ML20085E820
| ML20085E820 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 05/31/1965 |
| From: | SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
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| References | |
| FOIA-91-17 NUDOCS 9110210211 | |
| Download: ML20085E820 (35) | |
Text
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SUFPLuilliT 110. I to SAFEGUARIX3 REPORT FOR THE SAXT0!i REACTOR PARTIAL PLUT0!i1Uli CORE II May 1965 l
l i
I 9110210211 910~4 4 PDR FOIA fiEFOK91-J 7 PDR
,V
In order to prtvide a basis for evaluating thu conservatism of Question #1 the parameters used in the accident evaluation sections of the report, provide verification that the physica parameters measured in the critical experiment at VREC are at least as conservative as those assumed for the accident evaluations.
In addition, verify that the proposed loading vill be with a central plutonium region.
Answer The series of critical experiments outlined in the Safeguards Report for the Partial Plutonium Core 'JI is nov in progress at the Westinghouse Reactor Evaluation Cer er (VREC). Although the entire series is not yet completed, the results obtained to date show that experiment and nnalysis are in excellent agree-ment and verify that a conservative approach was followed in the design of the Partial Plutonium Core II. While additionni experiments and data processing and reduction are continuing.
the program is sufficiently complete to be able to state thatt (a) Any data and results obtained in the future are not expected to significantly alter the above conclusions and (b) The initial core loading vill be with the nine plutonium enriched fuel assemblies in the center of the core.
ins preliminary results of the criticals which are available are summarised below. The experimental prcgram and series of criticals being conducted at the WREC are outlined in Table 1-1, Predictions as to the number of fuel rods required for critica-lity, calculated k,ff and corresponding boron' concentrations are included in this tab 3e. The status of the experimental program of Table 1-1 is shown in the following list:
l-1
Configuration Zgge Status A
J0 -One Regi n Clean Core Completed 2
1 Fuo -UO -One Region Clean Completed except 2
Cor for 1(e) 2 Pu0
'O. U0 -Two Region Completed except CleknObrePbO-UO 2
2 in for 2 (a)
Inner 9egion Puo2-l1 -One Region Completed 3
2 Boratei Core b
Pu0 -UC -UO -IW7' E861on In Progress 2
2 2
Dorated Core, Pu(.."22 Inner Pagian 5
Two Regice, UO2 Fuel in Clean Core-Completed Inner Regaon, Clean and borated Core-In Borated Progress 6
Puo2-UO -One Region To be done 2
Clean Core, Larger Pitch Reactivity Experiment Resalts The results of two critical experiments are available for comparison with predicted results. A major portion of the experiments was done with the same H/Pu ratio that vill exist in the SaYton reactor at operating temperature (Tmod '
Fuel Rods Req'd for Criticality Configuration Fuel Pitch PJQ Analysis Experiment 1(c)
Puo -UO 0 56 in.-
355 343 2
2 A(3)
U0 0.56 in.
356 346 2
Using the same cross-section data and calculational methods employed in the core design, experimentally determined values of buckling vere used to chlculate the effective multiplication factors for various inttices and fuels.
I 1-2
Cn1_culgt ed k 7
Fuel "on fi gu rat i on.
Lattice Pitch LEOPAFit X-Y PDQ (Total Buckling) (Axial Buckling)
U)2 A(3) 0.560 in.
1.00k2 1.0045 UO A(2) 0.792 in.
0 9997 p
Corrected k ff LEOPARD X-Y PDQ (Total Buckling) (Axial buckling)
Pu0 -UO 1(b) 0 560 in.
0 9950 0 9966 2
2 Pu0 -UO 1(c) 0 792 in.
1.0063 2
2 For all of these experiments, the experimental k,ff was 1;0.
Evaluatior. of k,7f for the Pu0 -UO lattices included an allowance 2
2 of OD25 which is based on previous comparisons of analysis by these methods with experimental results of a number of ihnford mixed oxide critical experiments so that (Corrected k,77 =
Calculated k,f,,- 0.025).
The value of 0 025 van selected prior to completion of the experiment sp that its selection was not influenced by prior knowledge of the experimental results of the buckling measure-ments. The excellent agreement between the analytical predic-tions and the experAmental results shows that the allowance selected was a reasonable one. No allowance was included-in the evaluation of the UO results.
2
,From the standpoint of the Saxton core design, the results of the experiments lead to the following conclusions:
1 1-3 V
_ __.__. _._ _ ______.=._ _, __-
(a) There is no need to modify the expected core lifetime or installed renetivity predictions used in the reference i
e design of the Safeguards Report.
(b) The good agreement between analysis and experiment for a vide range of !!/Pu ratios indicates that one of the most important factors of the moderator temperature coefficier.t.
the density effect, is correctly calcultted by the analytical i
methods used in the core design.
Power Peaking Results Power peaking experiments in fuel rods adjacent to vater slots have been carried out in both single region and two region cores.
Only the results of the single region cores have been analyzed to date.
In the single region experiments, a water slot was formed by removing five center fuel rods from a square lattice.
The power level in the adjacent fuel rods was measured with I
and without the vt.ter slot. Experiments were also carried out with an aluminum slab in the water slot to displace some of the l
vater. Using the various lattice characteristics, PDQ-3 l'
analyses to predict the peaking effect have been carried out and are compared with experimental measurements.
Peaking Factor Ratio Analysis / Experiment Core g,0 Slot go+AlSlot Pu0 -UO 1.0779 1.0h00 2
2 UO 1.0555 1.010h 2
These resul+,s demonstrate that the analytical methods used in the Core Il evaluation are conservative in that they over-predict the power peaking effects in water slots. These results
(
1h
t are recrementative of the actual conditions which vill be present in Core II as installed in the reactor because the peak in the core occurs within the boundary of the Fu fuel region and is l
therefore more characteristic of a single region core than peaking at the boundary of a two region core. The results of this analysis demonstrate that the hot channel factors assumed in the core design are conservative and that the initial power level shown in the Core II Safeguards Report may be raised from 21.6 MWt, probably up to 23 5 MWt. Additional testing and low power experiments vill determine the actual hot channel 1
factors and initial power level for Core II.
Boron Worth Nesults Boro.1 vorth measuremente were made in the two region core of configuration h(b). The predicted boron concentration required for a full vater height critical was 1525 ppm. The experimental results extrapolated to full vnter height conditions showed a concentration of 1550 ppm which is in excellent agreement with the prediction.
)
Kinetic parameter Results The kinetic. characteristics of single region and two region cores are presently being investigated using pulse neutron techniques.
An additional experiment has been completed for a single region Pu core which measured the neutron lifetime by measuring the reactivity change for a small addition of boron (e/25 ppm) to the moderator.
Although all of the experiments being corducted to determine the kinetic characteristics are not yet complete.
these preliminary comparisons of analyses and-experiments are I
available:
l
(
1-3 l
m r
Tror.pt Neutron Lifetime. 1 (p see)
Fuel Lattice LEOPARD PDQ Boron Pulse (1/v Poison)_
AdditI'on.
Neutron One Re61on.
0.56 in.
8.5 19.h 15.8 20.5 (calculated from 8 = 0.003L and Puo,-UO measured 6/t =
166 see-1 r,
30 3 (Calculated One Region.
0 56 in.
15 0 20.h from 6 = 0.00795 and measured 6/t =
262 sec-1 As the table shows, the values of L if calculated for the experiment by LEOPARD are much shorter than those inferred trom the experiments.
This indicates that the actual values of t for Core II will be longer than those predicted by the LEOPARD calculation and reported in the Core II Safeguards Report.
1-B
Question #2 It is proposed that some of the Pu0 fuel in Core II will 2
operate at specific power levels of up to 16 Kv/ft. To enable us to evaluate any significant safety problems associated with operation at this proposed specific power, provide a discussior, of the results of such operation involving UO fuel at the Saxton reactor.
2 Ancver:
The peak specific power level of 16 Kv/ft is a conservative design limit based upon present Westinghouse fuel element design practice and techniques. This limit is believed to be a reasonable upper boundary for the initial operation of the mixed oxide, partial plutonium core for Saxton. A great deal of experimental data exists on the successful operation of test fuels of these types (sintered pellets and vibration compacted powder) at specific power levels greatly in excess of 16 Kv/ft and even, in some cases, with significant center melting of the fuel.
The limit of 16 Kv/ft is a reasonable step up from the maximum conditions so far experienced in the Saxton core (lk.5 - 15 Kv/ft) as less than tvc uszen rods of Core II would opernte above 14.5 Kv/ft if the peak rod vere to operate at 16 Kv/ft.
Because the Saxton reactor is an experimental plant, sustained periods of operation at the maximum rated power of 23.5 MWt have not been obtained in the past. The peak specific power in any fuel rod in Saxton is dependent on a great many factors; fuel enrichment, boron concentration, control rod position and reactor power level. Therefore, the peak specific power depends on the condition of the above parameters at the time the measurement is made.
I 2-1 i
_ _ _ _ _ - - _. - _... - - - - - ~ - - - - - - -
I With the reactor above 22 Wt, the maximum specific power level of the core is nominally 13-14 Kw/ft. This number is based on the same methods that would determine the 16 Kv/ft 2imit, that is, a 10% uncertainty in the measurements and an en6tneering hot channel factor of 1.045 The highest measured 1,;.ecific power has been 13.87 Kv/ft at a reactor power level of 22 9 Wt.
When extrapolated to 23.5 Wt, a maximun of 1h.56 Kv/ft is obtained from 12.16 Kv/ft at 19.63 Wt.
With the uncertainties involved, it is not possible to say that with the reactor at 23 5 Wt that specific powers in excess of 14.5 Kv/ft have been experienced in the Saxton core. All of the peak values referred to above have occurred in the central 9 x 9 which contains experimental fuel that is licensed to operate up to 16 Kv/ft.
Successful operat, ion of fuel at or above this level has been demonstrated by several Westinghouse experiments.
Six capsules containing three fuel rod samples from the CVTR core vere irradiated in the Westinghouse Test Reactor to a maximum pover rating of 2h Kv/ft.
The capsule configuration was a 5-inch column of UO pellets,.h30 inches in diameter, 9h 1 1 5% of 2
theoretical density clad with Zircaloy-2. The capsules were all successfully irradiated with no evidence of central melting.
Two additional capsules vere irradiated in the Westinghouse Test Reacter.(2) Or.e capsule contained three fuel rods with a 38-inen fuel length and was irrad.'.ated at peak -fuel rod power levelsof17to19Kv/fttoamaximumfuelburnupof3,h50f[.
The o'.her capsule contained four fuel rods with 6-inch fuel length. Average fuel rod power levels of > 18 Kv/ft were main-tained during irradiation to 6,250 The rods contained UO 2
pellets.h30 inches in diameter and 9h 1 1.5% dense. The capsules were clad in Zirealoy-2. The capsules were successfally irradiated and indicated that thermal reactors could be operated at these high rod powere safely and successfully.
2-2
l UO fuel capsules are being u radiated in the NASA - Plum 2
Brook heector as part of the High Power, High-burnup Irradiation Program.
Puel pins containing 0.3 inch diameter pellets 96% dence with a 6-inch fuel column are clad with 30L stainless steel. The capsules are being irradiated at power ratings of 20 to 60 Kv/ft, to a maximum burnup of 80,000 MTV.
g,D Four capsules hav; been irradiated to 10,000 at a paak MTU power rating of 39 Kv/ft. Three of these irradiations were completely successful; the fourth failed due to excessive fuel melting. Approximately seventy-five percent of the cross-sectional area of the pellets was molten. The failure occurred after long exposure at high rod power.
Three capsules were irradiated in the Plum Brook Reactor in a program designed to measure the thermal conductivity of UO2 at the columnar grain growth threshold temperature.
The pins were h-1/2 inches long and 1-1/k inches in diameter. They were successfully irradiated at rod powers of 20-2h Kv/ft.
Two vibratory comnacted pins and one pelleted fuel pin vere successfully irradiated in the GETR at peak rod power of 21 Kv/ft.
The pins were 5 2 inches long and had an active fuel diameter of 56 inches. The pelleted rod was 88.3% dense while the vipac vere 81.8% and 86.7%.
In addition, GE has run some very extensive, long irraniation high power level experiments in the GETR with fuel enriched to N 20% in Pu.
Two pelletized rods with no central voids were operated at peak specific powers of % 15 5 Kv/ft and s 17.8 Kv/ft for burnup of 23,100 and 17,600 respectively. The experiments vere very successful with no adverse effects due to these operating conditioas.
(
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2-3
Based on the: experimental evidence available, the possible operation of some rods in the Pu recien of Core II at 16 Kv/ft power levels vill present no significant safety problems in the operation of Core II and is a very conservative extrapolation from the power levels already experientesd in Saxton.
T 4
I 2-k
i 4
Beferences 1.
Duncan, R. 11., " Rabbit Capsule Irradiation of U0." CYNA-IL2, 1962.
2 2.
Duncan, R. ii., "CVTR Tuel Capsule Irradiation," CVNA-153,1962.
3.
WCAP-2500, 26bO, 2689, 2732.
L.
Balfour, M. G. and Terrari, !!. M., " Irradiation of Vitratory Compacted UO Fuel Elements " WCAP-2729, p
5 Gerhart, J.
H., "The Post-Irradiation Examination of a Pu0 -00 Fast 2
2 Beactor Fuel," GEAP-3833, 1961.
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We understand that new information concerning the conductivity of Question #3 uranium dioxide at high temperatures is available. Provide a curve of uranium dioxide conductivity as a function of temperature on which these new data points are incitded.
Answer:
The attached figure is to replace Figure III-7 of the Core II Safeguards Report.
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TE&iPERATURE, UC 0
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THERMAL CONDUCTIVITY OF URANIUM DIOXIDE Eluuns III - 7 Revision 1
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Question #L The Saxton reactor is the first licensed nuclear power reactor it, which a plutonium core loading is to be used.
To enable us to evaluate a possible manner in which plutonium might be released to the environs, provide a discussion of those operating procedures which vill assure that plutonium which may be in the containment building sa contamination vill not be transported to the remainder of the site or to the environs.
In addition, discuss wh/ the limits of sensitivity of the various monitoring equipment and health physics procedures proposed are adequate to assure that 10 CFR 20 limits for plutonium vill not be exceeded.
Answer Because of the conservatine assumptions and methods used in the plutonium fuel design and the rigorous testing and inspection performed on the fuel during its manufacture, the probability of fuel clad failure throughout the planned life of Core II is very small. In addition, the fuel rods and the fuel assemblies are monitored for alpha contamination prior to shipment to Saxton so thht there is little likelihaod that tramp plutonium vill cause a contamination problem during fuel storage and loading.
In the event some plutonium contznination should be present inside the containment, there are only three methods availabic for transporting plutonium contamination from the, containment building (a), Personnel Saxton's present radiation protection procedures have proven adequate to prevent the spread of contamination from the
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L-1
~._ - - -- - - - -
containment vessel. Access to the containment vessel is allowed only under the provisions stipulated by a radiation vork permit which specifies, among other thingo, protective clothing to be worn.
Step-off pads and storage for protec-tive clothing are provided in the air lock. Monitoring of personnel for alpha contamination prior to leaving the vessel vill be accomplished as required in the radiation vork permit.
(b) Ventilation Exhaust Sinct the containment vessel has no exhaust flow during reactor operation, the installed alpha monitoring system which will be added to the present containment air activity monitors vill give a reliable history of containment vessel air activity. At a time when entry is desired, the reactor vill be shut down and the containment vessel air activity will be knovn. Ventilation exhaust flow rate vill be adjusted, if necessary, to insure that any release to the atmosphere is within the limits established by 10 CFR 20.
it is expected that the containment vessel air activity attributable to plutonium vill be belov its MPC at all times and that it vill not be necessary to regulate the containment vessel air release rate.
l (c) Liquid Effluents Liquid effluents from the containment vessel vill be handled without any change to the present vaste disposal or chemistry sampling system.
The only procedural change vill be an l
increased monitorin6 of areas for sipha contamination.
Present procedures for monitoring effluents are adequate to assure that 10 CFR 20 limits for plutonium vill not be i
exceeded.
I k-2
.\\d
After discussions between Saxton personnel and personnel at the Plutonium Recycle Test heactor, we have concluded that the problems associated with radiation protection due to plutonium are no different from those which already exist, due to the presently installed uranium fuel. As quoted from U. S. Atomic Energy Commission Research and Development Report IN-83601, pROGPESS IN PLUTONIUM UTILIZATION by Hanford Laboratories:
" Plutonium fuels have been stored and handled in the same manner as uranium fuel, and irradiated fuels have been routinely handled for specini examinations and core changes without difficulty. No unusual procedural controls have been made necessary, nor has any specia-lized operator training been required specifically as a result of using plutonium fuels in the PRTH.
"The PRTR experience has shown that the effects of plutonium fuel failures are no different than those for uranium fuels.
Emissions have been virtually limited to fission gases with no evidence of particulate vashout.
Alpha contamination, usually of primary concern in fabricating plutonium fuels, is of little concern in reactor operation, as gamma contamination governs procedurec for almost all maintenance work."
The activity concentration requirement of 10 CFR 20 for Pu-239, Pu-2h0 and Pu-2kl for radiation vorhers exposed for k0 hours
-12 per veek, is a maximum airborne concentration of 2.0 x 10 uc/cc.
This activity level, defined as the re.dioactivity concentration guide for a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week (RCG/h0), represents that concentration of plutonium in air-to which a " standard man" may be exposed for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> p<r veek, 50 weeks per year for a total period of 50 years to that at the end of 50 years the total activity fixed in
(
the " standard man's" body vill not exceed the recommended maximum permissible body burden (MPDD) of 0.Ok pc of plutonium.
This MPBB as set by both the International Commissicn on Radio-logical Protection and the National Commission on Radiological Protection is defined as that amount of material which may be maintained indefinitely in the body of a " standard man" without producing any rignificant somatic or genetic effects throughout the life of the " standard man".
The sensitivities of the air particulate monitors, both the moving filter vapor container monitor and the fixed filter portable monitors, have been revised slightly from those given in the Core II Safeguards Report. The minimum sensitivity for these instruments for a 1-hour sample period and following a delay period (about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) to remove the Fadon-Thoron back-ground is given as 2.5 x 10-12 c/ce. As stated before, containment access is not possible during power so that detection of this level of activity is more than adequate to assure that centainment vessel purge prior to entry will not produce off-site plutonium levels above 10 CFR 20 levels. Containment vessel purge procedures can be altered, if required, if the containment vessel concentration is significantly above the limit of detection. Purge of the containment vessel prior to entry will assura adequate working conditions upon entry.
The portable air particulate monitors can be moved throughout various areas of the plant as required to sample for airborne activity. Alpha monitoring during such operations as main cuolant sampling in the sample room or analysis vork in the radiochemical laboratory is provided by these instruments.
These instruments are capable of detecting near RPG/k0 levels
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h-b
-___--_n
vith the 6-hour delay for Radon-Thoron decay. A more rapid readout of higher concentrations may also be obtained.
Following a one-hour sampling time and the presence of a high Padon-Thoron background (600 cpm) the minimum sensitivity is
~
about 2 x 10 pe/cc which is a factor of 100 above RPG/ko.
If this high plutonium concentration vere detected, vork in the area could be suspended and corrective action initiated. Workers exposed to these higher than RPG/k0 concentrations could be restricted from working in possibly contaminated areas for a period of time to allow averaging of this exposure.
For example, a one-hour exposure to 100 x RPG/k0 concentration is equivalent to about 2-1/2 vorking weekn at RPG/40 so that return to work with RPG/h0 concentrations would be permissible after 2-1/2 weeks of no exposure to plutonium.
Higher concentrations of plutonium can be detected in even shorter periods of time due to the fact that the count rate of the sample, due to Pu, increases linearly with exposure time and is proportional to the concentration. For a high Radon-Thoron background of 2h0 cpm
~9 a plutonium concentration of 1 x 10 uc/cc can be detected after a five minute sample time.
Exposure to 1 x 10'9 pc/cc or 500 x RPB/h0 for five minutes is almost equal to a LO-hour exposure to RPG/kO, so that one week of non-exposure to plutonium would then allow return to work in RPG/h0 levels.
The procedure of curtailing vork followinE exposure to levels above RPG/40 is a standard practice and where combined with the instrument sensitivities described will assure that personnel exposures are vell within the limits of 10 CPR 20.
I h-5 i
l
Question #5 In the accident analysis section of the report it is stated that each accident vat malyred using that combination of system parameters which would give the most serious conse-quences.
Indicate the manner in whics. it can be assured that the most adverse combinstion of parameters has been seJected, and provide the range of parameters considered for enth accident analysis.
Answer:
Two basic premises which underly accident and reactor transient analyses are to develop realistic yet conservative models and then to apply these models using reslistio yet conservative parameters. Analog computers are normally used to simulate the reactor. The selection of the basic parameters depends on 7
the transient being studied. The parametero are chosen on the basis of adding the most reactivity to the transient or prov ding the least help in limiting or preventing the transiunt.
As a a;ncific example, the detailed reasoning for the choice of paru eters of the control rod withdrawal at power accident are outlined below.
During this transient, heating of the fuel and the moder ator vill add negative reactivity to the systems and tend to depres the transieht. For this reason, the moderator coeffi-cient assumed was smaller than the expected value and would correspond to a boron concentration in excess of 2000 ppm. The Dcppler coefficient chosen was less than expected values.
Overpower scram initiation is ' set to trip at 115% of nominal full pouer and is a redundant circuit to assure reliability, llovever, errors in fixing set points and in power measurements are assumed to delay scram initiation until a power level of 122% is reached.
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5-1
Upon initiation of scram, an instrumentation delay of 0 5 sec.
is assumed to delay rod motion. Actual instrumentation delay times are less than 0 3 seconds.
A further delay in scram of 0.6 seconds is assumed for control rod motion 11. a region of small effectiveness and 0 9 seconds is assumed for completion of the rod insertion into the core. Actual measured control rod drop times for Saxton are on the order of 0 9 seconds or less so the actual scram completion time vill be about 1.2 seconds or less compared to the 2.0 seconds assumed in the analysis.
Control rod serm vorth upon insertion was assumed as 0.02 Ak/h.
The nominal operating conditions of this, accident, that is early in life with large hot channel factors and high boron concentrations (1500-2000 ppm), vill result in about 0.15-0.18 Ak/k reactivity in control rods out of the core.
Even if the most reactive rod (0.05 ok/k) were to stick, the reactivity insertion by control rods vould be about 0.10 dk/h. The only time that a reactivity insertion on the order of 0.02 ok/h would be possible vould be very early in core life at very lov boron concentrations (rodded control) which is a condition not compatible with the moderator coefficient chosen for the analysis.
A final conservative assumption is An the reactivity insertion rate of the control rods during withdrawal. The maximum insertion rate of the most reactive rod group (the two inner rods or the four outer rods) is 7 25 x 10-5 Ak/k/sec. anci assumes the control rods to be in the most reactive region and moving at the maximum withdrawal speed. The value of 2 5 x 10' Ak/k/sec.
which was assumed for this analysis __is_a much larger rate than could nossibly be experienced by the reactor during this transient.
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5-2 l'
)
The same general reasoning has been applied to the other transients and accidents analyzed. The following tables present a comparison of the parameters assumed for the analyses and those which might be expected to exist in the reactor.
R,o_d Withdrawal. Cold Startup I.
o Value Used Expectato Value 1.
Moderator Temperature Coefficient
+ 0 3 x 10' ak/k/*F 0.0 ok/k/'F (at 70'F, 2000 ppm boron) 2.
Doppler Coefficiet.t
- 1.1 x 10 Ak/k/'F
- 2.0 x 10~$ ok/k/*F
~$
3 Reactor Suberitical by 0.02 ok/h
>.05 Ak h.
Overpower Scram Initiation 122%
115%
5 Control Rod Drop Time 1 5 see.
< 0 9 sec.
6.
Scratu Beactivity Insertic. by Rods 0.02 ok/k 0.1 - 0.15 Ak/k
-5 7
Reactivity Insertion Rate 2 5 x 10 Ak/k/sec.
< 7.25 x 10 Ak/k/ sci II.
Hod Withdraval. Ret 9tartun 1.
Moderator Temperature Coefficient
- 2 7 x 10 ak/k/'F
- 3.0 x 10 ok/k/ F (at 530'F, 2000 ppm bcron)
-5 2.
Doppler Coefficient
- 1.0 x 10' ek/k/F
- 1.3 x 10 ok/k/ F l
Same as for Case I
- 3. Thru 7.
l l
l III.
Rod Withdrawal, At Pover_
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~
1.
Moderator Temperature Coefficient
- 2 7 x 10 ak/k/*F-
- 3.0 x 10 Ak/k/*F
~0
-5 2.
Doppler Coefficient
- 1.0 x 10 Lk/k/*F
- 1.1 x 10 ok/k/'F 3.
TrimaryCoolantPressure[aH-DNB 2050 psi
-2000 psi (For DNB Calcula.lons)
{Q-DNB 1950 psi
(=
5-3
_. _ __._ _. _ _____ ~.-
... _ ~ _ _ _ _ - _ _.
III. Rod Withdrawal. At Pover (Cont'd) i Value Used Expected Value-
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- l..
Instrument Delay Time 0 5 sec.
< 0.3 see.
Control Rod Drop Tine 1 5 sec.
< 0.9 sec.
5:
Reactor Power Level, % or Neminal 103%
95-100%
122%
115%
6.
Overpower Ceram In e
3 by Rods 0 02 AK/k 0.10-0.15 Ak/k 7.
Scram Etactivity 1r 8.
Maximum Specific Pw 16 5 Kv/ft 1h-15 Kv/ft t
IV.
Sleam Break k.0 x 10 Ak/k/'F h.1 x 10' Ak/k/'T 1.
Moderator Temperature Coefficient (Worst Case, End of Life -
O ppm Boron Concentration) 2.
Safety Injection Functions No Yes V.
Loss of Flov Accidgrd
~
2 7 x 10' Ak/k/'T 3 0 x 10 Ak/k/'F 1.
Moderator Temperature Coefficient t
l 2.
Control Rod Drop Time 1 5 sec.
< 0 9 Luc.
3.
Reactor Power Level - % of Nominal 103%
95-100%
k.
Scram Reactivity Insertion of Rods 0.02 Ak/k 0.10-0.15 Ak/k 5.
Maximum Fuel Power Density 16.5 Kv/ft ih-15 Kv/ft 6.
Primary Coolant Pressure foibDNB 2050 psi 2000 pai (For DNB Calculations)
J, Q-DNB 1950 psi l
.5-h
. v 4
.-r,,
"v-
.--,.<m
.u-1
-w-~
- - ~ - -
_ - ~ ~._.
1
+
- - In the report it is stated that the results of the chemical Quest 2cn #6 shim experiment program have demonstrated that a boron release accident as originally postulated is not credible and, accordingly, the requirements of an unexplained reactivity l
limit are no longer required.
Provide a descripticn of the resulu of the chemical shim work at Saxton so that we may evaluate the safety considerations of deleting this requirement.
Answer To answer this question, copies of WCAP-2599 "The Saxton Chemient Shim Experiment," are submitted herevida.
l l
t 8
j 6-1
=
-w
=v-
,aspm y.g.ar-e gwey9
.w---y c.y y en.m9 y
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-n-pey5,-
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p
-.-ey,w
,g yy
,.rf"9-%YWWWT-v"'
- W Ww'WM' Th wN'--w4*-Wh*Wew'G"W'-M-98W TYVW9"1'W--
Question #7 - Provide an estimate of the amount of plutonium that might be released to the containment in the event of the " maximum hypothetical accident" to enable a more definitive evaluation of the consequences of this accident.
In addition, provide an evaluation of the amount of plutonium that might subsequently reach the environs.
Ansver:
A concervative evaluation of the amount of plutonium oxide in the containment vessel following the maximum hypothetical.
accident has been completed. The maximum amount of PuO that 2
could be in the containment vessel vould be less than 50 mg and maximum smount available for leakage in the form of an aerosol vould be less than 35 mg.
These amounts vould result in a maximum two hour inhalation exposure at the site boundary of less than 10' of the permissible body burden or plutonium.
e Evaluation of the maximum hypothetical accident for the Saxton reactor partial plutonium Core II considered"a' condition in which the emergency systems to provide core cooling did not function following a less-of-coolant accident. For such a situation, decay heat generated in the core vill result in extensive melting of the clad and internal suppcrts and vill eventually cause the core to collapse into the bottom of the reactor vessel. This situation vill expose a large amount of fuel surface to the atmosphere in the reactor vessel and the high temperatures involved vill cause volatilization of the fuel, h
The amount of fuel which can be volatilized under these circum-stances vill be severely limited because of the geometry of the system, the presence of an air atmosphere and the fact that the fuel may be partially vetted by the molten clad or even partly submerged in a pool of molten cladding and structures.
d
(
T-1
As shown in Figure 7-1, experimental evide.nce(1' indicates that the vapor pressures of plutonium dioxide and uranium dioxide follow the same curne as a function of the reciprocal of the absolute temperature as measured in a vacuum. Also shown on this figure are the experimental data (3) for the vapor pressure of pug I"
2 an air atmosphere. As would be expected, the presence of an air atmosphere reduces the vapor pressure below that measured in a vacuum. For this calculstion, it will be assumed that pug 2 ""O UO have the same vapor pressure - temperature relationship in 2
an air atmosphere.
An empirical relationship has been developed which correlates the weight 1Lss rate, vapor pressure, absolute temperature and molecular weight for a system vaporizing a suostance in an insulated crucible with a small opening. The relationship is
}
as follows:
-9 P(atm) = 6.267 x 10 p/Ka (1)
P =
partial pressure of the effusing species, htm veight loss rate - mg/hr u =
Klausing factor (K = 1/(1 + 0 5 L/R)]
K
=
2 effective orifice area - cm (730 cm )
a =
absolute temperature
'K T
=
M molecular veicht
=
orifice length (assumed as 1 ft.)
L =
orifice radius (1/2 ft. )
R =
The Klausing factor is applied because the actual orifice has some finite physiccl dimensions while the correlation was developed for an ideal orifice. The molecular weight of the fuel vill be taken as an average of 271.
Using these constants, Eq. (1) becomes:
(
7-2 l
I l
4 p=
x 9 59 x 10 mg/hr (2)
If an average temperature of N 2h00'K is assumed for the core material which is sicviv heating throughout the meltdown, the corresponding pressure is N 10' atm. The weight loss rate is then:
10 11 u=
x 9 59 x 10 mg/hr 2h00 p=
196 x 10 mg/hr u=
19.6 g/hr The Pu0 in the core is 6.6 v/o of the central nine assemblies.
2
(
As there are 21 assemblier in the core, the average Pu02 "/
is 6.6 x
= 2 5 v/o.
If it is assumed that the volatilized material has the same weight fraction of PuO, then the 2
U 9
hr.
PuO2 The value of p 0.49 6m/hr vould be the limiting value if Pu02 the entire reactor vessel were at the temperature assumed for the hot fuel as us the case in the experiments of Reference (h).
Most of the reactor vessel vill be at temperatures considerably lower (500-600'F) than the k000 F used for the average of the fuel mixture.
Because of this situation, a great deal of the vaporized fuel material vill not leave the reactor vessel but vill plate-out on the relatively cold internal s Irfaces of the vessel.
k 7-3 I
4 The surface area of the inside of the reactor vessel which might 5
be available for plate-out is estimated at about 2 5 x 10 cm.
2 The cross-sectional area of the main coolant I.f pe is about 730 cm
-3 so that the ratio is about 3 x 10 Therefore, a conservative estinate of the rate at which the vaporized plutonium oxide lea.es the break would be 1% (three times the area ratio) of the rate calculated by Eqt.ation (2).
The rate at which Puo2 leaves the vessel is therefore k.9 mg/hr. In the unlikely event that a condition of no core cooling vere to occur, it is not expected that it vould exist for more than a few hours so that the total amount of Pu0 release to the containment vessel vould 2
be less than 50 mg.
Because of the large amount of relatively cold surface available in the containment vessel for plate-oat of the volatilized material, it is not expected that there vill be any significant airborne concentration of Puo which might cause an inhalation 2
hazard. As an upper limit on the evaluation, it vill be assumed that all of the Pu0 leaving the reactor vessel is of the proper 2
particle size to remain in the containment atmosphere as an aerosol.
Studies (E} on the reduction rate of the mass-concentration of aerosols indicates that a half life of h-5 hours is typical.
Assuming a half life of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, an equilibrium state for the amount of pug in aer sol f rm is soon reached. The equilibrium 2
amount is calculated as follows:
q
(
/
7-h k
6 R - AN (3)
=
N = amount of Pu0 in the aerosol, mg l; 2
release rate - mg /hr R
=
decay constant fF A =
Solution of equation (3) yields the familiar result N(g )=
{1 - e~Y ) + N (h)
~
g)e At equilibrium with N(g) = 0 f
(5)
N,q a As shown before R = 4.9 mg/hr A =
35.h mg b693 N
=
=
An equilibrium amount of 35.4 mg Puo gives a t tal weight of Pu 2
of 31.2 mg.
assuming that the Pu aerosol has the same isotopic concentrations as were present in the fuel, we have 2.*( ng of Pu-2hD and 26 5 mg Pu-239 These weights give activities of 0.6 1
-3
-3 x 10 curies of Pu-2h0 and 1 77 x 10 curies of Pu-239 or a total
-3 of 2.37 x 10 curies of Pu.
The original off site inhalation hazards for the Saxton maximum hypothetical accident have resulted in a Technical Specification containment leak rate limit of 0.k", of the contained volume per day. This leak rate is based on a design pressure of 30 psig existing throughout the accident. The design pressure was k
T-5
based on the total energy release of the reactor coolant at saturated water conditions and 2000 psi. The actual energy content of the reacto" coolant is considerably less than that nssumed previously. xiso, Figure 506.1 in the Final Hazards Report for Saxton indicats that the containment pressure vill drop very rapidly from the initial peak.
Using the Generalized Gaussian dispersions equation for a ground level point source and assumin6 Pasquill type "F" conditions with a vind speed of 1 meter per second, a dispersion factor X " = 6 x 10-3,-2 is obtained at the exclusion radius of 300 meters. Additional credit can be taken because of dispersion and, dilution in the wake of the containment building so that I " becomes:
q q
wo o + cA y "z
=
g building dilution factor cA
=
2 building cross section = 250 m A
=
factor ranging from 0.5 to 2 depending on the building, c
=
assumed as 0 5 Therefore:
- "= 3.h2 x 10-3
-2 m
Q Fall-out of particles as the plume travels vill also provide additional-reduction of the plume concentration. This reduction factor can be estimated for this case using the method proposed by Chamberlain, f
T-6 1
l l
The deposition reduction factor (DRF) is given by
~
o (DRF) = exp (- f [v t
dx) u z
For this case the release height, h = 0 V
(DRF) = exp (-
y h
dp) u z
300 For Pasquill "F" conditions hdxisabout300.
Data from Stewart (
indicates that the deposition velocity, v. for 0
plutonium odde with particle sizes to be expected in the aerosol size range is in the range of 3-5 cm/sec.
If a value of v = b cm/sec is chosen then:
l exp (- f 100 x 300)
DRF =
exp (- 7.64)
DRF =
~
h.92 x 10 DRF
=
The plume concentration of Pu at the site boundary in then given by:
-3 Q x 3 k2 x 10 x 4.92 x 10' x
=
300
-3 x
l'1
- 10
- /8" 2 37 x 10 x
Q
=
x 600 1.1 x 10' uc/sec Q
=
3 1.85 x 10 pe/m
=
X300 7-7
3 If an active adult bretsthing rate of 1.25 m /hr and an uptake retention factor of.25(9} are assumed, the two hour uptake of Pu is:
-0
~
D
= 4.85 x 10 x 2 x 1.25 x.25 = 1.16 x 10 pc g
The maximum permissible body burden of Pu is 0.0h ue 10),,
~9 the accident uptake is 2 9 x 10 below the permissible body burden. Because of the large deposition fraction within the exclusion radius, there vill be no significant plutonium released beyond the site boundar.
/
7-8
s Referene_es 1.
Ackerman, R., "The High Temperature, High Vacuum Vaporization and Thermodynamic Properties of Uranium Dioxide," ANL-5482 (1955).
2.
Mulford, R. N.
R., and L. E. Lamar, "The Volatility of Plutonium Oxide,"
Plutonium 1960, Cleaver-Hume Press., Ltd. London, 1961.
3.
Paprocki, S. J. et. al., "The Volatility of PuO in Nonreducing 2
Atmospheres," BMI-1591 (1962).
4.
Mulford, R. N. R. and L. E. Lamar, op cit 5.
Whytlaw-Gray, R. and H. J. Patterson, " Smoke, A Study of Aerosol Disperse Systems," Edward Arnold Co., London, 1932.
6.
Gifford, F. A., Jr., " Atmospheric Dispersion Calculations Using the Generalized Gaussian Plume Model," Nuclear Safety, Vol. 2, No, k, June 1961.
7.
Chamberlain, A. C., " Aspects of Travel and Deposition of Aerosol and Vapor Clouds," A.E.R.E. HP/R-1261 (1955).
8.
Stewart, K., "The Particulate Material Formed by the Oxidation of Plutonium,"
Progress in Nuclear Energy Series IV, Volume 5, The MacMillan Co.,N.Y. 1963 9
Stewart, K., op cit, pg. 575 10.
" Report of Co=mittee II on Permissible Dose for Internal Radiatica,"
(1959) Health Physics, Vol. 3, June 1960.
(
T-9 1
(
1 rigure 7-1 Vapor Pressure of Uranium and Plutonian Dioxidos vs Tmporature \\ \\A 5
\\
-3 1
e4 o
\\
,/
K
-5 N
-6 7
A 10, Vacuum, Bef. 1
\\
2
-7 M ' V"0"#*' P"I* 2
\\
A U
2 O pug ' Alf' E"I' 3 2
\\,
\\
-8 N
o A
-9 e
i i
i 2
3 4 '
5
'6
-4
-1 1/T x 10
'K i
N l
I Question #8 Since plutonium requires somewhat more stringent consideration of the reactivity requirements for fuel storage than uranium, provide an evaluation of the adequacy of the Saxton fuel storage facilities for plutonium fuel.
Answer:
Evaluatien of the adequacy c f the Saxton fuel storage facili-ties for the plutonium enriched fuel were carried out using PDQ-3 calculations to determine the suberitical multiplication factors of the UO and Pu0 UO fuel assemblies when installed 2
2 2
in the fuel storage racks.
The physical dimensions of the fuel storage racks consist of a 3 2-inch surface +to-surface fuel element separation in each row and n 12-inch separation between rows. Ambient water temperature conditions with 0 ppm of boren were assumed for the esiculation although the fuel stcrage water is actually borated.
The results of the calculatJons are shown below:
Fuel Calculated k 7
UO 0.838 2
Pu0 -UO 0.898 2
2 The calculated k,77 for the Pu0 -UO fuel includt.s a correction 2
2 to account for the discrepancy between the experimental results of the WREC criticals and the predicted analytical results.
From the data in the above table, it is concluded that there vill be no criticality problems or hazards in storing either type of fuel assembly at Saxton.
i 8-1
-