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P. Král, V. Pištora (ÚJV Řež):

Impact of ECCS Design of VVER Reactors on PTS Issue

Abstract In 1996, a study of Pressurized Thermal Shock (PTS) was started in the Nuclear Research Institute Rez (UJV) covering the NPPs operated in the Czech Republic. PTS analyses for NPP Dukovany (VVER-440/213) have been performed first and later a similar study was initiated for NPP Temelin (VVER-1000/320).

Our PTS analyses produced many interesting interim results concerning design of Emergency Core Cooling System (ECCS). The presented paper deals with these findings and summarizes possible improvements of ECCS from the point of view of PTS issue. In general, it can be said that design of ECCS system of VVER (or PWR) was previously focused, more or less, only to emergency core cooling and containment spraying, and no (or minimal) attention was paid to PTS issue.

However, the injection of cold water from the ECCS system (during an accident with ECCS actuation) is potentially a major contributor to rapid cooldown of reactor coolant system (RCS), where especially the fast and unsymmetrical temperature drop in reactor downcomer (DC) could challenge the reactor pressure vessel (RPV) integrity. The fast decrease of coolant temperature in DC would lead to a significant increase of thermal stresses in the thick wall of embrittled RPV. Consequently, these thermal stresses together with stresses from inner pressure could in conditions of relatively cold RPV material initiate a fast fracture from a flaw potentially existing in the RPV wall.

Nowadays it is obvious that the ECCS design should take into account not only the needs of emergency cooling of the reactor core (and containment spraying), but also the PTS issue. The ECCS design should be optimized from all these points of view. This is however not an easy task, because requirements of core cooling and minimization of PTS risks are usually contradictory.

Naturally, this challenging task would be easier to fulfill while designing a new NPP (VVER or PWR in general). However, it is possible to improve the ECCS design and thus mitigate PTS during hypothetical accidents also in the existing power plants. Measures like heat-up of ECCS tanks, hydroaccumulators and/or containment sump, reduction of hydroaccumulator pressure, bypassing of ECCS coolers, etc. have been applied at some NPPs with VVER reactors. In our paper, we discuss these potential NPP modifications – both those applicable at existing power plants and those applicable only in design of new VVER (or PWR) reactor systems.

Nomenclature DC (reactor) downcomer EBS Emergency Boration System ECCS Emergency Core Cooling System ESFAS Engineered Safety Features Actuation System HA hydroaccumulator HPIS High Pressure Injection System LBLOCA large-break loss-of-coolant accident LOCA loss-of-coolant accident LPIS Low Pressure Injection System MBLOCA medium-break loss-of-coolant accident

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MSLB main steam line break NPP nuclear power plant PRISE primary-to-secondary leak PRZ SV pressurizer safety valve PTS pressurized thermal shock PWR pressurized water reactor RCP reactor coolant pump RCS reactor coolant system RPV reactor pressure vessel RV relief valve SG SV steam generator safety valve SI safety injection TH thermal hydraulic Tk end-of-life critical temperature of brittleness Tk

a maximum allowable critical temperature of brittleness

1. Main objectives of ECCS system (design basis functions)

The purpose of Emergency Core Cooling System (ECCS) is to minimize consequences of postulated accidents.

Basic objectives of ECCS system:

a) Core cooling in conditions of loss-of-coolant accident (LOCA).

b) Ensuring subcriticallity by injection of boric acid into RCS in accidents connected with primary coolant (= moderator) temperature decrease (MSLB, LOCA).

c) Containment spraying (to keep containment pressure and temperature below design limits in course of LOCA or MSLB).

d) Further functions like residual heat removal in conditions of normal reactor shutdown (not in VVER-440 case), cooling of spent fuel pool, etc.

ECCS consists of the following main parts:

• High Pressure Injection System (HPIS),

• Emergency Boration System (EBS, not in the VVER-440/213 design),

• Hydroaccumulators (HA),

• Low Pressure Injection System (LPIS),

• Sump(s) and heat exchanger(s).

Also the make-up system should be considered while evaluating PTS risk of the cold water injection into RCS. From the PTS point of view, the following parameters are especially important: flow rate, shutoff head, capacity of tanks, water temperature, actuation logic, and exact position of the connections to primary loops.

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2. General overview of major ECCS modifications applied recently at VVER

The following major modification of ECCS have been implemented at the VVER power plants:

Examples of ECCS modifications applied lastly:

1994: NPP Dukovany with VVER-440/213 (Czech Rep.): • Heat-up of HPIS tanks to 55-60 °C (the same was done at some other VVERs)

1998: NPP Mochovce with VVER-440/213 (Slovakia, pre-operation design modifications):

• Heat-up of HPIS tank, LPIS tanks, and hydroaccumulators 1998: NPP Dukovany with VVER-440/213 (Czech Rep.):

• HA pressure reduction 6.0→3.5 MPa 2000: NPP Loviisa with VVER-440 (Finland):

• HA pressure reduction 5.5→3.5 MPa • HA level increase 6.0→7.5 m • Replacement of LPIS pumps (increased shutoff head 0.7→1.1 MPa)

2002: NPP Temelin with VVER-1000 (Czech Rep.):

• Heat-up of sump water to 50-55 °C • Bypass of ECCS cooler (cca 30% of total flow rate from sump)

(these 2 modifications were applied as pre-operational modifications suggested by Russian designer in the nineties; they were applied also at some other NPPs with VVER-1000)

Future modifications (under preparation, approximate dates): 2005: NPP Dukovany (Czech Rep.):

o HA level increase 6.2→7.25 m

2005: NPP Bohunice V-2 (Slovakia): o HA pressure reduction 6.0→3.5 MPa o Replacement of LPIS pumps (higher flow rate and shutoff head) o Installation of three-way valves into HPIS pumps discharge lines

2006: NPP Paks (Hungary): o HA pressure reduction 6.0→3.5 MPa and o HA level increase 6.0→7.1 m

Unfortunately, just minor part of these ECCS modifications (heat-up of ECCS tanks) was or will be focused on reduction of the PTS risk.

For example, the main objectives of the HA pressure reduction were to increase efficiency of HA injection during LOCA and to improve conditions for operator interventions during PRISE (HA pressure was shifted below SG SV pressure set-points).

Next, the purpose of HA level increase was again to enhance the efficiency of HA injection during LOCA plus to minimize the interval between the end of HA injection and start of LPIS injection (problem of low shut-off head of LPIS pumps in NPP with VVER-440/213).

Finally, replacement of LPIS pumps by pumps of higher shut-off head at NPP Loviisa should also have minimized or even canceled the time interval between HA and LPIS injection.

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3. Thermal hydraulic mechanisms leading to PTS

The main potential contributors to Pressurized Thermal Shock (PTS) can be divided into two groups – the phenomena leading to increased stresses (in RPV wall) due to inner pressure and the phenomena leading to increased thermal stresses in RPV wall.

Major mechanisms that can lead to increased thermal stresses in RPV wall:

• Fast RCS cooldown due to increase of heat transfer at SGs (MSLB etc.) = symmetrical cooldown or = asymmetrical cooldown

• Fast RCS cooldown caused by break outflow during LBLOCA or MBLOCA = symmetrical cooldown

• Primary temperature decrease due to safety injection during LOCA or MSLB = asymmetrical cooldown

The RCS cooldown from increased heat transfer in SGs during MSLB or from large energy break outflow from RCS during LBLOCA or MBLOCA can lead to very fast, but not extremely deep cooldown – the temperatures would not drop below 100 °C. However, as soon as the ECCS injection is initiated in these transients, the temperatures in RCS (especially at reactor inlet and in reactor downcomer) can drop considerably below 100 °C (in some cases even below 20 °C).

Major mechanisms that can lead to increased stresses due to inner pressure:

• High pressure safety injection – HPSI (or EBS injection)

• Pressure increase due to thermal expansion of primary coolant during later heat-up (typical for later phase of MSLB; dangerous especially in case of RCS full of liquid)

• Make-up water supply into RCS

Special attention should be paid to possible repressurization of the primary system after leak isolation. Usually, this is possible only in the case of “inadvertent opening of PRZ SV” (repressurization after reclosure of the PRZ SV), but in the VVER-440 NPPs, this is also possible in case of a break in primary loops (if the break is in separable part of the loop and operator closes the main gate valves in the loop).

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4. Basic information on PTS studies of NPP Dukovany and NPP Temelin

PTS study for NPP Dukovany (VVER-440/213) was started in 1996 and should be finished by the end of 2004. The analyses have been performed in UJV Řež [Ref.2, 7, 9, 12].

Also the PTS study for NPP Temelin (VVER-1000/320) is being elaborated in UJV Řež. It was started in 2001 and should be finished also by the end of 2004 [Ref.10, 11].

The deterministic approach was used in both PTS studies for Czech NPPs. Each PTS analysis consists of 3 basic steps:

1. Selection of initiating events relevant to PTS

2. Thermal-hydraulic (TH) analysis

3. Structural analysis

A more detailed “technological scheme” of the PTS analysis can look as follows:

1. Selection of relevant initiating events and scenarios

2a. System TH analyses for the given initiating event (with model of the whole NPP system; usually a set of variants is computed)

2b. Selection of the most adverse case(s) from the point of view of PTS (based on expert judgment of resulting system TH parameters)

2c. Detailed mixing calculation of the most adverse cases (only reactor downcomer and cold legs with SI are usually modeled)

3a. Computation of temperature fields in RPV wall

3b. Computation of thermal and pressure stresses in RPV wall

3c. Fracture mechanical assessment of postulated cracks (several types of postulated cracks are evaluated)

3d. Determination of maximum allowable critical temperature of brittleness TKa

In some cases (with symmetrical and “homogeneous” cooldown) the mixing TH calculation can be omitted and parameters from system TH calculation can be directly transferred to the structural analysis.

Conservative assumptions have been applied in compliance with IAEA guide “Guidelines on Pressurized Thermal Shock Analysis for WWER Nuclear Power Plants, IAEA-EBP-WWER-08, 1997” [Ref.1].

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As for the specific computational tools, the following computer codes were used in our PTS analyses for NPP Dukovany and NPP Temelin:

System TH analyses:

• RELAP5

• MELCOR (Dukovany confinement behavior)

• COCOSYS (Temelin containment behavior)

Detailed mixing calculations:

• REMIX/NEWMIX

• CATHARE 2D

• FLUENT

Structural analyses:

• SYSTUS (finite element code)

• COSMOS/M (finite element code)

• ORMGEN (mesh generator)

Fig.1 Nodalization scheme of NPP Dukovany RCS (only one of six loops is depicted)

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Example of results from PTS study– spectrum of MSLB analyses for NPP Temelin

In PTS study for NPP Temelin, four categories of PTS transients have to be evaluated. So far, 3 categories of initiating events have been completely analyzed – the MSLB events (13 final system TH analyses), PRISE (4), and LOCA (17). The work on the last category (consisting of inadvertent opening of PRZ SV and other transients) is in progress.

In the category “Main Steam Line Break” (MSLB), the system TH analyses of 12 scenarios were performed in 2002. Based on comparison of major TH parameters (relevant to PTS), the 3 worst cases were selected for further evaluation (mixing calculation, structural analysis).

Individual MSLB scenarios differ in:

• size and position of break (leak),

• initial reactor power,

• status of RCP in the transient,

• availability of ECCS (maximal of minimal).

In the three Figures below, one can see comparisons of the basis results obtained from the system TH calculations (reactor inlet temperature, primary pressure, flow rate in loop with SI) for all 12 scenarios. Based on expert judgment, the scenarios SLB1A, SLB1B, and SLB1C as the potentially most adverse ones were selected for further evaluations (mixing calculation, structural analysis). The SLB1A, SLB1B and SLB1C were selected because of the fastest temperature drop at reactor inlet from loop-1, for the most suitable conditions for cold plumes formation (early flow stagnation in loop-1 with SI), and for the reason of very high primary pressure, respectively [Ref.10, 11].

Fig.2 Comparison of reactor inlet temperatures in system TH analyses of MSLB

T e m p e ra tu re in C L 1 - re a c to r in le t

9 0

9 2

9 4

9 6

9 8

1 0 0

1 0 2

1 0 4

1 0 6

1 0 8

1 1 0

1 1 2

1 1 4

1 1 6

1 1 8

1 2 0

1 2 2

1 2 4

1 2 6

1 2 8

1 3 0

0 5 0 0 1 0 0 0 1 5 0 0 2 0 0 0 2 5 0 0 3 0 0 0 3 5 0 0 4 0 0 0 4 5 0 0 5 0 0 0 5 5 0 0 6 0 0 0 6 5 0 0 7 0 0 0 7 5 0 0tim e [s ]

[şC ] S L B 1 A S L B 1 B S L B 1 C S L B 1 D S L B 2 A S L B 2 B S L B 3S L B 4 A S L B 4 B S L B 5 S D A S D C

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Fig.3 Comparison of primary pressure in system TH analyses of MSLB

Fig.4 Comparison of loop flow rates in system TH analyses of MSLB

P re s s u re

7

8

9

1 0

1 1

1 2

1 3

1 4

1 5

1 6

1 7

1 8

0 5 0 0 1 0 0 0 1 5 0 0 2 0 0 0 2 5 0 0 3 0 0 0 3 5 0 0 4 0 0 0 4 5 0 0 5 0 0 0 5 5 0 0 6 0 0 0 6 5 0 0 7 0 0 0 7 5 0 0tim e [s ]

[M P a ] S L B 1 A S L B 1 B S L B 1 C S L B 1 D S L B 2 A S L B 2 B S L B 3S L B 4 A S L B 4 B S L B 5 S D A S D C

F lo w ra te in C L 1 - M C P o u tle t

-0 ,1 0

-0 ,0 5

0 ,0 0

0 ,0 5

0 ,1 0

0 ,1 5

0 ,2 0

0 ,2 5

0 ,3 0

0 ,3 5

0 ,4 0

0 ,4 5

0 ,5 0

0 ,5 5

0 ,6 0

0 ,6 5

0 ,7 0

0 ,7 5

0 ,8 0

0 5 0 0 10 0 0 1 50 0 2 0 0 0 25 0 0 3 00 0 3 5 00 4 0 00 4 5 0 0 50 0 0 5 50 0 6 0 0 0 65 0 0 7 00 0 7 5 00

tim e [s ]

[m /s ]S L B 1 A S L B 1B S LB 1 C S L B 1 D S L B 2 A S L B 2 B S LB 3S L B 4 A S L B 4B S LB 5 S D A S D C

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SLB1B – Results of Structural Analysis SLB1B = main steam line break by SG1, from hot zero power, min. ECCS (HPIS ⇒ CL3) Input parameters for the structural analysis (obtained from TH calculations: RELAP5+REMIX):

- primary pressure, - reactor DC temperatures and heat transfer coefficients.

Worst crack configuration: · axial crack, · location in weld No. 3, · a/c = 0.3, · crack depth = 20 mm, · worst position along the crack front: the deepest point. Warm prestressing approach applied. Critical time: 1980 s Tka = 108.6 °C

Fig.5 Temperature variations under loop-3 inlet nozzle in weld No.3 position (in SLB1B)

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Fig.6 Circumferential stress at 2000 s for SLB1B scenario

Fig.7 Dependence of KI on temperature for SLB1B scenario

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5. Possible modifications of ECCS to reduce the PTS risk

Generally, the major challenge in designing the Emergency Core Cooling System (ECCS) can be formulated as follows:

“Optimize the ECCS capacity and parameters to ensure emergency core cooling on the one hand while not endangering RPV integrity due to too fast cooldown on the other hand”

Maintaining other functions of ECCS like containment spray, etc. could be also a problem, but not so big as fulfilling of the two contradictory tasks mentioned above. So, when modifying ECCS parameters in favor of PTS, one must always check ECCS functionality from the point of view of emergency core cooling.

Improvements of ECCS applicable at operating NPP or in designing new NPPs:

• Heat-up of HPIS tanks.

• Heat-up of LPIS tanks (could be problematic if the tanks are common for LPIS and containment spray system – then the water temperature increase would lead to reduced confinement spraying efficiency).

• Heat-up of sump water (the same limitation from spray efficiency like in case of LPIS tanks).

• Heat-up of water in HA (it could be sufficient to heat up only HAs connected to reactor downcomer or to cold legs; heat-up of HAs connected to upper plenum or hot legs is not necessary because their impact on PTS is minimal).

• Filling of HA with warm water during NPP start-up (if not applying heat-up of HA during the whole operation, the filling of HA with warm water would at least prevent a state with ≈ 20 °C cold water in HA at beginning of cycle; later on the HA temperature follows air temperature in containment, i.e. 40-60 °C).

• Decrease in HA pressure (consequently the HA would inject into RCS at lower primary pressure and at lower ambient temperature in DC; again from the PTS point of view, it could be done only in HAs connected to DC or CLs).

• Optimization of ECCS cooler operation mode (control of essential cooling water flow rate, control of ECCS cooler bypass etc.).

• Improvement of ESFAS to cover all MSLB events (to avoid fast depressurization and cooldown from secondary system).

Improvements of ECCS applicable in new NPPs design:

• Optimization of ECCS cooler bypass piping design (to get inlet and outlet of the bypass piping at the same elevation and thus avoid effect of buoyancy forces that can lead to stopping of bypass or cooler flow in the conditions of low SI flow rate).

• Separate SI and spray tanks (to enable heat-up of SI tanks and retain cold water in spray tanks).

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• Prevent uncontrolled flooding of reactor pit (to minimize the risk of PTS at outer surface of RPV).

• Design of SI pumps (shutoff head etc.) optimized from the PTS point of view (to minimize the risk of PTS and cold overpressurization).

• Optimized layout of ECCS injection into individual loops and reactor inlet nozzles (plus optimized design of ECCS nozzles and design of flow baffles).

Further possible improvements of NPP to reduce PTS risk:

Applicable also at operating NPPs:

• Protection against cold overpressurization (through automatic opening of PRZ RV).

• Improvements of EOP (termination of SI, very careful approach to break isolation, preventive opening of PRZ SV/RV, influencing water temperature injected from SI systems, start of RCP to decay cold plumes, etc.).

• Annealing of RPV.

• Replacement of embritteled RPV (by a vessel made from better material and/or with more suitable weld positions; but even if replacing the bottom part only, and not the reactor head and internals, it would be an extremely expensive and complicated solution, unprecedented so far).

Applicable in new NPPs design:

• Doubling of PRZ RVs (for case of inadvertent opening/closing of one PRZ RV).

• Improved PRZ SV/RV construction (to avoid uncontrolled reclosure after inadvertent opening or to enable operator to stuck the valve in open position).

• Improved reactor design (from the point of view of PTS) (optimized fuel loading to reduce neutron fluence at RPV wall, thicker reactor downcomer, material of RPV less sensitive to embrittlement, elevation of RPV welds out of core region etc.).

• Design of reactor inlet section suitable from the PTS point of view (to prevent cold plumes merging and to reduce ECCS bypass by geometrical disposition and not by flow baffles etc.).

Example of ECCS modification impact on PTS results (quantified):

Within the PTS study for NPP Dukovany, the effect of heat-up of LPIS tanks (20 ⇒ 55 °C) was analyzed for LOCA with break D200 mm in hot leg [Ref.7, 8, 9]. According to the auxiliary analyses for 15 mm deep surface crack, such a heat-up would improve the PTS results (i.e. increase Tk

a) by approximately 30 °C. However, in NPP Dukovany (VVER-440/213) the heat-up of LPIS tanks is a problem, because water from LPIS tanks is used also for hermetic confinement spraying. Therefore, the proposal has not been accepted yet.

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6. Summary

The paper deals with ECCS design relevant findings obtained within the PTS studies that have been performed for NPP Dukovany (VVER-440/213) and NPP Temelin (VVER-1000/320). In general, it may be said, that previously the design of ECCS system of VVER (or PWR) was focused, more or less, only to emergency core cooling (and containment spraying), and no or minimal attention was paid to the PTS issue.

However, if the injection of cold water into primary system would lead (through high stresses in RPV wall and low temperature of RPV material) to loss of integrity of the reactor vessel, then any long-term core cooling would be nearly impossible. Solution of this problem is as follows:

The design of ECCS should be optimized to ensure emergency core cooling on the one hand while not endangering the reactor integrity due to fast cooldown and consequent

thermal stresses in RPV wall on the other hand.

This is however not an easy task, because requirements of core cooling and PTS issue are usually contradictory. Naturally, it is “easier” to fulfill it in designing a new nuclear power plant, but also the existing NPPs can be modified to reduce the PTS risk. Maintaining of other functions of the ECCS, such as containment spray, etc. could be also a problem, but not so big as fulfilling of the two contradictory tasks mentioned above.

A set of possible improvements of ECCS (and also other NPP systems) from the PTS point of view is listed and briefly discussed. Most of them are valid not only for VVER type reactor systems, but also for the PWR ones. Measures like heat-up of ECCS tanks, hydroaccumulators and/or containment sump, decrease of hydroaccumulator pressure, bypassing of ECCS coolers etc., have been already applied at some NPPs with VVER reactors.

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References

[Ref.1] Guidelines on Pressurized Thermal Shock Analysis for WWER Nuclear Power Plants, IAEA-EBP-WWER-08, IAEA, Vienna, 1997.

[Ref.2] Macek J., Muhlbauer P., Krhounková J., Král P., Malačka M.: Thermal Hydraulic Analyses of NPPs with VVER-440/213 for the PTS Condition Evaluation, NURETH-8, 1997.

[Ref.3] Král P.: NRI Řež Calculation of the IAEA Pressurized Thermal Shock Benchmark. Presented at the Second IAEA Meeting on the PTS Benchmark Exercise, April 1998.

[Ref.4] Král P.: RELAP5 System T/H Analysis of SBLOCA for PTS Evaluation of VVER-440/213 with 1-D and 2-D Nodalization of Reactor Downcomer, Fall CAMP Meeting, Bethesda, October 1999.

[Ref.5] Král P.: Thermal Hydraulic System Analysis of IAEA PTS Benchmark with 1-D and 2-D Nodalization of Reactor Downcomer, July 1999.

[Ref.6] Unified Procedure for Lifetime Assessment of Components and Piping in VVER NPPs, VERLIFE project of the 5th Framework Programme of the EU, final version, 2003.

[Ref.7] Král P., Parduba Z., Malačka M.: Thermal Hydraulic Analysis of LOCA with Break D200 in Hot Leg. Analysis Performed in Frame of NPP Dukovany PTS Study, UJV Rez, September 2000.

[Ref.8] Pištora V.: Integrity Assessment of the Reactor Pressure Vessel in NPP Dukovany for the Loss-of-Coolant-Accident in Hot Leg with Equivalent Diameter of the Break 200 mm (LOCA D200) Taking into Account Heat-up of TH Tanks UJV Rez, 2002.

[Ref.9] Krhounkova J., Kral P.: Thermal Hydraulic Analysis of Break D200 mm in Hot Leg, Analysis for PTS Study with Heat-up of LPIS Tanks, UJV Rez, June 2002.

[Ref.10] Král P., Mühlbauer P., Malačka M.: Overview of TH Analyses Results for PTS, Presentation at Czech-Austrian PTS Seminar, UJV Rez, May 2004.

[Ref.11] Pištora V.: Summary of Results from Integrity Evaluation. Presentation at Czech-Austrian PTS Seminar (SUJB), UJV Rez, May 2004.

[Ref.12] Pištora V., Král P.: Evaluation of Pressurized Thermal Shocks for VVER 440/213 Reactor Pressure Vessel in NPP Dukovany, Proceedings of 17th International Conference on Structural Mechanics in Reactor Technology (SMIRT 17), Prague, Czech Republic, August 17-22, 2003.

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Contact information:

Pavel Kral UJV Rez 250 68 Rez Czech Republic

mail: kra@ujv.cz tel.: +420-2-6617-2447