NEACRP-A- gxc

TO EZ-PUSLTSZ~ IN TiiE NUCiZ’;Ls CICiXEZXISNG 4XD DESIGX Topic-l.2

,

Analysis of Chernobyl Reactor Accident (I)

. . Nuclear and Thermal Hydraulic Characteristics

and Follow-up Calculation of Accident -

Toshio WAKABAYASHIY, Hiroyasu MOCHIZUKI*,

Hiroshi MIDORIKA\VA*, Yoshitaka HAYAMIZU*,

and Tanemichi KITAHARA*

* Power Reactor & Nuclear Fuel Development Corporation Address: Oarai Engineering Center, PNC, 4007. Narita, Oarai-machi, Ibaraki-ken,

JAPAN.

Abstract

A follow-up calculation was made on the accident of No.4 reactor of Chernobyl

Nuclear Power Plant based on the literatures and accident reports published by the

USSR. The analysis code system used had models peculiar to a pressure tube type

reactor, of which accuracy had been verified by the experimental facilities at 0-arai

Engineering Center and the tests .made at the “Fugen” Nuclear Power Plant. The

analysis data were prepared based ion plant specifications and its operation history

obtained from those published literatures and accident reports.

The analysis was composed of (1) a cakulation of the nuclear and thermal hydraulic

characteristics, and the graphite heating and temperature distributions which were the

basic data for the follow-up calculation of the accident, (2) an analysis of the plant

behaviors before .the test started, using these basic characteristics, and (3) a follow-up

calculation of the power increase which occurred after the test started. The analytical

results were found to agree well with the data published by the USSR.

It was confirmed from these analyses that the main factors causing the accident

were the increased enthalpy at the core entrance caused by the test made at low power

level and the increased void fraction due to reduced coolant flow rate,. in addition to the

nuclear characteristics and performance of the control system peculiar to Chernobyl

Nuclear Power Plant.

1. Introduction.

With the object of making clear the accident of Chernobyl Nuclear Power Plant

NO.4 reactor whic:h occurred on April 26, 1986, an analysis was made on the nuclear and

thermal-hydraulic characteristics and plant characteristics of this reactor before and

after the accident happened. This reactor is a graphite-moderated, light-water-cooled

pressure-tube-type reactor.

Power Reactor & Nuclear Fuel Development Corporation (PNC) has been develop-

ing a heavy-water-moderated pressure-tube-type reactor “ATR”, and has operated its

prototype reactor “Fugen” since 1979 (11. Regarding the nuclear and thermal-hydraulic

characteristics of the core, 0-arai Engineering Center of PNC has developed the

evaluation model,s and analysis codes of a pressure-tube-type reactor through the

experiments by Deuterium Critical Assembly (DCA) having a pressure-tube-type reactor

core, and the burn-out tests using a 14 MW full-scale heat transfer loop. These analysis

methods have been improved in precision, including that of the burn-up characteristics

analysis, by operation data of “Fugen”. Regarding the core cooling performance in loss-

of-coolant accident, the Center has also developed some analytical models and codes by

performing tests at the ATR Safety Experimental Facility simulating a pressure-tube-

type reactor.

The ATR, using heavy-water as the moderator, differs in it’s nuclear character-

istics from a graphite-moderated, pressure-tube-type reactor, but they have a similar

system as a pressure-tube-type reactor. Therefore, we applied the analytical method

developed for “ATR” to the analysis of the nuclear characteristics of the Chernobyl

reactor. The analytical results coincided well with the characteristics published by the

USSR and we could confirm the adequacy of using the characteristics obtained from our

analysis in the follow-up calculation of the accident. The results of the follow-up

calculation based on such reactor core charactersitics were compared with the plant

behaviors at the accident time reported by the USSR to the IAEA on August 25 to 29

111.

The data used in the analysis were prepared based on the literatures [ 3, 4, 5 1 laid

open by the USSR and its accident report to the IAEA (2). -I-

.

2. Analysis code system

Figure I shows the code system used in analyses of the nuclear, thermal-hydraulic

and dynamic characteristics of the Chernobyl Nuclear Power Plant and in the follow-up

calculation of the accident. It further shows the role of each code.

The codes used are outlined below.

2.1 ‘WIMS-ATR code (6, 71

The WIMS-D code is a general lattice cell program that uses transport

theory to calculate flux as a function of energy and position in a cell.~ The basic

cross-section library is in 69 groups with 14 fast, ‘13 resonance and 42 thermal

groups. The transport equation is solved by a collision probability method using

up to 69 neutron energy groups. The WIMS-D gives D, Z a, x rem 2 f and K-

infinitive :for the whole unit cell. The group constants are given in a.few energy

groups suitable for use with other computer codes such as the CITATION.

Besides producing few group cell-averaged constants, point-by-point reac-

tion rates over the entire energy range are calculated for detectors such as

plutonium and uranium.

Evaluation and adjustment to the WIMS nuclear data library for each

uranium and plutonium isotope have been preformed using Fugen operation data

and results of micro-parameter experiments in DCA. The code with new library

was named the WIMS-ATR.

2.2 CITATION code (8 1

The CITATION is a two- and three-dimensional reactor core calculation

code based on the diffusion theory, and the group constant obtained by the

WIMS-ATR code is used. The CITATION code is used in the analysis of the

reactivity coefficient and the control rod worth.

-2-

2.3 ANISN code 191

The ANISN is a one-dimensional transport equation code, which is used to

calculate the amount of heat generated in graphite by neutron and gamma ray.

2.4 TACZD code (101

The TAC-2D code is used to calculate temperature distributions within

graphite rings and the graphite moderator around the pressure tubes at the

steady state and transient. This code can calculate temperature distribution if it

is given the mesh position, the densities, specific heat, thermal conductivities

and heat amount corresponding to the substance within the mesh, and further the

amount of gaps existing between substances contacting each other and the

physical properties of the gas existing there.

2.5 HBALANCE: code

The HBALANCE code can analyze the heat balance in the reactor cooling

system of a pressure~tube type reactor, and can determine the heat balance of a

plant if it is given the operational pressure, flow rate, nuclear fission energy, the

incoming heat from graphite, the heat from pumps, the heat loss from piping, the

feed water enthalpy, etc. The main steam flow rate obtained by this code is

used as the initial condition of the FATRAC, which is a plant dynamic

characteristics analysis code.

2.6 FATRAC code 111, 121

The E’ATRAC code is used to analyze plant dynamic characteristics by

using the basic equations concerned with the change of neutron flux the change

of fuel temperature, the thermo-hydrodynamic behavior of plant components and

operation of control devices. The code can calculate dynamic characteristics of

each plant component taking account of necessary nuclear characteristics and

thermal flow models. In the thermal flow calculation, the flow area is divided

-3-

into both two-phase and single phase areas, which are differently treated to

speed up the calculation. Further control models simulating actual plant control

systems are incorporated there, which can arbitrarily have any characteristics by

selecting the control constants. As the typical systems and components, there

are a power control system, a feed water control system, a pressure control

system and a turbine control system. The precision of the calculation code has

been verified using the results of various steady state and transient tests

conducted in the start-up test of “Fugen”.

2.7 EUREKA-2 code 1131

The EUREKA-2 is a one-point approximation multi-region thermal/

hydraulic combined dynamic characteristics analysis code to analyze the nuclear

and thermal hydrodynamic behavior of a plant at the time of the reactivity

insertion. ‘The code can analyze the reactor transient due to the reactivity

changes caused by abnormal drawing out of control rods, coolant flow or

temperature changes. It is especially suitable for analyiing the reactivity

initiated accident. From the view point of nuclear physics, it determines the

reactor power by solving the one-point approximation dynamic characteristics

equation which assumes that the neutron flux distribution in space does not

change with the passage of time. The coolant’s thermal hydrodynamic behaviors

are analyzed by solving the equation of mass, momentum and energy conserva-

tion by the node junction method, which equation is based on the assumption of

one-dimensional uniform-flow thermal equilibrium flow (EVET).

2.8 SENHOR code (II, 121

The SENHOR code can analyze flow rate, pressure and fuel temperature in

the cooling system of a pressure-tube-type reactor during various transients or

pipe breaks., At the same time, it can also analyze operating characteristics of

emergency core cooling systems. The calculation of flow within the piping is

-4-

.

0;

0 l

conducted hased on the law of conservation of mass, momentum and energy,

using the one-dimensional IVIT with slip model and employing the character-

istics method. Of the thermal hydraulic correlations used in the codes, those

peculiar to the pressure tube type reactor have been developed by PNC and used

in this analysis. Their precision has been verified by full-scale blowdown experiments.

2.9 HEATUP code ( 11, 12 1

The HEATUP code cananalyze temperature and oxidation of each fuel pin

constituting the fuel assembly as well as the temperature changes of the

pressure tube and other components. In the temperature calculation, it can take

the convection heat transfer between fuel pins and coolant, and radiation heat

transfer inbetween fuel pins and between fuel pins and the pressure tube, into

consideration.

3. Analysis of the nuclear and thermal characteristics of the Chernobyl Reactor

3.1 Analysis of ,the nuclear characteristics

3.1.1 Analysis items

The following items, which were important nuclear characteristics in

the follow-up calculation of the accident of Chernobyl Nuclear Power Plant,

were analyzed.

(1) Coolant void reactivity coefficient

(2) Moderator (graphite) temperature reactivity coefficient.

(3) Doppler reactivity coefficient.

(4) Control rod worth.

(5) Composition of discharged fuel.

-5-

.

3.1.2 Analytical conditions

Table I shows the main specifications of the Chernobyl Nuclear Power

Plant, and Fig. 2 shows the core configuration.

The analysis of the coolant void reactivity coefficient, the moderator

temperature reactivity coefficient, the doppler reactivity coefficient and the

control and worth was made using the two-groups constants calculated by

WIMS-ATR, and using the two-dimensional (X-Y) model of CITATION. By the

way, the two-group nuclear constant of the control rod lattice used in the

analysis of the control rod worth was calculated by the multi-cell option

(141 of WIMS-ATR where the control rod lattice was surrounded by 15 fuel

lattices in order to simulate the control rod and fuel arrangement. The core

average ‘burn-up was set at 10.3 CW d/t as published by the USSR and the

burn up distribution in the core was assumed to be uniform.

3.1.3 Analytical results

(1) Coolant void reactivity coefficient

Figure 3 shows the results of the dependency of the coolant void

reactivity coefficient on the coolant void fraction and the burn-up, in

comparison with the values published by the USSR. At the burn-up of

10.3 GW d/t, the calculational value (2.0 x 10e44 k/k/% void) of the

coolant void reactivity coefficient at the rated power (40% void) agrees

well with the value (2 x 10e4d k/k/% void) published by the USSR. The

coolant void reactivity coefficient moves to the positive side following

the burn-up.

-6-

.

0

(2) Moderator temperature reactivity coefficient

Figure 4 shows the calculational values of the moderator tem-

perature reactivity coefficient in comparison with the value published

by the USSR. The calculational value (4 x low54 k/k/oC) at rated power

approximates to the value (6 x 10m5 d k/k/‘C) shown by the USSR.

(3) Doppler reactivity coefficient

Figure 5 shows the calculational value of the doppler reactivity

coefficient in comparison with the value shown by the USSR. The

calc:ulational value (-l.4x10-5dk/k/oC) at the rated power closely

approximates to the value (-1.2 x IO-?lk/k/‘C) announced by the USSR.

(4) Control rod reactivity

Table 2 shows the dependency of the control rod reactivity on the

cool,ant void fraction when all the manual control rods (163 PCS) are

inserted,‘in comparison with the value published by the USSR. The

analysis value (7.2% d k/k) of the control rod reactivity at the rated

power agrees well with the value (7.5% d k/k) shown by the USSR.

(5) Isotopic composition of discharge fuel

The coolant void reactivity coefficient largely depends on burn-

up. Therefore, the precision in calculating the isotopic composition of

fuel burn-up is important for evaluating the dependency of the coolant

void reactivity coefficient on burn-up. The composition fo discharged

fuel (235U, 236U, 239Pu, 240Pu and 241Pu) with a 20 GW d/t burn-up

revealed from the USSR was compared with our calculational values

obta.ined by WIMS-ATR. They agree well with each other, as shown in

Table 3.

-7..

.

It was recognized from these results that the WIMS-ATR/CITATION

code system could closely reproduce the nuclear characteritics of Chernobyl

Nuclear Power Plant. The analytical results of nuclear characteristics were

used as the input data to analyze the plant dynamic characteristics and the

accident.

3.2 Graphite temperature analysis

3.2.1 Analysis .items

The graphite moderator generates heat by neutron moderation and

gamma-r;xy, and the heat is transfered to the coolant through the pressure

tubes. Therefore the estimation of the heat transferred from graphite to the

coolant is important as it is used as a boundary condition for the analysis of

the plant: behavior. The following items were analyzed from this point of

view.

(1) The amount of the heat genarated in the moderator.

(2) Temperature distributions within the moderator at the rated power.

(3) Change of the heat transferred from the moderator to coolant consider-

ing the power change.

3.2.2. Analytical conditions

In the analysis, the square columnar moderator around the pressure tube

is treated as a cylinder having the same volume. Further, there is a I m

thick reflector around the reactor core of 11.8 m in diameter and 7 m in

height, and a 0.5 m thick reflector at the top and bottom of it. With such a

shape of the core taken into account, the amount of generated heat was

analyzed by ANISN, the steady and transient temperature distributions by

TAC-2D and the change of heat from the graphite over many hours by

HEATUP.

3.2.3 Analytical results

According to the literature (151, the maximum temperature reached

one year after the start of operation is estimated to be 73O’C. Therefore,

the gap of graphite rings and their gap conductance were determined by the

results of the TAC-2D calculation, using the amount of generated heat

calculated by the ANISN, so that the maximum temperature might become

73O’C. Figure 6 shows the results of analyses of the temperature distribu-

tions at rated power and 6% power using the analytical result for the amount

of heat obtained by ANISN. The amount of heat transfer between the

graphite and coolant in one day from the start of power decreasing (from I:00

of April 25) was analyzed by HEATUP. Figure 7 shows the result, which

indicates that the heat of 84MW has been discharged to the coolant just

before the test started.

-9-

4. Analysis of plant behaviors just before the accident

4.1 Analytical model

The analysis of the plant behaviors just before the accident was conducted

using the FATRAC code. The circulation systems handled by this code include,

the turbine system, the feed water system and the reactor purification system,

in addition to the primary cooling system. The code assumes that the primary

system is made of two loops, each loop including one hottest channel and 845

average channels. Regarding ~the size, etc. of the primary circulation system, we

followed the USSR’s report presented at the IAEA ( I!, and the literatures ( 2,

3, 4, 5 ) published by the USSR. Since no detailed information could be obtained

about control system, we assumed it to be the same as the control system in

ATR and analyzed plant behavior by using the model shown in Figure 8.

4.2 Analytical conditions

The ‘analytical results of nuclear characteristics in chapter 3.1 were used

as the various reactivity coefficients and dynamic characteristic parameters

which were necessary for the analysis of the plant behaviors. The analysis was

conducted under the following initial conditions from I:19 taken as the starting

point.

Drum separator pressure:

Drum water level:

Recirculation flow rate:

Main steam flow rate:

Feed water flow rate:

6.9 MPa

ordinary level - 600 mm

42,000 ton /h, 8 pumps

3: 7% of the rated flow rate

(5,800 ton /h)

Y 6% of the rated flow rate.

-lO-

The main steam flow rate was determined by the heat balance analysis

assuming that the heat generated by nuclear fission was 200 MW, heat transfer

from the graphite was 84 MW, the pump heat generation was 28 MW, the heat

loss from piping, etc. was 31 MW and the feedwater enthalpy was 200 kcal/kg.

Concerning the operation history, the feed water flow rate published by the

USSR was used.

4.3 Analytical results

According to the announcement by the USSR, the operators increased the

feedwater flow rate at 1:19 when the drum water level was found to be far below

the alarm level, and reduced it again 3 minutes later. As a result, as shown in

Fig. 9, the drum water level increased, but it caused pressure drop from 6.9MTa

to 6.3MPa, showing the tendency of rising again before 1:23:04 when the test

started.

The analytical results in figure 9 obtained by the FATRAC code have

predicted those values reported from the USSR with good precision.

- II -

.

5. Follow-up calculation of the accident

5.1 Analysis of the reactivity accident by EUREKA-2

5.1.1 Analysis items

EUREKA-2 was employed to analyze the changes in the power and

reactivities (the total reactivity, the doppler reactivity, the coolant void

reactivity and the control reactivity with automatic control rods (AC)

inserted), the enthalpy stored in fuel etc. at the Chernobyl accident. The

analytical results for the plant behavior (the changes in the entrace flow

rate, steam drum pressure, entrance ehthalpy, etc. with the passage of time)

before the accident obtained by FATRAC were used for the EUREKA-2

calculation.

5.1.2 Analysis conditions

(1) 1nit:ial analysis conditions

(a) Thermal Power: 200 MW

(b) Coolant flow: 42,000 t/h

(c) Peaking factor (the calculated values by WIMS-ATR/CITATION)

Radial peaking factor (RPF): 1.5

Axial peaking factor (APF): 1.3

(2) Plant characteristics

(a) Dynamic characteristics parameters:

Beff = 0.0057

e= 6.3 x 10-3 set

(The values were calculated by WIMS-ATR.)

(b) Coolant void reactivity coefficient:

Figure 3 was used.

- 12 -

(c) Control reactivity:

The reactivity corresponding to the inserted length of AC control

rod was obtained from the control rod S-shaped curves

(d) Changes in oressure, coolant flow rate and entrance enthalpy:

The calculated values by FATRAC were used.

(3) Analytical model

The object of analysis was limited to the core part,. which was radially

divi.ded into 4 regions (number of the hottest channels . . . 16, number of

the hot channels . . . 335, number of average channels . . . 654, and number

of ~1.0~ power channels . . . 654), with each region axially divided into 8

reg.ions.

5.1.3 Analytical results

Figure 10 shows the thermal power normalised at 200 MW.

Figure 11 shows the changes in the recirculation flow rate. Figure 12.

shows the total reactivity, the void reactivity, the doppler reactivitiy and the

control reactivity.

The following have been made clear.

(1) The peak power and the time when it occurred agree well to the values

(that of the first peak) published by the USSR.

The peak power is about 150 times greater than the rated power (100

times greater, as announced by the USSR).

The time when it occurred is 23 minutes 44 seconds past one (23

minutes 44 seconds past one, as announced by the USSR).

(2) The maximum enthalpy stored in fuel is about 370 Cal/g. (More than

300 Cal/g as announced by the USSR).

- 13 -

5.2 Analysis of the primary cooling system behavior

5.2.1 Analysis items

The primary cooling system behavior when the accident occurred, such

as the pressure, flow rate, etc. in the system, was analyzed by the SENHOR

and HEATUP codes using the analytical results (the reactor power) obtained

by EUREKA-2. The analysis was conducted from the time of 1:23:04.

5.2.2 Analytical conditions

(11 The initial conditions

(a) Thermal power: 200 MW

(b) Drum-type separator pressure: 6.5 MPa

(c) Re-circulation flow rate: 42,000 t/h

(d) Feed water flow rate: 342 t/h

(e) Feed water enthalpy: 200 kcal/kg

(2) The transient conditions

(a) Recirculation flow rate and feed water flow rate were used the

information reported by the USSR. Feed water temperature was

calculated by FATRAC.

(b) Turbine bypass valve flow rate: 0 t/h

(c) Safety valve capacity: 5,800 t/h

(3) Nuc:lear characteristics

The values obtained in Chapter 3.1 were used for the coolant void

reactivity coefficient, the doppler reactivity coefficient and peff.

- 14 -

5.2.3 Analytical results

Figure 13 shows the analytical results for the drum-type separator

pressure (and the flow rates at the entrance and exit of the reactor core. As

can be seen from the figure, the drum-type separator pressure rapidly

increases after the power increase and reached PMPa at 1:23:47. By rapid

generation of steam in the reactor core, the coolant was divided into the up

flow and the down flow, and its supply to the core ceased.

On the other hand,. as shown in Figure 14, the fuel temperature when

the accident occurs rises after the rapid reactor power increases. The fuel

dried out at 1:23:44 and, three seconds plater, reaches l,SOO’C, the meltdown

temperature of the fuel claddings. This analysis was made in an assumption

that the fuel pellets and cladding tubes maintained their original forms.

Acc:ording to the analytical results by the EUREKA-2 code in the

preceding: section, however, the enthalpy stored in fuel increased to 370 Cal/g

UO2, and we can suppose, judging from .the results of tests on the nuclear

safety research reactor (NSRR in JEARI), etc., that the pellets and claddings

were melted and crushed at the time point of power increase, generating the

impact pressure(l61.

On ,the basis of the present analyses, we think that this accident

occurred by the combined effects of such causes as (1) the large positive

coolant void coefficient, (2) that the test was conducted under the low power

conditions where steam void could be easily generated, and (3) that the

control rods were inserted too slowly (0.4 m/set).

- 15 -

6. Conclusion

We analyzed the nuclear and thermal hydraulic characteristics, the graphite

temperature, etc. for the follow-up calculation of the accident of Chernobyl Nuclear

Power Plant, using the ATR design and design evaluation codes. Our analyses were based

on the plant specifications and system data described in the literatures and accident

reports published by the USSR. The analysis results have agreed well to the fragmental

characteristic values published by the USSR and we could confirm the propriety to use

the various characteristic values obtained from our analysis in our follow-up calculation

of the accident.

On the basis of these characteristics obtained, we made an analysis of the’ plant

behaviors before .the test started and a follow-up calculation of the accident, and the

results have generally coincided with the values pubiished by the USSR.

By the above analysis, we have obtained the nuclear and thermal hydraulic

characteristics and plant characteristics for examining the cause of the Chernobyl

Nuclear Power Plant accident and considering the.measures to prevent similar accidents,

and we have confirmed the applicability of our analysis models and codes.

- 16 -

References

(11

(21

(31

(41.

(51

(61

(71

(83

(91

S. Sawai et al., Nucl. Eng. Int., 24 (1979) 33.

The Accident of Chernobyl Nuclear Power Plant and its Consequences, Information

Compiled for the IAEA Experts’ Meeting, 25 - 27 August 1986, Vienna.

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from Atomnaya Energiya, Vol. 43, No. 4, (1977), 235.

A.P. Alkersandrov et al., Physical and Power Startup of the First Unit of the V.I.

Lenin Nuclear Power Station, Leningrad, Translated in English from Atomnaya

Energiya, Vol. 37, No.2, (1974) 99.

Soviet Power Reactors 1974, ERDA-2, Report of the United States of America

Nuclear Power Reactor Delegation Visit to the Union Soviet Socialist Republics,

Sep. 19 - Ott; I, 1974.

T. Wakabayashi et al., Characteristics of Plutonium Utilisation in ATR, NEARRP-

A744 (1985).

J.R. Askew let al., J. of British. Nucl. Sot. 5 (1966) 564.

J.B. Fowier et al., ORNL-TM-249~6 Rev.2, USAEC July (19711..

W.W. Engle Jr., A User Manual for ANISN, CCC-Z!, RSIC Computer Code

Collection, ORNL (1976)

(101 S.S. Clark et al., GA-8868, USAEC, September (1969)

Cl 11 H. Kato et al., The Hitachi Hyoron 62 (1980) 724.

(121 Y. Hayamizu et al., The Hitachi Hyoron 67 (1985) 905.

(131 N. Ohnishi et al., EUREKA-2: A Computer Code for the Reactivity Accident

Analysis in a Water Cooled Reactor, JAERI-M84-074, (1984).

IIU! T. Wakabayashi et al., Power distribution and control rod worth in heavy-water-

moderated, cluster-type plutonium fuelled core, 27th OECD NEA-CRP, Aix-en-

Provence, (1984).

cl51 N.A. Dollez:hal’ and I.Y. Emel’ yanov, The Pressure - Tube Reactor, Moscow

Atomizdat (1980). (in Russian)

- 17 -

(16) M. Ishikawa and 5. Shiozawa, A Study of Fuel Rehavior under Reactivity Initiated

Accident Conditions - Review, J. Nucl. Hater. 95 (1980) 1.

- 18 -

T’ab le 1 Data of Chernobyl Reactor

rharmal Power

SueI Assembly

Number of Fuel Elements

Diameter of Fuel Elements

Diameter of Fuel Pellets

Fuel Enrichment

Mass of Uranium

Core

Fuel Burnw

Core Average Burnuo

Number of f’uel Channels

Number of Control Rods

Core Diameter

Core Heigh,t

Lattice P i tch

Vm/V f

Reactor Cool ins System

Flow (Rated Powei)

P ressure ( f?ated Power)

Inlet Temparature (Rated Power 1

Outlet Temperature (Rated Power)

Outlet Quality (Rated Power)

3200MW

t a .1 3.6mm

1 1.5mm

z.Owt%

1 1 4.7 kg

2.0 GWd/ t

1 0.3 GWd/t

1659

211

11.8m

7.0 m

250mm

3 0.1

37,5OOt/h

7MP a

27O’c

284’CCSaturated)

14%

Table2 Dependence of Control Rod Worth on Coolant Void Fraction at the Chernobyl Reactor ( Number of Manual Rods***164, Burnup.- 10.3 GWd/t >

Coolant Present Results Ini’ormat ion Reported by the

Void Fraction (%dK/K) USSR(% AK/K)

0 6.5

20 6.8

.40 7.2 7.5

60 7.6

8~0 8. 1

‘* 1 0 0 8.7

. .

Table3 Isotopic Composition of Discharged Fuel at the Chernobyl Reactor (‘Burnup -.- 20 C;Wd/t )

Nucl ide

235 u

231; U

2nnp L,

2,J”P u

2.4’ P LI

P resent

R es 14. I t s

4.8

2.5

2.4

1.8

0.6

(kg,‘tHM)

lnformat ion Reported by the USSR

4.5

2.4

2.6

1.8

0.5

I .

08

. .-.

<Nuclear Analysis>

JWIMS-IITRj-jZi-

Void Coefficient. Beff.L, etc.

<Heat Generation Analysis >

Heat Generation Distribution of Graphite of Graphite

TaIIW.

<Heat Balance Analysis>

pi-

Dynamic Analysis Accident Analysis

(Plant Dynamic Analysis>

Flow. Pressure, Level. Inlet Enthalpy. etc.

<PIant.Oehavior Analysis>

Pressure. Flow. etc.

l=ue I Tem~., Graphite Temp.,. Pressure Ture Temp., etc.

Figure] Code System for Analysis of Chernobyl Reactor Accident

Number of Detectors

8 DMEH 12 * : L_ocaI &tomatic sontrol Rod l Calibration T Chamber 225 ** : Local Emergency Protection

@ DMER 12 Rod

9 Detector fok LAC?LEP?System 48

@ Fission Chamber 4

Figure2 Core Configuration in’ the .Chernobvl Reactor

0 10.3GWd/t

A 15GWd/t ? I

I

// 1 SGWd/t ’

I’

/' , _A0 ’

L m-m- .&---

J 0.3 GWD/ t

0 Information Reported by the USSR

20 40 60

Void Fraction (% 1

80 1 0

Figure3 Dependence of Coolant Void Reactivity on Burnup at the Chernobyl Reector

e

0 Information Fiworied by the USSR

450 550 650 ‘5( I

Graphite Temperature (‘c)

Figure 4 Dependence of Graphite Temperature Coefficient on Graphite Temperature at the Chernobyl Reactor (Burnup **+ 10.3 GWd/t >

.

.

0

a SJ ? e Information Fiarorred by the

? USSR P

‘p z x -l.O-

00

Temperature ( C 1

Figure5 Dependence of Doppler Reactivity Coefficient on Fuel Temperature at the Chernobyl Reactor (Burnup--+10.3GWd/t >

0 s t

l ‘F

294 a-

Pr SL

/

-

rre

, 5

/ 3

/3

Graphite Ring

at Rated Power

Gas Layer

I ! I

_) 380

,.A’

54

Grzghita

Temperazure et Experiment

15

55.5 57 141.1

Distance form Center of Pressure tube (ur,nn)

Figure 6 Steady State Temperature Distribution in

the Coreof Chernobyl Reactor

. ’

3-

Itleat Generot ion Rate 0

!$j I I Power I I ‘, I I I \

50 I -L--- ;- - - _ :

I- -2 o’-----7- ---- , ___;_--- L-

I E -~ I I

I .-

5 - I I I

I I I I

; I I I I I I

I I I D 0. ------;------’ I__-__--_~_ _- ____ ;-- ----

I I I I

I # I I I

I I

I I I I

L 2 am

Time (I-lair)

Figure7 Heat Generation Rate History in Graphite from the Day before the

,Accident of Chernoby I Reactor

Neutron COUIXIX

(Turbine Control System) c - - - -

I-

Turbine Bypass Valve

!iO%X4

Feed Water Fecad Wator l:low Motor Control

L- -

Condl!rlsed Wa1or Pllrnr,

Figure 8 Plant Control Model of Chernol~yl Reactor

I

I ln,Format ion I- ---Reported by 8 the USSR I , - Calcalated I by FATRAC

I 1

!2

F---- - ___- -- -_----- -,-, --_-_---.-..-- - - --- --- -9 4 -cl

i I i

1:19 20 21 22 23

Time (min)

Figure 7 Plant Behaviar from 1: 17 to the Time before Experiment

I_ I_ b - b - I I I I I I I I I I I I I I I I I I

r ---------- I n.Format i on Reported by the USSR ---------- I n.Format i on Reported by the USSR

Calculated by EUREKA-2 it 1 I

“0 - f I’:

I I ’ I I ‘, I

“0 - :

i i I I I I I I IA IA I I I I I I I I ” ” 1:23:4 1:23:4 9 9 I4 I4 19 19 24 24 29 29 34 34 39 39 44 44 49 49 54

Time ( set) Time ( set)

Figure10 Neutron Power at the Accident o,f Chernobyl Reactor

. ~ 0

----------- ---=------

--O--==-~--- - a-

:- ,

0 5 I I ILL------J

14 l!l :4 2b 34 30 44 40 1:24:Oam Time (set)

I I I I I I I I I

-----, information Reported by the USSR

p resent Resu I t

Figure 11 Main Circulation Flow at the Accident of Chernobyl Reactor

. 7

. . I .

I _,

Void Fieactivi ty .I.

c

Tota I Reactivity

,\ Control Reactivity

\ ‘. Dow I er React ivi

Time (set)

Figure12 Reactivity Changes at the Accident 0.f Chernobyl Reactor

, I I .

c

,

.

I I - Colculatod by SENllOR

Yo Y- .__- --------L--------f-----. -I

- -- I > 1 1 IM- t

-Ii= I i 1 -d .---- ------I---------+------.---- I I

Time ( set )

Ficlure 13 Time History of Drum-Type Separator Pressure and Core Inlet and Outlel Flow Rate at the Accident of Chernohy I Reactor

.

;

L :i--.-------+--------

AZ I I I

I

Pel let center Y

.------,

21 adding .

if --- r

d 1

J

i

- 49

,------

%I

<

/ =-v

s/ 3 ressure Tube

I _i_

I I I I

I I

2 I z 5 + z---- ----_-- L---.-i.-

0 I I 1

.+

II c 14 1:23:4 19

Time ( set >

Figure14 Time H istory of Fuel Pet let, Cl adding and Pressure Tube Temperature at the Accident of Chernobyl Reactor

.

.