OFFICIAL USE ONLY

CC: Members of Reactor

Committee

COMMITTEE REPORT

Internal Design of Reactor

The program to date has consisted of determining best values of critical

numbers; these should allow the design to proceed with less fear of serious error.

Those chosen are:

- 1. Thickness of concrete shield
- 2. Reflector size
- 3. Thermal column length
- 4. Lead shielding for x-rays

Auxiliary information to the above includes:

- 5. Thermal neutron flux at reactor core surface; same for [gamma] rays
- 6. Attenuation of neutron flux in various materials; same for

[gamma] rays - 7. Tolerances

Flux: Two-group reactor theory^{a1} was applied to the water-graphite system as if

it were a sphere. Since this method lumps all fast neutrons into a single

"group", the predicted critical U235 mess of 0.96 kg (compared with the

value of .87kg) is rather good. The thermal flux at the edge of the core computed

was 3.75 x 1011 neutrons/cm²/sec as compared with tho

The theoretical attenuation in the graphite of the reflector and thermal column was

The

The [gamma]-flux may be estimated from the 10 kw power level, the number of [gamma]'s

per fission and the area of the container. A figure of 5.4 x 1011/cm²/sec of 2 MoV

energy is used.

Tolerances: The present accepted levels are: slow neutrons 1500/cm²/sec; 2 MoV [gamma]

rays 1050/cm²/sec. Those correspond to a 0.1 r/8 hour day.

Attenuation of [gamma] rays: Inverse-square spreading exponential attenuation is

assumed, with absorption coefficients ([mu] in o-[mu]x, where x

is in cm) as follows:

graphite | 0.064 | |

lead | 0.51 | |

0.09 | for ordinary type | |

0.19 | for "heavy" type |

Attenuation of slow and fast neutrons: Data from various sources diff or widely.

All predictions, however give safety with

a 6' Shield if the concrete is "heavy". Further investigation is needed.

Thickness of concrete shield: A tentative choice of a 5 ft diameter central cavity

filled with graphite

overall dimension of 17 feet, leaves a shield thickness of 6 feet. This is proposed

as fixed unless safety is questionable, which is not the case. From the standpoint

of [gamma] rays, the effects of reflector graphite, shield concrete (ordinary type) and

distance, reduce the flux from 5.4 x 1011 to less than 1 [gamma]/cm²/sec. From another

viewpoint the necessary thickness to reach tolerance levels is approximately 4 ft,

so a two foot safety margin is provided.

Thickness of reflector: In order to satisfy two requirements (a) a low critical

U235 mass and (b) a maximum thermal flux at the core surface,

the largest practical thickness is used. This seems to be 20". The rate of change

of critical radius and the rate of change of the ratio of wall flux to central flux

were computed by one-group theory. With this thickness, the critical volume differs

by only 3% from that for infinite graphite; the flux is within 5% of the ultimate.

Thermal column length: To achieve a length compatible with the Los Alamos reactor,

it is indicated that the end should be located near the

side

shield. If the Cd shield at the end is used, a built-in Pb block shield is needed

to stop the [gamma]'s from Cd; if Boron instead were the neutron shield, the Pb could be

eliminated. A 7 foot thermal column can easily be obtained in the space.

Load shield outside the reflector: If the thermal neutron flux from the end of the

column is to be free of reactor [gamma]'s a lead

shield must be inserted, presumably next to the reflector. One question that must

be decided is - What is [gamma] free? A slab 2" thick will cut the flux to [~=] 1400/cm²/

sec, slightly above health tolerance; a 4" slab will cut it to [~=] 110 or 1/60

tolerance. It is found that little thermal neutron absorption is encountered in

either case. Even in 4" of Pb the flux is reduced by only 6%. (The alternative, Bi,

absorbs 93% in a 4" section.)

Problems yet to be looked into further by means of calculations are listed:

- Type of concrete needed for neutron absorption (moderately heavy is probably

adequate) - Effect of open exposure ports (At present, estimates are no better than 80 grams

of U235 per port, comparing reactor cores with and without reflector, and

5 grams, on the basis of a fractional solid angle for escape. A guess would

give 20 grams.) - Decision on Cd vs. B for end of thermal column
- Neutron source strength, location and adjustment

Notes:

^{a1}