J. Birchley

Analysis of the Quench-14 Bundle Test With M5® Cladding

The QUENCH experimental programme at Forschungszentrum Karlsruhe (FZK) investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions, but where the geometry is still mainly rod-like and degradation is still at an early phase. The QUENCH test bundle is electrically heated and consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The cladding and grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO2 pellets. Experiment QUENCH-14 was successfully performed at FZK in July 2008 and is the first in this programme where Zr-Nb alloy M5® is used as the fuel rod simulator cladding. QUENCH-14 was otherwise essentially the same as experiment QUENCH-06, which was the subject of the CSNI ISP-45 exercise. It is also the first of three experiments in the QUENCH-ACM series, recently launched to examine the effect of advanced cladding materials on oxidation and quenching under otherwise similar conditions. Pre- and post-test analyses were performed at PSI using a local version of SCDAP/RELAP5 and MELCOR 1.8.6, using input models which had already been benchmarked against QUENCH-06 data. Preliminary pre-test calculations with both codes and alternative correlations for the oxidation kinetics indicated that the planned test protocol would achieve the desired objective of exhibiting whatever effects might arise from the change in cladding-material in the course of a transient similar to QUENCH-06. Several correlations were implemented in the models, namely Cathcart-Pawel, Urbanic-Heidrick, Leistikow-Schanz and Prater-Courtright for Zircaloy-4 (Zry-4), and additionally a new candidate correlation for M5® based on recent separate-effects tests performed at FZK on M5® cladding samples. Analyses of the QUENCH-14 data demonstrate strengths and limitations of the various models. Some tentative recommendations are made concerning choice of correlation and effect of cladding material.Copyright

Long Term SBO With Selected Mitigative Measures: MELCOR Parametric Calculations for a 2-Loop PWR

Analyses of three long term PWR Station Black-Out (SBO) scenarios with and without mitigation are performed using the US NRC source term code MELCOR. Refilling of the secondary side of the steam generator (SG) with fire water pumps was previously identified as a potentially important accident management measure. To assess parametrically the effect of the restored availability to refill of SG secondary, SBO sequences are analyzed both without any mitigation and with mitigation at different times into the accident representing different stages of the accident progression. The scenarios studied were (i) base-line SBO without any thermally-induced Reactor Coolant System (RCS) breach, and then sequences with assumption of (ii) Surge Line failure and (iii) thermally-induced SG tube rupture (SGTR). A detailed model was used for the description of flows in the hot legs and in SGs to enable us to simulate the counter-current natural circulation which is inherent in these types of scenarios, with the cold leg plugged by water in the loop seal. These were long-term simulations, some of them up to 5 days of the transient, with the analyses of the containment response and fission product (FP) releases to environment. For the relevant cases, the molten core-concrete interactions (MCCI) are modeled in detail in the complex lower containment geometry with several distinct volumes (cavity and other rooms with concrete walls) subject to ablation by the molten corium; the challenges to containment barrier by the late overpressurization and by concrete ablation are evaluated. With the Passive Autocatalytic Recombiners, PARs, installed the hydrogen is found to be of less threat to containment integrity. For the thermally-induced SGTR the impact of the chosen mitigation strategy on potential bypass FP release to environment is also assessed for different times of the refill. The results indicate that mitigation by the SG secondary refill can be very effective for the base-line SBO sequences. For the thermally-induced SGTR it is effective only when the refill is achieved early after the SG tube failure. Mitigation by the pressurizer-loop SG refill was much less successful in the case of the Surge Line failure sequences.Copyright

Spent Fuel Pool Under Severe Accident Conditions

The possibility of a spent fuel severe accident has received increasing attention in the last decade, and in particular following the Fukushima accident. Several large scale experiments and also separate effect tests have been conducted to obtain a data base for model development and code validation. The outcome of the Sandia BWR Fuel Project was used to define the flow parameters adjusted for the low pressure and the increased flow resistance due to the presence of the spent fuel racks which resulted in reduced buoyancy driven natural circulation flow compared with reactor geometry. The possibility of a zirconium fire, using the flow parameters obtained from the spent fuel experiments, is investigated in the present work. The important outcome of the study is that spent fuel will degrade if temperatures above 800 K are reached. In partial loss of coolant accidents the flow through the lower bottom nozzle is blocked and because there is no cross flow possible due to the spent fuel racks the coolant flow in the upper dry part of the spent fuel is limited by the steam production in the lower still wetted part of the fuel. This accident scenario leads to the fastest heat up in a postulated spent fuel accident. The influence of different kind of spent fuel storage (hot neighbour and cold neighbour) is investigated. An important factor in these calculations is the radial heat transfer to the neighbouring fuel assemblies. In the present work limits of the spent fuel storage under accident conditions (minimum allowed water levelin the pool) and total loss of coolant (maximum coolable decay heat per fuel assembly) are shown and explained.Copyright

Corium and Debris Coolability Studies Performed in the Severe Accident Research Network of Excellence (SARNET2)

The motivation of the work performed within the work package “Corium and Debris Coolability” of the Severe Accident Research Network of Excellence (SARNET) is to reduce or possibly solve the remaining uncertainties on the efficiency of cooling reactor core structures and materials during severe accidents, either in the core, in the vessel lower head or in the reactor cavity, so as to limit the progression of the accident. This can be achieved either by ensuring corium retention within the reactor pressure vessel or at least by limiting the corium progression and the rate of corium release into the cavity. These issues are to be covered within the scope of accident management for existing reactors and within the scope of design and safety evaluation of future reactors. The specific objectives are to create and enhance the database on debris formation, debris coolability and corium behavior in the lower head, to develop and validate the models and computer codes for simulation of in-vessel debris bed and melt pool behavior, to perform reactor scale analysis for in-vessel corium coolability and to assess the influence of severe accident management measures on in-vessel coolability. The work being performed within this work package comprises experimental and modeling activities with strong cross coupling between the tasks. Substantial knowledge and understanding of governing phenomena concerning coolability of intact rod-like reactor core geometry was obtained in previous projects. Hence the main thrust of experimental and modeling efforts concentrates mainly on the study of formation and cooling of debris beds in order to demonstrate effective cooling modes, cooling rates and coolability limits. Modeling efforts have been aimed at assessing and validating the models in system-level and detailed codes for core degradation, oxidation and debris behavior. The paper describes the work performed up to now and summarizes the main results achieved so far.Copyright