ML20150F885
| ML20150F885 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/30/1988 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | THREE MILE ISLAND ALERT |
| References | |
| CON-#288-6007 OLA, NUDOCS 8804060060 | |
| Download: ML20150F885 (21) | |
Text
-.w 00CMETED USNRC GT. ATED CORRESPO[tpg March 30,1988 -
'88 APR P7 :37
. UNITED STATES OF AMERICA
$'[Y " M' ~ l NUCLEAR REGULATORY COMMISSION
~ " ~.,
4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
GPU NUCLEAR CORPORATION
)
Docket No. 50-320-OLA
)
(01sposal of Accident-(Three Mile Island Nuclear
)
Generated Water)
Station, Unit 2)
)
t LICENSEE'S AN5WERS TO SVA/TMIA'S SECOND SET OF INTERROGATORIES TO GPU NUCLEAR Licensee GPU Nucle?r Corporation, pursuant to 10 CFR 2.740b, hereby submits the following responses to the SVA/TMIA interrogatories dated March 15, 1988.
The provision of answers to these interrogatories is not to be deemed a representation that Licensee considers the information sought to be' relevant to the issues in this proceeding.
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INTERROGATORY NO. 1 - In GPU Proposal, July 1986, you state "There is good j
confidence that the dispersion resulting from the 1985 data is similar to annual dispersion in recent years" (page 42, Section 4.2.2.1).
State the basis for this confidence.
State which year prior to 1985 provided the data which was used to determine dispersion prior to 1985.
Explain how this was different to the 1985 data.
ANSWER NO. 1 -
The actual annual dispersion based on the measured meteorological data from the onsite tower for the individual years of 1978 through 1985 were compared.
Of those eight years, the 1985 data indicated that the 1985 dispersion was actually slightly below average (which would yield a higher dose estimate).
i Dose calculations for actual plant releases are matched to the meteorology which existed at the time of the release.
Therefore each year is used in turn for the routine release calculations.
The method used for the water disposition calculations using the routine release calculation model and actual measured meteorology for a year provides i
an accurate, but still conservative estimate of tr.e actual dose which may be i
incurred.
INTERROGATORY NO.
2 In GPU Proposal, July 1986, you state that "the continuous tritium release rate to the atmosphere is limited to 570 pCi/sec 3
10-5 sec/M " (page 22, "Tritium").
Provide the j
based on an X/0 of 5.6 x mathematical calculations to provide the result of 570 C1/sec.
Does the X/0 value reflect the worse case short term dispersion characteristics?
Does the allowable release rate of tritium take into account the fact that other radionuclides are present in the effluent?
ANSWER NO. 2 -
As stated in the Proposal, the limit is derived from the Environmental Technical Specifications.
Tritium release rates Jre conservatively equated to noble gas release rates from Specification a. and Specification c., and the Bases on page 2-14 of the Environmental Technical Specifications.
l V
Specification-a.
states "The
.stantaneous release rate of gross gaseous activity except for halogens and particulates with half-lives longer than eight days shall not exceed:
01 1 1.5 x 1^+5 g3 sec"
/
(MPC1)
Specification c.
states "The release rate of gross gaseous activity shall not exceed:
01 1 2.4 x 10+4 3
M /sec (MPC1) when averaged over any calendar quarter."
The ratio of these release rates for noble gases is:
Mbsec=0.16or16 percent 2.4 x 10 1.5 x 10+5 g j3,c 3
i Thus, specifications a. and c. above, establish.an upper limit for the average release rate of gaseous activity at 16 percent of the instantaneous release limit.
4 A release rate limit was derived by using this noble gas limit applied to tritium, along with the annual average ground release level X/Q rrom the Offsite Dose Calculation Manual.
The MPC value to environs from tritium 10-7 uC1/ml of listed in 10CFR20, Appendix B, Table II, Column 1,
of 2 x j
air, was also used.
.i l
4 The resultant calculation for-the~ limiting ground release rate for tritium at the site boundary at the SSE. sector would be:
0.16 x 2 x 10- uC1/ml - 570 uC1/sec.
I 5.6 x 10-5 sec x M 3
6 M
10 cc Since the the value indicated is a continuous release rate
- limit, an-instantaneous worst case dispersion parameter is inappropriate and is not used for this case.
The continuous allowable release rate for tritium does not take other radionuclides into account.
Other radionuclides are particulates and are limited by release rates applicable to particulates.
INTERROGATORY NO. 3 - By what criteria was it decided to use a 100 foot stack to release the vapor into the atmosphere?
ANSWER NO. 3 -
The 100 foot height dimension, selected for the exhaust stack, was based on the height of the Auxiliary and Fuel Handling Building (AFHB) to which it will be attached and over which it must project.
The top of the AFHB is located at Elevation 387'-2", which is 82'-8" from grade located at elevation 304'-6".
The 100 foot stack will project over the AFHB approximately 17'-4",
the difference between 100' and 82'-8",
INTERROGATORY NO. 4 - List the uses and the location wherein the AGW will be used prior to the Unit 2 being placed in Post Defueling Monitored Storage.
ANSWER NO. 4 -
Some AGW will be used for various system flushes and decontamination purposes prior to PDMS. Current plans call for the following uses of AGW:
l 1
4 Fuel Cask Decontamination Reactor Building Basement Block Wall Flush AXOl6 (Cleanup Demineralizer) Cubicle Decontamination AX017 (Cleanup Demineralizer) Cubicle Decontamination AX114 (Makeup & Purification Demineralizer) Cubicle Decontamination AX115 (Makeup & Purification Demineralizer) Cubicle Decontamination AX117 (Makeup Filter) Cubicle Decontamination Cleanup Demineralizer Piping Flush Makeup & Purification Demineralizer System Piping Flush Liquid Radwaste System Flush Spent Fuel Pool Flush at Orain Down INTERROGATORY NO. 5 - Is hydrogen peroxide being added to the AGW in the RCS presently?
When was the last addition of hydrogen peroxide made?
Are the microorganisms in the RCS continuing to grow?
When was the first addition of a) Betz 1182 b) Betz 1192 c) Calgon 289 to aid in DWCS filter efficiency.
Has there been water sample analysis undertaken since these coaguients were added to determine whether or not their presence affects the ion exchange efficiency of the Epicor/SDS system.
State the results of these water sanple analysis and the dates when the analysis was undertaken.
ANSWER NO. 5 -
Hydrogen peroxide is added to the AGW presently located in the RCS.
The most I
recent addition was on March 22, 1988.
The hydrogen peroxide substantially 1
depletes the microorganisms.
Nevertheless, the microorganisms do regenerate to some extent.
The initial uses of coagulants in plant systems to improve filtering efficiency occurred as follows:
I a)
Betz 1182 8-87 (RCS) i i
i 1
e b)
Betz 1192 8-86 (B RCBT Test) c)
Calgon 289 19-87 (SFP-A)
Tests to determine the impact of coagulants on the efficiency of EPICOR/SDS system performance were performed before actual use of the product.
Since no negative or unsuspected chemistry problems were observed following use of each product, further testing was not warranted.
INTERROGATORY NO. 6 - Response to Interrog& tory S52 IC - page 37:
State the amount of nitric acid added.
State the amount of phosphoric acid added.
What is Triton-X-100? Describe its contcots and its purpose.
ANSWER NO. 6 -
Nitric acid - 6.7 gallons.
Phosphoric acid - 2.8 gallons.
Triton-X-100 is a
non-ionic surfactant, with the chemical name Isooctylphenoxypoly thoxyethanol ethylene oxide.
It is used as a degreaser and a decontamination agent for floors, walls, piping and mechanical equipment.
INTERROGATORY NO.
7 - List the storage capacity of each tank or storage location listed in Table 2-3, GPU Proposal, July 1986, page 12.
In Table 2-7, page l 'i, it is noted that SOS-T-l A (SDS Monitor) contents do not need to be reprocessed through EPICOR/SDS again prior to evaporation.
However, it is noted in Letter 4410-87-M-0349, June 22, 1987, tnat on June 18th, 505-T-1A's contents were reprocessed.
Please clarify.
ANSWER NO. 7 -
Tank Storage Capacity is as follows:
4-OPERAT20NAL DESfGNED CONTENTS (Gal.)
STORAGE DESCRIPTION CAPACITY CAPACITY (as of 3/25/88)
(Gal.)
(Gal.)
RCS Reactor Coolant System N/A N/A 66,624 PWST-1 Processed Water Storage 495,684 500,000 221,681 PWST-2 Processed Water Storage 495,684 500,000 442,291 CO-T-1A Condensate storage 240,500 250,000 200,764 WDL-T-9A Evap. Cond. Test Tank 9,834 10,602 9,834 WDL-T-98 Evap. Cond. Test Tank 9,834 10,602 5,994 CC-T-1 EPICOR II Off-Spec 85,208 85,978 3,586 CC-T-2 EPICOR II Clean 122.061 133,689 4,405 SFP-B Spent Fuel Pool 8 241,698 N/A 241,698 SDS-T-1A SDS Monitor 11,503 12,000 435 SDS-T-1B SDS Monitor 11,503 12,000 373 WDL-T-1A RC Bleed Holdup 74,850 83.400 39,066 WDL-T-1B RC Bleed Holdup 74,850 83,400 51,092 WDL-T-IC RC Bleed Holdup 74,850 83,400 66,730 BHST Borated Water Storage 459.589 459,589 429,544 WDL-T-8A Neutralizer 8,605 8,780 4,565 WDL-T-8B Neutralizer 8,605 8,780 1,930 WDL-T-2 Hiscellaneous Waste Holdup 19,800 20,000 2,312 WDL-T-11A Contaminated Drains 2,560 2,660 1,660 WDL-T-11B Contaminated Drains 2.560 2,660 880 Chem Cleaning Bldg. Sump 3,680 4.580 1,380 l
Auxillary Building Sump 9,01S 11.071 5,917 i
Reactor Building Basement N/A N/A 38,315 SFP-A Spent Fuel Pool A 205,234 N/A 205,234 Deep End of Transfer Canal 58,685 N/A 58,685 i
I 1
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Both SDS Monitor Tanks (SDS-T-IA and 18) are designed to provide a location for the collection and monitoring of water that was processed through the SDS. Af ter sampling of this water, it is transferred to either the EPICOR II system for polishing, the SOS reactor coolant system manifold for reprocessing through S05, or the SOS flush header.
Processed water may be transferred from the monitor tanks directly to the processed water storage and recycle system (PHST's).
Several volumes of water pass through the SOS monitor tanks at any given time.
The information presented in the July 1986 report is just an isolation of one particular moment in time.
The determination of which sources of water will require reprocessing, reported in July 1986, was based on a comparison of the Sr-90 and Cs-137 concentrations of the various waters to the Sr-90 and Cs-137 concentrations attainable by SOS processing followed by EPICOR II polishing (i.e.,
activity of water in the PHST's).
Those water sources that had concentrations significantly above this were designated as requiring reprocessing prior to evaporation. At the time Tables 2-3 and 2-7 of the July 1986 Report were compiled, the monitor tanks contained water that, based on the above
- criteria, would not require further processing prior to evaporation.
When actual continuous cycle evaporator operations take place, the water in the SDS monitor tanks at that time will go directly to the staging tank that feeds the evaporator if the concentrations of this water meet the influent criteria.
If the water in the SOS monitor tanks does not meet criteria, it will be reprocessed.
INTERROGATORY NO.
8-Is the AGH from one location mixed with the AGH from another location prior to processing through Epicor/SDS?
Explain the following in Letter 4410-87-H-0349, June 22, 1987, "Epicor II Batch Reports,"
including:
a)
F-N/A b)
K-27 4
c) 2K-18 d) sk e)
EOS f)
Including Island 011ution If the 'AGW is mixed from various locations prior to processing through Epicor II, are samples drawn from each location prior to mixing?
ANSWER NO. S -
- Yes, in some cases, in order to stage a complete batch of water for processing, more than one source will be used.
Explanations for the following from memorandum 4410-87-M-0349, June 22, 1987:
a)
F-N/A:
The EPICOR processing system consists of 3 positions (designated F,K, 2K) available for ion exchange material.
In this case, the "F"
position was not being used, indicated by "N/A" or not applicable.
b)
K-27:
This is the second or "K"
position of the EPICOR system and the
'27' refers to the twenty seventh consecutive demineralizer which was used in this position.
c) 2K-18:
This is the third or "2K" position of the EPICOR system and the
'18' refers to the eighteenth consecutive demineralizer which was used in this position.
d) sk:
This is not
'sk' but "5K",
a shorthand representation for 5,000 gallons of water processed.
e)
E08:
This is an abbreviation for "End of Batch" or completion of processing.
f)
Including Island Dilution:
This a conservatively calculated number showing what the concentration would be if discharged to the Susquehanna River with a minimum dilution factor of 3,600.
Intermediate AGW storage locations are not always sampled and analyzed when transferring water.
The control is maintained at the last storage location 1 -_.
e before processing so that the efficiency of the process can be monitored and evaluated.
INTERROGATORY NO. 9 - In the matter of sampling of water and analysis of
- sample, a) who undertakes calibration of the analytical procedures?
b) are split samples taken to confirm accuracy?
c) if answer to abcive is "yes", state who undertakes this confirmation and how often this is undertaken.
ANSWER NO. 9 -
a)
Analytical procedures used by the TMI-2 Chemistry Group are calibrated by the Group's chemistry technicians.
b)
Split samples as well as blind samples are analyzed on a set frequency to validate the analytical results.
c)
The Quality Control program has various modes for confirmation of analyses.
Confirmation is undertaken by the TMI-2 Chemistry technicians and supervisors, and two outside vendors.
The frequencies range from daily to semi-annually.
INJERROGATORY NO. 10 - How accurately does Table 2-3 reflect the contents of each of the storage locations presently?
(GPU Proposal July 1986)
In responding, please address each storage location separately and show the radiological and non-radiological contents of each location.
ANSWER NO. 10 -
Table 2-3 from the GPU Proposal July 1986 accurately reflected the contents of each source location as of the sample dates noted on the table for each source.
This table, when prepared, provided a snap-shot in time of the AGW inventory and was not presented or intended as a limit imposed on these contents by source location.
The volumes on March 25, 1988, are listed above In response to Interrogatory No. 7.
O The radiological and non-radiological content of each of these locations changes continuously during defueling and decontamination activities.
Although the radionuclide concentrations presented in Table I-? accurately depict the total radioactivity in the AGH, the specific source location for this inventory at a particular point in time is dependent on plant operations and does not alter the actual total source term.
INTERROGATORY NO. 11 Is the cresence and quantities of reactor corrosion activation products
- 65z, 58,60 o, 55,59 e and 54Mn-typical of water C
F produced la water at a normal operating nuclear power plant?
ANSWER NO. I1 -
Activated reactor corrosion products are oroduced as the materials of construction (such as 304 stainless steel, inconel 600, and carbon steel) undergo some oxidation on their surfaces and the metal oxides are carried into 1
tne core region and become activated by the high neutron flux of an operating reactor.
Below is a partial listing of the subject radionuclides for comparison of THI-2 with other operating nuclear power plants.
TMI-2 TYPICAL 58
<4.0 E-3 uC1/mi 1.7 E-3 uCi/mi Co 60Co 4.8 E-7 uC1/mi 1.3 E-4 uCi/mi 55 4.8 E-7 uC1/ml 1.0 E-6 uCi/ml Fe 59 Fe
<3.0 E-7 uC1/ml 6.0 E-5 uCi/mi 54Mn (4.0 E-8 uCi/ml 9.1 E-5 eC1/ml 65 Zn
<9.8 E-8 uCi/ml
<3.5 E-5 uC)/mi The quantities of activated corrosion products in THI-2 water are less than those of a typical operating nuclear power plant. -
4 INTERROGATORY NO. 12 - In GPU Proposal, Page 7, you state "Waste from the concentrated waste storage tank will not be considered for evaluation of disposition options since this material will be solidified directly for disposal as radioactive waste." Explain why this is 50.
ANSWER NO. 12 -
The waste from the Concentrated Haste Storage Tank (CHST) was excluded from the assessment of the disposition options because this waste is primarily chemical in nature and is treated independently of the disposition options addressing Accident Generated Water (AGH).
This waste, of varied chemical constituents, is collected from plant and Waste Handling and Packaging Facility (HHPF) chemical decontamination efforts and is staged in the CHST.
It is then periodically transferred to a waste disposal container and solidified with Portland cement for disposal at a commercial burial facility.
INTERROGATORY NO. 13 - What is the volume of AGH presently?
ANSWER NO. 13 -
Inventory of the ACH on March 25, 1988 was slightly greater than 2,100,000 gallons.
INTERROGATORY NO.
14 - State the basis and provide references for your statement on page 15, "Answers to SVA/ THIA's Interrogatories," "under the chemistry conditicns of this water, the carbon and lodine will be in chemical forms that are not volatile at these operating conditions."
ANSWER NO. 14 -
In the case of Iodine-129, the analyzed activity is less than 6 x 10-4 uC1/ml.
This calculates to less than 3.5 x 10 grams of Iodine per liter of solution.
In this very dilute condition in water such as AGW, the Iodine will be in chemical forms and ionized statar that greatly favor staying in the water phase.
Even at elevated temperatures (100 C), significant amounts of iodine must be present before the volatile lodine species would exist above the solubility concentrations at that temperature.
All of the very small concentrations, if any, of fodine-129 will greatly favor staying in solution i
~
in the evaporator bottoms rather than volatilizing or being carried over in I
i the distillate.
1 The large majority of tne carbon 14 identified in Table 3.2 of the System Description will be in various chemical forms other than carbon dioxide gas.
(The carbon 14 content of carbon dioxide dissolved in AGW is not expected to differ from the carbon 14 content of carbon dioxide in the atmosphere.
Th i t,
?
Is due to exchange of air across the water / air interface.)
The carbon not involved with a gas will remain in the evaporator bottoms as the distillation process takes place.
Evidence of aetailed measurements for PHR's indicates that virtually none of the carbon 14 released was associated with C0 '
2 INTERROGATORY NO. 15 - Will the operators of the evaporator receive training in the Licensee's responsibilities to keep releases of radioactivity as low as reasonably achievabl@,
If the answer is "yes," explain their training.
ANSWER NO 15 -
F Yes.
The operators of the evaporator will receive such training at the TMI l
Training Center prior to their site access.
The training will consist of the
[
a
]
following, among other things:
i i
RADIATION WORKER TRAINING i
Regulations i
I Biological Effects of Ionizing Radiation i
Reporting Radiological Deficiencies Location of Radiation Sources and Radioactive Materials o
Minimizing Radiation Exposure j
?
The Purpose and Function of Protective Devices i
I j
Responding to Radiological Postings, Warnings, Alarms, etc.
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,5 They will be trained on the TMI Radiation Protection Plan, the purpose of which is to control radiological conditions, to avoid accidental radiation exposures, to maintain exposures within regulatory requirements, and also to maintain exposures to workers and the general population as -low as is reasonably achievable.
INTERROGATORY NO. 16 - Has Licon Inc. installed evaporators at other nuclear power plants where the system is operated such that the vapor is released into the atmosphere as is planned at Unit 2?
If the answer is "yes,"
state at which plant the evaporator operates.
4 Describe all tests and provide research which clearly demonstrate that a system similar to the one to be used at TMI is a 'well proven technology."
(AGW Disposal System Description, 4410-88-L-0012/0335P, page 4).
Those tests l
and/or research mentioned must refer to a system operating so that the goal is to release the condensate to the environment.
ANSWER NO. 16 -
4 No ~.
The use of the term "Well Proven Technology" in thn AGW Disposal System Description made reference to the use of evaporators in nuclear power plants 4
throughout the United States, including TMI.
Whether the condensate from l
l these evaporators is released into the atmosphere or recycled for plant use is not pertinent to the fact that the use of evaporators to remove liquid from a solution or slurry by vaporization of the liquid is "Well Proven Technology."
k' INTERROGATORY NO. 17 - What is the expected date for completion of the final
)
design and fabrication of the evaporation-system?
j ANSWER NO. 17 -
l i
GPU Nuclear does not have a firm date for completion of the final design and fabrication of the evaporator system, but expects that comoletion will occur l
1 in the fourth quarter of 1988.
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INTERROGATORY NO. 18 - In the Systems Description, page 21, you state that "By using the evaporator as a pretreatment technique for certain of the volumes of water, pretreetment by one of the ion exchange systems and the resulting contamination of demineralizer resins would be eliminated."
Identify those "certain of the volumes of water."
In particular, state the present location of those volumes of water, quantities and radiological and non-radiological content.
Furthermore, clarify the relevancy of the "base case" in the EIS, Supplement #2, Table 2.2 should this process known as "Satch Cycle Operations" be adapted.
If the evaporator is used instead of the ion exchanger system, state how much waste would be developed for thes1 certain volumes of water, and what would be the radiological and non-radiological content of this waste?
ANSWER NO. 18 -
All AGW will be processed through the evaporator prior to release to the environment via vaporization.
The System Description, page 21, refers to operation of the evaporator decoupled from the vaporizer.
In this configuration, described as "Batch Cycle Operations," there is no atmospheric release.
Rather, the evaporator distillate is collected in a staging tank.
The AGW is recycled through the evaporator, as necessary, to achieve the desired effluent water quality.
- Then, the distillate is routed to the vaporizer for release to the environment.
Operated in this mode the evaporator may replace demineralization as a water pre-treatment method and provide a more effective means of particulate radionuclide removal.
The candidate AGW sources for this process will be determined by operational considerations, including efficiency of operations, and current influent criteria for continuous cycle operations as determined by the demonstrated carryover fraction of the evaporator during actual system operations.
The "Base Case" from Table 2.2 of the PEIS Supplement 2 refers to approximately 407. of the AGW inventory, which was the estimated quantity of AGW requiring pre-processing prior to open cycle evaporation (evaporator coupled with the vaporizer and operated in a continuous cycle) based on projected evaporation performance (i.e., carryover).
It is this volume of water, currently in use for cleanup activities, primarily-in the Reactor Coolant System (RCS),
Pools (A &.8), Transfer Canal and building sumps, that would be evaluated possible pretreatment with batch cycle evaporator-operations o
demineralization.
The volume of waste generated as a result of pre-processing this 40% inventory with batch cycle evaporation would be approximately the same volume as that generated from continuous cycle evaporation. See System Description, 2.3.7.
In the unlikely evert that all of this 40% inventory would be pre-processed by batch cycle evaporation, the non-radiological content of the was te form, approximately 62.5 tons of waste, would be composed of boric acid and sodium hydroxide.
- However, this volume of waste would be present regardless of SDS/EPICOR pre-processing due to sodium addition requirements for pH adjustment prior to evaporator operations and the fact that most boron is not removed by the SDS/EPICOR processing systems.
Based on the radiological contents of the waste form resulting from pre-processing the 40%
of AGW inventory by batch cycle evaporation, approximately 10% of the waste would be classified as Class A, the lowest level of waste form classification.
Up to 90% of the waste might be Class B, the second level of waste form classification.
INTERROGATORY NO. 19 - Who will provide confirmation of the sampling analysis of the product distillates?
ANSWER NO. 19 -
Confirmation of the sampling analysis of the product distillates will l
automatically be included under the Quality Control program already in place for the THI-2 Chemistry Laboratory group.
y__.
INTERROGATORY NO. 20 - Will each of the tanks now holding the AGW have to be moved to the pipe run located in the Unit 1/ Unit 2 corridor?
If so, explain how this is to be done.
ANSWER NO. 20 -
No.
g INTERROGATORY NO. 21 - How will you prevent water droplets from going up the exhaust stack?
ANSWER NO. 21 -
0 The combination of the 240 F temperature of the exhaust steam and the presence of an impingement screen in the exhaust stack will prevent the formation of water droplets during normal operations.
INTERROGATORY NO. 22(1st) - In response to Interrogatory 62 you stated that "The average decontamination factor is 1,000.
This is based upon an average system carry-over fraction of 0.1%."
What tests and research demonstrate this to be so when an evaporator is operated in an open cycle.
Provide examples of an evaporator operating in a mode identical to that to be used at Unit 2.
ANSWER NO. 22(1st) -
The average carryover fraction for this disposal system. 0.1%, is based on routine performance experience with typical evaporator systems, and not upon tests or research.
The evaporator portion of the Disposal System operates in the closed cyc'e mode.
The fact that distillate is thereaf ter vaportzed does not affect the performance (i.e.,
carryover fraction) of the evaporator itself.
No attempt has been made to identify an evaporator operating in a mode "identical" to the AGW Disposal System.
l I
INTERROGATO.RY NO.
22(2nd)
- What is the radionuclide content of the i
micro-organisms which have grown in the reactor core?
ANSWER NO. 22(2nd)
The radionuclide content of the microorganisms detected in the reactor coolant I
water has not been quantified.
Microorganisms make use of many different 1
chemical elements to carry on cell functions.
In this way, microorganisms may take in or excrete radionuclides along with non-radioactive elements.
The radioactive material, including that retained by the microorganisms, will still be detected in the routine analysis process.
INTERROGATORY NO. 23 - Is the Licensee aware of any technology which can remove the tritium from water?
If the answer is "yes," briefly describe that technology.
ANSWER NO. 23 -
GPU Nuclear is not aware of any commercial industrial-scale process for separating tritium from the accident generated water at THI-2.
t 1
UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION i
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
GPU NUCLEAR CORPORATION
)
Docket No. 50-320-OLA
)
(01sposal of Accident-(Three Mile Island Nuclear
)
Generated Water)
Station, Unit 2)
AFFIDAVIT OF F. R. STANDERFER County of Dauphin
)
)
Commonwealth of Pennsylvania
)
F. R. Standerfer, being duly sworn according to law, deposes and says that he is Director of TMI-2 and Vice President, GPU Nuclear Corporation; that Licensee's Answers to SVA/TMIA's Second Set of Interrogatories to GPU Nuclear are true and correct to the best of his information, knowledge, and belief; and that the sources of his information are the officers, employees, agents, and contractors of GPU Nuclear Corporation.
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/F. R. S tYn'd e r f e r
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Sworn to afd subscribed before me this t.go P day of March, 1988.
A f) b Y2/*R Notary Public
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My Commission expires esswaan u. gesen. smart w eBMtfWBIBOIL655455ConsW De eesnestett (WWS WAIN M.15m em e,emenman musslauen et asemim
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i April 1, 1988
'88 APR -4 P7 :37 j
fghk'7kr,yyh^y' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION bra ncy BEFORE THE ATOMIC SAFETY AND LICENSING BOARD I
In the Matter of
)
)
GPU NUCLEAR CORPORATION
)
Docket No. 50-320-OLA
)
(Disposal of Accident-(Three Mile Island Nuclear
)
Generated Water)
Station, Unit 2)
)
CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing "Licensee's Answers to SVA/TMIA's Second Set of Interrogatories to GPU i
Nuclear" were served this 1st day of April, 1988, by U.S. mail, first class, postage prepaid, upon the parties identified on the j
attached Service List.
Thomas A.
- Baxter, P.C.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
GPU NUCLEAR CORPORATION
)
Docket No. 50-320-OLA
)
(Disposal of Accident-(Three Mile Island Nuclear
)
Generated Water)
Station, Unit 2)
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J SERVICE LIST i
J Sheldon J. Wolfe, Esquire Richard P. Mather, Esquire Atomic Safety and Licensing Department of Environmental Board Panel Resources U.S. Nuclear Regulatory Commonwealth of Pennsylvania Commission 505 Executive House Washington, D.C.
20555 Harrisburo, Pennsylvania 17120 4
Mr. Glenn O.
Bright Ms. Frances Skolnick Atomic Safety and Licensing 2079 New Danville Pike Board Panel Lancaster, Pennsylvania 17603 U.S. Nuclear Regulatory Commission Ms. Vera L. Stuchinski Washington, D.C.
20555 315 Peffer Street Harrisburg, Pennsylvania 17102 Dr. Oscar H. Paris Atomic Safety and Licensing Dr. William D. Travers Board Panel Director, Three Mile Island U.S. Nuclear Regulatory Cleanup Project Directorate Commission P.O. Box 311 l
Washington, D.C.
20555 Middletown, Pennsylvania 17057 Stephen H.
Lewis, Esquire Colleen P. Woodhead, Esquire Office of tne General Counsel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Docketing and Services Branch Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555 I
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