The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are used in Tractebel to perform source term assessments and plant Thermal Hydraulic (TH) analyses in case of Severe Accident (SA) for the Belgian Nuclear Power Plants (NPPs).

These system codes were developed through intense and extensive validation activities which were mainly based on the verification of the capability of implemented physical models to reproduce the experimental results of Separate Effect Tests (SETs) and Coupled Effect Tests (CETs).

However, the SETs, and in less extension the CETs, do not take into account interactions and possible synergistic or antagonistic effects which can arise among the different physical and chemical processes occurring during a SA. Thus, if a code is able to reproduce correctly most of SETs or CETs, this does not mean that it can ensure the same quality of results for the more realistic and complex Integral Effect Tests (IETs).

Nevertheless, some contraindications exist also for the IETs, because of the scaling factor effect which, being difficult to quantify, could lead to development of models replicating correctly the experimental data but that are not totally reliable for reactor scale applications.

In addition, waiting for the dismantling of Fukushima Dai-ichi units 1–2 and 3, which can provide new information in the field of the core degradation phenomena, in the present only data on in-vessel core melt progression for the Three Mile Island, Unit 2 (TMI-2) accident are available.

Therefore, apart from simulating the TMI-2 accident scenario, there is no other way to assess models of in-vessel phenomena (H₂ production, corium pool/ debris formation and relocation, etc) implemented in the codes on reactor scale.

As a consequence, there are still uncertainties in the capability of these integral tools to predict the progression of postulated severe accident transients in a NPP, notably core degradation and ex-vessel phenomena.

In order to investigate the differences among the physical models adopted in MELCOR 2.2 and ASTEC V2.2 codes, and evaluate the impact that they could have on the assessment of Severe Accident Management Strategies, a crosswalk study was carried-out.

This paper describes the first phase of this work, which consists of a detailed comparison between these two SA tools on a well-defined plant (PWR1000-Like) with prescribed boundary and initial conditions.

Furthermore, to avoid additional and unwanted sources of discrepancies between code calculations, the two reactor models were developed in parallel following the recommendations suggested by developers for core and Reactor Coolant System (RCS) nodalization.

The comparative study is focused on the TH of the RCS and in-vessel phenomena, for two different accident scenarios (SBLOCA and CSBO) up to the moment of lower-head failure.

Ex-vessel behaviour will be examined in the second phase of this study.

The crosswalk results obtained have shown some differences and similarities on reproducing TH in the RCS and In-vessel phenomena.

Especially for these latter, the discrepancies predominate mainly due to how the codes treat corium behaviour, while thanks to the harmonisation of the initial steady-state and boundary conditions, the discrepancies on the prediction of the RCS behaviour have been minimized, at least before significant core degradation takes place.

在Tractebel中，使用了事故源术语评估代码（ASTEC）和泄漏估算和排放后果方法（MELCOR）系统代码，以在发生严重事故（SA）的情况下执行源术语评估和工厂热力水力（TH）分析。比利时核电厂（NPPs）。

这些系统代码是通过密集而广泛的验证活动开发的，这些活动主要基于对已实现物理模型的能力进行验证，以再现独立效果测试（SET）和耦合效果测试（CET）的实验结果。

但是，SET（更不用说扩展了CET）没有考虑到在SA期间发生的不同物理和化学过程之间可能发生的相互作用以及可能产生的协同作用或拮抗作用。因此，如果代码能够正确重现大多数SET或CET，则并不意味着它可以确保更现实，更复杂的整体效果测试（IET）的结果质量相同。

尽管如此，IET也存在一些禁忌症，因为比例因子效应难以量化，可能导致开发出能够正确复制实验数据但对于反应堆规模应用而言并非完全可靠的模型。

此外，等待拆除福岛第一核电站的1-2号和3号装置，这可以为岩心退化现象领域提供新的信息，目前只有三英里岛的船内岩心融化进展数据，第2单元（TMI-2）事故可用。

因此，除了模拟TMI-2事故场景外，没有其他方法可以评估反应堆规模的代码中实施的船内现象模型（H ₂产生，真皮池/碎片形成和迁移等）。

结果，这些整体工具预测NPP中假定的严重事故瞬变（尤其是堆芯退化和前驱现象）的能力仍存在不确定性。

为了调查MELCOR 2.2和ASTEC V2.2代码中采用的物理模型之间的差异，并评估它们对严重事故管理策略评估的影响，进行了人行横道研究。

本文介绍了这项工作的第一阶段，其中包括在定义好的边界条件和初始条件下，在定义明确的工厂（PWR1000-Like）上对这两种SA工具进行详细比较。

此外，为避免代码计算之间出现差异的额外和不必要的来源，根据开发人员对堆芯和反应堆冷却剂系统（RCS）节点化的建议，并行开发了两个反应堆模型。

对比研究的重点是直到下头部故障时两种不同的事故场景（SBLOCA和CSBO）的RCS的TH和车内现象。

在本研究的第二阶段将检查前容器行为。

获得的人行横道结果显示了在RCS和In-vessel现象中重现TH的某些差异和相似之处。

尤其是对于后者，差异主要是由于代码如何处理皮质行为，而由于初始稳态和边界条件的协调，至少在显着之前，RCS行为预测的差异已被最小化。发生核心退化。