety Systems Available & Functional During Event
Emergency Core Cooling System (ECCS)
Containment Spray System
Residual Heat Removal (RHR)
Standby Liquid Control
Automatic Depressurization
Diesel Generators
All systems available
One or more systems unavailable
I confirm all manual actions followed approved procedures and were within operational limits
Were any operational limits or license conditions temporarily exceeded during recovery?
This final section ensures proper event classification, regulatory notification compliance, and formal review by authorized personnel. All safety-significant events must be properly categorized and reported per nuclear safety standards.
Immediate Notifications Completed Within Required Timeframes
Internal Safety Board (within 1 hour)
Corporate Nuclear Oversight (within 1 hour)
National Regulatory Authority (within 24 hours)
International Reporting Obligation
Grid Operator Notification
Local Emergency Planning Committee
No notifications required
Provisional International Nuclear Event Scale (INES) Classification
Level 0 - Deviation (No Safety Significance)
Level 1 - Anomaly
Level 2 - Incident
Level 3 - Serious Incident
Level 4 - Accident with Local Consequences
Level 5 - Accident with Wider Consequences
Level 6 - Serious Accident
Level 7 - Major Accident
Does this event meet criteria for immediate regulatory reporting (e.g., 24-hour phone report)?
Follow-Up Actions Required
Root Cause Analysis Report
Corrective Action Plan
Equipment Failure Investigation
Procedure Revision
Training Review
Design Modification Evaluation
Independent Safety Assessment
No further action required
Estimated Overall Event Significance to Nuclear Safety
Upload All Supporting Documentation (Logs, printouts, photos, procedure references)
Shift Supervisor Digital Signature & Verification
Supervisor Sign-Off Timestamp
Independent Safety Board Reviewer Signature
Additional Reviewer Comments and Independent Assessment
Final Event Classification Approval
Approved as Reported
Upgraded Classification Recommended
Downgraded Classification Recommended
Pending Further Investigation