ML20236U835
| ML20236U835 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/25/1987 |
| From: | Standerfer F GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 4410-87-L-0170, 4410-87-L-170, NUDOCS 8712030404 | |
| Download: ML20236U835 (8) | |
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GPU Nuclear Corporation 3 Nuclear
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- 8o Middletown. Pennsylvania 17057-0191 717 944-7621 TELEX 84-2386 Writer's Direct Dial Number
(717) 948-8461 November 25, 1987 4410-87-L-0170/0270P US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Dear Sirs:
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 Post-Defueling Monitored Storage Environmental Evaluation Comment Responses NRC TMICPD Letter NRC/TMI-87-069, dated September 2,1987, provided 18 additional NRC comments on the Environmental Evaluation for Post-Defueling Monitored Storage submitted by GPU Nuclear letter 4410-87-L-0025, dated March 11, 1987.
GPU Nuclear letter 4410-87-L-0155, dated November 5,1987, submitted responses to NRC Comments 1, 3, 4, 8, 9,14,17, and 10 of the above referenced TMICPD letter. Attached are responses to comments 2, 5, 6, 11, 12, and 13.
Responses to the remaining comments are being prepared and will be forwarded under separate cover.
Sincerely,
. R. Standerfe Director, TMI-2 RDW/eml Attachments cc: Regional Administrator, Region 1 - W. T. Russell Director, TMI-2 Cleanup Project Directorate - Dr. W. D. Travers 9712030404B3j
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20 DR ADOCK O PDR I
GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation i
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ATTACHMENT 1 4410-87-L-0170 NRC COMMENT 2 Please clarify the location of groundwater entry, location of collection, maximum anticipated quantity and probability that contamination might be carried out of the building (s) by the same' path. The report also refers to collected precipitation.
What radionuclides concentrations are typical in this water?
GPU NUCLEAR RESPONSE GPU Nuclear letter 4410-87-L-0094, dated June 23, 1987, stated water inleakage currently occurs in the following areas of the plant and is collected as indicated:
o Fire Service Penetration; east wall of Turbine Building at the 300' elevation. Orainage is to the Turbine Building Sump, Water Treatment Sump, or the Condensate Regeneration Polisher Sump.
o Building Joint; between the Service Building and Air Intake Tunnel. This area does not have sump drainage.
It is pumped periodically, as necessary, to remove inleakage, o
Construction Joint; basement of the Auxiliary Building. Drainage is to the Auxiliary Building Sump.
o Electrical Penetration; southwest corner of the Control Building area at the 281' elevation. Drainage is to the Control Building Area Sump.
o An additional collection point for groundwater inleakage is the Tendon Access Gallery Sump at elevation 261'8".
Depending on the magnitude of the inleakage and the location of drains, water will initially collect in the respective areas and eventually collects in the respective sumps.
As stated in Section 3.3.2, " Liquid Releases," of the PDMS Environmental Evaluation the expected annual inleakage is approximately 5000 gallons.
Additionally, as indicated in the previously submitted response to NRC Comment 3, the radionuclides concentration of collected precipitation, based on 1987 records, is highest at the Tendon Access Gallery Sump where the range of radioactivity found is:
Gross Beta = 5.6 E-6 to 9.0 E-6 uCi/cc l
Cesium 137 = 5.2 E-6 to 8.1 E-6 uC1/cc Tritium
= 3.6 E-6 to 4.8 E-6 uCi/cc 1
Though not quantitatively determined, the probability that contamination might be carried out of the referenced buildings is unlikely based on the followina i
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ATTACHNENT 1 4410-87-L-0170
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o Inleakage would occur at times of high groundwater table elevations, when 1
J the table is sufficiently higher than the basement floor and provides an inward driving force. Most of the water would be retrieved by the floor drain / sump system making little, if any, water available for re-entry into the groundwater.
Water would not be expected to flow in both directions from the same location, j
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Groundwater elevations in the 13 measured stations in the immediate l
vicinity of Unit 2 (within or near the Unit 2 Protected Area Fence) do not normally fall below the 280' MSL elevation. This is three to four 1
feet above the bottom of the floor slab. Hence, the driving force is j
into the plant buildings.
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o An extensive groundwater monitoring program has been in effect since l
1980. No evidence of out-leakage of radioactive materials from plant i
buildings has occurred in those seven years, even though the extensive l
wet decontamination procedures used during that period provided far more than normal opportunity for out-leakage.
NRC COMMENT 5 The report has " conservatively" estimated that a maximum of 3% of the radionuclides in the concrete block might migrate to the outside. Please provide the basis or give your plans for a monitoring and mitigation program that will assure that the spreadable contamination does not exceed the estimate.
GPU NUCLEAR RESPONSE As noted in Page 10 of the PDMS Environmental Evaluation (reference GPU Nuclear letter 4410-87-L-0025, dated March 11, 1987), the 3% fraction was an assumed value which results in a total of about 1000 Ci of "suspendable" contamination in the vicinity of the block wall. For the purpose of this analysis, it was considered the largest single potential source for airborne concentration.
The 3% value was chosen to be unquestionably conservative. Dr. Hershal Godbee, from Oak Ridge National Laboratory and Chairman, American Nuclear Society Standards Committee on Concrete Technology (Working Group ANS-16-1),
believes that much less than 1% of the total activity adsorbed into the concrete will reappear on the exterior surface during extended storage. He estimated, for THI-2 concrete, that about 0.2% may return to the surface but recommended that a 1% upper limit be used to provide a conservative margin of safety.
In addition, the ratio of Cs to Sr was conservatively assumed to be 1:1.
This ratio was picked based on sump sediment samples evaluated against limited concrete samples. Based on gamma scans of more recent concrete samples, the
ATTACHMENT 1 4410-87-L-0170 Cs-137 to Sr-90 ratio averages 23.5:1.
This ratio gives a much lower total Sr-90 inventory than assumed in the source terms for off-site dose calculations. Taking into account these two (2) factors (i.e., 1% migration and the change in Cs-137 to Sr-90 ratio), the calculated off-site dose for the assumed fire would be lower. Thus, the analysis in the PDMS Environmental Evaluation is conservative and further confirms the conclusion that cleanup operations have progressed to a point where any threat to public health and safety has been eliminated. Although no revision of the analysis in the Environmental Evaluation, is being undertaken now, future analysis in support of a License change will take into account the latest available information.
1 NRC COMMENT 6 Your submittal states, "...the most significant radionuclides contribution to the off-site dose impact results from the transuranic listed on Table 2 and Sr-90 and will bound consideration of these other nuclides." What would the relative contribution from Cs-137 be.
GPU NUCLEAR RESPONSE The statement referred to in Comment 6 above was made to limit the range of radionuclides required to be considered for off-site dose impact. The dose commitment postulated in Table 4 of the Environmental Evaluation used the source terms summarized in Table 3 which included Cs-137. The relative contribution of each nuclide in the source term was not specifically determined. One techinque to estimate the upper limit of Cs-137 contribution is by assessing the calculated dose assessment for the liver which is the primary organ for dose commitment from Cs-137.
The routine releases via airborne pathways resulted in a liver dose from all isotope of about 10% of the total bone dose. Because Cs-137 bone dose is in the order of Cs-137 liver dose, the Cs-137 bone dose will be no more than 10%
of the total bone dose.
NRC COMMENT 11 An accidental fire might increase the airborne concentration of tritium which could pass uneffected through the HEPA filters. Please estimate (and provide the basis for) the quantity of this radionuclides that might be released by such an accident.
GPU NUCLEAR RESPONSE In the PDMS condition, the vessel and fluid systems will be drained, eliminating open water surfaces as an airborne source. There are currently no enclosed, non-ventilated areas where tritium could be accumulating.
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ATTACHENT 1 4410-87-L-0170 The amount of tritium that could be released from a fire would not pose any significant risk to the health and safety of the public.
In fact, combining the worst case assumption for the source (i.e., the amount liberated in a fire) and the accident meteorology, the resultant off-site exposure would be less than three (3) mrem to the maximally exposed individual via the inhalation pathway. Use of expected values for any one of the above factors would reduce the dose to the maximally exposed individual to less than one (1) mrem. An upper bound on the amount of tritium that could exist and be liberated in the Reactor Building can be estimated as follows:
After the TMI-2 accident, a large amount of water (i.e., 640,000 gallons, l
Reference 1) accumulated in the Reactor Building basement to a maximum level of 8.5 feet. This inventory contained a small amount of tritium, (i.e.,
.88 uCi/ml [ Reference 2]). Some portion of this tritium could have been absorbed into the concrete surface of the basement floor and structures.
Conservatively assuming that concrete could absorb water up to 10% of its weight, to a depth of one (1) foot, would result in 109 curies in the concrete (i.e., floor, D-rings, block wall (which is only 8" deep)), of the basement.
Assuming 100% of these curies are released in a fire and worst case accident meteorology, the resultant off-site exposure would be less than three (3) mrem to the maximally exposed individual.
REFERENCES:
- 1. GPU Nuclear Report, " Reactor Building Radiological Characterization," Volunie II, TP0/TMI-125, Revision 1, June 1987
- 2. "The Three Nile Island Accident Diagnosis and Prognosis,"
Edited by L. M. Toth, et. al., 293 ACS Symposium Series, May 1985 NRC COMMENT 12 Will operation of the ventilation system result in an exit velocity for the plume? If so, what velocity is expected?
GPU NUCLEAR RESPONSE The exit velocities (in feet / minute) for the ventilation systems for the Reactor Building (RB) Purge, Auxiliary Building (AB), Fuel Handling Building (FHB), and the RB Passive Breather are listed below. The RB Purge may be operated in combination with the AB and/or FHB Ventilation Systems in which case the exit velocity can be approximated by the summation of the respective system velocities.
The exit velocity reported for the RB Purge is based on maximum fan flow rate of 25,000 cfm. However, the flowrate for the RB Purge I
can be adjusted. The normal minimum flowrate is 10,000 cfm which would correspond to an exit velocity of 141 feet / minute, j
ATTACHENT 1 4410-87-L-0170 Normal Mode Maximum System (One Train)
(Two Trains)
RB 353 705 AB 459 918 FHB 254 508 Passive 0-2 0-2 Breather g
NRC COMMENT 13 What is the " Limiting Organ" shown in Table 5? Is it bone (as given in Table 4)? What are the units? What is the total inventory released on which E=
these values are based?
GPU NUCLEAR RESPONSES In response to the above comment, a revised Table 5 is included as Attachment 2 chowing the critical organs for the Appendix I group'.ngs. The revised table I
indicates that the limiting (i.e., critical) organ varies depending on the year and type of release. This Ic.ble also contains revised dose numbers which correct typographical errors and deletes the contribution of TMI-l to off-site exposures. The units in the revised table are expressed in millirem.
The total inventory that provides the basis for the value in revised Table 5 is included as Attachment 3.
This inventory was obtained from the Annual Environmental Monitoring Report for TMI for the respective years.
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ATTACHMENT 3 4410-87-L-0170 TMI-2 EFFLUENT RELEASES BY RADIONUCLIDES Total Release Per Year (C1)
Radionuclides 1982 1983 1984 1985 Liquid Cs-134 6.98 E-6 Cs-137 4.05 E-5 6.0 E-5 2.77 E-4 4.19 E-5 H-3 7.20 E-2 1.56 E-4 2.0 E-3 Sr-90 3.62 E-4 1.35 E-4 Tc-99m*
2.67 E-3 5.45 E-6 Gross Beta / Gamma 3.0 E-5 i
Gaseous l
Cs-134 3.94 E-6 1.37 E-8 Cs-137 5.39 E-5 4.0 E-7 4.55 E-6 2.56 E-5 00-60 3.40 E-8 14 3 112 48.7 14.3 19.8 Kr-35 914 191 246 Sr-90 9.22 E-8 1.64 E-7 2.02 E-5 Xe-133 1.33 E-1 Gross Alpha 1.11 E-7 8.0 E-7 4.3 E-7
- Medical Administration
- Kr-85 is not reported for 1985 as all values were less than the Lower Limit j
of Detectability (LLD). Prior to 1985, Kr-85 was assumed to be present at j
LLD.
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ATTACHMENT 2 4410-87-L-0170 TABLE 5 l
MAXIMUM POTENTIAL DOSES (mrem) CALCULATED FROM EFFLUENT MEASUREMENTS
- 1982 1983 1984 1985 App. I l
Limit **
U2 Liquid Releases Critical Organ Dose 5.6E-4 2E-3 1.4E-2 4.lE-3 10 L
Critical Organ Liver Liver Bone Bone Total Body Dose 3.5E-4 1E-3 6.8E-3 1.4E-3 3
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U2 Gaseous Releases q
Skin Dose 3.4E-2 9E-3 1.lE-2 1.3E-5 15 i
Total Body Dose 2.9E-4 8.E-5 9.3E-5 4.5E-6 5
j U2 Airborne Particulate / Iodine Critical Organ Dose 8.3E-3 1.lE-2 1.8E-3 2.2E-3 15 Critical Organ Skin Liver Total Total Body Body
- Source: Annual Environmental Monitoring Report for Three Mile Island Nuclear Station, Prepared by TMI Environmental Controls, GPU Nuclear Corp., Submitted on NRC Dockeb 50-320.
- Source: 10 CFR Appendix I