ML19320C603
| ML19320C603 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/09/1980 |
| From: | Hovey G METROPOLITAN EDISON CO. |
| To: | Jay Collins Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19320C604 | List: |
| References | |
| TLL-283, NUDOCS 8007170442 | |
| Download: ML19320C603 (88) | |
Text
o Metropolitan Edison Company g-g')
Post Office Box 480 Middletown, Pennsylvania 17057 717 9444041 (j
Writer's Direct Dial Number July 9, 1980 TLL 283 TMI Program Office Attn:
Mr. John T. Collins, Deputy Director U. S. Nuclear Regulatory Commission c/o Three Mile Island Nuclear Station Middletown, Pennsylvania 17057
Dear Sir:
Three Mile Island Nuelaar Station, Unit II (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 Submerged Demineralizer System Technical Evaluation Report Your letter NRC/TMI-80-89, dated May 16, 1980, requested additional inforroation to enable your evaluation of the subject report.
Enclosed please find our responses to your comments. We are prepared to discuss these responses with you or your designated represertatives at your convenience.
Sin erely, dm G.'
ovey r'ctor, TMI-II i
GKH:LJL: dad Enclosures cc:
B. J. Snyder i
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l 800 717 0 VVM Metrcpohtan Edrson Company is a Member of the General Pubhc Ubhties System
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GENERAL COMMENT
S.
QUESTION 1 Provide (1) a complete set of piping and instrument drawings (P&ID), (2) general arrangement drawings, (3) zeolite, cation, and filter vessel internal drawings, and -(4) the system description document to the latest revision for SDS.
RESPONSE
Piping and instrument drawings (P&ID) and general arrangement (GA) drawings are enclosed in accordance with the attached listing of SDS drawings. This listing includes the current revision number.
Zeolite vessel internal drawings are enclosed. Both the zeolite and cation vessel internals are of the same design.
The SDS system description document is surrently being internally reviewed.
Upon completion of that review, on or about July 15, 1980, the SDS system
- description will be forwarded.
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i SUBMERCED DEMINERALIZER SYSTEM DRAWINGS 527D-A-5001, Rev. 03 Supply Manifold, Submerged Ion Exchanger P & ID 527D-A-5002, Rev. 02 Contaminated Feedwater System P & ID 527D-A-5003, Rev. 02 Monitoring Tank System P & ID
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527D-A-5004, Rev. 03 Offgas and Liquid Separation System P & ID
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527D-A-5005, Rev. 01 Ion Exchange Flow Sheet 527D-A-5006, Rev. 02 Intermediate Sampling System 527D-A-5007, Rev. 00<-
General Layout Plan - Ion Exchange Equipment 527D-A-5008, Rev. 00-Utility Flow Diagram 527D-A-5009, Rev. 01 Beta Monitoring Manifold P & ID
-527D-A-5010, Rev. 00 Ion Exchanger Supporty Assembly 527D-A-5011, Rev. 01 Dewatering Station P & ID 527D-A-5012, Rev. 00 Pump Removal Instruction for Of fgas Pump Feed System 527D-A-5013 RCS Clean-up Manifo'd P & ID (SEE NOTE 1) 527D-A-5014, Rev. 00 Pool Floor Equipment Base Location Installation Dimensions 527D-D-5002, Rev. 01 Prefilter, 125 Micron 527D-D-5003, Rev. 02 Final Filter, 10 Micron 527D-D-5004, Rev. 00 Prefilter/ Final Filter Details 527D-D-5006, Rev. 00 Ion Exchanger with Agitator (SEE NOTE 2) 9 NOTE 1:
This drawing is currently in development and will be
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furnished when available.
l NOTE 2:
This drawing subject to change pending resolution of solidification requirements.
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GENERAL COMMENT
S OUESTION 2(a)
Provide'.a complete process flow diagram indicating (a) expected radioactivity concentration (pCi/ml) in each process stream including the final processed water for each radionuclide present...
RESPONSE
Oak Ridge National Laboratory has been tasked with the testing of the SDS process flow stream performance characteristics.
These performance characteristics include the determination of process stream node point radionuclide concentrations.
Attached to this response are the results of testing for the determination of radionuclido concentY1tions at process stream node points as submitted to GPU by ORNL on April 3, 1980.
These results characterize the SDS process stream as follows:
F - pre-filter and final filter Z1 - first zeolite bed Z2 - second seolite bed Z3 - third zeolite bed-C1 - cation bed C2, A, MB - polishing bed The model used for analysis of the process flow stream treats the polishing bed as though it is a combination of a separate cation bed, anion bed, and mixed resin bed.
While it is physically possible to provide final polishing of the effluent flow stream in this manner, there are many other methods that can be cmployed to accomplish the objective of the polishing bed. Present plans for the SDS polishing bed specify that the resins will be loaded into one vessel and that they will be " layered" to achieve the final polishing function.
Radionuclide concentrations at the defined node points, as determined during the testing program, are given. These results are based on preliminary testing and the final report from ORNL will be submitted to the NRC in August, 1980, depending on receipt by GPU by July 30, 1980.
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EXPECTED ACTIVITY CONCENTRATIONS IN SDS PROCESS STREAMS AFTER 200 BED VOLUMES THROUGH EACH ZEOLITE BED
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9 X 10CFR20 3H 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 333 60Co c
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5 E-5 5 E-5 0.1 89 b
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2.0E-3 2.2E-4 6.9E-5 5.9E-5 d
1.5E-5 1.1E-5 3.7 90Sr '
2.32 2.26 8.6E-3 9.4E-4 3.0E-4 2.6E-4 d
6.7E-5 4.8E-5 160 '
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1.0E-5 1.0E-5 0.1 103 c'
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106 Ru c
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2.1E-4 1.4E-4 14 i
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1.9E-2 1.9E-2 1.9E-2 1.9E-2 d
1.9E-2 1.9E-2 192 I34 Cs 2.62E+1 2.62E+1 2.9E-3 7.0E-6 6.6E-6 6.6E-6 d
6.6E-6 3.0E-6 0.03 l I37Cs 1.56E+2 1.56E+2, 1.7E-2 4.2E-5 3.9E-5 3.9E-5 d
3.9E-5 1.8E-5 0.9 144Ce c
c 5.1E-4 3.9E-4 2.6E-4 1.9E-4 d
1.3E-4 6.4E-5 6.4 a In pCi/ml as of July 1,1980.
bDifference 'in strontium concentration based on estimate of'1000 gal of solids in 650,000 gal of water.
cNot detected.
dNot yet determined.
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GENERAL COMMENT
S QUESTION 2(b)
Provide a complete process flow diagram indicating (b) expected concentrations (ppm or ppb) of nonradioactive chemicals in each process stream.
Include test results, references, and any supporting documents to support and substantiate the decontamination factors you assumed.
RESPONSE
Based on ORNL spark source mass analysis, the major nonradioactive chemicals in the SDS feed are estimated to be:
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- 2200 ppm C1 -
15 ppm Ca -
10 ppm Na -
1200 ppm These concentrations will be present in the process stream at the influent to the cation bed.
Some Ca and Na will be removed by the cation mixed bed.
Concentrations of these nonradioactive chemicals in the cation effluent / polisher influent will be:
B
- 2200 ppm C1 -
15 ppm 0-10 ppm Ca Na 1200 ppm The polishing unit will consist of anion, cation, and mixed resin.
Bed loadings, yet to be determined, will not be designed to deionize the influent water. It should be noted that the SDS design has been performed with the primary objective of removal of radioactive contaminants from the proceis flow stream. Any non-radioactive element removal from the flow stream is incidental to the process.
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GENERAL COMMENT
S
. QUESTION 3 Provide expected radioactivity loading (Ci) and throughput (gallons) in each pre-filter, final filter, zeolite bed, cation bed, and polishing demineralizer.
RESPONSE
To bound the filter loading estimate, two approaches were considered. The first estimate (worst case) is based on ORNL report dated September 12, 1979. Thise case assumes that:- (1) filter changeout occurs af ter 50,000 gallons throughput, i
.(2) all solids are loaded on one vessel, (3) no decay of Sr 89 since available data does not distinguish between Sr 89 and Sr 90 for solids. The results of this
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approach indicates a worst' case radioactivity loading of 554 curies on a vessel, f
of which 525 curies is Sr 89/90. -The second estimate (realistic case) assumes that:
(1) Strontium ratios in the solids are the same as in solution, (2) the 2 vCi/ml Sr 89 and.18 pC1/ml feed (solids) would contain approximately 1.3 X 10 Sr 90 as of October.1, 1980. These assumptions yield filter loadings of 2.5 curies Sr 89, 34. curies Sr 90,1.4 curies Cs 134,14 curies Cs 137, and 4.7 curies Sb 125.
The zeolite beds are planned to be operated for approximately 200 bed volumes prior to changeout of the first bed and movement of beds 2 and 3 to positions I and 2, respectively. This intended plan of operation yields a bed throughput of approxtmately,12,000 gallons, based on a bed volume of 8 ft3. The Technical 3
Evaluation Report identifies that the zeolite bed volume is 7 ft. Recent testing has indicated that optimum zeolite bed loading is achieved when the bed is filled with 8 ft3 of zeolite resins.
j Studies at ORNL have indicated essentially all the cesium will be deposited on the first and second. zeolite beds. Utlizing concentrations of 28 and 160 pCi/ml for Cs 134 and Cs 137 respectively, the first bed would contain an estimated 1270 curies of.Cs 134 and'7270 curies of Cs 137. Beds two and three would contain a slightly higher level of activity due to being in service for 400 and 600 bed volumes respectively. ORNL' column tests indicated that if feed solution contains 5.2 X 10-10_pCi/ml and 2.26 vCi/ml of Sr 89 and Sr 90, respectively, the first bed will i
contain approximately 23 curies of Sr 89 and 102 curies of Sr 90.
Smaller amounts
-of other isotopes such as Nb 95. Ru 106, and Ce 144, will also be present.
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RESPONSE TO OUESTION 3 (cont'd.)
i Based on ORNL column testing, the cation bed in service will contain approximately 3 mci of Sr 89/90 per 15,000 gallons of processed water. The polishing unit is estimated to contain the following per 15,000 gallons of process water:
ISOTOPE mci Sr 89 2.7 Sr 90 12.0 Ru 103
.48 Ru 106 44.3 Cs 134
.2 Cs 137 1.2 3 of zeolite It should be noted that zeolite column loading is based on 8 f t resins in the column.
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GENERAL COMMENT
S
' QUESTION 4 Provide expected solid loading capacity for pre and final filters and estimate number of spent filter vessels to be generated as a,recult of SDS operation.
RESPONSE
The design of the pre-filter requires the use of a 125 micron filter element with 2
40 ft of filter area. The final filter design requires the use of a 10 micron 2
filter element with 80 ft of filter area.
Load testing to determine SDS filter loading has been performed by AGNS using a synthetic slurry similar to the con-tainment sump water. This testing confirmed that the majority of the solids will be deposited on the final filter with only minor pre-filter loading. Assuming that the final filter loading is 1/2 pound per square foot and a total of 1000*
pounds of filterable solids pass through these filters, it is expected that approxi-mately twenty-five final filters will be expended.
I,t is also probable that one pre-filter will be replaced after the initial startup of the SDS in order to main-tain a lower DP across the two filters.
- NOTE: The 1000 pounds of filterable solids is based on 10 percent of the estimated solids on the reactor building containment floor being processed by the SDS.
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GENERAL COMMENT
S
- QUESTION 5
-Provide expected number of spent resin vessels (zeolite, cation, and pool purification mixed-bed) to be generated during operation of the SDS.
RESPONSE
Assuming thu: 200 bed volumes, approximately 15,000 gallons will be processed prior to zeolite bed removal and rotation,-67 zeolite beds will be expended to process 1,000,000 gallons of containment sump water.
Based on column
- testing, ORNL has estimated that SDS operation would generate approximately 1,000 gallons of cation resin.- This would expend approximately 17 cation beds. Assuming normal leakage, 2 to 4 pool purification vessels are anticipated to be utilized.
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GEN 5RAL COMMENTS OUESTION 6 Describe process flow and method for processing the reactor coolant through the SDS and/or EPICOR-II.
RESPONSE
The SDS is currently designed to interface with the Reactor Coolant System through the NDHR system. The processing method to be used for the decontamination of RCS water is similar to that method planned for the containment sump water.
The processing of reactor coolant system water by EPICOR-II is specifically pro-hibited by the Memorandum and Order issued by the Nuclear Regulatory Commission on October 16, 1979. However, NUREG 0591, Environmental Assessment on the Use of EPICOR-II at TMI-II, states, " Waste water that is acceptable for processing in the EPICOR-II system is that which has Iodine-131 and Cesium-137 concentrations of less than 100 pCiimi (inte-mediate level waste)".
Because the RCS water has Cesium-137 concentrations that are below the maximum permissible EPICOR-II influent concentrations, we are evaluating the use of the EPICOR-II system for processing RCS waste water.
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GENERAL COMMENT
S _
QUESTION 7 Describe the design and operational features of SDS polishing demineralizer.
,kESPONSE The design and operational features of the SDS polishing demineralizer has been identified as proprietary information under the provisions of 10 CFR 2.790.
Chem-Nuclear Systems, Inc. will forward design details under separate cover.
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GENERAL COM!!ENTS QUESTION 8 Provide the maximum expected tritium concentration in the TMI-II fuel pool water with anticipated routine leakage (coupling and decoupling operations of filter and resin vessels) as a result of SDS operation.
RESPONSE
Recent testing on the 1 " "Hanson" connectors had indicated that a maximum of 100 ml of liquid will escape during the uncoupling operation. The estimated contribution of tritium in the pool is calculated to be 1 x 10 " pCi/ml utilizing the following assumptions:
1.
Pool volume - 233,450 gallons 2.
Leakage tritium concentration - 1 pCi/ml 3.
Three zeolite beds and one cation are uncoupled each 15,000 gallons 4.
Two filters are uncoupled each 50,000 gallons 5.
Three disconnects per vessel are required 6.
Containment waste water is 1,000,000 gallons The above calculations assume that processed water from the monitoring tanks is utilized for the flush sequence prior to disconnect.
Furthermore, it is enticipated that the spent fuel pool will be initially filled with processed water that contains a tritium concentration of approximately 1 vCi/ml.
The contribution to existing pool tritium cencentrations from uncoupling operations is insignificant.
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s cENhRAL COMMENTS OUESTION 9 Provide expected fuel pool water temperature, water volume, evaporation rate.
and area ventilation flow rate, and resulting airborne tritium concentration in the TMI-I and TMI-II Fuel Handling Areas during operation of the SDS.
RESPONSE
If all the Cs 134 and Cs 137 contained in 1,000,000 gallons of containment water were deposited in vessels and stored in the "B" pool, and all decay energy is released to the pool, the rate of energy release is approximately 4000 watts.
Therefore, the pool temperature is estimated to be approximately ambient building air temperature, 80 F.
The pool water volume is approximately 233,000 gallons and will have an estimated evaporation rate between 25 to 50 gallons per day.
The Fuel Handling Building exhaust rate is 36,000 cfm.
16,200 cfm are exhausted from the fuel storage pool area.
Calculations have shown that if processed water containing 1 vCi/ml tritium is used to fill the pool, the resulting airborned tritium concentrations could range between 3 X 10 10 pCi/ml to 6 X 10 10 pCi/ml. As indicated in Ouestion 8, the additional tritium losses to the pool due to uncoupling operations would contribute an insignificant amount to increased spent fuel tritium, and therefore, will not significantly increase airborne tritium concentrations.
It is to be noted that the indicated airborne concentrations will exist only on the TMI-II side of the Fuel Handling Area. The TMI-I and TMI-II Fuel Handling creas will not communicate as a result of modifications being performed to accomplish unit separation prior to restart of TMI-I.
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GENERAL COMMENT
S 9
OUESTION'10-Describe the provisions provided for the removal of spent resin from the SDS polishing unit demineralizer.
RF3PONSE The SDS polishing unit demineralizer is a disposable unit. Our present intention is to dispose of the polishing unit at a shallow-land burial site in accordance with applicable regulations.
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SPECIFIC COMMENTS QUESTION 1 State if specific process features include processing of rinse water through the.SDS.
If so, estimate the volume and radioactivity concentration of rinse water.-
RESPONSE
Flush lines on each of the submerged SDS components are connected to a drain header as indicated on P & ID 527D-A-5001.
This drain line is connected to a shielded mist eliminator tank that drains to the offgas pump standpipe.
An offgas bottoms pump is utilized to transfer liquid back to the SDS feed standpipe for reprocessing.
Concentrations within the mist eliminator tank will vary from a small fraction of, to nearly equivalent to, the SDS feed stream depending on operating conditions.
It is estimated that approximately 200 gallons of flush water per submerged vessel will be utilized during a flush sequence.
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SP$CIFIC COMMENTS OUESTION 2 Based on past experience, project the Reactor Building water level for the next two years and list the critical instruments that may be affected by the water level you projected.
RESPONSE
2.1 The projected water level in the Reactor Building for July 1,1982 is 291.7 feet or a depth of 9.2 feed from the RB floor.
This projection is based on the current water depth of 7.7 feet (as of May 15, 1980) with an inflow leakage of N 0.1 gpm from the Reactor Coolant System (RCS).
This inflow leakage rate has been detected during the period of time that the RCS has been maintained at a pressure of N 100 psi (DVM).
As a bounding (worst case) estimate - a higher leak rate was also studied.
An inflow leakage of 0.2 gpm was indicated from sump level measurements during the time that the RCS was maintained at 4 300 psi.
Thus, an upper elevation of the water depth could be 10.7 feet by July 1982 should the RCS pressure be returned to 300 psi.
The projected depth at the postulated leakage rate of 0.2 gpm would be 293.2 feet.
A plot of the measured Reactor Building water level indicates that a fluctuation can be expected from deactivation of it.e Reactor Building Cooling Water.
This type of action allows the R.B. humidity level to approach 100%.
As condensation occurs on all components that are cooler than air, the sump water is transferred to higher elevation collection areas. When these collection areas become saturated or filled, the accumulation rate in the sump resumes at a rate matching the inflow leakage rate. When the RB :ooling water is activated the collected water evaporates due to the i
lower humidity and is returned to the sump by condensing at the cooling j
coils which are drained into the sump. The type of water accumulation fluctuation shown by this figure is expected to occur seasonally, thus, the projected straight line increase of 0.1 gpm will fluctuate depending on RB Cooling System operation, but the average leakage from the_
RCS will be 0.1 gpm.
2.2 The equipment that is located within the projected water level band which
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is r quircd to cparcts are rzdundant valvss DHV1 or DHV171. These valves, when opened or operated will allow the RCS isolation boundaries to be moved to valve DHV3 which is located outside of the Reactor Building. The
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bottom of the operators for these valves is located at approximately 292.5 feet while the valve bodies are at 291.0 feet.
Containment penetration R-401 is currently being used to measure the RB water level. Water level above the 291.8 feet elevation will allow sump water to fill the penetration sleeve. The measurement system installed through this penetration will still function with proper sealing of the penetrating instrumentation probes.
i A list of all electrical equipment whose operability would be affected by water up to elevation 293.2 is tabulated and is attached.
The listed items are not considered critical for use in plant operation, the maintenance of the safe shutdown of the reactor. However, DH-V-1 and DH-V-171 are critical components necessary for SDS processing of reactor coolant.
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, SPECIFIC COMMENTS QUESTION 3 You state that the radiation dose at the surface of the sump water measures approximately 120 R/hr. State expected radiation dose rate at the same elevation after removing all waste water in the reactor building.
ANSWER The activity present in the solids in the basement of the reactor building will serve as the principle source when the water is removed. The dose rates have been calculated to be between 2 and 10 R/hr. on contact.
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. QUESTION 4 You state that the construction of a vaste storage facility would exceed two years. Describe the waste stc. age facility considered and state estimated radiation doses from that facility.
RESPONSE
For purposec of-this study, the facility evaluated included the capability to store the radioactive liquid water in a long-term storage facility. This long-term storage facility would include not only the water retention capability but that shielding necessary to permit unlimited occupancy at the storage area boundary.
Irrespective of whether the facility is designed as an above-ground facility with adequate building wall thickness for shielding or the facility is an underground facility, it is estimated that design, construction and placing the facility into operation would require at least two years expenditure of effort.
For this facility, a detailed design was not performed.
Shielding requirements were not calculated. Therefore, radiation doses, as a result of the facility, are not available.
Recently, AGNS has performed a detailed analysis of the cost and schedule associated with the design, procurement, and construction of a single intermediate level liquid waste (ILLW) storage tank.
This analysis was performed under contract to DOE; the results are reported in AGNS Report 35900-4.2-26, dated November, 1979, entitled " Engineering Evaluations of Waste Handling and Storage Facilities". This report is based on the ILLW tank design basis of a single 300,000 gallon tank manufactured of stainless steel enclosed in a stainless steel lined 1sncrete vault. Applicable conclusions of the report are:
1.
Engineering, procurement, and construction for a single ILLW tank is projected to require 33 months to complete.
2.
Costs (in 1978 dollars) for a single ILLW tank is estimated to be
$11,000,000.
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SPECIFIC COMMENTS QUESTION 4'(Cont'd.)
3.
A'arage contained activity concentration in the tank is approximately 123 pCi/ml.
Using the results of the AGNS report for direct application to a TMI-II liquid waste storage facility, yields the following conclusions:
1.
Costs (in 1980 dollars) for a single ILLW tank (escalated at 10% per year) would be approximately $13,500,000.
- 2. ~ Assuming the requirement to construct and place into operation three ILLW tanks, proj ected costs would be $40,500,000.
Supporting facilities (such as a transfer system) and tie-ins would impose additional costs, both monetary costs and schedule costs, on the construction of the facility.
As a result of the foregoing discussion, the construction of a long-term storage facility to retain the containment sump water and the reactor coolant is not deemed to be a viable alternative for the following reasons:
f 1.
The schedule for placing the facility into operation is not commensurate with the goal of expeditious decontamination of the THI-II facility.
2.
Implementation of the long-term storage option does not provide for ultimate disposition of the contaminated wastes.
3.
Implementation of this option does not appear to have a distinct cost-benefit.
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' QUESTION 5 Include in your evaluation a solidification system that would utilize a large liner for direct solidification of containment sump waste water.
RESPONSE
Alternative III is devoted to direct solidification and disposal of reactor containment building waters.
A 55-gallon drum contains approximately 7 cubic feet of solidified material.
Should the LL-5100 cask be used (with a special liner), approximately 130 cubic feet of solidified material could be packaged.
The following discussion revises the analysis presented as Alternative III of the SDS Technical Evaluation Report based on use of the LL-5100 cask.
1.
The construction of an on-site solidification facility would be required.
2.
Based on 1,000,000 gallons of contaminated water to be processed, a 70 gallon availability of water volume in a 130 cubic feet liner, 70% availability of the solidification system, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> / day operation, and a two-hour cycle time, the processing time may exceed 40 months.
3.
Based on 1,000,000 gallons of contaminated water to be processed and a 70 gallon availability of water volume in a 130 cubic feet liner, the number of solidified liners that would be generated would exceed 14,285. Handling and transportation of this extremely large quantity of solidified waste would be prohibitive.
4.
The handling evolution required to solidify the contaminated water may involve substantial radiation exposure to personnel.
5.
The potential for leakage and contamination problems may be sub-stantial in operating a solidification facility for y ocessing this contaminated water in this manner.
CONCLUSION Because of the potential for a relatively high occupational dose commitment, in addition to the extremely large number of radwaste shipments, this alternative course of action is not deemed to be feasible..Furthermore, because of the long time for processing, this alternative is not considered a viable one in terms of expeditious clean-up of the TMI-II facility.
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9PECIFIC COMMENTS QUESTION 6 What is the " desired bed loading" for the first zeolite bed. Discuss the manner in which the loading will be determined during operation. Also, discuss loading in terms of integrated dose rate limitations and the potentic1 for gas generation while in store.
RESPONSE
The desired bed loading on the first zeolite bed is that resulting from processing approximately 15,000 gallons. This yields an estimated deposition of 1590 curies Cs 134 and 9085 curies Cs 137 based on 10 cubic foot zeolite beds.
Bed loading will be calculated from the routine sample analyses results. Samples will be drawn from the Hi-Rad Feed and Intermediate Level Sampic boxes, approximately once every four hours during system operation.
Integration of influent radionuclide concentration minus effluent radionuclide concentration yields an accurate estimate of bed loading.
Zeolite is an inorganic material composed of a silicate of aluminum that poses no known limitation in terms of integrated dose or gas production.
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' OUESTION.7 Is it the intent to store the organic cation bed underwater in the storage rack?
RESPONSE
It is our intent to dispose of the organic cation beds in accordance with applicable regulations. However, underwater temporary storage is available and will be used as necessary, s.<
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- QUESTION 8
~You state that cation resin will be limited to less than 75 Ci of strontium
-based on zeolite bed effluent monitoring. State strontium analysis capability at'the TMI site t.o sustain SDS process without delay.
RESPONSE
Radiochemical' preparation areas will be installed in'close proximity to the
.SDS operating area on the fuel handling floor.
Reagents and apparatus necessary to perform strontium separations will be available for use.
Following preparation for counting, samples will be analyzed for total strontium on gas flow or end window G-M equipment dedicated to support SDS operations.
This analytical plan will provide accurate strontium results in a minimum time.
In addition, the data gathered during initial operation may be compared to on-line beta monitor readings to assure further control of cation bed strontium loading.
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.c SPECIFIC COMMENTS QUESTION 9 Describe design and operational features of beta monitors in detail. Are they designed for cesium or strontium breakthrough?
RESPONSE
The beta monitors are end-window G-M tubes manufactured by LND, Inc. tube number 2
7232.
These tubes have a mica window of 1.5 - 2.0 mg/cm thickness and operate at approximately 1000 volts.
Five tubes are mounted in beta monitoring manifold to view zeolite bed effluents as well as cation influent and cation effluent sample lines. The tubes are mounted in holders, and shielded to reduce background. The holdete are positioned to view thin wall sections of teflon tubing.
The remaining monitors are mounted in line on the process input to the cation and the input to the leakage containment ion exchangers. They also view thin walled sections of process piping. Each detector is provided with an ASTON model 205 GM tube Pulse Discriminator to reduce false output pulse from low level noise.
A cabinet will house the High Voltage power supply, Bertan Model 313A, liner ratemeters, Mech-Tronics Nuclear Model 1254, and recorder Tracor Westronics Model MllE.
In addition, a radiation alarm, Ronan Model No. X15-10015 is provided.
The beta monitors are designed to provide assurance of proper bed function during initial operation and as a gross indicator of significant changes in process conditions. Correlation of beta monitor readings to analytical results from the sampling program will allow rapid estimates of process stream conditions so that i
SDS performance may be optimized and still assure that processed water is protected from recontamination in the event of an upset condition.
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SPECIFIC COMMENTS OUESTION 8 You state that cation resin will be limited to less than 75 Ci of strontium based on zeolite bed effluent monitoring. State strontium analysis, capability at the TMI site to sustain SDS process without delay.
RESPONSE
Radiochemical preparation areas will be installed in'close yroximity to the SDS operating area on the fuel handling floor.
Reagents and apparatus necessary to perform strontium separations will be available for use.
Following preparation for counting, samples will be analyzed for total strontium on gas flow or end window G-M equipment dedicated to support SDS operations.
This analytical plan will provide accurate strontium results in a minimum time.
In addition, the data gathered during initial operation may be compared to on-line beta monitor readings to assure further control of cation bed strontium loading.
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-SPECIFIC COMMENTS QUESTION 10 At a: loading of 75 curies.of strontium on the cation bed, what is the expected cesium loading? What is the basis for limiting integrated does rate to 10s rads?
RESPONSE
Based on ORNL column tests, Cs 134 and Cs 137 will not deposit in any significant amount on the organic cation beds.
s The integrated total dose of lo rads for organic resin has been selected because radiation damage studies have indicated that degradation of resin may start to occur at higher total doses.
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t SPECIFIC COMMENTS QUESTION 11 Provide estimates of the expected occupational exposures from the processing of sump and RCS water, and handling of all generated solid waste, including filt'ers. Provide the basis.for your estimates.
RESPONSE
The response to this question is currently being prepared.
It will be sub-mitted by July 31, 1980.
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SPECIFIC COMMENTS QUESTION 12 Identify those components which will require shielding, in addition to those in the fuel pool, to reduce dose rate to less than one (1) mR/hr. Also include components in the SDS offgas system.
RESPONSE
The SDS components above the spent fuel pool water level which are shielded are listed below. The piping associated with these components is also shielded as necessary.
1.
Filter, Beta Monitoring, RCS, Ion Exchange, and Feedpump Manifolds.
2.
High Rad Filter and High Rad Feed Sample boxes.
In addition, the polisher unit will be mounted in a CNSI 195-H cask (lid removed) to minimize radiation levels during processing.
The SDS offgas system is not shielded since only low radiation levels are expected. However, if required, temporary shielding will be added. The Mist E'_'. sination tank and associated piping with its portion of the offgas system is mounted in the surge tank pot covered with a concrete slab (SEE DRAWING 527-A-5007).
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, SPECIFIC COMMENTS OUESTION 13 Provide the basis for the 450 gallons water being released to the pool and for the 233,000 gallons of additional contaminated water.
RESPONSE
Paragraph 2.3 of the SDS Technical Evaluation Report provides a se::nnary of the unexpected occurrences that are evaluated in Chapter 7, Accident Analysis. As presented in Chapter 7, 450. gallons of contaminated water is released from the final filter effluent line at a rate of 30 gpm for 15 minutes. The water leakage is released directly to the spent fuel pool, which has a capacity of 233,000 gallons.
It is assumed that the leakage water mixes with, and is uniformly dispersed in, the spent fuel pool water.
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' QUESTION 14 You state, "It is possible that workers could be contaminated, however, prompt emergency decontamination procedures would prevent major radiation exposures".
Elaborate this statement and describe the location and the method to be used for prompt emergency decontamination procedures. Provide the basis for the 11R/hr. maximum exposure rate assumed.
RESPONSE
Personnel decontamination that may occur in the unlikely event of this hypothetical accident, as with any other possible personnel contamination events, would be performed in accordance with the existing TMI-II procedures for personnel decontami-nation. These procedures have been developed in accordance with the requirements of the TMI-II Radiation Protection Plan. Personnel decontamination would take place in the personnel decontamination area located within the TMI-II Health Physics area.
The basis for a radiation level of 11 R/hr. has been provided in our letter of May 27, 1980 (TLL 251).
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. SPECIFIC COMMENTS QbESTION 15 Discuss utlimate disposition of generated solid radwaste in greater detail.
Describe storage capability for spent liners, spent resin processing, cask availability for shipment, and acceptability of the processed waste at a low-level waste burial facility.
RESPONSE
Storage capabilities for storage of spent vessels has been addressed in our response to Specific Comment #17.
Our present plans identify the use of the CNSI 1-13e casks for shipment of the zeolite and cation resins and the dewatered filters.
Six of these casks are i
scheduled-to be made available in January, 1981 for exclusive use by TMI for disposal of SDS generated wastes. -The CNSI 195H cask will be used for shipment of dewatered polishing resins; two will be dedicated for exclusive TMI use.
Criteria which have been applied to all activities associated with the cleanup effort have included packaging of radioactive waste so that it can be transported i
to, and accepted by, waste disposal sites in compliance with existing regulations.
We recognize the concerns which exist es to the form, content and utlimate dis-posal of the material collected by the proposed operation of the SDS. We are also aware of the various proposals to modify the existing regulations. Because of these circumstances, we have mado provisions, in the design of the SDS and the availability of interim on-site storage for contaminated resins, to permit pro-ceeding with the immobilization of the fission products without foreclosing future options-for treatment to_ improve their suitability for long-term off-site storage or disposal. These features of our planning are key elements in our rationale for proceeding with the' design,' precurement, and installation of the SDS.
l Your letter indicates "further guidance" will be available concerning solid waste disposal as work is pursued on the PEIS. We urge that resolution of these issues be given'very hioh priority through an interagency task force with repre-centation.from all the concerned federal agencies.
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' QUESTION'16 You state that the results of recent experimental work with si. ulated and actual water samples-from TMI-II were used to support the selection of specific ion exchanger materials for the SDS. Provide a copy of such test resulta to the.
NRC for review.
RESPONSE
The attachment to this response is extracted from a preliminary report generated by the Oak Ridge National Laboratory. This report is entitled " Post-Accident Cleanup of Radioactivity at the Three Mile Island Nuclear Power Station", ORNL/TM-7081.
This attachment presents not only the report of specific ion exchanger selection tests, but the development of the SDS processing scheme. The final report will be available in August, 1980.
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o 18 3.4 Treatment of High-Activity-Level Water The largest volume of water generated from the accident can be regarded as high-activity-level water (HALW), which may be defined as' water with a radionuclide concentration in excess of 100 uti/ml 137
- based on the Cs content.
3.4.1 Quantity and characteristics Water in this category is primarily from two major sources.
The first source (90,000 gal) is the water contained in the primary loop coolant circuit which is used to maintain the reactor in a safe condi-tion by removing heat in a natural convection mode.
The second source (540,000 gal) includes (1) the water contained in the reactor containment structure which resulted from the release of water from the primary cool-ing circuit during the early phase of the accident, (2) the volume of liquid transferred into the building through the containment spray system, and (3) a large volume of water which was released through miscellaneous equipment (pump seals, space coolers, etc.) during the post-accident period.
Depending on the fission product removal efficiency of the Epicor-11 treatment system, the wa'.er in the three reactor coolant bleed tanks may be processed later, if necessary, in this system. Table 2 outlines the volumes of solution and concentrations nf the primary rh 'ionuclides contained in this material.
In addition to the increased radioactivit'y contained in this water, the ionic contamination mandates that the water be given special treatment.
The major ionic constituent is boron, present at levels up to approximately 2600 ppm, which is used as a neutron poison in the coolant circuit.
In addition to this element, sodium is also present as the result of spray activation with HaOH solution.
Table 3 presents significant information related to the chemical and ionic characteristics of the HALW solution.
_3.4.2 Necessity for emergency tankaoe Soon after the accident, it was recognized that additional tankage
'.'ould be h. quired to receive the HALW accumulating in the Reactor Contain-ment Building.
There was considerable concern that the water level in the e
O I
y ~. 7:...
.n.
7
19 Table 2.
Volumes of sc'ution and concentrations of principal nuclides a
contained in high-activity-level water and precipitate t
Primary Reactor Containment Building loopb Top Middle Bottor,~ ~
540,000 Volume, gal 90,000
=
Radionuclide content of water, uCi/ml 137Cs90-110 176 179 1 74 13'Cs 18-22 40 40 40 505r 17-19 2.7 2.9 2.8 e95r
- 305-330 43 41 42 d
d 0.076 d
0.080 1291 0.079 1311 0.5-0.6 0.012 0.012 0.013 l'0Ba/140La 4-5 0.09 0.078 0.14 3H 1.03 1.05 1.01 Gross a 3.4 1.60 1.2 2 1.3e 5.4 2 2.09 Radionuclide content of precipitate (on total ml of bottom sample)I,uCi/ml 9
95Nb 0.13 95Zr 0.05 103Ru 0.06-129mTe 0.40(: 0.17) 1311 0.013 134Cs 0.015 137Cs 0.064(:0.020) 14 OBa/1'O' a 0.04/0.11 89+905r 2.78 (one sample only) aRadionuclide content of surface deposit on painted steel plug, pCi:
95Nb 1.7 1255b - 0.5 1311' - 0.3 103Ru - 0.6 125mTe -
- 0. 5 13'Cs - 0.5 106Ru - 0.4 127mTe - 7.8 Id7Cs - 2.1 1135n - C.2 129mTe - 23.6 1*4Ce - 0.2 bAs of July 17, 1979.
As of August 28, 1979.
Samples from top and middle--light yellow solution; c
sample from bottom--greenish with dirty green precipitate.
All samples have a similar radiation level; pH is 8.1 (as measured in the top sample).
d[ nits are'ug/ml.
8
- nits are dpm/ml.
U f
Average of two samples.
9As of August-29, 1979.
l i
- =. *.-W enn e e d%. ms.es pe--..,e s.
'h, 9%
20
- Table 3.
Chemical and ionic characteristics of high-activity-level waste solution and precipitate Containment vessel liquid (pool)
Element Top Middle Bottom Sample concentration, pg/ml (ppm)
Al 3
3 3
B 1950 2200 1900
- Ca.
10 10 8
C1 10 15 8
Cs 0.6 0.7 0.7" a
a Cu 50.2 50.2 10 Fe 0.6 1.1 1.8 K
4 4
4 b
b b
Li 1.6 1.6 3,4 Na 1080 1200 1200 Ni 50.2 50.2 3
P 0.3 0.3 0.2 Rb 0.3 0.3 0.3 5
9 8
7 Zn 0.5 0.5 0.4 c
c c
Pu 0.010 0.011 0.033 c
c c
U 7
13 28 Solids from bottom sample, ug/ml (m1 of total bottom sample) 0.16 In Cu 54 A1 7
Ca 52 a ission product.
F bgg;,99; 7t$,
Cppb, not pg/ml.
-"W
- M
'am.s eAmy,,,,,
~
21-building might rise high enough to flood and thereby prevent operation of some of the vital instruments that monitor and control the reactor.
a.
These instruments had not been designed for operation in a hostile environment of high re11ation fields and submergence in liquids.
Sub-
- sequently, a tank farm was designed and installed in the fuel storage pool A contained in the Unit 2 Fuel Handling Building.
The original scheduled time for this project was 90 days; however, some acceleration of the schedule was required when the leakage rate increased and it appeared that 50,000 gal of primary coolant might have to be trans-ferred to one of the primary coolant bleed tanks.
But, two major leaks were repaired, and this transfer of primary coolant was deemed unnecessary.
The design of the tank farm '
arporated six tanks with a total capac-ity of 110,000 gal. Two 25,000-ga, tanks were installed near the bottom of the pool and connected to each other with a standpipe for which devices were to be designed later for sampling and solution transfer.
Figure 8 is a
~
photograph of these tanks taken during installation in the fuel storage
~.
pool.
Four 15,000-gal tanks, installed above the 25,000-gal vessels, were connected to a second standpipe.
Shielding for all vessels was provided I -
by installing concrete slabs on the top of the pool structure.
An inde-pendent off-gas treatment system was also installed on top of the tanks to decontaminate any gaseous effluent that might be evolved during tank opera-tions.
Fortunately, the tank farm system was not needed for its original purpose. This eliminated the potential risk of transferring additional liquid out of the Containment Building (into the tank farm) and, thereby, increasing the inventory of iodine in a structure external to the reactor system.
The decision had been made not to disturb the solution in con-131 tainment so the 8-day 1 could decay to safe limits (see Sect. 3.4.3).
The tank farm will be utilized as an integral part of the HALU trea.-
ment system.
3.4.3 Processing justification In order to proceed with the recovery of the reactor system, the liquid being held in the Containment ouilding 'will require removal and treatment.
Currently, this liquid is standing at a depth of approximately 7-1/2 to E ft l
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- u 23 and is covering several components, including instruments.
Although the leakage of water into the containment area has been minimized, the pos-f, sibility for increased leakage continues to exist.
In the early phases 131 f,'
of the accident, the water in containment had 1 concentrations esti-4 mated to be of the order of 10 pCi/ml.
Because of this factor, every effort was made to avoid disturbing this solution until the radioiodine had been allowed to decay.
Figure 9 shows the fission product decay curves, based on the analysis of primary coolant, for the radioactive nuclides that were of greatest concern with regard to treatment of the solution and indicates the most desirable processing period.
3.4.4 Flowsheet development Because of the unique nature of the water to be decontaminated, cer-tain phases of the development of a flowsheet were required. The first phase of the flowsheet development work involved the selection of a suit-able exchange medium for the processing of this solution.
Limited samples of actual TMI primary loop water were tested in an ORNL hot cell to estab-lish the characteristics of sorbeats believed to be selective for the pre-34-137 89-90 dominant Cs and Sr isotopes. A total of seven materials, both organic and inorganic ion exchange sorbents, were tested (Table 4).
Results from these tests established that one of the zeolites, AW-500 (an inorganic exchanger), was highly selective for the cesium isotopes under the conditions prevailing at TMI and that Dow HCR-5, an organic resin, would be suitable for selection of the strontium isotopes.
Following these studies, a series of small-scale column runs was made to verify the number of column volumes that could be expected from the various sorbents using solutions adjusted to TMI's water conditions.
Infonnation resultinc from these studies was then applied to a " scale-up" computer program at the Savannah River Laboratory to provide infonnation on design of the ion exchange columns.
Pertinent infonnation yielded by these studie.; is pre-sented in Figs.10 and ll.
0 3.4.5 Process design objectives and criteria Chem-Nuclear Systems, Incorporated (CNSI), was awarded a contract fror the General Public Utilities Services Corporation to provide a flowsheet for e
=
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24 ORNL Dwg 79-13239 A 108 v
y SUGGESTED PROCESSING 8311 PERIOD 103
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1979 1980 Fig. 9.
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ORNL Dwg 79-15328 t
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Table 4. Distribullon coefficients (Kg) for primary coolant
.Lt.
Phose Sr Ko Cs Kd
'O 'IO Second Second Enchanger 5 min 2 hr pass 5 min 2 hr pass Final pH (ml/g)
Primary water os received - 960 ppm No HCR-S 50 230 20 3700 8700
>330 3.6 27
~
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~O SO 39 280 1800 1400 144 m.
136 180 1300 AW-SOO
~2 100 Clino.
22 96 130 135 990 3000 8.6 67 Glass 96 250 54 160 440 400 8.6 68 Choicool 20 50 60 O
O.7 2
8.7 42 y
Primary woler + 50% NaOH to give 4000 ppm No
,i' HCR-S O
265 19 40 48 9000 8.5/3.7 56 55 l40
-l z_9or
~o AW-SC,0 27 54 86 238 1400 2200 9.7 54 Clino.
20 42 40 75 400 1300 10.4 60 Glass 55 270 38 70 350 320 10.1 56 l No Titanate 1500 4200 16 92 70 70 10.1 58
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28 the treatment of HALW.
CNSI then employed Allied-General Nuclear Services to assist them in the design of a suitable system.
A series of design objectives was drafted to provide a system that 1.
was totally integrated and as independent as possible from existing waste and off-gas systems at TMI, 2.
could remove the fission products in the water to a level that would meet designated TMI release requirements, 3.
could be operated with a minimum of exposure to personnel and a negligible risk to the general public, and 4.
would accomplish these objectives in a timely and cost-effective manner.
The following design concepts were used in order to accomplish these objectives:
1.
CNSI has designated an ion exchange system for the removal of re.'onuclides from the water.
2.
The ion exchange columns are to be located ur@rwater in the spent fuel pool, which will provide for containment and personnel shield-s ing.
3.
To the extent possible, all-welded stainless steel construction is specified to minimize the potential for leakage.
4.
Lead or equivalent shielding is provided for pipes, valves, and vessels (except those located underwater) that could carry con-taminated water.
5.
Contaminated water from the containment sump is to be filtered, collected in one of two batch tanks, sampled,'and then pumped continuously to the ion exchange system.
6.
Two parallel filters and ion exchange systems are to be used to provide capacity and to ensure continuous operatior..
7.
Continuous effluent monitoring systems are to be provided to ensure adequate control of the operation.
8.
The valving systems are to be designed to minimize the impact of operator error.
9.
The system is to be designed for a minimum of 2 years of service.
D-
---*g
. - - -. _..... ~ _,_ _ '
~
29 3.4.6 System and process description The conceptual system design consists of the following elements:
1.
a feed filtering system; 2.
four feed tanks with independent mixing, sampling, level-measuring, and feed-pumping capability; 3.
two parallel primary ion exchange trains, each comprised of 3
three 7-ft beds of zeolite exchange media; 4.
two parallel ion exchange beds containing organic cation resin for removal of strontium; 5.
a continuous monitoring system and sampler system for control of bed loading; 6.
a secondary containment system for the filters, primary and cation beds, and radiation shielding for piping, valves, sam-plers, and monitors; 7.
a carbon-cation " filter" bed for removal of trace fission products; 8.
a mixed-resin ion exchange polishing bed for removal of trace fis-sion products that are not trapped on the primary cr cation beds;
~
3 9.
a 195-ft ion exchange bed for removal of boron (should this be required);
10.
two monitoring tanks for collecting and sampling the treated water prior to discharge and storage; 11.
an off-gas system for treating and filtering gases and air from the system; 12.
associ~ated piping, valving, and structural' supports required for placement of the system components; and 13.
auxiliary systems, including underwater ion exchange column stor-age, a column dewatering system, a system for solidifying the resin beds with concrete, and analytical equipment.
Feed filtering system.
This system consists of two parallel under-water filter units.
Each unit is a cluster of 5-u cartridge filters pro-tected by perforated metal screens and contained in a 24-in.-diam x 54-in.-
~
long column.
Inlet, outlet, and vent connections are made with Hansen quick-release valved couplings which are remotely operated fr'.n the top "About 90 to 95% of the strontium and essentially all of the cesium will be loaded on the zeolite.
?
e E=-
,1 c
e e
30 of the pool.
A gamma monitor, located in a dry well adjacent to the
' filters, and inlet and outlet pressure gauges are provided to monitor and control solids loading.
Load limits will be based on available system pressure and/or the surface dose limit for the filter cas,k.
The capacity of each filter is about 15 gpm.
Feed will be pumped from the reactor containment sump, through the filters, and into the feed tanks on a batch basis.
Feed tank system.
This system consists of four of the 15,000-gal tanks located.in the emergency tank farm.
Each tank will have an air-lift system for circulating the fced solution and for raising it to a small pump head tank (approximately 8 in, diam x 3 ft long).
A sampler will be provided on each tank to pennit characterization of the feed prior to sending it to the ion exchange system.
Level-measurement in-struments are also provided.
Feed will be pumped to the ion exchange columns using one of two seal-less, magnetically coupled pumps rated at approximately 15 gpm each.
The samplers, headpots, and pumps will be located in a shielded enclosure.
Primary ion exchance columns. This system consists of six under-3 water columns (24 in. diam x 54 in. long), each containing 7 ft of Linde AW-500 zeolite resin.
Inlet, outlet, and vent connections are made with remotely operated Hansen couplings (similar to the filter con-nections).
The beds are arranged in two parallel trains of three each, with. piping and valves provided to operate either train individually or both trains simultaneously.
Loading will be controlled by feed batch size, operating time, effluent sample analyses, and continuous monitoring.
Flowmeters are provided for the total feed stream and the individual feed streams to the two parallel trains.
The maximum loading per column will be 65,000 Ci (cask limit).
These columns are expected to remove greater than 99% of the cesium and 90 to 95% of the strontium from the water.
When the desired bed loading is achieved on the first bed of the train,
- the flow to the train will be stopped.
The treated water will then be displaced with fresh water (through the downstream beds), and the first
~
bed will be disconnected and moved to the storage rack using the pool area crane.
The second bed will be disconnected, moved to the first position, and reconnected.
Likewise, the third bed will be moved to the second o
-+
-.-.L 2-
- _,,C,
31
[
position, and finally a fresh bed will be placed in the third position.
(voic: This approach is considered to be superior to a valving manifold which would permit rerouting of the streams becauce it minimizes the
, potential of a serious operator valving error.
A valve switching approach
~_
was considered; however, a valving error could result in totally bypassing the primary beds.
The current design prevents this from occurring.)
Cation columns.
Two additional underwater columns (of the same size i
and type as the primary columns) are located immediately downstream. These columns are loaded with organic cation resin (probably Dow HCR-5) for re-moval of residual strontium.
Column loading will be limited to 75 Ci of strontium based on primary column effluent monitoring and analysis.
The columns are arranged to be operated singly or
'multaneously in parallel.
The curie limit is based on restrictidy' the mtegrated radiation dose to 8
the resin to less than 10 rads.
Samoling and monitoring system. As previously described, the feed will be collected in batches, sampled and analyzed, and fed continuously to the ion exchange system.
Sampling lines are provided on the effluent e
streams of each of the primary beds and cation beds.
A continuous sample flow from each operating bed effluent wi'l be passed'through a beta scin-tilletion monitor.
A gama monitor is louated on the primary effluent line, and a local area radiation monitor is located near the valve box.
Each of the monitors will have an alarm system and a high-radiation trip point which will close an automatic valve of the main feed line, thereby stopping the operation in case of a leak or a serious breakthrough.
Shielded samplers are also provided to permit sampling at any time.
Con-tinuous sample effluent is routed to the off-gas separator tank.
Containment and shieldina.
The filters, primary beds, cation beds, and their associated remote connectors and flexible lines are contained in submerged boxes designed to confine any contaminated leakage.
Each box is connected to a pump manifold, and a continuous flow of about 10 gpm is maintained through each box. The containment pump effluent is discharged to an additional mixed-bed resin column and is released back into the pool.
~
A large, shielded box is provided for the various valves and piping which are required for feeding, flushing, and venting the columns. This box will i
,f
f r
32 e
be located on the edge of the pool, with shielded lines entering and exiting the box f rom the bottom.
Valves will be operated manually using extension handles that penetrate the top of the box.
The valves will also be serviced from the top of the box.
A separate shielded box will be provided for the column samplers.
All lines' which are normally radioactive, or could potentially be radioactive, are to be shielded.
Shielding will be designed to limit the operator dose to 1 mrem /hr at contact.
Wherever possible, pipe runs will be located under-water to minimize the neeci for lead shielding.
3
" Filter" bed. This is a 10-ft ion exchange bed containing carbon and cation resin.
This bed is intended to remove trace fission products from the water.
Load control will be based on the cation effluent moni-tor and the column radiation level.
3 Mixed-resin polishing system.
This system consists of a 195-ft bed of mixed anion and cation resins enclosed in a special shielded cask.
This unit, which is also intended to remove trace fission products from the cation effluent, will be located above the water.
Load control will be based on the cation effluent. analysis and the radiation level at the surface of the container.
3 Boron removal bed.
If desired, a 195-ft resin bed, enclosed in a shielded cask, is provided for boron removal.
Effluent sample analysis and surface radiation level monitoring will be used to control loading.
Monitoring tank system.
Two monitoring tanks of about 20,000-gal capacity will be provided'for collecting and sampling the treated water prior to disposal.
The tanks will be provided with mixing, level-measur-ing.and sampling systems, in addition to transfer pumps.
Piping will also be provided for recycle of the treated water in the event that it does not meet disposal specifications.
Off gas system.
The feed tanks, valve boxes, monitoring tanks, and column vents are connected to this system.
The system consists of a sepa-
-rator tank (which also serves.as a sampler effluent collector and valve-box drain tank), an electric off-gas heater, a roughing filter, a charcoal bed for iodine retention, two absolute (HEPA) filters, and a centrifugal blower.
The discharge of this blower will be monitored and routed to the reactor vent system.
A pressure control system is provided for the blower to 8
k
-~
r--*e e.~.g
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- * * - = ' *.
owsw
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s
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~
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position, and finally a fresh bed will be placed in the third position.
(vo:c: This approach is considered to be superior to a valving manifold which would permit rerouting of the streams because it minimizes the i-
, potential of a serious operator valving error. A valve switching approach was considered; however, a valving error could result in totally bypassing the primary beds.
The current design prevents this from occurring.)
Cation columns. Two additional underwater columns (of the same size and type as the primary columns) are located immediately downstream. These columns are loaded with organic cation resin (probably Dow HCR-S) for re-moval of residual strontium.
Column loadir9 will be limited to 75 Ci of strontium based on primary column effluent mclitoring and analysis.
The columns are arranged to be operated singly or simultaneously in parallel.
The curie limit is based on restrictidf the ir.tegrated radiation dose to 0
the resin to less than 10 rads.
Sampling and monitoring system.
As previously described, the feed will be collected in batches, sampled and analyzed, and fed continuously to the ion exchange system.
Sampling lines are provided on the effluent streams of each of the primary beds and cation beds.
A continuous sample flow from each operating bed effluent will be passed 'through a beta scin-
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tillation monitor.
A gamma monitor is located on the primary effluent line, and a local area radiation monitor is located near the valve box.
Each of the monitors will have an alarm system and a high-radiation trip point which will close an automatic valve of the main feed line, thereby stopping the operation in case of a leak or a serious breakthrough.
Shielded samplers are also provided to permit sampling at any time.
Con-tinuous sample effluent is routed to the off-gas separator tank.
Containment and shielding.
The filters, primary beds, cation beds, and their associated remote connectors and flexible lines are contained in submerged boxes designed to confine any contaminated leakage.
Each box is connected to a pump manifold, and a continuous flow of about 10 gpm is maintained through each box.
The containment pump effluent is discharged to an additional mixed-bed resin column and is released back into the pool.
l A large, shielded box is provided for the various valves and piping which
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This box will e
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Piping and equipment arrangement. The feed pumps, feed samplers, 7,
-and head pots will be located in shielded enclosures directly over the
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As previously described, the filters, primary beds, and cation beds are to be located underwater in special containment boxes.
These boxes and columns are to be supported along one side of the pool on a structural steel rack which hangs from the edge of the pool.
This rack, which also serves to support the system, will include an operating platform from which the remote connectors are to be handled.
Auxiliary systems.
A storage rack, sized to handle the projected number of filte/s#and ion exchange beds required to treat the containment sump water, will be provided.
A special rack will serve to displace the water from the columns and filters and to dry them prior to loading into the transfer cask.
This rack will be located in the cask loading pool.
A system for solidifying the filters and ion exchange columns with concrete will also be provided. An additional system will be included for
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analyzing feed and effluent samples.
Gamma and beta proportional counters and the necessary equipment for boron and sodium analysis will be provided as a minimum.
Figures 12-15 show details of the liquid waste (HALW) decontaminatior, system.
(The HALW treatment system is in no way related to either Epicor-I or.Epicor-II.)
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SPECIFIC COMMENTS QUESTION 17
'You state that only 200 bed volumes (approximately 2000 cubic feet or 15,000 gallons) of waste water are anticipated to be processed before the fir',t zeolite bed is removed and replaced.
This will require more than 60 zeolite bed replacements (excluding the subsequent zeolite beds, cation beds, and filters) to process one million gallons of waste water through the SDS. Justify the storage capacity provided for only 40 spent vessels in the fuel pool.
- RESPONSE At the time the SDS Technical Evaluation Report was written, system design required the construction of facilities within the fuel pool or sufficient size to provide for the underwater storage of 40 spent vessels in the pool.
Upon receipt of further guidance, as indicated in the NRC letter dated May 28, 1980,'
H. R. Denton to R. C. Aruold, we intend to process SDS resins such that resin disposition options are.not foreclosed. These options are:
- 1) ship dewatered SDS resins, 2) transfer SDS resins for solidification, or 3) transfer the SDS resins to interim on-site dry storage.
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. SIPECIFIC COMMENTS QUESTION 18 Provide the flow capacity of the pool secondary containment demineralis.ars.
RESPONSE
The secondary containment demineralizers have a flow capacity of 50 gpm.
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SPECIFIC COMMENTS QUESTION 19 Provide the administrative surface dose rate limit for change out of the SDS filters.
RESPONSE
Calculations from the ANISN computer code have indicated that contact surface exposure rates of 5000 R/hr on contact from Cs 134 and Cs 137 loaded on zeolite result in readings acceptable for shipment.
In vire of the fact that the major-ity of solids are Sr 89/90 beta emitters, surface exposure rates on filters should not become prohibitive before the maximum loading is reached.
However, underwater monitors are located in close proximity to the roughing and final filters within the leakage containments. Administrative limits will be placed at a f raction of the above exposure rate to assure that spent filter vessels may be safely loaded in the CNSI 1-13e cask.
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SPECIFIC COMMENTS QUESTION 20 Describe how the shipping cask and its contained liner will be drained prior to movement out of the pool.
RESPONSE
The CNSI l-13C cask is designed for underwater loading of vessels. With the drain plug removed and the vent open, the cask will be lifted above the pool surface and allowed to gravity drain.
The contained liner will be dewatered at the SDS dewatering facility prior to insertion it; the shipping cask.
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v SPECIFIC COMMENTS QUESTION 21 Describe how spent SDS filters will be changed out and packaged for ultimate disposal.
RESPONSE
SDS prefilters and final filters are packaged in vessels which are si'nilar in design to the ion exchange vessels.
These vessels are of the identical physical size and utilize Hanson quick disconnect fittings which allows the change out of vessels to be accomp3 ed through the use of remote underwater handling tools similar to the tools used for the ion exchange vessels. Once the filters are changed out, they will be dewatered in the dewatering station, (as depicted on CNSI drawing 527D-A-5011). Af ter dewatering, the filter vessels will be placed in the shipping cask (CNSI 1-13C) underwater.
The cask will then be decontaminated and prepared for shipment via truck to the disposal site.
Additional studies of filter media and techniques are currently under investi-gation which could offer additional options for processing.
The present technique is in compliance with existing Federal Reculations and the ability to adequately perform the dewatering task has been studied and testing accomplished, with the results published in the attached ACNS document dated March 14, 1980, entitled, " Dewatering and Drying Tests of the CNSI-TMI Prefilter and Final Filter".
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4 DEWATERING AND DRYING TESTS OF THE CNSI-TMI PREFILTER AND FINAL FILTER by James D. Iogue March 14, 1980 Allied-General Nuclear Services Post Office Box 847 Barnwell, South Carolina 29812 Prepa; d for Chem-Nuclear - rvic T,
Inc.
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r ACKNOWLEDGMENT The author gratefully acknowledges the assistance of Terry D. Bryson and Preston Fields of the Separations Production Department in the perfor-mance of these tests.
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ABSTRACT The dewatering and drying tests were conducted on the Submerged Demin-eralization System (SDS) to determine if they could be successfully dried. Dewatering and drying is required in order to meet existing and future criteria for shipment and burial of vessels.
The tests were started on February 8, 1980, and concluded on February 29, 1980.
This report contains an evaluation of test results and recommendations needed to satisfactorily meet drying criteria.
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TABLE OF CONTENTS PAGE ACKNOWLEDGMENT ABSTRACT G14SSARY
1.0 INTRODUCTION
1
2.0 CONCLUSION
S AND RECOMMENDATIONS 1
2.1 Conclusions..
I 2.2 Recommendations I
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3.0 DESCRIPTION
OF TEST 2
4.0 TEST RESULTS AND EVALUATIONS.
2 5.0 ATTACHMENTS 3
P TABLES:
Table 1 - Weights, Capacities, and Dry Storage Test Results.
4 Table 2 - Blowdown Times 4
f FIGURES:
Figure 1 - Blowdown Setup for CNSI-TMI Filters.
5 Figure 2 - CNSI TMI Prefilter Test Unit 6
Figure 3 - CNSI TMI Final Filter Test Unit 7
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1.0 INTRODUCIION Commercial low-level waste burial site criteria requires that wastes be free of liquids; that is for wastes received until January 1, 1981, the amount of liquid allowed is less than 1% of the total volume, not to exceed one gallon and 0.5% of the total volume for liquid wastes received after that date. Federal Transportation Regulations [49 CFR 173.393(G)] require enough abscrbent material to absorb at least twice the volume of liquids present in each package.
The present packaging design for spent Submerged Demineralization System (SDS) prefilter and final filter vessels does not allow for the addition of absorbent materfal; therefere, the SDS filter vessels mus t be dry prior to shipment.
The purpose of these tasts was to verify that the SDS prefilter and final filter vessels could be successfully dried; and that af ter drying, the SDS filter vessels cculd be stored for short periods of time under water without any in-leakage of pool water. These tests a M determined if acy weight must be added to the dry submerged filter vessels to prevent them from floating.
For these tests, each filter vessel was weighed empty and again after filling with water. To determine the length of time required to achieve dryness, the filters were blown down with dry air from 10 minutes to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Af ter each blowdown, any water which had not been removed from the filter vessel or that had drained back into the filter from the connecting piping was measured.
After drying, the filter vessel was pressurized with air and stored in the pool to determine if there was any in-leakage of pool water.
2.0 CONCLUSION
S AND RECOMMENDATIONS 2.1 Conclusions (1) The filters can be successfully dried and once dried can be stored for short periods of time under pressure without any in-leakage of water.
(2) The dried filters require the addition of ballast to prevent then from becoming buoyant.
2.2 Recommendations (1) The specific ballast requirements for each type of filter will be calculated and incorporated in the design of the units.
(2) To prevent in-leakage of pool water, the male hansen connector on the prefilter inlet nozzle should be connected to its female counterpart oefore attempting dewatering and drying. t
3.0 DESCRIPTION
OF TEST Two mock-up filter vessels were fabricated per CNSI Drawings (527-D-5010, Ion Exchange Mock-Up; 527-D-5002, Prefilter-125 Micron; 527-D-5003, Final Filter-10 Micron; and 527-D-5004, Prefilter/ Final Filter Details), except that instead of using dished heads on the filter top head, flat plate was used and a pipe plug was installed in the bot-ton of the filter housings. Plexiglas was used for the vessel walls.
The flat plate was used to reduce cost and fabrication times, the plug was installed to allow for posttest draining, and the Plexiglas was used to allow inspection of the filter internals (see Figures 3 and 4).
The volumetric capacity of each filter was determined by measuring the water needed to fill the filter to overflow. A weigh tank and scale was used for this measurement. After filling, the filter inlet nozzle was connected to a dry air supply and the filter blowdown nozzle was routed to the weigh tank.
Dried compressed air was used for the blowdown operation. The air was regulated to 40 psig and 10 sefm; the pressure was limited by the Plexiglas.
Af ter the initial amount of water was discharged, two blowdowns each at 10 minutes, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> were performed.
Af tnr each blowdown, the vessel was visually inspected for wetness and
- he plug was removed to collect and measure the remaining water.
To t iralate actual conditions, the filter vessel was tested at approxi-mately 20 feet below the weigh tank. A blowdown was also accomplished underwater to simulate remote conditions.
(Figure No. I shows details of setup.)
After the filter unit was dried, lead was addad for ballasc.
The extra weight enabled the filter to remain submerged after drying and added sufficient stability to the filter vessel to prevent it from shifting l
during hookup operations. The filter units were pressurized and sub-merged under approximately 20 f eet of water and stored for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Af ter storage, the filters were removed from the pool and inspected for water.
4.0 TEST RESULTS AND EVALUATIONS Following blowdowns of 10 minutes through 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the surface of the
-filters and the top and bottom plates remained wet.
Liquid drained from the filters when the plug was removed. To achieve the dryness necessary to ship the filter vessels, blowdown times of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> for the prefilter and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the final filter are required.
After the 16-and 24-hour blowdowns, the filter internals were dry and no liquid was detected when the drain plug was removed (see Table 2).
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e Dry compressed air at a temperature of approximately 68'F was used for these tests. The air was regulated to 40 psig and 10 scfm.
Heated air and air at higher pressure and volume would enhance the drying.
Increasing the pressure and heating the air could not be accomplished during these tests due to the pressure limitations of the Plexiglas.
The empty weight of the prefilter is 522 pounds and the final filter is 678 pounds. Extra ballast was required to submerge both filter vessels; in fact, 381 pounds of lead were added to the bottom skirt of each ves-sel. This additional weight allowed the vessel to be submerged and provided sufficient weight to prevent the vessels from shif ting during hookup operations (see Table 1).
During the test, it was found that the male hansen connectors would op en when submerged to a depth of 10 feet or more allowing in-leakage of pool water. After d Jing, the filters were pressurized to 15 psig and stored in tre pool submerged under 20 feet of water. The filters were removed after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and no water was detected (see Table 1).
Also, the check valves installed in the feed line of the prefilter will not hold pres-sure in the vessel and will allow in-leakage of water,unless the plugged female hansen connector is installed.
5.0 ATTACHMENTS
- - CNSI-TP-2, Mock-Up Filter Ballast and Drying Test - Additional Instructions for CNSI-TP-02, Mock-Up Filter Ballast and Drying Test.
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9-ATTACHMENT NO. 1,
PAGE 1 OP 4 98 0 REV DATE of THIS ILSUL PNEl'A";[3 SY PAGE CNSI-TP-02 0-o*Tc or P' lev. issuc J. I.ogue et inLt 1
HOCK-UP FILTER BALLAST AND DRYING TEST l
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1.0 INTRODUCTION
I
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Contiminated water at TMI will pass through a set of roughing and final f'ilters, before i
final purification in the ion exchangers. The filter units, af ter loading up, will be e
blown dry and shipped for burial. A mock-up filter unit will be used to simulate the drying process. The mock-up units will be filled with demin water and then blown dry.
Data collected from this test will deterinine the amount of time and air to accomplish drying.
2.0 PURPOSE l
I 2.1 This procedure provides the 3natructipss to collec5 the data necessary to deter-mine the time and amount of air required to remove enough water, so that the filters are considered dry and ready for shipment. Additional testing may 'be required.
2.2 Acceptance Criteria 2.2.1 This procedure will be complete whzn all steps are signed off; and
- 2. 2, 2, An evaluation is performed Eo determine if drying 'is sufficient.
3.0 SCOPE 3.1 Equipment mockup prefilter and final filter.
3.2 Area FRSS 4.0 PREREQUISITES Pressure tect of moc filters must be performed before units are used. ?ressure test complete O O 1.__
'2. / S/80 CohnizagEngineer T/D.
5.0 SAFETY PRECAUTIONS 5.1 Follows instructions as outif ned in the llNFP Industrial Safety Manual.
5.2 DO NOT EXCEED 40 PSI on filter unit.
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_6. 0 EMERCENCY ACTION PAGE 2 0F 4 None anticipated.
7.0 TEST EQUIPMENT 7.1 350 Kg Scale 7.2 55 Callon Drum i
7.3 Humidity Testor 7.4 Dry Ice or Liquid Nitrogen 8.0 PROCEDURE l
8.1 CALIBRATE 350 Kg scales per S-COP-90.-3.
a bper$ tor 7
T/D 8.2 Weigh Mockup filter units.
8.2.1 WEICl! empty units.
Profilter 7.'%.B Kgs Final Filter
'bO7. 49 Kgs Operator j
8.2.2, SUBMERGE filter unit into east CUP onto TMI test stand.
l 8.2.3 NOTE any difficulty to submerge.
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Remarks Adriad ?.b l th> of! cad -lo bo4h G 14r a Operator n % d m e-3 c.
8.2.4 RDIOVE test unit from CUP.
8.2.5 FILL UNIT's with water.
8.2.5.1 ATTACH water hose to fill line.
i 8.2.5.2 ATTACH hose to vent.
8.2.5.3 FILL unit with water until it comes out of vent and DETACH hose.
8.2.6 WEICHfilledunits.(wo<pl o f W e'%'" "*N Profilter '2 7-1 'h Kgs Final Filter
- 9. 4 *5. l Kgs Operdtor 8.3 SUBHERGE filled filter unit into cast CUP onto TMT. tent stand.
8.4 CONNECT regulated inntrument air supply to addition nozzle and ROUTE discharge line irom blowdown nozzle to drum and secure.
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4.5 START air to filter unit at,2' SCFM.
1nno 8.6 SilUTOFFaireftsr5cir.uteYe 10 m us ca C4u wa4=<-
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Prefilter '274 77 Kg Final Filter C 47. 05 Kg', Operator CA Y
8.8 REPEAT step 8.5 and measure humidity of discharge _41ri-- - O mW p '1A/&
Prefilter Final-Fnter Operator 8.9 SHUT OFF ai ft tes and leave 20 to 30 psig pressure in filter unit.
~fr 11ter psig Final Filter psig Operator
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8.10 REMOVE hose from nozzles and remove filter test unit from CUP.
8.11 INSPECT for water.
s.e 8.11.1 VISUALLY INSPECT internals for water.
Remarks Pa lb d
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7.- 9-10 Operator T/D i
8.11.2 REMOVE plug from test unit bottom and measure any liquid present.
Aucunt 1350 mi O'T'N
. Operator T/D 8.11.3 If unit appears to be dry, SET test unit aside for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and repeat
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8.11.2.
8.11.4 If unit contains water contact Cognizant Engineer for further instructions.
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DATA TABLES TABLE 1 WEIGHTS, CAPACITIES, AND DRY STORAGE TEST RESULTS Weight Filter Weights Required Dry Storage Vessel Empty Capacity for Ballast Time Press Liquid Pre 522 lbs.
274.3 1 381(1) 48 15 None Final 678 lbs.
245.1 1 381(I) 48 15 None (1) Maximum Weight - Limited space and weight configuration prevented adding additional weight.
TABLE 2 BLOWDOWN TIMES Volume Remaining Blowdown Pref 11ter Final Filter Times Run 1 Run 2 Run 1 Run 2 10 minutes 1350 m1(1) 290 ml 2156 ml 1920 ml 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 80 ml 70 ml 180 ml 175 ml 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 30 ml 22 ml 153 al 140 ml 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 9.5 ml 13 al 22 al 31 al 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (2) 0 0
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0
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(2) Visual inspection of the final filters during the 24-hour blowdown indicated wetness of the filter af ter 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. -- i2
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SPECIFIC COMMENTS QUESTION 22 The SDS feed tanks can.be considered as one feed tank with a capacity of 60,000 gallons since there are no ~ valves between these feed tanks and two monitor tanks,
~ which receive processed waste water have a capacity of 19,000 gallons each.
- Desc' ribe the criteria.for determining the batch volume without risking radio-activity contamination of a'. ready processed water in the monitor tanks. The in-line monitors may not provide. the sensitivity required to detect desired radioactivity breakthrough due to the radioactivity plateout and background radiation buildup.
RESPONSE
It is anticipated that 200 bed volumes (approximately 12,000 gallons) will be processed prior to zeolite bed removal and replacement.
Following SDS shutdown, the nearly filled monitoring tank will be isolated and the empty tank valved in prior to startup.
The beta monitors-will be monitoring the effluents of the first, second, and third zeolite beds. Additional monitors are provided for the cation influent and effluent as well as 'the in-line monitor before the polishing unit. These
= detectors coupled with the routine sampling and analyses program will provide early -detection of increasing levels at various points in the process stream. All beta monitors view a thin-walled section of piping that may be removed and replaced, should plateout restrict sensitivity.
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o SPECIFIC COMMENTS OUESTION 23 Provide estimated contact dose on spent filter, zeolite, and cation vessel cask.
RESPONSE
Activity loadings of the filter units has been estimated based on the expected '
activity of filterable solids in the 125 micron and 10 micron range and includes the amount of residual liquid remaining after dewatering. The coarse filter has been estimated to have a contact dose reading of approximately 3 R/hr. and the final filter to have a contact dose reading of approximately 400 mr/hr.
The contact dose rate on the zeolite bed has been estimated as 30,000 R/hr.,
based on a loading of N 9400 Ci of Cesium. The zeolite bed will only be transferred with the shield cask in place. The contact dose rate on the shield cask is less than 1 R/hr. with the above curie loading.
Column testing has indicated very little ion exchange of radionuclides on the cation bed, and, therefore, insignificant contact dose rates will be present.
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- s SPECIFIC COMMENTS QUESTION 24 You have provided an in-line monitor in a common recirculation line from all vessel containment boxes to the leakage containment ion exchangers. Describe leakage detection method to identify the source of leakage from each vessel containment box.
Describe also how you would distinguish accidental leakage from the leakages resulting from routine connections and disconnections of vessels.
RESPONSE
As indicated on P & ID's 527D-A-5001 and 5002, each leakage containment is provided with an isolation valve on the exit line.
In the event of a gross leak during system operation, the source may be isolated by selective valving and observation of the in-line monitor.
This monitor was provided to detect significant leaks during system operation.
It is provided with an automatic time delay isolation of the SDS feed. In view of the fact that submerged components will be flushed prior to disconnect and the large leakage containment dilution flow, it is anticipated that this moni-tor will be unaffected by vessel uncoupling.
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SfECIFIC COMMENTS QUESTION 25 Describe' leakage detection method provided within Hi-Rad Filter Glove Bar, Hi-Rad Feed Sample Box, Intermediate Level Sampling Glove Box, and Beta Nonitor Manifold.
-Describe (1) type of analysis to be performed on Hi-Rad samples, and (21 capability of such analyses on the TMI site.
RESPONSE
The Hi-Rad Filter, Hi-Rad Feed, Intermediate Level Sampling boxes, as well as the Beta Manifold, are contained shielded enclosures that are connected to drains and the SDS offgas system.
In view of the above, no leak detection system is provided. When the SDS is operating, the area vill be under constant surveillance and these enclosures may be visually inspected for leakage. In addition, a leak may also be indicated by a rise in the SDS offgas monitor. Routine radiological surveillance, in addition to the installed instrumentation, will provide early detection of minor leakage.
The types of analyses anticipated to be performed on Hi-Rad samples are as follows:
1)
Suspended solids
- 2) Gross Beta
- 3) Gamma spectral analyses 4)
Strontium 89 and 90
- 5) pH
- 6) Conductivity
- 7) Specific ions Sufficient chemical equipment will be installed to perform these studies within the shielded sample boxes. High level samples will also be diluted within the boxes prior to movement to chemical work areas for preparation of samples for counting.
Counting equipment has been purchased for dedicated use in support of SDS operation.
These include gas flow counters, end window G-M's, and a computerized Ge(Li) System.
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'5PECIFIC COMMENTS OUESTION 26 You state that the capability to solidify the spent resin vessels generated by the operation of the SDS has been developed. Describe the capability and process involved in detail.
RESPONSE
The attached drawing number 527-D S-1000, indicates the preliminary concepts for a 10ft3 vessel solidification facility.
Current plans are to install thir facility in the loading dock area of the Fuel Handling Building truck bay. The <.ubicle will be provided with a filtered exhaust system that will be ducted to existing station ventilation. Additional fIatures include closed circuit TV cameras, radiation monitors, and a motor driven cask cart.
A dewatered vessel will be loaded underwater in the CNSI l-13C cask and allowed to drain prior to movement to the solidification cubicle. When inside, the cask lid will be removed and the remote manipulator will be used to connect water, cement, and vent lines.
The proper amount of water and dry cement will be metered to the vessel and internal agitator blades will be rotated to assure proper mixing. Following remote disconnect, the cask lid will be installed and<the cask will be moved outside the cubicle for final preparation.
Pending further guidance and criteria development for the ultimate disposal of spent ion exchange media by the NRC, we have discontinued further work on the engineering design details and development of the SDS solidification facility.
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SPECIFIC COMMENTS QUESTION 27 Provide an estimate of dissolved KR-85 concentration and include it as part of the gaseous source term.
RESPONSE
Assuming the containment Kr-85 airborne concentration is 1 pCi/ml, the resulting 2
water concentration is appror.12nately 5 X 10 pCi/ml, the SDS process rate is 10 gpm, and that all Kr-85 is released to the offgas system during processing, the 8
resulting Kr-85 release rate unuld be 1.82 X 10 pCi/ minute.
Including this contributor to the gaseous source term does not significantly alter the conclusions presented in Chapter 6 of the SDS TER. The maximally exposed individual would l
receive an additional exposure to the lungs of 8.54 X 10 " mrem per year and
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-2 9.96 X 10 mrem per year to the skin, the critical organs for this isotope.
Additional annual exposures to other organs, listed below, are insignificant.
t Bone:
6.5 X 10"" mrem Thyroid:
4.46 X 10 " mrem Total _ Body:
5.6 X 10 " mrem GI Tract: 4.42 X 10 " mrem
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SPECIFIC COMMENTS QUESTION 28 Wouldn't the compartment cleanup and monitoring system for the pool water indicate a leakage contamination problem long before area radiation monitors in the SDS area would detect such leakage? What level are these area monitor alarms set?
Provide type, number, location, and setpoint for all area radiativu monitors.
RESPONSE
It is agreed that the leakage containment system will observe in-leakage from the spent fuel pool to the containment boxes. However, in the event of a leak on the effluent line, the time response of this radiation monitor relative to other anee monitors is sensitive to the location of the assumed leak.
The area monitor associated with the SDS which would detect an increase in spent fuel activity are listed below:
INSTRUMENT NUMBER TITLE
_ VENDOR RANGE CN-RE-1X03 SDS Manifold General Eberline 1-10,000 mR/hr Area Radiation Element CN-RE-VA06
.01-100 mR/hr Influent Radiation Element 3
' CN-RE-LC05 SDS Leakage Containment LND-INC 10 - 10' CPS System Influent Radia-tion Element The current plan is to set the alarm setpoints slightly above normal operational levels.
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~ SPECIFIC COMMENTS QUESTION 29 State if the reactor building water pump and SDS feed pump are provided with automatic trip upon detection of a present high radiation level by in-line radiation monitor (RE 527-13), fuel pool area monitors, and/or radiation monitors in the filter containment boxes.
RESPONSE
A high radiation indication at the in-line radiation monitor will cause valve V527-21 to close. This valve isolates the feed to both ion exchanger trains.
When this occurs, the SDS feed pump 527-G-01 goes into a bypass mode. The reactor building water pump (WGP-1) is not tied into the in-line radiation monitor RE-52 7-13.
Fuel pool area monitors and filter containment box radiation monitors are alarmed only.
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