IEEE Ttan4ction6 on NatCeQa Science, Vot.NS-25, No.6, DecembeA 1978NEW NEUTRON SIMULATION CAPABILITIES PROVIDED BY THE SANDIA PULSE REACTOR-III (SPR-III)

AND THE UPGRADED ANNULAR CORE PULSE REACTOR (ACPR)

Larry M. Choate and Theodore R. Schmidt

Introduction

This paper briefly describes the nuclearreactor facilities at Sandia Laboratorieswhich are used for simulating nuclear weaponproduced neutron environments. These reactorfacilities are used principally in support ofcontinuing R & D programs for the Departmentof Energy/Office of Military Application(DOE/OMA) in studying the effects of radiationon nuclear weapon systems and components. Assuch, the reactors are available to DOE andDOD agencies and their contractors responsiblefor the radiation hardening of advanced nu-clear weapon systems. Emphasis here will beplaced upon two new reactor simulation sources;the Sandia Pulse Reactor-III (SPR-III) Facil-ity which enhances the neutron exposure volumecapabilities over those presently availablewith the existing SPR-II Facility, and theUpgraded Annular Core Pulse Reactor (ACPR)Facility which enhances the neutron exposurecapabilities over those of the former ACPRFacility. The SPR-III reactor core was de-signed to permit exposure of large electroniccomponents (e.g., AF&F assemblies) within themaximum flux region of a fast reactor. Theupgrading of the ACPR was, in part, motivatedby a desire to increase the available fluxlevel for symmetric in-core exposure of largenuclear weapon components and for accommodat-ing asymmetric ex-core exposure of even largersystem components (e.g., bare nuclear assem-blies). The SPR-III, which first went criti-cal in August 1975 (SPR-II in March 1967), isutilized in meeting representative exoatmos-pheric neutron hardening requirements; whilethe Upgraded ACPR, which first went criticalin April 1978 (former ACPR in June 1967) willbe utilized to meet more representative endo-atmospheric neutron hardening requirements.Both reactor facilities are located at SandiaLaboratories Technical Area V (TA-V) on Kirt-land Air Force Base, East, in Albuquerque,New Mexico. Experimenter's manuals, contain-ing detailed operational and performance in-formation, are presently being drafted andwill soon be available for each reactor facil-ity.

SPR-III

Physical Description

The SPR-III is a fast-burst reactor whichwas designed and constructed by Sandia Labora-tories. An unreflected and unmoderated cylin-drical assembly of uranium, enriched to 93

percent 235U, is alloyed with 10 weight per-cent molybdenum, in order to ensure phase sta-bilization of the fuel material. The coredesign consists of eighteen stacked fuel plateswhich are mechanically fastened into two halvesof nine plates each. The mass of the indivi-dual plates varies between 6.8 and 15.4 kg,and the total mass of the core fuel is about259 kg. The nine upper plates are fastenedtogether and are held stationary by the coresupport structure. The nine lower plates(referred to as the safety block) are fastenedtogether and attached to an electromechanicaldrive mechanism. Four reflector-type controlelements are utilized: three are used for

*Sandia Laboratories 001Kirtland Air Force Base, EastAlbuquerque, New Mexico 87185

control, and the fourth is used as the burstelement. The control-element drives are stand-ard electromechanical drive assemblies usinga rack-and-pinion drive, with power to thedrive mechanism transferred through an electri-cal clutch which disengages upon loss of power.The burst-element is hydraulically driven bya pneumatically actuated piston to achieve thehigh rate of reactivity insertion required forpulsing. The safety block, normally eitherfully inserted or fully withdrawn, is not usedfor fine control. The control-elements areutilized to establish a critical configurationof the core.

The primary shutdown mechanism in thepulse mode is the inherent negative-temperature-coefficient of reactivity produced by fuel ex-pansion. Following the reactivity insertion,the fuel heats up rapidly. Simultaneously,the negative-temperature-coefficient of reac-tivity of the system terminates the pulse.The definition of the pulse size can be deter-mined very accurately by preselecting the totalreactivity of the assembled core.

The reactor can also be shut down byeither one of two protection channels whichinitiate shutdown signals to the actuatingdevices for the control-elements and safetyblock. In addition, the mechanical design ofthe reactor includes inherent characteristicswhich serve to make the reactor subcriticalafter a pulse. These characteristics are(1) the mechanical shock forces induced by thepulse, which cause the safety block to breakaway from the holding armature and free-fallto a full-out position, and (2) the thermalexpansion of the fuel due to its rise in tem-perature, which also can cause the safetyblock to break away from the holding armature.During normal pulse operations, a programmedscram from the fast-level scram system initi-ates a shutdown signal which results in therelease of the safety block; however, the in-herent electromechanical delay times in thisaction are such that the actual disassemblyoccurs after the peak of the pulse.

A central cavity, measuring 17 cm indiameter and extending through both corehalves, 36 cm, is the primary experiment fa-cility. In addition, experiments may bemounted around the periphery of the reactor.An aluminum shroud, covered with an adhesivemixture loaded with boron-10 is placed overthe reactor. This shroud provides a flowchannel for the nitrogen cooling gas and itdecouples the core from low-energy neutronsthat are scattered back toward the core fromthe reactor room. The reactor stand is mount-ed on an elevator which can lower the reactorinto a shielded pit, permitting rapid accessto the reactor room.

The SPR-III can also be operated at steadystate power levels for those experiments re-quiring high fluences where rate effects areunimportant. However, the cooling capabilityof the nitrogen system and administrative re-strictions effectively limit the time at power.Normally, steady-state power operations arelimited to 10 kilowatts or less.

8-9499/78/1200-1625$00.75 ( 1978 IEEE 1625

Performance Characteristics'

As mentioned previously, the SPR-III isused primarily in the nuclear weapon effectssimulation program, and as such, is designedto satisfy exoatmospheric-type neutron hard-ening requirements. In the pulse mode ofoperation (used primarily for simulationtests), the reactor can provide a maximum flu-

ence in the cavity of about 5.6 x 1014neutrons/cm2 with an associated pulse width(FWHM) of about 76 ps. The nominal perform-ance characteristics of the reactor in thepulse mode (maximum) are presented in Table I.

The neutron spectrum, within the centralexperiment cavity, is a fast, fission-likespectrum which is representative of the"harder" exoatmospheric neutron threat envi-ronments. The measured centerline spectrum,within the central exposure cavity, is pre-sented in Table II. The spectrum was obtainedfrom a spectral unfold of foil activation dataThis slightly softened fission spectrum withinthe central cavity is presented in Figure 1.

Upgraded ACPR

Physical Description

The Upgraded ACPR was designed and con-structed by Sandia Laboratories as a majormodification to the existing ACPR, which wasdesigned by General Atomic. With an over-allperformance improvement of about a factor ofthree over the former ACPR, the Upgraded ACPRadditions consist of a new core design, newcontrol console, and coded aperture imagingactive fuel motion diagnostic system, as wellas major modifications to the neutron radio-graphy facility and other experiment facili-ties. The reactor core is located in an openpool of water, which is 3.1 m in diameter and8.5 m in depth, such that the top of the coreis approximately 7 m below the surface of thepool water. Besides affording moreadequate radiation shielding, this pool ofwater provides cooling of the reactor core bynatural convection. The most prominent fea-ture of the facility is the large 23 cm diam-eter dry irradiation cavity within the centerof the core. Access to this cavity is bymeans of a dry, air-filled, 25 cm diameterY-shaped loading tube, which extends verti-cally upward from the center of the core. Thevertical leg of the loading tube is laterallysupported by the top of the mounting platform,which is attached to the top of the tank. Theoffset leg of the loading tube is supportedby the storage chamber at the side of the pooltank. The mounting platform also serves tosupport the control rod, safety rod, and tran-sient rod drive mechanisms. The lower end ofthe curved offset leg of the loading tubebranches from the straight vertical leg about2.4 m above the top of the core and has anoffset of 1.6 m before joining the lower endof the experiment storage chamber.

The reactor core forms a right annularcylinder and consists of a triangular latticeconfiguration of about 220 cylindrical fuelelements surrounding the hexagonal centralexposure cavity (23 cm flat-to-flat). Thecore is fueled with stainless-steel clad1626**As of October 1, 1978 the Upgraded ACPRwas renamed the Annular Core ResearchReactor (ACRR).

BeO/UO2 elements. The fuel elements are

spaced and supported by a 2.5 cm thick alumi-num top grid plate and a 5.1 cm thick aluminumbottom grid plate and are protected on thesides by a 0.6 cm thick aluminum shroud. The

fuel is enriched to 35 percent in 235U with21.5 weight percent UO2 and 78.5 weight per-

cent BeO. The height of the active fuel sec-tion of each element is 52.2 cm with a claddedelement diameter of 3.75 cm. The radial re-flector region consists of solid nickel ele-ments surrounding the fueled region.

The fuel design provides the reactor witha large negative-temperature-coefficient ofreactivity, derived from the doppler effectin the uranium, as an inherent safety feature.This means that any unwarranted increase inreactor power heats the fuel, thus causingthe fuel to immediately become less effectiveas a fission source. Consequently the reactorreturns automatically to normal operatinglevels.

The reactor has eleven moveable regula-ting rods for performing the two modes ofreactor operation (pulse and steady-state).Six control rods, two safety rods, and threeadjustable fast transient rods are used forthis regulating function. The six motor-driven control rods (and two safety rods) areused to regulate reactor power and are with-drawn in unison to prevent flux tilting with-in the core. These rods are of the poison/fuel-followed type wherein the reactivityeffect of removing the poison (boron carbideimpregnated graphite), as the rods are with-drawn, is augmented by the simultaneous in-sertion of the fuel-follower (BeO/UO2) sec-

tion. The electro-pneumatic driven transientrods are used for adjusting pulse yield andfor providing the high rates of reactivityinsertion required for pulsing. These rodsare the poison/air-followed type.

Performance Characteristics2

As mentioned previously, the UpgradedACPR is used in the nuclear weapon effectsprogram to satisfy endoatmospheric-type neu-

tron hardening requirements. In the pulsemode of operation the reactor can provide amaximum fluence within the cavity of about

7.4 x 1015 neutrons/cm with an associatedpulse width (FWHM) of 6.5ms. The performancecharacteristics of the reactor in both thepulse and steady-state modes of operation(maximum) are presented in Table III. Theneutron spectrum, within the central cavity,is a slowing-down spectrum, typical of the"softer" endoatmospheric neutron threat en-

vironments, and is slightly "harder" thanthat of the former ACPR. The calculated cen-

terline cavity spectrum is presented inTable IV. These calculations were performedusing the two-dimensional neutron transportcomputer code TWOTRAN. The 18-group analysiswith this code utilizes ENDF/B-IV cell weightedcross sections, generated using AMPX, whichincludes correct thermal scattering cross

sections. The cavity spectrum for the Up-graded ACPR is presented in Figure 1.

Because of the increased capabilities

which the ACPR Upgrade provides over the for-mer ACPR, it is anticipated that greater flex-ibility will be available for radiation hard-ening efforts. In addition, the reactor fac-ility can accommodate other experimental pro-grams of national interest, such as advancedreactor safety research programs. A briefcomparison between the simulation capabilitiesof the former ACPR and the Upgraded ACPR ispresented in Table V.

1. B. F. Estes, private communication,Sandia Laboratories, 1978.

2. P. S. Pickard, private communication,Sandia Laboratories, 1978.

TABLE I: SPR-Ill MAXIMUM PULSE MODE NOMINAL OPERATINIG PARAMETERS(CENTRAL CAVITY, HORIZONTAL AND VERTICAL CENTERLINE, FREE-FIELD)

TABLE III: UPGRADED ACPR NOMINAL OPERATING PARAMETERS'(CENITRAL CAVITY, HORIZONTAL AND VERTICAL CENTERLINE, FREE-FIELD)

PULSE OPERATION (MAXIMUM)REACTIVITY INSERTIONFUEL T-EMPERATURE RISEPULSE WIDTH (FWHM)REACTOR PERIODENERGY RELEASEPEAK POVERNEUTRON FLUEtNCE (>10 KEV)NEUTRON FLUENCE (ALL EtNERGIES)PEAK NEUTRON FLUX (>10 KEV)PEAK NEUTRON FLUX (ALL ENERGIES)GAMMA DOSEPEAK GAM1MA DOSE RATE

$2.95

1100°C6.5 MS

1.65 MS360 MJ

29,500 MW4.3 x 1015 NEUTRONS/CM27.4 x 1015 NEUTRONS/CM23.5 x 1017 NEUTRONS/CM2-S6.1 x 1017 NEUTRONS/CM2-S2.9 x 106 RADS (H20)2.4 X 108 RADS (1f20)/s

NEUTRON FLUENCEPEAK NEUTRON FLUXGAMMA DOSEPEAK GAMM&A DOSE RATEREACTIVITY INSERTIONREACTOR PERIODPULSE WIDTH (FWHM)TEMPERATURE RISEPEAK-TO-AVERAGE TEMPERATURE RATIOENERGY RELEASEPEAK POWER

5.6 x 1014 NEUTRONS/CM27.4 x 1018 NEUTRONS/CM2-S1.6 x 105 RADS ([120)2.1 x 109 RADS (H20)/s

$1.1023 ps76 ps400°C1.613 MJ170,000 MW

STEADY-STATE OPERAT ION (0AXIMUM)POWER

NEUTRON FLUX (>10 KEV)NEUTRON FLUX (ALL EtNERGIES)GAMMA DOSE RATEFUEL TEMPERATURE

2.0 MW2.4 x i013 NEUTRONS/CM2-S4.1 x I013 NEUTRONS/CM2-S1.6 x 104 RADS (H20) S

960C

*MEASURED PRELIMINARY VALUES (WITHIN + 10%)

* MEASURED

TABLE IV: CALCULATED UPGRADED ACPR NiEUTRON SPECTRUM(CENTRAL CAVITY, HORIZONTAL AND VERTICAL CENTERLINE, FREE-FIELD)

TABLE II: MEASURED SPR-III NEU'TRON SPECTRUM(CENTRAL CAVITY, HORIZONTAL AND VERTICAL CENTERLINE, FREE-FIELD)

NORMALIZEDGROUP LOWER ENERGY UPPER ENERGY GROUPNUMBER (MEV) (MEV) FRACTION

0.005580.048680.104100.121190.121230.161890.161540.110100.075020.036110.026680.011900.010230.004620.001110.00002Q.00000kQ1.00000

FLUENCE*(IIEUTRCNS/CM2-MIEV)

1.59+111.95+126.94+121.21+132.42+134.63+136. 46+137.314+137.50+137.22+136.67'135.96+135 .12+134.62+132.22+135.00+113.33+10

NORMALIZEDGROUP LOWER ENERGY UPPER ENERGY GROUP

NUPBER (MEV) (MEV) FRACTION

1234

S6789

101112

131415161718

4.4s9+01. 35+07.43-14.08-11.23-19.12-39,61-41.30-41.76-52.38-61.29-66.50-73.00-71.60-71.00-76.00-82.00-81.00-9

1, 49+14.49+01. 35+07.43-14.08-11.23-19,12-39.61-41.30-41.76-52.38-61,29-66.50-7

3.00-71.60-71.00-76.00-82.00-8

0.018610.147810.087840,078060.113910. 132370.085150.070190.062890,054110.017660.018860.020090.017230.016660.021810.02991

1.00000

'NORMALIZED INTEGRAL FLUENCE 10i4 NVT.

1627

1234S6789

1011121314151617

6.5+04.0+02.5+01.5+01.0+06.5-14.0-12.5-11.5-11.0-16.0-24.0-22.0-21.0-25.0-31.0-31.0-4

1.0+16.5+04.0+02.5+01.5+01.0+06.5-14.0-12.5-11.5-11.0-16.0-24.0-22.0-21.0-25.0-31.0-3

FLUENCE*

(NEUTRONS/CM2-MEV)1.79+114.71+121.45+132. 33+134.00+131. 16+141.04+158.45+155.62+163.56+171. 62+182. 95+185. 74+181. 23+192.78+195.45+197.48+193. 60+19

*NORMALIZED INTEGRAL FLUENJCE = 1014 NVT.

6ENE

TABLE V: ACPR NUCLE

QUANTITY

ERALORE FUEL MATERIAL

NO. OF FUEL ELEMENTSNO. OF CONTROL RODSNO, OF TRANSIENT RODSFUEL HEIGHT (CM)EXCESS REACTIVITY ($)

STEADY-STATE MODEMAXIMUM POWER-LEVEL (MW)MAXIMUM FLUX

(NEUTRONS/CM2-S)

PULSE MODEMAXIMUM ENERGY

RELEASE (MJ)MAXIMUM FLUEIJCE

(NEUTRONS/CM2)PULSE WIDTH

(MSEC FWHM)PEAK POWER (MW)

SPECTRUMNEUTRON FRACTION*

>0.1 EV> 1 EV> 10 EV> 100 EV> 1 KEV> 10 KEV>100 KEV> 1 MEV

EXPOSURE VOLUME1CENTRAL CAVITY

(CM3) tEXPOSIEXTERCOREI xWITHSOFTERELAT

EAR WEAPONS EFFECTS SIMULATION CAPABILITIES

ACPR UPGRADE ACPR1978 - 1967 - 1977

BEO/UO ENRICHED TO 35% U-ZRH ENRICHED TOWITH 21.5 WI. Z U02 20% WItH 12 'IT. Z U

220 1568 63 3

52.2 38.17.0 8.5

2.0

4.1 x 1013

360

7.4 x 1015

6.529,500

0.940.880.820.750.660.580.470.22

'CALCULATED ESTIMATES

2.2 x 1043URES CAN BE MADERAL TO THE REACTORW(ITH FLUENCE OF1015 NEUTRONS/cm2A SLIGHTLYENED SPECTRUM[IVE TO CENTRAL CAVITY.

0.6

1.2 x 1013

$4

0

$4

4-.

H

$4

E4

110

2.3 x 1015

515,000

0.890.750.680.62O.570.510.440.21

CALCULATED ESTIMATES

1.6 x 1O0EXPOSURES EXTERNALTO THE REACTOR COREARE LIMITED DUE TOLOW AVAILABLE FLUENCE.

0

IN

0$40'04)C3V404

04zu

NEUTIRON ENERGY (M4eV)

Figure 1. Central Cavity, Horizontal and Vertical ContrliTne,Free-Piceld Netutron Spectrum for the SPR-III and thcUpgraded ACPRI.

1628