c ORNURSIC-55

Early Test Facilities and Analytic Methods for Radiation Shielding

Proceedings of a Special Session for the Radiation Protection and Shielding Division at the American Nuclear Society Winter Meeting Chicago, Illinois, November 15-20,1992

1 Celebrating

P

‘, MARTIN MARIETTA ENERGY SYSTEMS, INC. ., iFOR THE UNITED STATES ‘: ‘DEPARTMENT OF ENERGY This report has been reproduced directly from the best available copy.

Available to DOE and DOE contractors from the Office of Scientific and Techni- I cal Information, P.O. Box 62, Oak Ridge, TN 37831; prices available from (6i5) 576S401, Frs 6263401.

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, com- pleteness, or usefulness of any information, apparatus, product, or process dis- closed, or represents that its use’would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily con&ii tute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. ORNL/RSIC-55

Engineering Physics and Mathematics Division

EARLY TEST FACILITIES AND ANALYTIC METHODS FOR RADIATION SHIELDING

Proceedings of a Special Session for the Radiation Protection and Shielding Division at the American Nuclear Society Winter Meeting Chicago, Illinois, November 1520, 1992

1 Celebrating Compiled by: D. T. Ingersoll Oak Ridge National Laboratory

and

J. K. Ingersoll Tee-Corn, Inc.

Published: November 1992

Prepared by the OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37831 managed by MARTIN MARIETTA ENERGY SYSTEMS, INC. for the U.S. DEPARTMENT OF ENERGY under contract DE-ACO5840R21400 Preface

This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting held in Chicago, Illinois on November 15 20,1992. The meeting is of special significance since it commemorates the 50th anniversary of the first controlled nuclear chain reaction, which occurred, not coincidentally, in Chicago. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting.

I must admit that throughout my school years, history was one of my least favorite subjects, owing in part to its seeming lack of relevance. Such is not the case here. As head of the present-day shielding section at ORNL, I feel a close professional affiliation with and a personal sense of gratitude toward the authors and the people whom they describe in their papers. They are individuals who helped to form the foundations of the discipline of radiation shielding and have all been appropriately honored by widespread recognition for their accomplishments. In their papers, they present a collage of facts and personal remembrances, which I find delightfully entertaining, fascinating and even inspiring. The picture presented here is by no means complete; many other talented and dedicated individuals have contributed to the history of radiation shielding. However, meeting time was limited and tough decisions had to be made.

The first paper, authored by Lorraine Abbott, could have opened with: “Inthe beginning...” She describes the earliest activities in radiation shielding research, which began immediately following the Chicago pile test. Frontiering programs grew from the insight and efforts of three key individuals: Everitt Blizard, Theodore Rockwell, and Charles Clifford. Abbott goes on to describe the major influence that Adm. Hyman Rickover had over those early programs and directions. The first shielding experiments at the Oak Ridge X- 10 Pile were in support of the Hanford production reactors, which later spawned shielding research at that site, as described by Wilbur Bunch in the second paper. Bunch provides a thorough description of their development and testing program, which focused primarily on iron/masonite shields and a vast range of special concretes. In an interesting aside, Bunch notes that not everything from those days survived, such as the “vigor” unit, which never gained the same level of acceptance as did the related “lethargy” tit.

The third paper, written by Norm Schaeffer, describes a fascinating test program for the Aircraft Nuclear Propulsion Project. The unique challenge of these tests is best reflectedin the factthatnuclear engineers participating in the test had to first be trained in parachute jumping. The limitations of the ground tests and the awkwardness of the flight tests led to a compromise solution: experiments conducted at the Oak Ridge Tower Shielding Facility, which allowed the reactor and shield to be suspended 200 ft above the ground. The design, construction and operation of the TSF is described in the fourth paper, authored by Buzz Muckenthaler. The TSF, which must be the longest surviving shield test facility, has supported a vast array of national programs, many of which are highlighted in Muckenthaler’s paper.

Switching from early shield test facilities to early shielding design methods, Dave Trubey describes in the fifth paper the original development and evolution of buildup factors. Point-kernel codes employing buildup factors were some of the first successful computational methods and am still in frequent use today. The method exemplifies the artistic nature of shielding design analysis due to the need to constantly balance speed and accuracy, a problem which persists even today. In the sixth paper, Kal Shure describes the early methods used at Bettis Laboratory for the Naval Reactor program. Cloaked by secrecy, many of the developments at Bettis paralleled work at other laboratories. Shure wittily places in perspective the “worthiness” of results computed with these early codes, and makes reference to “user-tolerable” codes - a term which unfortunately applies to even modem computing software.

. . . 111 The seventh paper, authored by John Butler, provides a thoughtful description of radiation shielding research as it began in the United Kingdom. With a primary focus on kernel and Monte Carlo methods, much of the U.K. development complemented U.S. activities. The same was true of benchmark testing activities in the U.K., which centered around the LID0 facility and later the NFSTOR/ASPIS facility.

In the eighth and final paper, Herb Goldstein begins by stating that: “The title tells it all.” In classic Goldstein style, he presents a wise and delightfully personal review of early computation methods, nuclear data, and even the early computers. He does well to point out the project-driven nature of shielding development, but goes on to describe some of those rare nuggets of fundamental theoretical research which have managed to “trickle” along the way. Goldstein provides a fitting conclusion to the session by stating that: “Now we have the tools...if anyone still wants to ask the questions, and is willing to pay for the answers.”

Unlike most meeting proceedings, which am meant to be studied one paper at a time, this collection of historical perspectives is meant to be taken as a whole and is best suited for a quiet evening and a soft armchair. So get comfortable and enjoy.

Dan Ingersoll Nuclear Analysis & Shielding Section Oak Ridge National Laboratory

iv Table of Contents

. . . Preface ...... lu

Acknowledgements ...... vii

Biographies ...... ix

The Origin of Radiation Shielding Research: The Oak Ridge Perspective Lorraine S. Abbott ...... 1

Shielding Research at the Hanford Site Wilbur L. Bunch ...... 9

Aircraft Shielding Experiments at General Dynamics Fort Worth, 1950-1962 Novnan M. Schaeffer ...... 17

Where Have the Neutrons Gone - A History of the Tower Shielding Facility F. J. Muckenthuler ...... 33

History and Evolution of Buildup Factors David K. Trubey ...... 47

Early Shielding Research at Bettis Atomic Power Laboratory Kabnan Shure and 0. J. Wallace ...... 55

UK Reactor Shielding: Then and Now John Butler ...... ,...... 59

A Very Personal View of the Development of Radiation Shielding Theory Herbert Goldstein ...... 75 Acknowledgements

We wish to especially thank the authors for meeting the tight schedule for completing their manuscripts. Since it was desired to distribute the proceedings at the session, the authors had to prepare their papers well in advance of the meeting. We also wish to thank Dave Trubey, who helped with the organization of the special session, and John Butler, who served as co-chairman. We am grateful toLorraineAbbottforvolunteeringtheservicesofherprivatecompanyforthecollectionand preparation of the final manuscripts, and to Bob Roussin, director of the Radiation Shielding Information Center, for his willingness to pay for reproduction costs. Finally, we nxognize the support of the ANS Radiation Protection and Shielding Division, and especially the RR&S Program Committee for their assistance in supporting this special session.

vii Biographies

Brief biographical sketches for each of the authors are given below in the order in which their papers are presented. With such a distinguished group of individuals, it was impractical to list all of their accomplishments and awards. Suffice it to say that they have all been widely recognized and honored. Also, not surprisingly, they have all been formally acknowledged by ANS for their technical excellence and distinguished service to the society.

Lorraine S. Abbott received a B.S. degree in chemistry from Maryville College in 1948. She combined her technical degree and her interest in journalism to become a technical editor and writer at Oak Ridge National Laboratory, specializing in recording radiation shielding research from 1948 to 1986. Of the numerous documents she has authored, including many as a ghost writer, two have probably been distributed to the widest audiences: Shielding Against Initial Radiationsfrom Nuclear Weapons (ORNLIRSIC-36, 1973) and Review of ORNL Radiation ShieldingAnalyses of the Fast Flux TestFacility Reactor (ORNL+-5027,1975).As an editor, she co-edited, with Everitt P. Blizard, the shielding volume of the U.S. Atomic Energy Commission’s Reactor Handbook, published in 1962. Upon retiring from ORNL in 1986, she formed her own company, Tee-Corn, Inc., based in Knoxville, Tennessee. Her company specializes in writing and publishing newsletters and reports and producing videos for technical organizations.

Wilbur L. Bunch received a B.S. degree (1949) and a M.S. degree (1951) in physics from the University of Wyoming. Being accustomed to the dry, open spaces, he took employment at the Hanford Site in 195 1. His original assignment at Hanford was in the shielding group, where he was given the responsibility for gamma-ray measurements. Later assignments at Hanford included reactor experiments to define fuel enrichments, reactor design modifications, and instrument development. The government obtained several patents based on his activities in this time frame. When the Fast Flux Test Facility project began in 1964, he returned to the field of shielding and became Manager of the Radiation and Shield Analysis group. His work on the FFTF brought him in contact with shielding personnel in many countries of the world. He is a Fellow of ANS and served as Chairman of the Radiation Protection and Shielding Division in 1977-78. Most recently, he was Technical Program Chair for the 1992 RP&S Topical Confer- ence.

Norman Schaeffer joined the nuclear group in Fort Worth after he received his Ph.D. in physics from the University of Texas in 1953. He was head of the shielding group at General Dynamics/Fort Worth from 1955 until 1962, when he organized Radiation Research Associates with several coworkers. E. P. Blizard of the Neutron Physics Division at ORNL gave RRA its first contract: some overflow work on a weapons radiation handbook. RRA successfully contributed to the shielding community for 28 years, with clients in the U.S., Europe, and Japan. Schaeffer is best known as the editor of the textbook Reactor Shieldingfor Nuclear Engineers, first published by the Atomic Energy Commission in 1973 and still in use in several graduate courses. He is a Fellow of ANS and a charter member of the Radiation Protection and Shielding Division. He is currently an independent consultant.

F. J. (Buzz) Muckenthaler received a B.S. degree (1947) and a M.S. degree (1949) in physics from the University of Kansas. He was initially employed by the Fairchild Engine and Airplane Corporation in Oak Ridge in 1949, and transferred in 195 1 to Oak Ridge National Laboratory. He joined the staff of the Tower Shielding Facility in 1955, but was later moved to lead the experimental activities at the Lid Tank Shielding Facility. He was transferred back to the TSF in 1962, where he has continued to conduct and lead numerous measurement programs supporting DOE’s advanced reactors programs and several DoD experimental shielding pro- grams. Most recently, he led a seven year program of measurements for a U.S.-Japan collabo-

ix ration for the Advanced Liquid-Metal-Cooled Reactor program. He also participated in several off-site measurement programs, including Operation BREN in 1962.

David. K. Trubey received a B.S. degree in physics from Michigan State University in 1953 and a M.S. degree in computer science from the University of Tennessee in 1988. He started his career in shielding at the ORNL Lid Tank Shielding Facility under the direction of E. P. Blizard. He was co-founder and past director of the ORNL Radiation Shielding Information Center, an organization which continues to provide an international focus for the shielding community. In addition to his work in developing new shielding data and methodologies, he served over 19 years as chairman of the ANS-6 standards committee. Trubey is a Fellow of ANS and was a former chairman of the Radiation Protection and Shielding Division. He has been recognized with numerous awards,includingtheANS/RP&STheodomRockwellAwardin 1992. Heretiredfrom ORNL in 1991 and continues to do consulting work for RSIC.

Kalman Shure is a consultant in Radiation Analysis at the Bettis Atomic Power Laboratory. Contributions during his 35-year career have been in the area of water-cooled reactor shielding and have included fundamental experimental research, exploration of new design techniques, and the development of practical shield design methodologies. Of special note is his work on the attenuation of neutrons in water-iron shields, refined methods for calculating the ductile-to- brittle transition temperature of steel as a function of neutron exposure, important contributions to the establishment of decay-heat and gamma-ray transport standards, and the application of point kernels, P, approximations, and discrete-ordinates codes to neutron transport problems. He has been a member of ANS 5.1 (Decay Heat Power in Light Water Reactors) and ANS 6.4.3 (Gamma-Ray Coefficients and Buildup Factors for Engineering Materials).

John Butler received his M.A. and Ph.D. degrees in physics from Oxford. He subsequently joined AERE Harwell and became Head of the Shielding Group with responsibility for developing computational methods for the U.K. Power Reactor Programme. In 1972, the group moved to Winfiith and he developed the ASPIS shielding facility in conjunction with the NESTOR reactor. He was a major contributor and promoter of integral benchmark experiments and the use of covariance data for the computation of nuclear data sensitivities and uncertainties in shielding applications. Mom recently, he has been involved with the exploitation of shielding methodology in industry, including the development of the EUROPA nuclear well log calibration facility in Aberdeen. He is now a consultant on business development for Petroleum Services and Reactor Services with AEA Technology.

Herbert Goldstein received his Ph.D. from the Massachusetts Institute of Technology in 1943. He joined Nuclear Development Associates in 1950, and six years later, originated the Computer Index of Neutron Data (CINDA), which may be the longest continuously active computer-based bibliographic index. He also led a development effort utilizing the moments method for computing gamma-ray transport. Since joining the faculty of Columbia University in 196 1, he has written numerous publications, including his famous textbooks, ClassicalMechanics and FunaimenraZAspects of Reactor Shielding. He has also shepherded 27 students through the doctoral degree program. He is Fellow of several scientific societies, including ANS, and was Vice Chairman and Chairman of the Radiation Protection and Shielding Division during the first two years of its existence. In 1990, he was honored as the third recipient of the ANS/RP&S Theodore Rockwell Award and also the Arthur Holly Compton Award. The Origin of Radiation Shielding Research: The Oak Ridge Perspective

Lorraine S. Abbott Tee-Corn Inc. Knoxville, Tennessee

Abstract

The discipline of radiation shielding re- the importance of secondary gamma rays as a search originated in 1947 when physicist Everitt radiation sourtits limitations prompted the P. Blizard, following the orders of his Navy design of a new Lid Tank Shielding Facility. boss, Captain Hyman Rickover, initiated shield- Placed in operation in mid- 1949, the Lid Tank ing measurements in a “core hole” through the utilized a fission source positioned over the 7-R-thick concrete shield of the X- 10 Pile in outer end of the core hole, which in turn was Oak Ridge, Tennessee. For that effort, he re- covered by a large tank of water attached to the cruited the assistance of Charles E. Clifford, a outside of the pile shield. Measurements made chemical engineer with experience in pile foil inside the tank provided shielding design data measurements. The first samples tested were for the Navy’s nuclear-powered submarines prepared by another chemical engineer, Theodore and for a number of stationary reactors, as well Rockwell, who had been developing high-den- as for the U.S. Aircraft Nuclear Propulsion sity concretes for potential reactor shields since Program. Blizard, Clifford, and Rockwell re- 1945. While the Core Hole Facility yielded mained associated with shielding research and significant results-including showing the ef- design throughout their careers, each making fectiveness of a spiral configuration in reducing significant contributions to the development of neutron streaming in ducts through shields, and the shielding discipline as it is known today.

*****************

The origin of radiation shielding research The X-10 Pile reached criticality for the -at least from the Oak Ridge perspectiveis firsttimeonNovember4,1943,only 11 months intimately tied to the operation of the world’s after the first controlled self-sustaining nuclear first “permanent” nuclear reactor. That reactor chain reaction had been achieved in the famous was, of course, the l-megawatt X-10 Pile con- Chicago Pile ((Y-1). Henry W. Newson later structed during the year 1943 on an East Ten- described2 how he and George Weil, bothmem- nessee hilltop that is now part of the city of Oak bers of the University of Chicago’s Metallurgi- Ridge. cal Laboratory, together with several Du Pont Built in secret as a pilot plant for the large- engineers, spent the early morning hours of that scale plutonium-production reactors that were day bringing the X-10 Pile to “just critical.” to be COnStrUCtedat Hanford, Washington, the Enrico Fermi arrived later in the day and or- X-10 Pile was an essential component of the dered the final additions of uranium in the United States’ new nuclear weapons program. presence of various notables, including Arthur The story is told that the construction crews at H. Compton, head of the Metallurgical Labora- the X-10 Pile conjectured among themselves as tory. to why any building would need a 7-ft-thick The Du Pont engineers present for that concrete wall.’ They had no inkling that the momentous event were in training for later wall, which consisted of 5 ft of barytes-haydite operation of the Hanford reactors. Among them concrete sandwiched between two 1-ft thick- was Charles E. Clifford, a young chemical nesses of ordinary concrete, was actually the engineer who with other trainees had dismantled world’s first nuclear reactor shield. the CP-1 (with their bare hands) and rebuilt it as

1 the second Chicago Pile (CP-2) at the site of the sion, received a telephone directive from future Argonne National Laboratory. Rickover ordering him to begin studying radia- A few months after the X-10 Pile began tion shields at the X-10 Pile. The call was no operation, Clifford joined Newson and others in doubt prompted by Rickover’s decision to push performing the world’s first reactor shielding for the development of nuclear-powered sub- experiment atop the pile. A 6-ft-square hole had marines, which would require a different type of been left in the top shield for testing a section of reactor shield than those planned for stationary the proposed Hanford reactor shield consisting reactors. In later years, Blizard described the of laminated steel and masonite. It was a suc- call as being extremely brief and very much to cessful test, with the shield’s radiation attenuat- the point-mother words, typically Rickover- ing characteristics found to be mom than ad- as well as one that changed the course of his equate; however, later, under actual operating career. On April 28.1947, he wrote to Columbia conditions, the decomposition of the radiation- University that “the exigencies of the project damaged mason&e presented problems. require that I work full time in a pile shielding program,” and he prepared to stay at Clinton In late 1944, Clifford departed for Hanford Laboratories indefinitely. and graduate school, only to return in 1947 to the X-10 Site, then known as Clinton Laborato- Blizard immediately began making plans ries. He joined the Technical Division, where for setting up a shield test facility at the X-10 another young chemical engineer, Theodore Pile. Because Clifford had already had experi- Rockwell, was developing high-density con- ence making gold foil measurements at the pile, cretes as potential reactor shield materials. it was decided even before he reported to work Rockwell had hired in at the Oak Ridge Y- 12 that he would be loaned by the Technical Divi- plant in 1943 and had transferred to the X-10 siontothePhysicsDivisiontoworkwithBlizard. plant in 1945. The location selected for the first shield test By that time the operator of Clinton Labo- facility was an approximately 2-l&square “core ratories was Monsanto Chemical Company, hole” that had been left in the middle of the and early in 1946 officials of that company, shield on the west face of the pile. The plan was realizing that the country should be educated on to insert shield samples in the hole and measure the promises of nuclear energy, had invited the neutron and gamma-ray fluxes penetrating members of the U.S. War Department to Oak through the samples. Ridge for a series of lectures on the subject.3 In a memorandum dated July 9, 1947, Among those responding to the invitation was Blizard outlined the program for the first shield- Captain Hyman Rickover of the U.S. Navy, ing tests at the new Core Hole Facility. After who became very enthusiastic about the possi- making measurements in a stainless steel tank bilities of utilizing nuclear power. In fact, he filled with water, the team would test various immediately promoted the organization of a concrete aggregates, including several devel- nuclear training school at the X-10 Site and oped by Rockwell. Among them were aggm- subsequently sat in on many of the classes.4 gates specially developed for a Brookhaven The school’s first class, which graduated in National Laboratory pile and several for a high- June 1947, consisted of 35 individuals ap- flux reactor then under development at the X- 10 pointed by various industries and universities Site. and by the U.S. Navy. One of the Navy appoin- Almost immediately it became apparent tees was Eve&t P. Blizard, a civilian physicist that a more intense fast-neutron source was of the U.S. Navy’s Bureau of Ships who had needed. To increase the fast-neutron flux inci- participated in the Bikini atom bomb tests early dent on the hole, Blizard and Clifford asked the in 1946. Arriving in Oak Ridge in the fall of that pile operators to move some of the natural year, he expected to remain in the city until the uranium fuel slugs distributed in the reactor’s following fall, during which time he hoped to graphite moderator to the outer edge of the perform research for submission as a PhD dis- graphite reflector, which was separated from sertation to Columbia University. the shield by a wide air-cooling plenum. In that However, even before he graduated from position, the slugs were in a straight line with the training school, Blizard, who was also work- the hole across the plenum. Clifford recounts ing in the Clinton Laboratories Physics Divi- that during this process of preparing the

2 hole-and for reasons that are vague to him instruments for measuring fast-neutron dose now-he crawled into the hole with a “cutie rates, thermal- and intermediate-energy neu- pie” (gamma-ray ionization counter) while the tron fluxes, and gamma-ray dose rates free from reactor was operating at low power andonly a 2- neutron effects, all in a water environment. in. thickness of lead plugged the inner edge of The first result of this instrument develop- the hole. ment effort, which became an important and These first Core Hole Facility tests also enduring part of the shielding rogram, was the showed that the stainless steel tank used for Hurst fast-neutron dosimeter. ti Other early in- water and solution measurements inside the struments were a boron trifluoride low-energy hole interfered with the measurements. More- neutron detector, an anthracene crystal gamma- over, the tank became so radioactive it pre- ray dosimeter, and a graphite-walled CO,-filled sented a handling hazard. These problems were ionization chamber. solved by replacing the steel tank with an alumi- num tank. In April 1948, Blizard first documented the type of new shielding facility he was consider- But a more serious problem arose. The pile ing. Writing to a colleague, he said “We expect shielding surrounding the core hole was less to try a lid experiment or two and may even plan effective than the shield samples being mea- to adopt this as our regular technique.” sured. This allowed neutrons from the pile shield to leak into the sides of shield samples, The “lid” was to consist of several short clouding the interpretation of the measurements. natural uranium rods from the X-10 Pile sand- wiched between masonite boards and placed Nonetheless, the facility yielded significant over the outside of the core hole. Thermal new results. Writing in January 1948 to Clark neutrons streaming from the hole would pro- GoodmanoftheMassachusettsInstituteofTech- duce fission in the rods to provide a source for nology, Blizard reported “We will...have some the experiments. Clifford suggested that a tank interestinginformation...onashieldperforation of water be positioned on the outside of the pile and gas conduction into or out of a reactor. I shield adjacent to the source, thereby providing have shown this to...NEPA and they were a vessel in which shield samples could be amazed, as I hope you will be.” measured and eliminating a host of problems Blizard was referring to a test on a concrete associated with background radiation and per- sample penetrated by a 6-in-diameter spiral sonnel safety. The facility, called the Lid Tank duct. Performed for the Air Force’s NEPA Shielding Facility, was indeed constructed, be- Program (Nuclear Energy for Propulsion of ginning operation the following year (in mid- Aircraft), it showed that the increased radiation 1949) under the direction of Clifford.6 (Note: In penetrating the sample was primarily due to the 1955, the original lid source was replaced with shield’s reduced density (that is, streaming a thin circular disk of enriched uranium, which through the duct was not apparent). doubled the source power from approximately Several Core Hole Facility tests were also 3 watts to 6 watts.) the first to reveal that radiation shield designers In the meantime, Captain Rickover was could not limit their considerations to the pri- keenly ‘interested in two significant events un- mary source of neutrons and gamma rays. The der way at the X-10 Site, by then officially production of secondary gamma rays by neu- known as Oak Ridge National Laboratory and tron interactions within the shield was found to operated by Carbide and Carbon Chemicals be an important factor. Company. One was the development of a suc- Still, the Core Hole Facility was increas- cessful technique for fabricating the highly ingly considered to be inadequate and cumber- enriched uranium-aluminum fuel plates called some to operate and the need for a new facility for in ORNL’s compact high-flux reactor de- became obvious. Also, the lack of sophisticated sign (later to be constructed as the Materials cotmting techniques was limiting the facility’s Testing Reactor [MTR] at Arco, Idaho). The capabilities. By early 1948, Blizard was search- other was the development of a technique for ing for suitable instruments throughout the coun- cladding the fuel plates with zirconium instead try. Finding none, he asked Clifford to work of aluminum. The design of a submarine nuclear with ORNL instrumentation groups to develop power plant was crystallizing.7

3 UNCLASSIFIED ORNL-LR-OWG 39663

ALUMINUM ALUMINUM CANS FILLEO AUXILIARY TANK WITH MATERIAL UNDER INVESTIGATION in- I-

CONFIGURATION TANK -J-

ft-thick CONCRETE REACTOR SHIELD

The ORNL Lid Tank Shielding Facility, located on the west face of the shield of the Oak Ridge X-10 Pile, was the first facility designed for experimental radiation shielding research. It operated from mid-1949 until the pile ceased operation in 1963.

Also in the meantime, Rockwell, who was shield physics could be as challenging as reac- heading up the ORNL Shield Materials and tor physics, if not more so, he had convinced Enginwtig Section in the Technical Division, some theoretical physicists to join the shielding continued his development of both generic and group. In early 1948, he visited several labora- specific shield materials for continuing tests at tories throughout the country to identify any on- the Core Hole Facility. Among them was a going shielding and instrument development special shield developed for consideration as a activities and to encourage joint programs. Hanford replacement shield. It was an oxy- Blizard also lobbied the Research Reactors chloride concrete, which according to Rockwell Section of the Atomic Energy Commission to “contained more water per cubic centimeter establish an AEC Shielding Advisory Commit- than pure water itself.” Also included were tee. Writing to the chief of the section in July tungsten carbide and boron carbide shields, as 1948, he said such a committee should include well as several shields proposed for non-reactor representatives from X-10, preferably Rockwell uses, such as isotope shipping containers.* and himself to ensure representation of both the During this period Rockwell also devel- shield engineering and shield physics groups. oped an aluminum and boron carbide mixture He also named the Massachusetts Institute of called “boral” that became well known as a Technology and the Bureau of Standards, the capture gamma-ray suppressor and has been only other organizations openly admitting that used in many shields over the years. (In 1958, they were interested in shielding research. He the monthly magazine Nudeonics named further suggested that members be named from Rockwell’s patent on the boral fabrication pro- Brookhaven National Laboratory and the Naval cess as one of the 27 most important patents in Research LaboratoMth having exhibited the history of atomic energy.)’ some interest in shieldineand from organiza- tions that would be expected “at some time or As these activities continued, Blizard was other to produce an efficient shield.” He listed working on several other fronts. By arguing that the latter organizations as Argonne National 4 Laboratory, the NEPA Program, Hanford Engi- most remembered was Welton’s concept of a neer Works, and Knolls Atomic Power Labora- single cross section to describe the “temoval”of tory. The group’s first meeting was scheduled neutrons traversing a heavr shield mixed with to be held in Oak Ridge on October 1,1948- hydrogeneous materials.’ The concept was immediately following a large three-day na- subsequently tested in the Lid Tank and was tional symposium that had been organized by utilized for many years in shield design, espe- Rockwell as the first conference ever conducted cially in submarine shield design. on the topic of radiation shielding.” Before the 1949 Summer Shielding Con- The national symposium brought together ference ended, word was out that Oak Ridge in Oak Ridge approximately 150 representa- National Laboratory would become a principal tives from 40 major institutions. Its announced player in the Air Force’s nuclear-propulsion purpose was to establish a “clearing house for program. In September, the AEC officially re- the future exchange of ideas with regard to quested that ORNL set up an Aircraft Nuclear shielding or protection against atomic radia- Propulsion (ANP) program. l2 The implications tion.” But one of the’ participants, Captain for the shielding program were obvious. On the Rickover, felt that the symposium should also following November 25, Blizard wrote to msultinanactionplan,andhecalledinRockwell Rickover, “It has been decided that a new and others to outline a list of 12 “subjects for shielding facility with source strength of 10 KW immediate attack.“The impact of that particular is essential for design of a mobile reactor shield. listisunknown, atleasttome, butinmany ways At present two propositions are being it corresponded to the shielding agenda that considetetipecifically, a ‘water boiler’ and a Blizard had outlined in a memorandum to the critical assembly using MTR fuel elements...If AEC (on September 2,1948) for the forthcom- no delays are experienced in obtaining ap- ing meeting of the AEC Shielding Advisory proval, we hope to be obtaining data next sum- Committee. mer.” That optimistic schedule was not quite At the second meeting of the Advisory met; however, the MTR-type assembly of Committee, at Brookhaven National Labora- ORNL’s well-known Bulk Shielding Facility didbe ’ its first operation on December 17, tory the following March, Blizard led the draft- l$” ing of a proposed AEC-sponsored shielding 1950. research program. He also suggested that an With the initiation of a large-scale ANP intensive shielding theory conference be held at program at ORNL, Blizard decided to remain in Oak Ridge that summer. He had already te- Oak Ridge permanently. In a second memoran- ceived assurance from Rickover that the Navy dum to Rickover on November 25, he resigned would sponsor the conference. (Concurrently, from the Navy tobecome an employee of ORNL. he was urging Rickover to sponsor needed That same month Rockwell resigned from measurements at another laboratory to deter- ORNL to accept employment with the Navy, mine cross sections and gamma energies for where he remained for 15 years, the last 10 years 0.5- to 8-MeV neutrons incident on iron, lead, as Rickover’s technical director. During the tungsten, and bismuth.) next several years, Rockwell and the ORNL With Gale Young of Nuclear Development group continued their collaborationonLid Tank Associates as the leader, the 1949 Summer experiments to test and analyze shield designs Shielding Conference was a tremendous suc- for the Nautilus, the Navy’s first nuclear-pow- cess. Participants included Herman Feshbach ered submarine. As the design progressed, of MIT, J. W. Butler of Argonne, Herman Kahn Rockwell recalls, many exotic shield materials of the Rand Corporation, and Frances Friedman were suggested, but he and the ORNL group and Gerald Goertzel, in addition to a number of successfully prevailed with their recommenda- other local people. John von Neumann of the tions that the shield consist of optimized ar- Institute for Advanced Study in Princeton acted rangements of an iron thermal shield combined as an advisor. Working closely with the group with water and lead. was Theodore Welton, who had recently moved In subsequent years, the three original from MIT to the University of Pennsylvania and “shielders,” known to their colleagues simply had been asked by the AEC to conduct a survey as Ted, Cliff, and Bliz, became well known for of shielding theory. From this shielding session their individual contributions to the field of emerged a number of ideas, but perhaps the radiation shielding. 5 Theodore (Ted) Rockwell, shown here in a re- Charles(Cliff) Clifford,nowretired,waspresent cent picture, pioneered the development of high-den- when the X-10 Pile first went critical and later partici- sity concrete shields and the material called boral pated in the world’s first shielding test atop the pile. In while at Oak Ridge National Laboratory. Later he 1947, he and Everitt Blizard began the first experimen- transferred to the Navy, where he acted as liaison with tal shielding program at Oak Ridge National Labora- theORNL shielding team performing submarine shield tory, testing samples in a “core hole” through the side design studies. In 1986, citing his 1956ReactorShield- of the pile shield. In 1949,he became the director of the ing Design Manual as a shielding “bible,” the RPS new Lid Tank Shielding Facility, and in 1952 he Division established the Rockwell Lifetime Achieve- assumed responsibility for the design (and subsequent ment Award. initial operation) of the Tower Shielding Facility.

In 1950, two years after the journal Nucle- air-filled ducts.17 One of his most important onics had offered to publish papers from the contributions was to establish unequivocally 1948 national shielding symposium in a special that the production of secondary gamma rays issue, Rockwell realized he could not get them was an overriding factor in shield design and declassified and instead collected them, along that the positioning of materials within a shield with several new papers, in a 334-page classi- was a critical consideration. The key experi- fied document that became known informally ment, performed for the submarine program, as ORNL-710.14 He was more successful in was with alternating layers of lead and borated 1956 when he edited the well-known Reactor water of equal thickness. Surprisingly, as lead Shielding Design Manual that was the first was added, the gamma-ray dose rate did not collection of shielding information ever made decrease as rapidly as had been anticipated. In available to industry in an unclassified fonn. l5 fact, Clifford found that a decrease actually Thirty years later, referring to the Manual as a occurred when the lead slabs closest to the shielding “bible” still in use, the ANS Radiation source were removed. He eventually optimized Protection and Shielding Division in 1986 es- the shield by adjusting the positions of the lead tablished the Rockwell Lifetime Achievement slabs until all were equally effective with regard Award, making him its first recipient. After to both the shield weight and the dose rate. leaving the Navy, he became a principal in the Clifford transferred from the Lid Tank soon firm MPR Associates, from which he is now after it became apparent that the new ORNL retired. His most recent accomplishment has Bulk Shielding Facility was inadequate for been the publication (in October 1992 of his ) some aircraft shielding experiments and that a latest book, titled The RickoverEfect. 6 new facility would be needed. The concept As the first director of the Lid Tank Shield- proposed by Blizard and others was that of four ing Facility, Clifford designed numerous ex- towers from which a reactor source would be periments to test shield performance and shield- suspended high above the ground. Clifford was ing theories, including the (Albert) Simon- given the assignment of overseeing its design Clifford theory of neutron streaming through and initial operation. The resulting facility, the

6 ORNL Tower Shielding Facility that began operation in 1954, is described in a companion paper presented at this session. I8 In 1955, Blizard was named director of the new ORNL Applied Nuclear Physics Division (later renamed the Neutron Physics Division), which performed experimental and theoretical studies of both reactor physics and shielding research. As director, he continued to person- ally lead the OFUVLshielding group (and, to a large extent, the shielding community at large). When the Aircraft Nuclear Propulsion program was cancelled in 196 1, he immediately concen- trated on studies of spacecraft shields, accelera- tor shields, and shields against nuclear weap- Oll!L Through the years Blizard was also achroni- clerof shield design techniques, reportedlargely in ORNL documents but also in chapters con- tributed to a number of books and in the shield- Everitt (Bliz) Blizard, who died of leukemia in ing volume of the AEC’s ReactorZ-Zana%ook,19 1966, is remembered as “the father of reactor shield- which he edited. In addition, he served as editor ing.“From the time he andClifford conducted the first shielding tests at the X-10 Pile in 1947, Blizard guided of Nuclear Science and Engineering from 1959 theoretical and experimental radiation shielding stud- until his death in 1966. ieS at Oak Ridge National Laboratory, coordinating But it is for his leadership and vision in them with many shielding programs at other institu- shielding research that Blizard is most remem- tions. The studies covered shields for stationary and mobile reactors, spacecraft, accelerators, and radia- bered. So strong was his influence that in a tion-hardened structures. The Franklin Institute post- memorial issue of Nuclear Science and Engi- humously awarded him the Elliott Cresson Medal for neering,20he was eulogized by his good friend his contributions to shielding technology, and the Herbert Goldstein as “the father of reactor shield- International Atomic Energy Agency dedicated a mas- ing,” a title reinforced by other honors. In 1966, sive shielding compendium to his memory. the Franklin Institute posthumously awarded him the Elliott Cresson Medal for “his many contributions to the technology of radiation shielding.” And in 1968, the International Atomic Energy Agency’s massive volumes on Such were the beginnings of the field of sbieldin@ngineering Compendium on Ra- radiation shielding. Admittedly, they have been diationShielding,editedbyR.G. Jaeger-were described with a biased Oak Ridge perspective. dedicated in his memory. However, I personally observed and recorded There was, of course, the fourth man who so ORNL shielding research from 1948 to 1986 influenced the three original shielders that he and worked directly with Blizard and Clifford can be counted as one of them. Captain during that period. From them and others I Rickoverqater to become Admiral Rickover- learned of Rockwell’s contributions and of knew what he wanted from shielding research Rickover’s strong influence. Therefore, I re- and adamantly insisted onquality performance, main confident that my bias is well founded and as he did in all other areas of research that he that few, if any, will remember the facts to be touched. otherwise. References

1. A. F. Rupp, “The First 15 Years,” Oak 13. W. M. Breazeale, “The New Bulk Shield- Ridge NationalL.&oratory Review 9(4), p. ing Facility at Oak Ridge National Labom- 4, 1976. tory,” Oak Ridge National Laboratory, ORNL-991 (May 8,1951).’ 2. H. W. Newson, “A Visit from St. Nucleus,” op. cit., p. 16. 14. Theodore Rockwell, III, and Arnold S. K&es, “Theoretical and Practical Aspects 3. Theodore Rockwell, personal communica- of Shielding,” Oak Ridge National Labora- tion to Lorraine Abbott, September, 1992. tory, ORNL-7 10 (September 29, 1950).

4. Oak Ridge National Laboratory Review, 15. Reactor Shielding Design Manual, op. cit., p. 67. Theodore Rockwell, III, Editor, Naval Re- actors Branch, Division of Reactor Devel- 5. G. S. Hurst, R. H. Ritchie, and H. N. opment, U.S. Atomic Energy Commission Wilson, “A Count Rate Method of Measur- Report TID-7004 (March 1956); repub- ing Fast Neutron Tissue Dose,” Rev. Sci. lished by McGraw-Hill and Van Nostrand, Instr. 22,981,1951. 1956.

6. C. E. Clifford, “The ORNL Shield Testing 16. Theodore Rockwell, The Rickover Eflect, Facility,” Oak Ridge National Laboratory, Naval Institute Press, Annapolis, Mary- ORNL-402 (n.d.). land, 1992.

7. Oak Ridge National Laboratory Review, 17. A. Simon and C. E. Clifford, “The Attenu- op. cit., p. 57. ation of Neutrons by Air Ducts in Shields, Part I. Theory,” Oak Ridge National Labo- 8. Internal memorandum by T. Rockwell, ratory, ORNL-1217 (November 1952); “ORNL Shield Materials and Engineer- Nucl. Sci. Eng. 1, 103 (1956). ing,” 1948. 18. F. J. Muckenthaler, “Where Have the Neu- 9. Nucleonics, October, 1958. trons Gone - A History of the Tower ShieldingFacility,“paperpresented at ANS 10. KnoxvilleNews-Sentinel, Knoxville, Ten- International Conference, November 18, nessee, September 18, 1948. 1992 (this conference).

11. R. D. Albert and T. A. Welton, “A Simpli- 19. Reactor Handbook, Second Edition, Vol- fied Theory of Neutron Attenuation and Its umeIII,PartB:Shielding,EverittP.Blizard, Application to Reactor Shield Design,” Editor, Lorraine S. Abbott, Associate Edi- Westinghouse Electric Corporation, Atomic tor, Interscience Publishers, New York, Power Division (November 30,195O). 1962.

12. Oak Ridge National Laboratory Review, 20. HerbertGoldstein,“EverittP.Blizard, 1916- op. tit, p. 63. 1966,” Nucl. Sci. Eng. 27, No. 2 (1967). Shielding Research at the Hanford Site

Wilbur L. Bunch Westinghouse Hanford Company Richland, Washington

Introduction

The original three plutonium production two additional reactors (DR and H) were built reactors (B , D, and F) constructed at the Hanford using the same shield design. At the request of Site in 1943-44 had shields consisting of alter- R. L. Dickeman, an experimental facility was nate layers of iron and a high-density pressed- included in the top of the DR Reactor to permit wood product called Masonite.* This design evaluation of shield materials. Concurrent with was the engineering response to the scientific the measurement of attenuation properties of request for a mixture of iron and hydrogen. The materials in this facility, a program was under- design mix was based on earlier studies using taken to investigate the structural characteris- iron and water or iron and paraffin; however, tics of various high-density Portland cement these materials did not have satisfactory struc- concretes. This research effort continued for hral characteristics. Although the shields per- over a decade and led to the use of these formed satisfactorily, the fabrication cost was concretes in subsequent reactor shields at the high. Each piece had to be machined precisely Hanford Site and elsewhere with significant to fit within structural webs, so as not to intro- savings in construction costs. duce cracks through the shield. Before 1950,

Shield Facilities

Most of the attenuation measurements were Although these two wells in the top of the made in the facility located in the top shield of DR Reactor provided the bulk of the shielding the DR Reactor. This facility consisted of a pair data at the Hanford Site, three other facilities of identical stepped openings through the bio- were also used to some extent. The first of these logical shield. They were centered about eight provided a small opening into the graphite feet in front of the core midplane and approxi- through the “E” test hole of the F Reactor. This mately two feet, nine inches on either side of the permitted short-term irradiation of samples core centerline. This location was dictated by through the shield and into the first few rows of the safety rod pattern through the top shield. As fuel and provided a basis for normalization of shown in Figure 1, each opening was 35 3/8 other measurements to the reactor power and inches square at the bottom, and 41 l/2 inches flux levels. square at the top, with an overall depth of 50 Another shield facility was the “A” test hole inches. Five steps were included to eliminate in the D Reactor. This facility consisted of radiation streaming along the sides. Each well stepped cylindrical openings approximately was lined with steel to provide a gas seat for the eight inches in diameter that permitted irradia- reactor atmosphere. The bottom of the well tion of samples in both the thermal shield and liner, in conjunction with the reactor thermal the existing iron-Masonite shield. These mea- shield, provided a lo-inch-thick iron shield surements permitted additional normalization between the well and the graphite reflector. of the DR shield facility measurements to the Additionally, a two-foot-thick graphite reflec- unperturbed iron-Masonite shield. tor existed between the thermal shield and the first layer of fueled process tubes, forming the The success of the measurements program core. in the DR facility inspired the inclusion of another shield facility when the C Reactor was *Masonite is a trademark of the Masonite Corporation. built. This facility was placed in the side shield

9 Figure 1. Cross section of Bulk Shield FacilitpDR Pile. of the reactor and consisted of a hydraulic ram preliminary measurement to normalize subse- that moved a “bucket” containing the test mate- quent tests to the basic iron-Masonite data was rial into position within the shield. The impor- the only use ever made of the facility because of tant feature of this facility was to permit evalu- the high levels of radiation that were encoun- ation of different thermal shield materials. A tered.

Measurement Methods

The nature of the facilities dictated the thermal neutrons from that associated with the types of measurements that were employed. epithennal neutron flux. The foils were cali- Based on half-life and cross section consider- brated in the Hanford Standard Pile, which ations, gold foils and cadmium covered gold consisted of a large stack of graphite with an foils were used to determine the distribution of embedded radium-beryllium source. thermal and low-energy neutrons through the shield. An adequate dynamic range could be Measurement of the high-energy neutron achieved using small foils in the high-flux distribution was by way of activation of sulfur regions and large foils in the lower flux regions. by the n,p reaction, which has a threshold at Gold has an approximate l/E activation cross approximately 1 MeV. Various methods of section, except for a resonance at about 5 eV. fabricating sulfur detectors resulted in the use of The use of the cadmium covers provided a pressed pellets. These pellets were also covered separation of the activation associated with by cadmium to prevent low-energy activation

10 of the sulfur and any impurities that would the calibration table, this indicated the pmdomi- generate undesirable background count rates. nate energy of the leakage spectrum was at Countratesfiomboththesulfurandgoldsamples approximately 24 keV, which is consistent with were taken several times over several days to the “window” in the iron cross section. establish the half life and permit accurate ex- trapolation back to the end of the irradiation. The gamma ray measurements were made The low activity foils were obviously counted using both ionization chambers and film. At first while permitting the highly activated foils that time, we did not have thermoluminescent time to decay to appropriate count rates. dosimeters (TLD) that could have been embed- ded in the shield to determine the gamma ray One other technique that was used in some field. Therefore, both graphite-walled and mag- of the measurements was the “hydrogenous nesium parallel plate guarded ring 10 cc ioniza- integrator.” This technique was based on the tion chambers were fabricated that could be principle that a monoenergetic source of neu- embedded within the shield test slabs in the DR trons would be scattered and slowed down in a wells. Type K radiographic film was employed hydmgenous material in a characteristic man- in the outer layers, where temperatures and ner. That is, the thermal neutron population radiation levels permitted. These could also be would peak at a depth related to the energy of the related to a one-liter gamma ray chamber at the source neutrons. Calibration measurements us- surface of the shield to provide further normal- ing paraffin were reported by Brookhaven Na- ization of the results. Both the fihn and the tional Laboratory, as illustrated in Table 1. We chambers were calibrated using a standard ra- used both paraffin and lucite to establish that the dium source. Sensitivity to neutrons was mea- peak activity of the neutron flux in an integrator sured to ensure that it was negligible in both the at the surface of the burned-out iron-Masonite chambers and the film. shield was at approximately one inch. Based on

Table 1. Calibration Measurements Using Paraffin

PARAFFIN INTEGRATOR CALIBRATION (*) Incident Neutron Energy Position of Maximum Flux (inches) 0.03 eV 0.3 24 keV 1.0 700 keV 2.1 4.1 MeV 3.0 15 MeV 4.5

*Reference 1.

Materials Studied

All of the initial measurements were fo- percent carbon, and 44.4 weight percent oxy- cused on the original iron-Masonite shield de- gen. The composition of the layered iron- sign, both to determine its characteristics and to Masonite shield is shown in Table 2. provide a basis for the evaluation of the subse- quent concretes. The iron-Masonite designcon- We began testing various high-density con- sistedofaltematelayersof4.5inchesofMasonite cretes in the DR facility as soon as theiron- and 3.75 inches of iron or carbon steel. The Masonite measurements were completed. How- Masonite had a density of about 1.28 grams per ever, concrete testing was interrupted to initiate cubic centimeter, and a composition that was additional tests on iron-Masonite. The operat- about 6.2 weight percent hydrogen, 49.4 weight ing power levels of the reactors were being

11 increased to take advantage of improved fuel process tubes a few rows from the poison. This designs and reactor cooling systems. These enrichment maintained the flattened central increased power levels significantly increased portion of the core, while creating a steeper the temperature of the Masonite in the shields, buckled zone at the edge. By using a poison which led to concerns of deterioration of this material that generated a useful by-product, material. Laboratory tests at the elevated tem- protection of the shields was effected at mini- peratures being encountered indicate4l that the mal cost. Through the use of this method, the hydrogen and oxygen would be driven from the power levels and production rates of the reac- Masonite in time, leaving a residue of only tors were increased by almost an order of mag- carbon. Using effective removal cross sections nitude (see Table 3). while maintaining shield of hydrogen, oxygen, and carbon as a basis, integrity throughout their operating lifetime. ‘Masonite was differentially removed from the The primary purpose of the DR test facility test slabs to simulate various stages of deterio- was to investigate the attenuation properties of ration of the shield. Measurements using these various high-density Portland cement concretes. “deteriorated” configurations indicated that ra- Table 4 summarizes the compositions of the diation levels through the shields would be- tested concretes together with that of iron- come intolerable if operation continued at the Masonite for comparison. In the case of both the higher power levels. iron-limo&e andmagnetite-limonit concretes, Because of the large production incentive two different mixes were studied. The first, associated with the increased power levels, a called “conventional,” was a mix that could be test was designed to load poison in the outer pouredinto forms. The second, called”pmpakt,” layer of process tubes to reduce radiation leak- was suitable for placing dry aggregate into the age to the shield. To maintain production, en- forms and pumping in the grout with the use of riched uranium was placed in some of the an intrusion aid. This resulted in slight differ-

Table 2. Average Composition of Iron-Masonite Shields

Element G/cm3

H 0.043 C 0.345 0 0.310 Fe 3.568

TOTAL 4.266

Table 3. Reactor Power Levels

DESIGN LEVEL MAXIMUM LEVEL REACTOR MEGAWATTS MEGAWATTS

B 250 2090 D 250 2050 DR 250 2015 F 250 2040 H 400 2140

*Reference 2.

12 13 ences in composition and density. There is mately one month, which was the typicallength significant uncertainty in the as-built composi- of an operating cycle for the reactor. The slabs tion of any concrete, because of the mixing and were then removed from the well, taken to a curing processes. laboratory, weighed, and placed in a large oven, At the same time that the attenuation char- where they were baked at the specified tempera- acteristics of these concretes were being mea- ture for a period of at least one month. Then the sured, their mechanical properties were being set of slabs was reweighed and returned to the tested. During this time frame, we encountered DR test wells for additional measurements. The the temperature problem in the existing reactor change in weight of the slabs as a result of the shields; therefore, the concrete tests were ex- heating process was attributed entirely to loss of panded to determine the effects of elevated water content, and the composition of the slabs temperatures on both the structural and attenu- was revised accordingly. Results of these mea- ation properties. Additionally, small concrete surements are published, in Volume II of the specimens were irradiated within the reactor to Engineering Compendium on Radiation try to determine the effect of radiation on the Shielding,3 and, hopefully, have been used in concrete. Control samples were maintaine4l in the design of radiation shields. Because of the an oven at the same temperature as those under- oven that was available, the maximum tempera- going irradiation. Within the accuracy of the ture achieved was 320’ C (608’ F). measurements, it was not possible to detect any Basically, it was found that the change in change in the irradiated concrete other than that neutron attenuation characteristics could be at- associated with temperature. tributed to the water loss, using effective m- Measuring the change in attenuation prop- moval cross sections. This conclusion recog- erties of each concrete as a function of tempera- nizes the difficulties in dealing with materials ture took a long time. Measurements on the as- such as concrete, which in two cases unexpect- cured test slabs were made in one of the DR test edly increasedin weight when heated from 175’ wells, with each irradiation requiring approxi- C to 320” C.

Calculational Methods In my opinion, we, were severely handi- One of the less successful techniques that capped at the Hanford Site by the lack of was developed was the presentation of the neu- scientific computer facilities. At the time of this tron results as a function of “vigor” rather than work, the Hanford Site was considered to be a lethargy or energy. Vigor was defined as Log E/ production site and the computers were selected E,, whereas lethargy is Ln Et/E and is useful in to maintain records. Nonetheless, at least two describing the slowing down process within the significant shielding codes were developed at core. Each unit increase in vigor corresponded the Hanford Site in that era. The first of these exactly to an order of magnitude increase in was the MAC code, which employed the re- energy. A plot of neutron flux, or neutron dose moval-diffusion theory to predict the attenua- rate versus vigor, yielded a linear description of tion characteristics of the various concretes. A the relative importance of the neutrons as a similar code was developed at the same, time in function of energy. No one else found any value England, using slightly different methods, but in the use of vigor. achieving similar success. An approximation to the neutron spectrum The second significant code written at the in the shield was made using the Westcott cross Hanford Site was ISOSHLD, which uses the section method. This technique divided the point kernel method. One important feature of lower energy flux into two components, the this code was the inclusion of specific built-in thermal flux and the slowing down flux. Based geometries that grossly simplified data input. A on the bare and cadmium-covered gold mea- second important feature was the inclusion of surements, the relative magnitude of these two the RIBD code to generate the fission product components was established. By neglecting any inventory associated with irradiated fuel. The losses in the slowing down flux as a function of included databases made it easy to use the code, energy, this component could be extrapolated to once the parameters were understood. the higher energy “fission’* distribution whose

14 magnitude was approximated by the activity of transport world using discrete ordinates and the sulfur foils under the assumption the fission Monte Carlo codes for current shielding calcu- spectrum still applied in the shield. In spite of lations. Such methods probably eliminate the known limitations, this method did provide an need for experiments such as those described; indication of spectral changes taking place within however, in their absence, the experiments pro- the shields. With the advent of more powerful vided a firm basis for the design of production computers, the Hanford Site has moved into the reactor shields.

Conclusion

The completion of the attenuation and struc- Additionally, the studies on heat damage to the tural measurements on the various high-density Masonite resulted in changes that permitted concretes provided a. database that could be increases in production, while at the same time usedin the design of shields for new reactors. At maintaining shield integrity. The decade of the the Hanford Site, the top shield of the C Reactor 1950’s was an exciting time at the Hanford Site was constructed of concrete, whereas the sides inthe fieldof shielding. Althoughit seemedlike were constructed of iron-Masonite. As more there was a lot of paper work in those days to get and more data were acquired, the later reactors, anything done, we safely completed the tests in KE, KW, and NPR, had shields of various a relatively short period of time and with very tested concretes. Using concrete in these shields few people involved. Those days are gone for- materially reduced the cost of the facilities. ever, but I hope the results linger on.

References 1. F. B. Oleson, “Neutron Monitoring With Battelle Pacific Northwest Laboratory, Indium Foils,” BNL-35 1, Brookhaven Na- Richland, Washington (1965). tionalLaboratory,Upton,NewYork(1955). 3. IAEA, “Engineering Compendium on Ra- 2. D. L. Deneal, “Historical Events -Reac- diation Shielding,” Volume II, Shielding tors andFuelsFabrication,“RLREA-2247, Materials, International Atomic Energy Agency, Vienna, Austria (1975).

Bibliography

1. W. L. Bunch and R. L. Tomlinson, “At- 4. E. G. Peterson, “Shielding Properties of tenuation Properties of Hanford Pile Shield Ferrophosphorus Concrete as a Function of Materials,” Pile Technology Section Engi- Temperature,“USAECReport,HW-64774, neering Department, USAEC Report, HW- Richland, Washington (1960). 36370, Richland, Washington (1955). 5. E. G. Peterson, “Shielding Properties of 2. W. L. Bunch, “Attenuation Properties of Ordinary Concrete as a Function of Tem- High Density Portland Cement Concretes perature,” USAEC Report, HW-65572, as a Function of Temperature,” USAEC Richland, Washington (1960). Report, HW-54656, Richland, Washington (1958). 6. E. G. Peterson, “Shielding Properties of Iron Serpentine Concrete,” USAEC Report 3. D. E. Wood, “The Effect of Temperature on HW-73255, Richland, Washington (1962). the Neutron Attenuation of Magnetite Con- crete,” USAEC Report, HW-58497, 7. L. L. Rice, “Construction Test High Den- Richland, Washington (1958). sity Concrete Shielding,” USAEC Report, HW-19563, Richland, Washington (1950).

15 8. H. S. Davis, “Reporton Development Work 14. E. G. Peterson, “MAC-A Bulk Shielding with High Density Concrete Performed in Code,” HW-7338 1, Richland, Washington 1952 at the Hanford Atomic Products Op- (1962). eration,” USAEC Report, HW-31213, Richland, Washington (1954). 15. H. S. Davis, “An Evaluation of Costs for Constructing High Density Concrete 9. H. S. Davis, “Troutdale Investigation of Shields at 105-l 1,” USAEC Report, HW- High Density Concrete,” USAEC Report, 33951(B), Richland, Washington (1955). HW-32456, Richland, Washington (1954). 16. R. L. Engle, J. Greenborg, and M. M. 10. R. L. Tomlinson, “‘Procedure for Radioac- Hendrickson, “ISOSHLD - A Computer tive Foil Preparation and Counting,” Code for General Purpose Isotope Shield- USAEC Report, HW-33121, Richland, ing Analysis,“BNWL-236, BattelleNorth- Washington (1955). west Laboratory (June 1966) and Supple- ment, Richland Washington (March 1967). 11. R. L. Tomlinson, “Two Techniques for Fast Neutron Detection Using the S32(n,p) 17. G. L. Simmons et al., “ISOSHLD-11: Code P32Reaction,“USAECReport,HW-33122, Revision to Include Calculations of Dose Richland, Washington (1955). Rate from Shielded Bremsstrahlung Sources,” BNWL-236, Sup. 1, Battelle 12 W. L. Bunch, “The Effect of Masonite Northwest Laboratory, Richland, Wash- Burnout on Shield Attenuation Properties,” ington (1967). USAEC Report, HW-38418, Richland, Washington (1956). 18. R. 0. Gumprecht, “Mathematical Basis of Computer Code RIBD,” DUN-4136, Dou- 13. R. D. Albert and T. A. Welton, “A Simpli- glas United Nuclear, Richland, Washing- fied Theory of Neutron Attenuation and Its ton (1976). Application to Reactor Shield Design,” USAECReport, WAPD-lS(Del), Westing- 19. D. R. Marr, “A User’s Manual for Com- house Electric Corporation, Atomic Power puter Code RIBD-II, A Fission Product Division ( 1950). Inventory Code,” HEDL-TME 75-26, Richland, Washington (1975).

16 Aircraft Shielding Experiments at General Dynamics Fort Worth, 1950-1962

Norman M. Schaeffer Schaeffer Associates Fort Worth, Texas

Background

In 1946, as a result of a nuclear propulsion diation source for shield testing in flight. The feasibility study called NEPA (Nuclear Energy former was called the GTR (Ground Test Reac- Propulsion for Aircraft), startedduringtheMan- tor) and the latter the ASTR (Aircraft Shield hattan Project, a program was initiated under Test Reactor). joint Atomic Energy Commission and Air Force This paper is a recollection of the aircraft sponsorship to develop nuclear powered air- shielding work done in Fort Worth from 1953 to craft. The program came to be called the ANP 1962, when I was a member of the shielding (Aircraft Nuclear Propulsion) program. Origi- group. Extensive radiation effects studies were nally two modified B-36 aircraft were planned, also performed by other groups. There were theX-6 andthex-9. ‘IheX-6was tobe anuclear also other shielding investigations done after propulsion test bed, and the X-9 was to be a 1962, until the facility was decommissioned in shield test vehicle. In 1950, the Air Force de- 1970. Since this session is devoted to shielding cided to establish a nuclear research facility at history, I will describe what were, in retrospect, Consolidated-Vultee Aircraft Corporation in the most interesting experiments in which I Fort Worth, Texas. The purpose was to investi- participated. Of course, the aircraft was the best gate the shielding and radiation effects ques- photo op, so it is Figure 1, the Nuclear Test tions that would have to be understood to de- Airplane in flight, carrying the ASTR in the aft sign, test and build a nuclear powered airframe. bomb bay for in-flight measurements, which Although the X-6 was canceled in 1953, the X- will be described below. 6 was retained for the airframe research pro- In 1954, an experimental shielding pro- gram. Responsibility for propulsion develop- gram was developed by B. P. Leonard and ment was assigned in separate contracts to myselfwhichincorporated air, ground and struc- General Electric and Pratt & Whitney. ture scattering experiments with three sources: The plant was subsequently renamed a large Co-60 source, the GTR, and finally, the Convair, Fort Worth; later, General Dynamics, ASTR. Shield penetration measurements were Fort Worth. The Nuclear Aircraft Research also planned with the GTR, and were also very Facility (NARF) was to have two reactors: a interesting, as we shall see. The program was copy of the Bulk Shielding Reactor at ORNL carried out from 1954 to March 29,196l. The for shield mockups and ground tests, and a final date is also of interest, and we will mention reactor especially configured as a variable ra- it again.

Ground Experiments The XB-36 aircraft had been stripped of reactors. Measurements were made throughout engines and equipment, and was available, so it the fuselage and inside a rubber and lead shielded was used for some preliminary measurements cylinder (a half-scale shielded crew compart- with a large Co-60 source. After making appro- ment mockup) mounted in the forward bay. priate measurements without the aircraft, the These measurements were then repeated with Co-60 storage cask was positioned beneath the the GTR replacing the Co-60 source at the same fuselage, and the source was lifted out of the approximate location. Figure 2 shows the GTR cask to the centerline of the aft bomb bay, a in an aircraft fuselage. position that would be employed later with the

17 The GTR was also used as a source for air materials were studied in a joint program with and ground scattering measurements by hoist- GE.23 Although some configurations were in- ing the reactor with a crane to a height of about tended as shield mockups, the majority were 30 meters above a concrete ramp.’ Figure 3 systematic studies of the effects on external shows the arrangement. Wide discrepancies dose rates of varying the quantity or position of between measurements and calculations of the a given material either by itself or within an gamma dose rates in air, and particularly in the array of another material. Typically, one might shielded cylinder in the ground level and early have 2 to 10 cm of gamma shielding, such as flight tests, helped to define the importance of tungsten, steel, or depleted uranium, in an array gamma rays from neutron capture, particularly of neutron shielding, such as lithium hydride, the 6 Mev gamma from N16capture. Compari- beryllium, beryllium oxide, with or without son of measurements made at GD in 1955, and boral layers adjacent to the higher density ma- at the ORNL Tower Shielding Facility, which terials. Borated stainless steel, zirconium hy- had just become operational, made the ANR dride, lead, and boron carbide were also tested. shielding community at ORNL, GE and GD For this program GE loaned us what was ptob- aware of the problem, and forced us to make ably the world’s largest concentration of exotic more accurate analyses of secondary gamma shield materials. rays produced in air. To give an idea of the materials studies that The absence of penetration data on bulk were performed, a series of slab arrays are shields lead to the design and installation of the shown schematically in Figure 5. This series GTR Outside Test Tank, shown in Figure 4. was designed to observe the effects of locating This was a water tank, about 4 meters in diam- several slabs of lead at various positions in a eter and height, with a central well for the GTR medium of LiH, boron stainless steel, and BeO. in its small water moderator tank, and with an The lead was sandwiched in boral plate.* Simu- adjacent rectangular oil-filled opening into which lated ducts were also tested, and samples of shield mockups could be placed. Thus, side many aircraft materials were placed between shield arrays of 35 x 35 inch slabs of candidate slabs in various neutron spectra to give a basis shield materials could be placed next to the for estimates of component activation. GTR moderator for neutron and gamma ray penetration measurements in air outside the tank. Using this arrangement, several hundred *Boral is a sandwich of ahuninum and boron carbide slab combinations of gamma and neutron shield used to suppress thermal neutrons.

Flight Experiments

The most exciting part of the program in- flight to alter the radiation distribution from the volved the experiments using the ASTR as a source. radiation source, since it included flight tests in Although many more configurations were a specially-modified B-36. The reactor, shown possible, it was determined in ground mapping in Figure 6, was composed of BSR-type fuel measurements that four configurations shown elements mounted horizontally, supported be- in Figure 7 represented the most interesting tween two grid plates in an 8 l-cm-diameter possibilities for the flight program: all tanks water moderator tank, surrounded by about 8 full, rear tank empty, one side tank empty, and cm of lead on the sides and rear, and 15 cm in the both side tanks empty. front. The lead jacket was enclosed by two concentric annuhrr water tanks on the sides, a A B-36H aircraft was modified for the cylindrical klc on the rear, and five cylindrical flightprogmm,andwascalledtheNTA(Nuclear tanks in the front. All outer surfaces were cov- Test Airplane). The nose was altered to accept ered with boral. The shield was completed by an a four-man crew compartment which was additional 15-cm-thick cylinder of lead mounted shielded with an inner layer of lead and an outer just forward of the reactor assembly. Each of the layer of 15 cm of borated rubber. Figure 8 shows water shield tanks could be drained and filled in the reactor and crew compartment in their rela-

18 tive positions on the ramp before installation in mine the contribution of the aircraft structure by the aircraft. Figure 9 shows the crew compart- extrapolating to zero air density. From such ment being installed in the B-36. The reactor extrapolations, the observed structure effect was suspended on the fuselage centerline in the (-10% to70%, dependingonlocation)isshown aft bomb bay. Special ground support loading for several detectors in Table 1 for neutrons and platforms were installed in a taxi ramp in the gamma rays. reactor area toloadthe ASTR before each flight, To complete the program, the reactor, half- and retrieve it after landing. scalecompartment, andcrewcompartmentwere The aircraft was equipped with neutron and shipped to the ORlVL Tower Shielding Facility gamma dose rate detectors, and thermal neutron for a set of measurements away from the ground detectors at six positions along the fuselage, in in the absence of the aircraft structure. This set the half-scale compartment in the forward bay, of data provided a direct measure of the struc- and at six positions in the crew compartment. ture effect. Figure 13 shows the reactor and Figures 10 and 11 show the detector locations. crew compartment at the ORNL TSF.’ Although some data was obtained manually in Younger members of the audience will be the crew compartment, most was recorded on wondering what happened to the ANPprogram. multitrack mag tape for post-flight analysis. President John Kennedy canceled the program Data aboard the NTA was supplemented March28,1961.Thisdateiseasyfortheauthor with additional measurements made aboard the of this paper to remember: he was co-program chase plane, a B-50 aircraft, which escorted the chairman of the first (and only) unclassified NTA on each flight. Data was recorded aboard symposium on Nucleonics in Flight, sponsored the B-50 at 640meters. Nuclear engineers aboard by the ANS North Texas Section, which started the B-50 to take the radiation data were required on the day following the cancellation. The to undergo parachute training and be prepared opening paper by the keynote speaker was to jump in the event that it became necessary to withdrawn, as were several other papers which jettison the reactor if a malfunction occurred. were to be the first unclassified descriptions of Fortunately, this never happened. nuclear aircraft engines under development. From February 1956 to March 1957, the The GD nuclear group continued research NTA made 17 flights; 16 at altitudes of 3050, in shielding and radiation effects supporting 5200,9250,and11,300meters(10,000,17,000, other nuclear applications until the facility was 26,000, and 37,000 feet), and one flight over the closed in 1970. Gulf of Mexico at 300 meters (1000 feet). ‘I4 Although the ANP program is now only a If all other parameters are constant, the dose footnote to the history of nuclear applications, rate at a given detector shouldvary linearly with it provided a wealth of shielding data and analy- air density. The variations with air density for sis methods for air transport, and shield penetra- neutrons and gamma rays at two detectors are tion, which has greatly benefited all subsequent shown in Figure 12. Empirically, one can deter- work in radiation shielding.

Acknowledgement and Caveat

The challenge in preparing an historical Stout, head of the G.D. Public Relations Office, paper is not to recall events thirty-five years and his staff, for their assistance in locating and ago, but to find illustrative materials that have providing the photographs for this paper. In survived periodic file-purging edicts. I am in- addition, the photo of the ASTR at the TSF was debted to R. E. Adams, W. Park, and W. B. graciously provided by F. J. Muckenthaler. Rose, former G.D. staff members, and to Joe

19 References (Readers are cautioned that some references are not readilyavailable.)

1. C. R. Tipton, Ed., Reactor Handbook,Vol. 477-482, Technical Information Center, I, “Materials,” Section 53.6, N. M. USDOE (1973). Schaeffer,“GTRandASTRScatteringMea- surements,” p. 1136, Interscience Publish- 4. W. B. Self and I. C. Roberts, “Analysis of ers, Inc., New York, 1960. Airborne Nuclear Shielding Data II,” Convair Report NARF-59-15T, FZK-9- 2. K. W. Tompkins, “Analysis of Multilayer 121, March, 1959. Shield Experiments III and IV,” Convair Report NARF-60-7T, MR-N-255, Septem- 5. W. P. Krauter, “ASTR-TSF Experiment: ber, 1960. NTA Station Measurements,” Convair Re- port MR-N-242, August, 1962. 3. N. M. Schaeffer, Ed., ReuctorShieZdingfor Nuclear Engineers, USAEC TID-2595 1, p.

20 Figure 1. Nuclear test aircraft.

Figure 2. GTR in B-60.

21 Figure 3. GTR on crane for air/ground scattering measurements.

22 Figure 4. Outside test facility.

23 Figure 5. OTF materials test configurations, lithium hydride series.

24 Figure 6. ASTR cutaway.

5 I6 CORE WATER Figure 7. ASTR configurations.

25 If! .II3 a Figure 9. Installing crew compartment in NTA.

27 INSTRUMENT CAPSULE CREW SHIELD ASTR- 7 7

241J h+4ALL CYLINDER

Figure 10. Detector locations in fuselage.

NPC 1691 r FUGHT ENGINEER 7 REACTOR OPERATOR r PILOT

RUBBER / PLAN VIEW \ LEAD 9.0 --y r 9.0

L- 9.3 L9.0 I- 0.25 NOTE: CENTER OFODENOTES DETECTOR CENTER OF DETECTION

SIDE ELEVATION

Figure 11. Detector locations in crew compartment.

28 NPC 8458

42

24

b

0 0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 0 0.2 0.4 0.6 0.8 I.0 1.2 1.4 AIR DENSITY (mg/cma) AIR DENSITY (mg/cm’) A. Station 201, Configuration 3 0. Station 281, Configuration 5 4.5

4.0

g? 3.5 L -E 2 3.0

‘?o 2.5 =. /$ 2.0 e i!JiB 1.5

4 Is0

02 0.5

0 0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 0 0.2 0.4 0.6 0.8 1.0 1.2 .’ 1.4 AIR DENSITY (mg/cmj) AIR DENSITY (mg/cm3) C. Station 202, Configuration 3 D. Station 282, Configuration 5

Figure 12. Variations with air density.

29 Figure 13. ASTR at TSF.

30 Table 1. Structure Scattering

FAST NEUTRONS

GAMMA RAYS

202 1 q-5) 22 7.0(-6) 4 3.0(-5) 1

212 2.0(-5) 17 1.5(-5) 7

222 1.3(-4) 42 1.4(-4) 29 1 .O(-3) 18

272 1.9(-3) 76 3.6(-3) 73 2.2(-3) 31

282 6.9(-4) 67 1.3(-3) 69 9.5(-4) 29

6 1.3(-6) 27 4.0(-7) 24 4.0(-5) 8 Where Have the Neutrons Gone -A History of the Tower Shielding Facility

F. J. Muckenthaler Oak Ridge National Laboratory Oak Ridge, Tennessee

Introduction

In the early 1950’s, the concept of the unit been the only reactor facility in the United shield for the nuclear powered aircraft reactor States (U.S.) designed and built for radiation- changed. to one of the divided shield concept shielding studies in which both the reactor where the reactor and crew compartment shared source and shield samples could be raised into the shielding load. Design calculations for the the air to provide measurements without ground divided shield were being made based on data scattering or other spurious effects. Although obtained in studies for the unit shield. It was the Aircraft Nuclear Propulsion (ANP) pro- believed that these divided shield designs were gram was terminated in 196 1, the remarkable subject to error, the magnitude of which could versatility and usefulness of the facility contin- not be estimated. This belief led to the design of ued to make it a valuable tool in resolving other the Tower Shielding Facility where divided- problems in shielding research. This paper dis- shield-type measurements could be made with- cusses that facility, its reactors, and some cho- out interference from ground or structural scat- sen experiments from the list of many that were tering. performed at that facility during the past 38 The Tower Shielding Facility (TSF) at the years. Oak Ridge National Laboratory (ORNL) has

Facility Development

In 1946, the U.S. Air Force decided to reasonable levels, the shielding should be di- implement an ambitious program to produce a vided between the reactor and the crew com- nuclear-powered long-range bomber. After some partment of the aircraft. Since it was not pos- early studies showed that one of the mom sible to do experiments on a divided shield in difficult problems to solve was that of shielding either the Bulk Shielding Reactor (BSR) or the the reactor for the crew’s benefit without sink- Lid Tank, a new research tool was proposed by ing the airplane, and realizing that the ability to ORNL, later to be known as the TSF. design shields with any degree of confidence was minimal, it was decided to enlarge the Design requirements were established to existing shielding program at ORNL. As a limit the ground and structure scattering of result of these concerns, ORNL management neutron and gamma rays into the crew compart- decided to increase the scope of the shielding ment of the aircraft to 15% of the air-scattered work under the direction of E. P. Blizard and dose for an airplane weighing up to 125 tons and others with support from the Air Force. One to have a reactor and crew compartment separa- result was the construction of the Bulk Shield- tion distance between 60 and 100 feet. ing Reactor in order to test the large shields that Several concepts were given consideration, would be required for a reactor that would but the first serious preliminary design for the power an airplane that could weigh up to a half TSF is shown in Figure 1, where the hoisting a million pounds. tower had two legs that were 200 feet high and Further design studies done by the NEPA were placed 200 feet apart. These legs sup project personnel strongly indicated, however, ported a bridge and two hoists for raising and that to reduce the weight of the reactor shield to lowering the reactor and crew shield. The swim-

33 ming-pool-type reactorwas tobeplacedina 12- operating hoists to position the reactor and crew foot-diameter tank of water which weighed 60 shield independently and to permit storage of tons. The reactor could be lowered into a 25 the reactor in the pool. This design met the foot-deep water pool for servicing. The crew nuclear requirements stated previously because shield, including aircraft components, could ‘there were no structures required near the reac- weight up to 75 tons. The operations building tor other than the lifting cables. was to be underground, covered with at least ORNL submitted a preliminary proposal five feet of dirt This building was to house the for the facility which was accepted by the AEC reactor controls, the data-taking equipment, and and the design work was started in July 1952. the hoist controls. This design was developed The constructionwas begun at a site shown in by C. Clifford as project manager with the aid of Figure 3 in March 1953 and completed in the Architect Engineering firm of Knappen, February 1954. The reactor and instrument in- Tibbets, Abbot, and McCarthy. stallations were completedin June of 1954. The The above design was evaluated for the total cost of construction was just slightly under effects of neutron scattering from the massive the two million dollar budget allowed. steel structure and the ground by A. Simon, who had joined Blizard’s group for this purpose. The first TSF reactor (TSR-I) was nearly a Simon predicted that the scattering from these duplicate of the first swimming pool reactor two sources would be too large for this struc- using removable ahuninum alloy fuel elements ture, which led to a second generation design, in a rectangular array. Figure 4 shows the reac- shown in Figure 2, that used Simon’s input to tor, which was supported from atravellingdolly reduce the unwanted effects. This design used that ran on tracks spanning a 1Zfoot water- guyed tower legs to minimize the structure filled tank with a spherical bottom. The reactor required for wind loads and extended the height could be positioned anywhere on the horizontal of the towers to 3 15 feet so that the experiment midplane of the tank, which when filled with could be lifted to 200 feet. A l/60 scale model water weighed 60 tons. The reactor was also of this configuration built by Clifford and Jack designed so that it could be removed from the Estabrook demonstrated the ability of the six tank remotely and be placed in a shield mockup. .

Early Shielding Experiments

Since the facility was built to perform ex- Prior to this experiment, all gamma-ray mea- periments with a divided shield free from exces- surements at the TSF were made using dosim- sive background from ground-scattered radia- eters. For this work, it was desirable to measure tion, it was fitting and proper that the first the spectra of these gamma rays. This was done shielding-typemeasurementshoulddemonstrate with a three-inch sodium iodide crystal using a that the facility did indeed provide that capabil- three-channel analyzerwith 1 MeV energy reso- ity. Results from these measurements indicated lution. This crude approach showed promise, the ground scattering component at 195 feet and the method was later improved by using a was about 2% of the total scattered neutrons. five-inch sodium iodide detector and advanced In late 1954, the first experiments using a electronics. specially designed reactor shield tank were ini- In 1957, this experiment was repeated us- tiated. The reactor was placed in the GE-de- ing a more controlled approach as shown in signed ANPR-1 vessel and both neutron and Figure 5. A collimated beam of neutrons was gamma-ray dose rate measurements were made allowed to escape into the atmosphere and the with the detectors in a square, water-filled tank. resulting gamma-ray production measured was By 1955, the program became more concerned with a well-collimated sodium iodide crystal. with performance of differential-type dose rate The end of the reactor collimator was either bare measurements. or covered with borated plexiglass to obtain In early 1956 a study was made of the spectra for both thermal- and fast-neutron inter- neutron capture gamma rays produced in air. actions with air.

34 Inearly 1958,anAircraftShieldTestReac- An experimental method for the optimiza- tor (ASTR) experiment was performed in coop- tion of a divided neutron shield was developed eration with Convair/Fort Worth to obtain data for the TSF using reactor shield and simulated for comparison with that obtained from their crew shield tanks, each having compartments own Nuclear Test Airplane program. For this that could be filled with water or drained re- mockup (see Figure 6) the ASTR was attached motely (see Figures 7 and 8). Results from these to an aluminum frame and the crew compart- measurements were used to determine the de- ment attached at the other end 65 feet away. gree of asymmetry of the reactor shield that This combination was then raised to 200 feet to would give a minimum shield weight. simulate flight conditions.

TSR-II Design

Because of the difficulties in calculating Director Alvin Weinberg suggested that the shield performance in the early years, it became new reactor should be a general purpose source, the practice to test shield designs using a known something that could be used to cover problems neutron source and full scale mockups of the as- of importance to shields for any reactor cycle designed shield. The effectiveness of this shield that might present itself. All of the requirements in conjunction with the reactor design would mentioned led E. P. Blizard to suggest that the then be predicted by making corrections for the reactor be spherical to provide an isotropic differences in source terms. The power source source. If spherical, reactor controls would selected for propulsion of the aircraft early in present a problem and C. E. Clifford proposed the program was a circulating-fuel, reflector- that the control plates be internal to the core to moderated reactor (CFRMR) being designed by minimize perturbation of the leakage flux. Its Pratt and Whitney Aircraft, something consid- design was directed by Clifford with the assis- erably different than the swimming-pool-type tance of L. B . Holland and Charles Angel. What reactor used at the TSF. This difference in materialized became known as TSR-II, a sche- reactor shapes and compositions would have matic of which is shown in Figure 9. The new made it very difficult to apply the proper correc- reactor became operational in February 1961 at tions. It was proposed that the shield measure- the approved maximum power of 100 kW. ments should be made with a source whose Eleven years later, approval was given to oper- radiation leakage was similar to that from the ate the reactor at 1 MW, the maximum power CFRMR. It was also suggested that a new achieved during its lifetime. The reactor was Shield Mockup Reactor be built having a beryl- used for only one experiment in the ANP pro- lium reflectorshapedliketheproposedCFRMR. gram, a divided shield mockup study for Pratt A design was submitted to Laboratory manage- and Whitney Aircraft Company, before the ment but this concept was rejected. Research ANP program wascanceled in June 1961.

Follow-On Experimental Programs

The loss of ANP support was not “deadly” and Office of Civil Defense in 196 l-62 to study to the TSF as its usefulness had already attracted the shielding effectiveness of the covers being attention from other programs. An agreement designed for missile silos that were under con- had been reached between the United Kingdom struction. Concern was centered on reducing and the U.S. Defense Atomic Support Agency costs while maintaining the safety of personnel (DASA) to measure the attenuation characteris- and equipment in the silos. Other experiments tics of specific military vehicles and research for DASA followed, including measurements models of mutual interest during 196 1. These of the angular dependence of fast-neutron dose measurements involved personnel from the U.S. rates and thermal-neutron fluxes reflected from Army Tank Automotive Center, General Dy- concrete, using a mockup shown in Figure 10. namics/Fort Worth, and ORNL. A specific se- Analytical methods that were verified in that ries of measurements were made for the DASA experiment were used to calculate the neutron

35 flux as measured in large concrete ducts con- was designed and fabricated by Atomics Inter- taining one, two, and three bends as shown in nationaLwent critical at the TSF in April 1967. Figure 11. Measurements were made of the neutron spec- About 1964, anNE-213 spectrometerbeing tra above 1 MeV transmitted through typical developed by V. V. Verbinski was introduced at SNAP shielding materials placed beneath the the TSF as a useful tool to measure neutron reactor. These materials included lead, =alJ, spectra above about 800 keV. Its presence was tungsten powder, hevimet, lithium hydride, and soon joined by that of the specially designed laminated slabs of those materials. This was Benjamin-type hydrogen-tilled detectors for followed by measurement of the gamma-ray measurement of the neutron energies from about spectra transmitted through these same materi- 40 keV to 1.5 MeV. With the advent of these als before the SNAP program at the TSF was spectrometers, the value of the data used for terminated in 1971. testing the analytical methods, such as F. R. From late 1970 to 1975, a large portion of Mynatt’s discrete drdinates code DOT, was the TSF’s time was devoted to the Fast Flux greatly improved. Later, in 1970, the Bonner Test Facility (FFIF) program as part of the ball detector system was developed to provide Liquid Metal Fast Breeder Reactor program. a measurement of the integral neutron flux. The Experiments were proposed to aid in the devel- BF,-filled spherical detector was surrounded opment of analytical techniques for calculating by spheres of polyethylene having different the neutron penetration through specific areas thicknesses, each combination being sensitive of the FFTF. Of particular interest were mea- to a given neutron energy region. The presence surements of neutron streaming through planar of this combination of detectors eliminated the and cylindrical annular slits in thick iron shields need for dose rate measurements. that represented the top head of the reactor Considerable time and effort was spent shield. Various integral experiments were done performing further programs for DASA that that included demonstration plant shield studies included: (1) the thermal-neutron capture proposed by GE and Westinghouse. Experi- gamma-ray spectral intensities from various ments were run to test the calculations by GE shielding materials; (2) investigation of the that radiation transport through shields of large minima in total cross sections for nitrogen, diameter stainless steel or boron carbide filled oxygen, carbon, and lead; (3) measurement of rods could be adequately described using a the angle-dependent neutron energy spectra homogeneous model. By 1974, that work was emergent from large lead, polyethylene, de- completed and further efforts in the LMFBR pleted uranium, and laminated slab shields; and program were applied to the Clinch River (4) gamma-ray spectra arising from thermal- Breeder Reactor (CRBR) program. neutron capture in elements found in soils, The experimental capabilities at the TSF concrete, and structural materials. were greatly improved in 1975 by addition of a In the late 1960’s, work was initiated in new reactor shield. Prior to this, the reactor was support of the Space Nuclear Auxiliary Power contained in the reactor shield ball that was (SNAP) program. Measurements were made of faced by concrete to minimize scattering and the scattered neutrons from support tabs and provide a flat surface against which the mockups beryllium control drums that would be ex- could be placed. The new shield permitted tended beyond the shadow shield when the placement of the mockups much closer to the SNAP reactor was in flight. This work was reactor and, with an increase in collimator di- extended to include measurement of the fast- ameter, the source strength incident upon the neutron transmission through an “infinite” slab mockups was increased by a factor of 200. The of lithium hydride and measurement of the fast- change altered the description of the source neutron dose rate transmitted through a coni- term from a previously considered point source cally shaped lithium hydride shield. to a disk source. In 1965,it wasproposedthattheTSFinstall The first experiment using the new shield a modified SNAP 2/10A reactor to provide was a measurement of the CRBR upper axial experimental results for comparison with Monte shield mockup. The full strength of the new Carlo calculations of the radiation penetration source term was used to measure neutron trans- through shadow shields. The reactor, which port through 18 inches of stainless steel and 15

36 feet of sodium followed by 35 inches of iron. During 1984, the TSF became involved in Measurements followed of the neutron stmam- studies of the radiation exposures received from ingin a CRBR prototypic coolant pipe chaseway the atom bomb explosions over Hiroshima and (see Figure 12) and the program ended with Nagasaki by mocking up a concrete structure comparison measurements of the neutron at- simulating a small single-story concrete block tenuation through stainless steel and inconel house and making internal measurements. The when used as a radial blanket shield. data provided verification of the discrete ordi- nates Three-dimensional Oak Ridge Transport By 1977, the Gas Cooled Fast Reactor Computer Code (TORT) being developed at program replaced the CRBR work with a desig- OFWL. nated program of eight specific experiments that continued until 1980. They included ex- In November 1985, the TSF began partici- periments on the grid-plate shield design, the pation in a cooperative experimental program radial blanket configuration, the shield hetero- betweenthe JapanesePowerReactorandNuclear geneity mockup, the exit shield, and the plenum Fuel Development Corporation (PNC) and the shield designs. U.S. DOE. The program, entitled JASPER for Japanese-American Shielding Program of Ex- In 1983, studies for the High Temperature perimental Research, was designed to meet the Gas Cooled Reactor program were initiated to needs of both participants. Eight experiments provide data for verification of the analytical wereplanned, withthefirsttworeachingcomple- methods used to calculate the radiation trans- tionbeforetheTSF wasshutdownin 1987.The ported through the bottom reflector and core remaining six experiments were completed this support structure and to predict the radiation September following resumption of reactor damage at the support structure. operation in 1990.

Acknowledgements

The success of the operation of the TSF However, with the loss of the ANP pro- throughout its lifetime can only be attributed to gram, this outside participation dwindled since the efforts of the people involved, both those a majority of the supporting companies were employed at ORNL and people from outside also left without ANP funding. Participation by vendors who became extensively involved in the TSF in new and diversified programs also various aspects of the many programs. Changes brought some changes in the vendors collabo- in personnel occurred as time went on, expand- rating in the TSF work, such as Westinghouse, ing and decreasing in numbers as the effort General Atomics (now called GA Technolo- required. Outside vendors, such as General gies), Atomics International (now known as Electric (GE), the National Advisory Commit- Rockwell International), GE, General Dynam- tee for Aeronautics, and the Boeing Airplane ics, Radiation Research, the University of Ten- Company, to name a few, provided personnel nessee, SAX, and personnel from the Japanese soon after the facility was built to aid in the Power Reactor and Nuclear Fuel Development experiments and performance of the data analy- Corporation (PNC). All who have participated sis. Support from outside the Laboratory con- in the programs have been generous in their tinued to increase as personnel from other com- efforts to promote the welfare of the facility. It panies like Consolidated Vultec Aircraft Cor- is to all of them that I want to say thank you for poration, Glenn L. Martin Company, Pratt and helping to make the facility what it has been Whitney, Convair/San Diego and Convair/Fort throughout all these years. Worth, the Lockheed Aircraft Company, and others added their contributions.

37 Bibliography

1. D. T. Ingersoll andN. Ohtani, “JASPER: A 8. R. E. Maerker and F. J. Muckenthaler, Joint U.S.-Japan Program of Experimental “Gamma Ray Spectra Arising from Ther- Shielding Research,” ANS Topical Confer- mal Neutron Capture in Elements Found in ence on Theory and Practices in Radiation Soils, Concretes, and Structural Materi- Protection and Shielding, Knoxville, TN, als,” ORNL-4382 (August 1969). April 22-24, 1987; &UC. Vol. 2, p. 346 (1987). 9. C. E. Clifford et al., “Neutron Transport Studies in Ducts and Single Shields,” 2. C. 0. Slater, F. J. Muckenthaler, and D. T. ORNL/TM- 1965 (October 1967). Ingersoll, “Experimental Verification of Calculated Neutron Streaming in the Bot- 10. W.E.Thompson,“HistoryofORNL(l943- tom Reflector and Core Support Regions of 63),” ORNLKF-63-8-75, LRD-RC (Au- a Large Scale HTGR Design Concept,” gust 1963). Nucl. Sci. Eng. 97(2), 123 (1987). 11. E. P. Blizard et al., “Determination of Ra- 3. Engineering Physics and Mathematics Di- diation Intensities in Missile Silos,” ORNL/ vision Progress Report, ORNL-62 14 (June TM-253 (1962). 1985). 12. Engineering and Mechanical Division, 4. F. J. Muckenthaler et al., “Verification “Completion Report on the Tower Shield- Experiment of the Three Dimensional Oak ing Facility,” ORNLKF-55- 1- 165 (Febru- Ridge Transport Code (TORT),” ORNL/ ary 1955). TM-9528 (December 1985). 13. E. P. Blizard et al., “Tower Shielding Facil- 5. L. S. Abbott, D. T. Ingersoll, F. J. ity Preliminary Proposal Number 18 1,” Muckenthaler, and C. 0. Slater, “Review of ORNLKF-52-3-125 (May 1953). ORNL-TSF Shielding Experiments for the Gas-CooledFast BmederReactorPmgram,” 14. Oak Ridge National Laboratory, “Tower ORNL-5805 (January 1982). Shielding Facility, Proposal Number 18 l- A,” ORNLKF-53-l-286 (January 1953). 6. F. J. Muckenthaler et al., “A Measurement of the Neutron Streaming in a CRBR 15. F. R. Mynatt, F. J. Muckenthaler, andP. N. Prototypic Coolant Pipe Chaseway,” Stevens, “Development of Two-Dimen- ORNL/TM-5901 (August 1977). sional Discrete Ordinates Transport Theory for Radiation Shielding,” Computing Tech- 7. L. S. Abbott et al., “Review of ORNL nology Center, Union Carbide Nuclear Di- Radiation Shielding Analysis of the Fast vision, CTC-INF-952 (1969). Flux Test Facility Reactor,” ORNL-5027 (July 1975).

38 , .

Figure 1. Early concept for the Tower Shielding Facility.

Figure 2. Final concept for the Tower Shielding Facility. 40 ,

_ -. -- CONE OF RADIATION FROM REACTOR COLLIMATOR

FIELD OF “VISION” OF DETECTOR

CYLINDRICAL ,/ REACTOR COLLIMATOR

FEET

:TOR COLLIMATING SLIT

Figure 5. Experihental configuration for measuring gamma rays produced &om neutron captures in air.

41 Figure 6. The ASTR suspended in air at the TSF.

42 REACTOROOLLI -1 \

U- l”LEICOOLI”O.!+IER

’ II- OUTLET COOLIHOYATER

- cowl! 1 ,COw*RTYCHTs ,.a,

CONE I (COYPARTYENTS 37.41) ,Cow*RTYENIs w.24,

CONE 6 ,COw*RTYENTs SbSO

Figure 7. Compartmentalized reactor shield tank.

Figure 8. Compartmentalized crew compartinent tank. ORNL-LR-DWG 32708R4 CONTROL TURRET--+-

t WATER 3UTLET

/TURBINE FLOW METER/

DRIVE GEARYL

FISSION CHAMBER I AN0 PREAMPLIFIER 3/----t

HELICAL COOLING-WATER TUBES -

INCHES

LEAD-WATER SHIELDING

‘/z-in. WATER ANNULI

‘h-in WATER ANNULUS

ARROWS INDICATE THE DIRECTION OF COOLING- WATER FLGW --

Figure 9. Schematic of second TSF reactor (TSR-II).

44 Figure 10. Experimental arrangement for measurements of fast neutrons reflected from concrete.

Figure 11.Experimental arrangement for measurements within a duct with two bends.

45 ORNL- DWG 7?- (0453

DIMENSIONS ian) : E-Q = 163.8 M-N = 33f.2 F-GH = 30.5 O-J= 45.7 J-F= 320.0 CHOKE ID - 152.4 L-O = 434.3 CHASEWAY HEIGHT- 289.6 L-R= 335.3

CHOKE NO. 1

55.9

L.

I-- (52.44

243.8

Figure 12. Schematic of pipe chaseway configuration.

46 History and Evolution of Buildup Factors

David K. Trubey Consultant Citrus Springs, Florida

Abstract

The term “buildup factor” has been in use but new results are available based on modem since the Manhattan Project to account for cross sections and which take into account the scattering in the representation of the gamma- various secondary radiations. Various fitting ray attenuation function. Buildup factors am functions representing the infinite medium data used extensively in point kernel codes for use in have been in use in the point kernel codes for shield design. The data from the Goldstein- many years. Fitting functions have also been Wilkins moments method calculations of 1954 devised to use infinite medium buildup factors have been the standard data in use until recently, for the design of multilayer shields.

Introduction

According to. Goldstein, ’ the gamma-ray scatteringisthemostfrequentpmcess forgamma buildup factor is a “term whose origin is lost in rays in the energy range 0.1 to 10 MeV, espe- the mists of the early history of the Manhattan cially in low-Z materials. A photon may easily Project.” Its introduction stems from the obser- experience five to ten successive Compton col- vation that the calculations for the uncollided lisions before its eventual outright absorption, photons, i.e., those which have arrived at R which most frequently occurs in a photoelectric without suffering any collisions, are usually a process. The photon energy decreases in each relatively simple matter, involving only an ex- scattering. The gamma-ray energy suffers ponential kernel. The buildup factor is then a thereby a progressive degradation. The average multiplicative factor which corrects the answer fractional energy loss in a Compton process which is proportional to the uncollided flux decreases in the course of degradation. There- density so as to include the effects of the scat- fore, the photons tend to accumulate in the tered photons. To define the term formally, let lower portion of the spectrum down to the the superscript 0 refer the tmcollided photons. energy range where outright photoelectric ab- Suppose what is sought is some functionalfly); sorption becomes predominant. then the buildup factor B with respect to w is The most common buildup factors are for defined by calculating the quantities dose and energy ab- f(v) = B f

47 B,WJ = [h (W-9~~r0 ml 1 where K is called the point kernel, J.L&?$)is the kerma response or flux-to-kerma rate function [h$‘(RJ9~~0 ml. (2) and p0 = p(EJ is the attenuation coefficient where p&I?)is a weighting or response function (total cross section). A number of point kernel at energy E and normally is the energy deposi- codes have been developed which calculate the tion coefficient for a particular material. The energy absorption, kerma, or dose at specific usual ones yield kerma for air, tissue, or the locations in a 3dimensional geometry by per- particular medium for which the flux density forming a numerical integration of Eq. 3 over a was calculated. source volume. The buildup factor can be evalu- For application to a point source in an ated from a fitting function, such as given in i&rite, homogeneous medium, the kerma rate Table 1 or from interpolation in a table of at a distance R from a source of 1 photon per buildup factor data. second of energy E. is W&l = pk (Ed expGpJ0 Bk~&EJl(4~2), (3)

Source of Buildup Factor Data Prior to 1954, values of the buildup factor similarlyasafunctionofenergy.Bythismethod, were estimated for shielding calculations by the values of ZeBfor various concrete mixtures approximatemethods.Forexample, somemanu- have been determined to be in the range 11 to als suggested using the absorption cross section 27; the lower values correspond to ordinary instead of the attenuation coefficient for the concretes and upper values correspond to the argument of the exponential, i.e., heavy concretes. Results of this type are given by Walker and Grotenhuis7 which provided B&W exp<-@I = exp(-@L (44) buildup factor data for concrete in use for many where p0 is the attenuation coefficient and p, is years until data became available from direct the absorption coefficient. As mentioned above, calculation. G-W published buildup factors and infinite mediumspectrafiommomentsmethodcalcula- Recently, ANS-6.4.3,* abuildupfactorcom- tions for point and plane sources in 1954. Addi- pilation intended to replace G-W as standard tional moments method results have been pub- reference data, has become available. ANS- lished by Spencer, Eisenhauer, etaZ.4Other data 6.4.3 data cover the energy range 0.015-15 have been published by many authors calcu- MeV to 40 mean free paths (mfp) for 22 ele- lated by various methods, such as the reports by ments and 3 mixtures: air, water, and ordinary Penkuhn Many of the authors published only concrete. Most of the data in the standard have a few values to validate their methods by com- not been published previously. The buildup parison to the G-W results. Unfortunately, very factor data for elements below molybdenum little experimental data am available due to a come from moments method calculations of lack of monoenergetic sources except for 137Cs Eisenhauer et aI. The data for the higher atomic and %o. Additional data for 6-MeV photons number elements are from PALLAS calcula- are available from the experiments of Bishop et tions of Sakamoto and Tanaka.’ Unlike most aL6 previous work, these data include important secondary radiations, particularly bremsstrahl- For mixtures, such as soil or concrete, G-W tmg and fluorescent radiation. Considerable recommended that an effective atomic number detail is given near the absorption edges where Z& be deduced by comparing the shape of the the buildup factor gets very large, mostly due to attenuation coefficient for the mixture as a the discontinuous photoelectric cross section, functionofenergywiththecormspondingcurves but also to the emergence of fluorescent radia- for individual elements of known atomic num- tion. In addition to the buildup factor data, there ber Z. In addition to general agreement in the are tables of correction factors to account for the shape of the attenuation coefficient curves, it is change in spectra near the shield-tissue inter- desirable that the curves of the ratio of the face and for the neglect of coherent scattering. Compton cross section to the total cross section These tables are from ASFIT calculations in (attenuation coefficient) should also behave h&a.lOvll

48 Table 1. Buildup Factor Fitting Functions

1. Infinite medium (x in mfp), point source Taylor B(x) = A, exp(a,x) + A, exp(-a& A,=l-A,

Linear B(x) = 1 + Ax

Quadratic B(x)= l+Ax+B2

Polynomial (Cap0 )

WA = : ~mGW m==o Berger B(x) = 1 + Cx [exp (Dx)]

Geometric-Progression (G-P) B&x) =l +@- l)(K”-l)/(K-1) forK#l and B(E;s) =l+(b- 1)x for-K= 1 K(E;s) =cx + d [tanh(x& - 2) - tanh(-2)] / [l - tanh(-2)]

2. Multiple slab or laminated media (each layer xi mfp thick) Kales (lead followed by water, slab geometry)

Wl;r,) = B2<“2> + [B,(x,>-~I[B~(x,)-~I-~[B2(~I+X2)-B2(~2>l

Bowman-Trubey (2 layer) l

B<~JC~> = B,(x,)B,(x,) exp(-x2) + B,(x,+x,) P-exp(-x,)1

Broder (N layers)

N N N B(Zx.J = Z B,( : xi) - C 6J?xi) ?l=l 1 n=2 1

Harima-Hirayama (multilayer, x1 includes all layers but last)

B(x,J~) = B1(xI>B2(x1+x2>.~x,JC2>

j(x,,x,) = 1 - xyb / a, for lead-water

j(x1,x2) = a / (a+xt) , for water-lead

log a = o(log x$2 + p(log x1) + y

log b = &(log x1) + e

49 Fitting Functions

The most popular fitting functions used in pmach has been very popular in a number of shielding applications are given in Table 1. codes because the coefficients p are also fit by Probably the earliest formula used for a buildup polynomials expanded in powers of l/E, mak- factor was the linear form. This form can be ing interpolation in energy very easy. Unlike all used for thin shields because most of the scat- the other formulations considered here, Capo’s tered flux is due to single scattering which has coefficients result in an expression which does a llrbehavior near the source. This implies that not reduce to exactly one for @ = 0. the scattered part of the buildup factor is propor- A two-parameter formula proposed by tional to r since the uncollided flux is propor- Berger16 and reintroduced by Chilton17 has the tional to l/i?. It is interesting to note that in Ref. simplicity of the liner form but fits the ,buildup 1, Goldstein pointed out that for the case of the factor data well over a long range. This formula energy absorption buildup factor, the value of has the advantages of simplicity and reasonable the one parameter can be determined by assum- accuracy and, like the Taylor form, can be ing conservation of energy and integrating over integrated over simple source geometries. A all space. He showed that the coefficient A is plot of the exponential parameter as a function given by of energy passes through 0 for some elements, whichmeans that the Berger form reduces to the where p,is the energy deposition coefficient for linear form in these cases. Trubey” determined kerma. the value of the parameters for the G-W buildup factors and made extensive comparisons with Even before the time of G-W, various pub- other forms. He found the Berger formulation to lications such as Rockwell12 listed parameters be preferable to the Taylor form, then in com- for the Taylor13 form. Its accuracy in reproduc- mon use, because the two terms are physically ing the published buildup factors is fairly good meaningful (separate uncollided and scattered for the higher energies and the middle to higher flux terms) and the parameters are slowly vary- atomic number materials, but fails to give ac- ing. ceptable values at lower energies and atomic The most recent form, the G-P19~20func- number. Interpolation of the parameters in en- tion, is the most accurate of all the popular ergy or atomic number is essentially impos- forms due to its having 5 parameters and having sible. It is the nature of a sum of exponentials some basis in the transport physics. It was that greatly different coefficients can give simi- selected forthe ANS-6.4.3 compilationbecause lar results. Foderaro and Hall14 proposed a 3- the buildup factors couldbe reproducedovertbe texm form which has been shown to be quite whole range of energy, atomic number, and accurate for water, a difficult case, but param- shield thickness to within a few percent. The eters are not available for other materials. parameter values can be interpolated in energy The use of a 4-term polynomial, capable of and atomic number with good results. The G-P good accuracy, became generally feasible when function, with interpolation in energy, has been Cape” published a rather complete set of coef- incorporated in the QA@l point kernel code as ficients for many materials in 1958. This ap- well as others,

Finite Media and Laminar Layers

All the forms discussed above reproduce factor may give a value that is too high. Cbilton infinite medium results, but most shields are et aLz give correction factors based on calcula- composed of layers of different materials and tions of Berger and Doggett.23 The correction have an outer boundary. It is rather surprising that the point kernel codes work as well as they factor,definedas(Bslab- l)/(Btiti,-l),canbe do. as small as 0.75 for water and 0.4 MeV but Many times with a shield of finite thick- increases to about 0.98 at 10 MeV. It is greater ness, the use of an infinite medium buildup than 0.97 for lead at all energies above 0.4 MeV.

50 For laminated shields, it was probably E. P. In 1962, BtoderaO proposed a complicated Blizard who first suggested that the last layer formula for any number of layers as shown in buildup factor be used but with the total number Table 1. For the case of 2 layers, it reduces to of mean free pathsU The gamma ray tends to forget the nature of the first materials. Chilton et B(x+~)= B2+ [Bl(xl) - B2(x1)l.(6) al. give a commonly used rule for laminations It can be seen that this is similar to Blizard’s of two different materials, of thicknesses xi and prescription, but with a correction term that x2 mfp, numbered in the direction from source depends on the difference of the buildup factors to detector as follows: If 2, c Z2, the overall for the two materials with the thickness of the buildup is approximately equal to the buildup first layer as the argument. factor B, for the higher-Z medium with the use d of x + x as its argument; if Z, > Z2, the overall As x gets large, Bnxler’s formula does not buil&.tp factor to use is the product B (x ) rimes account f or the final saturation buildup in the B (x ). The laminations should eachbe’at least last layer which should be approximately that of o6e mean-free-path thick, and the source pho the last material alone. Kita7ume31 therefore ton energy is used as the energy argument for all proposed to multiply the correction term by a tabulated values. decaying exponential, i.e., If the laminations are many and thin, the B(~J~) = B2 + [B,(x,) shieldmay be considered homogeneous and the - B,(x,)lexp[-orx,lt 0 effective Z method mentioned above may be used. Wood% describes a code to accomplish where 01 is a parameter to be determined by this. calculations or experiment. Btinemann andRich- A number of authors have examined the tes2 qxxtecl, based on calculations of Penkuhn values of the buildup factor in the case of and Schubart, that c1is in the range O-3 and that layered shields of different atomic number. the Kitazume formula showed good agreement Various functions, such as those shownin Table for water-aluminum-iron shields but poor R- 1, have been proposed. sults for a water-lead shield. In a 1967 article, Kitamme and Ogu~hi~~recommended adding Probably the first multilayer formulas found another correction term with 2 more parameters in the literature are the ones devised by Kales” for light-heavy shields and energies above 2 based on results of Monte Carlo calculations. MeV. In their experiments, they studied 6oCo The formula shown in the table is for transmis- gamma-ray penetration of water, steel, and lead sion through a lead-water slab shield and a shieldsarrangedinupto4layersandupto8mfp normal-incidence source. The water-lead for- in thickness. They found that their formulas fit mula (not shown here) is far more complicated. the experimental data to within 15%. In an independent set of Monte Carlo calcula- tions, Bowman and Trubeyz found that the In recent work, Harima and Hirayama34 Kalos formulas agreed with their results to introduce a new formula and give evidence that within 15% in most cases over a range of l-10 multilayer shields canbe represented by 2-layer MeV and 1-6 mfp for both lead-water and shields where the first layers are all represented water-lead shields. by the next-to-last layer in terms of their total mfp. Thus, their 2-layer form can be used for The Kalos formulas do not apply for energy any number of layers, and it will be incorpo- absorption (heating) in a slab. Bowman and rated into a new versipn of the QAD code. Trube~~~proposed the formula shown in Table 1 for this application, and it seemed to be Based on recent publications and ongoing adequate when compared to the results of their research, it is apparent that the venerable buildup Monte Carlo calculations. factor concept is alive and well.

51 References

1. H. Goldstein, Fundimental Aspects of Re- 10. D. V. Gopinath, K. V. Subbaiah, and D. K. actor Shielding, p. 162, Addison-Wesley Trubey,“GammaRayTransportinaShield- Publishing Co. (1959); reprinted by John- Tissue Composite System and Buildup son Reprint Corp. (1971). Factor Implications,” NucZ. Sci. Eng. 97, 362-373 (1987). 2. H. 0. Wyckoff, R. J. Kennedy, and W. R. Bradford, NucZeonics 3,62 (1948). 11. K. V. Subbaiah, A. Natarajan, and D. V. Gopinath, “Impact of Coherent Scattering 3. H. Goldstein and J. E. Wilkins, “Calcula- on the Spectra and Energy Deposition of tions of the Penetration of Gamma-Rays,” Gamma Rays in Bulk Media,” NucZ. Sci. USAEC Report NYO-3075 (1954). Eng. 101,352-370 (1989).

4. See, for example, L. V. Spencer and J. 12. T. Rockwell, III, Ed., Reactor Shielding Lamkin, “Slant Penetration of y-Rays in Design Manual, D. Van Nostrand (1956). H 0,” NBS Report 5944, National Bureau o I Standards (1958) or C. M. Eisenhauer 13. J. J. Taylor, “Application of Gamma-Ray and G. L. Simmons, “Point Isotropic Buildup Data to Shield Design,” USAEC Buildup Factors in Concrete,” Nucl. Sci. Report WAPD-RM-217 (1954). Eng. 56,263-270 (1975). 14. A. Foderaro and R. J. Hall, “Application of 5. See, for example, H. Penkuhn, “The Three-Exponential Representation of Pho- Gamma-Ray Buildup Factors in Ordinary ton Buildup Factors to Water,‘* NucZ. Sci. Glass and Heavy Lead Glass,” NucZ. Sci. Eng. 74.74-78 (1981). Techrwl. 1,809-830 (1979). 15. M. A. Capo, “Polynomial Approximation 6. See, for example, G. B. Bishop, “Buildup of Gamma-Ray Buildup Factors for a Point Factors and Scalar Flux Spectra for 6.2 Isotropic Source,” USAEC Report APEX- MeV Photon Penetration Through Two- 510 (1958). Layered Shielding Slabs,” Ann. NucZ. En- ergy 9,275-279 (1982). The experimental 16. M. J. Berger, “USNRDLReviews andLec- facility is described in G. B. Bishop, “A 6 hues No. 29,” Proceedings of Shielding MeV Gamma Photon Facility,**NucZ. fnstr. Symposium held at the Naval Radiological Methods 121,499-504 (1974). Defense Laboratory, p. 29 (1956).

7. R. L. Walker and M. Grotenhuis, “A Sur- 17. A. B. Chilton, D. Holoviak, and L. K. vey of Shielding Constants for Concrete,” Donovon,“Determination of Parameters in USAEC Report ANL-6443, Argonne Na- an Empirical Function for Buildup Factors tional Laboratory (196 1). for Various Photon Energies,” US Navy Report USNCEL-TN-389 (AD-249195) 8. AmericanNationalStandard,“Gamma-Ray (1960). Attenuation Coefficients and Buildup Fac- tors for Engineering Materials,” ANSI/ 18. D. K. Trubey, “A Survey of Empirical ANS-6.4.3-1991, American Nuclear Soci- Functions Used to Fit Gamma-Ray Buildup ety (1992). Factors,” ORNLZRSIC-10, Oak Ridge Na- tional Laboratory (1966). 9. Y.SakamotoandS.Tanaka,“QAD-CGGP2 and G33-GP2: Revised Versions of QAD- 19. Y.Harima,“AnApproximationofGamma- CGGP and G33-GP Codes with Conver- Ray Buildup Factors by Modified Geomet- sion Factors from Exposure to Ambient and ric Progression,” NucZ. Sci. Eng. 83, 299- Maximum Dose Equivalents,” JAERI-M 309 (1983). 90-110, Japan Atomic Energy Research Institute (June 1990).

52 20. Y. Harima, Y. Sakamoto, S. Tanaka, and 28. L. A. Bowman and D. K. Trubey, “Strati- M. Kwai, “Validity of the Geometrical fied Slab Gamma-Ray Dose-Rate Buildup Progression Formula in Approximating Factors for Lead and Water Shields,” Gamma-Ray Buildup Factors,” NucZ. Sci. USAEC Report ORNL-CF 58-1-41, Oak Eng. 94,24-35 (1986). Ridge National Laboratory (1958).

21. RSIC Code Package CCC-493/QAD- 29. L. A. Bowman and D. K. Trubey, “De,& CGGP, “A Combinatorial Geometry Ver- tion of Gamma Ray Heating in Stratified sion of QAD-PSA, A Point Kernel Code Lead and Water Slabs,” USAEC Report System for Neutron and Gamma-Ray ORNL-CF 58-7-99, Oak Ridge National Shielding Calculations Using the GP Laboratory (1958). Buildup Factor,” OakRidge National Labo- ratory. 30. D. Ll Broder, Yu. P. Kayurin, and A. A. Kutrezov, Soviet J. At. Energy 12, 26-31 22. A. B. Chilton, J. K. Shultis, and R. E. Faw, (1962). PrincipZesofRadiationShieZding,Prentice- Hall, Inc., p. 197 (1984). 31. M. Kitazume, Nippon Genshiryoku Gakkaishi (J. At. Energy Sot. of Japan) 7, 23. M. J. Berger and J. Doggett, “Reflection 496 (1965) (in Japanese). and Transmission of Gamma Radiation by Barriers: Semianalytic Monte Carlo Calcu- 32. D. Btinemann and G. Richter, op. cit., Sec. lation,” J. Res. NatZ.Bur. Stand. 56,89-98 4.3.2.3. (1956). 33. M. Kitazume and A. Ogushi, “Buildup 24. D. Btinemann and G. Richter, “Multilay- Factors for Gamma-Ray Penetration ered Shields,” in Engineering Compendium ThroughMultipleLayersofReactorShield- on Radiation Shielding, Vol. I, R. G. Jaeger ing,” Hitachi Hyoron 7, l-8 (1967) (in (ed.), Springer-Verlag (1968). Sec. 4.3.2.3. Japanese); translation published as USAEC Report ORNL-tr- 1906, Oak Ridge National 25. A. B. Chilton, J. K. Shultis, andR. E. Faw, Laboratory. op. cit., p. 198. 34. Y. Harima, H. Hirayama, S. Tanaka, and 26. J. Wood, Computational Metho& in Reac- M. Sugiyama, “The Behavior of Gamma- torShiekfing,PergamonPmss(1982),p.106. Ray Buildup Factors in Stratified Shields,” Proceedings of Topical Meeting, New Ho- 27. M. H. Kalos, “A Monte Carlo Calculation rizons in Radiation Shielding, April 26- of the Transport of Gamma Rays,” USAEC May 1,1992, Pasco, Washington, pp. 473- Report NDA 56-7, Nuclear Development 479, American Nuclear Society. Associates (1956). Also reported by H. Goldstein, op. cit.

53 Early Shielding Research at Bettis Atomic Power Laboratory

Kalman Shure and 0. J. Wallace Westinghouse Bettis Atomic Power Laboratory West Mifflin, Pennsylvania

Summary

ReminiscencesofshieldingresearchatBettis wisdom,asbriefquotes:“Themethods...inthis always have in the background the reason for .. . manual . .. have been used for, and tested on, the existence of the Bettis Laboratory - the teal power reactor shields... A fact . . . important design of efficient and safe reactors. Shielding in practical shield design is that water .. . has no is essential for personnel safety. However, the cracks.” only computational tools available in the early 1950’s were slide rules and desk calculators. Another famous warning (by Herbert Under these conditions, any shield design cal- Goldstein) at about this time may be para- culation accurate within a factor of two was a phrased as follows: “Beware of blind adherence good one and the phrases “Close enough for to incorrect formulae which have been put into shielding purposes” and “Including a factor for print, or results printed on computer paper, conservatism” became a permanent part of the which have thereby been sanctified.” shielding vocabulary. These warnings reflected the fact that the This early work instilled a respect for hand first crude digital programs were becoming calculations and the requirements that any re- available, whichmademorecomplicatedshield- sult, no matter how calculated, must meet the ingcalculationspossible. Someengineerstended test of being reasonable and in line with previ- to place too much confidence in computed ous experience. Even today, with sophisticated numbers, while others refused to accept that shielding programs available on the latest com- computations would replace most hand calcula- puters, calculated results must pass the same tions. There were early advocates of computer test. use in shield design, as well as those who were Significant improvements in hand methods reluctant to give up rules of thumb and “back- for shielding calculations were made by K. of-the-envelope” calculations. The specter of Shure, A. Foderaro, F. Obenshain, J. J. Taylor best estimate, reality, and experience was “alive.” and many others. These included calculations, measurements, and convenient functional rep Early shielding programs at Bettis included resentation of gamma-ray buildup factors, ex- the SPAN-2 program6 for the IBM-704 com- tensive work on point kernel methods, develop- puter. SPAN-2 calculated uncollided gamma- ment of better flux calculation formulae,‘-4 and ray flux in laminar geometry using ray-tracing much more work on all aspects of shield design and integrating over the source by 3-D Gauss without computers. All this knowledge was quadrature. “SPAN”is an acronym for “Shield- summarized in the Reactor Shielding Design ing Problems Ad Nauseam.” When neutron Manual5 edited by T. Rockwell and published removal cross sections became available, it in 1956. This book is still a useful reference. A became possible for the same computation problem unique to the shield designs done at scheme to be used to calculate fast neutron dose Bettis was the necessity to optimize the shield rates in laminar geometry. Someone looked at a weight. Methods for doing this were also devel- box of a popular cleaning product and named oped during the period, and the earliest shield the new neutron dose rate calculation program design became the benchmark by which the “SPIC.” The shielding engineers did not need next designs were judged. clever program names; they wanted improved The foreword of the Reactor Shielding results, and computational improvements were Design Manual is short but full of shielder’s rapidly made.

55 SPAN-2 was followed by SPAN-37 which and easy-to-use with practice; this was consid- calculated gamma-ray dose rates in more com- ered to be “user-friendly.” plicated geometries on the Philco computer. Some of these programs were written to Many other programs wete developed for shield- calculate the energy release from fission prod- ing applications, including a method of finding ucts. Experimental data of others was accumu- the thermal neutron flux in primary shields by a lated12 and put into a library suitable for calcu- one-dimensional Pl multigroup approach, com- lation. In 197 1, much of this information was bined with a point-kernel calculation for correc- used in the proposed ANS standard “Decay tion of the spatial dependence of the Pl results Energy Release Rates Following Shutdown for at large distances from the source.8 This method Uranium-Fueled Thermal Reactors.” was replaced in 1963 by a P3 multigroup ap preach which describes the neutron attenua- Improvements in the determination of prob- tion at large distances from the source reason- able radiation damage to ferritic steel were ably well. made by showing that the exposure time neces- sary to achieve a given property damage did Bettis shielders were associated with much indeed depend on the damage cross section other significant work during this period. One model. l3 Changing the neutron spectrum through important activity was the measurement of the reactor vessel wall changes the apparent gamma-ray dose rates and fast neutron fluxes in attenuation coefficient of the exposure. Such a air from a source in water, which resulted in damage model provides a physically meaning- production cross sections for calculations with ful measure of exposure. An estimate of the Nr6 and N17 sources. These were published in correlation of transition temperature change 1958 and reviewed in 1962.10*11 with neutron exposure was fitted to ferritic steel data, and a technique for assigning one-sided Bettis shielders furnished their expertise to tolerance limits was described. l4 programmers and worked with them to obtain the many different computer programs neces- There are more than an equal number of sary for efficient and accurate shield design other published papers and reports referring to calculations. By listening to their users, most shielding research and computer program de- programmers were able to write programs that velopment. Some of the programs were explic- were based on correct theory and which were at itly written for shielding or were written for least “user-tolerable.” Sometimes a program other disciplines but adapted for application to was written that was both theoretically correct shielding.

References

1. P. A. Roys, K. Shure, and J. J. Taylor, 5. ReactorShieMDesignManual,T.RockweU “Penetration of 6-MeV Gamma Rays in III, Editor, TID-7004 (March 1956). Water,” Phys. Rev. 95(4), 911 (1954). 6. P. A. Gillis, T. J. Lawton, and K. W. Brand, 2. P. A. Roys, K. Shure, and J. J. ,Taylor, “SPAN-2 - An IBM-704 Code to Calcu- “Penetration of 6-MeV Gamma Rays in late Uncollided Flux Outside a Circular Lead and Iron,” WAPD-T-170, Talk at Cylinder,” WAPD-TM-176 (August 1959). American Physical Society in Washington,. D. C. in March 1954. 7. W. H. Guilinger, N. D. Cook, and P. A. Gillis, “SPAN-3 - A Shield Design Pro- 3. J. J. Taylor, “Application of Gamma Ray gram for the PHILCO-2000 Computer,” Build-Up Data to Shield Design,” WAPD- WAPD-TM-235 (February 1962). RM-217, January 25,1954. 8. D. C. Anderson and K. Shure, “Thermal 4. G. L. Strobel, “Additional Exponential Rep- Neutron Flux Distributions in Metal-Hy- resentations of Gamma-Ray Build-Up Fac- drogenous Shields,” Nucl. Sci. Eng. 8,260 tors,” Nucl. Sci. Eng. 11,450 (1961). (1960).

56 9. K. Shure, “P-3 Multigroup Calculations of 12. K. Shum, “FissionProduct Decay Energy,” Neutron Attenuation,” Nucl. Sci. Eng. 19, WAPD-BT-24, p. 1 (1961). 310 (1964). 13. K. Shure, “Neutron Spectral Effects on 10. P. A. Roys and K. Shure, “Production Cross Interpretation of Radiation Damage Expo- Section of N16 and N17 ” Nucl. Sci. Eng. 4, sure,” Trans. Am. Nucl. Sot. 6,388 (1963). 536 (1958). ’ 14. K. Shure, “Radiation Damage Exposure 11. K. Shure, “The 016(n,p)N16 Reaction Cross and Embrittlement of Reactor Pressure Section,” Nucl. Sci. Eng. 16, 247 (1963); Vessels,” WAPD-TM-47 1 (November also WAPD-BT-25, p. 27 (1962). 1964); also Nucl. Appl. 2,106 (1966).

57 UK Reactor Shielding: Then and Now

John Butler AEA Technology Winfrith, UK

Then at the balance let’s be mute, We never can adjust it; What’s done we partly may compute, But know not what’s resisted Robert Burns, 1759 - 1796

Introduction

“Water has no cracks” was the maxim of distribution of low-energy fluxes within the Theodore Rockwell III in his 1956 book on the shield ‘Ihese,intum,producesecondarygamma- shielding of small pressurized water reactors in radiation, which usually determines the overall the American naval programme. In contrast to thickness. The removal cross section was identi- the compact steel-water shields of submarine fied with the fast-neutron transport cross sec- reactors, the shielddesignof the early gas-cooled tion, and values were obtained from published power reactors was dominated by radiation calculations based on the hard sphere model of stmamingthroughcracksbetweengmphitebricks the nucleus. The removal source distribution andstructuralcomponents~dalongthecoolant was coupled to a set of age-diffusion equations ducts which penetrated the reactor vessel and solved in one-dimensional cylindrical or spheri- biological shield. The design of shields for the calgeometryandthemethodwasthereforeanal~ early UK research and prototype reactors was gous to the conventional first-flight correction to basedontheChalkRiveratomicenergyptojects, Fermi age theory. with empirical extrapolation from GLEEP to ‘Ihissimpliliedmodel,embodiedintheRASH BEPG and hence to Windscale and Calder Hall. code written for the Ferranti Mercury computer, Suchmethods were, however, unsuitable for use. was first tested at Calder Hall by drilling a re- by the industrial consortia in their tenders to the entrant hole to a depth of some eight feet into the Generating Boards for the first civil nuclear radial concrete shield; the results are plotted in power programme. A special shield design Fig. 1. It is apparent that when the thermal- manual, AERE R3216, was accordingly pro- neutron flux comes into equilibrium with the duced by the newly established HarwelI Shield- removal neutron beam, both the slope and mag- ing team in 1959, which introduced the concept nitude are well predicted but there are discrepan- of the energy-dependent removal cross section cies at the inner and outer edges of the shield. The for calculating neutron penetration in biological peak in the distribution in the first few inches of shields. concrete is due to the rapid thermalization of In the Removal-Diffusion model, a point epithennal neutrons emerging from the steel exponential kernel was integrated over the fis- pressure vessel and thermal shield by hydrogen sion source distribution in the reactor core to present in the concrete. Later studies in experi- predict the flux of high-energy (MeV) neutrons mental bulk shields confirmed the hypothesis -the so-called removal flux-which penetrate that the estimated hydrogen content of the con- thick biological shields by virtue of the low cross crete(whichcouldnotbemeasured)wastoolow. sections and pronounced forward scattering at In the outer regions, the measurements revealed high energies. Neutrons removed from the beam the degree of conservatism which had, of neces- by scattering, with large changes in energy and/ sity, been incorporated in the original design or direction, are rapidly moderated by hydrogen since the thermal flux dropped by about two present in the concrete, generating a spatial orders of magnitude to a minimum at a depth of

59 about 2 ft. from the outside surface of the shield operational plants were used to test calculational Both these discrepancies underlined the difficul- methods. ties encountered when measurements made in

The LID0 Shielding Reactor

With the widely differing layouts of the first “standpipes” of Magnox reactors. However the commercial Magnox reactor designs -bottom mock-up of large gas ducts was clearly imprac- refueling, cylindrical and spherical steel pres- tical at LID0 and, notwithstanding the earlier sure vessels, and later in the programme, con- difficulties encountered at Calder Hall, a com- crete pressure vessels with internal heat-ex- prehensive programme of neutron and gamma- changer shields - it was clear that a broader ray measurements was conducted during the base of experimental validation was required. A commissioning of the Chapel Cross reactors in programme of shield mock-up experiments was 1960. A total of about 500 activation detectors accordingly initiated in the LID0 swimming and dosimetry packs was irradiated to obtain pool reactor at Harwell, which is illustrated in detailed flux maps within and around the pres- Fig. 2. ‘Ihe small enriched uranium core was surevesselincludingthelargeinletandoutletgas suspended in a large water tank surmunded by a ducts. The threshold reaction S32(n,p)P32, which biological shield. In addition to the beam tubes measures fast-neutron attenuation, was exten- penetrating this concrete shield, which provided sively usedinthis programme and the sensitivity sources of neutrons for shielding experiments, was increased by burning the irradiated sulphur LID0 was equipped with experimental caves cut in an alhum cup before counting the beta in the biological shield which were separated particles emitted by the phosphorus residue with fmmthepoolbylargealuminumpanelssetinthe greatly increased efficiency. By buming large wall of the reactor tank. These “panel facilities” blocks of sulphur weighing 10 kilograms, it was permitted broad beams of neutrons (and gamma possible to measure a fast-neutron flux as low as rays) to penetrate tanks located within the caves 1 neutron/cm2sec. Quite apart from generating a which contained experimental slab shields en- wealth of experimental data for validating the abling the fluxes of fast and thermal neutrons to shield design methodology, this progmmme es- be measured using small activation detectors tablished the key role played by measurements over some ten decades of attenuation. A variety madeduringlow-powercommissioningofpower of shield combinations involving steel, lead, and reactors when access is possible for the retrieval various concrete textures was studiedin this way of passive detectors. Such measurements proved and provided further confirmation of the accu- to be the essential complement to attenuation racy of the RASH code for treating power reac- studies in experimental shield facilities such as tor shields. LIDO; similar campaigns were subsequently undertaken on all Authority prototype reactors The beam-tube facilities in LID0 were used and also in collaborative programmes on the to study a variety ofneutron streaming problems plant operated by the Generating Boards - a such as the stepped annular shields or “muffs” practice subsequently adopted in power reactor located around the steel charge tubes or commissioning throughout the world

The Shielding of Advanced Gas-Cooled Reactors

By the time of the second nuclear power for personnel access to the pressure vessel at programme, based on Advanced Gas-Cooled shutdown for inspection and maintenance of gas Reactors, in the mid 1960s. the removal-diffu- seals and insulation. In order to reduce the sion method had been developed in the shutdown activation of steelwork above the core, COMPRASH code which predicted not only the internal axial shields were designed in which the thermal flux and gamma-ray dose-rates but also channels contained shield plugs incorporating the fast flux incident on steel pressure vessels. helical or stepped annular paths to attenuate the For AGR requirements, there was a requirement neutron flux with a minimum of coolant pressure drop. Shield mock-up studies were conducted in lar heat exchanger where, again access for the LID0 panel facilities and the robust model of maintenance was required at shutdown. The the COMPRASH code was further modified by methodology was further refined by the intm- the inclusion of “shield-plug” removal cross duction of simplified albedos for the treatment of sections.Thecombinationofline-of-sight~am- wall scatter, and a suite of kernel-a&do stmam- ing in gas-filled voids with removal penetration ing codes such as MULTISORD and throughthewallsprovedtobeapowe~method MULTICYN were developed to complete the for treating the multi-legged outlet gas ducts calculational capabilities for the design of com- which reduced neutronpenetrationinto the annu- mercial AGR’s.

Early Monte Carlo Developments

Work on the development of a Monte Carlo ing reduction in particle weight whilst only half code named McNID (Monte Carlo for Neutrons ofthosepassingintoregionsoflowerimportance in Ducts) had been initiated about the time of the were tracked, these being given an increased Chapel Cross experiments when it was recog- weight. The first comparison of McNID predic- nized that a more rigorous treatment of stmam- tions with experiment is shown in Fig. 3. The ing problems was required to underpin the sim- fast-neutron flux measured by sulphur detectors plistic experimental validation of the mmoval- in an air-filled reflector channel of the low-power diffusionmethod. The choice of the Monte Carlo graphite-moderated reactor GLEEP was accu- method was determined by the need for a proper rately predicted, although the statistics were not treatment of these complicated geometries en- shown-probably because they were unaccept- countered in gas-cooled reactors. The early ably large! progress with McNID was disappointing: the Ferranti MERCURY computer at Harwell in Furtherincxeasesinefficiencywereachieved 1960 was too slow and the code was transferred for duct streaming calculations with the to an IBM 704 at AWRE Aldermaston. The RANSORD/RANCYN codes, in which particle analogue Monte Carlo was again too slow for tracking within the surromding shield material practical calculations, but significant accelera- was represented by a single reflection at the wall tion was achieved by the application of impor- surface, using an energy and angular-dependent tance sampling, in order to increase the propor- albedo to obtain the energy and direction of the tionofthetimespentbythecomputerintracking emergent particle. The McNID code continued the “important” particles, i.e. those which were to be used-albeit at considerable expense-to going to contribute to the flux at the detector. The generate “theoretical experiments” which pro- greatest success was achieved by the application vided, inter&a, shield plug removal cross sec- of splitting and Russian Roulette, in which the tions, albedo data and build-up factors for gamma number of particles was doubled on passing into rays which were difficult if not impossible to a region of higher importance with a correspond- measure with the available detector technology.

Fast Reactors

In parallel with the AGR programme, work keV so the removal component of the spectrum had been in progress for several years on the is relatively unimportant the migration of these design of the Prototype Fast Reactor (PFR). intermediate energy neutrons was accurately Little information on shielding relevant to power reproduced by adjusting the number of the diffu- reactors was available from the Dounreay Fast sion groups to match the relaxation length of the Reactor because of the complexity of the top flux in typical fast reactor materials, principally rotating shield which allowed access for testing sodium, steel and graphite. For this purpose, a fuel subassemblies specifically for PFR. The large mock-up of a sodium tank with an outlet COMPRASH methodology again proved to be duct was installed in the LID0 panel facilities, surprisingly versatile: the breeder leakage spec- whichisshowninFig.4.Intheseexperimentsthe trum of a fast reactor peaks in the region of 300 breeder leakage was spectrum simulated by in-

61 m-posing a thick steel filter between the panel the PFR during commissioning, exploiting tech- andthe sodium tank. Whilst the thermal flux and, niques developed orlginally at Chapel Cross. surprisingly, the fast-neutron flux were accu- These measurements served to validate the rately reproduced along the axis of the main COMPRASH method for a variety of shielding sodium tank, major discrepancies were encoun- problems in various designs, which were used tered within the mock-up of the duct as shown in fortheplannedCommercia.lFastReactor(CFR). Fig. 5. The attenuation was, of course, domi- A key feature of all these layouts- as in all fast natedbylateralleakage andinordertoreproduce reactor designs - was the performance of the the measured fluxes it was necessary to use nucleonic instrumentation located in thimbles in buckling terms in the one-dimensional the front row of shield rods close to the outer COMPRASH code which were derived from boundary of the breeder. Neither LID0 experi- measured lateral flux scans. The problem was ments nor the PFR commissioning experiments solved by writing a two-dimensional diffusion could establish the accuracy of count-rate and code, SNAP, which was again coupled to the gamma-ray background predictions for flux- removal source of COMPRASH for fast-neu- measuring instruments located close to a breeder tron flux predictions. This example illustrates an in which complex distributions of plutonium important maxim for shield designers, namely build up during the life of the reactor loading. A that it is more important to model the geometry series of experiments was accordingly initiated of the shield correctly than it is to treat the on a mock-up of the inner shield region in the slowing-down process accurately. Thus a two- zero-energy critical facility ZEBRA. The spec- dlmensional diffusion calculation, notwithstand- trum at this breeder/shield interface was mea- ingthelimitations ofdiffusiontheory, may prove sured with a series of hydrogen-filled propor- to be more accurate than a one-dimensional tional counters which had been developed for transport method in which some empirical cor- measuring the spectrum within the reactor cone. rection must be made for lateral leakage. This was probably the first measurement of the isotropic flux spectrum within a reactor shield The steel filter was subsequently used in mock-up. Spectrum measurements had not been studies of a variety of shield mock-up experi- possible with LID0 because of the high gamma- ments for the PFR, including arrays of graphite ray background due principally to capture radia- rods enclosed in steel which were to be used for tion produced in the aluminum structure of the the internal radial shield in order to reduce core. Attenuation measurements of the fast, activation of the intermediate heat exchangers. epithermaIandthermalfluxesintheshieldmock- ups were again accomplished with activation ‘Ihe findings of the LID0 experiments were detectors. The whole system was analyzed with confinned when an extensive programme of flux the removal-diffusion method using the two- measurements wascarriedoutwithinthetankof dimensional code SNAP supported by McNID.

The Two-Tier Scheme of Shielding Codes

The next key development in calculation Russian Roulette, stems from the fact that alarge methods was published in 1967 when saving in execution time can be achieved by an COMPRASH was used in adjoint mode to gen- approximate importance function which may be erate approximate importance maps for the ac- in error by as much as a factor of two. Neverthe- celeration of the McNid code. The first test of less, it was recognized that an improvement in this method was the prediction of the thermal the accuracy of the diffusion method was desir- flux at the outside edge of a shield configuration able, not least because the forward solutions similar to that of Calder Hall irradiated by a were still required for design/survey calcula- plane source of fission neutrons. The results, tions. which areillustratedin Fig. 6, reveal the progres- sive improvement in statistical accuracy of the The removal diffusion model was accord- flux as the plane of the adjoint source is ap- ingly replaced by the ADC method in which the proached. The success of this method, which diffusion coefficients in a 28-group scheme were greatly facilitated the application of splitting and systematically adjusted against reference spec-

62 tra obtained from the published results of the points within the shield. McBEND was used, in method of moments. In due course, these refer- turn, to refine the constants adjusted in the ADC ence data were superseded by calculations ac- scheme and also to generate albedos, build-up complished with the McBEND code itself and factors and removal cross sections for use in the the two-tier scheme of design codes was estab- rapid survey calculations. This approach is still lished (Fig. 7). In this scheme the semi-empirical available in the current version of the code suite, methods, ADC diffusion theory and point-kernel McBEND8, although Monte Carlo is now used integration were used for survey calculations for most ifnot all shielding calculations, exploit- and also to generate importance maps for ing the inherent parallelism of the method on McBEND acceleration. TheMcBENDestimates parallel-architecture machines. served to peg the survey calculations at spot

The Impact of Parallel Architecture

Inordertoillustratetheimpactofcomputing Recognizing the inherent parallelism in hardware developments, the effective perfor- Monte Carlo, all McBEND calculations at mance for Monte Carlo calculations of the ma- Winfrithweretransferredfromthemainframeto chines available in the UKAEA (AEA Technol- Micm VAX-2 workstations in 1983, soon to be ogy) is plotted in Fig. 8 against the first year of followed by the introduction of SUN worksta- use in fuzzy time. The slope of the curve indi- tions. The considerable penalty in elapsed time cates an increase of a factor of 34 per decade. for a calculation was mitigated by the use of This trend has been maintained by architectural several workstations in parallel running over a innovationssuchasinstructionbuffers,pipelining week-end. The price/perfomrance of these mi- and some limited internal parallelism, which cro-processor workstations improved dramati- have apparently compensated exactly for the cally, perhaps doubling every 1 to 1 112 years. lower rate of technological improvement in the Absolute performance levels also increased very most powerful single processors. This leveling quickly, almoughthismay now be slowing down. off in performance of single processors suggests The first parallel-architecture machine with four a lower extrapolation curve for systems at con- Meiko T800 transputers was purchased in 1988 stantpricesuchasthatshowninthefigure. There giving some 10 MFLOPS performance with a are, however, new possibilities arising from higher MIMD system. After several enhancements, this degrees of internal parallelism offering rapid installation was superseded by the present ma- improvements in price/performance ratio with- chine, which employs six i860 processors in out any significant increase in the real cost of a parallel. The maximum performance of 360 topendsystem.Itremainstobeseenhowquickly MFLOPS claimed for this installation has not the transition will be made to the upper extrapo- yet been achieved, but within a decade, Main- lation curve postulated in Fig. 8, and the extent frame (CRAY) performance has been secured on to which the overall performance can be im- desk top with a saving in capital cost of the order proved without excessive investment of effort in of a factor 100. applications programming and system support

Gamma-Ray Shielding

During the 1970’s the need to improve the infinite parallel beam sources using the method accuracies of gamma-ray shielding calculations of moments. The solution of these problems in was recognized in all the reactor projects, re- complicated geometries was greatly facilitated flectingmorestringentlimitsimposedonperson- by the introduction of the RANKERN code (Fig. nel dose-rates. The traditional kernel methods 7), which utilized Monte Carlo methods to per- for calculating gamma-ray attenuation had re- form the integration of the conventional point mained virtually unchanged since the days of kernel over complex spatial distributions of point Rockwell and the early compilation by Goldstein sources. In addition to build-up corrections, in the USA of build-up factors for point and provisions were made to handle wall- and air-

63 scatter by means of a multiple scattering model shield slabs in addition to other more compli- in which the effects of secondary and higher- cated shield mock-ups. The study of cobalt acti- order scattering events could be accommodated, vation problems in AGR’s and all other reactor provided that the locations of these events were types was more conveniently undertaken in the known. The validation of RANKERN therefore second facility named ARCAS (Attenuation of called for experimental studies of bulk shield Radiation from Cobalt Activation Sources). In penetration with scattering at shield surfaces. this facility, provision was made to locate auto- Two facilities were accordingly constructed at matically a series of cobalt sources at pre-deter- NESTOR. The first, which was called PYLOS mined positions within a complex shield array (Photon Yielding Loop Source) after the island using a system of teleflex conttols and flexible where the legendary cave of NESTOR was tubes. The experimental shield was again housed ~~atedcomprisedawater-filledcircuitinwhich within a concrete cave which created the low activated nitrogen- 16 was pumped from the te- background conditions required to measure an actor to a concrete cave located in an area of attenuation of the order of 107. A variety of relatively low gamma-ray background remote experimental situations was analyzed using the from the reactor. The active liquid could be basically simple kemel/albedo/build-up model circulated through various combinations of cy- embodied in RANKERN, which is now in its lindrical and pipe-source configurations. Provi- thirteenth release and has become the most pow- sion was also made to utilize the standard ASPIS erful gamma-ray shielding code available.

The Evolution of Benchmark Experiments

‘Ihe word “benchmark” came into the U.K. required for the shielding of the cavity region and shielding vocabulary around 1970 when it was a half-size mock-up was mounted above the recognized that the accuracy of Monte Carlo experimental cave as shown in Fig. 9. A major calculations was being increasingly limited by programme of experiments was undertaken with erms in the basic cross section data as more this apparatus under the terms of the UKAEA/ stringenttarget accuracies fordesignparameters USNRC Collaborative Agreement to refine the weresought.LIDOwascloseddownin 1972and techniques for monitoring RPV damage. The a simple-geometry fission-source place was de- NESTOR Shielding and Dosimetry Improve- signed for benchmark experiments and built onto ment Pmgramme (NESDIP) complemented the the NESTOR reactor at Winfrith, which could U.S. Do&retry Improvement Program (NRC/ furnish very low gamma-ray backgrounds for DIP) and the REPLICA experiment was set up to spectrometer measurements. The new shielding measure spectra in the configuration of the PCA facility, which was named ASPIS (Activation pressumvesselsimulationexperimentsconducted and Spectroscopy In Shields), the word used by at Oak Ridge, benefiting from the clean-geom- Homer for the shield of NESTOR, is illustrated etry fission-source plate and the low gamma-ray in Fig. 9. Experimental shields with a total background in the ASPIS facility. A comparison thickness of up to 12 ft. could be accommodated of McBEND predictions and measurement of in the mobile tank which could be withdrawn the spectra made with hydrogen-filled propor- from the cave for the retrieval of passive detec- tional counters and NE-213 scintillators in the tors and the loading of shield components. Ac- position of the Void Box (simulated cavity) in the cess for spectrometers was provided by a series PCA configuration is shown in Fig. 10. of slots in the cave roof which, together with ports through the side of the shield trolley, en- The ASPIS experiments were all designedin abled a variety of complex shield arrays to be such a way that it was no longer necessary to assembled for the study of ducts with bends and make significant approximations in the repre- complex sodium-steel arrays for the commercial sentation of the geometry using the McBEND fast reactor designs, including a simulated oxide code. In these so-called benchmarks it was pos- breeder. sible, therefore, to attribute discrepancies be- With the advent of the Sizewell B PWR tween calculations and measurements to short- design, validated methods of calculations were comings in the basic nuclear cross-section data.

64 Nuclear Data

With the increasing use of Monte Carlo in DUCKPOND (Derivation of Unknown Con- practicalshieldingproblems andthecompilation stants from Known Perturbations of Nuclear of benchmark experiments in reactor materials, Data) was accordingly written for use with it became clear that the accuracy of the basic McBEND. Attempts were made to adopt the nuclear cross section data in the UKNDL was cross-section adjustment technique developed inadequate to achieve the target accuracies for for fast reactor core physics calculations, a shielding calculations which ranged from about generalized least-squares technique being used +lO% for reflector leakage fluxes to about a to adjust cross-section values within their as- factor of two for the dose-rate at the outside of a cribed uncertainties over energy ranges identi- biological shield. Moreover it was clear that the fied by the DUCKPOND sensitivity calcula- accuracies demanded could not be met by differ- tions. However, the range of extrapolation to entialmeasummentswithmonoenergeticsoumes practical shield designs was necessarily much produced by particle accelerators since the at- greater than for core performance and it proved tenuation in a shield is exponential in character to be impractical to utilize adjusted cross-section and an accuracy of approximately 1% would be sets for a wide variety of different shieldconfigu- nxpired in the relevant cross-section values to rations. Nevertheless the adjustments made to achieve an accuracy of a factor of two at 20 mean the iron cross sections, reproduced in Fig. 11, free paths. Similar problems were encountered proved to be in remarkably close agreement with in other countries and so the Nuclear Energy the values in the Joint European Data File (JEF) Agency (NEA) through their committee on reac- produced nearly a decade later. In general, the tor physics initiated a collaborative pmgramme sense andmagnitude of the adjustments provided of integral benchmark experiments. These were important pointers for the nuclear data evalua- designed on the ASPIS pattern, in which the tors. When new evaluations were produced, they neutron spectrum was measured as a function of could be checked by re-running the McBEND penetration distance through a well-defined slab calculations for the data-testing benchmark in shield array irradiated by fission neutrons from ASPIS and, if necessary, the iteration could be a well characterized uranium source plate. continued. European shielding calculations are Sensitivity calculations were initially ac- now based on the Joint European Data File, JEF- complished with the Oak Ridge perturbation 2, and data-testing benchmarks will continue to code SWANLAKE, but the restriction to one- be conducted in pursuit of the higher accuracies dimensional geometry introduced unacceptable now sought for some key parameters, such as errors in treating the smalI fission plate of the heating andiron displacement-rates in steel pms- ASPIS facility. A novel Monte Carlo routine for sure vessels, which are being reviewed for life first-order perturbation calculations named extension both in Europe and in the USA.

Concluding Remarks

This review has traced the evolution of U.K. that shielding calculations are once again being shielding methodology over a period of some carried out for the Magnox reactors, this time to thirty years. The early developments were driven refine estimates of pressure vessel fluences in by the need to solve the special streaming prob- support of embrittlement studies for plant life lemsofgraphite-moderatedgas-cooled(Magnox) extension. reactors. Now the wheel has turned full circle in

65 - Solid Line - Calculated flux normalized to 100 MW reactor power (fission energy release)

_ Circles - Experimental flux value8 (manganese foils) 10’

-I

Id’ I- t

Figure 1. A comparison of the thermal neutron flux predicted by the RASH code with measurements in concrete shield of a Calder Hall reactor.

66 67 Rqxesen&ion of the lndividud Fuel Cell for MC NID Cdc

-i

n : 50 Distance from End of Fuel element (cm)

Figure 3. Comparison of sulpher reaction-rates predicted by McNID with measurements made in a GLEEP reflector channel.

II / /i -, Rwctor Ahuniaiuur Panel I/ Iron Filter Backing

Large Sodium Tanks

Figure 4. Arrangement of sodium experiment LID0 Panel C facility.

68 10-2’

I o-22

I O-22 E 13uckling Terms Dcrivcd I’rwn Measured L;ueral V;\ri;\Iions

I o-24

Buckling Ternls Calculuted A’ssuming / Black Boundary Condilions for Ilie Duct Wall I()-” I I E

I o-2” _ - Theoretical Values

l Experimental Points

I O-2’ IOU I so

Distancerrtm Duct Mouth (cm)

Figure 5. Calculations of the cadmium-covered indium reaction-rate in the sodium-filled duct made in one dimension with different buckling terms.

69 Graphite Concrete 0

Relative Importance of Neutrons

Lemovul - Diffusion

(Thermal Flux)

-5

-10

Relative Importance of Neutron between -20 II I 2 per & 50 per

II I

0 I 00 200 300 Distance from Source Plane (cm)

Figure 6. Removal diffusion and Monte Carlo calculations of thermal flux at 348 cm from 6 MeV source.

70 UNITt&~GWAM EVALUATED SASIC NUXEAR DATAUBIURIES e LlwtARV ““T% DATA OJKNw 1 (ENDFiB)

I I I

Figure 7. Two-tier scheme of linked programs for radiation transport calculations.

IOT - Performance (FLOPS) IT-

IOOG-

IOG- CRAY Y P(8) IG- CRAY 2(4)J; , _ / IOOM / / -/ T800m ,u IBM RS 6OOOI‘550 IOM I ’ _/;dlIEC slatioll 5oW IM rnrr 11,

Ik

,9;0 A30

Year of firs1 use

Figure 8. UK Monte Carlo shielding calculations - past computing performance and possible future trends.

71 .‘,,. ,. /.. . \:: i :‘. . . \, \ $., .) ,\;, .;. .; ;:,.I’. . ;:: ,_, .,:.‘.’ -

\

( : i :

72 --- McBEND (8000 Group) - Meas-mt

Figure 10. Comparison of spectrum measurements and calculations in the void box of the PCA-REPLICA.

I r-7 I ‘i- -JEF --- DFN908 0 DFN 908 Adjusted

iR 10’

Energy (MeV)

Figure 11. Comparison of the iron DFN 908 adjusted and unadjusted nonelastic cross section with JEF values.

73 A Very Personal View of the Development of Radiation Shielding Theory

Herbett Goldstein Columbia UnivetSity New York, New York

The title tells it all. The personal bias is clearly admitted. What I have done, what I have urged others to do, and what I have seen done all get emphasis, most likely undue emphasis. Let future historians be objective!

Introduction

From the very start of the exploitation of the After ANP the shielding horizon was over- nuclear energy released in neutron fission, re- shadowed for quite some time with the needs of search in radiation shielding has been almost the LMFBR (Liquid Metal Fast Breeder Reac- entirely project-driven At any giventime, sbield- tor). Now that that project is no longer alive, at ing research has mostly been directed to satisfy least in the U .S . , the driving force in shielding- the pressing demands of whatever was the cur- type research seems to be the problem of radia- tent project of top priority. In the Manhattan tion induced embrittlement in reactor pressure project days, shielding research was mainly vessels. And all through these decades, until aimed at solving the problems connected with very recently, in addition to the variations in the the production reactors andthe associatedchemi- reactor shielding pattern, there has been the cal separation plants. In the immediate post-war continuous bass accompaniment of shielding period (and for some time after) the needs of the against weapons gamma radiation, whose te- naval reactors dominated the shielding scene. quirements have led to powerful and complex When that seemed to be well in hand, the calculational methods. program for ANP (Aircraft Nuclear Propulsion) But along with all this urgent press of came along to pose the high-priority shielding project-inspired shielding research there was questions. always a small trickle of what may be called (As an incidental side-note, the ANP pro- fundamental theoretical research. By which I gram provided the shielding designer with what mean research not aimed at answering this or were some of the most intellectually demanding that design question, but rather having the goal problems ever encountered, along with possible simply of studying how radiation of interest exotic solutions that were never again allowed penetrates through substantial thicknesses of within our range of vision. The fixed geometry matter. And the “radiation of interest’*primarily of reactor, crew and engines called for clever means gamma rays or neutrons generated in and skillful optimized shaping of components fission reactors and in their environment, par- of the shielding complex, with weight as the ticles whose capabilities for deep penetration principal constraint. On the other hand, cost was give rise to the most serious problems in radia- not considered to be greatly important, so we tion shielding. Beta rays and other charged could ponder the properties of neutron shields particles are also of interest, but most often they made of lithium hydride, and even toy with the appear as secondary particles in the transmis- idea of gamma shielding composed of separated sion or detection of the more penetrating radia- tungsten isotopes for critical parts of the shadow tion. This paper, therefore, is limited to the shield. Ah well, we came quickly down to earth history of fundamental theoretical research in again literally and figuratively with the demise the penetration of gamma rays and neutrons, in of ANP.) the context of reactor shielding.

75 Gamma Ray Penetration

For many masons, fundamental studies of energy, and large-angle scatterings are quite the type under discussion were historically first likely. So the straight-ahead approximation attempted for gamma rays. In the energy region failed disastrously, generally by predicting much of interest -up to about 10 MeV-the basic too large a transmitted dose. ’ interactionsofgammarayswithatoms hadbeen Another attempt was the direct calculation prettywellestablishedinthefirsteffluorescence of the flux of successive scatterings, the scat- of studies in the quantum theory of radiation in tered photons from a given order of collision the 1930s. While highly accurate numerical constituting the source for the next collision. In values for the interaction cross sections weren’t conception, the method is quite rigorous, and determined until a good deal later, even by the can work well when there are only one or two late 1940s they seemed to be well enough scatterings before absorption removes the par- known that detailed studies of deep gamma-ray ticle completely. But in a medium that is mainly transmission had a chance of being profitable. Compton scattering, many scatterings can oc- After all, over most of the pertinent energy curbeforeabsorptiontakesplace. Sothismethod region there were only three major mechanisms also failed disastrously in just those situations for gamma-ray interaction with matter - the where the penetrating flux is most affected by atomic photoelectric effect, Compton scatter- the buildup of secondary particles.’ ing, and pair production. Of these, only the Compton scattering provided major complica- A practical method for calculating gamma- tion in affecting the penetration because of the ray penetration, one which did not make unac- buildup of secondary scattered photons. Here, ceptable compromises with the physics of the at least, the kinematics of the scattering, relating interactions and was still amenable to numeri- the angle of scattering to the scattered energy, cal computations, first appeared in the method was absolutely certain and clear. Further, all of of moments devised by U. Fano and L. V. these interactions varied smoothly and rela- Spencer. The technique was spawned at the tively slowly both with energy and with atomic National Bureau of Standards around 1949, but charge number, 2.” There thus seemed to be the first publication (beyond internal memos) did tantalizing possibility of systematic studies of not take place until 1951.3 The method trans- value throughout the energy range and over the forms the integro-differential transport equa- entire periodic table based on only a quite tion into a set of linked equations for the mo- limited set of calculations. ments of the radiation flux relative to spatial By 1950, therefore, a number of attempts variables. Each member of the linked set is an were being made to calculate gamma-ray pen- integral equation in a single variable related to etration from first principles (as distinguished the particle energy. The linkages enable the from empirical formulae). Most of these, how- solutions to be carried out in a specific se- ever, involved some sort of approximation to a quence, depending on source and geometry rigorous calculation. For example, the straight- symmetry. In principle, there is no restriction ahead approach correctly included the slowing on the complexity of the scattering kernels, but down of the photon following Compton scatter- there is an intrinsic limitation to infinite media ing, but the angular deviation of the sqattered homogeneous in the type of scatterers. photon from its original direction was ignored. In addition, the solution is obtained in the This approximation had worked well in studies form of a finite set of discrete moments of the of cosmic-ray showers, dating from before World particle flux. The continuous behavior of the War II, in which the gamma rays are of such flux must be reconstituted from the discrete highenergy that most scattering angles are quite moments. Mathematically, such reconstruction small. However, in the context of reactor radia- has no unambiguous solution, but with some tion, the gamma radiation is much lower in physical insight - and a good deal of luck - useful reconstructions can be obtained. Despite ‘There is the obvious exception in the photoelectric these constraints, the method was so much of an effect of the discontinuous behavior in both energy and improvement over anything else then available, Z at the shell edges. But from the start this seemed easy to handle, mainly by breaking up the energy treatment that in 1950 the present author, acting on a into separate intervals at these discontinuities. suggestion of L. V. Spencer, proposed to the

76 AEC that a systematic program of moments concerned in carrying the project to completion method penetration calculations be undertaken were clearly justified. While we were of course for gamma rays. In the project, as subsequently gratified by the reception accorded the compu- organized under AEC support, the choice of tations, we did not foresee many of the subse- problems and subsequent analysis was under- quent developments in their use. For example, taken at NDA (Nuclear Development Associ- it appears that users have paid little attention to ates), with the actual implementation of the the predicted flux spectra. Most of the applica- moments method carried out by L. V. Spencer tions have simply used the tables of buildup on the then new NBS computer, the SEAC factors, which we had thought would be applied (Standards Automatic Eastern Computer). mostly to shields consisting of one medium. It was then the very dawn of electronic But many shields am inhomogeneous, frequently computation, and the full saga of the application with successive layers of materials having of computers to the project should really be told greatly different gamma shielding properties, by Spencer and his colleagues. only some e.g. lead-water combinations. For these, inge- highlights can be mentioned here. The SEAC nious empirical formulae have been derived (some of whose bones now rest in the bowels of leading to buildup factors applicable to the the Smithsonian) occupied several good sized entire shield. The development of these consid- rooms, [the CPU alone require 18 relay racks!] erations, along with that of analytic representa- and represented computing powerless than that tions of the behavior of buildup factors with possessed by many present day lap-top comput- shield thickness, have been detailed by D. K. ers. It involved computer technology now known Trubey in another paper presented at this ses- only to historians -magnetic drum memories sion.’ Because we expected more effort might coupled with faster mercury delay line memo- usefully be spent in reconstituting the spectra ries along with temperamental Williams tubes. from the computed moments, we distributed The clock time was 1 MHz, and the total sum of microfilm copies of the original moments out- the “fast” memory amounted to 45 kilobits. No puts to the then AEC Deposit Libraries, along high level languages were available, of course, with a hard-copy report to assist in reading and programming was done at the lowest ma- them.6 However, we don’+ believe anyone has chine instruction levels. Printers were teletype made subsequent use of these computer out- machines, and the output came out on paper of puts. Above all, we did not expect the longevity such size and finish as to remind one of the of the usefulness that the tabulated buildup coarser grades of paper used for more intimate factors have shown. They continued to be filnctions. quoted, compiled and applied without a full- scale replacement for almost 37 years until All told, some 280 combinations of materi- superseded by up-to-date calculations described als (8), source energies (9), and source geom- by Trubey in his paper referenced above. In- etries (up to 6) were studied. The computed stead, what we had anticipated was that there moments were reconstituted to predict scalar would be a gradually declining usefulness of the flux spectra at various distances from the sources. buildup calculations in the measure as com- From these primary results, integrated quanti- puter techniques improved for the full-scale ties, then described as buildup factors for dose combined neutron-gamma ray calculations of and energy, were calculated. These bald sen- reactor shields, so that gamma-ray penetrations tences conceal a process that was by no means would be calculated in situ, as it were. I believe straightforward, and often involved patchwork that this has indeed taken place. What was not interpolation and extrapolation based on physi- appreciated was the continued application of cal grounds not always of the utmost surety. A buildup factors to the more frequently occurring final report, authored by H. Goldstein and J. E. problem of shielding of isolated gamma-ray Wilkins, Jr., was issued from NDA in 1954 sources. On the other hand, the explosive devel- describing the calculational project in some opment of computer technology also was not detail, and presenting all of the flux spectra foreseen. The moments method treatment of reconstructed from the moments, along with gamma-ray transport survives today perhaps tables of the buildup factors4 mainlyintheformofahomeworkproblem tobe The results of the calculations rapidly performed by the student on his own personal achieved widespread use, and the efforts of all computer!

77 Initial Attempts at .Calculating Neutron Penetration

The success of the systematic calculations The most significant calculations were of of gamma-ray penetration inevitably led to sug- the penetration of neutrons from monoenergetic gestions of a similar project for neutron penetra- sources in water, lo which were undertaken to tion. But the factors that made such a program see what part of the fission spectrum wasimpor- feasible for gamma rays were all absent for tam in determining the penetration of fission neutrons, especially in the mid 1950s. Neutron neutrons in water.’ Figure 1 shows the most cross section data were only sparsely available, important conclusion of the calculations. especially in the MeV range that was known to be important Further, it was already clear that 1 neutron interactions not only often varied rap idly with energy, but were usually quite differ- ent from one nuclide to the next in the periodic table. Nonetheless, an attempt was made to carry out moments method calculations for a number 0.75 of media thought to be of greatest interest. The moments method was recoded for the neutron case, first for the UNIVAC7 and then again for the IBM 704 as one of the first major Fortran codes, entitled RENUPAK.8*bWith these codes 0.50 neutron penetration was studied in a substantial number of materials ranging from pure hydra- gen and water to several types of concretes.’ For many of these substances the chief quantity of interest in the calctiation was the second mo- I ment itself, because at that time there was 0.25 -R = 30 considerable interest in possible discrepancies betweenexperimental and theoretical values for the age of fission neutrons at epithennal ener- gies. But in regards to the deep penetration questions of importance for shielding, the cal- o- culated results had little contact with reality, u 2 4’ 4 8 10 12 14 except for hydrogeneous materials. This excep E, Mev tion, however, is an important one, for in sub- stances containing substantial amounts of hy- Figure 1. Relative contribution of source drogen, for which water is the prime example, energies to the fast neutron dose from a neutron penetrationis mostly determined by the fission source at various distances (in centi- neutron interactions with the hydrogen nuclei. meters) in water (from NDAlSC-60). Further, these interactions with hydrogen are relatively simple and were well enough mea- sured, even in the 195Os,to give some hope for Note that at a distance of only 30 cm fmm meaningful calculations. More useful, perhaps, the source the most significant source energy is than the actual numbers obtained was the under- already more than4 MeV, and beyond 60 cm the standing that the calculations led to the mecha- important source neutrons are greater than 6 nisms that govern the penetration of neutrons in MeV, involving only 2-3% of the fission neu- these media. trons. Of course, on a little thought this state of affairs becomes reasonable. A collision with

CNow-a-days perhaps we would simply do an adjoint %oth the machine and the code were so unstable that it calculation, but even so there is more information to be was desirable to have the programmer in attendance each derived from reconstituting the continuous source spec- time the program was run! trum from a set of monoenergetic sources.

78 hydrogen rapidly slows down the neutron, send- standing of physical penetration mechanisms ing it to an energy region where the hydrogen can come before the ability to go to the utmost cross section is even larger. At higher energies, decimal place. And that’s a great incentive to the hydrogen cross section decreases rapidly the fundamental research kind of theory. with energy, so scattering by the non-hydm- geneous component (which has relatively less It became clear to all at about this time effect on the neutron) becomes more prevalent. (somewhat before 1960) that further achieve- As a result, the penetrating component comes ment in research on neutron penetration de- from the relatively rare high-energy source neu- pended on the availability of better neutron data tron. Note that we can arrive at this picture of the and on the development of more capable trans- mechanism of the penetrating component even port calculational methods. Considerable if the cross sections are not known well enough progress on both scores took place in the next to enable the calculations to be done to high two decades, largely by our riding on the coat accuracy. So long as the cross-section behavior tails of the much better funded reactor design is qualitatively in the right ball park, under- community.

Advances Relative to Neutron Data

This is not the place to give a detailed now designated “low energy nuclear physics.‘d history of the efforts to provide the nuclear data Evaluated data also had to be compiled, most needed to develop the applications of fission preferably in a machine-readable form. It was a (and fusion) energy. In this country, the AEC considerable advance when a nearly universal and successor agencies mounted substantial format, known as ENDF/Bt was developed in programs to produce the nuclear data for the the U.S. for such compilations of evaluated design of reactors (and weapons), and the needs microscopic data. The final step in the manipu- of shielding managed to get heard in the process lation of the cross section data was the transfor- - occasionally. Similar efforts were under- mation of evaluated data into quantities directly taken in other countries, to varying degrees. inputtedintotransportcodes. Theseareusually, Cross section centers were set up in many although not invariably, multigroup quantities. countries (in the U.S. the role was taken by the Clearly this last step is particularly tied to the National Nuclear Data Center), and there was a transport calculational method to be employed, proliferation of national, regional and intema- unlike the previous procedures. tional data committees to stimulate and keep a These five steps (including the two compi- watch on the programs. Crucial to the success of lation productions) each have accumulated an all these activities was the realization that get- attendant host of computer programs, not to ting the data into the hands of the calculators is mention the organizations necessary for publi- a multistep process. First, of course there had to cation and dissemination of the final results. be experimental programs to undertake the Shielding has benefited greatly from all this needed measurements of microscopic nuclear activity. The present status of neutron data for data. Then the large volume of experimental transport calculations can be summarized as data had to be made available, usually through reasonably satisfactory. That is to say,if enough data compilations that were at first printed and trouble is taken, neutron data rarely forms a later made machine readable. There then had to bottleneck in the computation of flux and spec- be inserted a step that came to be known as “data tra needed even for shielding design numbers. evaluation” -typically resolving inconsisten- cies in the experimental data, and filling in the gaps in the measured data by means that ranged dBy 1960 low energy nuclear physics was considered “a from sophisticated nuclear theory to inspired dead subject” by “pure” physicists and relegated to the guessing. It took some time to convince the rear areas of physics research. Whatever vitality it has powers-that-be of the necessity of the evalua- had since then has been the consequence of the applied tion step, but by the mid- 1960s evaluation was data measurement and evaluation programs. recognized generally as important and calling ‘ENDF = Evaluated Neutron Data Format. Its origins for the highest level in understanding what is date back to about 1962.

79 The catch is in the taking of “enough trouble” to 6-12 MeV gap, where present neutron sources manipulate the nuclear data in adequate detail for cross section measurements are still not and computational accuracy, which has rarely if adequate. There seems to be little likelihood of ever been done in design calculations. How- substantial improvement in the situation in the ever, forfundamentalresearchinunderstanding foreseeable future. Still, illuminating sensitiv- how nuclear interactions affect the penetration ity and “what if’ calculations can now be con- of neutrons, the available data seem quite ad- ducted to see the effect of various possible cross equate. The only possible exceptions are in the section behaviors, if them is the desire.

Advances in Neutron Calculational Methods

By the late 195Os, the limitations of the available. After all,the applicationofthe Wiener- moments method, particularly the restriction to Hopf method to solve the transport equation in infinite geometries of only one physical me- certain simplified situations has a history that dium, had become intolerable, and successor goes back well before World War II.12 And in methods had to be sought. A seemingly irmu- 1960 K. Case published his most ingenious merable flock of neutron transport methods development of the method of singular eigen- have been proposed and developed to varying function expansions, initiating the field of what degrees. Distinction should be made between has been described as “Caseology.“13 But while empirical methods, whose main attraction has extremely sophisticated, these approaches am been their simplicity, and methods rigorously restricted mainly to infinite or half-infinite based on first principles. For present purposes media, monoenergetic transport, and simple not much attention need be paid to the empirical scattering laws. Attempts at extending the tech- methods. It will suffice to mention the once niques to more realistic situations have not ubiquitously popular removal cross section ap- progressed very far. Other than providing some proach. In its original formulation dating to rigorous analytic solutions against which more 1950n it was applied only to hydrogeneous numerical methods can be tested, I don’t know media, or layers of heavy material followed by of any applications to neutron shielding prob- hydrogeneous media. The idea was that the lems. main burden of determining the penetration was Mention should be made of one other ana- carried by the hydrogen content, and the heavier lytic attempt at solving a related problem. In the nuclei affected that penetration only by an ef- mid-1970s, Cacuci, building on some much fective absorption, or “removal”, characteristic older work of G. C. Wick and G. Placzek, of the neutron cross interactions at the dominant worked at an analytic solution for neutron slow- energy of penetration. These effective removal ing down and transport in a medium of constant cross sections of the heavier material could be cross section. Despite the virtuoso application estimated apriuri, but mostly had to be deter- of erudite mathematics, only the spatial mo- mined from bulk experiments. Later elabora- ments could be obtained.14 Neutron transport tions saw the introduction of energy-dependent results of use for shielding require, it must be removal cross sections and elaborate multigroup concluded, numerical calculations on a com- formulations for the removal flux, coupling to puter. Methods fitting this description are of multigroup diffusion calculations. During the two kinds - either stochastic simulation of decadeofthe60s,andwellintothe70s, removal particle transport, or deterministic solutions of approaches dominated the calculation of neu- the linear Boltzmann equation. tron attenuation for design purposes. In an era It may seem peculiarthat stochastic simula- when microscopic data were only scantily tion techniques - otherwise known as the known, and large-scale computing was expen- Monte Carlo method-did not make an earlier sive, there was some excuse for this approach. appearance in this history of theoretical shield- As these obstacles disappeared, use of removal ing research, either in connection with gamma methods deservedly withered away. rays or neutrons. After all, Monte Carlo is one Turning to more rigorous methods of solv- of the few approaches with the promise of ing the neutron transport problem, it is perhaps handling any geometry no matter how complex, natural to ask whether any analytic methods are and, in principle, has the capability of dealing

80 faithfully with the most complicated and de- above, Monte Carlo still has not achieved wide- tailed particle interaction data. In truth, how- spread usefulness. Part of the problem has been ever, Monte Carlo has had a variegated history the continuing need to bias the game. But most in relation to shielding calculations. In the important has been an almost intrinsic limita- earlier days, its application was considered tobe tion in the detail that can be obtained. The need so unreliable that the results were often looked to achieve reasonable statistical accuracy has on as of quite dubious accuracy. The trouble inevitably meant that the simulation has to stems from its stochastic nature, and from the concentrate on obtaining some particular inte- fact answers were always asked for rather im- gral quantity as an answer, e.g., a dose at some probable events-the small fraction of incident location. Thus, to determine the detail, for particles penetrating thick shields. It was there- example, of a flux spectrum is almost always fore generally impossible, especially with the out of the question. A deterministic method, in first primitive computers, simply to follow the contrast, will almost always furnish spectrum particles as they penetrate directly through the detail, usually with much the same accuracy as shield. To achieve satisfactory statistics it was an integral quantity. The future range of appli- almost always necessary to play actookedgame cation of Monte Carlo has the potential of being - biasing the details of the history so as to much wider, however. Monte Carlo lends itself increase the fraction of particles that succeeded easily, and to great advantage, to massively in teaching the desired detector. To do this parallel computation. Even primitive strata- usefully, one has to have a pretty good idea of gems, such as nmning a roomful of PCs on the what portions of the particle phase space are same problem continuously for a week or so, likely to lead to the desired penetration. In other can probably achieve more than the expensive words, it helps to know the answer ahead of straightforward use of a supercomputer. Some- time! Further, the statistical error of biased MY- or some agency -needs only to want sampling is often difficult to fix. There were to have it done. occasional instances historically, therefore, of Tubing to deterministic methods we am sensationally wrong answers obtained by Monte confronted with a bewildering variety of possi- Carlo, when very poor biasing had inadvert- bilities. Most suffered from limitations in ge- ently been chosen. A notion of what poor repute ometries or types of neutron interactions that Monte Carlo had in the shielding community in can be handled. We’ll limit historical discus- those days can be gleaned from a “shoot out” sion mainly to the present day clear winner - that the Shielding Division had arranged in discrete ordinates, considering first the fate of 1963. Some four types of neutron attenuation only one other method, as characteristic of the problems were specified, complete with cross also-ram. sections, and solutions were solicited from all About 1955, J. Certaine proposed a method possible participants. Results were presented for the numerical integration of the Boltzmann and analyzed at the 1963 Winter ANS Meet- transport equation, which found its fruition in a ing. ls Monte Carlo methods were poorly repre- code,knownasNIOBEffortheIBM7090series sented; in the benchmark problem of fission of computers.17 Basically, the transport part of neutrons in water, of the fourteen solutions the equation was evaluated in terms of the submitted, only one was the result of a Monte standard method of characteristic rays, while a Carlo computation. discrete energy treatment was used with the This situation gradually improved over the slowing down integrals determined by Gauss- years, partly from better understanding of the Legendre quadratures. The results of some nature of biased sampling methods, but espe- NIOBE-calculated problems have been pub- cially from the explosion in available raw com- lished18 and a good number more were per- puting power. Clever combinatorial and ray- formed in the early 196Os, but remain unpub- tracing methods were also developed to handle lished. After-that the method sort of faded away, complicatedgeometries. Inconsequence,elabo- and is practically unknown today. rate and sophisticated Monte Carlo programs Why? NIOBE is not restricted to any par- were produced whose use in shielding design ticular source geometry, and can potentially has become almost routine, and practically in- dispensable where streaming in ducts occurs. l6 JhIOBE = Numerical Integration Of the Boltzmann In fundamental theoretical research, as defined Equation.

81 handle any complication in the scattering ker- the discrete ordinate codes have been subjected nel. The reason for its failure to achieve wide- to intense scrutiny both as to the physics and the spread use can probably be explained by mak- numerical mathematics involved. Especial at- ing an analogy to agricultural endeavors. It is tention has been directed to the choice of angu- not enough to prepare the soil and sow the crop larmesh, and to the procedures for accelerating seeds. A lot of work must be done after these and ensuring convergence in the solution of the initial steps in tending the growing plants, till- algebraic equations. Notwithstanding the so- ing the soil, irrigating plenteously, energeti- phisticatedconsiderationsthathavebeenbrought cally weeding, applying insecticides liberally to bear on these questions, their successful (debugging?), and only after all this, setting resolution in practice remains as much an art as about with the harvesting. So too, with any a science. substantial computational system. After the ini- While Los Alamos continued to develop tial formulation and programming, a great in- the SN method, another center of development vestment in person-years must be put into the began at Oak Ridge early in the 196Os, and the subsequent development before the computa- shielding community since then has tended to tional system is generally seen as a dependable, use the codes that came out of Oak Ridge. First well-investigated and useful tool. NIOBE did there was ANISN (1967), a one-dimensional not receive this investment; the winner in the code, followed by the various versions of DOT deterministic transport sweepstakes, discrete for two-dimensional geometries. More recently ordinates, did, and is today almost universally a new generation of codes has appeared - the method of choice in the field. DORT for two-dimensional problems, and The present day procedure referred to as TORTforthme-dimensionalgeometries.(TORT “discrete ordinates” has nothing in common has probably not yet completed its developmen- with the classical method of the same name,19 tal stage.) Around these programs has accmted which has a much closer kinship with the spheki- a large panoply of subsidiary codes, primarily cal harmonics method, and shares many of the for problem preparation and analysis of the latter’s limitations. What we now call “discrete computer output. Especial mention should be ordinates,” or the SN method, as its creators made of linear perturbation codes which, to- preferred to name it, was pioneered at Los gether with solutions for the adjoint problem, Alamosm for their own purposes, and was first can be used for sensitivity analyses of the effect revealed to a wider audience at the 1958 Geneva of small uncertainties in the cross section in- “Atoms for Peace” Conference?1 The S,method puts. These codes are typified by the ORNL developed through at least two metamorphic program SWANLAKE, which has seen wide- forms. As presently employed, it involves di- spread use. The SN method is of course not viding the phase space of position and direction confined to neutrons; it is a trivial extension (in of the “transported” particle into a network of principle) to handle gamma ray penetration. cells, and integrating the Boltzmann equation Combined treatment of neutrons and gamma over each cell. The process reduces the transport rays has become routine, with the solution of the equation to a set of linear algebraic equations in neutron transport problem providing for at least averages over cell volumes and surfaces. Note part of the gamma ray sources. In the big that integration over phase space does not in- applications, e.g., handling full scale reactor volve the energy change upon scattering, for shields, such techniques spell finally the death which a conventional description in terms of a of the (now primitive) use of pre-computed multigroup formulation is one possible option. buildup factors in calculating the gamma ray The assumptions and procedures involved in component of the penetrating radiation.

Some Recent Fundamental Studies of Neutron Penetration

As examples of what has been achieved in above, it had proven possible, with ratherprimi- fundamental shielding research in recent years, tive calculations, to decipher the mode of deep some results will be presented of a number of neutron penetration in hydrogeneous media, studies undertaken at Columbia University over and to relate it to the characteristics of the the last two decades. As has been described hydrogen neutron cross sections. Spurred on by

82 this success, the aim of the later studies was to without the enormous overhead of rerunning try similarly to connect the neutron penetration cumbersome processing codes each time there innon-hydrogenous media to the vagaries of the is a change. The technique is most practically nuclear interactions in these media. The range applied in a hybrid forms in which conven- of materials considered ranged from artificial tional multigroup quantities are used except in elements with specially constructed cross sec- restricted energy ranges in which the point- tion behaviors, to carbon, oxygen, sodium, and energy picture is applied. (especially) iron. Relatively little use has been made, in this Most of the transport calculations were series of studies, of cross section sensitivity performed with discrete ordinate codes, but calculations based on linear perturbation tech- some still used the moments method, and one niques. This remark deserves some amplifica- thesis study involved an unusual application of tion, for it might be thought that such an elegant the Monte Carlo method. Some of the investiga- and sophisticated approach would be particu- tions employed modifications to the standard larly relevant to disentangling the connection discrete ordinate programs which greatly in- between cross sections and neutron transport. creased the information they provided about the However, there are two types of difficulties mechanisms of neutron penetration. For ex- with this approach. One is that the perturbation ample, it can be shown22 that introducing a very due to modifications of the cross section inter- small amount of absorption, as a percentage of actions may be nonlinear even at relatively the total cross section, makes it possible to small changes of input data. This is particularly deduce the average number of collisions it takes what happens in the presence of deep minima in for a neutron to reach a given position and the cross sections. The other problem is intrin- energy. Where certain types of collisions can sic to the simultaneous use of both forward and occur only rarely for a given neutron, it is adjoint transport calculations. A forward calcu- possible to do an order-of-collision computa- lation links a particular source configuration (in tion for that mechanism, by means of an itera- phase space) to all possible configurations of tive solution of the transport equation with a detector response (again in energy and posi- suitably modified scattering keme1.23And these tion). The adjoint calculation, on the other hand, tricks can be stratified, so that, for example, one links a specific detector configuration (in the could find out the average number of discrete same variables) with all possible source con- inelastic collisions occurring in a zone close to figurations. A linear perturbation technique, the source. By these techniques, deterministic which involves products of forward fluxes and transport calculations can yield information adjoint solutions, relates therefore only to a normally thought obtainable only from Monte specific source and a specific detector. Thus, to Carlo calculations. get a broader overall picture of what is happen- It has become customary in transportcalcu- ing one way or the other one must perform sets lations to describe energy behavior in terms of of multiple calculations. Accepting this neces- a multigroup formulation. Such a procedure has sity, it is usually simpler and more economical advantages, especially in handling energy ranges to perform sets of forward calculations to span large compared to the scale of the cross section both the relevant ranges of source and detector variations. But there are also disadvantages, configurations. particularly in the complexity of the cross sec- Something should first be said about the tion handling codes, and in the need to guess at lone Monte Carlo investigation. It would seem weighting functions. However, it is relatively that the Monte Carlo simulation provides a easy to show that any deterministic multigroup unique tool to examine what feature of a neutron transport code can be used to produce answers history distinguishes the “exceptional” neutron referring to a discrete energy grid, onl by using from the crowd with average behavior. By suitable point-energy cross sections. zy4There is “exceptional” is meant the particle that either then the advantage of specifically describing penetrates much further from the source than rapidly varying cross section features, such as the average, or on the other hand lives out its resonances, with greatly simplified scattering history much closer to the source than the kernel processing codes. The effect of different average. In 1976, L-p Ku reported on several sets of cross sections can then be examined such studies in different media.% One of these

83 was a simulation of the element sodium, but a neutron that finds itself in such a minimum with a “featureless” cross section, e.g., only could have a “free ride” to large distances from isotropic elastic scattering, with constant cross the original source point. Under certain circum- section at all energies. In retrospect, the re- stances, deep penetration of neutmns could be search mainly shows the limitation of what completely dominated b the properties of a P could be done in Monte Carlo even on a big single minimum. Preeg, 7 for example, used computer in that epoch. Very roughly, it can be ANISN calculations to show that with a 6 MeV said that the study examined the histories of neutron source in liquid oxygen, the neutron 10,000 neutrons, from a 10 MeV source, that flux at 10 meters was at least 4 orders of slowed down to the vicinity of 25 keV either by magnitudelargerbelowtheminimumthanabove 4 mfp. or by 20 mfp. from the source. It was (see Figure 2). Similar, if less drastic, effects difficult, with the external storage available, to could be constructed with iron. store the details of that many histories; even so, A series of subsequent investigations the number was too small for really definitive showed that in more realistic shielding situa- statistics. Within these limitations, no signifi- tions the presence of cross section minima cant differences could be found in the indi- might have local effects on the flux spectra, but vidual collisions between the exceptional neu- rarely dominated the nature of deep neutron trons and those with average life histories. The penetration. With neutron sources distributedin author notes that at both distances studied, the energy, so few neutrons are born in, or are flux and current density are rather well pm- scattered into, the usually narrow minima that dieted by age theory formulas (even though their effect tends to be swamped by other por- these neutrons presumably have statistically tions of the spectrum. This behavior is most exceptional histories). He goes on to conclude: clearly manifested in more recent researches of ..It seems likely that the different behavior Lieuw,23 on a fission source in pure iron, in of the “exceptional” neutrons arises...out which some 8 prominent cross section minima of (rare) correlations between successive between 1.0 and 2.0 MeV were arbitrarily in- correlations. .. . where the cross sections creased in value to twice their assumed value. It are the same at all collisions, moderately canbe seen from Figure 3, that in the immediate deep penetration. .. occurs as the cumula- vicinity of the minima this “filling in” of the tive effects of small modifications in many minima can have a drastic effect on the flux. But successive collisions which are difficult the effect does not persist at lower energies to to detect in statistical tests... anywhere the same extent. By 1.0 MeV, even In sum, with featureless cross sections, after penetrations through 1 meter of iron, the even reasonably deep penetrations are described minima do not seem to have increased the bystatistical,almostcontinuous,processes(such general level of the flux by more than 20- as age-slowing-down), with little distinction 30%. between one scattering and the next. The smooth The picture of the penetration of fission statistical process is disturbed only when the neutrons through substantial layers of iron can- neutron interactions are markedly changeable not therefore be described simplistically as the with energy. So, the question is what features of dominance of transport in cross section minima. an energy-dependent neutron interaction might Rather, there seems to be the confluence of the determine how far a neutron will penetrate? effects of several cross section phenomena. Attention was first paid to the influence of Source neutrons born well above the threshold sharp minima in the neutron cross section. The for inelastic scattering in iron (which in practice type example, so-to-speak, is oxygen, where means well above 1 MeV) are not particularly there is a prominent minimum at about 2.37 affected by elastic scattering. Nonelastic scat- MeV, going down to a cross section variously terings, however, drastically slow down the measured at 50 to 90 mb, or 5-10% of the neutron, often to below the threshold for inelas- average value in the neighborhood. Many other tic scattering. Thus, for a 14 MeV neutron, iron elements exhibit similar minima, if not always is a better slowing down medium than hydro- to such a marked degree as oxygen. Examples in gen. Once below the effective threshold for materials relevant to shielding include beryl- inelastic scattering, absorption is very rare, and lium, carbon, sodium andiron. It is obvious that the neutron can only wander through the iron,

84 lo-l2

-14 10

lo-l+

1s

12

6

3

.os 1 2 3 4 5 b Energy (MeV)

Figure 2. Cross section vs. energy in oxygen along with flux spectra from a 6 MeV source. (After Preeg, Ref. 27.)

85 0.001 1.4 1.6 Energy (MeV)

Figure 3. Effect of filling in eight cross section minima in iron between 1.2 and 1.9 MeV. (After Lieuw, Ref. 23.) Figure shows ratio of perturbed-to-original fluxes from a fission source due to changes in cross sections. At distances of 0.3,195,30.5,59.5,79.5, and 995 cm, diffusing by repeated elastic scattering until 35 or more by 1 m. from the source. About 15% (most likely) it exits from the medium. Neu- of deeply penetrating neutrons at 1 MeV are tmns from a fission source thus have the high born above 3.8 MeV, and about 25% between energy portion of the flux spectrum rapidly 1.5 and 3.8 MeV, emphasizing the role of eroded, while the region below 1 MeV steadily inelastic scattering at high energies in contrib- builds up. Detailed analysis of the flux spectra uting as a source of lower energy neutrons. ofpenetrating neutrons, with the tools described above, confirm this qualitative picture. Thus, In the shadowy Manhattan Project days, throughout most of a 1 m. slab of iron (with a nearly 50 years ago, shielding theory could fission source), neutrons at 1 MeV have suf- hardly be said to exist. Now we have the tools fered about 1.2 inelastic scatterings throughout to answer most of the questions that could be their history. Most of these have taken place asked of shielding theory, if anyone still wants within 1 mfp. of the source. However such to ask the questions, and is willing to pay for the neutrons suffer many elastic scatterings, up to answers.

I find that here I have been so involved with the unfolding drama, that I have neglected to talk of the many individual actors, for which my regrets and apologies. Especially am I sorry for not having the time and the words to acknowledge in detail the many participants in this history who have given me the golden gift of their personal friendship. The aid and comfort I have gained from their friendship has sustained me during the four decades of my career in shielding.

86 References

1. H. Goldstein, Furuiamental Aspects of Re- 10. R. Aronson, J. Certaine, and H. Goldstein, actor Shielding, Addison-Wesley Publish- “Penetration of Neutrons from Point ing Co., Reading, Massachusetts. (Hence- Monoenergetic Sources in Water,” NY0 forth referred to as FARS.) A detailed de- 6269 (NDA 15C-60), December 15,1954. scription of the method and variants is given here, with references to the original 11. FARS, Sections 6-9 and 6-10. literature. 12. B. Davison, Neutron Transport Theory, 2. FARS, pp. 168-172, and p. 239, Ref. 31 Chapters V and VI, Oxford (1957). therein. 13. K. M. Case and P. F. Zweifel, Linear Trans- 3. Theliteratureonthemomentsmethodinall port Theory, Reading, Massachusetts its ramifications is voluminous. For start- (1967). ers, see FARS, section 5-7, and Refs. 32 and 33, p. 239, along with Ref. 77, p. 241. 14. D. G. Cacuci and H. Goldstein, “Neutron (This last is listed in FARS “to be pub- Slowing Down and Transport in a Medium lished”; it appeared in 1959.) Further his- of Constant Cross Section,” J. Math. Phys. tory is given in Goldstein and Wilkins (Ref. 18.2436-2444 (1977). Very recently, Cacuci 4 below). has made some further progress on the problem: Nucl. Sci. Eng. 108,50-69 (1991). 4. Herbert Goldstein and J. Ernest Wilkins, Jr., “Calculations of the Penetration of 15. R. L. Ashley and W. E. Krieger, Eds., Gamma Rays,“NYO-3075, June 30,1954. Neutron Attenuation in Optically Thick (Section 1.2 describes something of the Shields, ANS-SD-l, April 1964. history of the project.) 16. In this country the earlier SAM-CE pro- 5. D. K. Trubey, “History and Evolution of gram may be mentioned, along with the Buildup Factors,” ANS Shielding Session, more current MORSE and MCNP pro- Nov. 1992. grams. For UK work, see J. Butler, “UK Reactor Shielding: Then and Now,” ANS 6. S. Preiser, C. Jensen, and H. Goldstein, Shielding Session, November 1992. “Guide to SEAC Printed Gutput Tapes of Gamma Ray Penetration Calculations,” 17. J. Certaine and P. S. Mittelman, “A Proce- NDA Memo 15C-83, Oct. 20,1955. dure for the Numerical Integration of the Boltzmann Transport Equation,**NDA lo- 7. FARS, p. 265ff andp. 339,Refs. 62 and 63. 161,1955.SeealsoECRS(Ref.9),pp. 155- 161. 8. J. Certaine et al., “RENUPAK, An IBM- 704 Program for Neutron Moments Calcu- 18. ECRS, pp. 287-293. lations,” NDA 2120-3, December,, 1959. 19. Davison and Sykes, Neutron Transport 9. A partial listing includes hydrogen, water, Theory, Chapter XIII, Oxford (1957). carbon, hydrocarbons, beryllium and the concretes. Results for some are still (de- 20. B. G. Carlson, “Solution of the Transport servedly) unpublished. See Engineering Equation by the SN Method,” LA 1891 Compendium on Radiation Shielding, Vol. (1955). 1, pp. 135-150, New York (1968). (Hence- forth referred to as ECRS.) See also A. D. 21. B. G. Carlson and G. I. Bell, “Solution of Krumbein, “Summary of NDA .. Results of the Transport Equation by the SNMethod,” MomentsCalculations...,“NDA92-2Rev., in Proceedings of the Second International August 1957. Conference on the Peaceful Uses ofAtomic Energy, Vol. 16, pp. 535-549.

87 22. L. Y. Huang, “Mechanisms of Fast Neutron 25. Y-W. H. Liu, “Neutron Transport Calcula- Penetration in Thick Layers of Sodium,” tions Involving a Mixture of Group and Doctoral Thesis, Columbia University Discrete Energy Fluxes,” Doctoral Thesis, (1975). Columbia University (1980).

23. Seng Liek Lieuw, “Mechanisms of High 26. Long-poe Ku, “Monte Carlo Study of the Energy Neutron Transport in Iron,” Doc- Mechanisms of Transport of Fast Neutrons toral Thesis, Columbia University (1984). in Various Media,” Doctoral Thesis, Co- lumbia University (1976). 24. J. Ching, H. Goldstein, and E. M. Oblow, “Application of a Discrete Energy, Discrete 27. W. E. Preeg, “Flux Sensitivity to Neutron Ordinates Technique to the Study of Neu- Data for Iron and Oxygen,” Doctoral The- tron Transport in Iron,” ORNL-TM-4235 sis, Columbia University (1970). (1974). Also, Trans. ANS, Vol. 14,No. 2, p. 835, (1971).

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