Nuclear reactor-design safety is one of the most important aspects of the discipline, addressing the safety of a nuclear power plant (NPP), from concept to detailed design, construction, commissioning, operation and maintenance, to waste management, decommissioning, temporary and permanent storage.
Nuclear safety of operating plants and new builds is the core business of the Nuclear Energy Division at ASCOMP. Our services currently cover:
Deterministic Safety Analysis (DSA) Expand
In the framework of the DSA our experts perform the calculations using both conservative and best-estimate approaches.
The analysis covers operational transients, design-basis accidents (DBA), beyond-design-basis accidents (BDBA), and severe accidents (SA). Detailed models of the NPP system are developed for codes such as TRACE, SCDAPRELAP and MELCOR addressing client needs and requirements.
Advanced CFD computations are also performed when small- or medium-scale analyses are required for studying special phenomena. See also Advanced Computational Thermal hydraulics. Furthermore supporting analysis may be made for Probabilistic Safety Assessment (PSA).
Probabilistic Safety Analysis Level 1, 2, and 3. Expand
In addition to DSA, probabilistic safety assessment may be performed by developing a realistic model of the system/s and or its components.
The PSA could be made at Level 1 for the evaluation of core damage frequencies, or at Level 2 for determination of the plant damage states, release paths and finally source terms, which are required for a Level 3 PSA. The Level 3 PSA model includes a detailed site model covering topography, meteorology, superficial and underground water, land use and agricultural product consumption and production. Source terms (outcomes of Level-2 PSA) and site model are used as basis for dispersion analysis, dose assessment, contamination level and decontamination strategy, and finally evaluation of the different countermeasures (such as evacuation strategies or use of Iodine prophylaxis). Level-3 PSA could be used for development or review of a country emergency planning.
Safety Analysis Reports, Preliminary and Final (SAR, PSAR, FSAR) Expand
The safety of a nuclear power plant needs to be periodically reviewed and proved. Any changes in design need to be examined and approved from the safety point of view. In addition to tests and inspections, detailed analyses using both deterministic and probabilistic approaches are required for approval of the changes, which are then reported in the NPP Safety Analysis Report (SAR) or in the case of a new buildits PSAR and/or FSAR.
Preparation, review, or update of the safety reports and the periodic safety review are supported by our Services.
Plant Specific Emergency Procedures (AOPs, EOPs, SAMGs, EDMGs) and Plant Emergency Plan
In addition to the services on SARs, Ascomp and partners develop or review (vendor procedures) all sets of plant specific procedures used in the Main Control Room (MCR), the Emergency Control Room (ECR) and the Technical Support Centre (TSC) during accidental DBA, BDBA and SA conditions: Expand
- Abnormal Operating Procedures (AOP)
- Emergency operating procedures (EOP)
- Severe Accident Management Guidelines (SAMG)
- Extreme damage management guidelines (EDMG)
Other supporting documents including AOPs/SAMGs setpoint studies and instrument uncertainty evaluation during “harsh” containment conditions based on state of art EPRI standards could also prepared by experts.
Furthermore, Ascomp is also ready to prepare and provide verification and validation, training course and exercises related to the mentioned procedures in DSA with detailed model developed for RELAPSCDAP or MELCOR code.
Reactor physics, In-Core and Ex-Core Fuel Management
Experts at ASCOMP have long experience in generation of the macroscopic cross sections starting from microscopic cross sections made available by the NEA or IAEA databases. The pin-wise or assembly-wise macroscopic cross sections are then used for generation of a detailed core model, which is the basis for the core parametric studies. Expand
Furthermore multi-parameter macroscopic cross-section databases may be generated for in-core and ex-core fuel management, loading optimization and core isotope inventory evaluations at an accident initiation (so called core reference inventory used for the PSA) or for fuel-pool inventories. Finally core calculations could be used for the evaluation of the radiations sources needed for radiation shielding and fluence analysis.
Radiation Shielding and Dose Assessment
ALARA principles and regulation puts limits on the exposure level of workers in a NPP or a nuclear facility. Fulfillment of such requirements need shielding calculations and dose evaluations in the region surrounding a component (source). Expand
Shielding calculations and the dose assessments are part of services provided by ASCOMP. The outcomes of such calculations are multi-particle radiation shielding, heating, and damage. In the framework of dose assessment, dose layouts, organ and effective doses can be estimated.
Thermal Performance Assessment
Evaluation of the thermal performance of secondary circuit (balance of plant, BoP) and its efficiency assessment for existing nuclear power plants is part of services provided by ASCOMP. Expand
The outcomes of such assessments are useful for increasing the plant electrical production and resolving technical problems on BoP components in accordance with state-of-the-art thermal performance methods and basing on EPRI standards.
Ascomp could also identify specific plant BoP issues and propose a plant thermal performance program and procedure in accordance with EPRI standards.
Design Modification Process
The reasons for plant modifications could be of different nature such as maintaining or strengthening existing safety provisions, recovering from plant faults, improving the thermal performance or increasing the power rating, increasing the maintainability of the plant, reducing the radiation exposure of personnel or reducing the costs of plant maintenance; and extending the design life of the plant [IAEA No. NS-G-2.3]. Whatever is the basis for the plant modifications, it needs to be reviewed thoroughly and performed according to well-established procedures. With this regards, ASCOMP and partners services cover: Expand
- Review of existing plant modification process and associated procedures in accordance with NRC 10CFR50.59 (Safety Evaluation) process and nuclear practice (NEI);
- Review of Design Modification Packages (DMPs) before implementation by independent DSA and PSA assessment and prepare the Preliminary and Final Independent Evaluation Report (PIER or/and FIER);
- Provide trainings and exercises on review and approval of DMPs.