NEACRP-A-973

Critical Experiments on Initial Loading Core of

Very High Temperature Reactor Critical Assembly(VHTRC)

Fujiyoshi Akino, Tsuyoshi Yamane,

Hideshi Yasuda and Yoshihiko Kaneko

Department of Reactor Engineering,

Japan Atomic Energy Research Institute.

Tokai-mura.Naka-gun,ibaraki-ken,Japan

September 12,1989

Abstract

A critical assembly VHTRC(Very High Temperature Reactor Critical

Assembly) was constructed by modifying the SHE(Semi-Homogeneous

Experiment) assembly in order to verify the nuclear design accuracy

of HTTR(High Temperature Engineering Test Reactor). The construction

was started on May 1983,and the initial criticality was achieved on

May 13 ,1985.

The initial core named as the VHTRC-1 core was made by loading fuel

rods which contained fuel compacts of the coated particles of the 4%

enriched uranium. The following measurements were carried out on

the VHTRC-1 core:

1) critical mass,

2) reactivity worths of HTTR mockup control rod and burnable

poison rod,

3) neutron flux distribution,

4) temperature coefficient of reactivity,

5) kinetic parameter 8,/,/A.

Calculations were performed with the SRAC code system using the

nuclear data based on the ENDF/B-IV. The agreements between calcula-

tions and experiments were fairly good for most experimental items

except the value of kinetic parameter. The results obtained satisfy

the accuracy requirements for fundamental nuclear design of HTTR.

-I- i518000?

I. Introduction

Since 1969,research and development for a High Temperature

Gas-Cooled Reactor(HTGR) have been continued at the Japan Atomic

’ Energy Research Institute(JAER1). In 1987, the Japanese Atomic Energy

Commission issued the revision of the Long-Term Plan for development

and utilization of the nuclear energy. The revision plan specifies early

construction of a High-Temperature Engineering Test Reactor(HTTR)”

proposed by JAERI, aiming to achieve criticality in 1995.

A core of HTTR comprises the prismatic pin-in-block type fuel

blocks using the uranium of 6% enrichment on the average. Such a core

structure is different from both of the pebble-bed type reactors, the

Arbeitsgemeinschaft Versuchsreaktor(AVR) and the thorium high tem-

perature reactor(THTR) in the Federal Republic of Germany(FRG) and

the multihole-type reactor, the Fort St.Vrain(FSV) reactor in the

United States.

All these reactors except the HTTR are based on the highly enriched

uranium and thorium fuel cycle. But, since the International Nuclear

Fuel Cycle Evaluation(INFCE) conferences, the low-enriched uranium

rue 1 cycle has been selected for HTGR, for non-proliferation

consideration. Therefore, such a situation enhanced unique progress

in the reactor physics study related to HTTR

Reconstruction program of the SHE(Semi-Homogeneous Experiment)”

to the VHTRC(Very High Temperature Reactor Critical Assembly) was

approved by Japanese government early in 1981. The object of the pro-

gram is to obtain experimental verification for the nuclear design

accuracy of HTTR with low enriched uranium fuel. The construction

was started in May 1983, and the initial criticality was achieved on

May 13,1985.

This paper describes the outline of experimental results on the

initial loading core,VHTRC--1,which is loaded with the coated particles

of 4% enriched uranium. The experimental items are the critical mass,

the reactivity worths of HTTR mockup control rods and burnable poison

rod, the neutron flux distribution, the temperature coefficient of

reactivity and the kinetic parameter. The analyses were performed with

the SRAC code system using the nuclear data based on the ENDFIB-IV data

file.

-2- 1518UoO2

. .

II. Critical assembly

The assembly is a horizontally-placed hexagonal prism in the shape.

made of graphite blocks and can be split into two half assemblies,

one of which is .fixed and the other is movable. The assembly has the

across flat of 2.4m and length of 2.4m. Each graphite block(across

flat:30cm; 1ength:lZOcm) is also the hexagonal prism with 18 holes

for fuel rod insertion and a hole for control or safety rod insertion.

The prismatic blocks are made of isotropic graphite containing very

low impurities of CO.1 ppm equivalent boron content. The whole core

covered with the insulator for heating up to 210°C using the electric

heaters. Outlook of VHTRC3’ is shown in Fig.1.

The driving speed for the movable half is slowed down near the

closing position. The fast-drive speed is 50 cm/min.from 150 cm tb

20 cm separations, and the slow-drive is 1.5 cm/min.from 20 cm

separation to the closed position. The distance between the two halves

is indicated by three position indicators; one is an analogue indicator

and the other two digital indicators. The position of the movable half

at the closed position is found to be settled within about 0.05mm

relative to the fixed one.

Two control and six safety rods driving mechanisms are mounted on

the supporting frames behind the two halves. These control and safety

rods are made of cadmium cylinders covered with stainless steel.

A radium-beryllium neutron source of 5OOmCi is used as a start-up

neutron source and is inserted by the driving mechanism from its coffin

through a guide tube to the radial reflector region in the fixed half.

A fuel rod consists of a stack of 20 fuel compacts packed in a

graphite sheath 732mm long, 5mm thick and 47mm in the outer diameter.

The fuel compact is the hollow cylinder of the BISO-coated particles

(2% and 4% enrichment) or the TRISO-coated particles(6X enrichment)

dispersed in the graphite matrix. The physical constants of the fuel

compact and the’graphite assembly were described in Ref.3.

III. Experiments

1. Critical mass

(1)Initial critical approach at room temperature

The initial core named BS the VHTRC-1 core was made by loading fuel

\ .

rods, each containing 20 fuel compacts of the 4% enriched uranium and

the core had radial and axial reflectors.

The core was loaded with the fuel rods step by step, keeping the

hexagonal symmetry. The neutron multiplication was observed at each

loading step by the source multiplication technique using six neutron

detectors. Four BF3 counters of 8mm diameter were distributed in the

core region. One pair of counters were placed in the fixed half and

the other pair in the movable half and two BF3 counters of 2.54cm

diameter were located outside of the fixed half.

After the step by step loading, the core reached criticality with

282 fuel rods at the ninth step, with all safety rods being fully

withdrawn and control rods being partially inserted. The fuel loading

pattern of this core is shown in Fig.2.

The reactivity worth of control rods was calibrated by the period

method. From the results of calibration, the excess reactivity was

determined to be 28.4*0.6 cents. The reactivity worth of a fuel rod

was measured with the calibrated control rods and found to be

7.11*0.26 cents. Using this fuel rod worth, the core loading of 282

fuel rods was corrected for the excess reactivity and for the reactiv-

ities of inserted materials” in the core, such as the safety and control

rods insertion holes, the gap between the fixed and movable half

assemblies, four in-core BF3 counters, etc., and also the temperature

difference between experiment and calculation. The number of fuel rods

measured at the critical state was obtained to be 269.7+0.6(4%

enriched) which gave the critical mass of 4.520*0.010 kgzz5U at 27’C.

At four loading steps in the initial critical approach, the effective

multipilication factor,k,,,. was measured by the pulsed neutron source

technique(PNS) using the four in-core BF3 counters. The data were

analyzed by means of the simple method based on integral version of

the SjGstrand’s area-method. This method d t e ermines the static reac-

tivity from one-point measurement by using 8 correction factor for

the value of area-ratio observed at the measuring position.”

The target of PNS(KAMAN A-801) was attached to the center of the

back-surface of the fixed half. Four. BF3 counters of 8mm diameter were

distributed in the core region. The data were collected using a mul-

tichannel analyzer(ORTEC 7010/7000-37) with the four independent input

in the multichannel scaling mode. The area.s of prompt and delayed

neutrons were obtained by fitting the decay curves using the ALPHA-D

code6’. This code determines prompt neutron decay constant taking

account of the decay of delayed neutrons.

The measured values of k,,, agree fairly well with those predicted

by a three dimensional diffusion calculation, as shown in Fig.3. This

result confirms that at early loading steps the neutron mulitiplication

would increase as expected.

(2)The heated core at 200°C

After the initial critical approach at room temperature, thk whole

core including the reflector was heated up to ZOO’C by using 30kw

electric heaters located in the radial reflector. The 14 K-type ther-

mocouples were placed in the core for measurement of the temperature

distribution. The average core temperature reached ZOO’C by about 50

hour continuous heating. and the spatial standard deviation converged

within 1.3’C.

At the temperature, about ZOO’C, the number of fuel rods was

increased stepwise to approach the criticality. The 2% and 4% enriched

fuel rods were added to make critical again. Then, the heated core at

200.8”C reached criticality with 453 fuel rods(4% enriched:309, 2%

enriched:144). The fuel loading pattern of this core is shown in Fig.4.

After correcting for the reactivities of a.11 inserted materials ‘)

including the reactivity of the heaters in ~the core. the number of the

fuel rods measured at the critical state was obtained to be

432.3*1.2(4% enriched:288.3*1.2, 2% enriched:144) which gave the

critical xnass of 6.043*0.019kg235U at 200.8”C.

2. Reactivity worth of HTTR mockup control rods

Reactivity worths of the HTTR mockup control rods(CRs) were

measured in the VHTRC-1 core with the pulsed neutron method.

A CR consists of two semi-rods, each having 46 neutron absorber

pellets(outer diameter:89.&nm: inner diameter:60.Omm; height:25.2mm

and boron content:38.3wt%) stacked in a stainless steel tube(outer

diameter:97.5mm; thickness:2.Omm; length:1150mm). The absorber

pellets are made of BaC particles dispersed in graphite powders.

Measurements were made for following five core patterns:

OC : The core was loaded with’274 fuel rods(4% enriched),without

the CRs.

1C : The 1C core was constructed from the OC core by inserting a

CR in the central column. The position of the CR was 9.5cm

radially distant from the core center. The arrangement of CR

-j- 15180005

is shown in Fig.5.

2C : The 2C core was constructed from the OC core by inserting two

CRs in the central column. The positions of the CRs were

9.5cm radially distant from the core center. The distance

between two adjacent CRs was 16.5cm. The arrangement of CRs

is shown in Fig.6.

2R : The 2R core was constructed from the OC core by inserting two

CRs in the reflector column. The distance between two

adjacent CRs was 16.5,cm. The arrangement of CRs is shown in

Fig.7.

12R : The 12R core was constructed from the OC core by inserting

twelve CRs in the reflector columns. A pair of CRs was

located in each of the 6 reflector columns. The distance

between two adjacent CRs was 16.5cm.

The arrangement of CRs is shown in Fig.8.

To determine the prompt neutron decay constants, the raw data were

fitted by the least-squares method to be a single exponential type

formula using the ALPHA-D code. The reactivity worth was determined

by measureing the prompt neutron decay constant with the revised

King-Simmon’s pulsed neutron method(PRKs)7). The measured values of

the reactivity worth of the CRs are listed in Table 1. In the case of

2c core, correction due to the change of neutron generation time in

the reactivity of revised King-Simmon’s formula was about 24% of the

reactivity of King-Simmon’s formula’). The neutron generation time

was obtained by the core calculation for each core pattern using the

three dimensional diffusion theory.

The reactivity worth for the case of 1C core. moreover, was measured

by the space integral version of the Sjostrand’s area-ratio method

(7s~). The areas of prompt and delayed neutrons were obtained by fit-

ting the decay curves using the ALPHA-D code. The reactivity worth

was measured placing the BF 3 counters in 12 fuel blocks of the fixed

and movable assemblies and 48 measured values were obtained. The

results are shown in Fig.9. The minimum and maximum values of the

experimental results were 13.85 and 28.0$. In analyzing these values

by a space integral method, the core volume was divided into

sub-volumes of 48, so that each contains one detector position. The

weighting function for the integration over the each sub-volumes was

assumed to unity. As the result, the reactivity worth of the CR was

obtained to be 18.6*0.4$.

-6-

. .

The measured reactivity worth determined by the space integral

method agreed well with that by the revised King-Simmon’s pulsed

neutron method for the case of 1C core.

3. Reactivity worth of HTTR mockup burnable poison rod

An HTTR mockup burnable poison rod(BP rod) consists of two

semi-rods, each having 20 neutron absorber pellets packed in a hollow

graphite rod. The absorber pellets 8x8 made of B& particles dispersed

in graphite powders. Three types of the BP rods have .been prepared.

They are different in the boron content in absorber pellets, the

diameter of the absorber pellets and grain size of B& particles. The

specifications of the absorber pellets are shown in Table 2.

For the measurement of reactivity worth of the BP rod, the graphite

block of the central column of the VHTRC-1 core was replaced by the

fuel block loaded with 12 fuel rods. The BP rod was inserted into the

corner of the prismatic fuel block of the central column. At first,

in the experiment, to obtain the critical state without the BP rod,

the number of loading fuel rods and the insertion depth of control rods

were adjusted. Next, to determine the BP rod worth, the core with

BP rod was brought to the critical state again by adding the several fuel

rods and adjusting the insertion depth of control rods. The BP rod worth

WBS obtai:ned as the sum of the added fuel rod worths and the compensated

worths of the control rods. The reactivity worth of a fuel rod was

obtained from the control rod worth curves which were measured by the

period method. The measured values of the reactivity worth of the BP

rod are shown in Table 2.

4. Neutron flux distribution

Distributions of the 63C~(n,7)6’C~ reaction rate in the radial and

axial directions were measured in the VHTRC-1 core at room temperature

and at 200°C by the activation technique.

The core was loaded with 282 fuel rods(4% enriched) at room tem-

perature and 453, fuel rods(4% enriched:309, 2% enriched:144) at 200°C.

Natural Cu foils and wires were prepared as activation samples. The

foils were used for the measurement of the radial,i.e., rod-wise

reaction rate distribution. Each foil had a thickness of O.lmm and

inner and outer diameters, 18 -and 36mm each, same as the fuel

compacts, and was put between compacts in every fuel rod positions of

S-3 and C-6 blocks, as shown in Fig.10. The axial position was 113mm

-l- 15180007

. ‘

in the fixed half from the midplane facing the movable and fixed half

assemblies. The axial reaction rate distribution was measured using

25 wires of 2mm diameter and 1Omm long, which were set on an aluminum

wire at the intervals of 50mm and were inserted at the position denoted

by “A” in Fig.10. The induced act-ivities of each samples were measured

with a Ge detector and a GM counter for the foils and the wires,

respectively. Experimental technique for measurement of the reaction

rate in the heated core at 200°C WBS the same as that at room tem-

perature except the following two points: (i) samples were placed at

every fuel rod positions of S-3,C-6 and D-2 blocks radially and along

graphite rods axially at the positions indicated as “A” and “B” in

Fig.ll., (ii) a NaI(T1) detector (03”x3”) was employed instead of a

Ge detector and a GM counter. The measured results of the radial and

axial distributions are shown in Figs.12 and 13.

5. Temperature coefficient of reactivity

After measuring the critical point at room temperature, the whole

core including the reflector was heated up by electiric heaters. This

core was constructed by loading with 300 fuel rods(4% enriched:288, 2%

enriched:lZ), four in-core BF3 counters and 40 electric heaters.

At four steps of the temperature 70,100,150 and 200°C. PNS experi-

ments were carried out to measure the subcriticality of the core due

to the temperature rise. ‘) The arrangement of the target of PNS and

BF3 counters was same positions BS the PNS experiment described in

Sec.1 in this chapter. The data were analyzed by the simple method

based on the integral version of the Sjostrand’s area-method.

Figure 14 shows the reactivity change due to temperature rise, where

the solid line was obtained by the least-squares fitting for the three

measured values between 1OO’C and 200°C. The measured values of the

temperature coefficient of reactivity are shown in Table 3.

6. Kinetic parameter B.,j/A

Kinetic parameter Be,,,/& is expressed with the prompt neutron decay

constant,a,,at critical, through the well-known inhour-equation:

where

oc is prompt neutron decay constant at critical,

0

0

-&

a

,li is the decay constant of the i-th delayed neutron

group.

The measurement of the &.,,,/A, in the VHTRC-1 core was made by

the PNS method. The technique adopted consequently was to measure

the prompt neutron decay constant in slightly subcritical state as a

function of the number of loaded fuel rods and to extrapolate the result

to the critical state point. The measured prompt neutron decay constant

is plotted in Fig.15 against the number of loaded fuel rods. The value

of (I, WBS determined to be 5.45*0.05 s-‘. Thus, the measured value of

8 ,,,,/A, was obtained as 4.8kO.l s-l.

IV.Comparison of experiments and calculations

1. Method of calculation

The critical experiments of VHTRC have been analyzed using the SRAC

code system”‘. Modeling of the unit lattice of the VHTRC-1 core is

shown in Fig.16. Calculations of group constants for the unit lattice

were made with the use of the nuclear data of ENDF/B-IV, taking into

account of the double heterogeneity of the fuel rod and the coated

particle, as precisely as possible.

The following conditions were adopted in the lattice calculation:

1) As the scattering law of thermal neutrons. the Young-Koppel’s

and the free gas models were used for graphite and the other

materials, respectively.

2) Isotropic reflective condition for’the outer boundary of the

lattice was adopted.

3) Resonance absorptions in the energy region from 1.1254eV to

130107eV were calculated by the PEACO routine“’ to

execute an ultra-fine group, multi-region lattice calculation

by the collision probability method.

The thermal neutrons in the energy region from O.OeV to 1.1254eV

were divided into 39 groups and the fast neutrons in the energy region

from 1.1254eV to IOMeV were divided into 22 groups. Using the neutron

spectra obtained by the lattice calculation. the cross sections of the

61 groups were condensed into th’e 24 groups(thermal:13 groups, fast:

11 groups) for the three dimensional core calculation by the diffusion

theory. The group constants for the reflector region were calculated

15180009

. I

by means of the asymptotic spectrum consisting of fission, l/E and

Maxwellian spectra.

The core calculation was performed with the three dimensional dif-

fusion code CITAT10N’2’ on the assumption of the reflectivity on the

midplane. A cross section of the graphite block of hexagonal prism

was expressed by 24 mesh points in the triangular model, and in the

axial direction, the half assembly of 120cm length was divided into

15 meshes. The effective, multiplication factor was calculated for

the two core temperatures 300K and 500K. The reactivity effect of

the thermal expansion due to the temperature rise was estimated using

linear expansion coefficients of graphite and frame material construc-

tion of the assembly. The estimated temperature coefficient of reac-

tivity due to the thermal expansion was about -8.4xiOe6d k/k/“C.

2. Critical mass

The calculated results of the critical mass of ‘?J and k,,, are shown

in Table 4 in comparison with the measured ones. The discrepancies

of k,,f between calculations and experiments are 0.4%Ak.,, at room

temperature and O.O7%Ak,,, at 200°C in the critical state. Differences

of the critical mass of 23sU between calculations and experiments were

3% at room temperature and 0.5% at 200.8”C, respectively.

3. Reactivity worth of HTTR mockup control rods

The core calculations were carried out by two different methods:

one was based on the 10 groups three dimensional diffusion theory and

the other was the Monte Carlo method. In the former method, 10 group

constants for the column with the CRs were prepared by 24 groups two

dimensional Sn calculation(TWOTRAN, X-Y geometry, SJ approximation)

taking into account of the precise geometry of the column. In the Monte

Carlo method, the VIM code’3’ was employed. The results are shown in

Table 1. The calculated values by both methods agreed with the

experimental cones obtained from the revised King-Simmon’s formula

within 6%.

4. Reactivity worth of HTTR mockup burnabl’e poison rod

In the lattice calculation for the BP rod containing 560pm BJC

particles, the double-heterogenity of the B,C particles in neutron

absorber pellets and the burnable poison rod in the fuel block was taken

into account by the collision probability. The calculated results of

-lO-

the reactivity worth of BP rod are shown in Table 2. The measured

and calculated values are found to agree with each other within the

experimental uncertainty of 5%. These’results indicate that the reac-

tivity worth of the BP rod can be estimated fairly good by the present

method.

5. Neutron flux distribution

The effective activation cross section of 63Cu was calculated using

JENDL-2 nuclear data14’. To consider the fine structure of the flux

distribution in the fuel block, the calculated radial flux distribution

was modified with the relative ratios of the fluxes at the outer fuel

rods to those at the inner ones in the lattice, which was estimated from

the results of the lattice calculation. The reactiohra.te was obtained

from the flux and the effective activation cross section of 63Cu which

was calculated with one dimensional slab and cylinder approximation

for the foil and the wire, respectively.

The radial distributions obtained finally are shown in Fig.12, where

the measured and calculated distributions are normalized to the average

value of the reaction rates of the all measured points. The calculated

values showed good agreement with the measured ones within 2%. The

axial distributions are shown in Fig.l3., where the measured and cal-

culated distributions are normalized to the integrated reaction rate

over the axial length of the half assembly. The calculated values

showed good agreement with the measured ones within 1% in the core

region.

6. Temperature coefficient of reactivity

The eigenvalue calculations were executed at 300, 325, 350, 400,

450 and 5OOK. The calculated reactivity change due to temperature rise

is shown in Fig.14 and compared with the experimental result. As

listed in Table 3, satisfactory agreements are observed between the

results of calculation and experiment for the temperature coefficient

of reactivity.

1. Kinetic parameter 8,),/A

The calculated values of Be,,< and A, were obtained with 24 groups

three dimensional diffusion method using the SRAC code with ENDFIB-IV

delaye’d neutron data. The calculated value of 8,//,/A, is

-ll- 15180011

. r

l7 o/,c 0.007249-- -=o,oo1335-5.428 5-l. AC (3)

Significantly large difference is observed between the measured and

calculated values of the kinetic parameter B,J~~IA, at critical. We

should,therefore. consider the refined technique of l~attice calculation.

V. Conclusion

A critical assembly VHTRC was constructed by modifying the SHE

in order to verify the nuclear design accuracy of HTTR.

The measurements of the reactor physics parameters: the critical

mass, the reactivity worths of HTTR mockup control rods and burnable

poison rods, the neutron flux distribution, the temperature coefficient

of reactivity and the kinetic parameter E,,//A, were performed in the

VHTRC-1 core which was loaded with 4% enriched fuel. Calculations l

were performed with the SRAC code system using the nuclear data based

on ENDF/B-IV.

Discrepancies of the effective multiplication factor between the

calculation and the experiment were 0.4%Ak,,, at room temperature and

0.07%dh,,, at 200°C in the critical state. Differences between caicu-

lation and experiment of reactivity worths of the HTTR mockup control

rods and the burnable poison rods, neutron flux distribution, tempera-

ture coefficient of reactivity and kinetic parameter were found to be

6%, 5%, 3%, 3% and 13%, respectively. The agreements between calcula-

tions and experiments are fairly good except the kinetic parameter.

The results obtained satisfy the accuracy requirement for fundamental

nuclear design of HTTR. 0 In the next step, we are planning to check the detailed design of

HTTR. At present. experiments are being conducted for the VHTRC-4

core which is loaded with Z-4-6% enriched fuel in the

axially-heterogeneous pattern.

Acknowledgments

The authors wish to thank to Dr.K.Tsuchihashi for comments on the

usage of the SRAC code system. Thanks are also due to the members

of the VHTRC for support in the experiment.

References:

-12-

l)K.Sanokawa and S.Saito :J.At.Energy Soc.Japan,Vol.29,p.603(1987).

Z)Y.Kaneko, et al.,:“Critical Experiments on Enriched Uranium

Graphite Moderated Cores”,JAERI 1257(1978).

S)H.Yasuda, et al.,: “Construction of VHTRC(Very High %Temperature

Reactor Critical Assembly)“,JAERI 1305(1987).

4)F.Akino, et al. ,:J.At.Energy Soc.Jspan,Vol.3l,p.682(1989).

5)T.Yamane, et al.,:“A simple method for reactivity determination

based on integral version of pulsed neutron area-ratio method”,

Proc.Int.Semin. on Nuclear Criticality Safety,Tokyo,(l987).

G)Y.Kaneko, et al .,:Nucl.Sci.Eng.,Vo1.50,p.173(1973).

7)F.Akino, et al.,:J.Nucl.Sci.Technol.,Vol.l7,p.593(19SO).

8)B.E.Simmons and J.S.King:Nucl.Sci.Eng.,Vol.3,‘p.695(1958).

S)T.Yamane, et al.,:To be published in J.Nucl.Sci.Technol.

lO)K.Tsuchihashi, et al.,: “Revised SRAC Code System”,

JAERI 1302( 1987).

ll)Y.Ishiguro and H.Takano :“PEACO:A Code for Calculation of Group

Constants of Resonance Energy Region in Heterogeneous System”,

JAERI 1219(1971).

lB)T.B.Fowler. et al.,:“Nuclear Reactor Core Analysis Code:CITATION”

ORNL-TM 2496(1969).

13)R.N.Blomquist. et al.,: “A review of the theory and application of

Monte Carlo methods”.ORNL/RSIC-44,p.31(1980).

14)T,Nakagawa(Ed) :“Summary of JENDL-2 General Purpose File”, JAERI-M

84-103(1984).

-13- 15180013

Table 1 Reactivity worths of H’ITR mockup control rods in the VHTRC-1 core

COl-e pattern

OC IC 2c 2R

12R

Measured ($) Calculated (S) i;;j PRXS Diffusion Monte Carlo -_-

- 0.17r0.03 __ 18.6-cO.4 19.3*0.4 18.7 19.2r0.8

- 30.4ztl.O 31.2 32.2r1.0 - 3.6+0.2 3.73 3.f3*0.9 - 24.5*1.3 25.8 25.9il.O

Table 2 Reactivity worth of H’M’FFtiockup burnable poison rod in the central column of the VHTRC-1 core

Type of Diameter of Neutron absorber pellet Reactivity worth($) B.P rod BdC particle Boron content Diameter HeightMeasured Calculated

(Pm) (wt%) hm) (mm) 0

N-8 <5 8.2 7.9 36.0 1.63kO.08 1.66 w-2.5 c.5 2.5 11.9 36.0 1.78*0.09 1.82 N-16 560 16.0 8.0 36.0 1.67kO.08 1.76

Table 3 Measured and calculated temperature coefficients of reactivity of the VHTRC-1 cc~re

Temperature coefficient of reactivity (lo-‘dk/k/’ C)

Measured ___-

-1.49*0.04 (25.5,71.2”C)

Calculated

-1.59 ( 27.77’C)

-1.80*0.04 -1.80 (100.9,150.5,199.6”C)(127,177,227°C)

-1.71*0.03 -1.73 (!25.5,199.6”C) (27,227’C)

Table 4 Measured and calculated values of eff’ective multiplications factor and critical mass of the VHTRC-1 core

-

Core Core Critical mass ,Effective multiplication

state Temperature of =‘U(kg) factor.k,,l

(“C) Expt. Calc. Expt . CESlC. --

Critical 27 4.520 4.378 1.0000+0.0003 1.0044 200.8 6.043 6.013 1.0000*0.0005 1.0007

- .

Fixed half assembly’ /----I

Movable half assembly

6 Heat insulaling cover

Neutron sovrce guide tube

Fig.1 Structural view of VHTRC.

0 Fuel rod(4XEU)

A Safety rod

E Control rod

@ BF,counter (Fixed half)

Fig.2 Critica fuel rod loading pattern of the VHTRC-1 core at 17.7’C

lU8(3015

. .

1.0

0

/I

I/

/ 6 Measured

_ / ---. Calculated

/ I

4 I

I

I I I

I I &--- e-+--lf - *OH- .-- ;T

Critical paint

I I I I

0 100 200 ; L Number ol fuel rods

0 Fuel rod(4iE

0 Fuel rad(Z%EU)

A Safety rod

(Fixed half)

Fig.4 Critical fuel rod loading pattern of the VHTRC-1 core at ZOO’C

)O

(Fixed hail)

Fig.5 Arrangement of the CR in the VHTRC-1 core (1C core pattern)

*(Fixed ha1 I)

Fig.6 Arranaement.ol the CRs in the VHTRC-1 core (ZC core pattern)

~Contrbl rod 0 CR

(Fixed half)

Fig.7 Arrangement of the CRs in the VHTRC-1 core (2R core pattern)

(Fixed half)

Fig.8 Arrangement of the CRs in the VHTRC-1 core (1ZR core pattern)

15180046

o”b”able half) (unit in I) (Fixed half)

The rsspective reaclivity value without or with bracket show that on the 15cm or 40cm distant plane from the midplans between the fixed and movable half.

Fig.9 Dependence of negative reactivity raluer on the detector position in the VHTRC-1 coce, measured by the araa:type pulsed neutron

method.

A Safety rod

l Fuel rod(4!4EU),m ,

n Control rod

_ v

*(Fixed half)

Fig.10 Arrangement of activation samples in the VHTRC-I core at the room Lemperature

151813019

@ Fuel rod( ZMEU)

q Control rod

A .

(Fixed half)

Fig.11 Arrangement of activation samples in the VHTRC-1 core at 200°C

. .

0 I, I I ) I I I,, I,, , I I 1 I , 1

l : Expt:

-I- : c a .I c

noon ,tenperature (Normalized point)

O-

Fig.14 Reactivity change in raising the core temperature