19th Symposium of AER on VVER Reactor Physics and Reactor Safety

Joint AER Working Group A on ?Improvement, extension and validation of parameterized few-group libraries for VVER-440 and VVER-1000? and AER Working group B on ?Core design? eighteenth meeting was hosted by ?KODA JS a.s. in Plze? (Czech Republic) during the period of 4th to 6th May 2009. There were present altogether 16 participants from 6 member organizations and 13 presentations were read.

19th AER-Symposium on VVER-Reactor Physics and Safety, 21.-25.09.2009,
St. Constantine and Elena resort, International House of Scientists, Bulgaria

Application of a Monte Carlo method for simulating of neutron transport in nuclear reactors has more than fifty years of history. Fast progress of computer power and development of more and more perfect algorithms, codes and nuclear data bases allow solving more challenging problems, including three-dimensional calculations of full-scale reactor cores. In a process of verification of MCU-PD code full-scale three-dimensional model of first unit of the Volgodonsk nuclear power plant has been developed. Some critical states were calculated using that model. Those are states at hot zero power level achieved during physical start-up. Power distribution in fuel assemblies throughout core and power distribution throughout fuel assembly height were calculated by MCU-PD and BIPR-7A codes. Concise description of full-scale model developed, annotation of algorithms and methods for the both codes and their results comparison between themselves and with experimental data are presented.

Dominance ratio, or more precisely, its closeness to unity, is important characteristic of large reactor. It allows evaluate beforehand the number of source iterations required in deterministic calculations of power spatial distribution. Or the minimal number of histories to be modeled for achievement of statistical error level desired in large core Monte Carlo calculations. In this work relatively simple approach for dominance ratio evaluation is proposed. It essentially uses core symmetry. Dependence of dominance ratio on neutron flux spatial distribution is demonstrated.

HELIOS2, is a 2D transport theory program for fuel burnup and gamma-flux calculation.
It solves the neutron and gamma transport equations in a general, two-dimensional geometry
bounded by a polygon of straight lines. The applied transport solver may be chosen between:
The Method of Collision Probabilities (CP) and The Method of Characteristics(MoC). The
former is well known for its successful application for preparation of cross section data banks
for 3D simulators for all types lattices for WWERs, PWRs, BWRs, AGRs, RBMK and
CANDU reactors. The later, MoC, helps in the areas where the requirements of CP for computational power become too large of practical application.
The application of HELIOS2 and The Method of Characteristics for some large from calculation point of view benchmarks is presented in this paper. The analysis combines comparisons to measured data from the Hungarian ZR-6 reactor and JAERI?s facility of Tanktype
Critical Assembly (TCA) to verify and validate HELIOS2 and MOC for WWER assembly
imitators; configurations with different absorber types- ZrB2, B4C, Eu2O3 and Gd2O3; and
critical configurations with stainless steel in the reflector. Core eigenvalues and reaction rates
are compared. With the account for the uncertainties the results are generally excellent.
Special place in this paper is given to the effect of Iron-made radial reflector. Comparisons to measurements from TIC and TCA for stainless steel and Iron reflected cores are presented. The calculated by HELIOS-2 reactivity effect is in very good agreement with the
measurements.
KEYWORDS: HELIOS2, ENDF/B-VII, method of characteristics, benchmarking,WWER1000, Iron reflector.

A new three-dimensional nodal model for neutronics calculation is currently under implementation into APROS – Advanced PROcess Simulation environment – to conform the increasing accuracy requirements. The new model is based on an advanced nodal code HEXTRAN and its static version HEXBU-3D by VTT, Technical Research Centre of Finland. However, several improvements for the model are made and the whole model has been reprogrammed. They don?t change the theory basement of the method, but rather makes the implementation more flexible. Currently the computational part of the program is ready and current work concentrates on testing and validation. User interface details and usability issues need also work in the future. In this paper, general information about the improvements of the theory is explained first. Then the latest validation results are given. Currently the dynamical characteristics are tested by calculating the AER?s kinetic benchmarks for VVER-440 reactors. In this paper, the results for the first benchmark are shown for two version of the code. The first version is fully HEXTRAN-comparable code to test that the basic structure works as wanted. The second version is the actual improved model for APROS.

Absorber element (AE) with compound absorber has been operating in WWER-1000 power units since 1995. AE design meets operating organizations requirements for reliability, service life (to 10 years) and safety functions. Extension of AE service life up to 20 ? 30 years by the complex of calculation and experimental work is an important problem of WWER new designs development.

The paper states that the regulatory documentation, as a rule, do not give explicit recommendations on formation techniques of engineering safety factors for design limited parameters of normal operation (
). The AER countries use different approaches to
evaluation (sometimes even one country in relation of various power units).

The methodology applied for the safety factor assessment of the VVER fuel cycles uses methods and terms of statistics. Value of the factor is calculated on the basis of estimation of probability to meet predefined limits. Such approach demands the special attention to the statistical properties of parameters of interest.

In the first decade of June 2008, during the power commissioning of the reactor at Mohovce NPP unit 1, the experiment with reducing the thermal power of core almost to the balance-of-plant (BOP) needs was performed. After the reactor has operated for seven hours at low power (about 200(220 MW (thermal)), its power was increased (at a rate of about 0.25% of Nnom/min) to the initial level, close to 107% (1471 MW).

Review of AER Working Group B Meeting in Plze? – Hotel ?U Pramen??, Czech Republic is given.

The VVER-440 reactors of Loviisa NPP are operated with 1500 MWth power and reduced core. During recent years a 3-batch loading scheme has been used. Loviisa-1 is currently running with BNFL fuel equilibrium cycle and Loviisa-2 with TVEL fuel equilibrium cycle. Our goal is to move to a 4-batch scheme with TVEL fuel for both reactors. To achieve this goal the U-235 enrichment has to be increased from the current designs used. The fuel to be used in the near future is 4.37 % enriched fuel with six Gd2O3 doped rods.

A nearly equilibrium five-year cycle has been achieved at Dukovany NPP over the last years. This means that working fuel assemblies (WFA) with an average enrichment of 4.25 w% (control assemblies (CA) with an average enrichment of 3.82 w%) are normally loaded and reloaded for five years. Operation at uprated thermal power (105% of the original one, increase from 1375 MWt to 1444 MWt) is being prepared by use of WFA with an average enrichment of 4.38 w% (CA with an average enrichment of 4.25 w%).

Development activities, presented in the paper, contribute to the evolution of gadolinium fuel for the VVER-440 reactors. Short overview of existing and already proposed Gd-assemblies is given. Two new assemby types with improved features are described. Multiplication coefficients and pin power peaking factors of all mentioned gadolinium assemblies are compared. Animations of power distribution behaviour in selected Gd-assemblies are shown.

!
#$% & ’

At the end of 2008 a transition procedure from the old Russian designed TVS-M fuel assemblies to the advanced TVSA fuel type was accomplished at Kozloduy NPP WWER-1000 Units 5 and 6. The old fuel assemblies had to be changed because of a distortion tendency observed several years before. The TVSA assembly distinguishes itself with much stronger construction and using a mixture of uranium and uniformly distributed Gd as a burnable absorber in 6 or more fuel rods.

The VVER-440 nuclear fuel vendor permanently improve the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. Therefore it is necessary to have the skilled methodology and computing code for analyzing factors which affecting the accuracy of flow redistributed determination through reactor on flows through separate parts of reactor core in case of parallel operation different assembly types. Whereas the geometric parameters of new manufactured assemblies were changed recently, the calculated flows through the fuel parts of different type of assemblies are depended also on their real position in reactor core. Therefore the computing code CORFLO was developed in VUJE Trnava for carrying out stationary analyses of T-H characteristics of reactor core within 60o symmetry. The CORFLO code deals the area of the active core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60o symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is calculated. Computing code is verified and validated at this time. Paper presents the short description of computing code CORFLO with some calculated results.

Replacement of ex-core neutron flux measurement system AKNT-2 at Bohunice NPP was finished at Unit 3 in 2007 and at Unit 4 in 2008. The previous system was replaced by new system AKNT-17R. Neutron power on-line calibration system AKE-02R was installed at Unit 3 in the last year 2008. AKE-02R was working in open-loop till to end of fuel cycle. AKE-02R system was also installed at Unit?4 during outage in 2009 (July 2009).

Developments performed lately at Paks NPP – such as power up rate, application of fuel assemblies with increased pin pitch or containing burnable poison ? caused power distributions within the assemblies differing significantly from the present ones. In-core measurements and preliminary calculations pointed that our codes modelling the coolant flow in the assemblies are not sufficient accurate and can not describe the processes detailed enough. This paper describes the caused problems and their nature and after that their solution by applying a new appropriate sub channel and thermocouple model. Both of them are simple and fast enough to be built in the online core monitoring system but yet they serve sufficient accurate results.

Detailed CFD model for the head parts of the VVER-440 fuel assemblies with burnable poison has been developed. The coolant mixing was analyzed in some typical assemblies with this model and the signals of the in-core thermocouples above the selected assemblies were calculated. The investigations pointed out that the mixing is intensive in these assembly heads but the coolant is not perfectly mixed before reaching the thermocouples. Significant differences between the outlet average coolant temperatures and the thermocouple signals were revealed in the case of the fresh fuels. These deviations can cause about 6% underestimations in the online monitored assembly powers unless a proper correction is introduced. The coolant mixing was also studied by means of numerical tracers and weight factors of selected rod bundle regions for the in-core thermocouple were determined. Using these weight factors and the outlet enthalpies of the assemblies? subchannels, the thermocouple signals can be corrected.

At the beginning of Cycle 22, Unit 4, Paks NPP the Iodine isotopes activity concentrations
raised irregularly in the water of the primary circuit. Analysis supposed that from 1 to 10 fuel
rods in one or more newly loaded follower assemblies had lost their integrity. Due to the fact
it was not necessary to shut down the reactor, but at the end of the cycle sipping tests were
performed for the entire core to find out the facts using a telescope sipping device supplied by
H&B Co., Germany.
This paper describes the circumstances of the emergence of the problem, the operational
inspection and limitation rules in Paks NPP, the theoretical analysis to estimate the scope and
kind of the problem, the sipping device and the measurement/evaluation methods applied for
the practical tests, fulfilment the tests, the results and their evaluation and the conclusions
regarding the event.

One of the limiting factors in term of nuclear power plant lifetime is reactor pressure vessel condition. The most important ageing effect on the reactor pressure vessel is embrittlement. Radiation embrittlement cause especially fast neutrons. In this work we are focused on describing neutron field in reactor pressure vessel area of reactor VVER-440/V-213 using simulation code MCNP 5. Calculation of the neutron field was realized by detailed full-core model of this reactor. The fuel composition is corresponding to the 23th fuel camping of 3th unit in Bohunice nuclear power plant and 7th fuel camping in Mochovce nuclear power plant. Computer simulation of selected neutron-physical parameters will be verified for a consideration of the data from ionization chamber, received from mode department of nuclear power plant in Bohunice and Mochovce. The goal of this work is to improve the assessment of reactor VVER-440/V-213 pressure vessel radiation degradation and following valuation possibility of its lifetime extension and comparison of neutron flux and neutron spectra in most loaded place of reactor pressure vessel and surveillance specimen area.

The AER Working Group D on VVER reactor safety analysis held its 18th meeting in ?e?, Czech Republic, during the period 18-19 May, 2009. The meeting was hosted by the Nuclear Research Institute ?e?. Altogether 17 participants attended the meeting of the working group D, 16?from AER member organizations and 1 guest from a non-member organization. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting.

Description of methodology of conservative RIA analysis with use of spatial kinetic reactor core model is presented. It is shown that their application yields more conservative assessment of reactor core parameters for which acceptance criteria for rod ejection RIA are established, in comparison with point-one-dimensional kinetic model.

A consequence of a total loss of AC power supply (station blackout) leading to unavailability
of major active safety systems which could not perform their safety functions is that the safety
criteria ensuring a secure operation of the nuclear power plant would be violated and a
consequent core heat-up with possible core degradation would occur.

In the traditional fuel-behaviour or hot channel calculations it is assumed that the fuel pellet is centered within the clad. However, in the real life the pellet could be positioned asymmetrically within the clad, which leads to asymmetric gap conductance and therefore it is worthwhile to investigate the magnitude of the effect on maximal fuel temperature and surface heat flux. In this paper our first experiences are presented on this topic.

The Budapest University of Technology and Economics, Institute of Nuclear Techniques (BME NTI) has been working since 2001 on the three-dimensional CFD model of the reactor pressure vessel of the VVER-440 type reactors. During this time period ? due to the development of the available computational capacity ? a very complex and detailed model of the RPV has been developed.

In this article, we study mixing inside TVEL fuel rod bundles used in VVER-440 -type
pressurized water reactors. The computational fluid dynamics (CFD) solver FLUENT 6.3 is
utilized. We are interested in the flow conditions under normal operation where the flow is in
one phase, the average temperature at the end of active fuel is approximately 315?C and pressure
inside the reactor pressure vessel 12 MPa. Our CFD model includes a 30 degree sector of the 2.5
m long fuel rod bundle. In this article the CFD results are presented for both the first and second
generation TVEL fuel assemblies, and some ways to further enhance mixing are studied. The
CFD results are applied in the calibration of certain parameters of the core performance
monitoring system, RESU, used in the Loviisa NPP. In this article, we present comparison
between the CFD and RESU results for both the first and second generation TVEL fuel
assemblies. We conclude that the changes in the geometry of the fuel assembly that have been
made after the year 2000, have only a minor effect on the mixing of the coolant between the
subchannels.

The latest version of SCALE system (SCALE 6) was tested for criticality calculation for VVER-440 fuel.

Codes for preparing master and working AMPX libraries and point-wise (PW) nuclear libraries for SCALE5.1 system have been created. Master and working libraries are constructed from multigroup library in matxs form which are produced by means of the NJOY code. The PW cross-section library is derived from pend files obtained also by NJOY. The AMPX libraries may contain neutron, gamma production and gamma transport data, as well. The produced master libraries can be used either with stand-alone functional modules or with control modules.
An assistant package of programs also has been developed in order to facilitate the usage of NJOY.

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated.

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A?large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an?analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor – SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10.

In the paper, different thorium nuclear fuel cycles are examined and compared under light water reactor conditions, especially VVER-440. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium U-233 content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. Different thorium fuel distribution in fuel assemblies is modeled – homogeneous and heterogenous.