ML19329E029
| ML19329E029 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/31/1967 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| References | |
| NUDOCS 8004090657 | |
| Download: ML19329E029 (21) | |
Text
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i CONTENTS APPENDIX 6 6A ANSWERS TO QUESTIONS i
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3 0043
'Amendraent 3 A6-1
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O QUESTION Discuss the ability to flood the primary cavity including (1) 6A.3 the level and size of the overflow drains and (2) the cavity volume.
ANSWER The reactor building sprays located in the reactor building Refer to dome discharge uniformly over the operating floor.
The dis-6.2.3 charge falls into the refueling cavity on the opcrating floor and in the steam generator enclosures.
The water falling on the operating floor drains to the refueling cavity and to the stair well.
All water collecting in the refueling cavity drains through two 10-inch drain lines to the reactor vessel cavity.
During normal plant operation these lines are open and protected by screens against debris.
During refueling operations these lines will be blanked closed with a blind flange.
Water collected in the reactor vessel cavity will overflow to the steam ger.crator enclosure through the annular space around n
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the primary c oolant pipe penetrating the primary shield.
W. iter collected in the cavity will achieve a level approximately 3 ft above the top of the core before it overflows to the steam generator enclosures. Two 1 in. drain lines in the reactor vessel cavity, one at the reactor vessel support and the other in the reactor in-core instrumentation trench, drain to the reactor building sump.
Under no conditions is there sufficient head on these drain lines to impair the ability to flood the reactor vessel cavity. The rate of water supply to flood the reactor cavity far exceeds the capacity of the two 1 in, drain lines. The reactor vessel cavity has a volume of approximately 6,150 ft3 to the bottom of the primary pipe.
This volume is approximately 50 percent of the primary coolant inventory and less than 15 percent of the borated water storage capacity.
0043 nn
_N U L/
O Amendment 2 6A-2
l Docket 50-312 Amendment No. I February 2, 1468 QUESTION What is your criterion with respect to the types and timing of 6A.1 operator action to be relied on af ter a design basis accident?
Consideration should be given to such things as switching to the recirculation mode and detecting and isolating a broken injection line. Where the action is vital to accident recovery and is required within a short time af ter the accident (~1 hour),
we believe that automatic action should be provided.
ANSWER A general criterion is to rely upon automatic devices to accom-Refer to plish all actions required immediately following an accident to Section provide protection of the reactor core and reactor building 6.0 integrity. Operator action is considered quite acceptable 15-20 min following an accident.
In fulfilling this criterion, activation of all engineered safeguards systems and isolation valves is accomplished automatically. Switching to the rec i r-culation mode is accomplished manually from the control room; and, depending on the number of pumps ir operation followinn an accident, recirculation is not necessary until at least 20 uin af ter the accident has occurred.
The above calcula tions incorporate the ef fect s of a broken lou pressure injection line for the amount of water that would he j
lost directly to the reactor building through a broken line.
Hence, detection of a broken injection line is not a critical item, and does not require automatic actuation for line isolation.
With regard ro the failure of a high pressure injection line--
when the final analysis is completed, it may be determined that it would be advantageous to add the high pressure injection water through smaller lines in order to assare that there is adequate time available for operator action. A decision en this matter will be made a f ter final calculatiors are performed which reflect the piping layout inside the reac tor building.
In arranging the systems according to the above criterion, it is believed that the engineered safeguards equipment and associated actuating circuitry is less complicated and hence more reliable.
QUESTION State your policy on continued operation of the reactor if it 6A.2 becomes necessary to valve of f an accumulator tank due to excess leakage or othar malfunction.
ANSWER If it becomes necessary to valve of f one of two core floodine Refer to tanks due to excessive leakage or other malfunction, the reactor Section plant will be shut doun.
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The special drain lines between the deep end of the refueling cavity and the reactor cavity will be protected from plugging by a protruding screened box sirailar to, but much smaller than, j
that in the reactor emergency building suep.
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i 0051 O
6 A-4 Amendment 1
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1 QUESTION Discuss the route which containment spray and injection water 6A.4 must take to reach the sump.
Indicate the size and location of drain lines to the sump, the criterion for sizing the drains and the method to be used to prevent plugging of the drains and sump.
ANSWER Except for those described in the answer to question 6A.3, there Refer to are no special drain lines provided to handle the flood of spray 6A.3 and injection water flowing to the reactor building sump follow-ing a loss-of-coolant accident.
Extra large openings, provided for conventional functions are relied on to carry this water to the sump with a mininum of holdup at any location within the building.
The discharge from the reactor building sprays is uniformly distributed over the refueling floor and fuel transfer pool, and the vented tops of the steam generator enclosures.
Water drains into the refueling cavity, from there via the special drain lines to the cavity, thun to the steam generator enclosure and finally, from the steam generator enclosure to the basement
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floor level and the reactor building emergency sump via the
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access opening. The building spray water which falls within
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the steam generator enclosure vent openings falls to the floor level of the steam generator enclosure, where it then flows to the reactor building emergency sump via the access opening between the steam generator enclosure and the basement area in the reactor building. The remainder of the water from the reac-tor building sprays flows to the sump via open grating and the stairwell openings on the north and south sides of the reactor building.
The injection water enters the vessel above the core and will flow out of the ruptured pipe to the reactor or steam generator enclosure, depending on where the pipe rupture occurred.
If the rupture occurred in the reactor cavity, the water will flow out of the annular openings to the steam generator enclosure and frcm there to the sump via the access opening.
If the rupture occurred in the steam generator enclosure, the water will only have to flow through the latter opening to reach the sump.
The reactor building emergency sump is provided with a rectan-gular mesh screen. This screen will be so sized to provide several times the required flow area to the suction of the decay heat removal and reactor building spray pumps. The mesh screen will ensure the continued operability of these pumps by preventing debris from plugging or entering the pump suction.
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0052 Amendment 1 6A-3
O In addition tc, the above, we will monitor unit vibration and motor temperatures to signal unusual conditions requiring other than normal attention.
QUESTION What criterion is proposed with respect to removal of engineered 6A.6 safeguards components for maintenance? For example, would the plant be shut down if one high pressure pump were unavailable for use since a single failure criterion could not be met in the event of an accident?
ANSWI.R All engineered safeguard systems contain sufficient redundancy Refer to so that any active rotating component may be removed for main-6.0 tenance. However, it may be necessary to take additional action to assure the availability of sufficient capacity to handle the design basis condition if a unit is taken out of service for maintenance.
The additional action might include an increase in the test frequency of the remaining equipment or or' ration in a standby mode.
The specific action to be taken in each case will be determined during the detailed design and specified in the technical specifications.
The station will not be shut down in the event of maintenance on any one safeguards pump unless that pump must be out of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. There is no such restriction
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on the high pressure injection pump since the make makeup pump is available for this service.
In the case of reactor building cooling safeguards, two full-capacity independent systems are available for maintaining the reactor building below design pressure.
In this case, a combin-ation of components from the two systems can adequately provide the required cooling capacity, and permit removal of a component from either system for maintenance.
CC2I' 0053 O1 6A-6 Amendment 2
QUESTION Provide the preliminary design for the fan coolers.
In 6A.5 particular, indicate the geometry and heat transfer coefficients that will be utilized to ensure that the units are conserva-tively designed to remove heat in the accident environment.
We require assurance that the design of the heat exchangers for the fan coolers will be fully capable of removing the specified heat load under the steam-air accident conditions. We under-stand that rather than submit a preliminary design, you prefer to specify that proof tests by vendors will be run under simu-lated accident conditions.
Specify the type of tests to be re-quired and indicate how degradation of performance over the lifetime of the plant will be taken into account.
ANSWER The design details for the fan coolers will be developed by the Refer to manufacturers of this equipment.
In normal practice, design 6.2.2 details become available only after an order for equipment is placed.
The purchase specification prepared for obtaining proposals
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from manufacturers will include as a requirement that the
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ability of the proposed equipment to perform be shown by theo-retical analysis and substantiated by test data.
When this information becomes available, we will be able to indicate the geometry and heat transfer coefficients that will be utilized to ensure that the units are conservatively designed to remove heat in the accident environment.
Test to be recuired: We will require manufacturers of the reactor building fan assemblies (cooling units) to test a rep-resentative sample of the cooling surface proposed to substan-tiate performance required under accident conditions. Test conditions will simulate the accident loading given in Section 6.2.
Measurements of air velocity and entering and leaving conditions of the air-steam mixture will be used to calculate the cooling performance per square foot of coil fan area.
Coils used in the fan assemblies will be sized from theoretical cal-culations and correlating test data.
When the information becomes available, it will be submitted to DRL.
Accountine for degradation: We will specify cooling coils with external and internal fouling factors of.001 to allow for degradation of heat transfer. Normal inspection and mainten-ance will correct degradation of such elements as filters and bearings.
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Amendment 1 6A-5
QUESTION Describe the test programs that will assure adequate performance 6A.9 of the engineered safety features in the post accident environ-(DRL 6.3) ment.
ANSWER Test to be required: SMUD will require manufacturers of the four Reactor Building emergency air coolers to provide verification by test of a representative sample of the cooling surface proposed to substantiate performs'ce required under accident conditions. Test conditions will simulate the accident loading given in Section 6.2 of the PSAR. The test air-stream mixture will enter the coils at 286 F and 74 psia (steam partial pressure at 53 psia, air partial pressure at 21 psia).
The test cooling water will enter at 95 F and be adjusted for a rise of 9C F (185 F leaving). Measurements of air velocity and entering and leaving conditions of the air-steam mixture will be used to calculate the cooling performance per square foot of coil face area.
Coils used in the fan assemblies will be sized from theoretical calculations and correlating test data.
The total capacity of each of the emergency air coolers under accident conditions will be at least 60 x 106 Btu /h, or 240 x 106 Btu /h for the four combined.
The coil test will be conducted with a fan of the same type that will be employed in the full-scale assemblies. The fan will handle the air-vapor mixture af ter it has passed through the cooling coils, making the fan-coil relationship the same as it will be in the full-scale assemblies.
All pertinent fan data will be recorded in order to evaluate fan performance under the test conditions of temperat' ire, pressure and humidity as described above.
It is anticipated that this test will be performed at an off-site facility and will not be a test of installed equipment.
Accountine for degradation:
SMUD will specify cooling coils with external and internal fouling factors of.0005 and.001 respectively to allow for degradation of heat transfer.
Normal inspection and maintenance will correct degradation of such elements as filters and bearings.
Totally-enclosed, water-cooled fan motors will be specified for the duty required.
The performance of the other safeguards will be assured by the tests outlined in the response to question 13A.3.
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6A-8 Amendment 2
QUESTION Update your PSAR with those design revisions described at our 6A.7 March 1, 1968 meeting.
(DRL 6.1)
ANSWER The PSAR has been updated to reflect those design revisions described at the March 1, 1968 meeting.
QUESTION Provide the anticipated post-accident radiation dose levels in 6A.8 the containment.
Compare the anticipated gamma exposures with (DRL 6.2) damage thresholds for the engineered safety features.
ANSWER Gamma radiation levels in the containment building were calcu-lated as a function of time after the Design Basis Accident.
The cumulative doses as a function of time are shown in Figure 6 A. S - 1.
The dose is essentially due to those fission products which remain airborne in the containment atmosphere after sprays.
It was conservatively assumed in the dose calculations that sprays remove only 507. of the airborne iodine.
It may be observed that the doses level off after approximately 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />
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to about 107 rads.
Consequently, the value of 107 rads con-
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stitutes the maximum expected exposure that the engineered
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safety features may experience in an accident.
Considering threshold damage from gamma rays to engineered safety features, the materials which are relatively sensitive to photon exposure are those of organic nature:
organic insulating mate-rials, lubricants and non-metallic motor seals.
In regard to organic insula ting materials. class H motor insulation has suf-ficiently high threshold daaage to insure against failure from radiation exposure. A preferred material for electrical cable insulation is cross-linked polyethylene which has a radiation 7 rads. For emergency fan motors, damage threshold of 2.5 x 10 silicons with a threshold damage of 3 x 109 rads will be used both for insulation and bearing grease.
In respect to motor seals, only metallic seals will be used to avoid radiation deterioration which may occur with organic seals.
In conclusion, with judiciously selected materials, an integrated 7
post-accident dose of 10 rads is not expected to detrimentally affect the engineered safety features.
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'O V J J J' SACRAMENTO MUNICIPAL UTillTY DISTRICT Amendment 2
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Tests recently reported from ORNL also show very rapid removal of elemental iodine.3 Spray solutions consisting of 1 w/o Na2 2 3, 3,000 ppm B, 0.153 M NaOH showed a removal half-time of S0 less than one minute for elemental iodine. Removal half-time for methyl iodide was 101, 94, and 101 for three runs with the same solution.
The spray system tests support the theoretically derived conclu-sions that elemental iodine can be removed very rapidly but that methyl iodide is less susceptible to current spray solutions.
Other work has suggested that iodine plating-out on surfaces within the containment prior to reaction with the spray chemi-cals may subsequently be desorbed in the methyl iodide form.
The ultimate decontamination effectiveness of sprays is apparently tied to methyl iodide. However, if the initial release contains little CH 1, the initial decontamination of an effective spray 3
must proceed very rapidly, with limited opportunity for surface absorption and desorption.
If the iodine reacts with the spray liquid it will stay in solution and a high decontamination factor will result.
The experimental ef forts referred to above have not yet established a firm decontamination factor, but the current results indicate that a value of 80 to 100 or more may be achieved. At the present, however, an overall decontamination factor of at least 5 is sub-stantiated and this value is generally accepted as a conservative figure.
Section 14.3 of the PSAR presents a discussion of the use of the reactor building spray water, augmented with sodium thio-sulfate, for removal of iodine contamination from the building atmosphere. Based on theory and the early research into removal of iodine by reactive solutions, a conservative calculation was made which resulted in a removal constant of 25 hr-1 (half-time =
1.7 minutes) for the exponential decay of an initial iodine release.
Iodine removal rates observed in the model containment tests referred to above are generally censistent with the removal rate calculated.
Research into spray system effectiveness has suggested alternate chemical additions to the borated spray water.
It is indicated that an increase in pH to 7 or higher by the addition of NaOH leads to an improvement in the methyl iodide distribution coef-ficient:
kd = conc. CH I in liquid / conc. CH 1 in atmosphere.
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Also, the solutions containing NaOH have improved stability and greater materials compatibility. Considering the recent 0058 FR g
Amendment 2 6A-10
QUESTION Provide an evaluation of the ultimate iodine removal capability 6A.10
.for the proposed spray systems that can be rigorously supported (DRL 6.4) by presently available. experimental evidence.
Include a dis-cussion of spray effectiveness in removing aerosols.
i ANSWER Radioactive iodine released to the reactor building atmosphere l
following the loss-of-coolant accident is expected to appear principally in the form of elemental iodine, 1 ; hydrogen 2
iodide, HI; Methyl iodide, CH 1; and perhaps small amounts of 3
higher alkyl iodides. These. contaminants may appear as vapors or maybe attached to particulates. The disposition of these contaminates depends upon subsequent processes taking place within the containment atmosphere.
e Iodine release experiments were carried out in the containment shell of. DIDO reactor at Harwell to test the efficiency of steam condensation in removing iodine from the atmosphere.t In the experiment, iodine laden air was mixed with steam and passed through a water-jacketed condenser.
Iodine constituents were determined in samples taken before and after the condenser.
It was concluded that the clean-up of iodine aerosols by condens-ing steam can effect the removal of about 977. of elemental
.ic/ine, about 507. or less of gaseous iodine compounds, and about 95 to 99% of iodine attached to particulate material.
i Less than 20% of the methyl iodide form was removed.
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This experiment shows that the post accident process of steam condensation within-the containment building would provide sub-stantial decontamination, even-without active sprays. Methyl i
iodide,.which is affected the least by sprays or by condensation, constituted only a small part of the iodine present in the con-i tainment atmosphere:
~5% at 30 minutes af ter iodine release
'and 1.57. after 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
It can be inferred from these experi-mental results that where initial methyl iodide constitutes no more than 57. of the iodine contamination and particulates are neglected, steam condensation alone can provide a decontamina-l tion factor of 6 to 10.
l Two large scale spray runs have been made recently in the Con-tainment Systems Experiment vessel at Pacific Northwest Labora-j tories.- The spray solution used was 3,000 ppm H B03 plus NaOH 3
to give'a pH of 9.5.
These runs, which are preliminary to an extensive test program, indicate spray removal half-times for
. elemental iodine vapor on the order of several minutes. Half-times for particulate iodine were approximately a factor of two higher.
4 s
me 3g 0959 Amendment 2' 6A-9 m
QUESTION Provide an analysis of the physical aspects of the proposed 6A.ll spray systems, including the fraction of the entire containment (DRL 6.5) volume directly covered by the sprays, the convection circulation into the spray pattern, the range of drop sizes, and the relative temperatures of spray and containment ait with their effect on iodine removal rate and efficiency.
ANSWER The spray system proposed for Rancho Seco Unit 1 consists of two independent trains, each rated at 1007. of design capability.
The two independent systems have overlapping coverage so that operation of either train provides nearly complete coverage of the containment horizontal cross-section. By omission of sprays over approximately 157. of the central cross sectional area, the sprays will induce air circulation within the build-ing assisting both the natural convection from the reactor and the emergency coolers. The flow of 1500 gpm from one train pro-vides a spray flux of 0.15 gpm/ft2 Spray nozzles are located in the dome region above the polar crane and all are directed downward. The volume directly covered by the spray and above the 45 foot operating level is 557. of the total free volume.
The induced air flow will draw the balance of the containment air within the spray pattern. Containment building cylindrical walls are washed by the outer ring of sprays.
The emergency air coolers will be placed so as to support a general circulation pattern downward at the cylinder walls and upward along the containment vertical axis.
A minimum of two coolers operating ensures circulation of 80,000 cfm through the sprayed region.
l Analytical model studies were undertaken by parsley to estimate the influence of temperature and drop size on gas-film deposition velocity and removal half-life. Temperatures of 30-1500C and drop sizes of 100-1600p were considered. The results show that the deposition velocities decrease approximately by a factor of 2 when the containment model temperature goes from 30 to 150 C.
However, interacting ef fects reduce the temperature sensitivity of the removal half-life.
Increasing drop size increases the half-life. The range of calculated half-lives were as follows:
100 u 1600 p 30 C 0.13 sec 111 sec 1500C 0.19 sec 113 sec Initial studies on drop size yere based on sprays of uniform-sized drops.
Subsequent work' considered a log-normal distribu-Ub0 6A-12 Amendment 2
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developments and the fact that the research is continuing, selection of the spray additive will not be made at this time.
However, three types of solutions will be regarded as candidate sprays:
Type 1 1 w/o Na3S,03 in borated water Type 2 1w/oNajS{03 in borated water with pH controlled at 9 to 9.5 by NaOH Type 3 borated water with NaOH added to give pH of 9.5 It is er:pected that the best chemical additive from all points of view will be identified by the continuing research programs prior to the completion of Rancho Seco Unit 1.
It is anticipated that the removal rate for molecular iodine at that time will be demonstrated to be at least as good as the value computed now.
For the purpose of computing off-site doses, it was assumed in the PSAR that the atmospheric contamination would remain at 57.
of the initial value (ultimate decontamination factor of 20).
The doses that result from this analysis are shown in Table 14.3-1.
If an overall decontamination factor of 5 is used, then the thyroid doses downwind from the facility will increase approximately by a factor of 4 above those indicated in the PSAR.
Thyroid doses following MHA and DBA and using an iodine decontamination factor of 5 are indicated in the table below:
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\\J THYROID DOSE, REM Site Boundary Low Population Center DBA 3.6 0.11 MHA 59.6 2.2 It may be noted from the table above that even with an iodine decontamination factor of only 5, the thyroid doses fall well below the AEC guideline values.
REFERENCES 1.
Stinchcombe, R.
A.,
and P. Goldsmith, " Removal of Iodine from the Atmosphere by Condensing Steam", Journal of Nuclear Enerev Parts A/B, 1966, Vol. 20, pp. 261 to 275.
2.
McCormack, J. D., "Large Scale S pray Tests in the Contain-ment System Experiment (CSE)", presented at Oak Ridge National Laboratory, Feb. 5, 1968.
(N 3.
Parsly, L. F.,." Safety Pilot Plant Spray Experiments", ORNL
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Nuclear Safety Resea rch and Development, advanced data release, March 1968.
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Amendment 2 6A-11 O O ~' ' ^
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QUESTION Discuss the extent to which reversible, competitive, and slow 6A.12 chemical reactions have been considered in the evaluation of (DRL 6.6) the effectiveness of the spray systems. Consider the contribu-tion of liquid film mass transfer resistance in the calculation of the overall mass transfer coefficient.
ANSWER In the event that Solution Types 1 or 2 (see Question 6A.10) are selected, excess of sodium thiosulphate will be addad to assure chemical reaction with any regenerated iodine. The research work of Patterson and Humphries reported in the ORNL Nuclear Safety Research and Development Bi-monthly Reports will be followed for indication of significant competitive reactions.
The mass transfer of radiciodine from containment air to spray droplets is governed by the well known relationship:
1 1
Cg 4
VG KG CKLL where Vc overall mass transfer coefficient based on
=
concentration Kg mass transfer coefficient in gas phase based on
=
concentration driving force, cm/sec.
KL mass transfer coefficient in liquid phase, em/sec.
=
CG concentration of radiciodine in the gas phase,
=
g-moles /cm3 CL concentration of radioiodine in the liquid phase,
=
g-moles /cm3 According to L. F. Parsly(l), the spray removal of two of the three forms of radiciodine under which it is present in the containment and air, namely 12 and HI, is a gas-film controlled process. Parsly found that the half-life of elemental iodine using a 17. solution of sodium thiosulfate sprays ranges from about 6 to about 181 seconds, depending on the droplet size and temperature of the sprayed liquid. The removal half-life for the acidic HI using an alkaline solution of sodium thio-sulfate should be even shorter. In the case of the third form of radiciodine, namely methyl iodide, the mass transfer from air to droplets appear to be a liquid film controlled process.
Since at this time it appears uncertain what the removal half-life of methyl iodide is in the presence of sodium t' osulfate sprays, analyses have been performed assuming that all methyl iodide remains airborne af ter sprays.
Considering reversible reactions which may take place during radioiodine spray removal, Parsly develops a decontamination factor for the radiciodine. The decontamination factor (DF) is given by:
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- 00b, 6A-14 Amendment 3
tion of drop size.
The results indicated considerable sensitiv-ity to the standard deviation. For a nominal 100p drop sized spray, the remal half-life increases by a factor of over 6 as the geometric standard deviation increases from 1 to 2.
Tentatively, the mean drop diameter requirement has been set at 700-p with spray system operating at conditions of minimum noz-zie pressure differential. A nozzle pressure differential increase, which will occur when the containment pressure decays or uhen recirculation starts, will result in smaller drop sizes.
Further details of the spray nozzle have not yet been defined.
Spray system development work is presently determining charac-teristics of commercial nozzles and is sufficiently advanced that one may expect with confidence that satisfactory nozzle designs will be identified and characterized during the Rancho Seco construction period.
At the beginning of the recirculation phase, the temperature of the sump water being pumped to the spray heads will exceed the temperature of the atmosphere. This temperature relation-ship will exist until equilibrated by the action of spray water heating the atmosphere and decay heat exchanger cooling of the sump water. Nuclear Safety Pilot Plant run 38 was performed to determine if there is an adverse effect when a relatively hot solution is sprayed into a colder atmosphere.3 The condi-
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l tions for the test were 74 C atmosphere temperature and 99 C
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solution temperature. The corresponding Rancho Seco conditions at beginning of recirculation are 90 C and 116 C respectively.
l3 The spray solution was the alkaline-controlled sodium thiosul-fate type. The test run demonstrated that evaporation from drops under the test conditions has no significant adverse effect on iodine absorption by sprays.
REFERENCES 1 Parsly, L.
F., ORNL Nuclear Safety Research and Development Report for May-June 1967, ORNL 1913.
2 Parsly, L. F., J. K. Fronzrel, ORNL Nuclear Safety Research and Development Report for September-October 1967, ORNL-TM-2057.
3 Parsly, L. F., " Safety Pilot Plant Spray Experiments," ORNL Nuclear Safety Research and Development, advanced data release, March 1968.
l
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0063 Amendment 3 6A-13
3l When recirculation starts, (the initiation of Phase 3) the sump will contain approximately 413,000 gallons of water con-sisting of 350,000 gallons borated water and 63,000 gallons of primary coolant.
Again depending upon the number and combi-nation of safeguards which operate, 18% to 30% of this water will have passed through the spray system.
Therefore, the sump water will contain 18% to 30% of the desired concentra-3 tion of spray additive.
Spray additive will continue to be injected in the spray circuit and can also be introduced to the sump water by means of the separate chemical addition and sampling system to bring the inventory of recirculating water to the desired composition for iodine retention.
Thus, the 5
water sprayed into the building atmosphere will at all times have the desired amount of additive and pH.
3l The core cooling solution will be slightly acid during phases 1 and 2 and the pH will be adjusted at the beginning of phase 3 and actively controlled to a value of 9 from that point for-ward.
During phase 3, the cooling water will be in contact with con-crete and therefore it is anticipated that there will be materials introduced into the water by leaching.
Preliminary considerations indicate that the substances involved will have no important effect upon the chemistry of the water.
However, it may be necessary to provide a low flow velocity area to allow particulate matter to settle.
In contrast to the insoluble matter, all fission products soluble in the spray liquid will be continuously recirculated during phase 3.
Eventually an equilibrium between the com-potent in the liquid and air phases will be reached.
QUESTION Provide a discussion of the extent to which exposure to the 6A.14 solution discussed in item 6.7 above will be factored into the (DRL 6.8) procedure for selection of materials for the engineered safety features for the facility. Discuss the systems that will be affected and the nature of the considerations that will be taken into account.
ANSWER The design of each engineered safety feature will take cogni-zance of the composition and pH of the solution recirculated in the post-accident condition. Material selection or pro-tective coatings will be specified so that corrosion products and chemically released hydrogen will be minimal. The design will provide compatibility with the recirculated solution so as to assure no loss of function for a long post-accident period.
J 0064 6A-16 Amendment 3
4
' p) 1 v
QL Co 1 + QG DF
=
CL total volume of gas in system, cm3 ~
where QG'
=
total liquid in system, cm3 QL
=
The decontamination factor is based on the assumption that a continuous recycling of the sprays will establish an equilibrium between the dissolved and the airborne iodine.
REFERENCES 1 Parsly, L. F. Jr., " Gas Absorption Theory Applied to Contain-cent Sprays," ORNL-TM-2002, January 1968.
4 QUESTION Provide an analysis of the composition and pH of the emergency 6A.13 core cooling solution as a function of time following the design (DRL 6.7) basis loss-of-coolant accident.
Consider spray system additives, l
soluble neutron poisons, fission and corrosion products, ele-ments leached from concrete, etc.
ANSWER The span of time following a loss-of-coolant accident can be divided into three significant periods with regard to the chemistry of the water flowing through the reactor core.
1.
The initial blowdown period, lasting less than one minute.
2.
_The period following blowdown in which cooling water is l
drawn from the borated water storage tank. This phase will last until the tank is emptied at from 35 to 70 minutes, l3 depending upon the number of safeguard systems operating.
3.
The remainder of the time in which it is necessary to pro-vide core cooling.
During this period the water is drawn from the reactor building sump _and recirculated.
During the first phase, the core cooling solution will be l3 simply the primary coolant and its chemistry will depend upon the time in core life when the accident occurs. The chemistry of the primary coolant is. described in section 9.2 of the PSAR.
In phase two, the core cooling solution will have the compo-sition of the borated storage tank water containing 13,000 ppm boric acid. From a time 35 seconds after the accident and until the end of phase 2, the building spray water will also 3
be taken from this source but a chemical additive in the spray circuit improves iodine removal by maintaining the pH h#~ j.
of the spray at approximately.9.
j
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Amendment l3 6A-15.
any of the solutions will form decomposition products which h
could interfere with system performance. As discussed in the answer to Question 6A.13.
Particles and other solid contaminants which are large enough to potentially interfere with core cooling should drop out in the low-velocity settling region around the sump.
REFERNCES 1 Zittel, H. E., ORNL Nuclear Safety Research and Development Program Bimonthly Report for July-August, 1967, Wm. B. Cottrell, Editor, ORNL-Ut-1986.
QUESTION Discuss both the time-dependent radiolytic and chemical hydrocen 6A.16 formation under post-accident conditions for the solution given (DRL 6.10) in item 6.7 above. Include an estimate of total a and S activity in both the core and in the liquid, and of the total expected irradiation dose characteristics. Indicate the extent of hydre-gen formation by chemical reaction (corrosion) with exposed reactor materials.
ANSWER The integrated dose to the emergency core cooling solution was calculated for the MIL \\ by assuming that the reactor had been operating at full power at 2568 MW(t) for an average fuel irradia-tion time of 620 days.
It was further postulated that 17. o f the solid fission products, 50% of the halogens, and 100% of the noble gases escaped instantaneously from the core. The gases were assumed to remain in the reactor building atmosphere and the halogens and solids were assumed to be immediately contained in the emergency core cooling water. The curve in Figure 6A.16-1 is the sum of both the beta and gamma doses to the solution from the radioactive materials in the core, reactor building atmosphere and pool of emergency coolant liquid at the bottom of the building.
The radiolytic hydrogen production resulting from a MILA dose to the recirculating coolant has been calculated on the basis of the experimental data reported by Zittel.1 His recent experiments indicate that the presence of dissolved oxygen has a strong effect in reducing the net release of hydrogen by immediately recombining with free hydrogen within the solution. Various spray compositions were given 60Co gamma irradiation while encap-sulated at a gas-to-liquid volume ratio of 25.
Under these con-ditions, each of the three solutions of c'irrent interest for O
ooss 6A-18 Amendment 2
.A
QUESTION Discuss the time, temperature, and radiation dependent stability
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6A.15 of the spray solution under both storage and post-accident recir-(DRL 6.9) culating conditions and indicate the possibility of forming solid decomposition products or precipitates which could poten-tially interfere with system performance.
ANSWER The thermal stability of potential spray solutions can be evaluated from ORNL spray technology work.
The acid sodium thiosulfate solution, Type 1, has shown poor stability to even short-term exposures to temperatures of 130 to 1400C. S ince these temperatures will be encountered in the initial phase of rec ircula tion, this characteristic appears unacceptable 1y poor at this time.
The alkaline controlled Type 2 solution has been tested at 140 C for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> and thus far shows good thermal stability. A loss of 10% of original iodine retention capabil-ity was noted which is small relative to the factor of 10 excess capability provided by the 1 w/o solution. Type 3 solution is thought to be free of thermal ef fects at spray or recirculation temperatures.
A study of the effects of high-energy radiation on various pro-posed spray solutions, including Type 1, 2, and 3 solutions (see Question 6A.10), has been reported by Zittel.1 In these tests, encapsulated solutions samples were exposed to 60Co gamma radiation. The radiolytic effects were followed in terms of 7s
- 1) iodine retention capability, 2) change in pH, 3) evolution of i
radiolytic H, and 4) identification and measurement of radiolytic 2
solids.
With regard to radiolytic effect on iodine retention, it was found that the Type 1 solution lost over 507. of its initial capa-bility, after exposure to 5 x 107 r.
Loss of all excess capacity occurred at 9 x 10' rads which, according to Figure 6A.16-1, would occur af ter approximately 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> of recirculation.
Solution Type 2 shows an initial increase of iodine reaction capacity as a consequence of decomposition into additional reac-tive species but, the long-term effect is a gradual loss of effectiveness. Sixty percent of the initial capability (6 times the required amount) will be available af ter 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> (108 r) and minimum required effectiveness will still be available after 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of recirculation. The Type 3 solution has not s h o wn a significant loss in effectiveness from radiolysis.
Solutions 2 and 3 show a slight but apparently insignificant reduction in pH.
This is not a consideration with the Type 1 solution since its pH is not controlled.
The evolution of radiolytic hydrogen is discussed in the answer to Question 6A.16.
Decomposition products, notable colloidal sulfur, were observed 7'^
following irradiation of Solution 1.
No decomposition products
(
have been noted in Solutions 2 and 3.
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