ML20008E414

Draft Proposed Amend 4 to Part B of Application for CP & OL, Superseding & Revising Previous Pages,Figures & Drawings

ML20008E414
Person / Time
Site:	Yankee Rowe
Issue date:	07/15/1957
From:	YANKEE ATOMIC ELECTRIC CO.
To:
References
NUDOCS 8101060870
Download: ML20008E414 (65)

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TABLE OF CONTENTS SUPERSEDED PAGES REVISED-PAGES 101:13 through 101:16 101:13 through 101:20 102:2 through 102:4 102:2 104:1 through 104:4 104:1 through 104:10 201:1 thrcugh 201:3 201:1 2131 through 213:5 2131 through 213:5 21931 through 219:2 219:1 thrcugh 219:4 408:1 408:1 501:4 501:4 through 50126 503: 2 through 503:2B 503:2 503:5 through 503:6 503:5 through 503:9 505:1 through 505:4 505:1 through 505: 6 506:2 506: 2 REVISED SUPERSEDED FIGURES AND DRAWINGS FIGURES AND DRAWINGS To Follow Following Nx Pare 2 Nc, Pare 8 101:13 8 101. 4 9 101:15 9 101:17 36 503:6 36 503:6 37 503:6 37 503:6 9699-QM-1 219:1 9699-QM-1 219:1 9699-FE-11 216:8 9699-FE-1A 216:8 8A 101:14 9A T01:18 9699-QE-1 219:3 cr ;_;7 1 Fig.12 (not revised) now follows Page 102:4 Fig.13 (not revised') now follevs Page 104:8 Fig. 41 (not revised) now follows Page 505: 2 i Fig. 42 (not revis~ed) now follows Page 505: 2 All other figures and drawings previously included in sections that have been revised will still fcllow same page as before. P00R BREN1 l l s ...a,, en & *.. .m ..s,.

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) ^ 101813 7/15/57 severr1 temperatures above room temperature and the operating e temperature. The value determined from the data in the operating temperature range is checked using the heterogeneous resonance escape probability formulation. The contribution to the temperature coefficient arising from neutron leakage is calculated using the equation, 1 Bk 3 leakage = E BT 1+8 8T where the partial derivative of the migration area [ is de-termined from a plot of migration area vs temperature in the room temperature and operating temperature ranges. In calcu-lating the contribution due to changes in thermal utilization with temperature, tus only nuclear density change which is large enough to be appreciable is that of water. The cross section variations which show up in thermal utilization are those which result from changes in "non-1/v factors" since, practically speaking, the 1/v variation of most neutron absorp-tion cross sections are cancelled out. The effect of chemical neutron absorber in the coolant moderator during plant varm-up is such as to decrease the negative j temperature coefficient. However, it is not anticipated that i enough chemical poison will over be present to cause the temperature coefficient to go positive. The concentration re-quired for a zero coefficient is 2.6 grams of boron per liter of 1 coolant in the hot reactor, or 2 3 grams per liter when the reactor is cold. On the other hand, 2.1 grams per liter provides ] 2 per cent shutdown of the cold clean core without any control rods. 4 P00R BRIGNu L =

101:14 7/15/57 Figure 8 shows the temperature coefficient of reactivity vs temperature at several boron concentrations. Figure 8A shows the temperature coefficient of reactivity vs boron concentration at three different coolant temperatures. It should be noted that the reactor is always suberitical at boron concentrations which could result in positive temperature coefficients, even with all control rods withdrawn. The result of these calculations are given in Table 3 Table 3 Temperature Coefficient of Reactivity Water Temperature 68F 508 F Contribution from: +.5 x lO-5/F +4 x 10-5/ Fast effect Resonance escape -2.5 -33 Thermal Utilization Without chemical neutron absorber .3 +6 Leakage .4 -4 Total, without chemical neutron absorber -2.7 x 10-5/F -27 x 10-5/ Pressure Coefficient of Reactivity l In the primary plant, reactor plus main coolant system, the nominal system pressure is 2,000 psia. Since the temperature controls for the pressurizer work over a finite range, and since l there are surges in the system due to changing flow of the soolant, the actual operating pressure may fluctuate above and below the nominal syst6m pressure of 2,000 psia. The pressure swings are calculated to be i 150 psi. With changes in system pressure, the density of the moderator in the reactor changes giving rise to l an increase or decrease in reactivity. The effect may be described l l l PDDRDRIGINL 4 U M -- < Y. 94 r. e a m

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101:15 ?/15/57 as a pressure coefficient of reactivity. Since this factor is a function, among other things, of the total neutron absorption in the moderator, two values have been computed, one without chemical neutron absorber in the coolara and one with chemical neutron absorber. The data are shown in Table 4. Table 4 Pressure Coefficient of Reactivity Water temperature 508 F System pressure 2,000 psia Without chemical neutron absorber +2.8 x 10-6 per psi With chemical neutron absorber (1.6 g boron per liter) +1.0 x 10-6 per psi The pressure coefficient of reactivity of the reactor is positive. Dttring plant transients, this coefficient opposes the temperature coefficient, since positive pressure su ges occur simultaneously with positive temperature surges. The pres-sure coefficient, being smaller never overrides the temperature coeffleient, but reduces somewhat its effectiveness. The pressure changes in the primary system due to changes in the temperature within the pressurizer, which result from the on-off type of control, are smaller than those asso-ciated with plant transients and, in general, take place over a relatively long period of time. It is difficult to see how any hazard could be associated,with pressure changes. Doppler Coefficient of Reactivity The reactor fuel is 2.6 per cent enriched uranium dioxide. Since this is a homogeneous fuel, that is, the U-235 and U-238 are intimately mixed, the temperature of the fission-g able (U-235) and fertile (U-238) materials are the same. As a s

f. 101:16 7/15/57 result, the 8radening of the neutron absorption resenance peaks in U-238 with incrpasing temperature is a rapid effect and results in a " prompt" negative temperature coefficient of reactivity. ~ The Doppler effect is caused by the spread in relative velocities between neutrons of a given vector velocity and uranium nuclei with various vector velocities in such a manner that the effective widths of absorption resonances are increased, thus de-creasing the self-shielding of uranium nuclei. The U-238 resonance integral has been measured to have a temperature coefficient of + 1 x lo-3/c (Nucleonics vol. lo, No. 5, 64, (1952). Differentiating the expression for resonance escape with respect to temperature, the following expression is obtained: 1 No) O Resonance Integral = P Gj 'B T s This expression is evaluated and the data are shown in Table 5. l %ew 5' t The Doppler Coefficients .7 x 10-7 er deg F, at 68 F p .8 x'10-7 per deg F, at 508 F F An additional " prompt" coefficient due to uranium dioxide density change with tempereture is estimated to be an order of magnitude smaller than the Doppler effect and is, therefora, neglected. a h- \\ - o s - L. s- ....w J. -w. d-~~-~ ~ - ~ ~ ~ - ' ~ - * ' - ~ ~

4 101:17 7/15/57 Void Coefficient of Reactivity Two of the basic assumptions in the design of the re-actor core are that local boiling, surface boiling of the sub-cooled' liquid, is permissible within the core but that bulk boiling is not allowed. The presence of local boiling does not alter the reactivity of the reactor provided it is restricted to a small region of the core. The reactor core is designed so that, in normal operation, bulk boiling does not occur even in the hottest channel. Under accident conditions, however, it is conceivable that bulk boiling may occur. The reactivity of the reactor then may be expected to be altered by the presence of steam voids. The effect of steam voids on reactivity is evaluated quantitatively r and expressed as a coefficient of reactivity. In the operating range where the average temperature of the coolant is 508 F, the void coefficient of reactivity is negative with a value of ~ -0 3% esk/k per % void. When there is no chemical poison.n the coolant the effect of voids on the keft of the core without l chemical poison is shown in Figure 9 in which curves are plotted for three mean core terperatures. As the ter.perature of the reactor is lowered, keft increases and more reactivity becomes available; therefore, a larger per cent void is required to shut l down the reactor. L

g 101:18 7/15/57 ( A change in system pressure may be expected to have an Given a set of initial conditions, if effect on the void volume. the system pressure were to increase as a result of reactor in-strumentation calling for heat to be added to the pressurizer, The time required by the the voids would be reduced in volune. pressurizer to go from the bottom of the dead band of 1,850 psia If a maximum 10 per cent void is to 2,000 psia is 16.5 minutes. assumed, which is an extreme estimate of the voids due to local boiling, and if -0.3% Ak/k per % void is used for the void coefficient of reactivity, the rate of reactivity addition will be 3 0 x lo-I ak par see. This is approximately 1/3 the maxi-zum rate of reactivity change associated with operation of the control rods. Although operation of the reactor is predicated on little or no boron in the coolant under power operating conditions, th effect of boron on the uniform void coefficient has been ir_vestigated. Figure 9A shows the effect of boron concentration on uniform void It should be noted' coefficient at three different temperatures. that the reactor is always suberitical at boron concentrations which could result in a positive void coefficient, even with all control rods withdrawn. Effeets of Plutonium Build-Up At the end of the core life it is anticipated that approximately one-third of all fissions take place in the plutonium that has built up following capture of neutrons by U-238. Since ( 1

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Plutonium has a wealth of resonanec F meture it should make.a sentribution to the Doppler coefficient as should U-235 which has a number of fission resonances. Information from fast reacter l projectshasindicatedthatbothU-235andPu-239give/apositive contribution to the Doppler coefficient. However, Pu-239 gives a smaller contribution than U-235. Also, the Doppler coefficient of a U-238 - U-235 mixture does not so to zero until the proportions are 1 to 1; thus the combined effect of U-235 and Pu-239 resonances 1 should be small in the Yankee reactor where their combined con-centrations will be less than 3 per cent of the U-238 present., l Therefore, it is concluded that throughout the core lifetime there l 1s no significant change in the overall Doppler eceffielent due to l the buildup of plutonium. The presence of higher plutoniu$a isotopes' should not change this conclusion because any positive contribution from fissionable Pu-241 should be negated by the purely absorption resonance in Pu-21+0 at 1 electron volt. Similarly, investigations indicate that the conttibution of resonance absorption in plutonium to the temperature coefficient is a negative effect. After 10,000 hrs of full power operation the magnitude of this effect is -4.6 x 10-3 n. P00RBRIBIN?L The effect of plutonium in the reactor on the delayed neutrons and the consequent effect upon accidents is being investi-gated. It is anticipated that there vill be negligible difference l between an accident with plutonium U-235 and U-238 in the reactor, 3 compared with only U-235 and U-238. The reasons for this are that no significant transient occurs until prompt criticality is passed and after prompt eriticality is passed the delayed neutrons have ,N M A e = = ' " "'d .~-,-, -,,., - - -....,,.v .-,n

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102:2 7/1 5/57 After shutdown from full power, the Ie-135 concen-tration rises to a peak at 8.1 hours. If the reactor is started during the peak )(enon concentration, the zenon burns out rapidly to below its equilibrium value, because no iddine precursor was formed during shutdown. When the reactor is started up, the iodine is formed again, and gradually comes back up to equilib-rium. During the initial stage, burnout is rapid. It must be established that available control rates can handle the maximum rate of decrease in neutron absorption cross section and the consequent increase in reactivity. Figure 11 shows this situation for the case in which a new reactor core is run to equilibrium at full power, shut-down, and then started at the time of maximum Kenon. This maximum )(pnen gives a 1, =.0071,e,-1 at 8.1 hours after shutdown, compared to.00415 cm-1 at equilibrium. Upon starting up again, EXe drops to about.0032 cm-1 after 8.9 hours and then rises towards equilibrium. The rate of change ( ' is plotted and l converted into dk/dt. ~ The rate of change of k,ff with time is l maximusk at start-up with a value of +3 5 x 10-65 A k per sec. k l This can be handled easily by the control rod system and thus presents no possibility of a runaway. 1 l The xenon instability problem, xenon tilt, as described in CRRP-657 by A. G. Ward of Canada, has been inve-tided. For the Yankee core, the migration area is appro.lrx. siy 56 square centimeters; the length of the core is 7 1/2 ft; and the l dore diameter is 6.2 ft. The square of the core height in feet v

I 10283 7/15/57 divided by thu migration area equals 1, and the square of the core diameter in feet divided by the migration area equals 0.68.. Based on these ratios, and according to Dr. A. Henry, a xenon oscillation in the Yankee reactor is possible, however, l not probable. The oscillation has a 30 hour time constant.. The oscil]dtion Secomes appreciable at 10 3 neutrons per square 1 centimeter per second and increases with higher flux. It is possible to consider that an oscillation will occur if enough reactivity is tied up by the xenon to equal the reactivity requirements of a second mode of the neutron flux. The problem is not an academic one since it has been observed in large reactors. This problem will be investigated for the Yankee reactor more thoroughly and instrumentation will be provided so that it can be both detected and, with the available control rods, controlled. The xenon instability provides a design condition mitigating against a reactor with only chemical control and no control rods since'the control rods are needed to modify the distortions of a neutron flux which may result from the xenon oscillations. The cause for concern with the xenon oscillations is the fact that the flux may be so perturbed that design hot channel factors are exceeded and thermal damage occurs to the core. t .n-,

102: 7/15/5 Handling of Puel Figure 12 shows the k,ff of groupings of v$hious numbers of fuel assemblies, and the k contributed by the last one added to the periphery of the group. A cylindrical geometry is assumed so that these values of k represent maximum values. Furthermore, immersion in cold water is assumed. Seven assemblies are required to achieve criticality. This limit applies only to assemblies being placed in a cluster; an infinite string stacked side by side would re7ain suberitical. In the Yankee plant, new fuel assemblies are stored dry in individual compartments which would be suberitical even with total water immersion. Spent assemblies are pulled up out of the core and,one by one, sent down a chute into a spent fuel pit located below the level of the reactor. In this storage pit, they are stored under water on 15 in, centers so that no critical configuration can arise. This conclusion has been verified experimentally by Dr. D. Callahan at Oak Ridge National 1 Laboratory. l l ( i y -,gr y=

10h:1 Draft 7/2/57 10h CONTROL General A prime objective of the Yankee project is to achieve the lowest possible nuclear fuel cost consistent with safety and reliability. In order to accplish this, every element of the nuclear fuel cycle must be examined for cost reduction possibilities. Fuel fabrication, processing and inventory charges are all important items and their final contribution to the fuel cost per kilowatt-hour is dependent on the length of time a core can remain in the reactor. The cere ic presently being designed for 10,000 hours at full power. Core lives af this duration have not yet been achieved in any operating reactor using slightly enriched uranium fuel. Experimental results, however, indicate that from the standpoint of irradiation da:nage to the fuel and structural materials and from the standpoint of corrosion 8 and thermal cycling, this result can quite possibly be attained. A core life of the order of 10,000 hours raises difficult prv.,lems of control, particularly in a pressurized water reactor with its large negative temperature coefficierA Approximately 19% excess reactivity must be provided in the clean cold core in order to remain critical at operating temperatures and with equilibriun pcisons present to the end of the 10,000 hours ' life. This excess reactivity requirement may be broken down as follows: Cold to hot 7% Fuel burnup 7% Equilibrium xenon and samarium 5% Total 19% k. In order to assure that the reactor can be rendered suberitical at ro am temperature with a new, clean core, a margin of 5% above this figure must be provided, and the tott.1 control range needed is, therefore, 2h%.

10L:2 Draft 7/2/57 In any reactor of this type, it beccres a problem to provide this amount of control entirely by me2ns of mechanical control rods. 'aten stain-less steel is used as a cladding and structural material, the fuel enrich-ment utst be increased slightly and the problem becomes se:newhat mere acute since the worth of a given control rod material and configuration is significantly less than when low cross-section materials, such as zirconium or aluminum, are used for cladding. In this reactor, a centrally located mechanical control red made of a material that is black to thermal neutrons only has a worth of approximately 2%. Additional control rods located at points away from the center have diminishing worths until those located near the periphery of the core have a value of only 0.1%. Accordingly, a very large number of control rods, possibly 75, would be necessary in order to cover the desired range of 2h%. If such a number of control rods were to be used--each with individual drive mechanism--it would mean 75 precision mechanisms, 75 penetrations through the vessel head, and would further mean sub-dividing the core itself which now consists of 76 assemblies into perhaps four times that number of smaller units. The large number of pene-l trations through the vessel head would seriously complicate fabrication and raise formidable questions of structural integrity. Some of these objections could be avoided by ganging a number of control rods to a single mechanism, but this suggestion has always met with limited enthusiasm because of l l the mechanical difficulties that arise and also because of the inability l l with ganged control rods to regulate various regions of the core through individual rod programing. l An additional disadvantage of a large number of control rods is the fact that to accommodate taez, about 6% of the fuel rods would have L to be omitted, thereby decreasing heat transfer area by the ss.ne amount. L

10h:3 Draft 7/2/$7 t Further heat transfer loss is encountered because of bypassing more coolant around the heat producing surfaces through the many control rod channels. With the same general core configuration, the dimensions of the core would have to be increased to remain at the same sverage and maximum heat flux levels. Because of these difficulties a control scheme, using a combin-ation of mechanical control rods and a chemical neutron absorber dissolved in the coolant-moderator, is proposed for this reactor. Reliance is placed on the natural stability inherent in a pressurized light water reacttr to handle short-term transients. Twenty-four mechanical control rods are used to control reactivity at operating temperatures. Space is provided for eight additional shim rods near the periphery of the core which are to be used if necessary to adjust initial reactivity of the co re. The control rods themselves can be programmed to attain favorable flux patterns during operation and, in addition, can be used under manual control to counteract any tendencies toward xenon tilt or instability. A homogeneous chemical neutron absorber is added to the coolant-moderator for cold shutdown and to hold the reactor suberitical in a clean condition at room temperature. While initially it is not intended to operate the reactor at power using the homogeneous chemical neutron absorber as a shim control, the ultimate possibility of such operati in is believed to offer many advantages. Chief of these is the fac' that if the excess reactivity can be counteracted by the dissolved chemical neutron absorber, it would permit operation at full power with all but one or two mechanical centrol rods fully withdrawn and, therefore, available as safety rods. ' Operating in this manner would increase the thermal and nuclear performance of the core

10L:L Draft 7/2/57 I while measurably reducing the duty on the expendable mechanical control rods and the wear and tear on their associated drives. The natural stability of a pressurized.;sor reacter lends itself to the slow reactivity changes provided by injection and dilution of the liquid neutron absorber. In addition, the use of a homogeneous shim offers the possibility of employing the entire volume of the core for heat production, thus realizing maximum heat transfer capability and minimizing the possibility of local het spots and fuel burnout. Borax III and E%"4 have been operated successfully at power for limited periods of time using a dissolved boren compound as a homogeneous shim. This experimental evidence is encouraging, but it is recognized that there are still many problems associated with operating a reactor in this manner and that these problems are not at this time well understood. The Research and Development Program now underway includes an extensive investi-gation of the behavior of boron compounds in solution with both in-pile and out-of-pile dynamic loop experiments planned. The results of this program, together with operating experience in the actual Yankea plant, may point the way to methods for safely using chemical neutron absorbers in the primary coolant during full power reactor operation.

10h:5 Draft 7/3/$7 Control Rods Mechanical control is provided by 2h cruciform control rods located in four concentric rings around the center of the mactor. Provision is also made for eight additional control or shim elements in the outer region of the core. By placing at these locations fixed elements of a neutron absorbing material, inert material, or fuel, the initial reactivity of the core may be adjusted to the desired level. A design objective for the first core is to provide sufficient control rod worth to render the reactor 3% suberitical with a clean core at operating temperature. To bring the reactor from this point to 5% sub-critical at room temperature, a chemical neutron absorber will be added to the main coolant water. The total control available fmm the present design using 2h silver-cadmium-indium control rods according to conservative calculations based on absorption of thermal neutrons only lier between 10% and 12%. Experiments with such rods in critical assemblies, however, indicate control rod worths higher by 30% than rods black to thermal neutrons only. This effect is thought to be due to additional absorptions at energies above the thermal r l range. Since the k rf of the hot clean reactor is 1.113 and total control e rod effectiveness is calculated at 10% to 12% on the basis of thermal neutron absorptions only, the control rods are not adequate to meet design objectives holding the hot clean core 3% suberitical. The Research and Development Program will reduce the uncertainty in these values. If experimental evidence shows that the control rods are inadequate, five possible procedures will be investigated for obtaining more control, as follows: l

10h:6 Draf?.7/3/57 In accordance with technical discussions between Yankee and Westinghouse, it has been agreed that a two-region core is a reasonable alternative design for the reactor. Since a two-region core has advantages associated with heat transfer, burnup and control, a considerable effort will M expended on this design so that it may be used for the first core. If two different enrichments are used for the first loading in the reactor vessel, the k rf e of the hot clean core will be approximately 1.06. This reduction in kerf would allow control rods of presently calculated worth to hold the core 3% suberitical in the hot clean condition. The possibility of leaving the chemical neutron absorbing compound in the main coolant during power operation is desirable from a nuclear design standpoint and would provide any additional control required. The undesir-able aspects of using chemical control during power operation are those asse :iated with the chemistry of the main coolant. If the use of chemical control during swer operation is adopted, it would probably require redesign of some of the pla. vstems, such as the waste disposal and purification sys tems. Additional control amounting t. 'oproximately 2}% can be gained by adding highly enriched uranium fuel to tu control rod followers with an equivalent reduction in the enrichment of the fue u. .ne core. This added control would probably meet design objectives. The eight outer control slots in the pr( 1ent core design, which it is con-templated might be used for fixed elements, could be provided with rods connected to mechanisms, and 0.8% additional control could be achieved in this manner. The incremental cost associated with such a small increase in control makes this change unattractive. l ( L

10h:7 l Draft 7/3/57 If all other nethods prove to be impracticable, additional control rods could be added to the nactor by redesigning the core and the reactor vessel head. This method does not appear to be desirable at the present time because of mechanical complications, structural difficulties and increased costs. The control rods are scram:ned into the core under the followirg conditions: 1 Excess neutron level Short period during reactor start-up Low main coolant loop flow i High or low main coolant loop pressure Manual scram When the reactor is at power, automatic run-in of the control rods is initiated by high temperature in any one of the four main coolant loops. Alarms tre provided for: High reactor outlet temperature Loss of turbine generator load Reactor period less than 20 seconds An interlock is provided that does not allow a loop to discharge into the system when the water temperature in that loop differs by more than 500F from the water temperature in the active loops at the reactor vessel inlet. This is accomplished by a permissive circuit coupled to the motorized valve. An alarm for this condition is also provided. Drop time of control rods at 0.6 the acceleration of gravity is 0.6 second which for a total rod worth of 10% provides a reactivity decrease of 16% per second. (

10h:8 Draft 7/3/57 Chemical Control The chemical control system is designed to shut down the cold clean reactor by a margin of approximately 5% in a k. with al'. the control rods inserted. This margin allows one or more centrally lecated control rods to de fully withdrawn for safety purposes and still have the reactor 2% to 3% suberitical. In the present design, this would require about 1.6 g of natural boren per liter of main coolant. The chemical compound which will be used in 1 he chemical control system has not been finally selected, although boric acid and ammonium pentaborate are possibilities as indicated by the reralts of a development program at Bettis. Boron compounds have good themal stability and have adequate solubility in the cold reactor. The solubility of boric acid at room temperature is $0 g per liter of water, which means that 9pproximatey 8 g of boron per liter can be retained in the coolant. The solubility is thus more than five times greater than required. The effective maltiplication factor for the present core design as a function of boron concentration in l the main coolant is shown in Figure 13 A concentration of 1.6 g of natural boron per liter of water is sufficient to reduce keft to unity at any i temperature above 225oF. even though all control rods are withdrawn. l The present design makes use of a bleed-feed system to change the 1 I concentration of the chemical neutron absorber. The present maximum rate of water injection for this system is about 100 gpm. Since the mechanical control rods e n handle xenon and samarium transients, there is no need for 1 faster action by the chemical control system. At a bleed-feed rate of 100 gpm in a 3,000 cu ft system, the maximum rate of change in reactivity is 0.0005% Ak/k Per second,.which is well within safe limits. ( l - ~.

10h:9 e Draft 7/5/57 r Questions that have been raised in connection with zwactor centrol through use of a hemogeneous chemical neutron abscrber in the main coolant water include the question of thermal stability of the chemical solution j at operating temperature and pressure, and possible interaction between the chemical control agent and other additives present in the water. Con-siderable work has been done in this field at Pettis for the PWR project. The following conclusions have been stated. 1. The boric acid is stable in solutions at high temperature and pressure. 2. Amonium berate solutions are likewise satisfactorily stable under these conditions. 3. The use of lithium hydroxide in combination with toric acid is probably satisuctory with very low quantities of lithium hydroxide. However, if the concentration of lithium hydroxide is comparable with the boric acid cencentration used, the combination may be unsuitable. The conclusions are based on experiments in autoclaves and loops. A possible adverse effect which could occur in a reactor is the precipitation of anhydrous lithium metaborate (lie 0 ). Experiments indicated a precipitation 2 out of lithium metaborate at the interface between the water and vapor phases. i This is known as the drying-up phenomena. Another question drich has been raised is the possibility of inverse solubility with temperature of boron compounds which might possibly be used in the chemicc1 control system. Lithium borate is the only compound which has been found to have this property. The solubility of boric acid increases rapdily with temperature. Solutions of lithium hydroxide and boric acid are sufficiently soluble at temperatures up to about 5000F. At higher temperatures i I I

10h:10 Draft 7/5/57 I the so-called drying-up phenomena can occur as desedbed above. The solubility of lithium borate decreases from.3 mole per liter at 5000F. to approximately .2 mole per liter at 6000F. Solubilities as a function cf temperature do not affect the present plan to operate the reactor from hot to cold with the chemical control system. Since concentration corresponding to the solubility of lithium borate at 7000F. is still adequate to control and the cold clean rer' 'or, it is more than sufficient to r.aintain subcriticality at temperatures from 5000F. to 700cF. The Research and Development Program for the plant, supported by the AEC under Contract No. AT(30-3)-222, includes four major projects which pertain to the problems of chemical control. Project 2.0 is concerned with calculations of the nuclear physics problems and effects of chemical control on reactivity coefficients. Prcject 3 0 is concerned with autoclave and dynamic loop out-of-pile studies of two reference water combinations with a chemical neutron absorber. Corrosion effects on materials as well as deposition and absorber injection and dilution problems are being studied. Project 3 0 also includes Van-de-Graaff irradiations of chemical neutron absorber solutions. Project 10.0 is the performance of a critical experi-ment which will experimentally check the nuclear calculations on chemical absorbers made under Project.2.0. Project 11.0 consists of in-pile pressurized water loop tests in the,MTR, some of which will use the reference chemical absorber selected from the out-of-pile experiments and other information available. At the conclusion of the Project 11.0 experiment, the character-istics of the chemical absorber (nuclear, corrosion, precipitation, etc.) should be well established.

f 20181 7/15/57 201 REACTOR PRES 3URE VESSEL The reactor vessel is cylindrical in shape, with a hemispherical bottom head and a removable closure head. It is approximately 31.5 ft overall height by 109 in. internal diam-eter, as shown in Figure 15. The cylindrical portion of the vessel is made of carbon steel plate approximately 8 in. thick; the bottom head is 4 in, thick; and the reactor vessel head is approximately 6 3/4 in. thick. All internal surfaces of the vessel in contact with coolant water are clad with Type 304 stainless steel. The vessel is designed in accordance with ASME Boiler and Pressure Vessel Code, Section I, " Rules for Construction of Power Boilers". The design pressure is 2,500 psia and the de-sign temperature is 650 F. Main coolant water enters the vessel through four inlet nozzles near the top, flows down through the thermal shield annuli, up through the core, and leaves the vessel through four outlet nozzles located at the same level as the inlet nozzles. The concentric, cylindrical, stainless steel thermal shields rest on local supports near the bottom of the vessel. Their purpose is to limit thermal stress in the reactor vessel shell during full power operation by absorbing radiation emanating from the core. All of the reactor vessel internal supporting structure is Type 304 stainless steel. The two thin stainless steel barrels that support and hold d wn the core are supported on a ledge near

201:2 7/1 5/57 the vessel top flange. All of the internals are held in place by the reactor vessel head which presses against the core hold down ring-top plate combinttion. The reactor vessel head is approximately hemispherical in shape with a heavy flange for bolting to the reactor vessel flange. Both the closure studs and nuts are applied and removed with an impact wrench. Special, dial-indicating, elongation gages are used to limit the tension in each stud. Leaktightne ss is secured from gaskets with provision for a backup seal veld. Operating experience will show whether seal w$1 ding of the reactor vessel head is required. The reactor control rod drive mechanisms are welded to the reactor vessel head and are handled as an integral part of the head. The fast neutron flux at the inside wall (attentuation of approximately 10 through wall) of the pressure vessel integrated 20 over 30 years of reactor operation is calculated to be 10 l neutrons per square centimeter. Experimental data exist at Oak Ridge l l which state tha.t changes in the properties of steel which has been 10 exposed to 2 x 10 neutrons per square centimeter are measurable. These effects have to do with increase in hardneas and decrease in ductility of the material. However, it has also been found l l that these effects can be annealed out in approximately 30 minutes at 600 F. Since the steel enclosing the main coolant loop for the Yankee reactor will be approximately 500 F, a diffusion calculation , l l has been made using the experimental point at 600 F as a check. t s, u~.

20183 - ?/15/57 1 ) This calculation indicatos that the irradiation effects will be annealed out as they occur and that there will be no serious 1 effect to any portion of the primary coolant-moderator container. h I a e i f I i l 1 I k O ^ 'ev ,.,n,..,.,. .,-r- ~,y,. n.- 4.- g..,._,.. y,.m,w. m.,n.,m,,,-. .w-v, ..-,.n.vw~we ,.,,.... -., - - -., - - - -,, +.

213:1 7/15/57 211 VAPLp CCNTAIT C T Funetion The vapor centsiner is a steel envelope which sur-rounds the main coolant equip-ent loops and enclosos all prer.surized parts of the main coolant system. It prevents the relence cf radioactivity to the atmosphere in the un]ikely event of an accident resulting from a rurture ;nd release of fluid from the main eco]ent system within the containment vessel. When the reacter is criticc1 or when the main coolant systen is prescurised with nuclear fuel in place, the vapor container is closed and pressure-tight. ,.11 access openings, vent connections, pipe lines not required for operetion, and the spent fuel chute are kept closed with tight shutoff valves or gasketed doors. The vapcr container, wSon closed, is maintained at a precsure level slightly higher than atmcspheric for cen-tinuous leekage indicttion, with allowance made for variations due to temperature change. Associated with the outer steel vapor container is an inner reinforced concrete structure which supports the main coolsnt loop equipment, attenuntes radiation fren the main ecclant loop to a tolerable level cutside the vapor containor, and acts as a stop for objects possessing kinetic energy. Thus concrete structure is not designed to centain pressure. Generel Descrittien The layout of the vapor centainer is shown on draw-ings 9699-FM-1A, 1B and IC. The vapor container is a steel spherical shell 125ftindicmeterandwithaminimumwallthicknessofh/8in. The sphorical shape is selected since it uses a =ini-um of material for a given volume and internal pressure. The spherical shapo permits the most accurate determination of secondary i .6ress and facilitates the design of the necessary penetrations. The p]ato material is ASTM Specificaticn A-3co, Class A-201 Grade E, firobox quality, a carbcn-silicon stoc1 of suitable quality for ferring and welding in pressure vessel service. The teasi]e strength is 60 Cef-72,000 psi with a minimum yield peint of 32,0cc psi. his atecs;horic temperature outside the uninsulated sphere occasione11y stpreaches -25 F, so that the she]I motel temporstures may be close te the freezing k peint during operatien. Specificatten A-300 saterisl is em-ployed fcr its surericr imphet value Lt low te-rerature, equiva-lent to 15 ft-lb st -50F. P00R BRER

213:2 f/15/57 l The vapor container is designed, built and tested in i accordance with the ASME Boilar and Pressure Yessel Code, Section VIII (Unfired Pressure Yessels), and the code stamp is l applied. The vapor container is not provided with a relief valve, in accordonce with special ruling, Case No. 1235, which states l since it is "It is the opinion of the Committee that intended that these vessels be designed and built to i i l safely contain all the lethal radioactive substanees that may be released in case of a maximum eredible accident affecting the reactor vessel or primary coolant circuit or both and because of the hasardous character of the waterials, which might be released, l pressure relief devices are not required." The stress permitted by the Code in the specified plate is 15,000 pai. The Code further spoeifies that the design stress shall be reduced by a factor of 0 9 when employing welded seams with 100 per cent radiographic inspection. The resulting design stress is 13,500 pai. The design pressure of the vapor container is 31 5 psi corresponding to a membrane stress of 13,500 psi _in a gage t 1.25 ft dian sphere with a minimum plate thickness of 7/5 in. The internal pressure of the vapor containar in the ev1nt of a major less of water accident is 34 5 pai gage. This t pressure includes the 10 per sent overpressure permitted by the Code under paragraph UG-125(e), which states " All unfired pree-sure vessels other than unfired steam boilers shall be protested by pressure relieving devises that vill prevent the pressure frca rising more than 10 per eent above the maximum allowable working pressure, except when the excess pressure is caused by exposure to fire or other unerpoeted soures of heat." A 10 per sent increase in t'io design pressure of 315 psi gage results in an allevable pressure of 34 5 pai gage which corresponds to the internal pressure developed in the sajor loss of water accident. The spherical vessel is supported on steel solussis, The pressurised equipment within the vapor container is surrounded by a reinforced sonoreto eylinderi the bottom of which is a segment of a sphere. Concrete wall thickness is 4 5 to 7 ft. Ordinary concrete is employed having a density of 150 lb per ou ft, except in several areas in which space re-strictions require high density concrete. The concrete structure is supt:orted on eight rein-l forced concrete piers which penetrate the spherical eentainer. l l These penetrations are oesled with stainless steel expension l joints. The joints are welded to a steel plate which passes i 300R ORG NAL

l 213:3 i 7/15/57 l completely through each concrete pier below the expansion joint and which is also welded to the interior reinforcing rods, thus The completing the metallic vaper seal of the centainer vessel. steel and concrete st support construction permits the I nove freely and independently of each other, thereby eliminating l temperature stresses resulting frem restraint. i l I'ipe lines, not required for normal operatirn, which f enter the vapor contsiner, are provided with valves 1ccated out-side the vessel wall and maintained in a closed position in order Pipe lines, to maintain the integrity of the vapor container. required for normal operation, which enter the containment vessel are each provided with two check valves, one inside and one out-Operating outgoint lines are each provided with side the shell. l' a closure trip valve arranged te close automatically on pressure rise in the container. Details of Vaoer Centatner Typical details of the vapor container are shown on drawings 9699-FM-11A and 12A. t All penetrations of the sphere are reinforced to the All shall sear.s are full strength value cf the =etal removed. t conpletely radiographed, as well as all velds in the penetrationsa= enable f All welds not wherever pcssible. nation are subjected to a magnetic particle inspection at every pass. All high temperature piring entering or 1saving the voluted expansion joint encased in a steel protective sleeve These expansion joints eliminate the necessity of heavily rein-forcing the spherical shell to contain the forces and moments resulting from pipe expansions. . Conduit fittings are welded in groups into the heads of special blisters which, in turn, are welded to the spherical This design facilitates construction, testing and any The cenductor is generally shell. required corrective rewryking. mineral insulated copper sheathed cable to the conduit to ensure leak tightness. The internal concrete structure consists of two concentric cylinders of 3,000 psi compressive strength reinforced These cylinders are tied together with five reinforced concrete radial valls so located as to provide an isolation com-concrete. partment for each main coolant loop and for an access way into The wall of the outer cylinder and the radial valls are perforsted with ports sised to limit the differential the structure. pressure across the concrete walls to a value of 6 psi at the t time of a major loss of water accident. I ?DDR ORG NAL ~

213 A ?/15/57 The inner concrete wall serves as the sup ort for the reactor vessel, the water-filled neutron shield tank surrounding the reactor vessel and as a shield tank cavity ab: ve the vesnol. The shield tank cavity, which is water-filled when handling fuel, is lined with a stainless steel menbrane to assure complete watertightness. When not otherwise cets) covered, the surft.ce cf the concrete is protected with a smooth, hard fi".ish pinstic paint to prevent absorption of centsvinated vapcr and to assist in decontamination. Varor Container Test s After the vapor container has been erected and all welding, radiographing and magnarluxing have been completed, in-eluding manhcle closures and shell penetrations, the vapor con-tainer is cema!ete3y closed and subjected to field acceptance tests, includinE an air pressure test, leakace detection test, and a leakage rate test. Air PSessure Test The vapor container is pressurired wit h air to one and one-quarter tiros the design pressure, or 40 psi gnge. This con-tro))rd pressure is hold fer o period of 6 hr. If leakage is de-tecte6 by a bubble test, the vessel is depressurized, the leak repaired and the vessel retested. The air pressure test establishes the design integrity of the complete vapor centeiner, including all penetrations and closures. Leakare Detection Test The purpose of the leakage detecticn test is to es-tablish the leak tightness of all welded jointe used in the eroc-tion of the vessel and gasketed closures required in the design, and to detect individual leaks frem the vapcr container in the order of.0001 cu ft per br of air at a test prescure of 15 psi l gage. A leakage detection test by tracer gas is considered to be the most suitablo, sensitive means of ensuring mnximum l l vaper container inteErity, and particularly for leaksgo around vspor container penetrations. The leakage detectitn test is conducte.1 with a halogen type lock detector equal to the Geroral Electric Company Type if-1. This is a sensitive instruennt capnble of detecting leakage rates l as low as.0001 cu ft per hr when the vaper container contains I per cent by volume of the tracer gas Freen-12. The vapor con-tainer is pressurized with air at 15 psi gtge during testing, and Freon-12 is introduced into the container. All welded soans, P00RORGiNi

21]s5 e3 7/15/57 penetration velded joints and closures of the veper ccntainer are hand probed with the leak detector. ftny leak detected is rep; fred and the area again retested. At the complet ion r f this test, the vapor centainer is ready for the final leakage rate test. Leakare intoj[ git Final evaluation of the vac;r contni.ar is based en a leakage rate test. The vapor centainer is pressurized with air to 15 psi gage, r nd the tersperature and pressure changes are recorded over a period of several days. The t*st pressure corre-spends to the average pressure anticipnted within t he vascr con-tainer during a 24 hr period following a major rupturo cf the rain coolant loop. When the averstre sir temperaturo vitlin the vapor container coincides, or nearly ceincides, with the init ial tem-perature conditions, the pressure change is recorded. If the leakage rate should be less then 0.1 vt per cent of the contained air during any 24 hr interval, corresponding to 70 cu ft per hr (STP), the vapor container is considered to be essentially lenk-tight. A leaktge of this amount corresponds to a f rossure decrease of 0.8 in, of wr,ter in 24 hr or a tomterature decreato of 0 5 F during the same peried. The negnitude of those seasured quantitics and the pessible ini.bi ity to -er sure the true average temperature of the centnined cir affect the accurney of the leak rate descnstration. Continucus Lenkane Indicatien During Operation In order to evaluste gunntitatively the Ic kage rate from the vapor centsiner during operation and to guar;i against the chance for gross learsge through improper closure after opening the container, the vsrcr contairer is continuously l monitored. A proposed system provides that, before the reector pitnt is mnde critictl, the vaper centsiner is closc3 and pres-surized tc chrut 1 psi gage by the staticn compressed air system. Thereafter, this pressure is controlled by a compressed air bottle, system connected to the vapor container through a pressure reducing valve. The weight loss cf air fr:n the compressed air bottles is determined over an extended period and is a measure of the leakage from the vapor centainer. Effects of pressure and ter.perature fluctuations from chanFing atmospheric cenditions balance out during long time intervals. 1 l P00R ORGINAL

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r,, c it 219:1 7/15/57 219 SAFETY INJECTION-SHIELD TANK CAVITY SYSTEM Function The functions of the safety injection-shield tank cavity system are to supply borated water for flooding the shield tank cavity during refueling operations, and to the reacter vessel for cooling the core in the unlikely event of a major loss of water accident. General Descrintion The system consists of a safety injection-shield tank cavity water storage tank, two dual purpose pumps and miscella-neous piping, valves and fittings, as shown on drawing 9699-QM-1. Remotely operated pumps and valves permit contrcl of this system from the control room. Basis for Desien The system is sized for handling 130,000 gal of ~ demineralized water containing 1.6 g of boren per liter as boric acid. This volume of water is sufficient for flooding the shield tank cavity te a depth of 25 ft, providing 15 ft of shield water over fuel assemblies while they are transferred to the spent fuel pit during refueling operations. One of the 1,200 gpm injection-fill pumps pkovides for miring the stored boric acid solution, filling the shield tank cavity in approxi-mately 1 1/1 hr, and pumping shield tank cavity water to the 3 waste disposal system for cleanup if it should become slightly contaminated when it is mixed with the main coolant in the shield tank cavity during the refueling operation. i i i

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21932 7/15/97 The safety injection function of the system is accom-plished by using shield tank cavity water storage and fill equipment. Safety injection is provided te each of the four main coolant lines outboard of the main stop valves in order tc ' cool the core following a main coolant system rupture of any size which can not be compensated for by the charging system pumps. Cooling is provided to prevent core meltdown due to decay heat. The safety injection system is started manually, but with partially automatic follow-thrcugh thereafter. To minimize the chances of erroneous start-up, a single covered starting switch is provided. System functioning will occur only when the reactor pressure falls below the shutoff head of the safety injection pumps, approximately - psi gage. Injection with two pumps at a rate of 2,400 gpm fills the reactor vessel to the top of the core in approximately 3 1/2 min. Assuming that two minutes are required for the initiation of the system, this action prevents core meltdown even after an assumed instantaneous loss of all main coolant. The injection flow rate is sized to provide for the loss of 25 per cent of the total pump discharge through a single ruptured injection line or main coolant pipe. Adequate missile protection is provided for the safety injection header, and the individual injection lines are divided compartmentally by reinforced concrete partitions. After the reactor vessel ( Le

219:3 7/15/57 is filled to capacity following the rupture, the 1,200 gpm injection flow from one pump is adjusted remotely by control valve arrangement to replace just the water in the reactor vessel that is boiled off into the vapor container by the release of decay heat. The 125,000 gal safety injection-shield tank cavity storage tank provides sufficient water to replace decay heat losses for approximately 300 hr after reactor shutdown. The tank is refilled, if it should be necessary, to continue borated water injection at rates less than 5 gpm for more than 300 hr. The vapor container is designed to hold 4,500,000 lb, approximately 580,000 gal, of safety injection water. Maximum system reliability is provided by independent power supplies to each safety injection pump as shown on drawing 9699-QE-1. One pump is supplied by bus section and transformer connected to the Harriman 115 kv transmission lin 1 1 and the other frem a similar bus section and transformer con-nected to the M111 bury 115 kv line. These power supplies are not only essentially independent of each other but are entirely separate from a third source of station power, a transforner connected to the turbine generator leads. Automatic switching is provided to pick up any section of the station service bus. in the event of a power failure in approximately one-third of a second. While details of the electrical diagram are not finally ' s. - - ~

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219:, 7/15/57 settled at this time, these concepts will be adhered te and the final scheme will be as reliable as that shown on drawing 9699-QE-1. The motor operated valves of the safety injection . system operate on 125 v d-c station battery supply. Operating controls of our valves and motors for the safety injection system are grouped on one starting switch so that a single operation energizes all components of the system. At periodic intervals, the system pumps and motor operated valves are individually operated and checked, and the safety-injection water sampled and analyzed for borax concen-tration. Throughout the period of system operation, an operator is available in the control room to exercise judgment and run the system manually, if in his judgment it is necessary, in con-formance with pre-established drills and procedures and as assisted by suitable plant instrumentation. O O u___...__.

LOF:1 7/15/?7 LO8 PLANT MAINTENANCE AND ACCESS During normal operation, the plant vapor container is closed. Temperature and radioactivity build up within the con-tainer are monitored. When access into the vapor container is desired, general safeguards apply, details of which will be in-corporated in the Plant Operators' Manual. The reactor is shut down. The shielding design of the reinforced concrete struc-ture supporting the reactor does not provide for access to the vapor container with the reactor at high power levels.

Likewise, the closed vapor container ventilation system limits access.

If only minor adjustment er maintenance is required, which can be accomplished without hazard from the high pressure primary plant system, the following sequence applies: Following reactor shutdown, the ventilation system is operated with monitored and controlled discharge to the plant stack. The main coolant system is not depressurized, but slowly drops from 2,000 psia because of radiation heat losses. Boric acid solution is injected into the main coolant in sufficient quantity to keep the reactor shut down. The main coolant pumps remain in intermittent service, as required to limit the rate of temperature drop

• equalize temperatures in the system.

The pressurizer is operated manually to mainta*:. tystem pressure at levels suitable for pump o, erat The closed television circuit is operated to determine condi' ions within the vapcr centainer. The maintenance crew entering the vapor container is outfitted with proper clothing and tank type gas masks. Their entry is preceded by a monitoring and decontamination man from technical services. Entrance is made by way of the double door personnel access hatch, and minor repairs are accomplished. Major maintenance within the vapor container, such as steam generator tube plugging, is performed on a deferred main-tenance schedule at the time of complete plant shutdown and de-pressurization. The plant power level is reduced by closing the inlet and outlet main stop valve in any defective main cool-ant 1 csp. All maintenance operations on contaminated -equipr ent or in contaminated areasadt supervised by technical serv

• that proper decontamination procedures and radioactivices to see ty safe-guards are observed.

W-mA n w.n= m.-

501:4 i 7/15/57 flow through the ion exchanger are limited to approximately an unscheduled cleanup of the neutron abscrber in the 100 gpm, primary system can not cause a significant increase in reactivity. Another accident is the loss of chemical neutron ab-sorber caused by leakage from the main coolant system slightly reactor temperature greater than charging pump capacity at a and with a core ccndition that requires dissolved cherical neutron absorber for reactivity control. A leak greater than 100 gpm results in depressurization of the plant and will r.e-cessitate shutdown. This is accomplished by injecting water containing chemical neutron absorber from the safety injection system. Another possible accident might be called a " boron hideout" accident in which a deposit of chemical absorter which has precipitated within the core is suddenly dislodged This veuld cause an increase in reactivity and swept out. This is screwhat equal to the amount tied up in the absorter. f similar to the reactivity tied up in the voids of a boiling A deposit of absorber would be essentially black tc reactor. thermal neutrons and would thus have the same reactivity effect as an equal surface area of control rod. A comparisen calcu-lation with control rod worths shows that only an 0.5% Lak/k effect would result from losing a deposit corresponding to a from the center of the reactor. Since a deposit of this size 4 sq ft neutron absorbing surface [is highly improbable and since I it corresponds approximately to prompt criticality, it is cen-cluded that no possibility of hazard'results from a boron l f hideout accident. w.

501: 5 7/15/57 t Rod Withdrawal at_ Power Continuous Another type of reactivity accident is continu'cus rod In this case, the reactor is initially withdrawal at power. operating at or near full power, and a continuous withdrawal of It is conceivable though control rods at design speed occurs. that such an accident could occur through a; highly improbable, combination of equipment and personnel failures. If a continuous withdrawal of rods occurs, power level increases and reactor temperatures rise as a result of the re-With the design reactivity addition rate of activity addition. per sec and einimum temperature coefficient of 1.03 x 10 4 fi kAC reactivity, with a chemical neutron absorber in the system, of !.h k/k per deg F, the temperature rises at the rate of 4 -1.6 x 10 At these slow rates, even if overtemperature: 0 38 F per sec. f control rod insertion devices and high neutron flux level scrams l fail to function, the operator still has ample time to shut down The scram circuitry, in-the reactor before e.ny damage results. cluding.that of, the manual scram, is independent of the circuitry which normally programs the rods and, hence, is not affected by l failures of the rod programming system. If the automatic centrols fail and if, in addition, the reactor operator does not prcmptly initiate a ranual scram, bulk boiling occurs in the reactor core, thereby compensating for fur-ther reactivity additions after the temperature of the water has With forced circulation, the boiling is ex-exceeded saturation. pected to be steady up to 1 per cent reactivity in the voids. - ( The void volume corresponding to 1 per cent reactivity is - l L

501:6 7/3 5/57 4 approximate]y 3 per cent, and boiling occurs only in a relatively small portion of the core. With the reactor initially at full ' power, bulk boiling begins in approximately 100 sec. The condition of smooth boiling is expected to persist for 100 see or up to apprcximately 200 see after the continucus rod withdrawal is initiated. This allows sufficient time for the operator to halt rod withdrawal or take other corrective action. Even with-out such corrective action, it is believed that the bulk boiling effect will limit the tra.'aient and terminate the accident at safe reactor temperatures. T i j l l l l I L ~. -

n ~ 503:2 7/15/57 t Since only the fcur pump failure without scram exceeds temperature.'. imitations, a four pump failure with scram has been analyzed by means of an analog computer, with the following additional assumptions: All control rods 60 per cent withdrawn at the beginning of the transient The control rods fall 5 ft in 0.58 see The reactivity decrease due to scram is .05 per cent esk/k Equilibrium decay heat present The temperature-time relationships are similar to those given for the previous cases and are shown in Fig. 33 There is no serious effect in this accident within the first few seconds. Assuming no heat transport to the steam generators as the pumps coast down, boiling occurs at the outlet cf the hottest channel in approximately 180 sec. Thereafter, temperatures rise slowly as decay heat centinues te be generated, until heat transport ecnditions are established. Heat transpcrt by thermo-l siphen circulation through the main ecolant loops to the shell side of the steam generatcrs is the subject of a study now in It can be shown, hcwever, on an overall conservation progress. of energy basis, that it takes approximately 4.3 hr for decay heat to evaporate all of the water on the shell side of the ~eam generators and apprcximately 7 1 hr te evaporate all of the water in the steam generators plus all the water in the main coolant system down to the level where the core would be ( S - ~., _

n, C : ;

• l/15/ 57 partially uncovered.

During the first 4 3 hr after the loss of all the main ecolant pumps, the evaporated water is discharged as steam to the atmosphere through the steam generater safety relief valves and the plant stack. During the pericd from 4 3 hr to 7 1 he, steam escapes frcm the safety relief valves in the pressure centrcl and relief system and is discharged te the low pressure surge tank. Initially, this steam is cuenched by the cool water in the 1cw pressure surge tank. Eventually, this water is heated to saturation pressure and temperature and steam is discharged thrcugh the 150 psi Fage safety relief valves on the low pressure surge tank inte the vaper container. Since there are three essentially independent sources of station service power and two of them are not affected by reactor scram, a tctal inter-ruption of power te all four pumps is highly improbable. If such an interruption did occur, however, partial service could be restored in a matter of minutes,and this is sufficient to get at least one pump back in service. Less of Water Accidents I General The effects cf less of water accidents withcut any insertien of borated water from the safety injection system, but including release of contaminated vapor from the flashing of fluid in the primary ceclant system, are considered from the follcwing points of view: Core'again becoming critical Core melting down when uncovered ~ Resultant pressure rise in the vapor container i t _ __ _ _ _ -- r -

L 503:2B 5/17/57 In any case involving loss of primary system pressure, auto-matic scram is effectee b'y the control system. To inver,tigate l l the possibility of a return to criticality, a series of breaks of increasing size is assumed, a small break equifalent to a lA in. diam opening, a medium break equivalent to a 1A ino to a 4 in. dian opening, and a large break equivalent to the rup-I turing of a 20 in. OD main coolant pipe. I ( i l L ^ "-A-- _--[ w - + 'E e-ws

• ^"'4" m.

.,.. -..---. -, --..--- --- ----- - - - - - - - - - - - - - - - - - - ~ ' - - - - ~ - - - ' - * * * - " ~ ~

• 4 503:5 7/15/57 The multiplication factor kort as a function of tem-porature is shown in Fig. 10.

kort as a function of void volume is shown in Fig. 9 k,ft as a function of height of water in the core is shown in Fig. 34 The change in these variables with respect to time has been calculated as a function of size of opening. Ereaks smaller than 1/4 in. constitute no problem. In a 1/10 sq ft, medium-size break, the keff goes below unity owing to void production and reactor scram. The korf with a clean core, drops to a value of less than 0 96 approximately 1 min after the rupture, as the temperature falls. After this, the reactor is held suberitical l by voids and control rods as core uncovering proceeds. With large openings, 3 sq ft, the rapidity with which the water is abruptly. blown out of the reactor causes the water level to drop % steam, The void coefficient, approximately 0.03% Ak/k per vol is the controlling factor, and there is no return to criticality. 1 l Decay Heat Following reactor scra:n, which should be completed within 2 see after a drop of sys';em pressure, decay heat will be given off at the rate indicated In Fig. 35 As long as the core remains covered, the decay heat will be extracted by boiling water. The rate of heat loss to the boiling water is such that the core will be cooled and the temperature of the fuel will drop. l Boiling will take place in the core and fuel tempera-tures will decrease as the water approaches saturation temperature during the pressure discharge. As the water level falls in the core, the temperature of the fuel tends to rise but can be held within safe limtts by use of the safety injection system. Yanor Contt M The vapor container is designed to retain all vapors, gases, liquids and solid materials released as a result of a loss of coolant accident. The maximum loss of coolant accident employed in the vapor container design consists of: Complete severance of one 20 in. main coolant line, with two open pipe ends Simultaneous rupture of one secondary main steam line inside the vapor container. The placement of each main coolant Joop in a separate concrete shielded com-partment ord the installation of e r^nreturn valve in the main steam line from each steam generator limit this part of the accident to the rupturing of a single secondary main steam line i 700R ORGE. t

u, 503:6 7/15/57 Detachment of an object or metal frag =ent from the pressurised system in such a way that it acquires kinetic energy, which, unless restrained or stopped by a barrier, might perforate the steel shell of the containment vessel, thus releasing contaminated vapor following the loss of water accident. Fig. 36 shows the initial pressure transient follow-186,000 lb of fluid frca the main coolant ing the release of system and one secondary coolant ci>cuit into the net volume The raximum differ-of the vapor container of %0,000 cu ft. setial pressure between the coneyete compartment and the vapor container is 6 pai, and this pressure is reached in 0.2 sec. A port area of 400 sq ft in any one loop shield compartment is provided to limit the pressure differential across the concrete The concrete walls are designed for a max-walls to this valva. inea differential pressure of 8 psi. All coolant is released from the main coclant system within approximately 18 see and equilibrium is at?ained inside the vapor centainer at a maximum pressure of 34.5 pei or 49 2 psia. The corresponding 6 vaportemperatureisbege9handtheenergyreleasedis94x10 Btu. i Fig. 37 shows the long-time effect after the release of During the vapor and initial p? essure rise to 34.5 psi gsge. first 2 hr, there is a marked decrease in pressure due to thermal l radiation and convtiction from the uninsulated vapor container shall and due to t'.1e diffusion of heat into the inner concrete Subsequently, there is a gradual decrease in pressure structure. with a small seen2dary rise, decay heat frca the reactor core. peaking in 4 hr to the continued release of The air-vapor mixture pressure with!n the vapor con-tainer after the maximum loss of coolant accident is based on the assumption that the total internal energy of the fluid remains This is based on the con-the same before and after the rupture. servation of energy relations ~ , Q = AW + M Where Q = Net heat release, Btu A = Reciprocal of mechanical equivalent W = Mechanical work performed, ft-lb AEeChangeofinternalenergy, Btu During the brief interval after the initial burst, it is assumed that there is no heat loss, or Q = 0. There is no werk done, since There-the fluid begins and ends in a state of rest, or AV = 0. fore, the internal energy before the accident is the same as that after the accident, or A E = 0 ~ PDDR DRGINAL b' s -n

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503:7 . 7/15/57 <AY The suraary of the principal data for the major loss of water accident is as follows: Main coolant pressure, uppor operating limit, psia 2,150 Average temperature main coolanti apper l operating limit, F 518 Total volume of water in main ecolant system, eu ft Reactor 1,600 Pressuriser 150 Steam generators 800 Piping $8 Pumps 20 Miscellaneous T2 Total 3,170-Total volume of steam in main soolant syste:s ou ft 110 l Total volume of water in one secondary loop, eu ft 570 Total volume of steam in one secondary loop, eu ft 190 Gross volume of vapor container, ou ft 1,020,000 Ret effective volume of vapor container, ou ft 840,000 Weight of fluids in main coolant system and one secondary aircuit, Ib 186,000 Internal energy of released fluids, Btu 94,000,000 l Yapor flashed from main so61ani, per sent 31 Final pressure, psia Yaper 29 2 Air

2QaQ, Total 49 2 Total, psi gage 34.5 Final temperature, F 249 I

P00R BRGlWL u-

l 505:4 Draft-7/10/57 l l i and beyond that point all buildings are shielded by hills. r All buildings east of the river and within one mile of the site are l owned in fee by Yankee Atomic Electric Company or New England Power Company and are considered to be under administrative control of those two companies. Vapor Container Leakage and Air-borne Radiati_o_n In the hypothetical accident, 20 per cent of the gaseous and volatile fission products are assumed to be homogeneously dispersed in the vapor container. Leakage from the vapor container at the assumed leak rate will release these fis-sion products to the atmosphere and, under certain meteorological conditions, they can be carried to populated areas where they may be inhaled or ingested. j Of the volatile and non-volatile fission products in the core, radio-l iodine and radio-strontium provide the controlling activities with respect to the 1 l inhalation dose, with iodine being selectively absorbed by the thyroid and stron-tium by the bone. For the purpose of this report, it has been conservatively assumed that 20 per cent of all the iodine and strontium are released from the core even though the release of strontium has been reported more nearly 1 to 5 per cent. The total activity of iodine and strontium assumed to be in the core is: Activity Curies ( Iodine - 131 1.0 x 107 Iodine - 132 1.6 x 107 l Iodine - 133 2 3 x 107 l Iodine - 134 2.7 x 107 l Iodine - 135 2.1 x 107 l Strontium - 89 1.3 x 107 Strontium - 90 2 4 x 105 Based on KAPL - 1178, it has been shown that the integrated 60-day dose to the thyroid from the inhalation of iodine - 131 is approximately a factor of 10 greater than the dose to the bone from the inhalation of a curie-equivalent ( of strontium - 89 Since the radio-iodine activity as icdine - 131 equivalent is approximately 1.8 x 107 curies as compared to the strontium - 89 activity of \\

e 505s5 Draft-7/10/57 7 / 1 3 x 10 curies, and since the dose to the bone due to strontium - 89 is compar-able to the dose due to strontiun - 90, the iodine - 131 dose to thyroid was selected as the controlling dose. The total radio-iodine activity amanating from the vapor container is assimied to have a concentration of 2.8 x 107 microcuries per eu ft. Based on the Sutton Continuous Point Source Equation and using in-valley meteorological condi-tions presented in Professor Austin's report the concentrations of radio-iodine, as iodine-131 equivalent,100 ft below the center of a radioactive cloud, which is over the nearest inhabited area 4,000 ft away, are as follows: Radio-Iodine Meteorological Wind Yelocity, Concentration, Condition Fps Microcuries/ml Inversion 33 1.6 x 10 Moderate Lapse 19.7 6.9 x 10-9 t l Unstable 16 4 2 3 x 10-9 i This tabulation and additional meteorological information indicate l that the highest concentration of activity would occur under an inversion condi-tion with a low velocity down-valley air movement. If the accident occurs under these conditions, the leading edge of the radioactive cloud reaches the nearest inhabited area approximately 20 minutes after release of fission products from l the vapor container begins. The once-in-a-lifetime off-site dose for ingestion and inhalation of air-borne radioactivity has not yet been established by the AEC, and there exists some difference of opinion on the subject. Lacking a definitive allowable dose, values suggested by K. Z. Morgan, W. S. Snyder, and Mary R. Ford in their paper, l Mexistaa Permissible Concentration of Radioisotopes in Air and Water for Short Period Exposure, presented in 1955 at the Geneva Conference on the Peaceful Uses of Atomic Energy, have been adopted. These are: l +-+v m. ,g--- -,y w-y

J 50586 Draft-7/10/57 / Mavinne Permissible Radio-Iodine Dose Criterion Concentration for 8 Hr Exposure, Following Exposure Microcuries/ml 8 0 3 rem in week 7.0 x 10 6 15 7 rem in year 1.7 x 10-150 rem in 70 yr 1.7 x 10-5 Dr. Shields Varren has stated that he believes a dose of 50 rem to the thyroid may show clinically detectable effects, while a dose of 15.7 rem would probably provide no clinical indication. On this basis,15.7 rem in the year following exposure has been taken as the off-site, once-in-a-lifetime internal dose. Based on the assumption that a person is 4,000 ft from the plant and 100 ft below the center of the radioactive cloud, and taking no credit for radio-active decay, the doses received under various meteorological conditions are as follows: ( Thyroid Dose, Meteorological rem in Year _, Condi_ti_on_ _ Following Exposure l Inversion 15 l Moderate Lapse 0.064 l Unstable 0.021 Comparison of this tabulation with the 15 7 rem dose limit adopted shows that in all cases the dose received in 8 hours is below the limit. Thus, from this analysis it is clear that, even under the worst meteorological conditions, the hypothetical accident does not result in excessive concentrations of radio-activity in the nearest inhabited area. / ^ \\

i 506:2 i Draft-7/10/57 i / s' Among the mechanical accidents that have been analyzed is oz.e caused by a break in a 20 inch main coolant line at the worst possible location and involving loss of all veter from the main coolant syst m. This is considered to be the==Munn_ credible accident. In none of these accidents is there any melting of the core, any release of gaseous and volatile fission prducts to the vapor container, nor any hazard to the public. However, an analysis has been made of a hypothetical accident in which core melting and fission product release are asstaned. An accident has been examined in which it is assmed that a large break occurs in the main } coolant syste; virtually all water is lost from the system; partial core meltdown occurs; and 20 per cent of the gaseous and volatile fission products are released to the vapor container. The analysis shows that there would be no hazard to the general public because of direct radiation fra the vapor container. Since the vapor container has a finite leak rate, see of the l [ fission products may escape to the atmosphere and, under certain meteorological conditions, the escaping fission products may be carried to nearby inhabited i areas. At the nearest commmity, however, an 8 hour exposure to the indicated concentration of radioactivity, under the most unfavorable meteomlogical oon-ditions, would result in less than tolerable once-in-a-lifetime inhalation and ingestion doses. Iankee Atomic Electric Capany, therefore, concludes that this reactor can be operated without undue hazard to the public health and safety. l t I i \\ ,}}