last modified: 05-MAR-1990 | catalog | categories | new | search |

NEA-0774 THALES.

THALES, Thermohydraulic LOCA Analysis of BWR and PWR

top ]
1. NAME OR DESIGNATION OF PROGRAM:  THALES.
top ]
2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
THALES NEA-0774/01 Tested 05-MAR-1990

Machines used:

Package ID Orig. computer Test computer
NEA-0774/01 FACOM M-380 IBM 3083
top ]
3. DESCRIPTION OF PROGRAM OR FUNCTION

THALES, which stands for "Thermal Hydraulic Analysis of Loss-of-coolant, Emergency core cooling and Severe core damage", is a computer code system for analyzing progression of core melt accident of light water reactors. The code was developed for Level 2 PSA (probabilistic safety assessment) and applicable to a wide range of postulated accident scenarios. Its outcomes are thermal hydraulic conditions in the reactor coolant system and the containment which are necessary for analyzing fission product release and transport behavior during the  accident.

The code system consists of following three member codes:

(1) THALES-PM for accident progression in the primary and the        secondary system of PWRs,

(2) THALES-BM for accident progression in the reactor coolant        system of BWRs, and

(3) THALES-CV for accident progression in the containment of PWRs        and BWRs.

The THALES-PM and the THALES-BM codes carry out two categories of analysis. The first one is overall thermal-hydraulic analysis in the reactor coolant system. The reactor coolant system is divided into multivolumes and each volume is further separated into a liquid region and a gas region by a movable mixture level. System pressure, mixture level in each volume, coolant temperature in each region, flow rate between volumes, etc. are calculated. The other one is core heatup and meltdown analysis. The reactor core is radially and  axially divided into many nodes. Fuel and cladding temperature, cladding oxidation rate, hydrogen generation rate, core melt fraction, etc. are calculated.

The THALES-CV code is for containment response analysis. It divides  the containment into multiple compartments, each of which is further separated into a liquid region and a gas region by a movable mixture level. Containment pressure, mixture level in each compartment, coolant temperature in each region, flow rate between compartments,  etc. are calculated. The code can treat coolant blowdown from the reactor coolant system, combustible gas burning, atmosphere- structure heat transfer, ventilation, containment spray cooling, etc. After the molten core penetrates the reactor bottom head, steam generation, concrete disintegration and noncondensible gas generation are calculated in the reactor cavity or the pedestal.
top ]
4. METHOD OF SOLUTION

Each of the THALES member codes first establishes the steady state conditions after reading input data. Then iterative time-dependent calculation is continued, taking account of various phenomena and events and their interactions which will occur in the course of a postulated severe accident. The transient calculations are iterated by the physical times specified  by input.
Generally the RCS thermal hydraulic analysis with the THALES-PM or THALES-BM code is first carried out and its results are transferred  to the following containment analysis with the THALES-CV code. Then  both results are transferred to a code for analyzing fission product release and transport behavior. Automatic data transfer is possible  in the case the JAERI's ART code is used for fission product behavior analysis.
In overall thermal hydraulic analysis, a new method is adopted aiming at sufficiently accurate estimation of mixture levels in the  reactor coolant system and the containment in a reasonable computer  time. The heat transfer calculation in the core is carried out based on the backward method.
top ]
5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Restrictions relating to storage allocation are:
(1) Maximum number of radial regions in the core : 10
(2) Maximum number of axial increments in the fuel rods : 50
(3) Maximum number of loops in the PWR primary system : 4
(4) Maximum number of volumes in the PWR primary system : 11
(5) Number of BWR recirculation loops: 2 (fixed)
(6) Number of volumes in the BWR reactor coolant system : 7 (fixed)  (7) Maximum number of compartments in the containment : 10
There is another restriction, which relates to time step size. THALES has a lot of models some of which use the forward-explicit method for transient calculations. Even in cases where a more stable method is used in some models, interactions between models are solved with the forward - explicit method. If the time step sizes are too long, the calculated results will be inaccurate, especially  for rapid transient.
top ]
6. TYPICAL RUNNING TIME

Expected CPU time for the analysis from the initiation to the termination of an accident depends on the accident scenario to be analyzed. Typical values for 2000 minutes of transients on the FACOM M-780 computer are:
(1) Large break LOCA with ECCS failure : 20 minutes
(2) Small break LOCA with ECCS failure : 30 minutes
(3) Transient with ECCS failure        : 30 minutes
(4) Transient with RHR failure         : 60 minutes
NEA-0774/01
NEA-DB ran the test cases included in this package on an IBM 3083 computer. The following CPU execution times were required: 1987 secs (THALES-BM); 2047 secs (THALES-CVB); 3301 secs (THALES-PM); 617 secs (THALES-CVP).
top ]
7. UNUSUAL FEATURES OF THE PROGRAM

The THALES code system is usually  used together with the ART code which analyzes fission product release and transport behavior during severe accidents. The objectives of the THALES/ART code package are almost the same as those of the USNRC's Source Term Code Package (STCP). However, the THALES/ART code package has a lot of special features compared with the STCP. For example, the reactor coolant system of PWR or BWR is divided into multiple volumes, molten fuel is relocated downward, void separation is dealt with also in the water in the containment,  and fission product transport carried even by liquid flow is analyzed.
With these models it has become possible to treat some of the phenomena which could not be analyzed easily with the USNRC's Source Term Code Package, including:
(1) Automatic "on" and "off" of ECCS reflecting water level      variation in the BWR vessel,
(2) Fission product scrubbing in the pressurizer when water is      retained there in the case of transient sequences of PWR,
(3) Variation of hydrogen generation rate and fission product  release rate reflecting the variety of meltdown progression      model, and
(4) Flashing of water in the sump or suppression pool followed by  contaminated droplet release to environment when the      containment fails and is depressurized.
top ]
8. RELATED AND AUXILIARY PROGRAMS

THALES uses several program libraries which were developed along with THALES. These libraries are designed to be used also in other computer codes. Their names functions are desribed here:
(1) UCL2 for unit conversion. With this library, all input data      and all results can be given in the user's preferred units.
(2) LOGIC1 for instrument and control system modelling. Most of the  standard logics for LWRs are provided here. Control circuit  failures can be modeled by changing a part of the logics in the  input. Human intervention can also be modeled easily by      describing such actions as control logics.
(3) SPLPACK2 for transient data plotting.
(4) PROPMG1 for steam and noncondensible gas property calculation.
top ]
9. STATUS
Package ID Status date Status
NEA-0774/01 05-MAR-1990 Tested at NEADB
top ]
10. REFERENCES

- K. Abe et Al.
"Overview of Development and Application of THALES Code System for    Analyzing Progression of Core Meltdown Accident of LWR's"
  Proc 2nd Int. Topical Mtg. on Nuclear Power Plant Thermal
  Hydraulics and Operations, Tokyo (1986).
- K. Abe et Al.
  "Development of Computer Code System THALES for Thermal-Hydraulic
  Analysis of Core Meltdown Accidents (1) Outlines of Code System
  and Analytical Models",
  J. of Japanese Nuclear Society, Vol.27, No.11 (1985).
- M. Nishi et Al.
"The User's Manual of THALES Code System for Analyzing Progression    of Core Meltdown Accident (Draft for discussion)",
  JAERI-memo 62-400 (1987).
- K. Abe et Al.
  "Development of THALES-CV2 Code for Analyzing Containment
  Temperature and Pressure Response during Core Meltdown Accident"
  Int. ANS/ENS Topical Mtg. on Thermal Reactor Safety, San Diego
  (1986).
- K. Abe et Al.
  "Thermal Response Analyses of PBF Severe Fuel Damage Tests with
  THALES-MX Code",
  JAERI-memo 59-355 (1984).
- M. Ida et Al.
  "Discussion of Reactor Pressure Vessel Meltthrough Timing in PWR
  Core Melt Accidents",
  JAERI-memo 61-311 (1986).
- K. Abe et Al.
  "Sensitivity Study on PWR Source Terms with THALES/ART Code
  Package and Effects of In-Vessel Thermal-Hydraulics Models",
  Int. SNS/ENS/ANS Topical Mtg. on Probsbilistic Safety Assessment
  (PSA'87), Zurich (1987).
NEA-0774/01, included references:
- K. Kobayashi et al.:
  Digital Computer Subroutine STEAM for JSME Steam Table
  JAERI-M 6967 (31 January 1977) (in Japanese)
- M. Nishi et al.:
  The User's Manual of THALES Code System for Analyzing Progression
  of Core Meltdown Accident.
  JW 219, Draft (October 1987)
top ]
11. MACHINE REQUIREMENTS

About two megabytes of storage are required.
NEA 774/01: To run the test cases on an IBM 3083, about 1,7M bytes of main storage is required.
top ]
12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0774/01 FORTRAN-77
top ]
13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:  FACOM OS IV F4 MSP.
NEA-0774/01
The test cases were run on IBM 3083 under MVS/XA with the VS VS FORTRAN Level 2.3.0 compiler.
top ]
14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

The
UCL2 program library uses "DECODE" and "ENCODE" programs in order to handle unit names given as a character string.
top ]
15. NAME AND ESTABLISHMENT OF AUTHORS

     Kiyoharu ABE
     Risk Analysis Laboratory
     Department of Reactor Safety Research
     Tokai Research Establishment
     Japan Atomic Energy Research Institute
     Tokai-mura, Naka-gun, Ibaraki-ken, JAPAN
top ]
16. MATERIAL AVAILABLE
NEA-0774/01
File name File description Records
NEA0774_01.001 Information file 412
NEA0774_01.002 THALES-BM source 3437
NEA0774_01.003 THALES-BM source 8851
NEA0774_01.004 THALES-BM source 2494
NEA0774_01.005 For unit conversion 243
NEA0774_01.006 For data plotting 5764
NEA0774_01.007 For modeling instrument and control system 1501
NEA0774_01.008 For unit conversion 3726
NEA0774_01.009 For steam table calculation 1586
NEA0774_01.010 For blowdown calculation 869
NEA0774_01.011 THALES-CVB source 6293
NEA0774_01.012 THALES-CVB source 11264
NEA0774_01.013 THALES-PM source 2411
NEA0774_01.014 THALES-PM source 8574
NEA0774_01.015 THALES-CVP source 5064
NEA0774_01.016 THALES-CVP source 10764
NEA0774_01.017 Source for include statement (COMMON) 63
NEA0774_01.018 Source for include statement (COMTBM) 164
NEA0774_01.019 Source for include statement (COMTPM) 165
NEA0774_01.020 LOCF and DATE assembler source 116
NEA0774_01.021 THALES-BM data (BT#ST) 208
NEA0774_01.022 THALES-BM blowdown data (BTIN) 10
NEA0774_01.023 THALES-CVB data (CBT#ST) 170
NEA0774_01.024 THALES-PM data (PS#ST) 276
NEA0774_01.025 THALES-PM blowdown data (PSIN) 10
NEA0774_01.026 THALES-CVP data (CPS#ST) 139
NEA0774_01.027 JCL of THALES-BM 55
NEA0774_01.028 JCL of THALES-CVB 28
NEA0774_01.029 JCL of THALES-PM 49
NEA0774_01.030 JCL of THALES-CVP 29
NEA0774_01.031 Sample output of THALES-BM (by Data Bank) 6720
NEA0774_01.032 Sample output of THALES-CVB(by Data Bank) 4215
NEA0774_01.033 Sample output of THALES-PM (by Data Bank) 7127
NEA0774_01.034 Sample output of THALES-CVP(by Data Bank) 3582
NEA0774_01.035 Sample output of THALES-BM (by authors) 5021
NEA0774_01.036 Sample output of THALES-CVP(by authors) 3695
NEA0774_01.037 Sample output of THALES-PM (by authors) 5427
NEA0774_01.038 Sample output of THALES-CVP(by authors) 2552
NEA0774_01.039 Non tested plotter source ($SPLIDA) 20
NEA0774_01.040 Non tested plotter source ($SPLPLOT) 5668
NEA0774_01.041 Non tested plotter source ($$STHTAB) 2685
NEA0774_01.042 Non tested plotter source ($SYMBOL) 5
NEA0774_01.043 Non tested plotter JCL 18
top ]
17. CATEGORIES
  • G. Radiological Safety, Hazard and Accident Analysis

Keywords: BWR reactors, loss-of-coolant accident, pwr reactors, reactor safety.