Based on icrp, ncrp, and frc guides for protection of individuals and the public against exposure to ionizing radiation, the aec necessarily has issued (and occasionally updates) its general regulations applicable to activities in this area, including nuclear power reactors. These are known as its “Standards for Protection against Radiation” (10 CFR Pt. 20). When it is desired to con­struct and operate a nuclear reactor, the Commission’s “Licensing Production and Utilization Facilities” regulations (10 CFR Pt. 36) come into play as a means of applying the general regulations to a specific reactor at a specific site. There follows a description of the latter process and a discussion of the isolation of regulatory and licensing responsibilities within the aec.

The icrp and ncrp standards for permissible human exposure to radioactive substances are based on the assumption that the permissible amount of radioactive substances accumulated within the body or in the critical organ should not cause the permissible annual dose to be exceeded. These figures are then translated into maximum permissible concentra­tions (mpc) of each radionuclide in air or water using a set of physio­logical parameters that describe the movement of each element to the critical organ, and the daily rate at which the contaminants are inhaled or ingested. In the case of ingestion, the aec regulations give only the mpc’s in drinking water. This is a defect, since ingestion may be by way of food or water. The Federal Radiation Council’s approach is different —and more logical, since their recommendations, called radiation protection guides, focus on the permissible daily intake of a given nuclide, regardless of the source.

Where several nuclides are present, the aec regulations provide a method for weighing the effects of each in relation to the others in such a way that the maximum permissible radioactivity of the mixture of nu­clides takes into consideration the contribution of the individual nuclides. In this case, the method errs on the side of safety. For example, if 131I and 90Sr are present in drinking water, the mpc of the mixture might allow 50 per cent of the 131I permissible concentration and 50 per cent of the 90Sr permissible concentration — despite the fact that one nuclide irradiates the thyroid, the other the skeleton, and the effects are not thought to be additive.

Another safety factor exists where long-lived radionuclides are con­cerned, because the mpc is taken as that concentration which will result in accumulation of the lifetime permissible body burden in 50 years. It can be shown from the mathematics of 80Sr accretion in the skeleton that this provides a significant additional safety factor.

Since the aec regulations are stated in terms of the mpc’s of radio­nuclides in air and water, the regulations implied for many years that if the mpc is not exceeded at the point of discharge to the environment, the dose to humans will not be exceeded anywhere beyond the site boundaries. The point of release in the case of a radioactive liquid effluent is the point at which the waste is discharged to the receiving body of water. In most cases, this is an enormously conservative assumption, since dilution up to several orders of magnitude can take place beyond the point of release. However, it is also possible for physical or biological concentration to take place, and when this occurs, the risk can be correspondingly increased.

Within the past few years, the aec standards have been modified to

allow for biological concentration. In the case of 181I, the mpc in air has been reduced by a factor of 700 to allow for the fact that exposure to man is increased by the tendency of iodine to deposit on forage and eventually pass to cow’s milk. Additionally, the regulations have been modified to require the licensee to demonstrate that accumulations in the food chain are not taking place. The discharges to the environment are considered to be excessive if the radionuclides ingested by a sample of the population by any route of exposure exceed one-third of the annual intake permitted for water and air.

The Commission has always had the right to place upon the prospec­tive licensee the responsibility for demonstrating that such concentration did not take place, and although the aec regulations were formerly silent on this point, no one who has followed the course of reactor licensing pro­cedures over the years ever doubted that the aec has meticulously probed into questions of biological concentration beyond the point of discharge.

Under the aec regulations, a licensee can discharge radioactive waste to the environment in concentrations greater than those permissible for immediate inhalation or ingestion if he can demonstrate the extent to which dilution does take place. Many utilities undertake micrometeoro­logical studies of a proposed site, and on the basis of data generated in this way the licensees are frequently permitted to take advantage of the natural dilution that takes place between the top of the stack and the site boundary. To my knowledge, however, reactor operators have not taken advantage of this approach in regard to liquid discharges. This is due to tha fact that the art of forecasting dispersion in the aquatic environment is not developed to the same degree as forecasting dispersion in the atmos­phere.

The aec requires the licensee to conduct monitoring programs in the vicinity of the reactor. This provides information about the concen­tration of radioactive substances in air and water and also in whatever food products may be grown in the vicinity. Thus, the question of human safety is not left to conjecture but is based on actual measurement of sam­ples collected from the environment. Some of the aec facilities, such as Oak Ridge and Hanford, have been collecting data for more than a quarter of a century; experience at these places has produced valuable information that in many cases is directly applicable to civilian power reactors.

For years, many of us in the field of public health and environmental protection have argued that, on balance, electrical generating stations powered by nuclear fuels make better neighbors than do stations using coal or oil. It is true that the current generation of nuclear plants discharge 40 per cent more heat to the environment and this places more stringent limitations on the use of water for condenser cooling, but regulations deal­ing with this problem are being promulgated in the various states for application to both nuclear and fossil fuel stations.

Much has been said about the ecological effects of radioactivity dis­charged to the environment, but there is no evidence that this occurs at levels of radioactivity permitted by the aec. Putting it more strongly, there is a considerable body of scientific data that demonstrates that such effects do not take place. In contrast, we do know that certain vegetation is adversely affected by traces of sulfur dioxide and possibly by other com­ponents of the combustion products of coal and oil (Stem, 1968). There have been millions of dollars spent investigating the ecological effects of low levels of ionizing radiation exposure — but there have been compara­tively few studies of the ecological effects of the chemicals in fossil fuel effluents, despite the fact that we know these effects take place and can be observed.

In most parts of the country, fossil fuels are the only practical alter­native to nuclear fuels. We know, beyond any doubt, that sulfur dioxide discharged to the environment by plants burning fossil fuels has been responsible for many deaths in the general population, particularly during periods of meteorological stagnation. Even the innocent gas carbon diox­ide, produced by combustion of fossil fuels, is accumulating in the earth’s atmosphere and is regarded as a long-range threat to the world’s heat balance, with the possibility of eventual climatic changes on a disastrous scale (Conservative Foundation, 1963). Finally, it is a curious fact that because radium and other radioactive substances are normally present in fossil fuels, the radioactive atmospheric emissions from fossil fuel plants are not insignificant compared with those from many nuclear plants. (Eisenbud & Petrow, 1964; Fish, 1969). These are among the reasons that some of us are convinced that nuclear reactors make good neighbors.

Additional reasons are to be found in the actual operating experience of the civilian power producing reactors. The atmospheric and liquid effluents are in most cases less than 1 per cent of the amounts permitted by aec standards, and the public health risks, though finite, are so small as to be more than offset by even the most modest of the benefits of in­creasing man’s available electrical resources.

Conclusions

From the foregoing, together with various additional information that has been presented by other contributors to this volume, it is possible to draw certain conclusions which constitute the thesis of this presentation and which argue that although the record of the aec has been a good one from the point of view of the public health official, changes in the present regulatory system are needed to reconcile differences between public attitudes and the aec that have not been resolved after 15 years of almost continuous debate.

There are obvious advantages to having radiation protection stand­ards that are applicable on a national scale, there being no reasons why the standards applicable in one state should be more or less stringent than in another.

The aec regulations are substantially compatible with the recom­mendations of icrp and ncrp. Moreover, they are both scientifically and philosophically compatible with evaluations of the state of our knowledge of radiation effects that have been undertaken from time to time by other national and international bodies, including the United Nations Scientific Committee on the Effects of Atomic Radiation, the National Academy of Sciences (Reports of the Committee on the Biological Effects of Atomic Radiation, 1956), and the British Medical Research Council (1956).

The aec regulations have resulted in a safety record that is probably unsurpassed for any new industry. In the 27 years that have passed since the first reactor went critical in December 1942, there has been time to evaluate the basic adequacy of the systems of control that have been derived.

Although there are ambiguities, inconsistencies, and perhaps even deficiencies in the aec regulations, they are sufficient to protect the public’s health. The standards contain enormous built-in conservatism.

There are mechanisms by which local government and individual citizens can bring to the attention of aec the need for changes in its regu­lations. The aec techniques of publishing new rules or proposed changes in rules and the public hearing associated with the licensing procedure are examples of how the thinking of local government or individual groups can be incorporated into the aec regulatory procedure.

The present system of aec regulation, which puts major emphasis on the maximum permissible concentrations of radionuclides in air and drink­ing water, should be changed in favor of specifying the maximum permis­sible daily intake from all sources. This is the method used by the Federal Radiation Council and is preferable because it automatically considers such factors as multiple sources of exposure and the ecology.

Neither ncrp nor aec is sacrosanct, but considerable weight must be given to the fact that the ponderous procedures of these organizations have produced a set of regulations that are workable, and that have suc­cessfully protected the public’s health for more than a quarter of a century.

An examination of 27 years of experience would seem to indicate that the aec has been fully prudent in discharging the responsibilities which the Congress bestowed on it in the health and safety field. However, this judgment is not shared by everyone. For reasons which are probably related to factors other than the excellent safety record it has achieved in the nuclear power field, the aec does not have the high degree of public confidence that is necessary for smooth development of the electrical gen­erating industry. There remains a credibility gap which has not been closed after more than 15 years of debate.

A significant factor in the credibility gap is the unusual dual respon­sibility of the aec for both development of civilian nuclear power and protection of the public’s health. I myself believe that the aec has an ex­cellent record of accomplishment in both areas, and has retained a high degree of objectivity in facing its responsibilities for health and safety, but the public is not fully convinced that this is so. For this reason I believe it would be in the public interest to begin active consideration of the means by which the regulatory responsibilities of the aec can be transferred to or shared with some other governmental agency. Only in this way can the public be assured that the present apparent conflict of missions is not operating to its detriment. However, a transfer of regulatory responsibility cannot be accomplished easily. The aec has well-developed regulatory machinery of a type that does not exist in any other branch of government. Although in theory it would be possible to transfer this entire organization to another agency, this would not be wise because interagency transfers are always disruptive of morale and working efficiency.

As a compromise, the Public Health Service should be given a more prominent role in the regulatory program. The Public Health Service rather than aec should promulgate the numerical standards of permissible exposure. The aec, with its highly developed capability to evaluate reactor designs, should continue to consider applications for new reactors and should continue to monitor construction and operation to assure com­pliance with the terms of the licensee. However, the Public Health Service, in its traditional collaborative relations with the states, should undertake the responsibility of effluent monitoring and ecological surveillance. By sharing its present statutory regulatory authority with the Public Health Service in this way, one may hope for the closing of the credibility gap that now exists between aec and many segments of the public.

REFERENCES

aec. Operational accidents and radiation exposure experience. 1943-1967. Wash­ington, D. C., 1968.

Conservation Foundation. Implications of rising carbon dioxide content of the at­mosphere. New York: the author, 1963.

Donaldson, A. W. The epidemiology of lung cancer among uranium miners. Health Physics, 1969, 16, 563.

Ehrenberg, Lars, Gunter von Ehrenstein, & Abne Hedgran. Gonad temperature and spontaneous mutation-rate in man. Nature, 1957, 180, 1433-1434.

Eisenbud, Merril. Environmental radioactivity. New York: McGraw-Hill, 1963.

——- & Henry G. Petrow. Radioactivity in the atmospheric effluents of power

plants that use fossil fuels. Science, 1964, 144, 288-289.

Environmental Radioactivity Exposure Advisory Committee, Department of Health, Education, and Welfare. Environmental contamination by radioac­tive substances. December 1, 1968.

Fish, B. R. Radiation in perspective — the role of nuclear energy in the control of air pollution. Nuclear Safety, 1969, 10, No. 2.

International Commission on Radiological Protection. Committee II. Report on permissible dose for internal radiation. Vienna, 1959.

——- . Committee I. The evaluation of risks from radiation. Health Physics,

1966, 12, 239-302.

Medical Research Council. The hazards to man of nuclear and allied radiations. London: H. M.S. O., 1956.

Penna Franca, Eduardo, et al. Status of investigations in the Brazilian areas of high natural radioactivity. Health Physics, 1965, 11, 699-712.

Rajewsky, B., & W. Stahlhofen. Polonium-210 activity in the lungs of cigarette smokers. Nature, 1966, 209, 1312.

Russell, W. L. Recent studies on the genetic effects of radiation in mice. Pediat­rics, 1968, 41, Suppl. No. 1, Pt. II, 223-230.

Stem, A. C. Air pollution, Vol. I, Ch. 12. New York: Academic Press, 1968.

United Nations Scientific Committee on the Effects of Atomic Radiation. Report. General Assembly Official Records: 19th Sess., Suppl. No. 14 (A/5814). New York: United Nations, 1964.

——- . Report. General Assembly Official Records: 21st Sess., Suppl. No. 14

(A/6314). New York: United Nations, 1966.

By far the largest source of this energy flux is the solar radiation in­tercepted by the earth. Of this flux, the possible channels amenable for use as sources for industrial power are: heat from direct solar radiation, water and wind power, and power derived from the stored energy of pho­tosynthesis.

The total power input from direct solar radiation amounts to 17.7 X 1016 thermal watts. However, this occurs at the low power density of only 0.139 watts/cm2 outside of the earth’s atmosphere, and at greatly reduced density over most of the earth’s surface.

According to Farrington Daniels (1964, Table 1, p. 22), the average solar power reaching the earth’s surface amounts to about 500 cal/cm2/ day. This, when averaged over a full day, amounts to about 2.8 x 10 2 watts/cm2. Large modern power plants have power capacities of about 1,000 megawatts, or 109 watts. Solar cells are capable of converting radi­ant energy to electrical energy with an efficiency of about 10 per cent. Hence, for a 1,000-megawatt solar power plant, it would be necessary to collect 1010 watts of solar power. For the average radiation on the earth’s surface, this would require a collection surface of 36 square kilometers, or a square area 6 kilometers to the side.

Although this is not a large area, the amount of electrical equipment

required to collect this amount of intermittent solar power and to convert it into a steady power output is formidable as compared with that for equivalent power stations using other sources of energy. As long as other sources of energy are available at much lower costs, solar power appears to offer little promise as a source of large-scale power. Solar power for special purposes, such as for rural telephone systems or for spacecraft, is practical, however.

In typical power reactors, engineered safeguards in the form of sev­eral separate and independent methods of cooling the core under a spec­trum of theoretical accidents are provided. Usually included are systems which will prevent the reactor core from overheating to any damage levels even in the event of a major rapid loss of normal reactor coolant water. The design basis of engineered safeguards is strongly influenced by the understanding and appreciation that a strong barrier keeps fission prod­ucts from being released from power reactors. If the nuclear fuel can be prevented from overheating to the point of melting during various loss-of — coolant situations, then the fission products will be kept principally in the fuel rods.

For purposes of example I shall pay particular attention to the net­work of engineered safeguards commonly referred to as the emergency core cooling network. Consistent with designers’ primary concern for maintaining a barrier for fission product releases, definite criteria were established very early in the design of engineered safeguards for the boil­ing water reactor. As the industry matured and more was learned about the phenomena associated with emergency core cooling, more exacting criteria evolved. Today, criteria for loss-of-coolant accidents are as fol­lows:

1. Fuel cladding temperature will be kept below maximum tempera­tures at which experiments have verified that fuel rod integrity would be maintained. Normal fuel cladding temperatures are about 2,000° F below this, so considerable safety margin exists.

2. For any size of break to the primary system causing the reactor core to lose coolant, at least two completely independent emergency core cooling systems shall be available to provide effective emergency core cooling.

3. The emergency core cooling network for the boiling water reactor will involve at least two methods for the cooling process. Today’s boiling water reactor uses the methods of both reactor core flooding from below the core and reactor core spraying from above the core. If there should be any unknown phenomena associated with either process, the other process will still operate to achieve adequate emergency core cooling.

4. Although there is usually very dependable off-site power provided to the emergency core cooling network, there shall be no reliance upon off-site power. Appropriate on-site diesel generators or gas turbines will be provided to supply the power to run the emergency core cooling net­work.

These are the four basic criteria upon which the emergency core cooling network for today’s boiling water reactor has evolved. Just how the above criteria are satisfied in today’s power reactor is graphically illus­trated on what is now known as the “boiling water reactor bar chart for emergency core cooling” (Fig. 4). For any break size found along the abcissa of the chart, there are always at least two bars representing indi­vidual emergency core cooling systems which could provide adequate protection in case of a loss of coolant. The two major systems on the boil­ing water reactor are the core spray system and the low-pressure coolant

injection system. Two full capacity core spray systems, each with its own pumping and power supply, are provided. Either of these spray systems can provide complete core cooling in case of loss of coolant. The low — pressure coolant injection system has a capacity such that even if one complete section of the system were lost, sufficient time is available to recool the core before any serious overheating. Other systems, such as the high-pressure coolant injection system and the auto-relief system, are also part of this network.

Notice that the entire network of systems is fully integrated. The systems work together as a set, providing protection for the smallest leak up to the hypothetical complete instantaneous severance of one of the main recirculation lines. There is no dependence on off-site power; and the entire system is fully automatic — it does not require operator intervention at any time during the initiation of the emergency core cooling systems. A feature unique to the boiling water reactor, which is a direct-cycle system, is that in spite of any nominally sized loss of coolant which might occur to the primary system, the reactor vessel itself is constantly being supplied with a large flow rate (from 5 to 10 million pounds of water per hour) directly into the pressure vessel for the purpose of steam gen­eration for the turbine. In general, this flow will overwhelm any small leakage which might occur. This is another inherent safety feature of the boiling water reactor, direct-cycle concept.

When the backup core cooling system has been preliminarily de­signed, it is subjected to detailed study to search for possible points of weakness or ways in which it could be improved. In the study, designers of course call upon the experience gained in the industry over the past two and one-half decades of operation of large nuclear reactors of various types. In addition, they are making increasing use of the highly developed techniques of reliability analysis and systems engineering which have been used with such success in the space program.

In the detailed reliability studies, such things as proper electrical power arrangements, proper sensing devices and sensing device arrange­ment, proper inspection programs, and proper redundancy requirements can all be evaluated by the disciplines associated with reliability technol­ogy. Reliability analysis in safeguards work for power reactors is being effectively used in Great Britain, Canada, and Switzerland and is now coming more and more into play in the United States as well. General Electric has employed and intends to continue to employ reliability analy­sis to assure that the highest levels of safety are achieved on power re­actors.

of emergency core cooling equipment, careful analytical investigations must be conducted for each and every type of accident in its full range of magnitude. Detailed analyses, using major digital computer programs, are conducted for entire spectrums of accident conditions. It is these analyti­cal investigations which are the subject of extensive audit during the period that a particular power reactor project is being reviewed by the aec. Another important aspect of the design of each emergency core cool­ing system is the extensive experimental programs that must be conducted to verify that the system performance claimed has indeed been achieved. For example, in each General Electric boiling water reactor in the current product line, the fuel bundles are all identical whatever the reactor size. The fuel bundle consists of a set of 49 fuel rods, each of which is 12 feet long, containing uranium dioxide pellets encased in zirconium tubing. Fuel rods are clustered together by appropriate spacers and tie plates and are encased in a channel box with appropriate nosepiece and upper handle. This individual fuel bundle has been simulated at full scale, with electrical heating in place of the nuclear heating, and the entire simulated fuel bundles have been completely tested to evaluate the performance of the core spray systems and core flood systems. In this system, each fuel bundle receives spray cooling. This was an extensive program, but each of the claims made for the emergency core cooling equipment has now been confirmed.

Containment. If any of the core fission products should be able to find their way through all of the barriers and into the air space outside the reactor process system, they would then encounter the further substantial barriers of the plant containment systems. The containment system com­ponents have no normal operational requirement for retention of radio­active materials. Thus, they are simply insurance and are needed only in case of simultaneous and significant failure of all of the process barriers and engineered safeguards itemized above. The containment structure totally encloses the nuclear steam supply system of a nuclear power reactor. On the boiling water reactor the entire nuclear pressure vessel is inside the primary containment. The feedwater lines from the turbine and the steam lines back to the turbine are the only major lines which pene­trate that containment. The containment structure is of high quality with very stringent leakage requirements placed upon it. The design basis for the containment is that even in the hypothetical event of the complete instantaneous severance of the biggest primary system pipe and the sub­sequent blowdown of the steam and water found in the primary system to the containment structure, the containment structure will remain within its design pressure. Therefore, the containment structure is another safe­guard in the design of nuclear power reactors.

A particular type of containment structure has been associated with the boiling water reactor. The containments of all the modem boiling water reactors to date employ a principle known as pressure suppression in the design of their containments (Fig. 5). By means of vent pipes in the large water source within the containment structure itself, large mag­nitudes of steam which would be released from the nuclear vessel in the event of a hypothetical primary system rupture are forced to condense in the large cold water supply found in the suppression pool. In this way, although the containment may see large pressure levels for a short period of time (still within its design rating), the suppression action of a large body of water soon condenses most of the steam released by the pressure vessel and the pressure level within the containment drops to very low levels. This has strong safety implications. Without suppression action all of the mass released to the containment would stay at high temperature and pressure, thus leading to potential increased leakage rates from the containment. By means of the pressure suppression principle and the rapid reduction of residue pressures within the containment, leakage from the containment is also minimized.

Another important feature of today’s boiling water reactor contain­ment system is the fact that a secondary containment also exists. Around

the structure known as the primary containment, commonly called the drywell and suppression chambers, the entire reactor building forms a secondary containment barrier. By the means of a ventilation system, the building can be held at a slight vacuum so that any leakage at all is from the outside into the building. This ventilation system sends the atmos­phere of the secondary containment through a standby gas treatment system should an accident occur. The standby system, which includes filters for fission products, provides still another barrier to fission product release outside the nuclear power plant.

Recognition of all of these facts points out the extreme measures which have been taken in the use of containment to ensure a maximum of public health and safety — again, a dramatic manifestation of a proper attitude toward reactor safety.

The licensing process is basically a two-step affair. A utility must first obtain a construction permit before it may commence construction of the plant. After construction is completed, the utility must obtain an operating license before it may begin operation of the plant. There is a thorough review of the health and safety implications at each of these stages, and there is no guarantee that even if the plant is constructed in strict accordance with the construction permit, operation will be licensed. As a practical matter, however, an operating license is unlikely to be de­nied under such circumstances, so that, realistically, issuance of the con­struction permit is the critical element in the licensing process. It is vir­tually a certainty once the construction permit is issued that aec will li­cense operation of a nuclear power plant at that site and that only tech­nical and operating details, if even these, will remain to be argued about at the operating license stage.

I focus, therefore, on the construction permit phase of the licensing process. This is initiated by the applicant’s filing of the license application, which includes a multi-volume preliminary safety analysis report. The ap­plication is studied at great length and in truly amazing detail by the aec regulatory staff. Voluminous correspondence flows back and forth, and the preliminary safety analysis report is amended and supplemented as a result of the staff’s searching inquiries and suggestions. While the staff is studying the application, a parallel study by the aec’s Advisory Commit­tee on Reactor Safeguards is under way. The acrs is a prestigious and conservative committee composed of scientists and engineers from outside the aec, representing several disciplines, who are asked to review and pass on the safety of each proposed nuclear power plant. Both groups are concerned with one principal, ultimate issue, Does the application meet the aec’s safety criterion for issuance of a construction permit? That is, is there “reasonable assurance that. . . the proposed facility can be con­structed and operated at the proposed location without undue risk to the health and safety of the public?”

When these two groups are satisfied as to the “reasonable assurance” test, which usually is at about the same time since their efforts are to some degree coordinated, each issues a report. Typically, the acrs report is terse and frequently includes suggestions for additional research and de­velopment or invites attention to safety aspects which the acrs believes warrant further thought or action. But the regulatory staff’s report, known as the safety analysis report, is long and comprehensive. Both reports are available for public inspection; indeed, a principal purpose of these re­ports is to inform the public about the health and safety aspects of the proposed facility.

Both reports invariably conclude with the judgment that there is “reasonable assurance that the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public.” I say “invariably” because ordinarily neither of these bodies will submit its report until it is satisfied as to “reasonable assur­ance.” In the light of experience, it appears that, if either body would emerge with a negative conclusion, the utility would undoubtedly with­draw its application.* Accordingly, this phase of the licensing process does not come to an end until the prestige of the acrs supports the appli­cation and the aec staff has been committed to the proposition that all health and safety standards have been met.

The next phase of the licensing process is a hearing before a three- man Atomic Safety and Licensing Board. Each Board is designated ad hoc and consists of three members drawn from a panel. Membership on the panel, except for the panel’s chairman and vice chairman, is a part­time avocation; two members of each board are “technically qualified,” and the third member, the chairman, is a lawyer. The principal function of the board is to determine whether the “reasonable assurance” test has been met.

The development of the nuclear industry in the United States has been different from most other industrial development. In a very real sense, it is one of the first deliberate attempts to understand and control the risks of an emerging large-scale technology. This approach has taken a great deal of planning, research and development, training and careful operations. This point was emphasized in a pioneering report by the Na­tional Research Council, National Academy of Sciences (1956), which stated: “The use of atomic energy is perhaps one of the few major tech­nological developments of the past 50 years in which careful considera­tion of the relationship of a new technology to the needs and welfare of human beings has kept pace with its development. Almost from the very beginning of the days of the Manhattan Project careful attention has been given to the biological and medical aspects of the subject.”

As a result of this approach, the United States atomic energy pro­gram has a record as one of the safest of industries, from the standpoint of radiation hazards as well as of ordinary industrial risks.