ML20214M932

Suppl a of FSAR on Nbs Reactor

ML20214M932
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Site:	National Bureau of Standards Reactor
Issue date:	10/01/1966
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• National Bureau of Standards Reactor l

NBSR 9A -, 7u o.. g.* "'y' ":~ L l E, c o m ' ~ muut iuu g NBS Reactor Radiations Division Institute for Materials Research I l October 1,1966 l NBS gNM"i js u. i ' \\ U.S. DEPARTMENT OF COMMERCE g ete-l NATIONAL BUREAU OF STANDARDS \\- t m h/ r., . b 'V / 'k[S[ t A PDR

I I I I THE NATIONAL IlUREAU OF STANDARDS The National Itureau of Standards in a principal focal point in the Federal Government for assur-ing msmimum application of the physical and engineering s<.iences to the advancement of technology in indu try and commerce, lin respon.ibilities include desclopment and maintenance of tha national standard of measurernent, and the prosi. ions of means for making measurements consistent with thoae standardet determination of phs.ical constants and properties of materialst development of methods for testing materials mechani.mg. and structures, and making such tente as may be neces-anry, particularly for government agencies; cooperation in the establishment of standar'd practices for incorporation in codes and specification *t advisory service to government agencies on scientific and technical problemnt invention and development of devices to serve special needs of the Govern. mentt a.$istance to industry, businc.s, and consumers in the development and acceptance of com-mercial standards and.implified trade practice recommendationst administration of programs in cooperation with United States business groups and standards organizations for the development of mternational standards of practicet and maintenance of a clearinghouse for the collection and dissemination of scientific, technical, and engineering information. The scope of the Bureau's activities le suggested in the following li. tine of its three institutes and their organizational units. Institute for llaale Standarda. Applied afathematics. Electricity. Metrology. Mechantes. Ifeat. Atomic Physics. Physical Chemistry. Laboratory Astrophysics.' Radiation Physics. Radio Standards Laboratory:' Radio Standards Physics; Radio Standards Engineering. Office of Standard Reference Data. Institute for Materiala Heacarch. Analytical Chemistry. Polymers. Metallurgy. Inorganic Mate-rials. Reactor Radiations. Cryogenics.* Materials Evaluation Laboratnry. Office of Standard Itefer-ence Materials. Institute for Applic.l Technologv, Duilding Rea, arch. Information Technology. Performance l Test Development. Electronic In=trumentation. Textile and Apparel Technology Center. Technical m Analynis. Office of Weights and Meanures. Office of Engineering Star.dards. OfLee of Invention and Innovation. Office of Technical Resoureca. Clearinghoune for Federal Scientific and Technical Information.* * 'Isated at flauldre, Calarada, in10t. "luated at 5285 l' wt Havel Hnad. 5prinnfield. Virginia 22171. I I I

e E NATIONAL BUREAU OF STANDARDS REPORT 9419 I I Supplement A of the Final Safety Analysis Report on the National Bureau of Standards Reactor NBSR 9A YN October 1, 1966 (66[ l' IMPORTANT NOTICE NATIONAL BUREAU'0F STANDARDS REPORTS are usually preliminary or progress documents intended for use within the Government. Before material in the reports is formally published it is subjected to additional evaluation and review. For this reason, the publication, reprinting, reproduction, or open-literature listing of this Report, either in whole or in part, is not authorized unless permission is obtained in writing from the Of fice of the Director, National Bureau ~ of Standards, Washington, D. C. 20234. Such permission is not needed, however, by the Government agency l g l g for which the Report has been specifically prepared if that agency wishes to reproduce additional copies for its own use. U.S. DEPARTMENT OF COMMERCE NATIONAL BUREAU OF STANDARDS I

I PREFACE The letter of August 2,1966 in the Docket No. 50-184 for the NBSR has requested additional information in a series of twenty-two questions. The following Supplement A to the report No. 8998 entitled NBSR 9, Final Safety Analysis Report on the National Bureau of Standards Reactor contains the re-sponse to these questions. Each response is preceeded by the I original question of the above mentioned letter. I

QUESTION 1. Provide an analysis of the likelihood and consequences of the following previously unanalyzed accidents: 1.1 Loss of cooling flow from throttle or isolation valve failure. I 1.2 Loss of cooling flow from pump shaft seizure or impeller failure. 1.3 Reactivity insertion from breaking of a shim rod or its shaft with the rod falling to the bottom of the core structure. 1.4 Reactivity insertion from securing of an inadvertently operating helium bubbler system. 1.5 Reactivity insertion from the seating of the spring-loaded fuel assemblies (in conjunction with a loss of flow). 1.6 Reactivity insertion from an " unrestricted excursion," using final core parameters. 1.7 Total loss of electrical power. 1.8 Total loss of instrument and service air. Each of these accidents, as well as accidents previously addressed, should be analyzed both with and without scram so that their most probable and upper limit consequences can be evaluated.

RESPONSE

1. INTRODUCTION Numerous accidents have been analyzed in Section 13 of NBSR 9, I Final Safety Analysis Report on the National Bureau of Standards Reactor, including the startup accident, loss of coolant, maximum reactivity insertion, and unrestricted excursion. In addition, certain accidents such as those due to beam tube flooding and cold coolant addition were briefly discussed and related to the most severe accident, that due to a fuel element handling accident. It was determined from this discussion that only the loss of coolant accident would lead to possible fuel element damage and release of 1-1

volatile fissiot product radioactivity. 1.1 There are, hewever, a class of accidents or circumstances which are of less serious consequence than those mentioned above which can be discussed more explicitly. Consideration has been given to the consequences of loss of flow from throttle or iso-lation valve closure where the reactor is operating at maximum power of 10 Mw. This question has been faced previously and partially discussed in Section 4.7.8 and 4.7.9 of NBSR 9. The actual mechanism for total loss of flow from valve failure is subject to question because of. redundancy and fail-safe instal-lation of operators; but it is, of course, credible that power may be lost from the main D 0 circulating pumps. In the event 2 that power is lost, one of two shutdown pumps should start

• since each is supplied with both an AC and a DC motor. Flow in the primary loop is monitored with three separate flow measurement systems, each of which supplies indication, alarm, and automatic reactor scram.

Since these systems as well as the reactor scram system are fail safe to the scram condition in the event of loss of power, it would appear to be incredible that the reactor would not scram in the event of loss of flow. If it is assumed that the reactor does scram in a time short compared to the pump coastdown, then a calculation can be made of the heat transfer conditons which obtain for the fuel element sit-ting ta the core surrounded by reactor moderator. It was assumed that the fuel element could lose heat only by heat transfer from the side plates to the moderator, no credit being taken for loss of heat to tre coolant within the channels between plates of the element during or af ter pump coastdown. At the moment of shut-down the heat source from fission product decay is approximately

67. of the operating power prior to shutdown assuming the normal fuel irradiation cycle. Approximately one-half of the energy is in the form of gamma radiation, some fraction of which would be expected to escape the element.

If it is assumed, however, I

• As discussed more fully in NBSR 9, Section 4.7.8.

1-2

that all decay heat is deposited within the element then the maximum heat flux frora the side plates to the moderator will 2 be approximately 25 watts /cm. This heat flux is approximately a factor of ten less than required to reach the condition of film boiling, i.e., the heat flux is still well within the nucleate boiling region. If a calculation is made of the ap-proximate element centerline temperature required to cause this heat to flow to the sideplate, the centerline temperature would be less than 200 C above the side plate temperature which is not far from the moderator temperature. The element therefore is in no danger of melting even at the heat flux condition which obtains directly at shutdown. If it is further considered that the heat capacity of the element and the water within the channel will delay the onset of the condition which has been calculated until the decay heat release rate has fallen by more than a factor of two, then clearly the element is not likely to be damaged. As stated in Section 4.7.9 of NBSR 9 these predictions are amply borne out by observations

• on the ORR reactor which has been run at even higher power densities with similar plate elements, and shutdown with simultaneous loss of flow without fuel element damage.

In fact, boiling was not observed except at a shutdown power level greater than 15 Mw. The above stated report gives an empirical equation for calculating the heat transfer coefficient at the onset of observed boiling. Although it is difficult to be certain from the nature of the report that the equation is directly applicable to calculations for the NBSR, the heat transfer coef ficient computed f rom this empirical ex-pression is within a factor of two of that used in the above stated calculations. If one considers the question of loss of flow without simultaneous reactor shutdown then different conditions cer-tainly obtain. Clearly the heat flux rate is sufficiently T. E. Cole and J. A. Cox, " Design and Operation of the ORR," Vol. 10, 86, Second Geneva Conference Report (1958). 1-3 w

high to reach film boiling at some point as the flow coastdown takes place. It is expected, however, that the very boiling process will lead to coolant temperature increase and to channel void and consequent decrease of moderator density, resulting in increased neutron leakage or decreased reactivity. Experiments

• at DIDO reactor at Harwell, England confirm these assumptions.

As. stated in the referenced report, the contrib-utions of these effects to reactor stability have been studied both by simulator and on the DIDO reactor itself. An experiment was performed to demonstrate the effect of a failure of both main heavy water pumps while the reactor was operating at 6 Mw. Since this condition would trip the reactor scram it was neces-sary to bypass these safety features. The power trace following pump failure shows a rapid decrease to 50"/. of initial power in I a time comparable to pump coastdown with a further slower de-crease afterwards. The fuel plate temperature rises rapidly to a maximum at the saturation temperature of the coolant before decreasing as the power diminishes. Since the temperature and void coefficients of DIDO are very similar to the NBSR corresponding results would be expected. An exact calculation is difficult in the transient condition. A simple calculation I shows that the NBSR temperature coefficient alone would put the reactor on a 5 second negative period by the time the moderator reaches boiling, i.e., the reactor power would decrease on a period which is comparable to the flow coastdown period. This is approximately what is observed in the DIDO cxperiment. Further confidence in the inherent stability and self-shutdown capability of this class reactor is given by the long period Sport II transient tests. Transient No. 00392 of the l previously referenced report ** shows an example of an unrestricted 364 millisecond excursion for a core configuration very close to K. Q. Dagley, et.al., "Research Reactor Utilization," Vol.10,, 3, 2nd. Geneva Conference Report (1958).

• • V. W. Goldsberry, IDO 16990, September 1964 1-4

I the NBSR. The transient was performed without reactor flow and was observed to self limit at 12 Mw. The reactivity compensated af ter the peak power was reached was greater than 0.57, and the I resultant decay period was of the order of 0.5 sec. The total energy released af ter the peak power was reached was less than I 5 Mw-sec. The fuel element temperature was observed to exceed the saturation temperature of 100*C only very slightly and very briefly. If it is considered that a reactivity increase is associ-ated with loss of flow because of seating of fuel elements as discussed in Section 1.5 below, then the worst circumstance 1 cads to an insertion rate of approximately 0.47. reactivity per second. This insertion rate would put the reactor on a period of approximately 5 seconds. Scram set points would be passed in less than i second, thus giving a fourth independent mecha-nism for automatic scram. In any case the resultant reactor period is much longer than was discussed above as leading to inherent self-shutdown in the case of Spert II. It is safe to assume, therefore, that no fuel element damage is likely with a loss of flow accident in its most severe form with no reactor scram. Clearly no structural damage is possibic. Even if fuel cladding should fail for some unexpected reason, the primary system would remain intact I and fission products would not leave the reactor vessel. No danger to the of f site public can be foreseen f rom such a cir-cumstance as discussed. 1.2 A consideration of loss of cooling flow from pump shaft seizure or impeller failure is a lesser accident condition than discussed above in Section 1.1, since the main circulation pumps are redundant and simultaneous failure is highly improbable. In any event the previous discussion covers the condition for complete loss of flow. I

I 1.3 1. INTRODUCTION The possibility of a mechanical failure of a shim am or shim am drive shaf t in such a way as to cause a reactivity insertion is extremely remote. The CP-5 reactor at Argonne I has used such a system without this type of failure for over 15 years. The DIDO and Pluto reactors at Harwell have about 10 years each of successful experience, and more recently this type of system has been in use at the Georgia Tech reactor. In all these accumulated years of experience, there has been no case of such failure. Nevertheless, several types of failures, which would introduce the most severe conditions, will be discussed. First, the result of shim am drive shaft failure will be investigated and next the result of a failure in the shim am itself. I E DRIVE SilAFT FAILURE A drive shaft failure in the form of a complete break allowing tl a f ree rotation of a shim arm will be considered under operating conditions and under shutdown conditions. 2.1 Failure During Ope rat ic a. The free rotation of a shim arm during operation would result in a large negative re-activity insertion. This is due to the fact that the shim am can not mechanically go significantly beyond its full-I in position. The shim am guides used to guide the shim am during installation are located such that they prevent the motion of the shim ams more than a few degrees beyond the full insertion position. Even without the guide, the shim am would be prevented f rom f alling out of the core by hit-ting the bottom grido plate. Thus, this type of fatture does not cause a positive reactivity insertion, but instead would cause a reactor scram. 2.2 Failure During Shutdown. For the reasons cited above, the failure would result in a rotation of the shim arm of only 1-6

1 a few degrees beyond the full-in position. Reference to Figure 4.20 of NBSR 9 indicates that this would result in a negative reactivity insertion. Ilowever, even if the result was positive, I as would result from the complete removal of the arm, the re-maining three shim arms would always be adequate to maintain the reactor subcritical. 3. SilIM ARM FAILURE I If the shim arm were to fail so as to be completely broken into two pieces, the break would occur near the hub where the I greatest strain is. Such a failure is considered under shut-down and operating conditions. 3.1 Failure During Shutdown. As discussed above, even the complete removal of a shim arm during shutdown would never be sufficient to cause the reactor to go critical. 3.2 Failure During Operation. If the shim arm were to break, it would fall into the core being guided by the fuel elements and prevented from slipping out of the core completely by the lower D 0 tank. It would probably hang up on one side and re-2 main in a highly negative reactivity position causing reactor shutdown. It might, however, fall completely through the core and rest along its whole length on the bottom grid plate. Cal-culations using the EQUIPOISE 3A program show that the arm in I this position is worth -1.07. Ap. If complete latitude of shim arm configurations is allow-I ed, it is possibic to achieve a situation in which almost all the excess reactivity of the reactor is held down by a single shim am. This would be the case if three ams were fully withdrawn and the remaining arm inserted as necessary. Under this condition the inserted arm would be holding down all the excess reactivity which might be as great as 127.. Thus, it is desirable to limit the shim arm configurations allowed under normal operating conditions. During most of the operating cycle, the availabic excess I reactivity is less than 57.. During startup, however, before 1-7

I xenon and samarium have built up and while the reactor moderator I is still cool the available excess reactivity might be as high a s 127.. Although this would be an unusual situation, and pre-sent only during the start of a new operating cycle, it will be examined as the most severe situation. The reactivity controlled by any one shim arm can be mini-mized by requiring all four arms to operate as a bank such that all arms are in the same position. It is clear, however, that if the reactor is sub-critical with all ams at the same setting, the further insertion of one or more arms decreases the excess I reactivity which could be inserted by the failure of any one arm. Furthermore, any two shim arms, fully inserted, can shut-down the reactor. So if three arms are fully inserted, the fourth may be positioned as desired since a failure of any arm l would still leave two fully inserted, assuring reactor shut-dowa. Based on these facts, the following restrictions on allowed shim arm configurations will result in minimizing any potential accident that could result from a shim arm failure. 3.2.1 Allowed Shim Arm Configurations. Only the follow-ing will be allowed during normal reactor operation (1) Three shim arms fullyinserted; the remaining arm I may be moved and positioned as desired. (2) All four shim arms may be moved together as a bank with no arm position differing from the mean position by more than 1.- (3) Each arm may be moved and positioned as desired, provided no arm is withdrawn beyond the position at which the banked set of four rods would allow the l reactor to be just critical. 3.2.2 Maximum Excess Reactivity Due to Shim Arm Failure. As discussed above, the maximum excess reactivity that can be inserted into the reactor under the conditions stated in sub-section 3.2 1 would result f rom the f ailure of one of the four banked arms. If the banked shim arms were holding down 127. I excess reactivity (reactor critical), the breaking of one am 1-8

such that it fell through the core and lay along the bottom grid plate would result in the maximum possible excess re-activity insertion. Calculations using the EQUIPOISE 3A program show that the reactivity change due to the removal j of the highest worth arm would not exceed 2 07.. If this arm also happened to be 1 more fully inserted than the mean, an additional.47. Ap must be added to give a maximum reactivity change of +2.47.. The fact that removal of the highest worth I arm introduces only +27. Ap, a value less than the average apparent arm worth of one-fourth of 127. or 37., is due to the marked interaction between shim arms in the top reflector. A single arm at that position would be worth almost half of the whole 12%, and the reactivity change due to the addition of each successive shim arm would decrease as each arm was added. Hence the addition of the fourth arm, or conversely, its re-moval would change the reactivity significantly less than 37.. This qualitative argument is substantiated by the calculations I mentioned above. This condition is not as marked for fully in-serted arms in the core region where the medium absorption is I' inuch higher and the interaction between shim arms is, consequently, much less. In this case, the worth of a single arm is only one and one-fourth as great as one-fourth the worth of four arms, and the last arm about three-fourths this value. Since the fallen shim arm is worth -17. Ap, the net re-activity insertion would not exceed 2. 4-1. 0=1. 47.. This value is well below the 27. excess reactivity insertion discussed in response to question 1.6. 1.4 A discussion of possible consequences of reactivity in-sertion f rom inadvertently securing of a helium bubbler systerr j leak requires knowledge of what leak might go undetected. In i the first place, regardless of a helium bubbler leak detector, t it is certain that sustained accumulation of helium to the helium sweep system would be noticeable from gas holder level g 1-e 1

I increase. Secondly, it is df fficult to imagine that more then a fraction of one percent reactivity could be held in the form of helium void during routine normal operation without being noticed by Reactor Operations. Only 2.3 percent is available to provide uranium burn-up during a cycle and deviation from expected reactivity should be clearly noticeable. Examination I of the helium bubbler system indicates that a redundant system of pressure measurements and associated alarms should assure I no helium leakage without alam and automatic scram, (see the discussion in response to Question 12 of this submission). Finally, however, a helium flow meter indicates helium flow when it occurs. If it is assumed that the mini::.um flow that this meter can reliably detect is flowing and suddenly stops without prior operator action, a reactivity accident can be postulated. The minimum detectable flow would be approximately 6 cfm which corresponds to approximately 0.75*/ reactivity. If this helium flow stops, the corresponding helium void uculd I rise from the core in a time of approximately three seconds. The reactivity insertion rate would thus be approximately

0. 25*/ pe r sec.

This rate is less than that previously discussed above. In any event the ultimate reactor period wculd be approximately 0.600 sec. The reactor rundown and scram interlocks would be actuated in less than 1 second. In case of failure of reactor scram, the situation is similar to that discussed in the event of loss of flow above, with fuel element seating. A transient would take place with the above stated period which I is clearly within the margin discussed previously as tolerable on the basis of Spert II experiments. No fuel element damage would be expected before self-shutdown would teminate the excursion. In any event, the primary system would remain intact and no fission product release would be expected. 1.5 Under normal flow conditions each fuel element is forced up tightly against the top grid plate. If flow should cease, however, each element will drop slightly and sect in the bottom grid plate. The total motion will not exceed 3/16". The drop I in the core is equivalent to a small withdrawal of the shim arms. 1-10

Thm equivalcnt incr::s2 in r activity can be deduced from Figure 4.20 of NBSR 9 and the geometry of the core and shim arms. The result of the calculation shows that the maximum reactivity insertion possible is 0.47. op and during normal operation it I would be only 0.27.. This is less than the worth of the regulation rod. I Since this. reactivity change is caused by the movement of fuel elements and since each fuel element would fall at a slightly dif ferent moment, the reactivity insertion is probably spread I over a period of about a second. However, even if a step in-sertion is assumed, any resulting excursion would be much smaller than the one analyzed for the 27. reactivity insertion. The effect of the fuel element seating during a loss of flow accident is discussed in the answers to Sections 1 and 2 of Question 1. 1.6 An analysis of the consequences of a maximum reactivity insertion has been made utilizing the data from Spert II tran-sient tests.* The final NBSR core configuration is very similar to that of the so-called Spert II BD-22/24 expanded core. In fact, this particular configuration was chosen in the Spert II tests because it is " typical of that used in research reactors consisting of highly enriched uranium-aluminum I alloy, plate-type fuel assemblies in heavy water (e.g., MITR, DIDO, CP-5, etc.)"** The major difference in fuel element design, other than central core gap, between the NBSR and Spert II BD-22/24 is the number of plates per element and the thickness of the aluminum clad. Each of the twenty-four Spert II elements contain twenty-two plates, each plate having identical fueled area to each of the seventeen plates of the twenty-four NBSR l elements. From the point of view of transient heat transfer and shutdown void formation the increased surface area is almost exactly compensated by the increased clad thickness of the Spert II l plates and consequent increased energy release for a given moderator void formation fraction. IDO-16990 op. cit. p. 3 I 1-11 ~

I As stated in NBSR 7 and NBSR 9, the original transient calculations were made by first fitting transient studies to Spcrt I results for an H O moderated reactor core. By adjusting 2 calculations to t!ese results, an effective heat transfer coefficient could be arrived at for use with the NBSR. A new series of calculations have been performed using, the latest core parameters for the NBSR and the results fitted to the transient No. 00582 of Spert II using the same analog computer techniques discussed in NBSR 7 and NBSR 9, Previous results indicated that approximately 24 Mw-sec would be released in 27. reactivity transient for the NBSR. The latest parameters indicate that the NBSR lifetime should be increased over that used in the calculation by approximately 40 percent. At the same time, however, the void coefficient should be reduced by approximately 55 percent. These changes I are of opposite effect in terms of total energy released in a reactor transient since the longer the lifetime in a prompt critical excursion the longer the transient period and, con-sequently, the less the energy release. Vice-versa, the smaller the void coefficient the greater the energy release for a given transient period. A parameter study in which the void co-efficient.was varied in the analog computer program indicates that the total energy release for the NBSR transient corresponding to a 2 percent reactivity insertion is calculated to be I approximately 40 Mw-sec as compared to 26 Mw-sec measured for Spert II. Since the expected error in these calculations and measurements is probably of the order of 20 percent *, the prior expectation of no major difference seems adequately borne out. The total reactivity compensated in this excursion was greater than 6 percent so that there is ample margin to account for the somewhat different void coefficient for Spert II and the NBSR. The maximum fuel plate temperature was less than 250 C and this was reached only momentarily. No melting or damage to the core was observed. It is certainly clear that the margin of safety in the NBSR is considerable if it is remembered that I

• This estimate is based on the degree of fit of the data over a range of parameter variations.

1-12

I 34 Hw-sec is required to bring the core to melting if no heat is lost to the coolant. It is just the heat that is lost to the coolant, however, which provides the large margin of shutdown reactivity fedback to the system to terminate the excursion. 1.7 Total loss of electrical power is a highly unlikely event as discussed in NBSR 9 since storage battery power is available in addition to standby diesel generator power. In any event, total loss of power would mean reactor scram since the shim safety arms are maintained in a standby in'sertion mode by an electromagnet clutch engagement. The conditions which would I obtain in the reactor are exactly those of the loss of coolant flow discussed above. As previously analyzed, the elements cannot be damaged in such an accident and fission products would not be released. The one additional consideration would be the time required to restore some degree of coolant flow to remove core after-heat. Since the primary and emergency cooling systems would remain intact, several hours are available to restore some form of diesel or back-up battery power to the emcrgency shut-down pumps. Partial or total core unloading to the storage pool could be considered and would be feasible with only simple I emergency procedures. As a last and ultimate resort, if none of these situations are manageable, H O could be added from the 2 I standpipe source to the reactor vessel and heat removal by boiling would be adequate. 1.8 Total loss of instrument and service air is an accident of even less consequence for the NBSR than any previously dis-cussed. The reactor instrumentation system is essentially completely electronic so that no unsafe condition will obtain following loss of air. Numerous valve operators are, however, air controlled so that a total loss of air will have specific and clearly definable consequences. It should be noted that complete loss of air is probably incredible since certain of the critical valves have air accumulators as a stored energy I back-up in event of loss of primary air supply. I 1-13 J

An analysis of the situation which follows complete loss I of all air, however, shows that most primary valves fail to the open position. This includes the top reflector dump valve which will indirectly cause reactor rundown upon opening. This latter fact is assured since both reactor overflow and reactor level alarms and interlocks will be initiated following dump valve operation.. In any event, the loss of the top reflector would finally terminate reactor operation. Since reactor flow can continue no cooling problem would be involved and no fission product release would take place. I I I I I I I I I I I I 1-14

I I QUESTION 2 Provide the results of an analysis which determines the maximum I reactivity insertion whose energy could be contained without ruptur-ing the reactor vessel or the primary system. The question of maximum system pressure generated by a reactor transient has been examined by reference to the 'ransient pressures t measured in Spert II tests of the so-called expanded D 0 core. As I 2 previously discussed, this core has characteristics very similar to the NBSR. A summary of transient pressure data is given on page 128 of the previously referenced report.* The design pressure of the NBSR reactor vessel was 50 psig and it has been tested to 75 psig. It can be seen from Figure 208 of the reference that the maximum pressure observed in the 0.062 sec period transient discussed in Section 1.6 of this analysis was less than 25 psig. An extrapola-tion of the Spert data would indicate that a transient whose period is approximately 0.030 see would be required to reach the test I pressure of the NBSR vessel. Such a transient would require a re-activity insertion in the NBSR of approximately 3.1% reactivity. This is approximately 1.1% greater than the maximum reactivity insertion of 2% assumed in the accident analysis of NBSR 9. The margin of safety seems especially conservative when it is considered that those pressures which were observed were highly transient pressures, persist-ing for fractions of a second, and, therefore, would not be expected to be representative of the sustained vessel wall pressures that were generated in the vessel test procedures. I

• ID0-16990 I

I I 2-1

I QUESTION 3. What is the maximum positive reactivity that could result from selective fuel melting and redistribution. Discuss how this po-tential reactivity source could affect previously analyzed ac-cidents. 1. INTRODUCTION Any melting and redistribution of fuel in the NBSR is ex-ceedingly unlikely. Two cases, however, which could possibly result from other unlikely but assumed accidents will be dis-I cussed. One is the redistribution of fuel during complete core meltdown and the other is the local hot spot melting which, although very unlikely, might result from the accidental ex-cursions discussed in the accident analysis. 1.1 Core Meltdown. In spite of the extensive emergency cooling features of the NBSR, a complete core meltdown has been consider-ed. This would not be possible in the presence of water and if water were not present no redistribution of the melted fuel could be critical. One feature of the NBSR, however, is a tank en-closing the lower half of the core which in all probability would keep the lower core section immersed in water and there-I fore keep it from melting. Under this condition, it is con-ceivable that the uncooled top section might melt and flow into the lower section. To analyze this case, the following assump-tions were made: (1) Only fuel melted and entered the lower core. (2) The fuel distributed itself uniformly within each I element of the lower core. 235 (3) The total U mass was used, and no credit taken for any disadvantage factor. I (4) The fuel was homogenized throughout each annular region of the core as was done in the normal calculations. 3-1 1

E (5) The retaining tank was filled completely bringing the water level to the top of the fuel. Under these conditions, the resulting multiplication factor, k, I was 0.829, about 17% subcritical. Therefore, this redistribution could not contribute to any accident. 1.2 Local Hot Spot Melting. Although fuel melting is not ex-pected from even an uncontrolled excursion resulting from a reactivity insertion as great as 2% the assumption is made that such an excursion results in local melting at hot spots. The most severe type of redistribution would result in fuel being introduced into the gap region. If it is assumed that the equivalent of as much as the top half of one full fuel element found its way into the gap, a reactivity increase of 1.17. would result. This calculation is based on perturbation theory and I the normal and adjoint flux data available from the EQUIPOISE 3A code calculation. In this type of an excursion, the total void generated is several times that needed to terminate the excursion. Thus, by the time the fuel could have melted and moved into the gap, the void would have introduced a negative reactivity significantly greater than the positive 1.17. reactivity caused by the re-distribution. Therefore, even this large fuel redistribution l would not contribute to the accident. I I l l I I l 1I I 3-2

I I QUESTION 4. Extend your evaluation of mechanisms for pressure differential leading to building outleakage following a loss of coolant ac-cident to include: I a. release of stored heat energy in the spilled coolant ~ and the system metal; b. core decay heat; c. solar heating and cooling. Re-evaluate your dose calculations to include any additional fission products which could be released through these mechanisms. These calculations should also include considerations of smaller barometric changes than previously analyzed using associated more stable meteorological conditions. 1. SOURCES OF_ PRESSURE INCREASES Following a loss of coolant accident, there are several sources of heat which might result in warming the air within the building causing an increase in building pressure which would necessitate additional pumping by the emergency system in order to maintain the desired pressure differential across the building walls. These sources are the heat energ'y stored I in the D 0, the core decay heat, and solar heating of the 2 building. Another source of increase building pressure would be the saturation of the air by the spilled D 0 if it were lef t 2 exposed to the air. The contribution from each of the sources is discussed below. 1.1 Heat Energy Stored in D 0. The D 0 inlet temperature is 2-2 l 100 F and its outlet is 112 F. The outlet goes directly to i the heat exchanger where it is cooled to 100 F. About half of the water is in the tank and the other half in the external system. Thus the average D 0 temperature is about 106 F com-2 pared to a room temperature of 75 F. The total water mass under consideration is about 50,000 kg. This large mass and = 4-1

I the high specific heat of water makes the D 0 heat content much I larger than that in any of the mechanical structure. So only the D 0 heat content will be investigated. 2 The area under the D 0 piping is surrounded by a curb so 2 any leaked D 0 is contained and would be pumped into the storage 2 tank. The transfer would require up to two hours during which I time some heat would be transferred to the surrounding structure. Once in the storage tank which is isolated in a sunken pit, its I rate of heat transfer would be greatly reduced and the cooling ef fect of the secondary water in the heat exchanger plus the cool water in the storage pool would readily compensate for the remaining D 0 heat transfer. 2 During the time the D 0 remained on the floor, it would 2 loose heat to the thick concrete floor slab in addition to heat-ing the air. It could only heat the concrete slowly, however, and so its temperature would be only slightly lowered in two hours. Therefore assume that its temperature remains unchanged I for its two hour stay on the floor. (The heat capacity of the water far exceeds that of the air in the building.) The only remaining question is then the heat transfer to the air. This would surely take some time and would probably not be completed in two hours. If, however, we assume con-servatively that negligible time is required for the air tem-perature to reach equilibrium with the water, the average tem-perature rise of the air in the whole building can be estimated. The air over the water is confined to the sub-pile room I and is only mixed with the air in the rest of the building by the emergency reefrculation fan. This fan would add warm air from the process room to the main building at the rate of 1000 cfm. Thus in two hours, 120,000 ft of the 106 F process room air would be exchanged with the rest of the building. The 3 average temperature increase in the 120,000 ft of exchanged air would be 31 F. The volume of air in the process room all of which is assumed to have reached 106 F is about one-fourth of the 600,000 ft building volume. Thus the average temperature I rise in the building air would be 14 F which would give rise to a 4-2

I pressure increase of 2.6*/. resulting in an additional 2.6*/. of the building volume that would have to be pumped out. 1.2 Core Decay Heat. The melted core would be confined by I the reactor structure to the interior of the reactor vessel and thermal shield. Thus, its heat would have to penetrate the thermal shield before it could warm the air. There might I be some small exchange of air with the sut;-pile room below the reactor, but this is a tightly closed room with thick concrete walls that would not transfer a significant amount of the sub-pile room heat to the exterior building air. Thus, essentially all the decay heat would be removed by the thermal shield cooling system whose pumps are supplied with emergency power. Therefore, the core decay heat would not contribute to a pressure rise in the building. 1.3 Solar Heating. The best data available on the effects of solar heating are the temperature data taken during a building leak test run from noon on July 23 to 1430 hours July 24, 1966. I The atmosphere was very clear during this period, so the solar heating was close to the maximum available. The temperature curves are shown in Figure 5.4 in the response to Question 5, and the internal temperature is shown on an expanded scale in Figure 4.1. The temperature rose approximately linearly from 78.8 F at noon until 2000 hours. It then gradually began to level off reaching a plateau of 81.6 F at midnight which lasted until noon the next day when the temperature began to rise again. The available data was not sufficient to determine the I second rate of rise or even to confirm that the second rise would continue significantly. The data can be interpreted as showing a second temperature rise starting about noon with a rate approximately the same as that observed the day before. There appears to be very little correlation between the outside and inside temperatures. The inside temperature con-tinued to rise long after the outside temperature had dropped below the inside temperature. The inside temperature correlates much better with the solar cycle. It rises for 14 hours, remains steady for 10 hours, and then begins to rise again. This pattern 4-3

I is displaced from the true solar cycle by about 7 hours. This I hysteresis can be attributed to the mass of concrete that must be heated before heat is transferred to the air within the building. The fact that the internal temperature does not drop significantly during the night is due to the fact that the ex-ternal concrete surface temperatures generated by the sun's rays exceed the internal temperatures by much more than the internal temperatures exceed the night external surface temperatures. This explanation of the data shows that the temperature would continue to rise each day until the difference between the I internal temperature and external night surface temperature was great enough to compensate for the large solar surface heating. In summary, the experimental data show that the maximum temperature rise in a day would be close to 3 F. The data do not allow us to predict how much less the temperature rise would be on successive days, but they would surely decrease as the tem-perature of the air within the building rose. Thus, it is con-servative to assume that the air temperature rises 3 F per day. This would result in a pressure rise of less than 0.67./ day re-I quiring an additional 0.67 of the building volume to be ex-hausted each day. I 1.4 Vapor Pressure. The relative humidity in the building is maintained at about 507.. Therefore, if the D 0 were to saturate I 2 the air, the pressure increase would be one-half the vapor pres-sure of water at the temperature of the air. If the air temper-ature is conservatively assumed to be 80 F, (normally it is maintained below 78 F) the pressure increase would be 13 mm of Hg. This would result in a 1.77. pressure rise during the first day provided the two hours is assumed to be sufficient time to completely saturate the air in the process room. 1.5 Summa rv. Under the conditions outlined above, an additional 4.97, of the building volume would have to be pumped out as shown in the Table below: I 4-4 I

I Summary of Pressure Increases I Source pressure rise Ileat content of water 2.67. i Core decay heat 0.0 Solar heating 0.6 Increased vapor pressure 1.7 Total

4. 97.

I After the first day, the additional exhaust rate required would be 0.67. of the building volume. I The calculations above are based on conditions associated with the worst credible accident. Although failure of the thermal shield cooling and a situation resulting in the inability to pump the D 0 into the storage tank would not be associated 2 with the worst credible accident, the results of such conditions will be discussed briefly. If the D 0 remained on the floor, all the air in the build-2 ing would reach the 106 F of the D 0 as compared to 89 F for the 2 conditions above. This would result in a op of 6% instead of I 2 67.. If the thermal shield cooling failed, the decay heat would still have 'to heat up a large mass of concrete before it could start warming the air. Then the warming air would be cooled by the storage pool, the building walls and other structures. Assuming that this is equivalent to heating up one-fifth of the structure within (but not including) the walls (a volume ap-proximately equal to the biological shield and closely associated structures), the air temperature would rise at the rate of about I 3 K/ day based on the heating rate at the end of one day.

Thus, there would be an additional continuing pumping requirement of about 1% of the building volume per day. This rate would de-crease with time falling to about 1/27. per day at the end of a 30 day period.

l If the spilled water were lef t exposed to the air, and if its temperature were to rise at the same rate as the air, the I 4-5 I 1

I vapor pressure would rise at the rate of 6 mm/ day contributing 0.87. Ap/ day. The solar heating, of course, would be unaffected. 2. DOSE RECALCULATION In response to the question of dose calculation re-evaluation in view of the pressure sources outlined in 1.5 above, reference should be made to Section 13 of NBSR 9 and to the documents in the record of these proceedings entitled, " Statement of Position of National Bureau of Standards on Order to Show Cause" dated May 10, I 1963, " Addendum to Statement of Position of National Bureau of Standards on Order to Show Cause" dated July 12, 1963, and "AEC Staff Analysis of Population Growth and Hazards in Connection I with the NBS Reactor Site" dated August 6, 1963. If the combined 4.97. pressure rise in the first day is added I to the 6.257. discussed on page 13-7 of NBSR 9, the dose to the thyroid for a person standing on the nearest site boundary for a period of two hours would rise to approximately 18 rad under the neutron diffusion conditions accompaning the rapid barometric fall. If this dose is then added to a revised 30 day dose which would result from the most persistent wind conditions, a combined effective dose would be approximately 63 rad. The revised long term dose of 45 rad results from the conservative assumptic,n that an additional 0.67 leakage rate is added to the average rate of 1.25% Secause I of a continued solar beating for the entire 30 day peri ad, an unlikely extrapolation, as discussed above. If consideration is given to the effect of the revised leakage rates upon dose calculations for atmospheric conditians which are more stable than for the above calculation, then reference should be made to Table 1 and Table 2 of the AEC Staff Analysis mentioned above. If the Icakage rate is to be increased for the strong inversion condition it is reasonable to assume that only 4.3% increase is effective since strong solar heating would not be expected to accompany strong inversion. If this assumption is made the dose, as calculated at 1050 meters in Table 2, is increased I from 22 rad to approximately 54 rad. If the long term persistent wind calculation is combined with this initial strong inversion 4-6 J

I condition the resultant dose at 1050 meters is slightly less than o the 63 rad total of the previous calculation above.* In regards to the consideration of potential dose to persons in nearby high rise buildings, previous discussion of this question I occurs in the record. (See for example page 15 of the AEC Staff Analysis previously cited). The nearest high rise building planned would be farther away than the 1050 meters used in the calculations above. (See Question 20 of this submission). The dose in the center line of a plume at the elevation of the reactor stack would not be greater than a factor of two different from the above mentioned calculation. The previous discussion in the record of whole body dose resulting from the hypothetical accident conditions obtaining in the above calculations are not altered by the reconsiderations of pressure effects. The whole body dose for a person on the site boundary is due entirely to direct radiation from the reactor building and is not effected by release rates. Whole body dose I due to exposure in a stack plume is negligible compared to direct radiation exposure. I I I I I I see footnote 6 of the "AEC Staff Analysis of Population Growth and Hazards in Connection with the NBS Reactor Site" dated August 6, 1963. 4-7

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I I QUESTION 5. Provide sufficient information and evaluation of the confinement building, including where needed, details of equipment and schematic diagrams to verify that the building can reliably perform all of its intended functions following an accident. In particular, provide: a. Test results which demonstrate that all areas of the building can be maintained at a negative pressure by the emergency ventilation system. I b. Test results which demonstrate the margin existing in the emergency ventilation system in the event that one cf the automatic building closures fails to operate properly or building leakage exceeds design. c. Test results which verify that the internal recirculation system is capable of processing the entire building volume I rather than continuously recirculating some small fraction of the total volume. d. A discussion of what corrective measures can be taken by an I operator in the emergency ventilation control station and a quantitative summary of the values or conditions under which he would take these measures. Include a description of all controls and indicators at this station.

RESPONSE

a. On July 23,1966, an operational test was performed on the re-actor building emergency exhaust system. The purpose of this test was to demonstrate the adequacy of the system in the normal emergency operating mode. The test was conducted over a 24-hour period from 1200 hours on July 23 to 1200 hours on July 24. The building was closed using the manual " major scram" button; thus, isolating and sealing the building in the normal emergency situation. The exhaust fans functioned properly to reduce the internal pressure to a negative 0.25 inches H O with respect to 2 exterior atmospheric pressure. Special test manometers at various locations internal to the building (see Figures 5.1 and 5-1

I 5.2) were used to measure internal-external differential through-out the 24-hour period. As can be seen by Figure 5.3, the interior pressure remained negative at all locations through-out the test. Figures 5.4 and 5.5 show the temperature and ambient pressure variations for the test period. As seen from Figure 5.4, the. confinement interior temperature increased from solar heating but stabilized about the 13th. hour of the test. I The test occurred under stable atmospheric conditions and generally a rising barometer, thus the " harmful" effects of I a rapidly falling barometer could not be observed. At one point in the test, however, at 1530 on July 23, the barometer did show a sharp decrease of 0.4 mm of Hg in a 30 minute period. The exhaust system apparently had no difficulty in keeping up with this drop since there was no uniformly measured decrease in building interior vacuum. Figure 5.6 shows the hourly wind speed and direction for the duration of the test. As shown, the wind was generally from the SSE at speeds varying from 16 to 2 mph. The building vacuum data I plotted in Figure 5.3 shows that over these varying wind conditions the internal building pressure was everywheres clearly negative. b. The second portion of the ventilation test was performed to I observe the building internal pressure subsequent to introducing leaks igto the evacuated building. This test was conducted from 1200 to 1435 hours on July 24. Three modes of leakage were used. The first test was designed to measure the effects of the swinging fire doors and vestibule doors at the NE personnel entrance. These doors were closed and the seal of that sliding door deflated. Building internal pressure stabilized at a negative 0.15 inches H O with one emergency 2 exhause fan in continuous operation (see Figure 5.7).

Next, I

the fire doors and vestibule doors were opened with the sliding 5-2 I

E* door seal deflated. The interior pressure of the building was maintained at approximately minus 0.05 inches of H O 2 with both emergency fans in continuous operation (see Figure 5.8). The last and most severe test was conducted with the seal deflated on the truck door to the reactor building. I All other doors were sealed. Interior building pressure continued to remain negative by 0.03 inches H O utilizing 2 both exhaust fans (see Figure 5.9). At one point in this test, the indicated pressure at the NW location did drop to zero but at no time did it go positive. The wind was constantly from the SSE at speeds varying from 6 to 14 mph at the time of this test. In addition to the pressure measurements, a definite inward draft was observed using cigarette smoke with the truck door seal deflated. This I inward draf t was also detected with only one emergency fan in operation. In summary, it has been demonstrated that the emergency ventilation system is capable of maintaining the reactor building at a negative pressure even with certain sealing devices having failed. The test was not performed to obtain exact leak rates, etc. but was a practical demon-stration'of the relative margin built-in to the emergency exhaust system. c. The internal recirculation system services all three levels of the reactor building, primarily through separate ducts specifically for this system. The first floor of the reactor building is purged with a 2,000 CFM flow entering the room through a 36" x 10" supply register on the east wall at elevation 446'-3" and returning through an 18" x 28" register located in the southwest corner of the room I at elevation 436'-6" approximately 80 feet away from the supply. The room volume is approximately 155,000 cubic feet giving one volume change in approximately 78 minutes. Flow rates were verified usiag an Alnor Type 3002 velometer to be approximately I 5-3 I

I 800 ft/ min at the supply register and 570 ft/ min at the return I register. 2,000 CM4 of air is supplied to the second floor of the reactor I via a 36" x 10" register in the southeast corner of the room at elevation 476'-9". Return from the room is through a 36" x 14" register approximately 95 feet from the supply at elevation 470'-0" in the southwest corner of the area. This 2,000 CFM services a volume of approximately 284,000 cubic feet giving 142 minutes for a volume change of air. Face velocities at the supply and return register were measured at approximately 800 ft/ min and 570 ft/ min respectively. Air flow to the process room is by way of an 18" x 10" register which exhausts 1,000 CFM of air directly from SF-19 to the Mezanine level at elevation 420'-0". This air is then drawn into the normal process room duct at the discharge of SF-11 via an 18" x 12" automatic damper. This air is then supplied and I distributed to the process room through eight supply registers varying in sizes: three at 20" x 8", two at 36" x 20", two at 30" x 16" and one at 20" x 16". All supply registers are located at elevation 422'-10". Return is through a single 18" x 18" register and butterfly valve ACV-11 located at elevation 419'-9" in the southwest corner of the process room. Flow velocities were verified using a velometer to assure proper balancing of the supply dampers. The area containing approximately 64,000 cubic feet would experience a volume change in approximately I 64 minutes,

d. The emergency ventilation control station is intended to be a remote monitoring point for the reactor building closure and emergency exhaust systems.

This panel has the following instruments, switches, and indicating lights: I There are "open" and " closed" indicating lights for (1) ventilation valve, ACV-11. This valve opens on a signal to start the internal recirculation fan SF-19 to return air from the process room for cleanup. Refer I 5-4 s

I to Figure 12.7 of Question 12. No valve control switch is provided since there would be no reason to open the valve without SF-19 running. Should this valve fail to open cleanup of the process room air by SF-19 would be impaired; however, supply air from SF-19 to the process room would continue, being exhausted through EF-5 and EF-6 emergency exhaust fans via ACV-10. (2) "Open" and " closed" indicating lights are provided on all normal ventilation intake and exhaust valves, these include ACV-1, (see Figure 12.4), first and second floor air conditioning intake valve; ACV-2, reactor basement I intake valve, ACV-3, (refer to Figure 12.8), reactor basement exhaust valve ACV-6, (Figure 12.12), ir-radiated air exhaust valve, and ACV-7, (Figure 12 11), first and second floor exhaust valve. These valves operate from contacts in their respective supply and exhaust fan starter circuits. No control functions are provided for these valves. Should ACV-3, ACV-6, or ACV-7 fail to close air could be discharged through absolute filters up the exhaust stack. If ACV-1 or I ACV-2 failed to close, air could be exhausted, with a positive pressure in the building, to the intake areaway; however, this air would first have to diffuse into the intake duct system before escaping. (3) Valve position lights are provided for ACV-10. (see Figure 12.17). This valve permits 25 CFM of air to be drawn from the process room by the emergency ex-haust fans, EF-5 and EF-6. It is operated from contacts in the AC and DC starting circuits for EF-5 or 6. Should the valve fail to open no air would be exhausted directly from the process room; however, the recirculation I system SF-19 would continue to operate to cleanup the process room atmosphere. 5-5 I

I (4) The reactor building vacuum relief valve ACV-12, can be opened, closed and placed in automatic operation from the emergency ventilation control panel. "Open" and " closed" lights are provided. Normally this valve operates in the " auto" position (see Figure 12 21). a pressure controller, PS-151, will open ACV-12 on a sensed -2 5 inches H O pressure in the 2 I reactor building and reclose when the vacuum is de-creased. This controller operates only this valve. Should it fail to operate, the reactor operator I could open and close, the vacuum relief valve, thus regulating building negative pressure. He can observe building internal pressure on a separate indicator also located on the panel as discussed below. (5) Recirculation cleanup fan SF-19 can be operated from the panel. An " automatic" "off" "on" switch and indicating lights are provided (see Figure 12.7). In the " auto" position the fan is started by an open contact. in the "FSR" relays on a major scram signal. Should the fan fail to start, the reactor operator would place it into operation to recirculate e and cleanup internal air. If ACV-ll failed to open, and release of radioactive materials was only to the process room, as determined from a large difference from normal in readings on RI-1-1 and RI-1-8 (see below), SF-19 could be stopped, thus limiting the spread of radioactive materials to other portions of the building. (6) Emergency exhaust fans EF-5 and 6 are provided with AC and DC power switches as well as "off" "on" lights on the panel (see Figures 12.15 and 12.16). Each AC and DC switch has an " auto", "off" and an "on" position. In the " auto" position for AC to EF-5, this fan will start on a major scram signal to reduce I the internal building pressure to -0.25 inches H 0. 2 5-6 l J

Should it fail to start EF-6 AC will start automatically. If AC power is not available, the DC power will start EF-5. If DC power fails to start EF-5, the other ex-haust fan EF-6 will start. The reactor operator can start and stop either fan using AC or DC if the auto-matic operating features fail. He would operate the fans to reduce internal pressure to -0.25 inches of H O as sensed on PI-10 indicator (see below). 2 (7) Position lights " Opened" and " Closed" are provided on each of EF-5 and 6 suction and discharge valves (see Figure 12.17). These indicators on ACV-4, ACV-5, ACV-8, and ACV-9 inform the reactor operator that the valves I are functioning properly. Should a valve fail to operate when its respective fan is started the operator can place the other fan in service. (8) An "On" "Off" switch and lights are provided for EF-2, I the dilution exhaust fan (see Figure 12.19). This fan normally would continue to operate after a building closure to provide additional dilution air. Should one of the exhaust valves ACV-3, ACV-6, or ACV-7 fail to close this fan would evacuate the reactor building through the open valve. If the reactor building pres-sure decreased without EF-5 and 6 in operation, the dilution fan would be turned off under these circum-stances to limit the amount of air passing through an I open valve. (9) The reactor operator can observe the building internal pressure using an installed pressure indicating channel, PI-10. Its range of -4 to +10 inches H O covers the 2 operating range of the building. Based on this indicator, he can determine the correct effect of the emergency exhaust system to reduce and hold the internal pressure negative. 5-7 I

(10) FI-19, emergency exhaust flow rate, is displayed at the panel. This instrument would supply the operator with supplemental information on proper operation of the emergency exhaust fans, EF-5 and EF-6. Should flow be excessive with one fan in operation, it could possibly rnean a malfunction of the filter system for that fan. The operator would use this information to decide to place the other fan in service. This instrument would I also supply information on the tightness of the building since flow rate and time of discharge could be used to calculate the total amount of air discharged. (11) Area radiation monitors in the process room and on I the first floor of the reactor building have remote indicating meters on the panel. This information would inform the operator of the effectiveness of the internal recirculation system by indicating the direct radiation in these areas. The process room monitor has a range of 0.1 to 100,000 mr/hr and the first floor monitor covers 0.01 to 10,000 mr/hr. (12) Position lights for RW-13 and RW-16 are included on tbc panel, informing the operator for their automatic closure. Should they fail to have closed, he would manually isolate the liquid waste system in the waste I storage pit where the emergency ventilation panel is located. (13) A wind direction and speed recorder has been mounted near the emergency panel to inform the operator of these atmospheric conditions. This recorder whose anemometer sensor is located atop a 20 foot tower on the roof of the reactor building would be used by the operator to determine where to monitor for released contaminates. Transparent overlays of possible dis-persion and fanning patterns of released activity would be placed over a map of the surrounding areas to determine the best monitoring locations. 5-8

(14) In addition to these instruments and controls both outside telephone communications and loud speaker communications with the reactor and office buildings have been installed. 1 I r I

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E I QUESTION 6. Provide a complete listing of those tests which must be completed prior to initial fuel loading.

RESPONSE

The following tests of systems and equipment are considered to be essential and must be completed prior to the initial fuel loading of the reactor. a. Flushing, hydrostatic testing, and functional operation of all primary and auxiliary process syn.tems utilizing light water and 'I helium. b. Functional test of the confinement closure system and reactor building leak rate test to include various internal and external applied pressures to the building and operation of the emergency exhaust system under varying modes of " induced leak paths". c. Ventilation system tests to include: (1) Determination of minimum and maximum flow rates. (2) Hold-up volume delay times. (3) "In-Place" D0P testing of all absolute filters. (4) Operability of building closure devices, doors, dampers, 1 valves, etc. d. Complete operational tests of fuel transfer system. e. Instrumentation tests to include: (1) Pre-operation tests, calibration, and establishing set points and alarms. (2) Response times for all scrams and rundown modes. (3) Calibration of nuclear instrumentation using an installed neutron source. (4) Functional operational tests of all instrumentation operating 1 on a light water-helium system. (5) Recalibration of all D 0 systems level instrumentation for 2 use with D 0. 2 f. Operational tests of the emergency power system by loss of power simulation. 6-1

I g. Operational tests of the void shutdown system to include response times, flow rates, and reliability utilizing both helium and air. h. Operational performance of liquid waste system to include: (1) Hydrostatic test of all embedded piping. (2) Calibration of liquid effluent monitoring system. (3) Monitoring system response observation. (4) Fail safe o'peration of diversion valves. I 1. Calibration of all installed radiation monitors.

j. Emergency D 0 cooling system performance tests to observe I

distribution, drain times, and flow rates under both modes of 2 injection, to the plena and to the inner reserve tank, also sump pump flow rates. k. Shim arm and regulating rod performance tests to measure withdrawal and drop times and speeds for a minimum of 50 cycles per unit. 1. Evaluation of the secondary water treatment program through I visual inspection of the shell side of the primary and purification heat exchangers. Functional tests of loss of compressed air systems in progressive m. steps first to simulate loss of the 150 psig system, the 90 psig system, and then the 10 psig system. Data to include how all essential equipment failed with each successive step. Draining, drying, and 4 psig helium leak test of all D 0 systems. n. 2 Filling of D 0 systems with heavy water and installation of o. 2 neutron source. I p. Functional retest of all process systems using D 0 ""d h 1i""* 2 q. Flow coast-down measurements on primary coolant system under pump loss of power with and without shutdown pump action, pneumatic valve closure, and motor driven valve closure conditions. r. Moderator dump times. s. Functional test of shutter control mechanisms. t. Operational test of pneumatic irradiation system. Alignment of reactor automatic control system utilizing a reactor u. simulator. I 6-2

I QUESTION 7. Provide a description of the test program to be conducted following initial fuel loading, including low power physics tests and assent to power tests, to assure that reactor parameters are in agreement with calculated values. Include: (a) description of the tests which will be conducted; (b) the information to be obtained; (c) the acceptance limits of important paraceters.

RESPONSE

1. INTRODUCTION The initial tests and operating procedures are described in Section 11 of NBSR-9. In this response to question 7, those tests and measurements which determine the true values of the reactor parameters will be described in more detail and the acceptance icvels which the measured parameters must meet will be established. 2. INITIAL FUEL LOADING 2.1 Procedure. The initial fuel loading and approach to criticality will follow the procedure outlined in Section 11.2 of NBSR-9. Records will be kept of the precise location of each element at all times. 2.1.1 Pre-loading. The initial neutron count rate due to the source (C ) will be determined with no fuel in the reactor 9 and with all shim arms and the regulating rod fully withdrawn. In preparation for fuel insertion, two shim arms will be fully in-serted. 2.1.2 Loading. One hundred and seventy gram fuel elements will be loaded in steps of 6, 3, 3, 2, and thereafter 1 element at a time until criticality is reached (calculated to be just over 16 elements). The two inserted shim arms are withdrawn af ter each step loading and the corresponding neutron count rate, C, determined. The two shim arms will be reinserted before cach step loading. 2.1.3 Determination of Criticality. After each step, the ratio C,/C will be plotted against the number of fuel elements I 7-1 I

loaded and an extrapolation made to estimate the number of elements required for a critical mass. At no step will the number of elements added exceed one half the number of additional elements estimated to be necessary to reach criticality with the exception, of course, of the addition of singic elements. 2.2 Limits. During low power tests and the initial approach to full power there will be no more than three 170g elements loaded above the critical loading. This is the additional' loading required to compensate for fission product buildup and 3 week fuel burnup at full power. In no case, shall the loading during this period I exceed 24 elements. The value of 24 170g elements is based on the maximum estimated uncertainties in the calculation of the critical mass required to reach 10 Mw operation. If 24 elements should be insufficient, hidden sinks for neutron absorption would be indicated. A careful analysis of all core components for neutron adsorption would be undertaken. If this did not reveal the explanation for the large mass requirement, a more detailed calculational program would be undertaken to explain it. Only af ter a satisfactory explanation had been found for the deviation from the original I calculation, would the approach to full power be undertaken. 3. REGULATING ROD CALIBRATION The regulating rod calibration curve will be determined each time a new core configuration is introduced. This would include the critical core and the more heavily loaded one (calculated to be 19 elements) which is required to go to full power. 3.1 Method 3 1.1 Positive Period Method. This method will be used to determine the absolute value of the reactivity worth of sections of the rod. 3.1.2 subcritical Multiplication Method. This will be used to determine the detailed shape of the regulation rod curve from relative reactivity values. 3.1.3 Regulation Rod Curve. The regulation rod curve determined by the above methWs will include the entire range of motion of the 7-2

I rod. The curve will be used to determine the most desirable I settings for the normal upper and lower limits of motion. These normal limits will be about 15" of travel. 3.2 Acceptance Limits 3.2.1 Minimum Acceptable Worth. The minimum value for the regulation rod worth is 0.5% 6p. A lower value would not allow sufficient flexibility for manipulation of experiments. 3.2.2 Maximum Acceptable Worth. The maximum allowed value is 0.87. Ap (S for the NBSR). Since the regulation rod is automatically W controlled and is adjusted much more frequently than the shim arms, it is desirable to limit its value to less than S, so that novement of the regulation rod cannot make the reactor prompt critical. 323 Alternate Action. If these conditions aren't met, the original rod will be replaced by a new rod of modified design, as required, before zero power tests are completed. The new rod will, of course, be calibrated, and shown to meet the above limits. 4. SHIM ARM CALIBRATION Calibration curves will be determined for selected individual shim arms as well as for the bank of four. The curves will be determined for the minimum critical core. The calibration of the bank of four will be checked for the core configuration required for the first approach to power and for the equilibrium core. The calibration of a single shim arm will be done with the other three fully withdrawn. Thus, a single arm would control all the excess reactivity in the core. This is small for the just critical core, but similar measurements for more heavily loaded cores may I be neither necessary nor desirable. 4.1 Method. The subcritical multiplication method will be used to determine the shape of the shim arm calibration curves. The absolute normalization of the curves will be determined by: (1) Comparison to regulating rod over top portion to normalize scale. I (2) Positive period method to determine scale of top portion. 7-3

I 4.2 Acceptance Limits 4.2.1 Lower Limit. Any two (2) shim arms with the others I fully withdrawn must be capable of maintaining the most highly reactive core allowed (127. available reactivity) subcritical by

47. Ap. This would be sufficient to maintain the reactor suberitical by more than the regulating rod worth should another element be inadvertently added.

If this condition is not met, the shim arms must be replaced by ones of a modified design, or additional control rods of a different design must be added. 4 2.2 Upper Limit. The maximum limit on the worth of the four banked arms is determined by the maximum reactivity insertion The "startup accident" based on a maximum reactivity insertion rate. -4 rate of 7 x 10 op per second has been analyzed in Section 13.2 of NBSR-9 and shown not to disrupt the core. The maximum shim arm withdrawal rate of.04 degrees per second combined with this reactivity insertion rate leads to the maximum allowable slope for the shim I arm curve of 1.757 op per degree. Comparison of this value with the calcula ted value in Section 4.6.6 of NBSR-9 of 1.57. Ap/deg. indicates that the maximum allowable value for the total worth of the shim arms would be 547. Ap. If the value of 1.75% for the maximum slope of the shim arm curve were exceeded, a new evaluation of the startup accident might prove the higher rate to be acceptable. If not, the mechanical withdrawal rate would be decreased. 5. VOID COEFFICIENT The void coefficient will be determined for the just critical Additional measurements will be made for the first power core core. (19 element core), and later for the equilibrium core to confirm the predicted values for these cores. 5.1 Fuel Element Void Coefficient. 5.1.1 Method. A fuel element will be prepared such that it is voided in the region of the fuel plates. The change in reactivity caused by replacing a normal element with the voided one will be measured for each fuel element position except that only one ^^ I

1 measurement is required for symmetrically identical fuel element j positions. These measurements, combined with void coefficient measurement made in the gap region external to the fuel element will be used to determine the average void coefficient in the fueled I regions of the elements. 5.1.2 Acceptance Limits. The minimum acceptable limit for the void coefficient in the fuel is.0457. Ap per liter. A correlation with SPERT II data, similar to that discussed in response to Question I 1.6, showed that for a 27. reactivity insertion, -the maximum surface temperature of a fuel element plate would reach the melting point of aluminum if the fuel void coefficient should be as low as.0457. ap/ liter. The calculated value of.047./ liter given in Table 4.6-1 of NBSR-9 was the average value for an homogenized core. A comparison of the NBSR with other similar reactors (MITR, CP-5, Georgia Tech, and SPERT) for which measured void coef ficients in the fuel elements are available indicates that the most probable value for the average void coefficient within the NBSR fuel elements is 0.17. Ap/ liter. If the measured value should be less than.0457. op/ liter, a I re-evaluation of the correlation with SPERT data would be undertaken and different core configurations would be investigated to increase I the void coefficient. 5.2 Experimental Position Void Coef ficients. Measured reactivity worths against the regulating rod curve, will be made for various size installed voids (up to 1 liter) in the several experimental positions and blank fuel element positions. The void effect of several voided thimbles similar to that described in Section 11.3 of NBSR-9 will be measured in the central position. These will vary in diameter from about 1" up to the thimble shown in Figure 11.1 of NBSR-9. The purpose of these measurements is to determine the I effect of neutron streaming on the void worth as a function of tube diameter. The void worth of the largest thimble is expected to be worth less than 5 As stated in NBSR-9, these axial thimbles are not pemanent and will be removed before the end of the low power testing program. I

The only limit on the void coefficients in these positions since they don't enter into the excursion shutdown mechanism is that they be negative. The negative requirement is determined by the use of the void shutdown system. No mechanisms are I available to generate positive void coefficients for well thermalized D 0 reactors such as the NBSR, so failure to meet 2 this requirement is considered very remote. A positive void coefficient would not allow the use of the void shutdown system. 6. TEMPERATURE COEFFICIENT 6.1 Methods 6.1.1 With Pump Heat ( < 100 Wa t t s ). e (1) Measure the reactivity worth, against the Regulating Ro d cu rve, of increasing the temperature of the moderator / coolant by adding heat by pump heat and maintaining the reactor power level by adjusting the Regulating Rod. (2) From this determine the Temperature Coefficient of the reactivity for the core. 6.1.2 With Nuc1 car Heat (at 1 KW and above). (1) Measure the reactivity worth, against the Regulating Rod curve, of increasing the temperature of the moderator / coolant by allowing the nuclear I heat to raise the water temperature. Maintain the reactor power 1cvel by adjusting the Regulating Rod. (2) From this determine the Temperature Coefficient of reactivity for the core over a temperature rise of 80 - 120 F. 6.2 Acceptance Limits. The requirement that any moderator temperature increase must not result in a reactivity increase, requires the temperature coef ficient to be negative. 7. UPPER REFLECTOR REACTIVITY WORTH 7.1 Measurements to be Made on Minimum Critical Core I 7e

7.1.1 Top Reflector Borch Curve. With the banked shim arms set for criticality, the water level will be lowered in I steps and the change in reactivity measured for each step. Since the shim arms will be almost fully withdrawn for the minimum critical core, the resulting reflector worth curve should be a good approximation to the unperturbed curve. 7.1.2 Top Reflector Worth as a Function of Shim Am Position. The top reflector will be dumped for each of two or three shim arm bank positions (each position subcritical), and the reactivity change determined as a function of shim ann position. 7.2 Methods. The top reflector worth curve will be determined over the top portion of the regulating rod. The remainder of the curve will be determined by the subcritical multiplication method. The same factors that were used to normalize the shim arm subcritical multiplication curve can be used here. This method will also be applied to the measurements described in Section 7.1.2 above. 7.3 Measurements on other Cores. The top reflector worth will be determined for the 19 element core and for the equilibrium I core (~ 24 elements) for a single shim arm bank position. The position will be the critical position for each core at the beginning of a cycle when the top reflector would have its minimum worth because of the greater shim arm insertion. 7.4 Acceptance Limits. It is anticipated that the top re-flector will in all cases be worth at least the 4*/. Ap which has been previously sighted as a desirable shutdown margin. Any shutdown margin, however, would be acceptable since the top reflector dump is only an emergency backup to the shim arm system. 8. IlELIUM VOID SilUTIX)WN TEST 8.1 Method. With the minimum critical core operating at less than 100 watts and with the automatic dump and scram bypassed, the power decay rate af ter the initiation of the helium void will be measured. The rate of negative reactivity insertion and the reactivity worth will be determined from the decay curve and the steady shutdown counting rate. 7-7

82 Acceptance Limita. Tha minimum acc:ptable vclu2 for tha n:g:tiv: reactivity inserted by the helium void shutdown is 27. Ap. This is the value of the maximum excess reactivity considered in the maximum excursion accident. The void shutdown system is designed to give about 87. ap since it is substituting briefly for the top reflector dump which is of the order of 87.. If possible, the helium flow will be adjusted to give about this value, but, in any case, it will be adjusted to give at least -27. op. 9. EXPERIMENTAL POSITION REACTIVITY WORTHS

9.1 Methods

9.1.1 Beam Tubes and Grazing Tubes. (a) Measure the reactivity worth, against the Regulating Rod curve, of these tubes, when flooded with D 0 2 and when flooded with H 0. 2 (b) Measure the reactivity worth, against the Regulating Rod curve, of various sized installed " black" and " gray" poisons in these tubes. 9.1.2 Central Thimble, 3-1/2" Thimbles, 2-1/2" Thimbles, Dummy Fuel Element Positions and V-1 through V-7 Thimbles. Measure the reactivity worth, against the Regulating Rod curve, of various sized installed " black" and " gray" poisons in these thimbles. 9.1.3 Cryogenic Port. (a) Measure the void reactivity worth of the cryogenic I port by draining the D 0 from the water tank 2 described in Section 8.1.4 of NBSR-9. (b) Measure the reactivity worth, against the Regulating Rod curve, of various sized " black" and " gray" poisons installed in the 6" beam tubes both with and without the water tank drained. 9.2 Acceptance Limits. With the exception of the cryogenic port which is discussed more fully in response to Question 22, the maximum reactivity that a beam tube or grazing tube should have is. 57.. Should this limit be exceeded, either the beam tube volume will be decreased by the insertion of solid material, I or a second barrier to flooding may be provided in the form of an inserted water tight tank. 7-8

I 10. FLUX MAPPING 10.1 Measurements. The follow measurement.s will be made for the 19 element core. As many as necessary will be repeated for the equilibrium core to confirm predictions based on the I earlier detailed measurements. The flux measurements will be made by scanning activated wires and foils The neutron densities actually measured will be converted to flux determinations by assuming the average I neutron temperature equals the moderator temperature. In most cases, gold, cobalt, or indf um wire or foils will be used and in some cases they will be compared to cadmium covered wire or foils. (a) The vertical flux shapes and absolute values will be determined for each experimental position in the core and reflector by scanning activated wires. The facilities are the 2-1/2" thimbles and 3-1/2" thimbles in the core, and the V-1 through V-7 thimbles I in the reflector. (b) Activated foils will be used to determine the flux in each beam tube. (c) Wires will be activated in several channels of several selected fuci elements to determine the neutron flux spatial distribution within the elements. 10.2 Acceptance Limits. The flux distribution within the fuel elements will be compared to the calculations of the vertical flux distribution which were used to determine the hot spot factor. The calculated peak-to-average value is 1.67 as given in Table 4.6-3 in Section 4.6 of NBSR 9. A 157. allowance I for the uncertainty of the power distribution cal (ulation was included in the hot spot factors calculated in Section 4.7 of NBSR 9. Therefore, the measured peak-to-average flux ratios must not exceed 1.92. If this value should be exceeded, new heat transfer calculations would be made to detemine new minimum flow requirenents. If necessary, a third primary pump currently planned as a standby could be used to achieve the required flow. I

Th rediel flux distribution will be comparad to the radial power distribution measurements discussed below. 11. RADIAL POWER DISTRIBUTION FEASUREMENTS 11.1 Method. After operating at 1 KW of power, the radial power I distribution in the startup core (~ 19 elements) will be determined by gamma scanning the fuel elements. This will be done by placing a detector in a thimble in a fixed location in the top reflector. Then each element will be raised in turn and placed in a fixed location relative to the detector and counted. Since the fission product inventory is directly proportional to the power in each fuel element, the comparison of the several fuel element activities will determine the relative power distribution in the core. If the background activity in the region above the core should prove too great, these measurements could be done with I a detector in the sub-pile room outside the transfer chute and each element placed in the chute for counting. I If this technique proves practical and versatile enough, it is anticipated that it will be applied to the equilibrium I Core. 11.2 Acceptance Limits. The relative power distribution I between the elements fed by the inner and outer plena will be used to determine the optimum relative coolant flow through each plenum and the flow will be adjusted accordingly. 12. APPROACil TO FULL POWER In addition to the tests and measurements described above, the following will be performed at the appropriate steps in the approach I to full power. 12.1 Evaluate Shielding. Starting at the 100 KW 1evel, the I radiation levels around the biological shield, top floor and, where possible at the lower levels, the process room will be measured. Gamma fields and both fast and slow neutron fields will be detennined. The results will be used to predict the fields resulting from the next power step. I

The radiation fields at each step must not exceed tho:e in 10 CFR 20 and the predicted fields for the next step must also conform to 10 CFR 20 regulations. Wherever these conditions are not met, steps will be taken to improve the shielding as required. During the lower power runs, selected beam port plugs will be removed to measure the efficiency of the beam port shutters. These measurements will be done under the close surveillance I of a health physicist. They will be started at the lowest power that gives measurabic results, and will be terminated as soon as significant results are obtained. If required, temporary I shielding in the vicinity of the unplugged beam port will be provided. 12.2 Evaluate Temperatures in the Thermal and Biological Shields. At each power step, the thermocouples embedded in the themal and biological shields will be read. The results will be used to predict the temperatures to be expected at the next power step. The maximum acceptable temperature differential between any two points within the biological shield is 50 F. If this should be exceeded, a more detailed analysis of the structural properties of the NBSR's re-enforced concrete shield would be undertaken to determine the maximum temperature differential I consistent with structural soundness. If the result of this detailed investigation showed the measured AT to be too large, reactor power would have to be limited accordingly. The anticipated maximum temperatures in the thermal shield is 150 F in the lead and 145 F in the steel. The maximum thermal I shield temperatures must not exceed the melting point of lead (622 F). If the temperatures are significantly above the anticipated ones, the flow through the thernal shield will be increased and adjustments in the coolant distribution would be made. 12.3 Fuel Element Fission Product Decay IIcat. The procedure for measuring the temperature reached by a fuel element during transfer and detemining the minimum pemissible decay time before transfer are discussed in the answer to Question 8. 7-11 l

W 12 4 Powir Channel Calibration. While operating at a fix::d power as determined by the thermal power recorder physically adjust (move) the power range chambers so each channel reads in agreement with the thermal power recorder within + 5*/.. Care must be taken while adjusting the channels that the comparator circuits do not trip and alarm. The thermal power recorder reading will be checked against the power calculated directly from the measurements of flow and coolant temperature rise. 12.5 Xe and Sm Reactivity Measurements. The reactivity ef fect of Xe and Sm will be measured as a function of time af ter startup against the shim arm calibration curve. The Xe and Sm buildup will also be measured as a function of time af ter shutdown. These data will be useful in timing startup and estimating available reactivity at the beginning of each cycle. The reactivity worth of the poison buildup I can be measured by dropping the reactor power to a low level. As the poison builds up, the regulating rod and shim arms are adjusted to compensate and the reactivity change determined from the shim arm calibration curve. 13.

SUMMARY

The tests and measurements discussed above are listed below I and are correlated with core configuration and power level in Table 13.1. As pointed out in the text, certain of the measurements and calibrations shown as being repeated at different stages will not be complete duplicates of the original procedure. Instead they will be used to confirm the predictions obtained from the original, more detailed procedures. I (1) Regulating rod calibration. (2) Shim arm calibration. (3) Void coefficient measurement. I (4) Temperature coef ficient measurement. (5) Upper reflector reactivity measurement. (6) llellum void shutdown test. (7) Experimental position reactivity worths. (8) Flux mapping. 7-12

I (9) Power distribution measurements. (10) Shielding evaluation. (11) Temperature measurements in biological and thermal shields. (12) Fuel element fission product decay heat. I (13) Power channel calibration. (14) Xe and Sm reactivity measurement. I I I I ' I I I I I I I I I I

M M M M M M M M M M M M M M M M M M Table 13.I SCHEDULE OF TESTS AND MEASURDfENTS Test and Min. Critical 19 element core 24 element Measurement Core < 100 W < 100 W 1 Kw 10 Kw 100 Kw 1 Mw 5 Mw 10 Mw core 1 X X X 2 X X X 3 X X X 4 X X X X 5 X X X 6 X X X y E 7 X X 8 X X 9 X X X 10 X X X X X 11 X X X X X X 12 X X X X 13 X X X X 14 X X X l l l l

I QUESTION 8. Provide a description of the test program to be conducted following initial fuel irradiation to establish: a. The cooling requirements of a fuel element in the fuel trans fer chute ; b. The decay heat level at which uncooled fuel element under any handling conditions will not reach unacceptable I tempe ra tures.

RESPONSE

The test program to develop the cooling requirements of a fuel element in the fuel transfer chute and to determine the decay heat level at which an uncooled element will not reach unacceptable tem-peratures will consist of thermocouple temperature measurements I made at increasing fuel element decay heat levels from 500 watts to 2000 watts. Each series of measurements will be made first in the upper reflector region above the core and then, if satisfactory results are obtained, the measurements will be repeated in the fuel trans fer chute. Approximately one-half hour af ter shutdown f rom 1 Megawatt operation, the hottest element in the NBSR core will produce about 500 watts. This 500 watt element will be withdrawn from the core into the upper reflector region of the reactor vessel and the tem-perature rise and equilibrium temperature measured and recorded. Upon completion of the measurements in the refIcctor region, the I element will be confined in the fuel transfer chute and the tem-perature measurements wilI be repeated. Prediction calculations for the equilibrium temperature of a 1000 watt element will then be made using the 500 watt test results I as a guide. A 1000 watt element will be obtained by selecting the hottest I element in the NBSR core approximately eleven hours af ter shutdown f rom 5 Megawatt operation. The same basic test procedure discussed 1 8-1 3

I above will be employed to measure temperature rise and equilibrium temperatures first in the reficctor region and then in the confines of the fuel transfer chute. Once again the predicted and measured results will be correlated. The test results obtained at the element power level of 1000 watts will be confirmed by selecting the hottest element approximately sixty-six hours after shutdown from 10 Megawatt operation. This element will again produce about 1000 watts. The tests made after 5 Megawatt operation will be repeated to accomplish the desired confirmation. At the next shutdown from 10 Megawatt operation a delay of I approximately twenty-two hours will result in an element which will produce about 1500 watts. Once again the tunperature measurement series will be conducted and the results correlated with predictions based on heat transfer calculations. The final test series will be scheduled at the third normal shutdown from 10 Megawatt operation. At this time, an eight hour delay before removing the element will result in a 2000 watt element. Temperature measurements and successful correlation with calculated values will complete the program. Throughout the temperature measurement program, limiting values of maximum temperature and temperature rise will be enforced. These limits have been chosen to provide a safe margin well below the temperatures known to cause structural damage to the materials in the NIISR fuci element. The limits are these: a. The steady state temperature above the reactor must not exceed 700'F. b. The steady state temperature in the fuel transfer chute must not exceed 800 F. c. The rate of temperature rise in the fuel transfer chute I must not exceed an average of 5 F per minute from 700*F to 800 F, i.e., 50 F in ten minutes. 8-2

I QUESTION 9. Provide a list of all operating and emergency procedures for the facility. Provide a summary description of the contents of all emergency procedures. Also include a discussion of the provisions for assuring that local authorities are, and will remain, properly appraised of actions which could be required of them in the event of fire or accident to the facility.

RESPONSE

The NBSR Opera ting Manual consists of eleven volumes designed to describe all aspects involved in the normal and emergency operation of the NBS reactor. Volume III of this manual, " Annunciator and Emergency Procedures" describes the planned course of action to initiate in the event of abnormal or emergency conditions. Section 1.0 of these procedures governs the evacua tion of the I building, site, and surrounding area, should it be required. Medical and fire emergency plans are also included. The following is a list of these procedures: 1.0 Emergency and Evacua tion Procedures 1.1 Nuclear and Radioactive 1.1.1 Local Area Evacuation - EI 1.1.1 1.1.2 Building Evacua tion - EI 1.1.2 1.1.3 Site and Surrounding Area Evacuation - EI 1.1.3 1.2 Medical I 1.2.1 Non-Radioactive Injury - EI 1.2.1 1.2.2 Radicactive Injury - EI 1.2.2 1.3 Fire 1.3.1 Non-Radioactive Ma terials - EI 1.3.1 1.3.2 Radioactive Ma terials - EI 1.3.2 Those procedures in Section 1.1 govern progressively serious incidents ranging from a local spill or high radiation to the entire evacuation of the site and surrounding area. Local procedures call for evacuation to the nearest telephone outside the affected area when local radiation monitors alarm. I Af ter notification of the Shif t Supervisor and Health Physics personnel, all involved personnel will remain at the telephone until cleared by the Health Physica Section, g e-1

I Operations personnel will take immediate action to prevent or minimize further contamination and to protect personnel. If the reactor is in operation, and it is determined that the accident condition could affect continued operation, the reactor will be shutdown. Health Physics personnel will investigate and determine the extent of radiological hazard and recommend necessary emergency equipment and protective clothing. Follow-up action requires clean-up dose evaluation and reports of the incident to be filed in accordance with 10 CFR 20. Should accident conditions progress to the point where evacuation of the building is required, personnel in the laboratory-office building will be notified by an evacuation siren and the loud speaker system. All non-essential personnel not required to terminate or minimize the accident will proceed immediately to the Linac building where Health Physics personnel can evaluate all dose received and account for personnel. Operations personnel will secure the reactor and assure the reactor building is I scaled, then proceed to the underground emergency ventilation control panel. Proper operation of the emergency exhaust system will be confirmed and the NBS Cuard Force instructed to post road barricades at both entraces to the reactor facility ground. The Shif t Supervisor, being advised by IIcalth Physics personnel, will evaluate the hazard and decide if conditions warrant more extensive measures to protect the health and safety of the surrounding populous. He will also notify higher icvel management and summon additional aid if required. This decision will be based on analysis of air samples from portable monitoring equipment, and direct surveys of areas for which high concentrations are indicated. Should concentrations exceed those in 10 CFR 20 when averaged I over a year or should direct radiation levels exceed those in 10 GTR 20 for an unrestricted area, immediate evacuation is required. The Of fice of Safety and Civil Defense for NBS would be called upon to direct the evacuation of Bureau personnel from other buildings, the Montgomery I County Police and Maryland State Highway Patrol would be notified to regulate traffic on county and state highways passing the site, and should evacuation e-2 g

I I of area residents be required, the Montgomery County Civil Defense would be notified. Section 1.2 of these emergency procedurea describes the action to be taken in the event of personnel injuries involving either radioactivity or non-radioactive materials. Reactor Operations personnel have attended a Bureau sponsored first aid course and would be available tc administer immediate first aid to any injured person, while outside aid was arriving. NBS maintains a medical of fice during normal working hours staf fed by a full-time doctor and registered nurse. These people would be summened during normal hours. Should ambulance service be required, the Montgomery County Rescue Squad would be Hospitalization services for potentially conta ninated personnel have I notified. been assured, utilizing facilities a t the Cancer Institute of the National Ins titutes of Health in Bethesda, Marylana. Section 1.3, dealing with fire emergencies, describes the proper manner of coordinating the Montgomery Cot.nty Fire Departments with NBS Fire Of ficial and reactor personnel. The reactor and laboratory-office sections of the building all contain both manual pull-bor.es and automatic heat detection equip-ment to notify personnel of a fire. These devicca scund an evacuation bell throughout the building as well as sound a remote fire alarm a t the NBS Central Guard Office. The alarm bell will bring immedia te evacuation of ncn-essential I personnel f rom the building. Operations and Health Physics personnel will investigate and determine consequences of the fire. A check of the alarm panel in the front corridor of the laboratory-office building will pin-point the area of the alarm and indica te if radioactive materials may be involved. Upon receipt of the alarm, the Central Guard Office will notify the Montgomery County Fire Control Center, with immedia te response by local engine companies. The fire departments will respond to the front entrance of the laboratory-office building. They will not enter the building uncil directed by the Shift Supervisor on duty whose judgement will be based on Health Physics recommendations. I If radioactive ma terials could be involved, respira tory equipment and protective clothing will be utilized. Health Physics will control exposure of all personnel I through measurement of radiation levels and limiting working times. I 9-3 I

I r I These procedures have been coordina ted with local officials of fire, I medical, and civil defense agencies, and the Administrative Cervices Division 4 of the Na tional Bureau of S tandards. Medical emergency plans have been co-ordinated through the Office of Radiation Safety for NBS. Area fire department of ficials and firemen have been invited and have toured the reactor facility to familiarize them with the building. These activities have been coordinated through the services of the NES Fire Chief and his staff, a part of the Adminis- ) tra tive Services Division. Copies of these procedures have been given to each organization which would be called upon. Each procedure has the approving official's signature and the date of issuance. Any revision to these procedures will bear the i revision date and approval with revised copies sent to all involved organizaticas. Volume II of the NBSR Operating Manual describes all phases of normal system and equipment s tartup, opera tion and shutdown. Those procedures in Section 1.0 are " Master Procedures", tha t is, they are a composite of all procedures required to startup operation or shutdown the reactor. Procedures in subsequent sections govern individual systems. Listed below are the titles of these operating procedures: 1.0 Master Procedures 1.1 Reactor Startup (0.I.-l.1) I 1.2 Reactor Normal Opera tion (0.I.-l.2) 1.3 Reactor Shutdown (0.I.1.3) 2.0 Heavy Wa ter Systems 2.1 Primary Cooling System (0.I.-2.1) 2.2 Purifica tion Sys tem (0.I.-2,2) 2.3 Emergency Cooling System (O I.-2.3) 2.4 Thermal Column Tank Cooling System (0.I.-2.4) 2.5 Experimental Cooling (0.I.-2.5) 3.0 Light Wa ter Systems I 3.1 Secondary Cooling System (0.I.-3.1) 3.2 Thennal Shield Cooling Sys tem (0.I.-3.2) 3.3 Storage Pool Cooling System (0.I.-3.3) I l I l 9-4 i

I 3.4 Secondary Water Experimental Cooling Sys tems (0.I.-3.4) 3.5 Demineralized Water Experimental Cooling Systems (0.I.-3.5) 3.6 Water Treatment System (0.I.-3.6) 3.7 Chilled Water System (0.I.-3.7) 3.8 Domes tic Wa ter (0.I.-3.8) I 3.9 Steam and Condensate (0.I.-3.9) 3.10 Hot Wa ter Hea ting (0.I.-3.10) 3.11 Fire Protection (0.I.-3.11) 3.12 Ho t Wa s te (0.I.-3.12) 4.0 Gasous Systems 4.1 Helium Sweep Gas Systems (0.I.-4.1) 4.2 Void Shutdown System (0.I.-4.2) 4.3 Irradia ted Air (0.I.-4.3) 4.4 CO Sys tem (0.I.-4.4) 2 4.5 Pneuma tic Irradia tion Sys tem (0.I.-4.5) 4.6 Compressed Instrument and Emergency Air Systems (0.I. -4. 6) I 4.7 Laboratory Gas Systems (0.I.-4.7) 4.8 Reactor Building Ventila tion Systems (0.I.-4.8) 4.9 Reactor Building Emergency Ventilation Systems (0.I.-4.9) 4.10 Reactor Building Leak Rate Test System (0.I.-4.10) 4.11 Laboratory and Office Building Ventilation System (0.I.-4.11) 5.0 Electrical and Instrument Systems 5.1 Electrical Dis tribution (0.I.-5.1) 5.2 Diesel Genera tor (0.I.-5.2) 5.3 Radia tion Monitoring (0.I.-5.3) I 5.4 Nuclear Ins trumenta tion Sys tem (0.I.-5.4) 5.5 Leak Detection System (0.I.-5.5) 5.6 Shutter Lif t Control System (0.I.-5.6) 1 I I I I l l 9-5

I QUESTION 10 What information is available to the operator to guide his judgment on the location of a primary system pipe break so that he knows what I point of emergency cooling water injection to select? What are the possible consequences of his making a misjudgment? I l

RESPONSE

The emergency cooling system for the NBSR is described in Section 7.1.1 of NBSR-9. If a break in the primary system should cause the water to drain from the vessel, the inner emergency I cooling tank will direct a stream of water into the top of each element. As described in NBSR-9, this is a direct physical response to low water level and is independent of operator action or location of break. Even if no action at all is taken by the operator, the inner emergency tank will supply cooling for 25 minutes. Thus, ample time is provided for a considered decision to be made by the operator. The operator has two choices: he may use the D 0 in the 3000 2 gal main emergency cooling tank to replenish the inner tank and maintain the flow into the top of the element; or he may direct the flow from it into the two plena below the fuel elements. The latter procedure is preferred when possible because the full pressure due to the elevation of the main emergency tank is available to force I water up through the elements assuring that all fuel plates are immersed in water. The only condition that would cause the operator to use the first procedure (top feed) instead of the seccad pro-cedure (plena feed) is a break in the piping between check valves DWV-114 and DWV-118 (see Figure 7.1 of NBSR-9) and the plena. This section of piping is completely within the sub-pile room and of minimum length and extra strength to minimize the possibility of the break occurring here. Thus the normal procedure would be for the operator to initiate emergency flow into the plena, and then determine whether the break was in the plena region which would require a switch to the top feed I procedure. If the break were in the plena feed system, the water 10-1

I would drsin out the break and not cool the elements. Since, initially, there is always top feed from the inner emergency tank, the failure of the plena feed does not endanger the elements. As described below the integrity of the plena feed system can be de-I termined quite easily. Thus, the failure of the plena feed system could be determined well within the grace period provided by the inner emergency tank. During this short period of time, the only consequence of having initiated flow into a broken plena system would be the loss of less than 107,and probably no more than 27. of the main emergency coolant supply. The integrity of the plena feed system can readily be determined from the pressure in the plena. If the system is working properly, the plena and the fuel elements are full of water causing the pressure at the emergency cooling inlet to the plena to read about 7 psig. If the break were in the plena piping, the water would, of course, drain I out and the pressure would be 0 psig. Thus, the emergency procedures will specify the range on the instrument sensing the plena pressure which indicates proper operation of the plena feed system. If the instrument should read outside the specified range, the operator switches to the top feed procedure. The final determination of the specified operational range will be made when D 0 is in the system. 2 Thus, in addition to the information provided by normal level instrumentation and leak detectors, the operator is provided with a simple, straightforward criterian on which to base his decision. I I I I 10-2

I QUESTION 11. Describe the provisions for assuring proper protection of experi-menters and visitors to the facility. Include information con-I cerning orientation lectures, access restrictions, and evacuation plans. Discuss the equipment and procedures provided to minimize I the potential for exposure from a scattered or unattenuated beam or from experiments. Calculate the maximum dose rate which could result from direct exposure to an unattenuated beam. 1. INTRODUCTION Experimenters utilizing NBSR facilities are considered in two groups: those permanently assigned to the reactor facility and those " guest workers" or transient personnel. The perma-I nently assigned experimenters, as they report, will be issued Volume VII of the NBSR Operations Manual, " Health Physics Procedures" and Section 1.0 of Volume III, " Annunciator and Emergency Procedures". These procedures describe all aspects of the precautions necessary to maintain an adequate NBSR radiation protection program as well as a planned course of action to initiate in the event of emergencies. These pro-cedures with a brief orientation program conducted by the Health Physics staff are designed to protect all aspects of work practices and procedures. Guest workers and transient personnel will also be given I an orientation session to familiarize them with the facility. Since these personnel will be working closely with permanently assigned occupants, they will be guided and directed primarily I by these people. Visitors to the facility will be required to register with the receptionist and are provided with appropriate personnel monitoring equipment. All visitors are to be escorted by a member of the permanently assigned staff at all times they are 11-1

in the regulated zone or any radiation area in the building. Groups of visitors in excess of five people will be given a brief orientation session prior to entering the regulated zone of the facility as well as provided with one set of personnel monitoring equipment. 1.2 Open Beam Radiation Fields. The most intense beams are obtained when no type of soller slit collimators are used in the beam, and the solid angle subtended at the beam exit in-cludes the full 6" diameter source area at the end of the beam tube. The radiation fields at the exit of a beam tube under these conditions and for 10 Mw reactor operation are summarized in Table I. Table I. Radiation Fields at Beam Exit Source Intensity Dose 10 Thermal Neutrons 2 X 10 n/cm -sec .8 X10 r/hr 9 5 Fast Neutrons (E=0.1 Mev) 6 X 10 n/cm -sec 1.8 X10 5 Gamma Rays 6 mw/cm .6 X10 5 Total 3.2 X10 r/h Since dase is a meatare of energy density deposited the numbers in Table I do not change with beam size provided the whole 6" diameter source area remains unchanged. The total number of particles emitted is, of course, proportional to the area of the exiting beam. If beam holes were left com-pletely unplugged, the exiting beam would have an area of about 1 ft Some concept of what this would mean for such a com-pletely unshielded beam can be given by noting that if this beam were distributed over an area equal to the floor of the reactor building, the average dose would be about 40 r/hr. All of these doses are much too high to tolerate and the beams were never intended to be used in an unshielded manner. Typical uses and shielding arrangements are discussed below. I I 11-2

1.3 Neutron Beam Shielding. No beam will be ope'ned into the reactor without first giving careful consideration to the shield-ing. The most intense beam that is likely to be used will have radiation fields no greater than those in Table I and a cross I 2 sectional area no greater than 4 in. For the purpose of illustrating beam shielding, the most intense beam presently being planned will be discussed. The beam wiH be used in a neutron diffractometer facility and Because s'ller slits will be will be about 1" x 2" in size. o used to restrict the angular divergence of the beam to 1/2, the beam intensity will be 1/4 that given in Table 1. The beam will pass through a shielded tunnel about 2-1/2 feet I long and impinge on the monochromating crystal located in the center of a 5 foot shielding drum. The tunnel is designed so the beam does not hit the sides, but a small fraction of the beam is scattered. The tunnel is lined first with 1/4" boral followed by 1-1/2" of lead. This absorbs the scattered thermal neutrons and gamma rays. The lead is followed by about 1-1/2 feet of borated paraffin to moderate and absorb the scattered fast neutrons. The actual thickness of the paraffin is deter-mined more by neutron scattering from the drum cavity than from I the incident beam itself. In the drum itself, the incident beam is stopped by boral and several inches of lead to absorb the slow neutrons and gamma rays. The remainder of the 2 foot thick drum is masonite and an outer steel shell. The shield which has been approximately described here is a dupli-cate of one currently in use at HFBR at Brookhaven. The effective-ness of the shielding has been checked up to 30 Mw operating power at the HFBR. At this power the beam intensity at the Brookhaven facility is about four times that of the beam con-sidered above. Thus, the ability to adequately handle beams with intensities of the same order of magnitude of those available from the NBSR has been thoroughly developed and checked. I 11-3

1.4 Precautionary Procedures. Although the methods and shield-ing techniques necessary to handle intense radiation beams have been thoroughly developed, not all shields at the NBSR will be simple duplicates of those tested at other reactors. Each ex-perimental facility will have its own individual requirements I and individual shields will have to be designed. These designs will, of course, be based on the above mentioned developed techniques, but in addition a certain amount of judgement based on experimental and engineering experience is required. Thus, any new shield must be tested thoroughly before it is accepted as complete. The procedures to check experimental facilities is outlined below: 1. No unchecked changes in a facility shield which would significantly increase the radiation level I in the vicinity of the facility will be permitted while the reactor is at power without the express permission of the shift supervisor and must be accomplished under the surveillance of a health physicist. 2. All major shielding modifications and new in-stallation will be accomplished during reactor shutdown. 3. All facilities will be visually checked before the reactor startup to be sure no obvious breaches of shielding exist. 4. All facilities will be checked for radiation at low reactor power. Particular attention and check-ing will be given to new facilities. 5. Checks will continue to be made as the reactor comes up to power. 6. Before the direct control of beam tube shutters is given to an experimenter, its opening and closing must itave been checked and approved by Reactor Operations and Health Physics. 11-4

I QUESTION 12. Provide detailed information, including schematics if necessary, to assure compliance with the following criteria in the reactor safety system, building isolation system, and the emergency ventilation control systems: (a) No single component failure or circuit fault should result in loss of automatic protection against any reactor accident. (b) No single component failure or circuit fault can both initiate an accident and reduce the degree of protection. (c) Sufficient initial and periodic testing of systems and components must be conducted to assure that the proper operation of that component or system can be anticipated when it is required (initial in situ tests should include variations of voltage, temperature, and humidity over the maximum permissible limits of these parameters). 1. REACTOR SAFETY SYSTEM l.1 Nuclear Instrumentation Failure Analysis. The Nuclear Instrumentation System design employs redundant channels so that no single failure of a complete channel will result in loss of reactor protection. All the analog channels employ at least two channels as shown on Figure 9.8 of the NBSR 9. The safety system employs a redundant system consisting of 3 completely separate bi-stable trips feeding two separate and isolated logic cards, 07-62, 07-63. These logic cards in turn supply two separate isolated busses, A and B. The two isolated busses also feed the NOD Logic Card (NOR DIODE LOGIC, 07-65) which energizes a relay K07-14. (This relay is operated in a fail safe fashion and opens normally open (N.O.) contacts E-12-1

I connected in the relay safety system. Ref: DWG E-70 009 )* The relay K103 in turn opens contacts which de-energize the relays K2001 and interrupts the power supply (-10 v) to the clutch power supplies 07-53, 07-54, 07-55, and 07-56. The clutch power supplies (output units) consist of two transistor switches connected in series with the power supply to the inverter. The inverter transforms the -10 v DC to a 2 Kc square wave which is stepped up by the transformer TRI I to approximately 28 volt AC. The 28 volt AC is rectified, filtered and supplied to each clutch independently. (Ref: DWG D-70-004, Figure 9.11 of NBSR 9) Pairs of the output units as shown in Figure 9.11 of NBSR 9 are connected in parallel to share the output load of two clutches so that the failure of one clutch power supply will not de-energize the rod clutches. However the unsafe failure of one supply will not prevent de-energizing of the I other two rods which will safely shutdown the reactor. In addition the relay scrams K2000a, K2000b, K2001a, K200lb will de-energize the power input to the clutches (-10 v) and de-energize the -10 v supply to the failed pair in this case. In response to part (a) of the question if it is assumed as a worst case that one logic circuit fails unsafely, e.g. 07-62, a scram would be prevented from occurring on "A" Bus which in turn would not turn off the SW1 transistor switch in each output unit module. However, upon receipt of a scram signal by logic circuit 07-63 the SW secti ns of the output 2 unit would de-energize the inverter and turn off the clutch g power supplies. In addition the NOD logic circuit, NOD 10, W would receive the scram signal, 0 volts, and de-energize the 1 relay K07-14. A set of N.O. contacts of the relay K07-14 would then open and de-energize the relay K103 which would in turn de-energize the relays 2001a and b to de-energize

• The functional drawing equivalent to this is Figure 9.12 l

of NBSR 9. 12-2

the -10 v supply to the clutch power supplies. (Ref: D-70-004)* 1.2 Relay Safety System. The function of the relay safety system is shown on DWG E-70-009 sheet 2.** The relay scram system consists of the contact string K103 (circuit 4) and a duplicate set of scram contacts in the K2000 circuit (circuit 35). The final scram relays K2000a, K2000b, K2001a, and K200lb have contacts connected in series with the shim rod clutch power supplies as shown on DWG D-70-004.*** It may be observed in response to part (a) of the question that if I, any one set of contacts weld closed in a scram string the other string is still operable and will de-energize the clutch power supplies upon receipt of a scram signal. As a final backup, contacts of the manual scram push button, S2, are connected in series with the -10 volt supply to the clutch power supplies. If the K2000, K2001 relays fail to de-energize on a scram condition, and the electronic scrams have failed, then the clutches may be de-energized by the manual scram push button. I In response to part (b) of the question no single component malfunction in this system can initiate an accident and reduce the degree of protection. 1.3 Nuclear Instrumentation Testing. The nuclear instrumentation has undergone a 4 month testing program at the Leeds and Northrup (L&N) plant before acceptance. This program is documented and the results are available for inspection. The below listed tests were conducted. a. Calibration b. Response Times I Figure 9.11 of NBSR 9. See Footnote on Page 12.2

• • • Figure 9.11 of NBSR 9.

I 12-3 1 't

I c. Drift g E d. Elevated Temperature e. Component Failure f. Voltage 4 g. Frequency I 1.3.1 Calibration. All the analog channels and bi-stable f } trips (input units) were checked for calibration by employing signal generators to simulate the input conditions. In addition a neutron source was used with 200 f t of cable to check the source range channels for noise and ability to operate on statistical signals. 1.3.2 Response Times. Response times of the entire safety system were checked by introducing step changes in current into the input of the various channels. The overall time was measured from initiation of the signal to the clutch power supply until zero current in the clutch. 1.3.3 Drift. Four readings were taken over a minimum of 4 hours and the drift noted in calibration and setpoint. 1.3.4 Temperature. The entire nuclear cabinet was maintained at 40'C for a 4 hour period and permitted to cool to approximately 32 C. Checks were made for shift in cali-bration and any failures that may have occurred. No failures were noted and equipment functioned normally. Temperatures during this test wre measured at 8 selected points in the I cabinet. The highest temperature in the cabinet at the 40*C ambient temperature was 59.5 C. The components in the safety system attained a temperature of 40 C to 42 C. (The 59.5 C temperature was measured at the only vacuum tube equipment in the system, namely the General Electric pico ammeter) 1.3.5 Component Failures. Rather then simulate individual component failures, individual module failures in the safety system were simulated and the effect on the safety system noted. 9 12-4

t Each module was failed in two directions either by tripping or not tripping, i.e., "0" and "1" state. These effects were simulated in both operational modes of the safety system, l 1.e., 1 of 3 or 2 of 3 scram logic. During this testing regime the ability of the automatic monitor to detect the simulated failures was noted. 1.3.6 Voltage. The effect of varying the line voltage j upon the equipment from 107 to 127 volts was checked. The output of the system power supplies (-10 v and +10 v) were checked and did not change over this range. Calibrations of the analog channels and bi-stable trip setpoints were not affected. 1.3.7 Frequency. A variable speed motor generator set was used to vary the frequency from 57 to 60 cycles. The effects upon period indication, channel calibration, safety system and -10 v and +10 v power supplies were checked. These frequency variations did not affect operation of the system. 1.3.8 Operating Experience. The Nuclear Instrumentation I was placed in service April 1, 1965 and has been in almost continuous cperation since that date. During this period over 11,688 hours operation has been experienced by the equipment. Several failures have occurred during this period of time but none have been unsafe failures. Records have been kept of the failures and repairs made. 14N has made all repairs to date. 1.3.9 Pre-operational Testing. The pre-operational testing phase will duplicate the tests performed at the factory with the addition of testing with a neutron source. The procedures will be documented in the form of an instrument procedure. All the tests performed at the factory will be duplicated I with the exception of elevated temperature, voltage and r frequency tests. Prior to operation internal calibration signals permit a complete check to be made on all components in the system. 12-5

I One of the prerequisite conditions prior to startup is the completion of a check list which requires a calibration check of the analog channels and testing of all the bi-stable trips and their associated alarms and interlock functions. 1.3.10 Operational Testing. The nuclear instrumentation has been designed to incorporate internal calibrate and test i signals to verify the operation of any power range channel I during operation without causing or preventing a scram (2 of 3 coincidence only). A front panel switch on the nuclear panel permits the feeding of current calibration signals into NC-6, NC-7, or NC-8. Thus, in a 2 of 3 mode a test current may be put into i any one channel and the calibration and trip points of the channel may be checked. The setpoints of the bi-stable trips and operation of the associated annunciators may be checked at any time. The position of the calibrate operate switch is maintained by an annunciator AN 4-27. The relay K500 has contacts con-nected to the startup permit circuit to prevent reactor start-up if one of the switches is inadventently lef t in the calibrate ' position. All the nuclear scram circuits are continuously monitored during the automatic self testing mode of operation. The functioning of the "self-tester" is outlined in Section 9.5.2.3.1 of the NBSR 9. I i If the automatic monitor detects a fault a control room i annunciation occurs. The operator is warned that a fault has l been detected. The faulty module may be replaced in a matter I of a few minutes. Thus it will be possible to maintain the I reliability of the nuclear safety system by locating faults and repairing immediately. 1.4 Process Instrument Safety System I 1.4.1 Introduction. It may be observed by examining i Figure 9.10 of NBSR 9 that the rpocess instruments are not I electrically redundant. However it may be shown that a I 12-6

I redundancy exists in other instrumentation and other para-meters different than the one being measured. 1.4.2 Void Shutdown Flow. This system has been re-designed to obtain redundancy since none was present in the previous system. The Section 9.4.1.3 in the NBSR 9 must be revised. The pressure in the bubbler pipe is normally maintained to be just sufficient to keep the bubbler pipe dry, approximately 5 psig. Two pressure indicating alam switches monitor this pressure and open contacts when the pressure in this line exceeds a pressure great enough to generate bubbles. These contacts in turn will de-energize the relays Ka-01-299 and K116, and in turn de-energize the K2000 and K2001 relays which removes the -10 volts from the shim rod clutch power supplies. An examination of the circuit for the part (a) portion of the question has been made. If it is assumed that channel PIA-40 fails to de-energize Ka-01-299, then, since the PIA-40a channel is completely independent, the associated relay Kil6 contacts will open and de-energize che K2000 and K2001 relays. In response to part (b) of the question, if the pressure rises high enough to produce bubbles then a scram will be initiated by either channel PIA-40 or PIA-40a. Therefore, no single unsafe component malfunction in this system can now initiate an accident. 1.4.3 Flow Scrams. The flow channels are redundant. The main coolant flow measured by channel FRC-1 is the sum of the flow of the two channel FRC-3 and 'JRC-4. An examination I of the Figure 9.10 of NBSR 9 indicates that the loss of flow in any one of these channels will scram the reactor. Therefore, a failure of one of the channels in the unsate direction will not reduce safety to any significant extent since the other two channels are independent and will scram tb.e reactor. For example: Assume the unsafe failure of channel FRC-1. The channels FRC-3, FRC-4 will de-energize their corresponding relays Ka-01-285 and Ka-01-286 upon receipt of a scram signal. Again in turn I 12-7 A

I the K2000 and K2101 relays will op:n the -10 volt supply to the shim rod clutch power supplies and scram the reactor. In response to part (b) of the question a malfunctioning of these channels cannot initiate an accident since they I provide no control function. I 1.4.4 Level Scram. The level scram circuit is described in Section 9.4.2.3 of NBSR 9. A component failure in this channel may result in less of automatic protection from this I-channel. However, the channel FIA-2 would produce an alarm that the level has dropped and the overflow has stopped. The operator could then manually scram the reactor. Eventually the loss of level would result in loss of reflector as shown in Figure 4.22 and as discussed in Section 4.6.9 cf NBSR 9. If a serious rupture occurred which would cause a loss in level, then the leak detectors would alarm also. 1.4.5 Temperature. The fundamental variables of control for the NBSR are power level and coolant flow. As long as these parameter.s remain within bounds the reactor core is relatively safe. This condition obtains because the bulk I temperature rise is relatively low, being approximately 12.8 F as rated flow. The hot spot temperature is thus relatively insensitive to coolant inlet temperature. Nevertheless, inlet and outlet temperatures are measured and annunciation is made in the event pre-set bounds are exceeded. The reactor is subject to scram on the basis of high differential temperature between outlet and inlet temperature, however. This parameter provides redundant power level information when combined with coolant flow rate in a multiplier which reads out directly in I power. Power level is therefore measured in two ways, both by neutron density level and by process variable. Since I process variables are only relatively slowly changed, primary power level control is by neutron density level. Explicit redundancy exists however. As long as adequate flow is maintained the core is protected against fuel element burnout. As previously discussed the flow measurements are multiply rudundant. No single component failure in the temperature circuitry can endanger the core. Such failure would immediately be 12-8

I evident in discrepant power level information and temperature difference measurements. Implicit in the above discussion is the redundance of differential temperature measurements with inlet and outlet temperature measurements. 1.4.6 Process Instrument Testing. In response to part (c) of the question environmental testing of the process instrumentation was not conducted initially by NBS since the manufacturer could furnish certified test data on the components used in the system. The manufacturer, Honeywell, certifies that the AP and pressure transmitters have been tested over a temperature range of -40 F to 250 F and a humidity range of 107, to 907.. (Voltage does not apply here because the transmitter receives power from other instruments). I The temperature transmitters, MV/I, have been tested at a +40 F to +120 F temperature, 107. to 907. humidity, and 107 volt to 127 volt line variation. The balance of the equipment associated with the safety system and mounted in the control room has been tested at +40 F to 120 F temperature, 107, to 907. humidity for a 1% accuracy, and line voltages of 107 to 127 volts. 1.4.7 Operating Experience. The process instruments and their associated safety system were placed in operation March 1, 1965. Since that time the equipment has accumulated approximately 13,000 hours operation. The only failure en-I countered so far that affected the safety system was the loss of one of the 42 volt power supplies. The failure of the 42 volt process instrument power supplyoccurredafterapproximately6920hioursoperation. A series pass transistor failed by thermal runaway. The failure was repaired and no further failures have been encountered in any of the safety system components. g 12-e

I

I 1.4.8 Pre-operational Testing. Procedures have been developed whereby all the process instruments will be checked for calibration, trip settings, operation of annunciators, response times of safety instrumentation, etc. These pro-cedures are now in the process of being performed and I documented. These procedures may be repeated at a future date to verify the operation of each channel. 1.4.9 Operational Testing. The process instrumentation has not been designed to permit operational testing without shutdown. However prior to each startap the safety system I process instrumentation SCRAMS will be checked for calibra-tion, response time and periormance of its intended design function. Procedures have been evolved so that this infor-mation will be documented. It has been decided to eliminate the ability to test the process instrumentation during operation. Therefore the Section 9.4.2.6 of NBSR 9 does not apply. 2. BUILDING ISOLATION SYSTEM AND EMERGENCY VENTILATION CONTROL SYSTEM 2.1 Introduction. The confinement closure system for the NBSR is operated by six parallel relays designated "FSR" and "DSR" relays on attached Figure 12.1. These FSR and DSR relays are actuated by contacts 5-6 and 7-8 of the major scram relay K 114, which is energized during normal reactor operation. These relays: (1) serve to turn off normal reactor building ventilation, (2) place the emergency and internal recirculation ventilation systems in operation, (3) close the reactor building doors and inflate their seals, (4) close waste discharge valves and (5) close the reactor building service valves. 2.2 Reactor Building Normal Ventilation Fans, SF-1 SF-3 and SF-12. The reactor building 2nd floor fan, SF-1; 1st floor fan, SF-3; and basement air conditioning fan, SF-12 I 12-10

I are supplied electrical power motor control centers A-3 and B-4. Their electrical schematics are given in Figure 12.2. These fans are not supplied with emergency power since their operation is not required during emergency closure conditions. I No valves are directly associated with the operation of these fans ; however, they are turned off automatically by opening of FSR relay contacts in their motor starting circuits. 2.3 Reactor A-C Fresh Air Supply Fan, SF-2. This supply fan is automatically turned off on a " major" scram signal and its intake valve ACV-1 is closed. The electrical schematic and flow diagrams are shown on Figure 12.3 with the pneumatic diagram chown in Figure 12.4. Upon receipt of a closure signal, the FSR relay contact opens to de-energize relay M-44, I opening relay contacts in the electrical supply to a 4-way solenoid valve which supplies operating air to ACV-1. Figure 12.5 is typical of all ACV valve pneumatic operators. The 4-way solenoid in the energized position supplies air to the ACV operator through ports A and D with port B vented to exhaust C. In the open-energized position ports F and G of the 3-way pneumatic valve are open to vent the other side of the ACV operator. This 3-way pneumatic valve requires no extra supply air since it operates from the differential I pressure between ports E and F and G and F. In the de-energized or loss of electrical power condition to the 4-way solenoid, air is supplied to close the ACV through ports A and B with port D vented to exhaust C. The 3-way penumatic valve supplies pressure to close the ACV through ports E and G. Should normal air pressure fail, an air to close-spring to open pneumatic valve opens to supply operating air from a I 12-11 I I

I I high pressure reservoir attached to each ACV pneumatic operator. A ball check valve prevents this emergency air supply from escaping through the 4-way solenoid. The 3-way 1 pneumatic valve would supply this emergency air to the ACV cylinder through ports E and G. Power supply for this fan comes from MCC B-4. 2.4 Reactor Basement H & V Supply and Exhaust Fans, SF-11 and EF-27. These fans are operated similar to SF-2. They are automatically turned off on a " major" scram signal and their intake valve, ACV-2 and exhaust valve, ACV-3 are closed. Their flow diagram is shown in Figure 12.6 with their electrical schematic shown in Figure 12.7. Their pneumatic control system is shown in Figure 12.8. Upon receipt of a closure signal, the FSR relay contacts open to I de-energize relays M68 and M69, which opens relay contacts in the electrical supply to the solenoids for ACV-2 and ACV-3. The ACV operators for these valves are identical to that of ACV-1, discussed above. Power supply for these fans comes from MCC A-3. 2.5 Internal Recirculation Fan, SF-19. This fan is automatically started on a " major" scram signal from the FSR relay. Process room return air valve ACV-ll is opened by this signal as well. The flow diagram is shown in Figure 12.6. Figure 12.7 shows the electrical schematic while Figure 12.8 outlines the l pneumatic controls. The " major" scram signal closes the FSR contact in the " auto" start circuit to energize relay, M. l Contacts in this relay energize the air solenoid to open ACV-11. l l When relay M is energized, the relief damper M-19 from the t mezzanine is opened by pneumatic relay R-19. This allows 1000 CFM of air to be supplied to the process room through SF-11 supply duct. Return is back to SF-19 via the opened ACV-11 and return ductwork. ( Electrical power is supplied to this fan and its controls from emergency motor control center B-6. I 1-t

I 2.6 Reactor Normal and Hood Exhaust Fans, EF-3 and EF-23. FSR g M. relay contacts open, to automatically stop exhaust fans EF-3 and EF-23 as shown on the attached electrical schematics in Figure 12.10. This de-energizes relay M70 in EF-3 starting circuit, de-energizing the air supply solenoid for ACV-7. This butterfly valve is equipped with identical closure mechanisms as described for ACV-1, thus assuring its closure on loss of air or electrical power. Flow diagrams and pneumatic controls are shown in Figures 12.9 and 12.11. Figure 12.9 describes the auxiliary operation of pneumatic dampers internal I to the system. Power is supplied to EF-3 from emergency motor control I center A-5. EF-23 is supplied by MCC B-4. 2.7 Irradiated Air Exhaust Fan, EF-4. Relay M71 in the start-I ing circuit for EF-4 is de-energized by a open contact of the FSR relays or receipt of a major scram. (See Figure 12.13). This causes contacts to open thus de-energizing the air supply solenoid to close ACV-6. Flow diagrams and pneumatic controls are shown in Figure 12.12. ACV-6 is equipped with an operator identical to that described for ACV-1. Power is supplied to this fan from emergency motor control center A-5. 2.7.1 Reactor Emergency Exhaust Fans, EF-5 & 6. The sequence of operation for these fans is given on Figure 12.14. These functions are set into motion by the closing of the FSR I contacts in the starting circuit for EF-5. (See Figure 12.15). With the control room switch in the " Auto" position for EF-5 contacts 1 and 2 of CS-6C are closed and the PCX relay con-tacts are closed by static pressure controller SPC-150 (See Figure 12.17). This allows relay M42 to be energized and the fan to start. Energizing of M42 also energizes relay X42. The air supply solenoids for ACV-4, ACV-8 and ACV-10 are then energized by closed contacts in X42. I 12-13 I

I Assuming EF-5 did not start or has stopped from a failure, I EF-6 will be started. This is accomplished with the EF-6 con-trol switch in the " Standby" position (See Figure 12.16). The FSR contact in the AC control circuit would be closed, contacts 3 and 4 of C5-6D are closed, and the PSX relay contacts would close to energize time delay relay TD2. After a delay of 15 sec, TD2 contacts will close. With PCX contacts closed M43 will energize and start EF-6. In the event that AC power is lost, EF-5, relay RA is de-energized to close contacts in the DC starting circuit which energizes relay DA and starts EF-5 DC motor. Similar operations will start EF-6 on DC power should EF-5 fail to start or stop on a failure. Contacts in either X42, X43, DA and DB relays energize to close ACV-10, the return from the process room. ACV-4, 5, 8, and 9 are equipped with operating mechanisms similar to ACV-1, described above which fail in the "open" position on loss of air or electrical power. (See Figure 12.17a). ACV-10 will fail "as is" on a loss of air and will I fail closed on a loss of electrical power. Fan EF-5 and 6 are supplied with AC electrical power from emergency motor control center AS. DC power is supplied to both fans from the 125 v DC emergency motor control center. 2.8 Reactor Building Relief Valve. This valve whose controls are shown in Figures 12.20 and 12.21 is operated automatically by PS-151 to open at -2.5" H O and close at 2 -1.5" H 0. In the " automatic" position PS-151 relay contacts 2 will close on a sensed -2.5" H O to energize the solenoid 2 opening ACV-12. The valve is equipped with a closure mechanism identical I to that of ACV-1 to close the valve on loss of air. 2.8.1 Reactor Building Truck Door No. 1300/1. The reactor building truck door is not equipped with a motor drive. In the cases where a " major" scram signal or when I the reactor withdraw permit (K-104 relay) are present, the I 12-14

I seal on the truck door cannot be deflated. To accomplish this function, relay CR-1 (See Figure 12.23) is energized by either the DSR closed contacts or K104 closed contacts. With an energized CR-1, contacts are I closed to energize air solenoid DV-1A, thus inflating the seal. Depressing the " Deflate" button under these conditions energizes solenoid DV-1B, however, with DV-1A energized I simultaneously the seal does not deflate. DV-1A and DV-1B are actually two coil, 3-way solenoids. Referring to Figure 12.22, ports A and B are common with DV-1A energized or with both DV-1A and DV-1B energized while ports B and C are connected with DV-1B energized. Pressure switch R-1 operates to monitor seal pressure in the control room. 2.9 Reactor Building Doors, No. 1205/2, No. 1212/2 and No. 049/2. These doors unlike the truck door are equipped with motor driven closures. The electrical control schematic typical for these three doors is shown in Figure 12.24. As shown the door is assumed to be closed with power "off". Starting in this I condition, assume power is restored by closing the line circuit breaker. Contacts 5 and 6 of S-2 are closed and CR-1 relay contacts are closed (since CR-1 remains de-energized) to energize SV-1A. If one depresses the open button, relay CR-1 is energized through closed contacts 3 and 4 of limit switch S-1. With CR-1 energized, contacts close to deflate the seal through SV-1B. Pressure switch, " SEALS" closes when the air is exhausted from the seal causing relay MO to energize thr.ough closed contacts of de-energized relay MC. M0 contacts I close in the power supply to the door motor, which opens the door. When the door reaches its open position contacts 3 and 4 of S-1 open and contacts 1 and 2 of S-2 close. This de-energizes MO relay to prepare MC relay for door closure when the "close" button or the DRR relay is closed. Contacts 5 and 6 of S-2 open preventing air supply to the seals while the door is open. 12-15

I If the "close" switch is depressed or the DSR relay contacts are closed on a " major" scram, MC relay closes the door and S-2 contacts 5 and 6 close to energize SV-1A and inflate the I seal. On a closure the safety " knife" switch shorts to de-energize SER relay opening contacts in the MC relay circuit. The door will reopen and then reclose if the obstruction has cleared the doorway. Power is supplied to all three doors from emergency power panel P-9 which receives its power from emergency motor control center MCC A-5. Air is supplied from the building 150 psig system with 60 psig NBS plant air backup. 2.9.1 Reactor Building Door, No. 1302/2. This door is normally kept closed and is not interlocked with the major scram circuit. Its circuitry is designed to open the door when the " panic" bar is depressed. Once open, it automatically recloses and inflates the seal. The electrical schematic for this door is shown in Figure 12.25. With power available, relays TR-1 and TR-1A energize to close contacts and energize CR-1. CR-1 closed contacts energize SV-1A to deflate the seal. When air from the seal I is exhausted, the seal pressure switch closes and relay OC energizes to close contacts in the motor circuit, thus opening I the door. On opening, limit switches LSLB-A opens and LSLB-B closes preparing TR-2 to be energized when LSO switch closes just as the door reaches the full open (90 ) position. When TR-2 and TR-2A are energized contacts close to energize the closing relay CC and TR-3, which de-energizes opening relay OC. Upon closing LSLB-A recloses and the seal is inflated j g by SV-1B. l 1 The pneumatic system is identical to that shown in Figure i 12.22, however, this door is hydraulically operated. The j motor shown in the electrical schematic drives a pump which operates the door utilizing " opening" and " closing" hydraulic I cylinders. I 12-16

I l 2.10 Building Services Closure Valves, RUV-1, 2 and 3. These three valves supply hot water, gas, and cold water to the re-actor building. They are pneumatically operated valves as shown in Figure 12.26. On a major scram signal DSR relay contacts open to de-energize SV-225 and close RUV-1, 2 and 3. 2.11 Reactor Building Waste Discharge and Areaway Valves, RWV-13, 14; 15, and 16. These valves serve to isolate liquid waste discharge lines from No.1 and No. 4 sump, and to close external arcawcy drains which drain into reactor I building sump No. 1. DSR relay contacts open to de-energize air supply solenoids, thus closing the valves. Limit switches on each valve light "open" and " closed" remote indicating lights. 4 I I I I I I I I I I I 12-17

I L GG E Lf D I A LOCATED ON EQu iFM EN T OTMER TWAN ELE C T R I C A L I A L o CAT E D A T c e W e A ve M cT o se A LOCATED IW con T eoL paw EL. Sc A mro g O 'o c-T e o AT em aec-si wc r' e x u x us 7 FAN CON T e oL PAN E L. I O Loc A 7so IN cow T e c t. c e s w- - AIR G A G-G I ACV ALE CoWolTioWEe VALVE CS CON TEOL SW i TCW DA M c To R STA e 7 OPERAT I N G Co l L. g DB MoTOE STAE T OPEEATlWC-Co!L. EP E X.H A LI S T F= A M GP G LGCTEIC FNGd M ATIC D EV IC E Pse FLo A T sw I Tc M REL.Ay l M STAR TEE O PER A.T I M G Co i L. Rc. W oe hd AL C Los E O l Nio. NORM AL C P E bJ I O.L. OVER L c A.tn FC X PREssu e s con T E C L LEle Au AIL L AR Y R EL AY l FSX PReseuRE Sw iTc M A u x l LL AR= Y RE L AY g SF SUP PLY PAN SV SoLENo!O VA LV E l T DC T i k1 E D E L A Y O M CLo s t NG tl 12 ,, c, \\j A c / c,o ~ E (R*)hh COMPRESSED AIR SU P PLY (PSI) I

W W M M M M M M M M M M M M M M M M l i i 1 --l E G V A C - C P I T I C A '_ PONeR.=' !J iE L C P R R C 3-4 RRC 3-5 M AJOR SCRAM (Kil4) .E4 D T Y P (FS R') TYP CSR p. o _g 4, V ME'M! ZED (18 Fl E Q D (1 6 P EE Q D ', E g DURINC, NOPZ AL o -o o-RESET REACToE 5-M O F E R'-sT :aN. o n Y I I, - 12 :, V D.C. PAN E L *DC P ?

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(+) b (_. ) SF-IN.O. DOOR 049/2-IN.C. & IN.O. H O SF-2N.O. 1212/2-N.C. O . o V SF-3N.O. 1205/2-N.C. O SF-11N.O. TRUCK DOOR 1300/1 - N.O. FSR RELAY 9 # U SF-12N.O. HOT WASTE LINE SV-217(RW-13)N.O. g y g c. c 4_3 O5R R6 LAY 6 O EF-3N.O. AREA DRAIN SV-222 ( RW-14 )N. O. EF-4N.O. AREA DRAIN SV-223 (RW-15 )N. O. 3H O EF-SN.C. SUMP #4 DISCH SV-224( RW-16 )N. O. EF-SN.C. COLD WTR. LINE (RUV-1)N.O. g gg;4 -o EF-6N.C. HOT WTR. LINE SV-225(RUV-2)N.O. iy EF-6N.C. GAS LINE (RUV-3)N.O. EF-27N.O. 6 SPARES - 3N.O. 3N.C. M A J O F: S C R A M. R E -..- f SF-19N.C. EF-23N.O. C0M T ROL 5 SPARES - 4N.O. IN.C. ( F S R - F/ ?. 'i.. D O R - D O D E' E' NOTE: . d i i ".9 h V / / cE.T !. i _ ' ' E '. ~ B L D Q

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S (1) FURNISH 3FSR & 3DSR REL. EACH WITH 6 CONTACTS MAX. (AUTO RESET) (2) ALL RELAY CONTACTS SHOWN IN THE DE-ENERGIZED " MAJOR" SCRAM CONDITION. Figure 12.1 L

M M M M M M M M M M M M M M M M M M NOTE: R E L. AY. CO N TAC T S S W C W N IN1 D E-E N E F.:C--l E E D " M A. J c R S C RA ld co N r.:> i Tio N. SW ITC H Co t IT AC TS I N T~ H E

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ll @g @^ M45-M 41 @3 @^ RE AC TOR '" FL. PAhl 5F-3 R E AC TOR 2"P FL. FA k SF-1 REACTOR RSMT LAR A/c FMs SF-l/l Figure 12.2 l

M M M M M M M M M M M M M M M M M M [l V2 V2 A 6_. T T-g ACV-l I -l-- 0 ~I NTAKE O C = LEV. W 4 V t 2A (b I f AREAWAY S P-2 Room 4 37 FILTER Stt<UENJr. vF L,PEiU, TION /reezcstat TS-2 operates on a fall in gg 7 b te.nperature below its predetermined i L.O A -7 settino to de-enersize the unit fan. STOP F5R TC (2 On a rise in temperat ure, the reverse g g g ,g @}d- -_.]. oecurs. o w o o j,4 ' M44 when the unit fan is det-energized, the .ac u t i t.o coil. vch, s are opec rna e 29-1 ( g i4 .l. _d. g g [ W ..se: L.w.ma t tan is energi. zed, i.JV-1 is opened and the control nyntec. opcrates M44 AC V-1 cs follo. b f - 7ture controller T0-2 controls in sequence the re-heat coil valve V2B, A pre-heat coil valve V2A, and cooling y coil valve V2 to maintain a satis-C M 44 6 factory discharge air temperature. On loss of air or electric signal d AdV-1 closes. N O T E

• fd? E L.AY CONTAC75 cs uow N iw r w e ' os-sueRc izerf RGACTOR % FRESU Alf FAW SF-2 w xe s c em com m,o s.

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Z3 TS ? Y T T-2 h 99 99 40* F T N -Z F j l--] hh Tc-Z 53*F D' A-lu SERIE6 WITH l I F AM ST A RT8 N G l i(G/(,o C' =c u i r, n j pe iv. 44 y ' I L _ q p _ _I II 4 7 = l g j'7 i 0 vsv C_ O Acv.i RC. $is NOT E : __l m m u R E L AY C oN TA c-r H } STEAM ~ ^ g l e SHOW N IN DE-EHERC,12ED vee veA C0HIRol "MAJo R "Sc RA M, Ac v-i 7,"[ ew-pe# cataso 2 l " o-r l CLose o PositTioN MTER d$ d 6 l REACTOR A/c FRESlJ AIR PAN 5F-2 Figure 12.4 \\

m m m m em m M M M M M M M M M M M I I WORM A

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-.- -.b-l 6 ( o m l B A. L L WIPRESsues C W E C kt. C esse!Rvot R (3 VA LVE s y. s i r l iD l . _ L ;L, E X. W A.d 5 T I / EE-F- E x. L4 AU ST = 3 G O P EhJ== L ~~~ ~~ = CLOSE l -- -- A C V CPE li%TC E SWoWN [ kJ VALVE OPEN PostTioN TYPICAL ACV VALVE PM EU ki ATic O P E E /A T o lE Figure 12.5

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/ 7 g ^b ~ C ese N V i'. m k CV'3 m E P.27 pg=om E L EVATOR op Ef:-E 'd AC I4 tN E RM FROM i N' NTAKE Dil Tt pc SFil / D 7- ' AREAT[k ,/ _k

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~ F L o o R -, r lud FROM E E Ac TOR _ / g es> To iST 4 E" FL. eu man v AC V-O g pje; SEQUENd6 0F UPERATIt.N Room thermostat T-ll controls in sequence the heatint, coil valve V-ll and the fresh air and return air dampers to maintain a satisfactory space temperature. Low limit controller T-11A operates on a fall in temperature below its predetermined settin to 3radually close the fresh air damper and open the return air damler and to redually open T-ll to maintain a minimum dischar c air o e o temperature. 'then the supply fan SF-19 is energized, the relief damper frota the mechanical equipment is opened and the AdV-ll in the air stream from the reactor basement is opened. When this fan is de-energized these dampers are closed. ilhen the supply fan SF-li is ener31 zed, the dis.:nar,e air draaper is opened and AdV-2 in the fresh air intake is c opeaed and AdV-3 octueen the hold-up chamber and the suction of fan EF-2 is opened. .Jhen this fan is de-energized, these dampers and AdV-2 and AdV-3 are closed. Static pressure indicating controller PC-27 shall control the static pressure damper on the enterin side of the hold-up chamber to maintain a satisfactory static pressure between the basement floor reactor area and the first e floor area. Un loss of air or electric s.ignal, AdV-2 and A;V-3 are closed. iPG AC TOR S5MT W 4 V AN D IN TER'N AL K ECIRCU L ATION FANS, EF-21,5F-(1,$ 5F-19 Figure 12.6 l ~

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I QUESTION 13. Coordinate the discussion of the reactor safety system and the accident analyses so that the maximum trip settings, response times, and other applicable requirements for each of the safety system parameters is established. This discussion should show I the sensitivity of the accident consequences to variations of safety system response or trip points. In coordinating the discussion of the reactor safety system and the accident analyses it is necessary to consider two types of occurrences. One is the ef fect of the process variables on the fuel I while running at or near the safety trip setting, and the other is the rapid, short-lived occurrences studied in the startup and reactivity insertion accident analysis. The safety system process trip settings were chosen to give as great a differential from the minimum acceptable conditions i shown in Tables 4.7-1 and 2 of NBSR-9 (in the safe direction), as was consistent with practical operating practices. These minimum acceptable conditions were predicated on the complete absence of boiling at the core hot spot, and individually, result in burn-out ratios better than 2 The individual trip settings therefore have an even greater margin of safety. Of the process variables monitored by the reactor safety system, there are four which effect the fuel temperature and the burn-out margin; reactor power, primary flow rate, core dif feren-tial temperature, and reactor vessel level (as re flected in the I hot spot saturation temperatures). Each of these variables is monitored by separate instrumentation and each has its assigned sc ram t rip point. The detailed description of each of these instru-ments is given in Section 9 of N13SR-9. Table 13-1 of this response summarizes the trip points for the variables in question. Since each of these process variable trip functions operates independently of the values of the other variables, consideration y 1,.i

%P was given to the effect on core temperatures and burn-out ratios of all possible combinations of these process variable levels at the trip points. For the purpose of this analysis, only the hot i spot temperatures and heat flux conditions were considered. As pointed out in Section 4.7.3 oi NHSR 9, the fuel temperatures and the corresponding heat fluxes are calculated on the assumption that all possible adverse condi tions occur simultaneous ly at the same (hottest) spot in the core. l }I Table 13-1 l NBSR Safety System Trip Points ( Sc ram ) 1 1 Function Iligh Flux Leve l 130"; rated power l l Low Primary System Flow: Reactor Out let 4590 gym Reactor Inlet Startup Core Outer Plenum 3000 gpm t Inner Plenum 1600 gpm Equilibrium Core I Outer Plenum 3500 gpm Inner Plenum 1090 gpm High Corc Dif ferential Temperature 15.5 F i l Lcw Vessel irvel 168" D 0 2 While it is possible to show a wide margin of safety for each j trip setting, the best twasure of the true significance of these settings from the standpoint of an accident milysis is to simply consi de r a ll four var iables at the respec ti t.e t rip level while the reactor continues to operate. Th t :> at tuatim wuld then call for a reac tor powe r 1307, above norrna l, a ce re differential t enpe rature l l of 15. SF a s opI'ose d to 12.2'F normal, the ve%el levei <!an t1 168 inches frem 175 inches no rma l, anu t he p r i ru ry system 1. Aced to 1600 cpm and 3000 g;n,

csel t<> 1745 m n and 1165 gpr I

it-( 1 . ~ -.

the inner and outer plenums respectively. Further, this unfortunate combination of events must be assumed to occur during startup, prior to the build up of fission poisons. This results in the use of the highest power per fuel channel to be found in the life of the NBSR core, and the burn-out ratio thus calculated will be the lowest possible margin of safety for the NBSR. The resulting I burn-out ratio is 1.7-Examination of the instrument accuracies and response times as given in Section 9 of NBSR 9 in relation to the very wide margins of safety reflected by the burn-out ratios shows that both the I rated accuracies and response times make only small differences in the burn-out margin, and that,as a measure of the significance of these settings, doubling the rated accuracies and response times still would not violate the minimum acceptable conditions, i.e., individual burn-out ratio >2 As for deviation f rom the trip points, a safe margin of 77. of the trip point parameter setting exists between the most sensitive setting (inner plenum flow at startup) and the miminum acceptable condition. (Trip setting 1600 gpm, minimum acceptable point 1494, nonnal flow 1745 gpm). This safe margin ranges up to 87. of the parameter trip setting for the reactor outlet flow. The second type of occurrence to be considered is the rapid short-lived excursion. For this discussion the detailed accident I analyses given in Section 13 of NBSR-9 offers the best measure of relationship between instrument response times and settings, and the significance of a given accident. The occurrerces shown in these studies are extremely conservative upper limits to any possible extent of fuel damage. Power trip levels over 137, higher than the actual trip settings were used in the startup accident analysis and a rod drop delay double that actually measured was used to the fuel element insertion accident. Although burn-out at the hot spot is calculated by these conservative means to be reached for approxi-I mately 70 milliseconds for the startup accident, and.7 seconds for m'aximum reactivity insertion, actual clad penetration would be 13-3

I limited, if indeed at all allowed to occur, to an extremely small point by the thermal capacity of the fuel element structure. At I the actual settings of the NBSR then, it is felt that no fuel damage would occur. I I I I I I I I I I I I I I 13-4

I QUESTION 14. What monitoring and control over gases and vapors evolving from the evaporizing of liquid wastes is provided?

RESPONSE

14.0 The question of monitoring and controlling gases and vapors f rom liquid wastes is basically answered in NBSR-9. All air which leaves the reactor building is examined by three detectors, namely the particulate filter monitor, the gaseous effluent monitor, and the stack monitor. In particular, the hot waste sump vapors are collected in the irradiated air system and subjected to the three monitors RD l-11, RD 3-4, and RD 4-1 in that effluent stream. I Gases and vapors from the wam laboratory hoods are monitored at the particulate filters. I I I I I I I

I I

I s 14-1

I QUESTION 15 Discuss the provisions for tritium control for both on-site and off-site personnel during normal operation or in the event of an accident, in-cluding a heat exchanger tube bundle failure which causes carryover of all primary water to the secondary cooling system. In the event it is necessary to isolate the main heat exchanger to effect tritium control (or for other reasons) discuss how core cooling would be accomplished. EESPONSE The exposure control method for tritium at the NBSR is based on a body uptake of 1 mC1, resulting in a whole body dose of approximately 85 mrem the first week and a total integrate.1 dose of about 210 mrem. -6 The maximum permissible concentration in air is 5 X 10 Ci/cc for a 40 hr/wk. The total body burden is 2 mci and the body tissue burden I is ImC1. Tritiated water vapor can enter the reactor confinement building atmosphere whenever irradiated heavy water is exposed or the helium system is opened. This will occur when: a. Fuel elements or experiments cooled directly by the heavy water are loaded and unloaded from the reactor; b. The D 0 or helium circuits are opened for maintenance or 2 modification; c. Samples are taken of the D 0 or helium; 2 d. There is any accidental spillage of D 0; and 2 e. There is any leakage of helium from the sweep gas system. I Precautions to prevent or minimize exposure to tritium will include: a. Monitoring the air (1) with cold trap samples of water vapor evaluated in a liquid scintillation counter, and (2) with air samples taken in an evacuated ionization chamber and evaluated with a vibrating reed electrometer; b. Monitoring the surfaces with smears counted in a windowless gas flow counter; Restricting exposure time for tritium concentrations in excess c. of MPC; I d. Requiring that impervious plastic suits and air breathing equip-15-1

L F 5 ment be used for levels much greater than MPC; and c. Using a local exhaust system for ventilation around maintenance operations on the D 0 and helium systems. 3 The only means (at the present time) of determining a worker's actual tritium uptake at the NBSR is by urinalysis using liquid scintillation counting. Therefore, routine sampics from persons working with tritium or tritium contaminated materials are necessary. Non-routine samples will also be collected after suspected uptake of tritium. The tritium concentration in the urine as a result of a one millicurie uptake is about 23 nCi/1. If the bioassay result is above 40 nCi/1, the individual is to be removr' from all work involving radiation until the result is below 10 nCi/1. If the result is between 20 and 40 nCi/1, the individual ~ is to be removed from all work involving tritium until the result is below 10 nCi/1. Experience at the Savannah River Plant has shown this to be an effective control following a tritium uptake. ~ The three most probable modes of tritium release to the environ-ment are through the liquid waste disposal system, effluent release from the stack and release to the secondary cooling system, thus the atmosphere via the cooling tower, from a D 0 to H O heat exchanger tube leak. 2 2 D 0 can enter the liquid waste disposal system only through the 2 reactor building sump pumps. To reach this sump (1) a rupture must occur in a system containing D 0 and (2) the spilled D 0 must drain or be pumped 2 2 over a 24 inch high curb which surrounds the D 0 systems. Should a pipe 2 rupture occur, heavy water leak detectors located throughout the D 0 systems 2 would sound an alarm, thereby alerting the reactor operator. No water I would be released from the reactor building sump to the sewer system under these conditions until the water was first sampled and analyzed for tritium using a liquid scintillation counter. Gaseous release via the stack is to be controlled at the point of origin; that is, at the local maintenance or D 0 sampling activity. Re-2 leased concentrations of tritium would be controlled so as to limit instantaneous H -6 I release to 10 times MPC for unrestricted areas (2 X 10 c/cc) accounting for dilution by sir in the exhaust system. Air samples will l be taken in an evacuated ionization chamber and evaluated with a vibrating reed electrometer. Tritium concentrations in excess of approximately 15-2

1 X 10-c/ml can be detected. In addition to grab sampling for H activity, a remote system has been installed. This system consists of I an air pump which can draw air samples from nine locations within the reactor building through a parallel network of solenoid valves electrically operated from a 24-point recorder in the control room. At periodic intervals air flow through each solenoid is singularly passed through an on-line vibrating reed electrometer capable of detecting H in con- -6 centrations of approximately 3 X 10 ac/ml and greater. The reading is printed out of the control room recorder, which then steps to the next point. The following points within the building can be continously monitored using this device: (1) Reactor Basement Exhaust at ACV-3; (2) Normal Exhaust at EF-3 discharge; I (3) Irradiated Air Exhaust at EF-4 discharge; (4) Process Room Exhaust at EF-27 discharge; (5) Second Floor A-C System at SF-1 suction and a 20' hose for monitoring atop the reactor during refueling operations; (6) First Floor A-C System at SF-3 suction; (7) Process Room D 0 area - two 3T hoses for remote monnodng 2 during D 0 system maintenance. 2 (8) Reactor Sub-Pile Room; and (9) Monitor Room. I An installed N mn r in secoMary cooHng system wouM 16 warn of a rupture in the primary to secondary heat exchangers. A sample of the secondary water when analyzed on a scintillation counter would be 3 used to verify H concentrations and all contaminated water disposed of at a controlled rate below concentrations for unrestricted areas. Periodic sampling of the secondary cooling system will also be performed to detect any long range increases in H concentrations caused by a leaking heat exchanger tube. These samples will also be analyzed using a liquid scintillation counter. I Should the primary heat exchanger HE-1 develop a leak, the reactor could be cooled in the shutdown mode through the use of the purification heat exchanger HE-3. Once entry could be gained to the process room, HE-1 would be manually isolated by closing valves DWV-95 and DWV-100 (refer to Figure 5.2 of NESR-9). Approximately 350 KW of heat could be continuously removed from the reactor core by using the D20 storage I 15-3

tank pumps to force coolant through HE-3 and into the primary system via DWV-39. D 0 would then pass up through the core and flow out the I 2 normal overflow line through DW-10 to the storage tank. This system is capable of removing the core decay heat almost immediately after shutdown while maintaining coolant temperatures in the 100 to 115 F l temperature range. 1 I I ,I I

I lI l

!I } lI i

I iil

!it il I I i 'I 15-4

I I QUESTION 16. Describe the methods and procedures you have considered for monitoring I iodine releases and their relative advantages and disadvantages. Justify the method (s) selected for both routine and accidental releases.

RESPONSE

IODINE AND RELATED DETECTION NBS has recently reinvestigated the state-of-the-art of iodine stack monitors. This was done by searching authoritative lite rature reviews such as are found in the AEC journal of " Nuclear Safety," I and by contacting experts in the field. The detectors employed are of two general types:

1) the " moving tape", which is prompt but nonquantitative ; and 2) the " cartridge",

which is quantitative but non-prompt. The combination of promptness I and quantitativeness has not yet been realized; moreover, the dynamic range required for the detection of large iodine release rates does not exist in any of these instruments. The best of the stack detectors indicate reasonably promptly the presence of iodine in excess of some preset low level. They do not measure quantitatively the total stack release, nor do they possess a dynamic range adequate to assess a large and variable release such as would pertain to a core meltdown. In view of the above, the NBS position with respect to the technical feasibility of these devices is that they are not adequate to yield meaningful information in the application for which they were suggested; I namely, to evaluate of f-site dose in the case of the worst hypothe-tical incident. The iodine dose referred to above was established previously and cited in this context, especially on pp. 8-10 of "Brief of I National Bureau of Standards in Support of Exceptions Taken to the Initial Decision," dated April 13, 1965. Its value recently has been reestimated taking into account additional factors which could g 1e-1

augment the effluent release. These have been outlined and analyzed in the response to Question 4 of this submission wherein it was shown that a dose to the thyroid of ~ 63 rem might be possible. Though a dose of this magnitude is still very small, NBS regards that in the event of such a release it would be advisable to search downwind with a detector which will measure the iodine concentration at ground level. For this purpose, NBS plans to utilize the so-called "Staplex" device which passes air at the rate of 10 cfm through a filter pad that retains most of the iodine. l The pad is then observed with a NaI-Tl scantillation detector set -3 for iodine 131 y-rays. This technique is sensitive to 10 C l iodine. Thus, if the collection time is ten minutes, an iodine -0 concentration in the air of ~ 4 x 10 C/cc can be detected. Correspondingly lower concentrations could be detected for longer collection times. In a persistent wind, NBS would operate several of these detectors for various collection times and repeat the process. j Considering the ten minute collection time as typical, the -10 concentration of 4 x 10 C/cc if breathed continuously for a period of 30 days would result in a thyroid dose of ~ 0.5 rem. This value l is a factor ~ 125 times less than that cited for the worst hypothetical incident. Clearly then such a technique could appraise NBS of an unanticipated high concentration and allow it to pass judgement on the need to alert off-site persons of this hazard. It should be pointed out that steady measurable concentrations of this magnitude would not be expected in an actual situation. Either the filters function and/or the wind direction fluctuates considerably more than assumed for the most adverse meteorological conditions. Either of these factors would obviate a steady concen-tration of the magnitude cited. A negative detection result is, of course, acceptable as long as the sensitivity of the detection technique to steady wind conditions is adequate to indicate a I hazardous situation. That this is the case has been demonstrated above. Accordingly, NBS concludes it has more than adequate iodine detection safeguards. g 1e-2 l

I There is no routine release of iodine, only A which is detected by the stack monitor described elsewhere. Lesser accidental releases than result from a core meltdown are contained in the primary system. For example, if a pin hole, crack, or other rupture develops I in one of the fuel plates, noble fission gasses would be released to the helium sweep gas and promptly detected by monitor RD 3-2 as discussed in Section 9.7.3.2 of NBSR 9. I I I I I I I I I I I I I 16-3

I QUESTION 17 Provide quantitative information to show how radiation monitor RD4-3 in the liquid waste system can be set so as to assure that the Commission's limit of one curie per year for releases to sanitary sewer systems (10 CFR 20-303) is not exceeded. Consideration should be given to: (a) variations in flow rate; (b) variations in composition of the fluid being monitored; (c) minimum sensitivity of the monitor employed; (d) the unknown nature of possible activity released; (e) the need to include tritium and other low energy emitters. How is the necessary functional or physical redundancy provided in this automatic system to assure its operation when required?

RESPONSE

A design to effect the liquid waste control philosophy stated in NBSR 9 has been developed. As previously stated, the purpose of the system is to protect the drainage system from the reactor and warm laboratory areas from accidentally discharging to the main laboratory sewer significant Jevels of radioactive liquid waste. All drains from this area are treated as suspect of radioactivity although none of these I drains are permitted to be used for waste disposal purposes. There should be no reason to expect other than small accidental spills to any of this system. All suspect waste drains lead to a monitor and holdup volume as shown in the Figure 7.19 of NBSR 9. Flow rate in this system is monitored by a variable area flow meter. Valve RW-2 will be normally open while valve RW-3 will be normally closed. A holdup tank which permits a 6 minute delay before discharge at the maximum possible flow rate of approximately 30 gpm is included before valve RW-2. If the counting rate of monitor RD 4-3 is integrated over the delay time associated with the holdup tank, a decision as to the radioactive content I of the effluent can be made with reasonably good statistics to give assurance that maximum concentration levels are not being exceeded. I 17-1 I

I If pre-set levels are exceeded, RW-2 will close and RW-3 will open to cause all effluent to be diverted to the 1000 gallon retention tank. The retention tank is sized to give approximately 30 minutes for opera-tion personnel to effect a stoppage of all effluent to the system even under the unlikely condition of maximum flow of approximately 30 gpm. Normally it would be expected that less than 1000 gallons of effluent will have been accumulated that is contaminated. The contents of the retention tank can be sampled by means of an ejector system located at the hot waste control panel in the cold lab basement where the radiation monitor and flow meter indicators and recorders are located. Controls of pumps and valves are located at this panel to permit transfer of the contents of the retention tank to I either a 5000 gallon holdup tank for storage and/or dilution, and subsequent decay, or to one of either of two batching tanks. The contents of the batching tanks can be sampled before pumping to the hot waste treatment area in the hot laboratory. Tank levels and tank transfer controls can be effected from remote panels. Finally, the contents of the 5000 gallon holdup tank can be sampled after which its contents can be returned to the effluent control system and discharged to the sewer. This larter procedure would be accomplished only after safe radiation levels have been achieved; but in any event, the discharge I of this tank would be treated like any suspect effluent and counted by detector RD 4-3. A development project is underway to extend the technique of monitoring with the RD 4-3 detector and ratemeter. Since the flow rate can be highly variable, a preferred basis for diverting flow to the retention tank is the integral of the product of flow times count rate, integrated over a period determined by the time integral of flow which is equal to the volume of the delay tank. Such an arrange-ment can reduce the frequency of unnecessary diversion when high activity I concentration accompanies a momentary low flow condition. This latter technique is in the experimental stage, however, and is not a necessary technique to prevent excess activity being released. The detector RD 4-3 is a Tracerlab Model MD-12B thin walled beta I gamma geiger counter and has been calibrated in its operating configura-tion for numerous possible effluent activities. The isotopes used for 17-2

I calibration to date are listed in Table 17.1. It should be noted tha t 90 of the five isotoped listed, only Sr can probably not be detected above background at MPC levels.* I It should also be noted that of the four other isotopes used for 137 calibration, the detector was least sensitive for Cs In this case, howner, the signal to be expected from an activity level equal to MPC is four times greater than the natural background of the I counter. The detector measures a counting rate level one hundred times background for Na a c t. ivi ty. Operational history of this system has begun to be available for a quantitative discussion of the conditions it must satisfy to meet the requirements of 10 CFR 20.303. To date records have been maintained I for flow and count rate. A summary of flow data is presented in the attached Figure 17.1. This figure gives a histogram or frequency distribution of the total number of gallons per day of reactor and warm laboratory effluent. It should be noted that the mean total flow per day is approximately 2290 gallons and the maximum observed total flow per day, 4000 gallons. At this mean flow rate, if the system were to -5 continuously pass the level of 9 x 10 c/ml of Appendix B the total activity released per year would be 0.003 curies, a factor of 300 less than a one curie limit. A further calculation can be made to show the capability of the I system in meeting the conditions of 10 CFR 20.303. If it is assumed that a spill occurs to the system which is ten times the level of Appendix C for the isotope Cs for which calibration test indicated 33 the lowest sensitivity, then the response of the detector can be predicted. Assuming the spill consists of 10 pc and this enters the detector whose sensitive volume is approximately 2.5 liters moving at flow rate of the mean observed rate of 2290 gals / day, then the a detector signal would reach a level 40 times background levels for the

• For this reason it is proposed that administrative procedures will be written to exclude from use in the laboratories, unless separate special provisions such as dry glove-box operation is included, all of the isotopes listed in the table of note "C" of Apg24, Raendixg26, of

Sr90, 29, 10, At211, Ra223, Ra I

10CFR2Ogamely30,Pa Ac227, Ra , Th , Th and natural Th. 17-3

I period of transit past the detector of approximately 25 seconds. Since this isotope is restricted to the lowest level of Appendix C, excluding the isotopes previously mentioned as administratively controlled, it seems clear that a censiderable margin of safety is inherent in the system. As backup information on the nature of the effluent from the reactor building, a routine sampling procedure has been established whereby the contents of the delay tank has been examined for activity. I In a sequence of approximately daily samples taken since the middle of July, no activity level greater than 5 x 10' c/ml has been observed. I The system has not been designed to detect tritium in the hot waste effluent. It should be noted, however, that in previous dis-I cussions the tritium hazard has been analyzed. In any case the system is closed and sealed from the only source of tritium contamina-tion, a D 0 spill in the reactor basement. As further precaution, 2 however, in the event separate D 0 leak and tritium monitors alarm 2 the control valves of the Hot Waste Collection System will be interlocked to divert all flow from the reactor building sump to the retention tank, so that the effluent can be retained and sampled before release to the sanitary sewer. I I I I I I I I

I I I I I TABLE 17.1 HOT WASTE DETECTOR RD 4-3 CALIBRATION l ACTIVITY NET DETECTOR i NB LEVEL COUNT RATE ISOTOPE uc/ml cpm -5 Sr 1 x 10 51 C1 1 x 10~ 25 -5 Na 1 x 10 30 -4 Cs 1 x 10 30 Ce-Pr 1 x 10~ 812 I I I I

• e cxe e - 30 ce.

st ce e tr c 1 3 I I I 1,-5 I

!l .m N O TI 0 U A 0 B R :: I l 0 I : a" l 4 TS a l D Y

^' e.

C " 0 N" 0 E :" N ^;e " 2 = 5 U 3 'O = Q: "n ": I e ER : = F : = 0 . 00 3 YA l D l R E I 0 P ~ 0 1 A 5 S 7 E 2 N 1 O hi' L l iiI I I 1 i l L E A R l G U = 0 G = 0 F I 0 O F 3 R E B M I U 0 N 1 0 5

== 00 I 0 1 005 i e 4 3 2 m my:s Eys \\lIl l l1 ll

I QUESTION 18 Describe the type of charcoal to be used in the emergency exhaust and internal recirculation filters. Calculate the fission product burden each of these filter system can accommodate without heating to the point of off-gassing or burning. Discuss the expected efficiencies for these filters based on the most recent filter testing I data. Also, discuss the testing to be conducted to assure initial and continued proper operation of particulate filters. 1. CHARCOAL TYPE The emergency exhaust and internal recirculation system for the NBSR both utilize coconut shell charcoal. This charcoal is as-received Barnebey-Chaney charcoal with a proprietary treatment I to increase the surface area and is designated as MSA25725. This charcoal has a retention for iodine greater than 1100 mg per gram of carbon and has particle sizes ranging from 8-14, Tyler mesh size, I with a hardness of approximately 95*/.. 2. FILTER HEATING The internal recirculation system for the NBSR contains three banks arranged in series, each having twelve 2' x 2' x 1" treated I charcoal filters. These filters are of the type described above. The emergency exhaust system contains two parallel filter systems, each having three 2' x 2' x 1" treated filters arranged in series. In response to the question of fission product burden reference W should be made to the previous discussion of this question as found on page 2 of "The Reply of the NBS to the Letters of the Hearing Examinor Dated April 21, April 28, May 6, 1964" in docket #50-184. Recalculations have been perfonned using the actual physical characteristics of the final in-place filters. The assumptions are restated for clarity. First assume, for the purpose of maximum filter heating, that all available iodine is collected on one bank of the recirculation system (twelve 2' x 2' x 1" charcoal filters). 10M. of beta activity and 50"'. of gamma is considered as a possible heat source. 1 l l l l 18-1

I Considering 1007. release of Xe-Kr and 507. release of I, and a 2 hour cycle time for the recirculation system, the maximum deposited I power would occur 3 hours after the release and is approximately 1190 watts. This heat source would be divided equally among the 12 filters for a power of approximately 100 watts each. To arrive at the worst conditions, the air flow is assumed to cease at this time, thus stopping forced convective cooling of the filters. Utilizing one dimensional conductive heat flow to the outside walls of the filter frame and duct by the screening which holds the charcoal granulars in place, the maximum temperature reached in the charcoal I is approximately 179 C. With an ignition temperature of 360 C for charcoal as reported by Durant 190 watts of power can be collected on a single 2' x 2' filter before burning could be I experienced. The emergency exhaust filters also contain charcoal for iodine I absorption. It is actually only these filters which could release or pass iodine to the atmosphere exterior to the building. If the emergency exhaust system is operative at the time of the accident, these filters will also accumulate a heat source of iodine. The heat source accumulated in these filters is approximately 25 watts if both the recirculation and exhaust systems are operative. The .naximum filter temperature under the assumption of loss of air flow at the time of maximum heat is approximately 60 C. If the assumption is made that the recirculation filter system does not I operate while the emergency exhaust filter system does, the heat accumulation in the charcoal is altered. The heat source would be expected to reach a maximum approximately 40 hours after the accident assuming constant operation at the maximum flow rate, a condition which is very conservative since the average rate is expected to be approximately 207. of the maximum rate. Under these assumptions the filter would accumulate a source of approximately 90 watts. The maximum filter temperature is, the re fo re, less than the 179 C calculated above for the recirculation system. I I

• W.

S. Durant, " Performance of Activated Carbon Bids in SRP Reactor Confinement Facilities," DP-1028, January 1966. 18-2 1

I 2 Recent work as reported by Milham indicates that disorption or off-gassing of iodine which has been collected on charcoal filters is negligible at temperatures up to 250 C, which is well above any expected temperatures of NBSR filters. 3. FILTER EFFICIENCY Recent work as described in ORNL-TM-1291, "The Release and Absorption of Methyl Iodine in the HFIR Maximum Credible Accident," by Atkins and Eggleton has identified organic compounds which penetrate charcoal filters. The largest component was identified 5 as methyl iodide. Further, work by Collins has defined the mechanisms by which methyl iodide is generated in a reactor accident. I These include direct release from fuel melting and desorption from surfaces within the containment building. Peters has compiled estimates of CH 1 riginating fr m the descrption process for three 3 installations at Oak Ridge. These range from negligible to li, of the iodine inventory. He has further stated factors which contribute to this desorption process: (1) Bare ferrous metal surfaces I (2) High surface to volume ratios exposed to the released iodine (3) Reducing atmospheres (4) Presence of organic gases (5) Long Aging periods (6) Moderately high temperatures (7) Moisture or steam (8) High deposition velocities (9) Low gas phase iodine concentrations I

• R. C. Milham, "High Temperature Carbon for Iodine Absorption,"

DP-MS-66-45 September 1966. I 3* R. D. Adams, W. E. Browning, Jr. et.al., "The Release of Absorption of Methyl Iodide in the HFIR Maximum Credible Accident," ORNL-TM-1291, October 1965. I 4* Ibid, page 10 of Footnote 3.

5. Ibid, page 11 of Footnote 3.

Ibid, page 12 of Footnote 3. 18-3 /

Based on these factors Adams, Browning, et.al. state "...there is no evidence which implicates painted surfaces as contributing significantly to methyl iodide production." They further conclude that methyl iodide formed by desorption would be negligible at the HFIR, however, I less than 5% of the iodine inventory would exist as methyl iodide released from the fuel during the MCA, this being somewhat higher than similar estimates by others. "Whereas the mechanism for removal of elemental iodine is similar to chemical reactions, the removal of CH I is more of a mechanical 3 or physical rather than chemical, it is therefore very sensitive to conditions such as temperature, humidity, gas velocity and char-coal depth. In general, methyl iodide will be retained more efficiently by charcoal the lower the temperature, the lower the gas velocity, the shorter the duration of gas purge and the greater I the depth of charcoal." Adams and Browning reported that a one inch bed of treated MSA charcoal would be 94% efficient in removing CH after urs of 3 operation and a two inch deep bed would be 99.8% ef ficient af ter this time of collection. These quoted efficiencies were observed I at a temperature of 23-25 C, a relative humidity of 65-70%, and a gas velocity of approximately 40 ft/ min. Efficiencies of non-treated charcoal at like coaditions are quoted at 85.2% and > 99.99% (dropped to 99.6% in 2.5 hours). Gas velocity variations under like conditions were observed to reduce efficiency for ce,1 f,om ee.e% at 1e ft, min to es.e% at Se ft,m1n.t111 1ng un. treated charcoal. The type of charcoal used in the NBSR installation would be expected to have an efficiency between these quoted values since the NBSR charcoal has been treated but has not been impregnated. If one conservatively assumes the NBSR charcoal to be untreated and I unimpregnated, linear interpolation between stated efficiencies for 7. Ibid, page 12 of Footnote 3. Ibid, page 23 of Footnote 3. 9. Ibid, page 25 of Footnote 3. t 18-4 /

a varying gas velocities yields a value of approximately 95.57. af ter 2 hours for the 25 ft/ min, in the emergency exhaust system and 757. for the 104 ft/ min in the internal recirculation system. These efficiencies apply to a 1 inch bed of charcoal with a 2 hour collection time. The NBSR installations provide for three one-inch chracoal beds in series, giving an installed efficiency of 99.9+7. for the emergency exhaust system and 98.4r7. for the internal recir-culation system. 4. TESTING MET 110DS Recent tests by Adams and Browning aboard the NS SAVANNAH are presently under review for possible application to in-place testing of both the recirculation and emergency exhaust systems at the NBSR. The results of numerous tests aboard the NS SAVANNAH utilizing two techniques have been reported. One method injects 131 radioactive I into the duct upstream of installed filters with air samples taken on each face of the filter. Knowing the activity of collected iodine upstream and downstream of the filters, one can calculate the efficiency. A second method utilizing elemental 127 I and activation analysis techniques has also been employed. 27 The latter method using I appears to be the most promising for use at the NBSR. Testing of particulate " absolute" filters as stated in Section 3.6.3 of NBSR 9 was performed at the Oak Ridge I Filter Test Facility prior to installation. In-place testing of these filters was conducted after their initial installation using the dioctyl plitholate (DOP) method. Periodic retesting will be conducted at approximately 3 month intervals. An evaluation of a series of these periodic tests may prove that less frequent retests are required to assure continued efficiency. I I I

10. R. E. Adams, and W. E. Browning, Jr., " Iodine Vapor Absorption Studies for the NS SAVANNAll Project," ORNL 3726, February 1965.

} 18-5 L

} I QUESTION 19 Provide the results of additional comparisons of meteorological data between the NBS weather station and the Washington National I Airport weather station to confirm the correlation factors obtained from the April 1962 data. The weather station for the NBSR was reinstiituted during the spring of 1966. An "Aerovane Wind Transmitter" identical to the wind instrument used in the previous study

• was mounted on the I

southwest corner of the reactor confinement building roof parapet and placed in operation on February 10, 1966. The wind recorder chart speed was reduced for these measurements by a factor of two since the period of measurement was intended to be more extensive and the consequent data handling procedures were somewhat simplified. The shelter which accommodates the hygrothermograph and micro-barograph was not available until late spring but barometric data was taken from a building interior position until the shelter could be located. An analysis of data from the months of April, May and June, 1966, is presented herein. The weather parameters which were re-studied were temperature, barometric pressure, relative humidity, wind direction and speed. The main purpose of this work was to pro-vide further direct comparison between the weather data taken at NBSR and at the weather station of Washington National Airport.** Since the data from WNA is no longer published on an hourly basis, special arrangements with WNA were necessary to obtain data I for a particular day. It was possible to correlate temperature and relative humidity data on an hourly basis for a period of one day I only. June 30, 1966 was chosen. A temperature correlation coefficient of 0.975 and a relative humidity correlation coefficient of 0.879 was I

• NBSR 7C, Preliminary Hazards Summary Report, unpublished report
  • Hereinafter designated WNA 19-1

calculated for this day. The previously reported

• temperature correla-tion for April 1, 1962, yielded a coefficient equal to r

= 0.924. For relative humidity the hourly data of April 1, 1962 at NBSR and WNA l correlate to the degree r = 0.891. It is clearly evident in this comparison that the temperatures correlate to a very high degree while the relative humidity shows somewhat less correlation. Barometric pressure data was available at the NBSR station for a more extended time and, therefore, it was possible to use daily average pressures for an entire month to correlate the data from NBSR and WNA. The comparison of daily averages for the entire month of April 1966, yic1ds a value of r = 0.974. A scatter diagram (Figure 19.1)has been provided to illustrate the relationship between the barometric pressure data from the two weather stations. Since abundant data was available, daily averages for the entire month of May 1966, were used I to calculate another correlation coefficient r = 0.991. In the XXI 2 1962 study, a similar comparison was made for the barometric pressure I during April of that year, and a correlation coefficient r = 0.9993, again shows the high correlation for this parameter. Of most importance to this study are the wind data. In the former study, these were the only data which failed to yield a high correlation on an hourly basis. In the present study, both wind speed and direction were tabulated on an hourly basis for a period of forty-two separate days during the months of April and May. The remaining days were neglected because of incomplete records on the part of NBSR. Since the 1962 study, the United States Weather Bureau has changed their method of reporting wind data. Instead of giving the daily average wind speed and prevailing direction, they now report the resultant wind speed and I direction. This method of analyzing wind data is more complex and in-volves taking the vectorial sum of wind directions and speeds divided by the number of observations. In order to make comparisons on this new basis, it was necessary to convert NBSR wind data into daily resultant directions and resultant speeds comparable to WNA data.

• NBSR 7C 19-2 I

When correlation coefficients were finally claculated, it was found that r = 0.961 for the resultant direction and r = 0.860 for X1& Xi4 the resultant speed. The daily average wind direction during April 1962, also yielded a very significant r = 0.963. XI Xe It is now clear that each of the five weather parameters compared between NBSR and WNA have repeatedly shown high correlatien coefficients. One additional task, however, was undertaken in this recent study. Due to its importance to wind dispersion, it was desirable to make another f requency distribution of wind direction ranges over fifteen minute intervals. Figure 19.2 shows such a frequency plot of wind ranges taken hourly for the month of April 1966. When I this is compared to the related frequency plot for April 1962*, it is clear that they are nearly identical. It should be observed that the mean range of each plot are within two degrees of each other. I I I I I I I I

• NBSR - 7C, page 25 I

19-3

E A F DAILY AVERAGE BAROMETER l APRIL 1966 I l 770 - I a I E 760 I l e x g 750 R lI i m $ 740

I s2 Qz g

l I I f 730 740 750 760 770 GAITHERSBURG READINGS (mm. Hg) I I FIGURE 19.1 I

m M M M M M M M M W W M M m M W M M 120 - ^ '= 81 5 100-DIREC oN RANGE oV R 15-MIN. IN ERVALS 80-APRIL 1966 60-40-20-E ~ ~^ ~~~ di A ~' o 2 40 80 80 ioo iao 4 e e o RANGE (*) FIGURE 19.2 l.

I QUESTION 20 Update your population estimates to include latest zoning changes and high rise building construction. Calculate potential doses to the nearest present or planned high rise buildings giving considera-tion to the ventilation stack height being less than the height of these buildings.

RESPONSE

A tabulation of the most recently reported notices of change of zoning are given in Table 20.1 including a legend of abreviations utilized in the table. In response to the question concerning dose calculations to persons in the nearest present or planned high rise building, it should be noted that the nearest such building is one that is planned for location at a distance of approximately one mile to the WNW direction f rom the reactor site. This distance is greater than that for which dose calculations have been made and reviewed in Question 4 of this submission. Reference should also be made to the record where this question has also been discussed. (See for example page 15 of the attachment to the letter dated August 6, 1963 entitled "AEC Staff Analysis of Population Growth and Hazards in Connection with the NBS Reactor Site.") I I I I I I 20-1

L Table 20.1 3 Distance from N8SR Site Direction Zoning Change Area Application No. Resolution No. (mi) from Site From To (acres) C-1287 5-1401 1.2 NtM R-R C-2 8.5 g C-1288 5-1402 1.3 N!M R-R R-20 26.2 C-1289 5-1403 0.9 PM R-R C-P 44.9 I C-1290 5-1404 1.5 fM R-R R-150 79.6 I C-1291 5-1405 1.1 WrM R-R R-20 16.9 C-1292 5-1406 0.8 W R-R R-T 9.6 C-1294 5-1407 1.0 WIM R-R R-11 16.5 C-1293 5-1411 1.3 W R-R I-3 47.4 C-1435 2.5 E R-R I-1 3.9 C-1455 2.9 NE R-R R-T 4.0 l C-1480 3.0 E R-R I-1 29.2 C-1481 3.0 SE R-R I-1 24.6 C-1513 2.0 SE R-R I-1 1.6 1 I E-218 1.3 IM R-R I-1 30.6 E-221 2.0 SW R-R C-1 5.6 E-222 2.0 SW R-R C-1 1.7 E-228 2.5 E R-R I-1 1.1 E-250 3.0 E R-R I-1 .6 E-275 3.0 E R-R R-30 6.8 E-283 2.0 SE R-R I-1 1.1 E-327 2.3 NE R-R T-S 1767.3 1 E-419 1.2 SW R-R C-P 56.0 E-720 1.2 SW R-R C-P 36.2 E-721 1.2 SW R-R C-P 6.2 E-730 2.0 SE R-R I-1 1.8 E-731 2.0 SE R-R I-1 1.9 I I 20-2

CODE LEGEND R-ll Multiple - Family liigh Rise Residential 1,000 square feet per dwelling unit P-T Town llouses 3,500 square feet per dwelling unit ~ R-R Rural Residential 20,000 square feet per dwelling unit R-10 Multiplc - Family liigh-Density Residential i 1,000 square feet per dwelling unit 7 R-20 Multiple - Family Medium-Density Residential 2,000 square feet per dwelling unit R-30 Multiple - Family Low-Density Residential 3,000 square feet per dwelling unit i R-150 One-Family Controlled Density 15,000 square feet per dwelling unit I-1 Light Industrial I-2 lleavy Industrial I-3 Industrial Park C-1 Local Commercial C-2 General Commercial I C-P Commereial Office Park l T-S Town Sector 1,500 Acres minimum area. Maximum density of 15 persons per acre. I I g 2e->

I ~ QUESTION 21 Provide the minimum qualifications and experience for each of the 1, engineering and supervisory positions on the operating staff (i.e., shift supervisors and above) and for the health physics staff. Provide resumes of experience for each individual to be initially aasigned to these positions.

RESPONSE

1. MINIMUM QUALIFICATIONS AND EXPERIENCE FOR OPERATING STAFF I The aninamum qualifications and experience of the operations staf f supervisor are given below and resumes of those initially I assigned are attached. 't.1 Chief Nuclear Engineer er The Chief Nu:. lear Engineer should be a graduate engineer and have professional experience in the following areas: (a) Minimum of four years in reactor operations, having held a Senior Operators' license or its equivalent. (b) Demonstrated ability in nuclear and general engineering. (c) Experienced in plant maintenance and large equipment j performance. (d) Minimum of one year experience in plant management or administration; commensurate with the position. I (e) Experience in the development and safety review of pro-cedures, experiments, technical specifications, safety analysis reports, and other documents associated with reactor operations. (f) Compliance requirements of applicable AEC regulations, to include administration, knowledge of required records and other associated " mechanics" involved with such regulatlons. I 21-1 I

I 1,2 Deputy Chief Nuclear Engineer The Deputy Chief Nuclear Engineer should be a graduate I engineer and have professional experience in the following areas: (a) Minimum of three years in reactor operation, having held a Senior Operator's license or its equivalent. (b) Demonstrated ability in nuclear and general engineering. 4 (c) Familiar with plant maintenance and operation of large plant equipment. (d) Experience in the development of procedures and other documents associated with reactor operations. 1.3 Reactor Supervisors The Reactor Supervisors should have completed some form of formal training in reactor technology and reactor operations. They shall have had operational experience in the following areas: (a) Minimum of three years in reactor operations, having held a Reactor Operators' license or its equivalent and be capable of obtaining a Senior Operators' license for the NBSks (b) Experience in the operation and maintenance of large plant equipment. (c) Some experience in the supervision of technicians and/or reactor operators. 2. MINIMUM QUALIFICATIONS AND EXPERIENCE FOR HEALTH PHYSICS STAFF The minimum qualifications and experience of the health physics staff supervisors are given below and resumes of those initially assigned are attached. I 21-2 I i

L r a 2,1 Chief. Health Physics Section U The Chief of the Health Physics Section should be certified by the American Board of Health Physics, In order to be certified he must meet the following requirements: (a) Must have a Bachelor's degree in a physical science or a biological science with a. minor in physical sciences. (b) Must be at least 28 years old and have had at least 6 years of responsible professional experience in Health Physics. (c) Must be capable of making a satisfactory evaluation of the following installations or operations: (1) Radiographic Installation - Industrial or Medical (2) Fluoroscopic Installation l (3) Therapy Installation (4) Radioisotope Laboratory (5) Air end Nater Sampling and Environmental Survey (6) A Radioisotope Chemical Separation Plant l 1 (7) Reactor (8) Major Decontamination Operation (9) Accelerator I 2.2 Supervisory Health Physicist, (Reactor)_ I The Supervisory Health Physicist for the NBSR should have a Bachelor's degree in a physical science or a biological science with a minor in a physical science and have a minimum of three years professional experience in health physics, of which additional education may be substituted for experience up to one year. The experience should be in the following areas: I 21-3 1

(a) Development and safety review of procedures and experiments. { (b) Ccmpliance requirements of applicable AEC regulations, to include administration, knowledge of required records, and other associated " mechanics" involved with such regulations. (c) Personnel and environmental radiation monitoring methods. i (d) Application of protective clothing and associated l equipment. l (e) Calibration and evaluation of fixed and portable radiatic,a instrumentation. (f) Radioactive source and waste handling. 5 l I. t I ,I I I I I I 21 4

I 3. RESUMES I 3.1 LARRY EUGENE SMITH, Chief Nuclear Engineer I EDUCATION: N. C. State University (B. S. Nuclear Engineering) 6/60 University of S. C. (worked on M. S. Nuclear Engineering) 2/62 - 6/62 Ca tholic University of America (worked on M. S. Nuclear Engineering 2/65 - 2/66

• total of 12 semester hours earned i

NBSR Training Program (125 class hours) 5/66 - 8/66

• also included in this was a 2 day trip to N. C. State University to I

operate their reactor. Carolina-Virginia Nuclear Power Association Training Program

• AEC license OP-1323 received 10/62 and SOP-179 received 8/64 Materials Testing Reactor Training Program
• Qualified Reactor Engineer 3/61 N. C. Sta te University Reactor Training Program
• AEC license OP-789 received 9/60 EXPERIENCE:

(10/65 - Present) Chief Nuclear Engineer, National Bureau of Standards a. Directing operation and maintenance of the NBSR Facility b. Collaboration with AEC-DRL personnel in facility license matters c. Recruiting and training new operating personnel I d. Write technical specifications and opera ting procedures. (11/64 - 10/65) Deputy Chief Nuclear Engineer, National Bureau of S tandards a. Assumed responsibilities of the Chief Nuclear Engineer during his

absence, b.

Wrote plant opera ting procedures c. Scheduling of plant testing program d. Directed the initial testing of operation of facility process i systems. I 21-5

I LARRY EUGENE SMITH Pa ge 2 (9/63 - 11/64) Assistant Operations Engineer, Carolina-Virginia Nuclear Power Association a. Coordinated plant operations planning and scheduling of tests. b. Conducted reactor opera tor training program, c. Directed startup test program, d. Technical and safety review of all test and operating procedures. I (3/63 - 9/63) Shif t Srneyvisor, Carolina-Virginia Nuclear Power Association a. Supervised activities of reactor technicians and other operating personnel, b. Responsible for all nuclear plant operations on assigned shif t. c. Supervised fuel loading and preparation for initial approach to criticality on initial plant startup. (6/61 - 3/63) Shif t Engineer, Carolina-Virginia Nuclear Power Association a. Conducted training program on plant systems for operating personnel, I b. Assisted in preliminary checkout of plant systems, c. Wrote initial startup procedures, d. Assisted in preparation of plant operating procedures. (9/60 - 6/61) CVTR Shift Engineer, Working for CVNPA at Materials Testing Reactor a. Operation of 40 Megawatt Materials Testing Reactor, b. Opera tion of reactor in-pile loop experiments. c. In-core experiment changes, I d. Fuel additions and rota tions in core. (6/60 - 9/60) CVTR Shif t Engineer, Working for CVNPA at N. C. State University Reactor a. Trained for and obtained AEC license for N. C. State reactor, b. Assisted in the initial flux mapping and power calibration. I 21-6

l e l 3.2 JAMES FRANKLIN TORRENCE, Deputy Chief Nuclear Engineer I EDUCATION: 5 N. C. State University (B. S. Mechanical Engineering) 6/61

• Also audited all course work necessary for B. S. Nuclear Engineering.

P NBSR Training Program (125 class hours) 5/66 - 8/66

• Also included in this was a 2 day trip to N. C. State University to operate their reactor.

L N. C. State Reactor Training Program 8/63 - 11/63

• AEC license SOP-430 received 2/64

" Fundamentals of Operating Nuclear Test Reactors" at Materials testing Reactor (400 class hours) 3/62 - 5/62

• Qualified Reactor Engineer 8/62 EXPERIENCE:

(4/66 - Present) Deputy Chief Nuclear Engineer, National Bureau of Standards a. Assume responsibilities of the Chief Nuclear Engineer during his absence. b. Set up and conducted reactor opera tor training program. { c. Write startup test procedures and supervise the conduction of these

tests, d.

Scheduling of plant testing and experimental programs. B (3/65 - 4/66) Reactor Systems Engineer, N. C. State University a. Responsible for inhouse planning, design and specifications of thermal I and mechanical systems for a new 2 megawatt research reactor to be built a t N. C. S ta te. b. Responsible for review and acceptance of 30,000 curie Co-60 Gamma Irradia tion Facility. 21-7 I

JAMES FRANKLIN TORRENCE Page 2 (10/62 - 3/65) Reactor Opera tions Engineer, N. C. S ta te University a. Immediate supervision of reactor operating staff, b. Prepared technical specifications for modification of reactor to 500 KW. c. Supervised neutron activa tion analysis program. d. Acting Reactor Supervisor (10/63 - 7/64) (1) Administrative supervision of the Reactor Facility. (2) Certification to the AEC of competence and safety of all licensed ope ra tors. e. Wrote original technical specifications for the 10 KW reactor. I l (6/61 - 10/62) MTR Reactor Engineer, Phillips Petroleum Company a. Operation of 40 megawatt Materials Testing Reactor b. Opera tion of reactor in-pile loop experiments. c. In-core experiment changes g W d. Fuel additions and rotations in core. l (12/59 - 6/61) Reactor Technician (Part-Time 20 hr/wk) N. C. Sta te University a. Construction and installation of reactor control instrumentation. j b. Assis ted in the construction and installation of the 10 KW Heterogeneous Research Reactor and its associated auxiliary equipment, c. Performed the thermal power calibration of the reactor. I 1 B 1 2,-.

l 33 ARTHUR LEWIS CHAPMAN, Reactor Supervisor (Instrument and Electrical) EDUCATION: Shortridge High School, Indianapolis, Ind. 1944 i NBSR Training Program (125 class hours) 5/66 - 8/66

• Also included in this was a 2 day trip to N. C. State University to operate their reactor.

AFRRI TRIGA Mark F Training Program 8/63 - 12/63

• AEC license OP-1684 received 2/64.

U. S. Navy Nuclear Power Training Unit 12/53 - 7/55 L

• Navy Reactor Operator's Certificate #85 (U.S.S. Nautilus)

Westinghouse Atomic Power Division School 7/53 - 12/53 U. S. Navy Submarine Electrical School 4/51 - 5/51 U. S. Navy Submarine School 2/51 - 4/51 U. S. Navy Cyro Compass School 2/48 - 7/48 U. S. Navy Basic Engineering School 11/44 - 1/45 EXPERIENCE: (5/66 - Present) Reactor Supervisor (Instrument and Electrical), National Bureau of Standards a. Direct shif t personnel in the operation of NBSR and its associated L equipment. b. Supervision of all changes in the reactivity of the reactor, whether in the form of experiments or fuel addition. c. Assisted in the final installation and check-out of the reactor l control instrumenta tion. (9/65 - 5/66) Reactor Technician, National Bureau of Standards a. Assisted in plant electrical system check-out. (8/63 - 8/65) Nuclear Reactor Operator, Armed Forces Radiobiology Research Institute (U. S. Air Force) a. Operated TRIGA Mark F pulse reactor. 21-9

ARTRUR LEWIS CHAPMAN Page 2 b. Assisted in design and construed on of an improved reactor control system console. c. Assisted senior scientists in utilization of the pulse reactor. (5/57 - 8/63) Nuclear Reactor Operator, Wright-Pa tterson Air Force Base (U. S. Air Force) a. Assisted in construction and installa tion of Air Force Test Reactor, b. Supervised fuel loading team in first dummy core loading, c. Wrote operating procedures and assisted in training of new men. (7 /55 - 4/57) Reactor Opera tor, U.S.S. Nautilus a. Was eleventh man in U. S. Navy to be designated a " Reactor Operator". b. Supervisor responsible for pre-critical check-out of reactor systems before startup. c. Trained new men. 21-10 J

I I .4 3 JOSEPH MABRY MC SPADDEN, Reactor Supervisor (Nuclear) EDUCATION: Fairfax High School, Fairfax, Virginia 6/56 NBSR Training Program (125 class hours) 5/66 - 8/66

• Also included in this was a 2 day trip to N. C. State University to opera te their reactor.

Elk River Reactor Training Program

• AEC license OP-1467 received 4/63 I

U. S. Navy Electronic Counter-Measures School 10/61 U. S. Navy Nuclear Power Training Unit (S3G) 6/60 U. S. Navy Nuclear Power School 9/59 U. S. Navy Advanced Sonar Electronics 11/57 U. S. Navy Submarint; School 11/57 U. S. Navy Electronic Technician School 8/57

• Reactor operator lice.nse #416 for S3G received 1/60 EXPERIENCE:

(11/64 - Present) Reactor Supervisor (Nuclear), Na tional Bureau of Standards a. Direct shif t personnel in the operation of NBSR and its associated equipment, I b. Supervision of all changes in the reactivity of the reactor, whether in the form of experiments or fuel additien. c. Assisted in the construction of the reactor facility. (10/62 - 11/64) Operations Engineer (4 - 5' 7 ' River), Allis-Chalmers Mfg. Company - - Elk River, Minn, a. Shift supervisor responsible for sate operation of Elk River Boiling Wa ter Reactor. b. Assisted in startup test program and calibrations, I c. Received AEC license OP-1467 4/63 5 21-11 I

I JOSEPH MABRY MC SPADDEN I Page 2 (9/62 - 10/62) Opera tions Engineer (a t APPR-1), Allis-Chalmers Mfg. Company - - Schenec tady, N. Y, s. Operated critical facility reactor (APPR-1) and loaded reactor cores, b. Initiated operating procedures. c. Assisted in writing crew training manual for PM-3B Pressurized Wa ter Reactor. I (6/60-8/62) Electronic Technician, U. S. S. Skipjack (S S (N) 585) a. Operated S5W Pressurized Water Reactor, I b. Trained new reactor operators. (9/59 - 6/60) Electronic Technician, U. S. Navy Nuclear Power Training Unit. a. Operated S3G Pressurized Water Reactor. 2. Assisted G.E. personnel in plant testing and core physics tests. I I I I I I I I E u-u

l 3.5 RAY AUGUST MESCHKE, Reactor Supervisor (Chemical) EDUCATION: Hirsch High School, Chicago, Ill. 1935 Roosevelt University, Chicago, Ill. 1946 to 1951

• About 90 semester hours earned NBSR Training Program (125 class hours) 5/66 - 8/66 Argonne National Lab on-job training in reactor technology 1957 - 1960 I

EXPERIENCE: (7/63 - Present) Reactor Supervisor (Chemical), National Bureau of S tandards a. Direct shif t personnel in the operation of NBSR and its associated equipment. b. Supervision of all changes in the reactivity of the reactor, whether in the form of experiments or fuel addition. c. Establish the facility chemistry program and controlling process systems chemistry. d. Assisted in the construction of the reactor facility. (9/46 - 7/53) Assistant to Reactor Facility Supervisor, Argonne National Laboratory a. Assisted in developing a process for extraction of Actinium from irradiated slugs of radium, b. Assisted in installation and testing of Mass Spectrometer at CP-5 and worked in D 0 quality control. 2 c. Coordinated reactor operation scheduling with experimental programs. d. Acted as liaison between Reactor Operations Division and groups using the reactor for research. e. Instructor in the International School of Nuclear Science and Engineering. i 21-13

3.6 ROBERT WAYNE WHITEHOUSE, Reactor Supervisor Odechanical) EDUCATION: USAFI (High School) U. S. Navy 1957 L NBSR Training Program (125 class hours) 5/66 - 8/66

• Also included in this was a 2 day trip to N. C. State University to operate their reactor.

L Elk River Reactor Training Program

• AEC license OP-1443 received 4/63 F

Carolina-Virginia Nuclear Power Association Training Program 9/61 - 12/61 U. S. Navy Nuclear Power Training Unit (S4G) 1958 l U. S., Navy Nuclear Power School 1957 U. S. Navy Submarine School

• Reactor operator license #68 for S4G received 11/58 EXPERIENCE:

(6/64 - Present) Reactor Supervisor (Mechanical), National Bureau of Standards a. Direct shif t personnel in the operation of NBSR and its associated equipment. L b. Supervision of all changes in the reactivity of the reactor, whether in the form of experiments or fuel addition. c. Assisted in the construction of the reactor facility. (10/62 - 6/64) Operations Engineer (at Elk River), Allis-Chalmers Mfg. Company - - Elk River, Minn. a. Shif t supervisor responsible for safe operation of Elk River Boiling Water Reactor. L b. Assisted in startup test program and calibra tions, c. Received AEC license OP-1443 4/63. l 21-14 l

l ROBERT WAYNE WHITEHOUSE Page 2 (9/62 - 10/62) Operations Engineer (a t APPR-1), Allis-Chalmers Mfg. Company - - Schenectady, N. Y. a. Operated critical facility reactor (APPR-1) l b. Assisted in writing training manual and test procedures. l (9/61 - 8/62) Reactor Technician, Carolina-Virginia Nuclear Power As socia tion j a. Assisted in writing training manual and pre-operational test procedures. b. Assisted in pre-operational test prior to fuel loading. (2/61 - 8/61) Engineman, Firs t Class Petty Officer, U.S.S. Skate (SSN578) a. Participated in reactor refueling. b. Overhauled and maintained all primary and secondary systems. l l (7/58 - 2/61) Engineman, First Class Petty Officer, U.S.S. Triton (SSRN586) a. Overhauled, maintained, and tested reactor systems and steam systems. b. Assisted in writing operating procedures and training new men. I I I I I 'I 21-15 5

I 3.7 BILLY JOE YOUNG, Reactor Supervisor (Nuclear) I EDUCATION: McCurtain High School, McCurtain, Okla. 6/42 I NBSR Training Program (125 class hours) 5/66 - 8/66

• Also included in this was a 2 day trip to N. C. State University to opera te their reactor.

Carolina-Virginia Nuclear Power Association Training Program

• AEC license OP-1498 received 5/63.

U. S. Navy S5W Reactor School 1/59 - 2/59 U. S. Navy Nuclear Power Training Unit (S3G) 7/58 - 1/59 U. E. Navy Nuclear Power School 1/58 - 7/58 I U. S. Navy Electronic School 6/53 - 6/54

• Reactor operator license #288 for S3G received 1/59 I

EXPERIENCE: (2/65 - Present) Reactor Supervisor (Nuclear), National Bureau of Standards a. Direct shif t personnel in the operation of NBSR and its associated equipment. b. Supervision of all changes in the reactivity of the reactor, whether in the form of experiments or fuel addition. (7/62 - 2/65) Reactor Technician, Carolina-Virginia Nuclear Power Association a. Opera ted reactor and associa ted heavy wa ter sys tems. l g W b. Assis ted in plant check-out, original fuel loading, initial criticality, and reactor calibra tions, c. Received AEC license OP-1498 5/63. I I I 21-1e l I

BILLY JOE YOUNG Pa ge 2 l (2/59 - 7/62) Chief Electronics Technician, U.S.S. Abraham Lincoln J a. In charge of Reactor Control and Instrumentation Section. b. Wrote operating procedures and trained new men. I I 1 I 1 5 1 3 5 lI I E u.17 I

il 3.8 ABRAllAM SCINEBEL, Chief, Ilealth Physics Section it EDUCATION: University of Maryland (Ph.D. Organic & Physical Science) 9/49 - 6/51 Chicago University 6/44 - 12/44 Virginia Polytechnic Institute 6/43 - 1/44 I Columbia University 1/42 - 6/42 New York University (M. S. Chemistry) 6/36 - 6/38 Brooklyn College (B. S. Chemistry) 6/31 - 6/35 Teaching experience in Radiochemis try a t George Washin?, ton University, American University and NBS Graduate School. I EXPERIENCE: (3/46 - Present) Chief, llealth Physics Section, National Bureau of S tandards I (4/45 - 2/46) Research Technician, Fercleve Corp., Oak Ridge, Tenn.- Assigned by U. S. Army I (1/44 - 12/44) Jr. Chemist, Metallurgical Lab., University of Chicago - Assigned by U. S. Army (1/41 - 1/43) IIead Chemis t, Wa ter Service Lab. Inc., New York, N.Y. I I I I I I

I 3.9 HUCH EDWARD DESPAIN, Supervisory Health Physicist (Reactor) EDUCATION: I University of California (M. S. Bioradiology) 1964 i Memphis S ta te University (B. S. Ma thema tics') 1962 NBS Training Programs t (5/22/66 - 5/27/66) Observed meth:ds for monitoring tritium in reactors and associa ted facilities and opera tion of a heavy wa ter modera ted reactor. The reactors observed were at Georgia Tech., Carolinas Virginia Tube Reactor, and the Savannah River Plant. I (7/13/65) Observed health physics procedures at the MIT Reactor, Cambridgq Mass. (1964 - 1965) Visited the Naval Research Laboratory and the Armed Forces Radiobiology Research Institute. (12/6/64 - 12/11/64) Health and Safety Workshop U. S. Atomic Energy Commission, New York, N. Y. I (10/3/64 - 10/21/64) Training to monitor the CP-5 type reactor, Argonne, Ill. I University of California 9/62 - 8/64

• AEC Special Fellowship in Health Physics EXPERIENCE:

(3/64 - Present) Physicist in Charge of Health Physics for NBS Reactor during design and construction phase, National Bureau of Standards (6/62 - 9/60) Sunner employment, Na tional Bureau of Standards

• Assisted in designing and testing of neutron spectrometers.

I Prior to these dates he attended under:;raduate school and served in the U. S. A. F. as a Radar Operator. I 21-19

I " 10 THOMAS G. HOBBS, Superviaory Health Physicist (Reactor) EDUCATION: Graduate courses at American University and University of Maryland (Physics) 9/59 - Present I Vanderbilt University, Vashville, Tenn. 9/49 - 8/59

• AEC Special Fellowship in Health Physics Lincoln Memorial University, Harrogate, Tenn. (B. S. Math) 6/55 - 5/58 EXPERIENCE:

(9/59 - Present) Supervisor Health Physicist at the NBS Linear Accelerator, National Bureau of Standards I I I I I I I I I I 21-20 I

I I QUESTION 22. Provide a discussion of the planned experiment program, including an analysis of the potential consequences of accidents involving planned experiments and experiment facilities, and the identifi-I cation of an envelope of safety conditions within which the program will be conducted. Where applicable, at least the following general types of considerations should be analyzed: a. Radioactive inventory b. Potential reactivity effects c. Tempe ratu re d. Pre s su re e. Chemical reaction effects f. Explosive potential I g. Mechanical integrity h. Control and instrument requirements 1. Administrative controls 1. INTRODUCTION The experimental facilities and service available at the NBSR are described in Section 8 of Final Safety Analysis Rc-port of the NBSR (NBSR 9), and the types of experimental pro-I grams are briefly discussed in sub-section 3.1.1.2 of the same report. The purpose of the present section is to discuss more fully the types of experiments anticipated and to estab-lish the criteria that each experiment must meet to assure the safety of the reactor and personnel working in the area. Since the very nature of basic research prevents the forecasting of detailed experimental programs, no attempt will be made to pre-sent an all inclusive listing of anticipated experiments and procedures, but, ra the r, the emphasis will be on the estab. I lishment of an experimental envelope and the criteria to be used in judging the safety of exnerimental installation and procedures. I 22 1

The discussion will be divided according to the several types of experimental facilities that are available, preceded F l by general criteria applicable to all types of experiments. s 2. GENERALLY APPLICABLE CRITERIA e 2.1 Reactivity Limits. Any creditable failure of an exper- + imental installation or component shall not cause a rapid reactivity increase greater than that which can be controlled by the regulation roo, (0.57 ap). Wherever correlation be-tween failures in two or more experiments is creditable, the correlated failures shall not lead to a total rapid reactivity increase greater than 0.57. ap. 2.2 Experiment Failure. No creditable failure to an experi-mental component shall endanger personnel or cause any damage to any reactor system component or interfere with its proper operation. In particular, the results of loss of power shall be carefully analyzed and shown not to endanger the safety or operation of the reactor. 2.3 Explosives. No combination of explosive ingredients which if exploded would rupture the primary experimental container will be permitted. Furthermore, a second barrier between the reactor core and the explosive must be provided which is capable of withstanding any creditable rupture of the primary container. 3. BEAM TUBE EXPERIMENTS 3.1 Introduction. The great majority of these experiments will simply use a neutron beam extracted through collimators in the beam hole. Externally, the interaction of beams with materials will be studied. All the components of such systems will be external to the inner face of the reactor thermal shield and therefore have no ef fect on reactor reactivity. In some I cases supplemental shutters or experiments will be placed in the beam tubes within the thermal shield. These two categories are discussed separately below. 22-2

I 3.2 External Beam Experiments. Physically no component of I these experiments extends within the inner surface of the thermal shield and the reactivity effects are negligible. 3.2.1 Typical Experiments. The experiments described be-low are typical of the type currently planned. The list makes I no attempt to be all inclusive, and the omission of any parti-cular experimental setup is not intended to exclude it from future installation. Neutron Diffraction: A beam of about 1" x 2" is defined by collimators placed in the beam hole external to the shutter. A rotating plug may allow the choice of several collimators and a filtering material may be placed in the beam. The beam is then scattered from a carefully shielded monochromating crystal which produces a monochromatic beam of neutrons which is in I turn scattered from a sample. The scattered neutron intensity is measured as a function of angle to investigate the structural properties of the sample. Time-o f-F light : The beam extracted from the reactor is chopped by a rotating neutron absorbing disc which contains slits which allow short bursts of neutrons to pass through. These are scattered by the sample and the time required for the I neutrons to reach the detectors is measured to determine their e ne rgy. I Fission Physics: A small sample of fissionable material (<100 gm) is placed in the external beam and the distribution in the mass of fission fragments is determined using energy sensitive detectors. 3.2.2 Safety Criteria. Since these experiments have no reactivity worth, the areas of prime concern are physical interactions with the reactor system and proper radiation shiciding. Thus all plugs, filters, cryogenic systems, etc. inserted into the beam tube must be reviewed to assure that they are not heated to the point where their physical in-tegrity is impaired and that they in no way damage the stain-I less steel liner of the beam tube. This includes assurance 22-3

I that objects inserted into the beam holes cannot become stuck. Similarly any systems inserted into the beam holes such as cryostats, pressurized gases, etc. which might be subject to the sudden release of stored energy must be such as to reduce the effect of such an accident to the point where no part of the beam tube could be damaged. Rotating equipment such as neutron choppers 'which can contain large quantities of stored mechanical energy will be carefully analyzed. Extensive tests of rotating equipment, including tests at speeds exceeding the maximum operational speed, will be performed to prove out the I system design before installation is permitted. If engineering analysis and preoperational tests indicate that there exists a creditable although remote possibility that a rotor may fail in such a way as to breach the rotor housing, external mechanical shielding will be provided to protect personnel and reactor components. The external beams constitute an intense source of radiation even to the extent that air scattering of the beam presents a significant shiciding problem. In general, the neutron shield-I ing requirements of the experiments themselves are more stringent than biological radiation protection standards; therefore, every experiment will be thoroughly studied for adequate shielding. The minimum requirements, however, will be that the radiation in all readily accessible areas, not specifically designated by the Health Physics section as high radiation areas, must be below 2.5 mrem /hr. The beam shielding problem has been dis-cussed more thoroughly in Item 10. Currently, five experiments of the external beam type are I firmly planned and one more is included in near future programs. Five of these are neutron diffraction experiments and one is a polarized neutron nuclear physics experiment. Each of these will utilize a shield plug-collimator combination in the beam port and at least one will include a liquid nitrogen cooled filter. All of these assemblies will be external to the shutter. I

The total radiant power entering the region external to the shutter is only 26 watts for direct y-radiation and 54 watts for thermal neutrons absorbed in iron. The neutron capture energy may be reduced by about a factor of 3 by the use of boral. Thus, no special cooling of the collimator plugs is necessary. l None of these experiments have any reactivity effect and the only radioactive inventory involved is the activation of I the inner ends of the collimator plugs. An electric power fail-ure would cause the shutdown of the experiments, but is completely l unrelated to the equipment inserted into the reactor shield. Thus, an electric power failure creates no hazard to reactor l components or operation. f 3.3 Internal Beam Tube Experiments. These types of experiments must satisfy any applicable external beam requirements, however, a closer look at their relation to the reactor core and vessel is necessary. 3 3.1 Typical Experiments. Typical of components or experiments which may be installed within the thermal shield are: (1) Water tanks which can be filled or drained periodically ,I to act as neutron beam shutters. (2) Loop type experiments in which samples are placed in a controlled environment, cryogenically cooled, for example, and instrumented. (3) Moderators may be placed in the end of beam tubes to change the neutron beam spectrum. Typical moderators wc uld be D 0 or graphite. Both would require adequate 2 cooling. (4) Pneumatic or hydraulic tube systems for the irradiation I of samples. (5) A scattering sample placed in the center of one of the through tubes to act as a neutron source for an experiment requiring unusually low background. (6) Similarly, a capture y-ray source placed in a through tube. i = 22-5 I

I To illustrate the problems and their solutions associated with internal beam tube experiments, three specific cases will be examined followed by a discussion of general criteria to be followed in evaluating these types of experiments. Water Tank Shutters: Referring to rigure 8.1 of NBSR 9, a water tank would be placed in the region between the shutter I and the reactor core. Tubing to circulate and drain such a shutter tank are shown in phantom in the figure. The shutter uses demineralized H O from the experimental demineralized 2 water cooling system. During normal use the water is drained and the volume filled with helium or CO. When beam atten-2 uation is desired, the tank is filled and the water circulated. The length of the water tank is controlled by several considerations: experimental requirements, reactivity worth, and heating of the empty tank. The tank will be made no longer I than sufficient to supply the desired shielding. Since it is planned to fill and drain this tank, the re-8 activity worth of the water volume will be kept below 1/47. Ap, a value consistent with maximum pneumatic tube insertion. Re-ferring to Table 4.6-7 of NBSR 9, the total worth of a flooded beam port is less than 0.17. Ap. Thus, if D 0 were used, this 2 would place no limit on the volume of the tank. The negative ef fect of 110 placed in a beam tube close to the core was cal-2 culated to be.147. Ap using the perturbation results of the I EQUIPOISE 3A code. This is less than the 1/47. Ap criterian. The actual reactivity worth will, of course, be measured to I assure compliance with the 1/ 47. Ap. So no limit is placed on tank size by the reactivity worth of either H O or D 0. 2 2 The radiation heating of an aluminum tank is calculated to range from 1/2 w/gm at the core face to.025 w/gm at the reactor vessel edge. Thus, the heating in a 1 cm thick end plate would vary from 240 watts (1.35 w/cm ) at the core face 2 to 11 watts (0.062 w/cm ) at the vessel edge. These rates are sufficient to make convection cooling through the CO 2 conduction to support points inadequate so continuous cooling I. would be required. This would be supplied by cooling tubes 22-6

I around the tank and supplied from the experimental demineralized water cooling system. Failure of this cooling system would nec-essitate reactor shutdown. The cooling flow or cooling outlet temperature or both will be monitored and high temperature or I low flow will alarm in the control room. Automatic rundown will not be required because the.5 w/gm heating rate at the hottest spot would still require 16.7 minutes for the aluminum to reach me lting temperature. This will allow the operator ample time to check instrumentation and take any necessary precautions in-cluding shutting down the reactor. As an added safety feature, it should be pointed out that an aluminum tank with I cm thick walls which penetrates the vessel no more than 14 cm has a max-2 imum heat aroduction rate of no more than.13 w/cm of outside surface. An aluminum temperature of 980 F would radiate heat at this rate and so require no additional cooling in an emer-gency. Of course thinner walls would present less of a problem. Moderator Tanks: Such a tank would normally use D 0 to 2 improve the thermal to fast ratio or to define a smaller beam thereby increasing the intensity. These tanks will always have D 0 circulating in them for cooling. Wherever desirable, they 7 will be supplied with emergency auxiliary cooling coils. In general they will be examined in the same way as the water shutter I tanks and indeed they are a sub-class of such systems. Pneumatic Tube: Pneumatic tubes placed in radial beam ports will be similar to those already described in sub-section 8.1.6 of NBSR 9. The reactivity worth of a sample would be limited to less than 0.1% ap as discussed for the existing system in sub-section 4.6.5 of NBSR 9. Its use would be co-ordinated with all other pneumatic systems to limit the total worth of pneumatic or hydraulic system samples inserted simul-taneously to.257.. For very short sample irradiations, a pneu-I matic system might be inserted in one of the througl. beam tubes. Its sample reactivity worths would be coordinated as discussed above. These tubes will be cooled as necessary and the same general conditions for monitoring the cooling and taking action in case of failure would apply as for the water tanks discussed I

above. 3.3.3 General Criteria. The discussion in the previous sub-section indicates that the c.ajor problem with the insertion of samples within the region defined by the thermal shield is one of heating. Thus most such experiments must be cooled. The only exception being those in which the heating rate is low enough to allow adequate cooling by conduction, by convection in the CO atmosphere in the region, or by thermal radiation. 2 Normally these experiments would be cooled by the experi-mental demineralized water system or the D 0 system. Each of I 2 these systems is powered by two pumps one of which is normally a standby. The D 0 pumps are supplied with emergency power, 2 I and the system has emergency cooling provisions as discussed in sub-section 7.2 4 of NBSR 9. Although failure of these systems is very unlikely, failure in individual experimental system plumbing must be allowed for. Thus, wherever loss of coolant might cause structural failure of an experimental system in the reactor that could lead to damage to any reactor component, the outlet coolant would be monitored for flow, temperature or both and connected to the annunciator panel in the control room. In most cases, as shown for aluminum, the complete loss of all cooling of any type still allows adequate time for operator action and would not require an automatic reactor rundown. However, provision is provided at each beam hole to generate an automatic reactor scram signal whenever the type of possible failur warrants it. A shutdown of the reactor does not, of course, remove all heat sources, but they are greatly reduced. Calculations based I on the decay gamma heating rate compared to the operating gamma heating rate show, for example, that if 5 minutes of full power operation under no cooling conditions is required to produce structural damage, then at least 45 minutes is required under shutdown conditions. By this time the heating rate has dropped more than a factor of 10 and natural convection, thermal radiation, or conduction may be sufficient to prevent damage. Each I

1 I experiment is analyzed from this point of view and em2rgency provisions provided for shutdoun cooling if required. I The potentfal reactivity effects of beam tube experiments have been shown to be small, and therefore no difficulty is en-I countered in meeting the criteria of <.257. op for relatively slow changes (e.g., draining a water tank). The reactivity that could be introduced by failure of semi-pemanent experiments 9 (those that can't be removed during operation) will be included in the 2.6% op total allowed for experiments (see sub-section 4.6.5 of NBSR 9. The radioactive inventory of these types of experiments will be almost entirely in the form of activation of the structural material. Removal of most internal experiments will require with-drawal mto a shielded transfer flask and temporary storage in the wait plug' storage facility or the storage pool. The development of significant pressures in most experiments is unlikely. Wherever significant pressures are a possibility, 3 tress analysis and, in some cases (based on engineering judgment), experimental evidence will be required to show that the experimental container will not rupture. Furthermore, the potential results of a rupture or leak will be investigated to eliminate danger to the reactor vessel. I All experiments will be examined for the possibility of significant chemical reactions. Most experiments will not in-volve strong chemicals, but where such chemicals are used, a full study will be required to assure that no uncontrolled chemical reaction that could structurally damage or corrode the reactor components is creditable. _4. COLD NEUTPON FACILITY The purpose of this facility is to provide an intense source of low energy neutrons. This wil'. be accomplished by 3 placing about 1 ft of D 0 ice cooled to 25 K in the large 2 beam tube whose location is shown in Figure 4.2 of NBSR 9. A more detailed drawing is shown in Figure 8.3 of NBSR 9. This figure shows the beam port filled with a simple water tank. I

This water tank will be used until replaced by the facilities described here. The facility is divided into four main parts: the cryostat [ containing the cold moderator, the shielding tip which shields the cryostat from gamma radiation, the two beam tubes which allow beams to be extracted from the cold ice, and the cryogenic re-frigeration system. The cryostat is supported on the end of a square plug that slides inside another square plug which holds the shield tip. Both concentric plugs are penetrated by match-ing holes which are concentric to the beam port center lines shown in Figure 8.3 of NBSR 9. Each component is discussed separately below. 4.1 Shielding Tip. The purpose of the shielding tip is to shield the refrigerated ice from the main gamma-ray heating. The shield will be a 2" to 5" thick shell and will be mounted on the end of the large rolling plug in the same manner as.the water tank shown in Figure 8.3. It will reduce the core and other external gamma-ray intensities at the cold moderator to the point where it is less than the capture gamma-ray in-tensity from the structural materials inside the shielding shell. The shielding shell will be a combination of lead and bismuth encased in aluminum. It will be cooled with D 0. }, 2 r r The heat generation in the shield has been calculated to be 27 Kw at 10 Mw reactor operation. This calculation is based on the type of gamma-ray flux calculation described in sub-l section 4.7.7 of NBSR 9. It includes capture gamma-rays f rom the aluminum structural materials. This rate is such that only 1 or 2 minutes of complete thermal isolation of ne shield during 10 Mw operation would raise its temperature to the melt-ing point of the shielding material. Thus, loss of coolant must cause reactor shutdown. Furthermore, the core decay heat is significant and emergency cooling will be provided to remove it. As a further precaution, the system shell will be designed to contain the shielding material even if it should melt. I 22-10

The maximum reactivity effect that the shiciding tip would have is calculated to be -1.57.. This is a long range installation which would be removed only if repairs were required. The neg-ative reactivity is caused by the neutron absorption in the shielding material, so there is no way accidentally of introducing the 1.57. reactivity into the reactor operation. Thus, the re-activity tied up in the shielding tip presents no problem to i reactor operation and will not be included as part of the 2.6% ap allowed for experiments. The shielding tip will essentially i constitute a permanent facility of the reactor system. The radioactivity content of the tip will almost completely be due to the neutron activation of the structural material. To the extent that bismuth is used, the polonium produced by neutron capture must be considered. It presents no radiation -5 problem since it emits almost no gamma rays (~10 gammas / dis-integration), but due to its alpha emission it represents a significant contamination problem. Therefore, any bismuth must be enclosed in a gas tight container. 4.2 C ryos tat and Moderator. The moderator will be largely D 0 ice with up to 107. H O ice mixed in. It will be 14" in 2 2 diameter and 10" deep yielding a volume of 9 ft A vacuum cryostat will thermally insulate the ice. The cryostat is mounted on the end of a square shielding plug which slides inside the larger, rolling plug which carries the shielding tip. The reactivity of the assembly is estimated to be negligible since it is separated from the core by at least 4" of shielding material. At 10 Mw operation, the rate cf heat generation in the cryostat and moderator has been caltoi:ted to be less than 400 watts. Most of this comes from gamma radiation due to neutron capture in the structural material of the cryostat and inner liner of the shielding tip. This heat will be removed by cooling coils in the moderator through which 25 K helium gas will be circulated. The capacity of the helium refrigerator I is 1000 watts which allows an ample margin of reserve cooling I 22-11

I capacity. The primary helium refrigerator is backed up by an emergency helium recirculating system which provides helium gas at liquid nitrogen temperatures. The f ailure of the cryogenic cooling presents no signif-icant problem as far as high temperature structural damage is concerned since the heating rate is low and the mass large (~26 Kg). In addition to the time it takes to reach the melt-ing point of the ice, it would require about 6 hours to melt it and another 7-1/2 hours to reach 100 C. During this time the melted moderator could easily be drained and the remain-ing structure cooled by breaking the vacuum to let the heat leak out to the surrounding medium. The only significant potential hazard from the moderator would be due to the possible adverse effects of a radiolytic decomposition of the moderator by the radiation field. Only I simple radicals such as H*,

• 011, are produced and their re-combination rate within the moderator is large. While the ice temperature is below 80 K, these radicals are rather im-mobile, and are only released from the solid in the form of gases as the moderator is warmed above this temperature to the melting point.

It is estimated that the maximum number of free radicals of H* within the moderator available for re-lease on warming would give rise to a free gas mass of hydrogen about 0.3 gm and only about 1/10 as many 0 I 2 molecules would be released. The moderator will be constantly vented and could be swept or exhausted into a vacuum to remove any gases released. Any potential explosion would be contained within the heavy shielding tip. The system will be thoroughly tested before installation and the anticipated radiolytic decomposition will be correlated with experimental work on a similar facility which has been completed at Argonne National Laboratory and is currently awaiting a reactor (CP-5) shutdown of sufficient length for I its installation. The radiolytic products will be studied as a function of radiation time, and techniques developed to properly handle any problem they may present. We are in constant g 22-u

I touch with the Argonne program, and the actual radiolytic mea-surements will be participated in by a member of the NBS Re-actor Radiations Division. 4.3 11e lium Re f rigerator. The helium ref rigerator is presently installed on the main floor of the reactor building. It con-sists of the main system and an emergency backup system. The main system contains, within the reactor containment building, a liquid nitrogen heat exchanger, a turbine expansion engine, I and various other valves and heat exchangers to complete the system. Most of these components are located in a large vacu-um chamber for thermal installation. External to the reactor building, in the secondary pump room, is the 300 SCFM com-pressor to drive the helium through the refrigerator. The emergency system consists of its own helium blower and liquid nitrogen heat exchanger. The refrigerator is connected to the in-pile cryostat by vacuum transfer lines. The mechanical details and theory of operation have no I bearing on reactor safety and so won't be discussed here. In-stead we shall limit the discussion to those aspects of the system which bear on reactor confinement and the potential release of gas into the building. In addition to the compressor, the following major items are located outside the confinement building: Capacity llelium gas holder 300 SCF llelium pressurized storage tank 400 SCF Rack for 10 helium bottles 2000 SCF Liquid nitrogen storage tank 120000 SCF The lines which penetrate the confinement building walls for this system are listed below. They all pass through the 12" opening listed in Tabic 3.5-2 of NBSR 0 for " cryogenics service feed through", except wires which pass through a separate conduit. The opening is sealed around the lines by polyurathane. 22-13 I

I Type Number Size ( 1._D._ ). Liquid N Supply 1 1" 2 Liquid N v nt 1 1" 2 Main lie flow in 1 2" Main lie flow out 1 3" 3 Make up lie 1 1" lie line to gas holder 1 2" lie line to emergency skid 1 1" I lie line to gas turbine 1 1/2" Pressure gauge lines 6 1/4" Electrical cable, multi-conductor 1" All these lines form closed loops within the confinement building and are thus really part of the confinement system. With the exception of the nitrogen vent line all the external lines are also closed loops. The nitrogen vent line is a vac-uum jacketed line which vents the nitrogen boil off from the liquid nitrogen heat exchangers. Since there is no reason to expect a correlation between the breaking of this line and a reactor incident, this line presents no creditable hazard. I In the unlikely event that it were broken, it could be readily plugged. If any of the helium gas lines were broken, helium from the system would enter the building. The maximum volume of helium available (gas holder, helium bottle, cold helium in system, and pressurized storage tank) would not exceed 3000 I SCF which is only about 1/2% of the building volume. The liquid nitrogen input is a vacuum jacketed stain-less steel line and therefore doubly contained. In the highly I unlikely event that both the jacket and inner line were broken, a manual valve located external to the confinement building at the liquid nitrogen storage tank would be closed manually. A dif fusion pump in the system is cooled by water which is taken from the domestic service and returned through a shutoff valve to the storm sewer drain. ~ I 22-14 I

I 4.4 Crvostat Beam Tubes. As shown in Figure 8.3 of NBSR 9, two beam tubes look at the cold neutron source. These beam tubes and the associated programs fall into the class of ex- ? periments already discussed in sub-section 3.2. They are external to the thermal shield and so do not interact with the reactor core. A time-of-flight type of experiment is I 4 planned for one beam with a neutron filter placed in the beam hole. 5. PNEUMATIC TUBE EXPERIMENTS The pneumatic tube system is described in sub-section 8.1.6 of NBSR 9. The system is designed so that the tubes can be removed from the re-entrant tubes in the reactor I vessel. The system consists of concentric tubes where they enter the reactor. The sample carrier travels in and out through the inner tube, and the CO propellant flows through 2 the outer. The tubes are water cooled by water from the de-i mineralized water system, and in an emergency the flow of CO2 through the system would also serve as a coolant. 5.1 Sample Restrictions. The system will be used for the irradiation of small samples. The most extensive program will be in activation analysis. Many different types of materials will be studied. Any material will be allowed provided it l l meets the following criteria: (1) Reactivity worth <0. l*/. = (2) Expected activity <25 mci (Co equivalent) (3) Non explosive (4) Does not amalgemate with aluminum or in any way interact chemically in such a way as to endanger aluminum structures (for example, Hg). (5) If not a firm solid, it must be encapsuled sep-arately from the rabbit itself to prevent sample contamination of the rabbit. I (6) Will not overheat if left in reactor for significant periods of time. 22-15 f I

I Samples that fail to meet the above conditions are not automatically rejected, but must be separately reviewed. All I irradiations must meet the standards set for beam tube ex-periments in respect to the incredibility of damage to reactor I components. The activated samples are returned to lead lined receivers which are located in hoods. The thickness of the lead is 1-1/2" except for the doors which are about 1-1/2" Pb equivalent. Thus, a 25 mci cobalt source would give rise to a radiation field no greater than 50 mr at l' from the sample in the receiver. The procedures for handling and analyzing the samples af ter return to the lead lined receivers does not directly effect reactor safety or operation, but it would be reviewed I and continually checked by Health Physics. Any unusual pro-cedures (those which, in the health physicist's opinion, might fail to meet the appropriate AEC guidelines) would be reviewed by the NBSR safeguards committee. 5.2 System Safety. Two potential problems that are common to all uses of the pneumatic tube system will be discussed here. One is the problem of adequate cooling including failure of the water cooling, and the other is the problem arising from a stuck rabbit. 5.2.1 Pneumatic Tube Cooling. The pneumatic tube tips are thin walled (1/16" thick) concentric tubes terminated in a solid aluminum plug. The tips enter re-entrent beam ports in the reactor vessel with a clearance of 1/8" on the radius. The space in between is part of the CO jacket around the vessel. 2 Four water cooling tubes run between the inner and outer tubes I 5 and pass through the aluminum plug at the end to which they are welded. The heating rate in the hottest tip is about.7 w/gm which produces 85 watts of heat in the aluminum end plug and an additional 170 watts in the last 8" of the tube. The total of 255 watts plus the heat produced at a much lower rate in the remainder of the assembly is easily removed by the water flow, I 22-16

I If the flow should fail, the assembly could still be cooled by the CO2 gas which is used to propel the rabbits. I This should be quite adequate although it might prove rather awkward. An additional safety factor, however, is present in the fonn of radiant cooling and conduction of heat from the outer l tube through the 1/8" CO Sap to the reactor vessel. These 2 straight forward heat transfer calculations show that even without any forced circulation of water or gas, the maximum temperature any part of the tip would reach during 10 Mw operation would be less than 930 F, well below the melting point of aluminum. This temperature would, of course, be much lower if the reactor were shutdown. The reactor would not be started up if the pneumatic tube cooling system was not functioning properly, and any but very brief failures of the system would be caused for shutting down the reactor. 522 Stuck Sample. If a sample should become stuck in the system its location would be determined. If uncertainty existed as to whether it was in the reactor or not, it would be assumed to be in the reactor. Unless it were a sample whose exposure time was limited by radiation heating, an immediate shutdown of the reactor would not be necessary. Time would be available for an assessment of the situation, location of I the rabbit, and attempts to extract it. If che rabbit re-mained stuck in the reactor, it would be necessary to shut-down the reactor and disassemble the pneumatic tube. This can be done by removing shielding plugs and unbolting and withdrawing the tube tip through a plug hole and into a shield-l ing cask. 6. VERTICAL THIMBLES The vertical thimble facilities are described in sub-section 8.1.7 of NBSR 9. They are designed for a small sample irradiation including irradiation in controlled environment. Typical experiments will be examined here to I u

indicate the types of experiments anticipated and the attendant safety considerations. 6.1 Typical Experiments. The following examples illustrate thimble use and safety considerations. 6.1.1 Radiation Effects--Non-Instrumented. Small samples of solids are irradiated at ambient temperatures for periods of several weeks. They are inserted or removed only during the re-actor shutdown. Normal water flow is sufficient for cooling. A typical example would be small (<1 gm) copper crystals encapsulated in an aluminum shell. l 6.1.2 Radiation Effects--Instrumented. These experiments are similar to those described above, but will be instrumented to measure such things as temperature, resistivity, dielectric constants, etc. The sample will be mounted at the end of a long thimble through which the instrumentation wires would be I lead through the top plug to the sample. 6.1.3 Radiation Effects--Controlled Environment. This 1 combines such things as temperature control with instrumentation. I For example, a sample might be cooled with low temperature i helium gas to control its temperature as desired between room temperature and 20 K. Studies of the various properties men-tioned above might be made as a function of temperature. High temperature studies might also be made using radiation as a heat source and controlling the temperature by coolant flow or g conduction to the water cooled shell. g 6.1.4 Source Production. Long lived sources can be pre-pared by irradiation in the core thimbles. Since the fuel element transfer system does not interfere with samples in the experimental thimbles, they may be left in as long as desired. In other cases, I short lived isotopes may be prepared for various experimental programs. Very short lived isotopes would be prepared in the pneumatic tube system, but intermediate ones would be made in the vertical thimbles and withdrawn during operation. I

I 6.2 Safety Considerations. Experiments in the vertical thimbles must, of course, satisfy the criteria set forth in section 1.2. In addition, several special problems associated with the use of the vertical thimbles will be examined to indicate the general I approach to such problems and develop additional criteria for the safe use of thimbles. I 6.2.1 Sample Cooling. The heating rate in the core is as high as 1.4 w/g in most materials and as high as 1.7 w/g in I aluminum. The rate in the reflector positions is much lower, so only the in-core problem will be examined here. Each experimental thimble has a flow of at least 8 gpm of water. If a 50 F temperature rise were acceptable, a total of 60 Kw of heat could be removed. This capacity is much greater than any expected requirement, so heat transfer is clearly the problem. Heat transfer from a sample cladding to the water can be made very good by designing the container so that the water must pass through narrow openings at high velocity. That this I can be donc can be illustrated by considering an equivalent fuel element channel inserted into the experimental thimble so that all the water must pass through it. Under these con-ditions, the normal 8 gpm flow is reduced by less than 2 gpm to 6.25 gpm. This can be compared to the actual flow through an element channel of around 10 gpm to see that adequate heat transfer is available. Actually the 6.25 gpm flow would yield a heat transfer rate, at saturation temperature, of about 1 Kw/in of channel length. The problem remains, of course, I to transfer the heat from the sample to the cladding. Since no experiment is detailed yet, this won't be pursued. This problem, however, is more one of satisfactory experimental design than one of reactor safety. For example, if the sample were lead and contact with the container was inadequate, the lead might melt upsetting the experiment, but the container would still be easily cooled. Presumably, in this case, the melted lead would make good contact and resolidify. In any case, the good heat transfer from the container to the water I assures that its structural integrity will not be destroyed I 22-19 m

I 6.2 2 Reactivity Considerations. Again the criteria are I used that any creditable failure of the experiment must not in-troduce more than 0.57. Ap, the worth of experiments that can be removed during operation wo.ild not exceed 0. 57. Ap, each, and the total worth of all in-core experiments would not exceed 2.67.. To give an idea of the limits set by these restrictions, several examples will be considered. A hollow container with an I.D. of 3" and a length of 10" would add less than 0.27. if flooded. If the sample were a 10 gm copper crystal, it I would have a worth of less than.017. Ap. Similarly 1 gm of 2B highly enriched U would have a positive reactivity of about . 017.. These exanples show that, for the type of experiments I planned for the NBSR, the reactivity restrictions will r.ot be difficult to meet. 6.2 3 Instrument and Controlled Environment Experiments. Any leads or tubes extending to the vicinity of the core must be so arranged that radiation streaming is prevented. Usually all such leads and tubes will be enclosed in an aluminum tube extending from the top plug to the core which is part of the support structure of the experiment. Void paths up such tubes will be avoided, and direct paths through the top plug will be prevented by bends or steps. The types of instrumentation anticipated have already been I mentioned. This instrumentation, although perhaps difficult to achieve successfully, does not react with the reactor or its operations and so doesn't present any significant safety question. On the other hand, controlled environment requires circulating fluids which must be examined more closely. The most common fluids will be D 0 and H O for temperature control, and cold 2 2 helium gas for low temperature work. In some cases liquid helium may be used in the future but the technical complications at present preclude its use. As far as reactor safety is con-I cerned, the only special requirement on the use of liquid helium would be that a loss of thermal insulation resulting in vaporization of the helium would not result in the dangerous overpressure of its container. In the case of circulating I 22-20

aii

- Ei M water, its purity must be maintained to minimize activation, M and the lines entering the top plug must be adequately shield-U ed. This is achieved by using D 0 and H O from the experimental _]_ 2 2 cooling system described in sub-section 7.2.4 of NBSR 9. Con-7 M nections are available to these systems in the trench at the -e -.= ] second floor level. As described in sub-section 8.2.2 of NBSR 9, interconnecting tubes to these systems can be run below floor level, so they are always rhielded by at least 2" of M_- M steel. The negative reactivity of the H O coolant is negligible g 2 since even a liter of H O has a negative worth of only.027. Ap. 2 Any helium used in a refrigeration cycle will be of very high b purity and so would present a negligible activation problem. _g It is possible that other circulating fluids may be desirable for some future experiment. If so, their use will be analyzed 9 to be sure that criteria on reactivity, activation, heating, pressure and chemical reactions are met. g 4rmi 6.2.4 Sample Activation. The degree of sample activation liii -N that will be allowed is primarily determined by the sample handling 9d procedures af ter removal. Thus, any experiment will be examined ya to assure that it can be safely removed from the reactor. Hot =5i M cells capable of handling 10 kilocuries of activity are planned E for the future. So this is the upper limit on the strength sources that would normally be handled. Of course, most samples would have a much lower activity of the order of 50 curies or E less. The top floor plate over the reactor, and trench cover -mmm plates were designed to allow the removal of strong sources, _m (~10 kilocuries). The floor plate is 6" of steel to shield U the source as it is withdrawn from the top plug through the 2 opening below the floor plate. In addition to the floor plate, E .a the lead shielding flask into which the source and its support-ing structure is withdrawn will have a broad flange, about 6" _ 1_ _g thick, extending from its base to add further shielding in the immediate vicinity of the source withdrawal. In -= q 22-21 =d A 5 _Q -um

I 6.2.5 Sample Removal During Operations. In some cases I it may be desirable to irradiate samples for times shorter than a normal reactor cycle. In such cases, provision must be made for the removal of the sample from the core. Two cases will be considered to illustrate the basic concept. The simpler one is the case where the sample needs only to be removed from the neutron radiation field and can wait for reactor shutdown for removal from the reactor tank. The other case requires complete removal from the reactor tank during operation. In either case, the reactivity change will be limited to less than I 0.57 and usually be much smaller. In the case where it is only necessary to remove the sample from the neutron radiation field, the sample can be withdrawn a few feet into the region above the core. For the large ex-perimental thimble positions, this can be into the poisoned sleeves where both fast and thermal neutron flux would be very small. This motion can be achieved simply by a rod supporting the sample being partially withdrawn through an o-ring seal in the top plug. The requirement of completely removing a sample from the reactor is more difficult to satisfy. There are several methods that might be used. Two will be discussed. Consider a sample which can be withdrawn through a 1" diameter hole, which is allowed to fill with D 0 as the sample is removed. The sample 2 itself must be withdrawn into a transfer cask. When the cask is removed, a hole, 1" in diameter, is left through the top shield plug to the D 0 water surface. The hole would then be 2 filled with a plug. During the time the hole was open, the I radiation field would be only 6.6 mr/hr in the beam defined 2 by the hole. The beam area would be less than 2 in at the floor plate over the reactor. This small field would present no problem during removal of the cask and insertion of a plug. Another method is to adapt the sample support so the sample can be removed by the fuel element transfer mechanism and lowered through the transfer chute. I 22-m

7. TilERMAL COLUMN The thermal column can be seen in Figure 4.2 of NBSR 9. The graphite region is the only region accessible for experi-ments. As can be seen, it is well separated from core and re-actor vessel. Thus any experiment utilizing the thermal column will have negligible reactivity effect. The thermal column will be used for experiments requiring a high cadmium ratio such as the development and standardization of thermal flux measuring techniques. It may also be used for capture gamma ray studies using coincidence techniques where high thermal intensity is required and the sample must be observed simultaneously at two different angles. Any experiments in the thermal column must satisfy the criteria already set forth for the other types of experiments. No anique problems are associated with its use and, in general, problems are similar to other experiments but simpler. E E E E C E E 22-23}}