Archived Web Page - GD–99.2: Guide to the Reporting Requirements for Operating Nuclear Power Plants: Compliance Monitoring

Archived information

Archived information is provided for reference, research or recordkeeping purposes. It is not subject to the Government of Canada Web Standards and has not been altered or updated since it was archived. Please contact us to request a format other than those available.

Archived

This page has been archived on the Web.

Preface

This guidance document provides details of the information that nuclear power plant (NPP) licensees are to report to the Canadian Nuclear Safety Commission (CNSC) as specified in conditions of applicable operating licences.

This document provides direction and clarification to the mandatory reporting requirements and information important to the public, that requires immediate notification. Information for lower safety significance situations or events is also captured for follow-up reporting, scheduled reporting, management control and necessary actions, audit and program review. Reporting requirements are expected to be fulfilled using best management practices and objective-based and risk-assessment principles, in alignment with RD-99.2 and industry norms.

This guidance document serves as a companion to RD-99.2, Reporting Requirements for Operating Nuclear Power Plants: Compliance Monitoring. RD-99.2 sets out the types of reports and the frequency and the applicable time frame for reporting. Reporting timeframe is determined by the safety significance of the event; for example, high safety significance events are reported immediately to the CNSC.

This document contains guidance, explanatory information, forms, and templates provided by the CNSC to assist users in meeting the requirements of RD-99.2. The reporting program provided in this suite of documents is intended to enable the CNSC and industry to track, trend, and analyze situations and events, and to develop improvement initiatives.

This document is a part of a suite of regulatory documents providing event and routine compliance monitoring reporting requirements, and the public information and disclosure requirements for nuclear facilities. Each regulatory document is accompanied by a complementary guidance document.

Key principles and elements used in developing this document are consistent with national and international standards. The complete list of standards is included in the Additional Information section and includes CSA-N285.4, Periodic Inspection of CANDU Nuclear Power Plant Components from the Canadian Standards Association (CSA) and the Boiler & Pressure Vessel Code from the American Society of Mechanical Engineers (ASME).

Nothing contained in this document shall be construed as relieving any licensee from pertinent requirements. It is the licensee’s responsibility to identify and comply with all applicable regulations and licence conditions.

Table of Contents

1 Introduction

1.1 Purpose

This document provides value-added direction and best practice guidance on the implementation of the reporting requirements set out in RD-99.2, Reporting Requirements for Operating Nuclear Power Plants: Compliance Monitoring. Rationale and examples are also provided.This guidance document has a dual purpose:

• to provide guidance on the extent of compliance monitoring reporting, the expected level of detail, and the appropriateness, completeness and timeliness of information that is reported to the Canadian Nuclear Safety Commission (CNSC) by licensees of nuclear power plants (NPPs) pursuant to RD-99.2, Reporting Requirements for Operating Nuclear Power Plants: Compliance Monitoring
• to help interested persons, such as applicants for operating licences for NPPs, licensees and CNSC staff understand the extent of the programs that may be required for collecting and reporting information in accordance with regulatory requirements

1.2 Scope

This guidance document clarifies regulatory requirements by providing supporting how-to information, process and procedural guidance, and examples of industry best practice; it does not create additional requirements beyond those contained in RD-99.2.

1.3 Relevant legislation

The following provisions of the Nuclear Safety and Control Act (NSCA) are relevant to this document:

• Subsections 24(1) and 24(5) of the NSCA give the CNSC the authority to establish classes of licences and to include in those licences any terms or conditions that the CNSC considers necessary for the purposes that are laid out in section 3 of the NSCA. Pursuant to this authority, the CNSC has established operating licences for NPPs and, for the purposes of the NSCA, proposes to incorporate this document into these licences.

2.0 Reporting Requirements

Reports are submitted to the CNSC in order to:

• help the CNSC collect information needed to ensure that an NPP is complying with regulatory requirements
• enable the CNSC to take timely and appropriate action in response to licensee non-compliance or unforeseen circumstances
• permit CNSC staff to compile the data necessary to perform trend analysis

2.1 General requirements

The CNSC requires scheduled reports for the purpose of determining if the licensee is in compliance with regulatory requirements. The scheduled reports will provide information regarding the status and performance of the NPP with respect to key safety and control areas. Safety and control areas include:

• management system
• human performance management
• operating performance
• safety analysis
• physical design
• fitness for service
• radiation protection
• conventional health and safety
• environmental protection
• emergency management and response
• waste management
• security
• safeguards
• packaging and transport

Table 1 lists the documents containing reporting requirements for scheduled reports. If these documents are referenced in the operating licence, the reports are to be submitted according to the requirements in the respective document.

Cross-reference of the safety and control areas with scheduled reports is given in table 2.

The term “NSCA” in this document refers to the Nuclear Safety and Control Act and to any regulations made under the NSCA.

The term “calendar year” is, as defined by the Interpretation Act, a period of 12 consecutive months beginning on January 1.

Where RD-99.2 requires reports to be filed or submitted to the CNSC, the “CNSC” is to be interpreted as the designated CNSC staff contact(s) or a person authorized by the CNSC.

2.2 Requirements for scheduled reporting

2.2.1 Quarterly reports

2.2.1.1 Operations report

For RD-99.2 section 2.2.1.1 items (a) and (b), the licensee may report low safety significant contraventions of the NSCA, Regulations, and licence in the quarterly reports as per General Nuclear Safety and Control Regulations (GNSCR) 29 (3). However, high or medium safety significant contraventions are still reported as GNSCR events under RD-99.1.

For RD-99.2 sections 2.2.1.1 item (d), summary statements regarding the level of conformity with sections 12.0, 13.0 and 14.0 of RD-204, Certification of Persons Working at Nuclear Power Plants, as well as non-compliance with program and process requirements in Part 1 of RD-204 would be acceptable.

For RD-99.2 sections 2.2.1.1 item (e), all situations or events where the shutdown clock is initiated are to be reported whether the repairs are completed within the necessary timeframe or not.

For RD-99.2 sections 2.2.1.1 item (f), it is not necessary to include RD-99.1, table 1, item 11(a) events.

Security-related information that belongs to section 2.2.1 of RD-99.2 should be provided in the quarterly security report referred to in section 2.2.1.3 of the same document.

2.2.1.2 Performance indicators report

Specifications for the performance indicators set out in section 2.2.1.2 of RD-99.2 are in Appendix A.

The data sheets for the performance indicators set out in section 2.2.1.2 of RD-99.2 are available on the CNSC Web site at nuclearsafety.gc.ca.

The performance indicators provide compliance monitoring information for the safety and control areas as shown in tables 3 and 4.

The Performance Indicator Data Sheets are provided for illustrative purposes. Licensees may choose to use the data sheets herein, develop their own, or use a combination of both. However, it is essential that all information requested on the data sheets be provided.

2.2.1.3 Security report

No further guidance is needed at this time.

2.2.1.4 Report on the degradation of nuclear power plant pressure boundaries

The events reported in RD-99.1 that did not include the information provided in the requirements of RD-99.2 are to be included in this report.

Typically, the following information is considered for pressure boundary degradation assessments:

1. date first reported and report number if reported under RD-99.1
2. identification of affected systems
3. identification of equipments and components involved
4. material type and code classification of the affected component
5. design and hydrostatic test pressure of the system if applicable
6. magnitude, size or quantification of the degradation (e.g., deformations shape and size, crack length and depth, approximate leak rates, deviation from set point, etc.)
7. code, standard or methodology used to assess the safety significance of the degradation
8. measures taken or planned to correct, to reduce, and to avoid reoccurrence of the situation
9. identification of supporting documents or relevant information

2.2.2 Annual reports

2.2.2.1 Report on environmental protection performance

No further guidance is needed at this time.

p>2.2.2.2 Report of occupational radiation protection performance

No further guidance is needed at this time.

2.2.2.3 Report on the progress of research and development (R&D) activities

Where a common industry strategy and report exist, mechanisms may be utilized to submit a single report with appropriate references among the licensees.

2.2.2.4 Report on the risk and reliability of the NPP

For RD-99.2 section 2.2.2.4, item a), a brief explanation should be provided for changes from previous years, in the list of systems important to safety or in the assigned reliability target. The licensee provides the information included in the template provided in Appendix B.

For item e) iii, the licensee may choose to use bounding evaluations for specific impact calculations. The cumulative impact of test deferrals needs to account for all test deferrals for the system over the year.

For item g) i, examples of abnormal plant configurations to be reported include deviating from the probabilistic safety assessment or safe operating envelope.

2.2.2.5 Report on fuel performance

The licensee provides the information included in the template available on the CNSC Web site at nuclearsafety.gc.ca.

2.2.2.6 Report on conducted training, examination, and testing of personnel for assignment to positions requiring CNSC certification

This section refers to certification training, initial examinations and requalification testing conducted in the previous calendar year. The required dates include the start and end day, month and year of each distinct activity, course or stage.

2.2.2.7 Planning report on scheduled training, examination, and testing of personnel for assignment to positions requiring CNSC certification

This section refers to the proposed dates or estimated timeframe established by the licensee for planning and scheduling purposes. The CNSC expects these dates to be adjusted as operationally required. As these changes are made, the licensee is required to inform the CNSC through a letter or via email.

2.2.2.8 Report on reactivity management

This report includes:

1. brief description of reactivity management occurrences of situations or events that were included in the estimation of the reactivity management index, including the date(s) of occurrence and the associated circumstances, causes, consequences and corrective actions
2. brief description of frequency of unplanned Adjuster Rods (AR) withdrawal and percentage of operating time ARs were withdrawn; this should include the date(s) of occurrence and the associated circumstances, causes and consequences
3. brief description of spatial control performance and number of occurrences for individual zone levels less than 20% full or exceeding 75% full, bulk control performance (percent of time Average Zone Level is at target average level) and frequency of Liquid Zone Control related problems, such as level oscillations; this should include the date(s) of occurrence, the associated circumstances, causes and consequences, and any corrective actions
4. brief description of frequency of post-fuelling deratings, and frequency and percentage of operating time that the core average irradiation limit assumed in the safety analysis was exceeded; this should include the date(s) of occurrence, the associated circumstances, causes and consequences, and any corrective actions
5. name and address of the sender of the report, the date of completion of the report, and the name and signature of the designated representative of the licensee

2.2.3 Other reports

2.2.3.1 Accounting and reporting of uranium, plutonium and thorium

No further guidance is needed at this time.

2.2.3.2 Updates to facility descriptions and final safety analysis report

When submitting an Update to the Final Safety Analysis Report according to RD-99.2 Section 2.2.3.2, the licensee includes at the beginning of the report a Safety Report Revision Summary Sheet highlighting the differences between the existing and the updated safety reports.

The revision summary is to include:

1. list of sections with changes in existing analyses or any new analyses
2. reasons for updating the analysis
3. summary of key changes in methodologies, major assumptions or reactor operating states
4. changes in findings or results that may cause a reconsideration of the safety report
5. list of affected operational documents

2.2.3.3 Station security report

No further guidance is needed at this time.

Appendix A: Performance Indicator Specifications

A.1 Accident Severity Rate, Accident Frequency Rate

1.0 Purpose:

• To indicate the accident severity rate and accident frequency rate at NPPs.
• To monitor performance in meeting nuclear industry standards in the area of worker safety.  
• To compare Canadian NPP performance internationally.

2.0 Definition:

• The accident severity rate (ASR) is the total number of days a person is unavailable for work due to any disabling injuries per 200,000 person hours worked at a NPP. Accident frequency rate (AFR) is the number of disabling injuries per 200,000 person hours worked at a NPP. The value 200,000 person hours is equivalent to 100 person years worked.

3.0 Calculation:

ASR = Number of days disabled  / Number of person hours x 200,000

AFR = Number of disabling injuries / Number of person hours x 200,000

4.0 Notes:

4.1 A disabling injury is one that prevents an employee from reporting for work or from effectively performing all the duties connected with the employee’s regular work.

4.2 An employee is any individual (contractor, temporary staff) performing work at the NPP.

4.3 The Canadian federal reporting requirement for severity includes non-work days not worked. For example, a person is hurt on the last regularly scheduled shift and then is away for two days that were regularly scheduled off. If the person would not have been able to work those two days, but was able to return to work on the first regularly scheduled day, those two days would be counted as lost days.

4.4 Recurrent injuries are attributed back to the originating accident. For example, if an injury from an accident that resulted in a lost time injury occurred in 2008 and then recurred in 2010, with no new accident, the lost days would not appear in the 2010 totals. These days are attributed back to 2008.

4.5 Permanent (or partial) disability resulting from a disabling injury can be assigned equivalent lost days by the licensee. Submit appropriate information showing conversion data for each disabling injury.

A.2 Chemistry Index

1.0 Purpose:

• To indicate long-term unit control of important chemical parameters.
• To monitor performance in meeting licensee's requirements in chemistry.
• To compare performance between Canadian reactor units.

2.0 Definition:

The average percentage of time that the selected chemical parameters are in specification during the quarter.

3.0 Calculation:

Chemistry Index (%) =  x 100%

where:

a_i = the number of hours that parameter “i” is in specification during the quarter  

A = the number of hours the plant is in an operational state during the quarter, as defined by licensee-specific documentation

a_i/A = the fraction of time that parameter “i” is in specification during the quarter

m = the number of parameters monitored during the period, usually the 15 parameters on the list below

  is the sum of individual time-in-specification fractions for each parameter monitored in the index.


All data is dimensionless. The Chemistry Index (CI) results will range
between 0% and 100%.

4.0 Parameters Monitored:

Primary Heat Transport System:                    Steam Generators:

- pHa (calc)                                                                    - [Cl^-]

- dissolved D₂                                                                - [SO₄^2- ]

- chloride                                                                        - [Na⁺]

- fluoride

- conductivity

Annulus Gas [O₂]

Feedwater:                                            Condensate Extraction Pump:

- dissolved [O₂]                                      - dissolved [O₂]

- total iron           - pH

- total copper

- hydrazine

5.0 Notes:  

5.1 The Chemistry Index is reported as the percentage of time in specification and is calculated as follows for each parameter:



The initiation of an out-of-specification event occurs with the first result measured outside the range of the specification, as indicated in licensee-specific chemistry program documentation.

Termination of the event can only be achieved by reducing the control parameter within the specification range. Calculate the duration of the out-of-specification condition as the time between the first out-of-specification sample and the next measured in-specification sample. The % time in-specification is then calculated as 100% - S (% time of measurement period out-of specification).

The total operating hours in the period refers to the total operating hours for the system that the chemical parameter pertains to.

5.2 Parameters included in the indicator but not measured (because the monitoring capability did not exist or whose measurements were not obtained during the period, e.g., an instrument not available) will be reported as being out-of-specification. In cases where the parameter is out-of-specification due to the unavailability of a facility, the parameter is reported as being out-of-specification.

When the safety of chemical technicians or employees could be adversely affected by new hazards during normal execution of their tasks, or when the status of the plant is such that the chemical measure is useless or unrepresentative, the representative period will be adjusted without penalty. Such measures will be qualified "void". The data is to be auditable.

In some cases a temporary exemption is granted for measurement of a parameter, or deviation of a specification from the range specified in program documentation. The chemistry program authority grants this exemption. It is acceptable to indicate ‘not applicable’ for the parameter for the specific time period. If temporary exemption applies for less than one quarter, calculate the time-in-specification for the parameter as the time that the temporary exemption does not apply. Temporary exemptions are intended to be used when instruments or facilities are not available for a significant period of time, where there are temporary modifications to licensee procedures due to new concerns regarding unsafe conditions, or when short duration trials are being carried out. Records pertaining to the temporary exemption are auditable.

Whenever the parameter is deemed not applicable (N/A), a short explanatory note, as well as a reference to the licensee’s documentation for any temporary exemption, should accompany the submission on these performance indicators.

In cases where the parameter is deemed N/A for a particular period of time, the number of parameters in the equation in Section 3 should be adjusted to reflect the number of parameters actually tracked during the specific quarter.

5.3 For those systems whose performance is reported only for unit operating conditions, the following guideline is to be utilized:

If a parameter is in (or out of) specification before a shutdown, it is considered to remain in (or out of) specification once the system is back in service until it is re-analyzed and found to be otherwise.

5.4 Report performance for all Chemistry Index and Chemistry Compliance Index parameters using the specifications documented in the most current revision of the licensee’s Chemical Specifications Manual. Performance must be reported for all time periods when the system is considered to be in an operational state, as defined by licensee-specific documents.

5.5 The reference chemical specifications and sampling frequency for each parameter is as documented in the most current revision of the licensee’s Chemical Specifications Manual. Note any deviations from these reference values for reporting on these performance indicators in report submissions. Document any changes to the specifications and sampling frequency in the Chemical Specifications Manual and supporting documentation.

The minimum sampling/monitoring frequency is determined by the licensee's current requirements.

5.6 Each station will determine whether results from grab samples or on-line instrument readings will be used to calculate the performance. On-line instrument readings are the preferred method if an adequate quality management program or management system is in place to ensure accuracy.

Where on-line monitoring equipment is available, the success ratio is calculated as the ratio of time where the monitoring is on-line and valid data is available and within range over total time. When monitoring equipment fails, it is permissible to replace the monitoring with manual sampling techniques at a reasonable frequency.

5.7 For multi-unit sites, the unit performance is the average of the performance of the individual control parameters. On an operating unit basis:



5.8 The station result is the time-weighted average of the operating units' Chemistry or Chemistry Compliance index values. This ensures that units that were operating for only part of the period are not given the same weight as those that operated for the whole period.



5.9 Performance need not be reported for parameters during short duration trials or tests being conducted to optimize chemistry that affects those parameters.

5.10 The CNSC reviews parameters making up the list of the index, and the definitions of time-in-specification and voiding.

A.3 Chemistry Compliance Index (Non-GSS and GSS)

1.0 Purpose:   

• To indicate unit control of safety-related chemical and radiochemical parameters.
• To monitor performance in meeting regulatory and licensee requirements in chemistry.
• To compare performance between Canadian reactor units.

2.0 Definition:

The average percentage of time that the selected chemical parameters are in specification during the quarter.

3.0 Calculation:

Sampling frequencies and specifications are defined in the licensee’s operating documentation. The method for calculation of the Chemistry Index also applies to this Chemistry Compliance Indices (Non-GSS (Guaranteed Shutdown State), GSS). Parameters are based on safety and selected as compliance parameters in accordance with most OP&P requirements.

4.0 Parameters Monitored:

Non-GSS Operating Conditions

[Gd] in Liquid Injection Safety System Poison Injection Tanks

[Gd] in Moderator (unit in poison outage, SDS2 actuated)

Moderator D₂O Isotopic

Moderator ³H

Moderator Cover Gas D₂
Moderator Conductivity

Primary Heat Transport System D₂O isotopic

Primary Heat Transport System ³H

Primary Heat Transport System ¹³¹I

Primary Heat Transport System D₂O Storage Tank Cover Gas D₂

Moderator to Primary Heat Transport System D₂O Isotopic Purity Difference Check

Annulus Gas System Dewpoint

End Shield Cooling Water pH

End Shield Cooling Cover Gas H₂ (for Point Lepreau, Gentilly-2, Pickering B)

Emergency Coolant Injection (ECI) System High Pressure (HP) Water Tank(s) pH

ECI HP Water Tank(s) Hydrazine Concentration

Liquid Zone Control System Cover Gas [H₂]

Liquid Zone Control System Conductivity

GSS Conditions

Liquid Injection Safety System Poison Injection Tanks pH_a (when SDS2 is available)

[Gd] in Moderator

Moderator D₂O Conductivity (except for Gentilly-2)

Moderator D₂O pH_a

Supplementary parameter(s) sampled

5.0 Notes:

5.1 The Chemistry Compliance indices (Non-GSS and GSS) is to be reported as the percentage of time in specification and is calculated as follows for each parameter:



The initiation of an out-of-specification event occurs with the first result measured outside the range of the specification, as indicated in licensee-specific chemistry program documentation.

Termination of the event is achieved only by reducing the control parameter within the specification range. The duration of the out-of-specification condition will be calculated as the time between the first out-of-specification sample and the next measured in-specification sample. The % time in specification is then calculated as 100% - S (% time of measurement period out-of specification).

The total operating hours in the period refers to the total operating hours for the system that the chemical parameter pertains to.

5.2 Parameters included in the indicator but not measured (because the monitoring capability did not exist or measurements were not obtained during the period (e.g., instrument not available)), will be reported as being out-of-specification. In cases where the parameter is out-of-specification due to the unavailability of a facility, report the parameter as being out-of-specification.

When the safety of chemical technicians or employees could be adversely affected by new hazards during normal execution of their tasks, or when the status of the plant is such that the chemical measure is useless or unrepresentative, the representative period will be adjusted without penalty. Such measures will be qualified, "void". The data is auditable.

In some cases, a temporary exemption is granted for measurement of a parameter or deviation of a specification from the range specified in program documentation. This exemption is granted by the chemistry program authority. It is acceptable to indicate not applicable (N/A) for the parameter for the specific time period. If the temporary exemption applies for less than one quarter, the time-in-specification for the parameter is calculated as the time that the temporary exemption does not apply. Temporary exemptions are intended to be used when instruments or facilities are not available for a significant period of time, where there are temporary modifications to licensee procedures due to new concerns regarding unsafe conditions, or when short duration trials are being carried out. Records pertaining to the temporary exemption are auditable.

Whenever the parameter is deemed N/A, a short explanatory note, as well as a reference to the licensee’s documentation for any temporary exemption, should accompany the submission on these performance indicators.

In cases where the parameter is deemed N/A for a particular period of time, the number of parameters in the equation in Section 3 should be adjusted to reflect the number of parameters actually tracked during the specific quarter.

5.3 For those systems whose performance is reported only for unit operating conditions, the following guidelines are to be utilized:

If a parameter is in (or out of) specification before a shutdown, it is considered to remain in (or out of) specification once the system is back in service until it is re-analyzed and found to be otherwise.

5.4 Performance must be reported for all Chemistry Index and Chemistry Compliance Index parameters using the specifications documented in the most current revision of the licensee’s Chemical Specifications manual. Performance must be reported for all time periods when the system is considered to be in an operational state, as defined by licensee-specific documents.

5.5 Document the reference chemical specifications and sampling frequency for each parameter as per the most current revision of the licensee’s Chemical Specifications manual. Note any deviations from these reference values for reporting on these performance indicators in report submissions. Document any changes to the specifications and sampling frequency in the Chemistry Specifications manuals and supporting documentation.

The minimum sampling/monitoring frequency is determined by the licensee's current requirements.

5.6 Each station will determine whether results from grab samples or on-line instrument readings will be used to calculate the performance. On-line instrument readings are the preferred method if an adequate quality management program or management system is in place to ensure accuracy.

Where on-line monitoring equipment is available, the success ratio will be calculated as the ratio of time where the monitoring is on-line and valid data is available and within range over total time. When monitoring equipment fails, it is permissible to replace the monitoring with manual sampling techniques at a reasonable frequency.

5.7 For multi-unit sites, the unit performance is the average of the performance of the individual control parameters. On an operating unit basis:



5.8 The station result is the time-weighted average of the operating units' Chemistry or Chemistry Compliance index values. This ensures that units operating for only part of the period are not given the same weight as those that operated for the whole period.



5.9 Performance of parameters need not be reported during short duration trials or tests being conducted to optimize chemistry and which affect those parameters.

5.10 The CNSC reviews parameters making up the list of the index, as well as the definitions of time-in-specification and voiding.

A.4 Change Control Index

1.0 Purpose:

• To indicate the control over changes in equipment and procedures for safety-related systems.
• To monitor performance in the management of change for safety-related systems.

2.0 Definition:

Change control index is the count of the number of temporary changes used to maintain control over NPP equipment and procedures.

3.0 Calculation:

3.1 Total number of pages of temporary procedural changes.

3.2 Total number of pages of temporary procedural changes over six months old.

3.3 Total number of temporary equipment changes.

3.4 Total number of temporary equipment changes over six months old.

3.5 Total number of incomplete permanent equipment changes.

4.0 Notes:

4.1 All equipment or procedural changes for the indicator are to include the safety-related systems identified by the NPP staff.

4.2 Permanent changes are considered incomplete until all testing, design, installation, and operating documentation have been amended.

5.0 Data is collected on a unit basis. Include data for common unit (unit 0) for multi-unit NPPs.

A.5 Radiological Emergencies Performance Index

1.0 Purpose:

• To provide a measurement of the performance of an NPP’s emergency preparedness plan during radiological emergencies.

2.0 Definition:

This index is the percentage of all the successful performance opportunities to the total number of performance opportunities identified during the previous eight quarters.

3.0 Calculation:

Radiological Emergencies Performance Index

4.0 Notes:

4.1  Performance opportunities are:

• categorizing a radiological emergency  
• notifying offsite authorities
• providing decision making information to local authorities
• developing protective action recommendations

4.2 A performance opportunity is successful when both the timeliness and the accuracy criteria are fulfilled.

4.3 The timeliness criteria and the accuracy criteria are specified in the licensee’s emergency preparedness plan.

4.4 Included details of any failed performance opportunity in the explanation of data section of the data sheet.

4.5 Emergencies, drills evaluated by the Emergency Response Organization (ERO), exercises, and other simulated emergencies which are assessed and that interact with one or more of the facilities or functions listed in 5.6 are to be included in this indicator.

4.6 The ERO consists of the following facilities and functions:

• Control Room
• Technical Support Group, Technical Advisory Group, Site Management Centre
• Operations Support Group
• Emergency Operations Facility
• Emergency Response Teams
• Field Monitoring Teams
• Damage Control Teams
• Joint Information or Local Media Centre
• Offsite Governmental Authorities

4.7 Training practices cannot be included in this indicator.

A.6 Emergency Response Organization (ERO) Drill Participation Index

1.0 Purpose:

To track the participation of Emergency Response Organization (ERO) personnel in drills, exercises, or events within an NPP.

2.0 Definition:

This index is the percentage of the total available ERO personnel who have participated in proficiency-enhancing drills, exercises, practical evaluation opportunities, or in events during the previous eight quarters, inclusive of the quarter being reported.

3.0 Calculation:

ERO Drill Participation Index = [A / B] x 100%

where:

A =        number of ERO personnel fulfilling designated ERO positions that have participated in a qualifying drill, exercise, practical evaluation, or event during the previous 8 quarters

B =        total number of ERO personnel fulfilling designated ERO positions during the previous 8 quarters

4.0 Notes:

4.1 Designated ERO positions are those performing the following functions:

• categorization of event
• offsite notification
• plant operations
• corporate resources
• radiological monitoring
• dose projection

4.2 Count only for the most recent participation of ERO personnel who have participated in more than one event during the previous eight quarters.

4.3 Reflect any changes in the number of designated ERO personnel in both the numerator and denominator of this index.

4.4 Multiple assignees to a given designated ERO position may each be counted for their individual participation in performing the designated ERO position at different times in the same proficiency-enhancing drill, exercise, practical evaluation opportunity, or event during the previous eight quarters.

A.7 Emergency Response Resources Completion Index

1.0 Purpose:

To indicate the level of verification of emergency response equipment and facilities dedicated to emergency preparedness at an NPP.

2.0 Definition:

This index is the percentage of preventative maintenance items, tests and checks completed for the emergency response equipment and facilities over the total number of preventative maintenance items, tests, and checks scheduled during the quarter.

3.0 Calculation:

ER Resources Completion Index = [A/B] x 100%

where:

A =       the number of preventative maintenance items, tests and checks completed during the quarter

B =       the number of preventative maintenance items, tests and checks scheduled during the quarter

4.0 Notes:

4.1 Licensees are to provide to the CNSC a complete listing of preventative maintenance items, tests and inventory checks pertaining to dedicated equipment and facilities that are used for emergency preparedness, including:

• fixed systems
• portable instruments
• communications equipment
• other equipment identified in the licensee’s Emergency Preparedness Plan required to be in a state of readiness
• other dedicated emergency equipment and facilities identified in the CNSC regulatory guide G-225, Emergency Planning at Class I Nuclear Facilities and Uranium Mines and Mills

4.2 Measure schedule compliance using predetermined dates for the preventative maintenance items, tests and checks. This index measures the number of preventative maintenance items, tests and inventory checks performed and compares them with the predetermined fixed schedule. It is not concerned with whether a test has failed or whether equipment was missing.

A.8 Non-Compliance Index

1.0 Purpose:

• To indicate the number of occurrences where the operation of the NPP failed to comply with its licence conditions, or with the NSCA and regulations.
• To compare performance between Canadian NPPs.

2.0 Definition:

Non-compliance is a failure to comply with the NSCA and regulations or any condition of the licence, including documents referenced therein.

3.0 Calculation:       

Non-compliance index = total number of situations or events that have at least one non-compliance in the following categories

Non-compliance sub-index = total number of non-compliances for each of the categories

Categories:

• number of non-compliances of the Operating Policies & Principles referenced in the licence
• number of non-compliances of the radiation protection requirements referenced in the licence
• number of non-compliances with the minimum shift complement referenced in the licence
• number of other non-compliances with the licence
• number of non-compliances with the NSCA and regulations

4.0 Notes:

4.1 A reportable situation or event is as defined by RD-99.2.

4.2 The values used in the calculation will be based on the date of reporting of the situation or event.

4.3 For each situation or event reported, the licensee must indicate which non-compliance category applies. Use only the non-compliance category which most accurately describes the situation. Do not double count.

4.4 In multi-unit NPPs, the licensee must indicate the applicable unit(s) of each non-compliance.

A.9 Number of Pressure Boundary Degradations

1.0 Purpose:

• To indicate the number of pressure boundary degradations that have occurred at the NPP.
• To monitor the performance in meeting nuclear industry codes and standards.
• To compare performance between Canadian NPPs.

2.0 Definition:

Pressure boundary degradation: A degradation of a pressure boundary that exceeds any relevant limit specified in the applicable design analysis, design codes or standards, or inspection codes or standards.

Pressure boundary: A boundary of a pressure-retaining vessel, structure, system, or component of a nuclear or non-nuclear system, where the vessel, structure, system, or component is registered, or eligible for registration, under boiler or pressure vessel legislation or where the vessel, structure, system or component is part of the NPP containment system.

3.0 Calculation:

The sum of the number of pressure boundary degradations in each of the following categories:

a_n1 =      number of pressure boundary degradations in Class 1 nuclear systems reported under RD-99.2 in a given quarter

a_n2 =      number of pressure boundary degradations in Class 2 nuclear systems reported under RD-99.2 in a given quarter

a_n3 =      number of pressure boundary degradations in Class 3 nuclear systems reported under RD-99.2 in a given quarter

a_n4 =      number of pressure boundary degradations in Class 4 nuclear systems reported under RD-99.2 in a given quarter

a_nc =      number of pressure boundary degradations in conventional systems reported under RD-99.2 in a given quarter

4.0 Notes:

4.1 For examples of pressure boundary degradations, see main body of RD-99.1.

4.2 In the above categorization, classes 1, 2, 3 and 4 refer to what the code class of the failed or degraded component would be under the requirements of the CSA standard N285.0, General Requirements for Pressure-retaining Systems and Components in CANDU Nuclear Power Plants.

A.10 Preventive Maintenance Completion Ratio

1.0 Purpose:

• To monitor the effectiveness of the preventive maintenance program in minimizing the need for corrective maintenance activities.
• To compare performance between Canadian NPPs.

2.0 Definition:

Preventive maintenance completion ratio is the ratio of preventive maintenance (PM) jobs completed divided by the preventive maintenance plus corrective maintenance (CM) jobs completed.

Preventive maintenance is an action that detects, precludes or mitigates degradation of a functional SSC to sustain or extend its useful life by controlling degradation and failures to an acceptable level. Preventive maintenance may be periodic, planned or predictive.

Corrective maintenance is an action that, by means of repair, overhaul or replacement, restores the capability of a failed SSC to perform its defined function within acceptance criteria.

3.0 Calculation:

4.0 Notes:

4.1 Preventive maintenance jobs are jobs performed on the safety-related system equipment in the field that is in working order when the job commences. Preventive maintenance jobs include those that are frequency or condition based.

4.2 Corrective maintenance jobs are performed as a result of a reported failure of safety-related system equipment. The maintenance jobs cannot include design modifications.

4.3 Work orders on safety-related systems are work orders that are written during the quarter to correct deficiencies. Work orders must have undergone a preliminary review by the NPP work management group and be designated as valid to be included in the count for a quarter.

4.4 Report data by unit, including the common unit for multi-unit NPPs. Jobs are counted by work order issued to each discipline, not on a task basis. Count a work order that covers repetitive jobs for multiple equipment as one order for each separate piece of equipment.

A.11 Radiation Occurrence Index

1.0 Purpose:

• To indicate the number and weighted severity of radiation occurrences that have taken place at the NPP.
• To monitor performance in meeting the CNSC’s expectations in the area of worker radiation protection.

2.0 Definition:

A radiation occurrence is an occurrence where one or more of the following has occurred:

• fixed body contamination exceeding 50 kBq/square metre (1.35 Ci/m²) has been detected
• an unplanned acute whole body dose (resulting from an external exposure) exceeding 5 mSv (500 mrem) has been received
• an unplanned acute intake of radioactive material resulting in an effective dose greater than 2 mSv (200 mrem) has taken place
• an acute or committed dose in excess of any of the limits specified in the Radiation Protection Regulations Part 1 (Ontario Power Generation), Radiation Protection Regulations (Bruce Power), Directives de Sant& et Normes de Radioprotection (Hydro-Quebec), or the Radiation Protection Regulations (New Brunswick Power) has been received

3.0 Calculation:

Radiation occurrence index   = a + 5b + 5c + 50d

where:

a =        number of occurrences in a quarter, after decontamination attempts, where fixed body contamination in excess of 50 kBq/square metre (1.35 μCi/m²) was measured

b =        number of occurrences in a quarter where an unplanned acute whole body dose (resulting from an external exposure) exceeding 5 mSv (500 mrem) was received

c =         the sum of the effective doses, normalized to 2 mSv, for occurrences in a quarter where an intake of radioactive material resulted in an effective dose greater than 2 mSv (200 mrem)

d =        number of occurrences in a quarter where an acute or committed dose in excess of any of the limits specified in the Radiation Protection Regulations Part 1 (Ontario Power Generation), Radiation Protection Regulations (Bruce Power), the Directives de Santé et Normes de Radioprotection (Hydro-Québec), or the Radiation Protection Regulations (New Brunswick Power) was received.

4.0 Notes:

4.1 The values of the weighting factors (1, 5, 5, and 50) required to differentiate between the least safety significant radiation occurrences and the most serious ones are not risk-based. They are an approximation of the significance for each type of situation or event.

4.2 The total for ‘c’ above is normalized to 2 mSv. This means that the actual dose received as a result of the occurrence(s) is divided by 2 mSv. For example: two unplanned exposures as a result of separate intake incidents, one of 3 mSv and one of 4 mSv, would result in a value for ‘c’ =  3 mSv/2 mSv + 4 mSv/2 mSv = 1.5 + 2 = 3.5. This value ‘c’ would then be multiplied by the weighting factor 5.

4.3 Only the higher level of consequence will be calculated for a single occurrence to prevent double counting. For example: if a single occurrence results in an unplanned acute whole body dose exceeding 5 mSv (500 mrem), and in the affected worker’s whole body dose exceeding the legal limit, the occurrence would be counted as a “d” type of occurrence (the most serious type) and not as one “b” type and one “d” type.

4.4 As an example of the calculation of the Radiation Occurrence Index, given the following values:

a =        2 occurrences in a quarter, after decontamination attempts, where fixed body contamination in excess of 50 kBq/square metre (1.35 μCi/m²) was measured

b =       1 occurrence in a quarter where an unplanned acute whole body dose (resulting from an external exposure) exceeding 5 mSv (500 mrem) was received but did not exceed the plant radiation dose limits specified in their radiation protection regulations

c =       3.5 example from 4.2 above for two unplanned exposures of 3 mSv and 4 mSv

d = 1 occurrence in a quarter where an acute or committed dose exceeded any of the limits specified in the plant radiation protection regulations and not also included in the count for “b”

Radiation Occurrence Index = a + 5b + 5c +50d
= 2 +5(1) + 5(3.5) + 50(1)
= 74.5

A.12 NPP Radiation Dose

1.0 Purpose:

• To indicate the total dose due to ionizing radiation received by all individuals working at the NPP and its related facilities.
• To monitor the performance in keeping NPP collective effective dose as low as reasonably achievable.
• To compare Canadian NPP performance internationally.

2.0 Definition:

The sum of the whole body doses due to ionizing radiation received by individuals (including permanent, part-time and temporary staff, external contractors, consultants, etc.) working at the NPP and its related facilities, over the year.

The sum of the total estimated external dose from NPP Outages.

3.0 Calculation:

Total Collective Effective Dose = internal + external dose for the NPP

Average dose per unit = Collective Effective Dose / Number of units operating or being rehabilitated

Average dose per worker = Collective Effective Dose / Number of workers at NPP receiving non-zero dose

Total estimated external dose per outage = Sum of estimated external doses during unit outages (planned or forced)

4.0 Data Presentation:

4.1 Dose data to be presented includes:
• collective effective dose:
• external
• internal
• total
• average dose per unit
• average dose per worker
• status of the reactor units
• outage collective dose:
• planned outage dose – external
• forced outage dose – external
• total estimated external dose per outage
• outage unit number
• outage duration (days)
• reactor under normal operation collective dose – external dose
• individual dose distribution (grouped according to the following distributions: 0 to 1.0 mSv, 1.01 to 5.0 mSv, 5.01 to 10.0 mSv, 10.01 to 15 mSv, 15.01 to 20.0 mSv and greater than 20.0 mSv)
• median worker effective dose
• maximum worker effective dose
• neutron dose summary (i.e., number of individuals assigned a neutron dose, maximum neutron dose received by an individual, collective neutron dose)
• extremity dose summary (i.e., number of individuals assigned an extremity dose, maximum extremity dose received by an individual, collective extremity dose)
• skin dose summary (i.e., number of individuals assigned a skin dose, maximum skin dose received by an individual, collective skin dose)

4.2 For the work group dose data, the information to be presented for each work group is the work group title, the number of persons, the collective dose (external, internal and total), the maximum dose, the median dose, and the percent of the total station dose.

5.0 Notes: 

5.1 A unit outage is considered “planned” when preparations and the decision to shut down the unit have been made in advance, regardless if the unit enters into the outage at an earlier time than the target date due to a forced shutdown.

5.2 Dose related to work during unplanned power outages, forced shutdowns or power manoeuvres, or short outages such as poison prevent operation, are to be reported under “forced outages”. Also doses from activities on a unit that is decommissioned, mothballed or in lay-up are not to be reported under “refurbishment activities or safe storage”.

5.3 Doses attributed to an outage include the doses directly related to the preparation of the outage, the start-up and related power manoeuvres.

5.4 Doses from neutron exposures should normally be accounted for, in this indicator.

5.5 For the purposes of the CNSC performance indicator program, whole body doses can be considered equivalent to deep doses.

5.6 Total NPP Collective Effective is to include all doses (outage and operation) received by individuals (including permanent, part-time and temporary staff, external contractors, consultants, etc.) working at the NPP and its related facilities, over the year.

5.7 Dose commitments to staff resulting from common NPP services such as Unit zero, heavy water upgrading, or from activities in facilities directly related to the operation of the unit such as laundry, decontamination, fuelling, waste handling, are to be included in the “Total Collective Effective Dose”.

5.8 Doses to visitors or from initiatives related to public information programs are to be excluded from this indicator.

A.13 Number of Missed Mandatory Safety System Tests

1.0 Purpose:

• To indicate successful completion of tests required by licence condition, including those referenced in documents submitted in support of a licence application.
• To monitor performance in meeting regulatory and licensee availability requirements.

2.0 Definition:

The number of missed safety system tests is the sum of those tests that are not completed for each of the three groups of safety-related systems (i.e., the Special Safety Systems, the Standby Safety Systems, and other Safety-Related Process Systems) reported under RD-99.2 in a quarter.

3.0 Calculation:       

Number of Missed Tests = a + b + c

where:

a = number of missed tests for the Special Safety Systems

b = number of missed tests for the Standby Safety Systems

c = number of missed tests for the Safety-Related Process Systems

4.0 Notes:

4.1 For the purpose of this performance indicator, the following is to apply:

• Special Safety Systems: SDS1 (Shutdown System 1), SDS2 (SDSE (Shutdown System Enhancement) for Pickering Nuclear Generating Station-A), ECC (Emergency Core Cooling), and Containment
• Standby Safety Systems: Boiler Emergency Cooling, Emergency Power Supply, Standby Generators, Emergency Filtered Air Discharge, Emergency Water, Inter-Unit Feedwater Tie
• Safety-Related Process Systems: Reactor Regulating, Heat Transport, Moderator, Class I, II, and III, Auxiliary Boiler Feed, Service Water

This list may be expanded in the future.

4.2 Missed tests refer to those not completed, as opposed to those that fail.

4.3 Tests conducted beyond the maximum allowable time interval (test interval + grace period) will count as missed test, unless they have been deferred in accordance with procedures that are permitted by the licence or approval has been obtained from the CNSC to extend the test interval.

4.4 For multi-unit stations, station-wide tests are to be reported under unit 0.

4.5 For benchmarking, report the total number of tests performed for each category, a, b, c.

A.14 Number of Unplanned Transients

1.0 Purpose:

To indicate the number of reactor power transients due to equipment failures or operator errors while the reactor is not in a guaranteed shutdown state.

2.0 Definition:

The unplanned transients are the situations or events that result in a change of reactor operating states due to:

(a.) unplanned reactor setbacks and stepbacks, both automatic and manual, which occur while the reactor is not in a guaranteed shutdown state; these reactor setbacks and stepbacks are the events resulting from the corrective actions taken by operator or the internal plant equipment failure, spurious signal, human error or external events such as severe weather, earthquake, aeroplane strike, grid instability, railway explosion, etc.

(b.) unplanned reactor trips, both automatic and manual, which occur while the reactor is not in a guaranteed shutdown state; these reactor trips are the result of the events arising from corrective actions taken by operator or internal plant equipment failure, spurious signal, human error, or external events such as severe weather, earthquake, aeroplane strike, grid instability, railway explosion, etc.

3.0 Calculation:       

The total number of unplanned transients in a quarter for a unit. The total number of hours during which the reactor is either being placed in GSS or is in GSS.

4.0 Notes:

4.1 The manual reactor trips, setbacks or stepbacks that are required by planned (as opposed to forced) outage maintenance or routine testing are not to be included.

4.2 If a situation or event results in a combination of a reactor setback, stepback and/or trip in sequence, then the total number of transient will be counted in as one.

4.3 If a situation or event results in a reactor trip on both shutdown systems, the number of reactor trips is to only be counted as one.

4.4 After a reset of reactor setback, stepback and/or trip by operator and the reactor power is allowed to increase, if another transient occurs again due to failure to correct causes of the initial transient, then the subsequent reactor setback, stepback and/or trip is to be included in the calculation of the number of unplanned transients.

A.15 Unplanned Capability Loss Factor

1.0 Purpose:

• To indicate how a unit is managed, operated, maintained in order to avoid unplanned outages.
• To compare Canadian NPP performance internationally.

2.0 Definition:

This indicator is defined in WANO (World Association of Nuclear Operators), Implementing Guideline 19.1 (1993), addendum September 1996.

3.0 Calculation:

Reference Energy Generation (REG) = Unit Capacity x referenced period (MW-hrs)

Total Unplanned Energy Loss per quarter (UEL) = Total unplanned energy loss over referenced period (MW-hrs)

Unplanned Capability Loss Factor (UCL) = UEL/REG x100%

4.0 Notes:

Referenced WANO document, WANO IG 19.1 ANR1.OR, September 1996.

A.16 Corrective Maintenance Backlog

1.0 Purpose:

• To indicate one aspect of plant materiel condition through the number of outstanding corrective maintenance work orders on structures, systems and components.
• To compare Canadian NPP performance.

2.0 Definition:

Corrective maintenance is required when an SSC has failed and can no longer perform its design function. Corrective maintenance backlog consists of all corrective work generated through work order requests and appears in the work management system as uncompleted work.

3.0 Calculation:       

The indicator consists of total number of on-line corrective maintenance work orders at the end of the quarter which appears as uncompleted work. The indicator is in the form of work orders per unit.

4.0 Notes:

4.1 Value should be the same as that used in the NPP Stations Equipment Reliability Index.

4.2 For multi-unit stations, Unit 0 is considered a separate unit.

4.3 Outage corrective maintenance is not included in this indicator.

A.17 Elective Maintenance Backlog

1.0 Purpose:

To indicate one aspect of plant materiel condition through the number of outstanding elective maintenance work orders on structures, systems and components.

2.0 Definition:

Elective maintenance is planned when structures, systems or components have been identified as degrading but still capable of performing their design function. Elective maintenance backlog consists of all elective work generated through work requests and appears in the work management system as uncompleted work.

3.0 Calculation:

The indicator consists of the total number of on-line elective work orders at the end of the quarter, which appears as uncompleted work. The indicator is in the form of work orders per unit.

4.0 Notes:

4.1 Value should be the same as that used in the NPP Stations Equipment Reliability Index.

4.2 For multi-unit stations, Unit 0 is considered a separate unit.

4.3 Outage elective maintenance is not included in this indicator.

A.18 Preventive Maintenance Deferrals

1.0 Purpose:

• To indicate the extent to which completion of the preventive maintenance program is being respected. Deferring preventive maintenance can lead to negative plant performance.
• To compare Canadian NPP performance.

2.0 Definition:

A preventive maintenance deferral is a preventive maintenance work order that exceeds its late date and has gone through an approval process that includes an engineering assessment.

3.0 Calculation:

The value is the sum of all predefined preventive maintenance work orders deferred over the quarter. Preventive maintenance that is performed monthly or more frequently may be excluded. The indicator is in the form of work orders per unit per quarter.

4.0 Notes:

4.1 The value is the same as that used in the NPP Stations equipment reliability index.

4.2 For multi-unit stations, Unit 0 is considered a separate unit.

A.19 Average Work Week Schedule Adherence

1.0 Purpose:

• To indicate how effective the NPP is in planning and scheduling work. Good work schedule adherence is also an indication that the maintenance program is not reactive in nature and therefore more effective.  
• To compare Canadian NPP performance.

2.0 Definition:

Work week schedule adherence is the percentage of work completed by the end of a given work week as compared to the total scheduled work for that work week.

3.0 Calculation:

Schedule adherence is the total completed scheduled activities divided by the total scheduled activities expressed as a percentage. The indicator value is the average for all weeks in the quarter. The units of measure are average percentage per unit.

4.0 Notes: 

4.1 Value should be the same as that used in the NPP Stations Equipment Reliability Index.

4.2 For multi-unit stations, Unit 0 is considered a separate unit.

Appendix B: Report on the Risk and Reliability of the NPP

B.1. Summary

Major findings or major changes during the calendar year should be summarized. These include:

• changes in the list of systems important to safety or their reliability targets
• overall performance of systems important to safety, including statistical summary of performance
• changes having major impact on reliability models
• major update of reliability models for systems important to safety
• generic discussion on completion of required surveillance activities, including tests, predefines, and operator routines
• the number of initiating events causing actuation of mitigating systems
• major changes in failure modes and/or failure rates

B.2. List of Systems Important to Safety

RD-99.2 section 2.2.2.4 a): List all identified systems important to safety and, for each system, include the assigned reliability target. Changes from previous years, in the list of systems important to safety or in the assigned reliability target should be discussed and explained.

Table B.1 List of Systems Important to Safety and Reliability Targets

System Important to Safety	Reliability Target

This section may list structures and components important to safety and their reliability targets (if these structures and components have been identified).

B.3. System Performance

Insert a section for each system important to safety according to the format below:

B.3.1 Systems important to safety

RD-99.2 section 2.2.2.4 b): For each system important to safety of the nuclear power plant, include a comparative assessment of the reliability and reliability target for the system. This includes:

• the predicted reliability of the system, where the predicted reliability is calculated using the system up-to-date model and current data
• the observed reliability of the system during the previous year
• a comparison between the predicted reliability, observed reliability and reliability target of the system
• an explanation of changes in the predicted reliability of the system from the predicted reliability reported in previous years
• specific reliability indices for major components such as Class III power systems, and emergency or qualified power systems that include the failures to start (in failures per demand) and failures to run (in failures per hour) for each generator

This section describes comparative reliability indices for the system important to safety identified in section B.2. This section should take into account the following system specific situations:

• one system important to safety may require several different reliability assessments to account for different failure criteria
• the reliability assessment should choose the limiting unit if the system is shared by multiple units and the unitized parts have differences
• the modeled system boundary should be specified for all considered cases (e.g., emergency coolant injection system model for small break loss of coolant accident may include a steam removal system for crash cool down)

B.3.1.1 Predicted reliability

Table B.2 Predicted Reliability

Failure Criteria	Predicted Reliability				Target
	Previous Years		Present	Future
	Year X – 2	Year X – 1	Current year (X)	Year X + 1

This section reports the future reliability predicted using current data and compares it to the value obtained for the present and previous years, as well as to the target. The reliability assessment needs to be re-evaluated annually using the latest relevant failure data.

This section should include the information regarding the assessment such as computer code, cut-off value, electronic file name, revision information and the report number, if available. If this information is presented in Report Appendix B.6, this section may refer to it. If the reliability assessment uses the supporting data different from the data presented in section B.5, the rationale should be described in this section.

B.3.1.2 Observed reliability

This section is intended to capture the trends in the systems important to safety reliability.

Table B.3 Observed Reliability of System

Failure Criteria	Unit no. _	Unit no. _	Unit no. _	Unit no. _	Previous Year Predicted Reliability	Target

Table B.4 Standby Generator Reliability Indices

A. Test Results
SG#	Start		Running
	Attempts	Failures	Hours	Failures
1
2
3
4
Total
B. Outage Statistics
SG#	Maintenance		Forced
	Occurrences	Hours	Occurrences	Hours
1
2
3
4
Total

Table B.5 Standby Generator Reliability Indices Trend

Year*

Running Failure Rate

Mean Time to Failure (MTTF)

Mean Time to Repair (MTTR)

Running Unavailability Index (U)

SG1

SG2

SG3

SG4

SG1

SG2

SG3

SG4

SG1

SG2

SG3

SG4

SG1

SG2

SG3

SG4

Year X – 5

Year X – 4

Year X – 3

Year X – 2

Year X – 1

* Current year = X.

Table B.6 Emergency Generator Reliability Indices

A. Test Results
EPG#	Start		Running
	Attempts	Failures	Hours	Failures
1
2
Total
B. Outage Statistics
EPG#	Maintenance		Forced
	Occurrences	Hours	Occurrences	Hours
1
2
Total

Table B.7 Emergency Generator Reliability Indices Trend

Year*	Running Failure Rate		Mean Time to Failure (MTTF)		Mean Time to Repair (MTTR)		Running Unavailability Index (U)
	EPG1	EPG2	EPG1	EPG2	EPG1	EPG2	EPG1	EPG2
Year X – 5
Year X – 4
Year X – 3
Year X – 2
Year X – 1

* Current year = X.

B.3.1.3 Incidents

RD-99.2 section 2.2.2.4 c): Identify and briefly describe, for each system important to safety, (i) incidents during the calendar year where the system was unavailable to perform its function and the dispositions for these incidents.

This section describes incidents (also known as major impairments) of the systems and the effect on the system reliability. The incident means any system fault reducing the effectiveness of the system such that it would fail to perform its safety, even if the system would still operate.

B.3.1.4 Minor impairments

RD-99.2 section 2.2.2.4 c): Identify and briefly describe, for each system important to safety, (ii) the occurrence, nature and duration of any impairment of the system over the calendar year and the effect of the impairment on the reliability of the system. Impairments caused by failure or unscheduled removal of equipment in support systems to the system important to safety should also be included.

This section describes any minor impairment of the systems important to safety and the effect on the system reliability. A minor impairment means a fault that reduces the redundancy of the system or increases the possibility of serious process failure or initiating event. In other words, it is a fault that cause degradation of a system but in which the system would still have met its design and performance specifications.

If multiple minor impairments overlapped, the overlapping effects should be assessed.

Table B.8 Minor Impairment and Effect on the System Reliability

Fault	Component/ Primary Event	Failure Mode	Failure Duration	Applicable Model	Instantaneous Reliability

Report Appendix B.4 should report the fault information for the nature of the impairment in detail.

Scheduled removal of service of equipment that are reflected in the reliability model do not need to be described.

This section also describes any situation that results in impairment of systems important to safety, but the failed or out-of-service component does not belong to the system under consideration. For instance, a loss of a Class II electrical bus supplying control power for an ECI component must increase the failure probability of ECI system. However, the intrinsic modelling approach may overlook the impact of the situation on the reliability of the system important to safety. The failed or out-of-service component may belong to a system not important to safety, even though the situation increases the failure probability the system important to safety.

If the event also causes impairments of other system(s) important to safety, these should be described in this section, shared dependent event.

B.3.1.5 Changes

RD-99.2 section 2.2.2.4 d): For each system important to safety, describe changes during the calendar year that had an impact on reliability due to (i) the design, (ii) an operating or maintenance practice, and (iii) models used to assess the reliability. The impact of the change on the reliability of the system should be evaluated and described.

This section reports changes to design or operating and maintenance practice affecting the reliability of the systems (e.g., the duty cycle of equipments may affect the test frequency of the equipments). Only permanent changes should be reported in this section, while the temporary changes are reported in section B.3.1.4.

The section should also discuss changes to the model and the effect on the reliability of the system.

B.3.1.6 Performance of surveillance activities

RD-99.2 section 2.2.2.4 e): Provide the following information:

1. summarize all missed surveillance activities required by the operation licence during the calendar year
2. list any scheduled activities to inspect, monitor, test or verify the reliability of a system important to safety of the nuclear power plant that were not completed on schedule during the calendar year
3. indicate the impact on system reliability of the scheduled activities not being completed on schedule

This section reports any missed and postponed scheduled activities credited in reliability assessment. The reasons for missing or postponing the scheduled activities and the impact on the system reliability should be discussed. These activities include required tests, predefines (call-ups) and operator routines.

B.3.1.7 Correction of previous reports

The section describes any corrections in previous reports; the correction may be a mistake or new findings over the reporting calendar year. The reason of correction should be clearly provided.

For instance, when a design defect of a component is found over the calendar year and the design defect has been dormant since the installation or modification of the component, the previous reliability indices should be modified. If an impairment is revealed by a test and the test period is long enough to affect the previous reliability indices, the previous reliability indices should also be reported.

B.4. Plant Performance

B.4.1 Initiating events

RD-99.2 section 2.2.2.4 f): Describe initiating events that occurred during the calendar year at the NPP. Provide the current probability assigned to each initiating event. Indicate the reason for each significant change.

The initiating event includes any following events:

1. serious process failures or potential serious process failures
2. unplanned shutdown system operation
3. unplanned power transient

The frequency of the initiating event should be re-assessed and a comparison with the initiating event frequency used in the risk analysis should be provided if a plant probabilistic safety assessment (PSA) model is available. The description of the event should indicate if it affected the plant mitigating systems capability.

The impact on the plant safety goals of the initiating event should be provided. The analysis should include any equipment malfunction that occurred during the response to the initiating event. Any of these mitigating system failures should be clearly identified.

B.4.2 Abnormal plant configuration(s)

RD-99.2 section 2.2.2.4 g): Describe the occurrence, nature, duration, and individual and cumulative impact on safety goals of:

1. abnormal plant configurations during the calendar year, either planned or unplanned, resulting from either degraded conditions or discovery of safety analysis issues (an abnormal plant configuration occurs when the plant is operated outside the normally approved operating procedures)
2. events and incidents that had the potential to significantly increase the predicted risk of core damage or release

“Abnormal plant configuration” refers to state of the plant being outside approved operating procedures either due to planned or unplanned activities or equipment failure (e.g., operating with less redundancy in safety or safety-related systems, operating with degraded conditions).

The description should include the date and duration of the occurrence, the reason for the occurrence, the mitigating actions taken (if applicable) and the impact on safety goals. If Level 1 and Level 2 PSAs are available, provide the impact on both core damage frequency and large release frequency as a minimum.

B.5. Supporting Data

RD-99.2 section 2.2.2.4 h): Provide the data that supports the licensee’s assessments over the calendar year of the reliability of the systems important to the safety of the NPP. Relevant data shall include:

1. assumed rates of failure of system components
2. significant trends in component failure rate data
3. input data regarding human reliability
4. list of scheduled activities to inspect, monitor, test or verify the system’s reliability
5. data regarding the impairment (failure, incipient failure or degraded ability) of one or more system components as a direct result of a shared or common cause

B.5.1 Component failure data

B.5.1.1 Failure rates

This section describes major updates of the component failure rate data and the specific changes in failure rates from the failure rate database experienced over the calendar year. The failure database used for reliability model should be presented in Report Appendix B.3. This section should describe any significant failure rate change or significant trend. The definition of significant failure rate change or significant trend should be clearly described.

B.5.1.2 Failure modes

This section describes any failure mode that occurred during the calendar year and is not modelled in the system fault trees, as well as its impact on the system reliability.

B.5.2 Human reliability data

This section reports human actions that could affect the reliability of the system important to safety. Some human actions that directly affect the reliability of the systems important to safety may be included in section B.3.1.3, but a comparison to the human errors considered in the reliability model should be performed.

B.5.3 Dependent failures

B.5.3.1 Shared cause

This section reports events resulting in inter-system shared dependent impairments that cause unavailability of multiple components belonging to different systems and can be included in the reliability models explicitly. For instance, a failure of an instrument air manifold that supplies air to multiple components in more than one independent system would result in an inter-system shared cause dependent failure and should be reported in this section.

Intra-system dependent impairments having impact on one system only are described in the section of individual system (e.g., if the manifold supplies air to multiple components but in only one system, the failure would be identified as an intra-system shared cause dependent failure and might be described in the section of individual system).

B.5.3.2 Common cause

This section reports events that may result in inter/intra-system dependent failure or degradation, but reliability model cannot include the failure cause explicitly. Even though section B.3.2 reports human induced common cause failures, this section should identify these events as common cause failures.

B.5.4 List of scheduled maintenance and surveillance activities

The list of scheduled activities to inspect, monitor, test or verify the system’s reliability should be included here or in an appendix.

B.5.5 Other plant specific data

This section includes other plant specific data not included in sections B.3.1 and B.3.2. The reliability statistics of special equipment can be included here.

B.6. Report Appendices

B.6.1 Report Appendix B.1 Definitions

Provide in this appendix definitions for technical or licensee terms used in this report.

B.6.2 Report Appendix B.2 List of Acronyms and Abbreviations

Provide in this appendix acronyms and abbreviations used in the report.

B.6.3 Report Appendix B.3 Component Failure Data

See section B.5.1.1 for details regarding the information to be placed in this appendix.

B.6.4 Report Appendix B.4 Fault Records

See section B.3.1.4 for details regarding the information to be placed in this appendix.

B.6.5 Report Appendix B.5 Test Program Summary

This appendix lists the scheduled activities included in the reliability models and describes the changes that are not included in the individual system sections, including routine tests, predefines, operator routines and main control room (MCR) panel checks.

B.6.6 Report Appendix B.6 Models of the Systems Important to Safety

The information presented in this appendix may vary according to the reliability program being implemented by individual NPP (e.g., one plant may apply a single cut-off to all systems, while another plant may apply different cut-off values to the systems). Therefore, the example below defines the content, not the format.

Table B.9 Reliability Models Information

System	Reliability Model				Report
	File Name	Cut-off	Rev.	Date	Report #	Date
Computer Code*	CAFTA version 5.01a (September 2003)			CAFTA Cutset Generator Version XX (September 2000)

* Common for all system models.

Tables

Table 1: Documents Containing Reporting Requirements

No.	Report Title	Time to Submit Preliminary Report or Notify the CNSC	Source Document
1.	Periodic inspection of CANDU NPP components	For items inspected during a scheduled maintenance outage, file periodic inspection report within 90 days of completion of the outage. For items not inspected during a scheduled maintenance outage, file periodic inspection report within 90 days of completion of the planned inspection period.	CSA N285.4
2.	Periodic inspection of CANDU NPP containment components	For items inspected during a scheduled maintenance outage, file periodic inspection report within 90 days of completion of the outage. For items not inspected during a scheduled maintenance outage, file periodic inspection report within 90 days of completion of the planned inspection period.	CSA N285.5
3.	In-service examination and testing requirements for concrete containment structures or CANDU NPPs	As specified in the standard.	CSA N287.7
4.	Report on the review of the Integrated Ageing Management Program effectiveness and improvements	This report is submitted annually to the CNSC for review and assessment.	RD-334

Table 2: Categorization of Reports by CNSC Safety and Control Areas

Functional Area	Safety & Control Area	Report #	Report Title
Management	Management System	2.2.1.1	Operations report*
	Human Performance Management	2.2.1.1 2.2.2.6 2.2.2.7	Operations report* Report on conducted training, examination, and testing of personnel for assignment to positions requiring CNSC certification Planning report on scheduled training, examination, and testing of personnel for assignment to positions requiring CNSC certification
	Operating Performance	2.2.1.1	Operations report*
Facility and Equipment	Safety Analysis	2.2.1.2 2.2.2.4 2.2.2.5 2.2.3.2 2.2.2.8	Performance indicator report* Report of the risk and reliability of the NPP Report on fuel performance Updates to facility descriptions and final safety analysis report Report on reactivity management
	Physical Design	2.2.2.5 2.2.1.4	Report on fuel performance Report on the degradation of NPP Pressure Boundaries
	Fitness for Service	2.2.1.2 2.2.1.4	Performance indicator report* Report on the degradation of NPP Pressure Boundaries
Core Control Processes	Radiation Protection	2.2.1.1 2.2.1.2 2.2.2.2	Operations report* Performance indicator report* Report of occupational radiation protection performance
	Conventional Health and Safety	2.2.1.2	Performance indicator report*
	Environmental Protection	2.2.2.1	Report of environmental protection performance
	Emergency Management and Fire Protection.	2.2.1.1 2.2.1.2	Operations report* Performance indicator report*
	Waste Management	Reports required by the in-plant waste management program and by CSA N294: Decommissioning of Facilities Containing Nuclear Substances when the standard is referenced in the operating licence.
	Security	2.2.1.3 2.2.3.3	Security report Station security report
	Safeguards	2.2.1.1 2.2.3.1	Operations report* Accounting and reporting of uranium, plutonium and thorium
	Packaging and Transport	---	---

Notes:
1. * The indicated reports contain information from a multiple number of safety and control areas as shown.
2. R&D activities can involve all 14 safety and control areas, therefore report number 2.2.2.3, “Report on the progress of research and development (R&D) activities” can be submitted for each area.

Table 3: Categorization of Reports by CNSC Safety and Control Areas

No.	Safety and Control Area Performance Indicator	Safety and Control Area Number (see Table 4)
1.	Accident Severity Rate, Accident Frequency Rate	1–5, 8
2.	Chemistry Index	4–6
3.	Chemistry Compliance Index
4.	Change Control Index	1–6
5.	Radiological Emergency Performance Index	10
6.	Emergency Response Organization (ERO) Drill Participation Index
7.	Emergency Response Resources Completion Index
8.	Non-Compliance Index	1–7
9.	Number of Pressure Boundary Degradations	5, 6
10.	Preventive Maintenance Completion Ratio	4–6
11.	Radiation Occurrence Index	7
12.	NPP Radiation Dose
13.	Number of Missed Mandatory Safety System Tests	1–3, 6
14.	Number of Unplanned Transients	3, 5, 6
15.	Unplanned Capability Loss Factor
16.	Corrective Maintenance Backlog	5, 6
17.	Elective Maintenance Backlog
18.	Preventive Maintenance Backlog
19.	Average Work Week Schedule Adherence

Safety and Control Areas not included in the Performance Indicators given in table 3 are:

• 9. Environmental protection
• 11. Waste management
• 12. Security
• 13. Safeguards
• 14. Packaging and transport

Table 4: Safety and Control Areas (SCAs)

No.	Safety and Control Area
1.	Management system
2.	Human performance management
3.	Operating performance
4.	Safety analysis
5.	Physical design
6.	Fitness for service
7.	Radiation protection
8.	Conventional health and safety
9.	Environmental protection
10.	Emergency management and fire protection
11.	Waste management
12.	Security
13.	Safeguards
14.	Packaging and transport

Glossary

abnormal plant configuration

Plant configurations not covered by the normal operating policies and principles or the normal operating procedures.

impairment 

A failure such that the safety-related system would operate with reduced redundancy or margin of safety, or would fail to meet its design intent. Level 1 impairment is used to describe a system state that is impaired to the extent that it would provide inadequate protection. Level 2 impairment is used to describe a system state that is impaired to the extent that it would provide some, but not complete, protection for a worst-case process failure. Level 3 impairment is used to describe a system state where the level of redundancy or margin of safety is reduced, but the system is still fully capable of meeting its design intent.

safety-related system

Those systems, and the components and structures thereof, which, by virtue of failure to perform in accordance with the design intent, have the potential to impact the radiological safety of the public or plant personnel from the operation of the nuclear power plant. Those systems, and the components and structures thereof, are associated with:

1. the regulation (including controlled startup and shutdown) and cooling of the reactor core under normal conditions (including all normal operating and shutdown conditions)
2. the regulation, shutdown and cooling of the reactor core under anticipated transient conditions and accident conditions, and the maintenance of the reactor core in a safe shutdown state for an extended period following such conditions
3. limiting the release of radioactive material and the exposure of plant personnel and/or the public to meet the criteria established by the licensing authority with respect to radiation exposure during and following normal, anticipated transient conditions and accident conditions.

systems important to safety

Structures, systems and components of the NPP associated with the initiation, prevention, detection or mitigation of any failure sequence that have the most significant impact in reducing the possibility of damage to fuel, associated release of radionuclide, or both.

Additional Information

The following legislation and regulations are relevant to this guidance document:

• Nuclear Safety and Control Act, S.C., 1997, c.9
• General Nuclear Safety and Control Regulations, SOR/2000-202
• Radiation Protection Regulations, SOR/2000-203
• Class I Nuclear Facilities Regulations, SOR/2000-204
• Nuclear Substances and Radiation Devices Regulations, SOR/2000-207
• Packaging and Transport of Nuclear Substances Regulations, SOR/2000-208
• Nuclear Security Regulations, SOR/2000-209
• Interpretation Act, R.S.C. 1985, c. I-21

The following documents provide additional information pertaining to reporting requirements for operating NPPs:

• RD-99.2, Reporting Requirements for Operating Nuclear Power Plants: Compliance Monitoring, Canadian Nuclear Safety Commission, 2011 (TBC)
• P-325, Nuclear Emergency Management, Canadian Nuclear Safety Commission, 2006
• S-294, Probabilistic Safety Assessment (PSA) for Nuclear Power Plants, Canadian Nuclear Safety Commission, 2005
• S-98 Rev-1, Reliability Programs for Nuclear Power Plants, Canadian Nuclear Safety Commission, 2005
• S-210, Maintenance Programs for Nuclear Power Plants, Canadian Nuclear Safety Commission, 2007
• CMD 03-M68, Criteria for Preparing Significant Development Reports (SDR), Canadian Nuclear Safety Commission, November 2003
• CMD 05-H32, Licence Amendments to Strengthen Regulatory Controls on Sealed Sources, Canadian Nuclear Safety Commission, August 2005
• RD-204, Certification of Persons Working at Nuclear Power Plants, Canadian Nuclear Safety Commission, 2008
• RD-334, Aging Management for Nuclear Power Plants, Canadian Nuclear Safety Commission, (TBC)
• G-225, Emergency Planning at Class I Nuclear Facilities and Uranium Mines and Mills, Canadian Nuclear Safety Commission, 2001
• G-274, Security Programs for Category I or II Nuclear Material or Certain Nuclear Facilities, Canadian Nuclear Safety Commission, 2003
• G-208, Transportation Security Plans for Category I, II or III Nuclear Materials, Canadian Nuclear Safety Commission, 2003
• S-296, Environmental Protection Policies, Programs and Procedures at Class I Nuclear Facilities and Uranium Mines and Mills, Canadian Nuclear Safety Commission, 2006
• RD-336, Accounting and Reporting of Nuclear Material, Canadian Nuclear Safety Commission, 2010
• CSA N285.0/N285.6 series, General Requirements for Pressure Retaining Systems and Components in CANDU Nuclear Power Plants/Material Standards for Reactor Components for CANDU Nuclear Power Plants, Canadian Standards Association
• Periodic Inspection of CANDU Nuclear Power Plant Components, CSA N285.4, Canadian Standards Association
• CSA N285.5, Periodic Inspection of CANDU Nuclear Power Plant Containment Components, Canadian Standards Association
• CAN/CSA-N286.0, Overall Quality Assurance Program Requirements for Nuclear Power Plants, Canadian Standards Association
• CSA N286, Management System Requirements for Nuclear Power Plants, Canadian Standards Association
• CAN/CSA N289.5, Seismic Instrumentation Requirements for CANDU Nuclear Power Plants, Canadian Standards Association
• ISO 17025, Quality Manual Template, International Organization for Standardization, 2005
• ASME Boiler & Pressure Vessel Code, Section III, Division 1, Rules for Construction of Nuclear Power Plant Components - Subsection NB – Class 1 Components, American Society of Mechanical Engineers
• TS-R-1, Regulations for the Safe Transport of Radioactive Material, International Atomic Energy Agency, 2009