ML20085D909
| ML20085D909 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 02/28/1965 |
| From: | SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
|
| References | |
| FOIA-91-17 749, NUDOCS 9110170260 | |
| Download: ML20085D909 (15) | |
Text
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SUPPLMENT NO.1 to i
SAFQUARDS REPORT FOR Tile SUPERCRITICAL TECllN0 JGY PROGR/Ji 0F SAXToti NUCLEAR J
EXPERIMENTAL CORPORATION FIVE-YEAR --
RESEARCH AUD DEVELOPMEltf PROGRAM s
February 1965 9110170260 910424 PDR FOIA DEKOK91 --17 oDR
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Q,uestion 1:
If operatica of the loop with its fuel generating 136 W of power is contemplated, describe the criteria which vill be used to select the operating parameters.
Answer:
Operation of the supercritical loop vith its fuel generating 136 KWt is not contemplated at the present time.- The maximum output of the 21 v/o enriched fuel assembly under normal operating conditions with the reactor at 23 5 Wt is expected to be 80 Wt.
However, in order to accommodate a possible increase in the Saxton power level from 23 5 Wt to 28 Wt -
(28.0/f.35=1.19)plusallowforaloop-powerlevelincrease due to operation of the loop in the hot standby condition (20%
increase), the maximum power level requested for the loop technical specification is 115 Wt (80 x 1.20 x 1.19).
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Question 2:
Under the condition of (1) loss of flow, and (2) loss of coolant, provide a graph of " hot spot" fuel cladding tempera-ture, loop flow and loop pressure as a function of time assuming normal operation of both the loop safety injection system and the reactor control and safety systems. Provide sufficient infomation to verify that clad temperature vill decrease following initiation of the safety injection system.
If, under the above circumstances, the safety injection system were to fail, state the amount of time before fuel chd failure vould occur.
Answer:
(1)
The attached figure gives 1.he requested time responses of the " hot spot" clad temperature, loop flov and loop precsure for normal functioning of the safety injection.
The figure applies to either the loss of flow or loss of coolant accident and shows that the safety injection system would provide cooling rapidly enough to prevent clad failure at the hot spot in either (defined as rer :h-ing a clad temperature of d 2000*F). The following assumptions were used as a basis for obtaining the re alts:
a)
The fuel assembly generates full power during the first four seconds after the accident.
b)
The fuel rods act as adiabatic rods and absorb all heat generated for the first three seconds (no heat loss from rods).
c)
Safety injection flow reaches the fuel in three seconds. One second of the delay is piping delay and two seconds is for instrumentation and valve opening delays. It is expected that the measured delays vill be less than three seconds. -
2-1
1 d)
The safety injection flow rate vill be at least a minimum of 5 gpm (2500 lb/hr) for all accidents.
If the accident is a loss of flow, the standby pump (capacity # 5 gpm) vill start in less than two seconds so that the total flow vould be 10 gPm.
For a loss of coolant break downstream of the pressure tube, the flow in the tube vould continue because of the continued operation of the loop pumps.
For a loss of coolant break upstream of the pressure tube (the vorst case analyzed), flow in the tube vould continue for 0 7 seconds during blowdown.
No credit is taken for any cooling effect of this flow in determining the temperature rise of the clad hot spot during the first three seconds.
(Boiling would begin at es 0.02 second following loss of coolant. )
e)
Reactor scram is completed between the third and fourth seconds after the accident. Three seconds is assumed as the delay time from initiation of 1he scram signal until the rods begin to move in a region of effectiveness and one second ir assumed for com-pletion of rod motion. Measured delay times are expected-to be less than these values.
f)
Following the four secords of full power, the rods generate an average of 15 per cent of full power _for the next 10 seconds and as average of 7 per cent for the next 90 seconds. The o.verage bulk temperature of the fluid in the pressure tube is taken as 1000*F for the first 9 seconds and es 500*F therea'fter. Tbds is conservative as the acfety injection water is at about 120*F.
2-2
a g)
Safety injection flow will be available for a mirtimum of 180 seconds after which the system is automatically switched to service water. The service water can be continued or the pressur9 tube isolated and the emergency condenser put into service. Tnis condenser is designed for 6 per cent of full power at 300'T so tha c it could be placed into service after about 50 seconds. For the loss of flow accident, the flow from the standby pump would be more than enough to cool the fuel after depletion of the safety injection reservoir.
(2)
For the case of malfunctioning of the safety injection system, the time until fuel fa$1ure vould be about 5 seconds in the vorst case (lor.s of coolant). Fuel failure is defined as the clad reaching a temperature of about 2000*F. For the case of loss of flow, the time vould be somewhat longer, but not significantly.
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Question 3:
With reference to the Reactor Head Nozzle Cooling System, if the 500 lb/hr coolant flow vere to stop, at what rate vould the temperature of the pressure vessel head nozzle increase?
lbw much time vould be available to take corrective action in this situation, and what actions vould be taken?
Ansver:
For a condition of complete loss of bleed flow, the maximum rate at which the temperature o' the reactor head nozzle vould increace is approximately 30*F par minute. Under these circicatances, a period of about three minutec vould be available to take corrective action. The bleed controller indicator (FIC-X9) has been modified to include a low flow alan, at 50%
of set point and heater shut off cignal at 0% of the 500 lb/hr set point. The corructive action required in this situation is supplied by the automatic function of this instrument.
Upon initiation of this alarm, the loop operator will check to see if the bleed flow has gone below the alarm set point and if the nozzle temperature is rising. The heater pmier vill be reduced if the low flow condition continues and the nozzle temperature increases.
3-1 i
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Quection L:
Su=nri:'e the pre-operational tests which vill be performed on the super:ritical loop.
Answer:
A ary extensi"e and thorough program of pre-operational 1 stinC is planned for the supercritical loop. A brief sumary of the testing is given below.
a)
A complete hydrostatic test of the high and low pressure sections of the loop viu be conducted includin6 a pneu-ntic test of t he pot;itive pressure portions of the off-gas system. All 1 op high pressure velds are radio-graphed. All loop low pressure velds are 100% dye enecked sind 10% are radiographed in addition. All loop rafety valves vi]l be tested. Pre-installation tests of the loop pumps and letdown valve vill be conducted on test cts.nds. The instrument air supply system vill be pneumatica11y tested.
o)
The pressure tube vill receive external and internal l
hydro tests o" a test stand prior to openttion in the loop. The pressure tube connection to the reactor vessel vill be hydro tested and the connections to the Saxton Auxiliary S stems vill be hytiro tested and all connection y
welds dye checked.
c)
All aev vapor container penetrations and all penetrations that have been repaired or modified will be leek tested in accordance with Section B-4 of the Technical Specificatione d)
All $nstruments and controls and circuits vill be given continuity and functional checks in addition to calibra-f tion checks and response tire measurements. All electrical systems vi n e tL'n fm proper grounding, All alarms s
vill be chec).d for proper functioning and calibration.
All record < s vill be checked for proper responses to signals h-1
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from inctrumeidation. All valver, matorc and purpc will be elecked ta.d tected for proper functioning and responce to operation producing cignalc.
e)
Continuity and responce timeo of all loop reactor ceram cignals vill be checked. Seram signal initiation vill be checked against crecified set pointo.
fi Loop contrrl cystems vill be given thorough checho to accure that ttrol set pointo correspond to calibrated and meucured parameterc.
g)
Following ?r concurrent with the above outlined tecto, the complete loop vill be subjected to hot functional tects. The first tect vill be made with a jumper pipe inrtalled in place of a preccura tube (co that there ic no connection between the loop and the reactor priusry cycter). The cecond tect vill be nede with an unfueled precsure tube inctalled in the reactor. The loop cam,oling cyctem vill be functionally t ented. Simulated losc of coolant and loss of flov tectr will be conducted to check the integrated functioning of the safety injection cyctem.
The emergency condenser performance vill alco be checked d *1ng this period.
h)
' ne operations deceribed above vill be used to check out at' modify, if necessary, the nomal operating and emei ency procedures which are developed for the loop.
4-2
Q,uestion $1 Describa checks and measurements that vill be mde 6 urin 6 the initial operation of the loop with nuclear pover.
Answer:
The checks and measurements that vill be made during initial operation of the loop with nuclear porer fall into two categoriest nuclear and thermal arid hydraulic.
a)
The nudear tests and measurements to be mad? for the initial operation of the loop with nuclear power vill be limited to those tests and measurements necessary to verify the predicted reactivity coefficie-ta for the pressure tube as quoted in the Supercritical Loop Safe.
guards Report (flux vire meacurements and reactivity follov). Because of the small difference in the fuel loading between the er ren supercritical fuel rodo and the 21 Saxton fuel roan they reple.cc and the peripheral lochtion of the presrure tube, it is expected that there vill be no significant variations in the reactor's nuclear parameters.
b)
The initial approach to full power'in the reactor aid the loop vill be accomplished in a stepwise fashion in order to monito* the loop parameters and operation at lov power levels. Loop operating parameters vill be set for each step and then the loop pressure, temperature and flow parameters-vill be monitored to check for.any unexpected conditions. A heat balance computation vill be performed at each new power level step to determine-the_supercritical fuel ansembly power level.
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l Question 61 In the event of a supercritical loop loss of coolant, state the amount of fission products that tright be released to the containment vessel assuming (1) normal operation of the loop safety injection system, and (2) malfunction of the loop safety injection system.
Ancver:
(1)
Based on the analysis presented in the answer to Question 2, thc only activity which would be released to the containment following a loss of coolant accident with proper opemtion of the safety injection system would be the activity contained in the loop coolant.
With a maximum operating level in the coolant of 20 pe/cc (15 mitute degassed activity), the total release vould 6
be approximately 6 x 105 pc of gases and 3 7 x 10 pc of solids and halogens.
(2)
Aesuming tatal malfunction of the safety injection system and complete loop fuel assembly meltdown, the maximum
- t. mount of fission products that might be released to the containment atmosphere are listed below. The numbers are-based on the relet.aes quoted for the maximum hypothetical accident for the Saxton reactor with a correction factor applied for the number of fuel rods involved and the average power level in the rodo. (The value of the corree-tion factor is approxicately 350.)
Fission Product Releas,e, e
Activity Released to the Gat:::n EnerEy Grouy Contaitment - curien O.8 MeV 5.83 x lONb 1.6 Mev 1.11 x lo P. 5 Mev-0.43 x lok Iodine Isotopes 3
131 1.12 x 10.
132 1.65 x 103 133 2 54 x 103 134 2 92 x 103 135 2 30 x 103 6-1
Question #7 Provide the basis for the statement on page IV-3: "In the event of such an occurrence, the Caxton reactor would be able to with-stand such a failure and still be able to shut down and prevent extensive core damage".
Answer:
Although the conservative design basis of the pressure tube essentially precludes any pressure tube failure, two cases of gross pressure tube rupture were investigated. The first case assumed an instantaneout tupture of the side of the pressure tube and the formation of a two inch diameter hole. The supercritical loop coolant then leaves this hole with a density equal to operating conditions and with a velocity equal to the speed of sound in the loop coolant. The resultant reaction forces on the adjacent Saxton fuel rods were then calculated to determine if the rods vould fail.
The results of the analysis show that the adjacent fuel rode vill not shear off but vill deform and bend. There are no control rods adjacent to the special fuel assembly that contains the supercritical pressure tube ao that deformation of a few fuel rods and possibly even the fuel assembly can vould not cause enough movement to cause the closest control rod to deform and stick.
The reaction force from the coolant jet leaving the pressure tube break decreases very rapidly with distance from the tube as a result of dispersion and quenching of the jet in the reactor coolant so that deformation of more than a few fuel rods is not possible.
Therefore, the shutdown capability of the reactor would not be affected by thi'J accident.
The second case investigated was the complete and instantaneous severance of the bottom end of the pressure tube.
In the event of such a failure, the severed piece of the pressure tube would be accelerated toward the bottom of the fuel assembly, Because of the small clearances available and the deformation and rapid 7-1
deceleration of the severed piece in the reactor coolant, it is highly improbable that the end of the pressure tube vould be forced through the bottom nozzle of the fu.al assembly but more likely that the piece vould be jammed into the nozzle. Follow-ing the severance of the end piece, the. resultant large pressure drop across the fuel in the pressure tube during blowdown of the loop coolant vill exert substantial forces on the fuel and the internals and vill possibly cause the fuel and bal.'les to be torn loose from this support structure and jammed into the bottom of the fuel assembly.
This damage to the pressure tube and its fuel and internals could possibly cause some deformation of the fuel in the Saxton fuel assembly but as in the first case, the damage vould be limited to the fuel in the immediate vicinity of the pressure tube and would not affect any control rods or shutdown capability.
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Question 8:
Provide calculatious of the amount of liquid and gaseouc vactes which raight te released from the S!ZC fa lity as a consequence of nomal supercritical loop operation.
Ansver:
Norml operstion of the supercritical loop vithout any clad-ding defects vill produce on'y snnll amounts cf activity in the coolant as a result of activation of loop corrosion products and fission products f rom tramp uranium, and therefore, no measurable additional amounts of radioactive liquid or gaceo'.s vastes vill be released from the SNEC facility.
Operation of the supercritical loop vith defective cladding would result in an increase in the coolant activity level and the release of measurable amounts of radioactivity above the
'd from the S!EC facility. The follow-amounts normally r-w ing list gives the amount of radioactive gases released to the S!EC vaste disposal system during six veeks of operation with four of the seven loop ft,el rods dcrective.
Activity Releaned to the Isotope Wacte risposal System 3
Ar-41 1 50 x 10 pc Kr-85 8.7 x 100 pc YJ-85 (m) 17 x 107 y Kr-87 33 x 10' pc i
0 Kr-BS
~ 47 r.10 pc 0
Xe-133 8.7 x 108,ac Xe-135 1.23 x 10 g, Waste gases from the supercritical loop can be stored in the gas decay tanks to allow decay prior to release from the SIEC facility to the environment via the plant stack. The follow-ing list gives the amounts of those gases that would be avail-able for re] ease to the atmosphere following a two-veek decay period.
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Ar-41 0
K1-85 8.7 x 10 pc re.85(m) 0 Kr-37 0
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0 Xe-133 1.0 x 10 pe Xe-135 0
Additional calculations were performed to detemine the maximum amount of iodine that could be relected from the SNEC facility due to leakage of supercritical loop coolant into the containment atmosphere. The a sumptions used in this calculation vere:
1)
A loop coolant leskage rate of 100 ce per hour with the iodine activity at equilibrim and the leak at a point upstream of the loop deadneraliser.
e.
2)
The loop coolant activity st, ns maximum permissible i
level of 20 pc/cc of 19 minute degassed activity.
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Fifty per cent of the iodine released from the loop coolant is assumed to deposit on the internal cu. faces of the containment.
Natural decay is assumed during the six week operation prior to purge of the containment.
4)
None of the undeposited iodine is removed by the contain-ment air recirculation filters or by the exhaust duct filters.
t Based on these assumptions, the total amount of radioiodine isotopes that could be released to the etvironment is 2.4 x 10-3 f
curies. Iodine-131 represents approximately 73% of this tote.l.-.
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