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NEA-1313 BWRDYN.

BWRDYN, Thermal Hydraulic Analysis of a BWR Plant

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1. NAME OR DESIGNATION OF PROGRAM:  BWRDYN.
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2. COMPUTERS
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Program name Package id Status Status date
BWRDYN NEA-1313/01 Tested 14-MAR-2002

Machines used:

Package ID Orig. computer Test computer
NEA-1313/01 Many Computers PC Pentium III,DEC ALPHA W.S.
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3. DESCRIPTION OF PROGRAM OR FUNCTION

BWRDYN performs analysis of transient responses induced by operating mode changes, system malfunctions etc. in a commercial boiling water reactor (BWR) plant. Simulated model includes the reactor system and the balance of plant (BOP) system from the turbine inlet to the feedwater heater.
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4. METHODS

The Euler method is used for integration and first order lag, and the Runge-Kutta method is used for the steamline dynamics.
Major assumptions and simplifications are summarised in the following;
(1) Pressure is assumed to be uniilmn in the reactor vessel. Differential pressures can be calculated optionally in two regions, the lower part of the free surface and the rest of the vesseL
(2) In the thermal-hydraulic model for reactor vessel, saturated coolant region (from core to feedwater sparger) are divided into five regions, core, upper plenum, separator, steam dome, free surface region from feedwater sparger to free surface).
(3) A bubble separation model is used in the subcooled coolant region downcomer, recirculation pipe and lower plenum) when coolant becomes saturated.
(4) The core is divided into stacked several nodes to take into consideration the void distribution in the axial direction.
(5) Empirical correlations on slip ratio for normal coolant flow and slip velocity for slow coolant velocity after recirculation pump trip are used to simulate the void behavior.
(6) Coolant mass change in the separator is calculated from water level, inlet steam and total flows by the correlation obtained from the experimental data.
(7) Recirculation loops can be simulated with or without jet pumps.
(8) Core inlet flow and recirculation flow are calculated on the basis of momentum equation obtained by the analogy with electrical circuits. Other flows in the saturated regions are calculated from mass and energy equations.
(9) First order lag representations are used in order to apply static correlations to dynamic behaviors.

Other characteristics of the code are:
(1) A point kinetic model is used for neutron flux calculation.
(2) In the fuel temperature calculation, the fuel rod is divided into several nodes in the radial and axial direction, gap heat capacity is ignored and the clad is treated as one region in radial direction.
(3) The temporal momentum change of steam flow in main steamline Can be optionally considered.
(4) The control systems fix pressure, feedwater and recirculation flow are simulated.
(5) The thermal-hydraulics in hot channel including MCHFR (Maximum Critical Heat Flux Ratio) can be calculated separately from the calculation of those in whole core.
(6) Emergency core cooling system (ECCS) flow can be calculated.

Typical transients which can be analyzed by BWRDYN are as follows:
I) Pumptrip
(1) Recirculation pump trip
(2) Feedwater pump trip and restart
II) Valve position change or failure
(1) Turbine main stop valve
(2) Turbine control valve
(3) Turbine bypass valve
(4) Relief valve
(5) Safety valve
(6) Main steam isolation valve
(7) Feedwater control valve
(8) Drain level. control valve of BOP model
III) Control system characteristics
(1) Pressure set point change of pressure regulator
(2) Water level set point change of feedwater control system
(3) Flow set point change of recirculation flow control system
(4) Power set point change of recirculation flow control system
IV) Others
(1) Feedwater enthalpy change
(2) Reactivity insertion
(3) Feedwater line isolation of BOP model
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Most part of the reactor system and BOP system are modeled as fixed except for the reactor core region, the main steam line and annunciators. The allowed number of core nodes is less than 25. The number of annunciators is less than 100. The number of steam line nodes is less than 10.
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6. TYPICAL RUNNING TIME

Running time is from one tenth to one twentieth of the simulated transient time using IBM ThinkPad.385D (Pentium MMX 152MHz with 80MB main memory) with Windows 98 SE.
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7. UNUSUAL FEATURES

Simple models adjusted with actual BWR plant records realize a fast program running capability with acceptable accuracy of simulated results. This feature makes it easy to perform various sensitivity studies and predictive simulations of transient including small coolant leakage from reactor vessel.
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8. RELATED OR AUXILIARY PROGRAMS
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9. STATUS
Package ID Status date Status
NEA-1313/01 14-MAR-2002 Tested at NEADB
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10. REFERENCES

- Yokobayashi,M. and Takahashi,Y. :
"Verification Study of Transient  Analysis Code BWRDYN Using Startup Test Data of Tokai Unit 2 BWR Plant"  
ANS Proceedings of the Plant Transient Meeting pp.727-738 (1983)
NEA-1313/01, included references:
- M. Yokobayashi, K. Yoshida and K. Fujiki:
BWR Plant Dynamic Analysis Code BWRDYN
JAERI-M 89-070 (1989) (in Japanese)
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11. HARDWARE REQUIREMENTS
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-1313/01 FORTRAN-77
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13. SOFTWARE REQUIREMENTS

Any operating system on which a compiler for FORTRAN77 and above is available.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHORS

- YOKOBAYASHI Masao, YOSHIDA Kazuo and FUJIKI Kazuo
Japan Atomic Energy Research Institute
2-4 Shirane, Shirakata
Tokai-mura,Ibaraki-ken 319-11
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16. MATERIAL AVAILABLE
NEA-1313/01
\SOURCE directory Source codes
Abstract.doc Abstract in Word format
Data1.d Sample data
Data2.d Sample data
RCPTRIP.TXT Sample data RCP TRIP
NEA_1313_1.PDF Documentation in PDF
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17. CATEGORIES
  • G. Radiological Safety, Hazard and Accident Analysis
  • H. Heat Transfer and Fluid Flow

Keywords: boiling water reactor, hydraulics, thermal analysis.