Physics Simulation and Configuration Control in support of the Operation and Ageing Management of SAFARI-1

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1 Physics Simulation and Configuration Control in support of the Operation and Ageing Management of SAFARI-1 J.I.C. Vermaak SAFARI-1 Research Reactor NECSA, Church Street, PO Box 528 Pretoria South Africa ABSTRACT The SAFARI-1 reactor is an ageing Material Testing Reactor (MTR) in operation since It originally operated exclusively as a nuclear research facility but as of the 1990 s its commercial utilization increased significantly to the extent that its continued operation represents a key component of the future commercial endeavors of the South African Nuclear Energy Corporation (NECSA). Consequently, the facility is required to both; operate within the modern environment, and to comply with modern commercial nuclear facility standards, for which the key components are safety and reliability but also includes market driven utilization (competitiveness), compliance to modern requirements, and ageing management. This paper presents some aspects of the use of simulation and analysis tools and processes that are employed in new utilization initiatives and ageing management projects. Throughout the paper there is a strong focus on the use of proper configuration management from design through to simulation and analysis. Systems, Structures and Components (SSCs) due to be changed through new utilization initiatives or Ageing Management projects often have to be evaluated first, in order to determine their original design intent. Due to the aged nature of the reactor facility, design reconstitutions are frequently required before such an evaluation may be performed. This paper touches on the importance of a proper configuration management system in facilitating this process and to provide a sound designinformation baseline from which to proceed with new utilization initiatives and ageing management projects. To illustrate this point, a brief overview of SAFARI-1 s configuration control implementation is presented with a focus on the information structure design and how information is fed into the configuration management system. Another important aspect to consider when contemplating changes to SSCs, particularly in the context of the facility as a whole, is the interaction between the SSCs to be changed and other systems in the facility. In order to evaluate the impact of such changes on the facility and to determine if the proposed changes will deliver the desired results within the constraints of safety and reliability, proper physics simulation tools and methodologies have to be employed. This has to be done through seamless integration with the configuration management system and requires that the physics simulation tools, processes and results are also under configuration management. Physical processes involved in the operation of the plant, tools used to simulate the underlying physics and SAFARI-1 s strategy for the implementation of control measures on the acquisition, development, maintenance and Verification and Validation of simulation tools and models are presented. The application of the simulation tools and the processes related to simulations and analyses are also touched on with the inclusion of some examples. 1 INTRODUCTION Over the considerable period of operation of the SAFARI-1 reactor (since 1965) the facility played witness to a number of leaps in technology. Periodically, some technologies created such a demand in the market that it over-shadowed the demand of that already implemented, consequently the associated technical equipment became obsolete and unavailable. This is, however, a natural phenomenon that occurs in any long-operating plant. The critical issue

2 surrounding the change of a technology is not necessarily that the newer technology replaces the older one but that, when it happens, either the newer implementation or the older implementation is not sufficiently documented. This issue is very prevalent at SAFARI-1 and is gaining importance in light of the modern requirements placed on the facility as well as the implementation of Ageing Management. In 1998, the facility adopted an Integrated Management System (IMS) based on the requirements of the ISO 9001 standard, which marked the start of a structured manner in which documents or sources of information were controlled. By the end of 2010, the focus on Ageing Management, as well as other requirements, increased the demand for information structures capable of integrating the flow of information involving: Design reconstitutions (historical information), Current baseline information, Information used for evaluation of the system (i.e. calculations), Information being generated by developments (i.e. projects and operation), Maintenance information and Information relating to requirements management. In 2011, a strategy for Configuration Management (CM) was developed and is currently being implemented. The SAFARI-1 Configured Information Management Structure (CIMS) is discussed in section 2. An important part of the CM information structure is information used for evaluation of the system, more specifically it is information translated from detailed information for use in the plant Evaluation Model, i.e. the Evaluation Model is the collection of Calculation Models, implemented in various simulation tools, and constitute the theoretical evaluation of the plant system. Up until the end of 2011, SAFARI-1 supplied as input to service providers (performing calculations) full-detail information, i.e. engineering drawings, of the plant or the associated physical process to be modeled. This approach inherently resulted in a translation of detailed information to sub-inputs, which were then used by the associated service provider, and together with the lack of strict requirements on the output of calculations, the risks on SAFARI-1 s liability for the correctness of plant evaluations increased. In order to address this risk, a Calculation Process was implemented and a single control point assigned. The Calculation Process firstly required the break-down of the plant Evaluation Model into a collection of Calculation Models. Calculation Models are analytical representations or quantifications of a real system, used to predict or assess the behavior of the real system under specified conditions. Most often, a Calculation Model is associated with a specific Software Product, which is then also contained within the Calculation Model. Within a Calculation Model, different software codes or programs can be operated within the Software Product, each with an associated model of the part of the system it is intended to simulate or assess. Therefore, a Calculation Model may comprise a Software Product and a collection of System Models. An illustrative decomposition of an Evaluation Model is shown in figure 1 below together with the associated documentation. The primary physics simulation tools, used within the SAFARI-1 evaluation model, are detailed in section 3 as well as representations of their accuracies. These tools also have dedicated configuration control structures within the SAFARI-1 CIMS. Software products have both control items for the product itself and V&V of the product (qualification). Hardware configurations that successfully completed hardware V&V tests (as identified in the software V&V) as well as authorized users of the software, are controlled by means of configuration control registers.

3 Calculations and Calculation Models also have items which can be linked to other CM-items, specifically to V&V items, Hardware items, Software items and Data packs. Figure 1 Breakdown of the definition of the SAFARI-1 evaluation model. Items of the same colour are of a specific Configuration Management item type. 2 SAFARI-1 CONFIGURED INFORMATION MANAGEMENT STRUCTURE The initiation of a need for the generation, change or removal of information from the facility's information structure can originate from various sources. Fundamentally, the facility needs to fulfill its operational functions which may be elicited from the basic mandate of the facility, or by certain requirements. The complete CIMS structure is shown in figure 2 below. The functions are controlled within the Functional Breakdown Structure (FBS) and interfaces directly with the Requirements Management Structure (RMS). The RMS is used to control all of the imposed requirements on the facility which can range from economical requirements to regulatory requirements. Together, the FBS and the RMS can define a need, and once it is decided to act on this need, one or more projects are initiated for which the information is controlled in the Project Information Structure. Projects ultimately develop work activities which are then arranged in a Work Breakdown Structure (WBS). The operation and maintenance of the facility can, however, also initiate work activities which are controlled within the WBS. See figure 2 below. During the execution of work activities, information is generated which needs to be configured in a structure than can facilitate the different levels of Configuration Management activities. These are: planning (partly contained in the WBS), identification (making information reference-able), change control (version control), status accounting and auditing. The initial stages of per-item -generation, -checking and -approvals are facilitated within the Deliverable Management Plan Structure (DMP Structure) and allows items of information to be developed before being incorporated into new revisions of the System Breakdown Structure (SBS). The SBS, and its revisions, form the collective structure of system information at a specific moment in time (i.e. baselines) and is ordinarily checked and approved as a whole. The subordinate structures of the SBS include:

4 The Configured Operational Information Management Structure (OPS-CIMS) which contains all the configured information as part of the operation of the facility, The Maintenance, Repair and Overhaul Structure (MRO Structure) which contains all of the configured information relating to maintenance activities, and The Hardware Breakdown Structure (HBS) which contains all the technical or engineering information of the facilities SSCs. Figure 2 Schematic of the Configuration Management Information Structure showing the different interactions and links. A breakdown of the initial levels of the SAFARI-1 SBS is shown in figure 3 below.

5 Figure 3 The first levels of the SAFARI-1 System Breakdown Structure (SBS) as contained in the Product Lifecycle Management software Siemens-Teamcenter. 3 PHYSICS SIMULATION TOOLS First and foremost, the SAFARI-1 reactor core requires the simulation of nuclear physics, specifically neutron and photon transport, criticality, heat generation and neutron activation. For this purpose, SAFARI-1 utilizes OSCAR-3[1] and OSCAR-4[5], MCNP5/MCNPX[6], FISPACT[3] and ORIGEN-S[4]. FISPACT and ORIGEN-S is used primarily for neutron activation calculations and are normally supplied with values calculated by MCNP5. As a core-follow tool, which simulates the flux and power distribution in a 3D nodalization with a size of approximately 8 cm, SAFARI-1 utilizes OSCAR-3. This Software Product was developed by NECSA s own Research & Development group and contains the nodal diffusion solver code MGRAC[2]. The MGRAC code allows the estimation of important nuclear safety parameters such as Beginning Of Cycle (BOC) excess reactivity, shutdownmargin and Power Peaking Factor (PPF). Other parameters obtained from the software include prediction of cycle length and fissile inventory verification, all of which forms part of the cycle-to-cycle regulatory core approvals. In 2010 NECSA s Research & Development group released OSCAR-4, which introduced many improvements. The software is currently still undergoing Verification & Validation. As part of a preliminary calculation in order to develop specifications for replacement gammadetectors (for SAFARI-1 s gamma safety channels), the OSCAR-4 inventory of fuel and control assemblies were transferred to a MCNP5 model of the SAFARI-1 core. In order to validate the MCNP5 model s representation of the core, axial thermal neutron flux profiles were compared to experimental data obtained by irradiating copper wires over the axial length of the fuel assemblies and control assembly fuel followers. The results for fuel assemblies are shown in figure 4 while results for control assemblies are shown in figure 5.

6 Figure 4 Comparison of axial flux profiles, calculated with MCNP5 and OSCAR4, with the inventory as calculated by OSCAR-4, to measurements from activated copper wires. The results are for a fuel assembly showing the least amount of deviation (left) and a fuel assembly showing the most deviation (right). Figure 5 Comparison of axial flux profiles, calculated with MCNP5 only, with the inventory as calculated by OSCAR-4, to measurements from activated copper wires. The results are for a control assembly fuel follower exhibiting the greatest deviation and exclude OSCAR-4 values due to differences in nodalization. The model allowed for the calculation of ex-core neutron and photon fluxes as shown in figure 6 below. The 3-dimensional nature of such results allowed for the development of surfaces that could be imported into a solid modelling computer package which added considerable value to the design process. The OSCAR-4 and MCNP Calculation Models of the SAFARI-1 core are configured within the System Breakdown Structure item (discussed in the following sections) designated Reactor System. Including these models in the Reactor System rather than in the Reactor Core System is done because of its interfaces with the Reactor Cooling System which contains the thermal-hydraulic Calculation Model.

7 Figure 6 Three-dimensional representation of photon-flux iso-surfaces at ex-core regions of the SAFARI-1 core and outside the vessel (160 cm diameter). The values were calculated with MCNP5 using importance based variance reduction techniques. A critical safety consideration for SAFARI-1 is the ability to cool core internal structures and for this RELAP5/SCDAP Mod 3.4 [7] is used for thermal-hydraulic analysis. As part of the design of Isotope Production Rigs (IPRs), SAFARI-1 utilizes detailed models for modelling in RELAP and validates these models with an experimental flow-loop simulating the characteristics of a single core position. The procedure for such analyses is to first include a first order model in a detailed SAFARI-1 core model, for which the core pressure differential (ΔP) can be determined within 5%, and then to modulate the flow-loop s coolant flow to match the calculated ΔP (i.e. both for the calculation and for the experiment). For this configuration, the relevant cooling characteristics are determined (i.e. pressure, velocity, etc.) after which they are used to determine whether or not the relevant Operating Technical Specifications limits (OTS-limits) are adhered to. For the conceptual design of an IPR, involving a motorized housing with a fuel follower - similar to a control assembly which reduced the effective reactivity worth, a detailed flowloop experiment was conducted to validated the associated RELAP models. A schematic of the rig arrangement is shown in figure 7 below. A comparison of rig coolant velocities, between RELAP calculated values and flow-loop measured values, is shown in figure 8. The differences were approximately 7% for the entire range of core differential pressures.

8 Figure 7 Schematic of a conceptual Isotope Production Rig (IPR) design in which a fuel follower is utilized as part of a motorized assembly with the purpose of reducing the effective reactivity worth of the rig. Figure 8 Comparison of RELAP5 calculated coolant velocities, in Isotope Production Rigs of significant complexity, to values obtained by means of an experimental flow-loop representing a single core position. The RELAP Calculation Model is configured within the SBS under the Reactor System structure. Commutative to the neutronic models (OSCAR-4 and MCNP), the model is not contained under the Reactor Cooling System structure because it directly interfaces with the neutronic models. For structural and thermal analyses, SAFARI-1 uses ANSYS workbench thermal-structural which allows for the calculation of structural stresses during the analysis of designs. Also, the software allows for vibrational frequency analysis which was of particular importance for evaluating the conditions of the SAFARI-1 core gridplate, a component which locates the

9 core assemblies, for eventual replacement. Due to the fact that structural stresses are difficult to measure, no validation information for the use of ANSYS is presented. A deformation contour of the SAFARI-1 grid plate is shown in figure 9 below. Figure 9 Visualization of the deformation of the SAFARI-1 core grid plate under the static load of the fuel as well as the flow induced loads. The replacement of the SAFARI-1 core grid plate is an ageing management project and is currently being executed. In figure 10 below, a Configuration Management structure is shown for a neutronics calculation relating to the ageing management of SAFARI-1 s gamma safety channels. The structure represents the Calculation as it is contained in the SAFARI-1 Product Lifecycle Management Software Siemens-Teamcenter.

10 Figure 10 The Configuration Management structure of all the items of a calculation. This example calculation is for an ageing management project addressing the replacement of the gamma safety channels. 4 CONCLUSION Since the Calculation Models, constituting the SAFARI-1 Evaluation Model, are configured within the SAFARI-1 System Breakdown Structure, a great degree of traceability has been achieved. This is because all of the source information for the Calculation Models is derived from SSC-information configured within the SBS. Therefore, whenever changes are made to the system, or new information is available, the interface is immediately identified and the affected Calculation Model can be updated to accurately represent the changed SSC. Ultimately, the controller of calculations can be made responsible for: Deriving inputs to Calculation Models from the SBS, Packaging of inputs, together with relevant requirements, for dispatch to service providers, Collecting and ensuring the Verification & Validation (V&V), of Calculation Models, Collecting and ensuring the V&V, of Calculations, Configuring Calculation Information within the Configuration Management System, and Ensuring that Calculation Models are up to date with the latest declared SBS baseline.

11 In comparison to the procedures that were available before the implementation of the SAFARI-1 Calculation Process, the new procedures enforce greater traceability. Particularly, the following explicit changes were implemented: Software now has to undergo verification in order to ensure that its mathematical models are correctly implemented (i.e. without fudge factors) after which these models have to be validated for use with a specific physical process. It was demonstrated on numerous occasions, that software products can produce different results on different hardware configurations and therefore software products now have to maintain a list of authorized and tested hardware and software platforms. Calculations now have to specify the exact hardware and software configuration used to generate the results, which was particularly important for traceability of results. System models have to be verified to correctly represent the associated SSC. Records have to be kept of the verifications, and identified outstanding issues need to be addressed appropriately. The system models need to have a complete and comprehensive descriptions which include description documents, acceptable quality of in-code annotations (i.e. commenting of input files), appropriate user manuals and procedures for change control. Calculation Models are required to have approved V&V documentation as well as independent review of both the Calculation Model development and the V&V process. Calculations are required to reference only approved Calculation Models. In the broader view of nuclear quality requirements, the SAFARI-1 Calculation Process introduces no new concepts but merely enforces a culture of safety, quality, accountability and confidence. 5 REFERENCES [1] E.Z. Muller, G. Ball, W.R. Joubert, H.C. Schutte, C.C. Stoker, F. Reitsma. Development of a core follow calculation system for research reactors. Paper presented at the 9th Pacific basin nuclear conference, Sydney, Australia. May [2] E.Z. Muller, D. Tomasevic. MGRAC Versatile Core Simulator for Research Reactor Core Tracking and Operation Planning. Paper presented at PHYSOR [3] R.A. Forrest and M.R. Gilbert, FISPACT-2005: User manual, UKAEA Report, UKAEA FUS 514, [4] O.W. Hermann, R.M. Westfall. ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms, NUREG /CR-0200-Vol. 2, Revision 4. (1990). [5] G. Stander et.al. OSCAR-4 Code System Application to the SAFARI-1 Reactor. Paper presented at International Conference on Reactor physics, Nuclear Power: A Sustainable Resource, PHYSOR2008. Interlaken, Switzerland [6] MCNP5: MCNP. X-5 Monte Carlo Team, A General Monte Carlo N-Particle Transport Code, Version 5, LA-UR Los Alamos National Laboratory. [7] NUREG/CR-5535/Rev 1-Vol I. "RELAP5/MOD3.3Beta Code Manual Volume 1: Code Structure, System Models, and Solution Methods". Information Systems Laboratories. May 2001.