ML20031A437
| ML20031A437 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/15/1981 |
| From: | Englehart R NUS CORP. |
| To: | |
| Shared Package | |
| ML20031A420 | List: |
| References | |
| NUDOCS 8109230517 | |
| Download: ML20031A437 (24) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION TWATED CORRESPONmmen BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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PENNSYLVANIA POWER & LIGHT COMPANY
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and
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Docket Nos. 50-387
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50-388 ALLEGHENY ELECTRIC COOPERATIVE, INC. )
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(Susquehanna Steam Electric Station, )
Units ' l and 2)
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E RICHARD W. ENGLEHART
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T September 15, 1981 8109230517 810915 '
PDR ADOCK 05000387:
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September 15, 1981 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEF07E THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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)
PENNSYLVANIA POWER & LIGHT COMPANY
)
l l
)
and
)
Docket Nos. 50-387
)
50-388 i
ALLEGHENY ELECTRIC COOPERATIVE, INC. )
)
(Susquehanna Steam Electric Station, )
Units 1 and 2)
)
TESTIMONY OF RICHARD W.
ENGLEHART l
ON CONTENTION 1 (TECHNETIUM) 1.
My name is Richard W.
Englehart, and I am Manager, Radiological Programs Department, Environmental Services Division, NUS Corporation ("NUS").
My' business address is 4 Research Place, Rockville, Maryland.
The purpose of my testimony is to aldress the Technetium-99 portion of f
Contention 1 in this proceeding.
1 2.
Contention 1 alleges, in relevant part, that the l
quantity of Technetium-99 ("Tc-99") which will be released l
during the waste management and reprocessing activities of the l
fuel cycle required for the Susquehanna Steam Electric Station
("Susquehanna") has not been, and should be, adequately assessed, and the radiological health effects of that l
l
technetium should be estimated and these estimates f actored into the cost-benefit balance for the operation of Strquehanna.
My testimony will quantify the Tc-99 releases attributable to Susquehanna and the radioactive dose commitments caused by such releases, and will demonstrate that those doses are insignifi-cant.
I.
Introduction.
3.
Technetium has no stable isotopes and is rarely fcund in nature.
Some naturally-produced technetium is generated by spontaneous fission of other elements and by cosmic neutron bombardment of natural molybdenum.
Tc-99, the longest-lived fission product isotope of technetium, is produced by the fission of uranium-235 and by the neutron activation of molybdenum-98 [1].1 4.
Tc-99 is a metal with a half-life of 220,000 years.
It decays by emitting low-energy beta particles (with 0.29 MeV maximum energy) to stable ruthenium-99.
No other radioactive emissions (alpha or gamma) are involved in the decay of Tc-99.
5.
Due to its low beta energy, Tc-99 poses no l
significant external exposure hazard.
Thus, two feet of air l
separation from a Tc-99 source will totally absorb all radia-tion; the shielding provided by work clothing and gloves would l
l protect an individual from external doses due to Tc-99.
l 1
References are listed at the end of this testimony.
l 1
6.
The potential health hazard associated with Tc-99 arises from internal exposure because decay of Tc-99 within the body will result in deposit of its energy on tissue.
Accord-ingly, the exposure pathways for Tc-99 are those that may lead to its ingestion or inhalation.
II.
Technetium emissions during the uranium fuel cycle.
7.
Tc-99 does not exist naturally in any significant amount.
It appears in the uranium fuel cycle when it is produced by fission and activation in an operating reactor, and is subsequently carried into the "back-end" of the fuel cycle.
8.
When uranium-235 undergoes fission during reactor operation, it yields, with a fission yield of 6.2%, Tc-99 at a rate of 0.84 kg of Tc-99 per metric ton of uranium at an irradiation level of 33,000 MWD /MT [2].
This production rate corresponds to about 14.3 curies ("Ci") of Tc-99/MT of uranium, or 500 Ci per reference reactor year ("RRY").
9.
Because the reactor fuel is encapsulated, and due to the long half-life of Tc-99, essentially all of that isotope produced by fission remains in the spent fuel from the reactor.
Currently, spent reactor fuel is being stored at the reactor or in interim storage facilities.
All the Tc-99 in spent fuel (500 Ci/RRY) is in elemental form and is retained within the l
2 Here and throughout this testimony, I have adopted the RRY definition in WASH-1248 [3], which is a 1000 MWe commercial reactor, assumed to be operating at 80% of its maximum capacity for one year and requiring 35 MT of enricted uranium per year. --
fuel ass emblies during the interim storage stage, with no releases occurring during that period.
III.
Long Term Fuel Cycle Options.
10.
Two fuel cycle options have been considered in the S-3 proceeding 3:
"once-through" and " uranium-only" recycle.
The once-through fuel cycle consists of packaging and disposal of spent fuel in a federal waste repository without recovery of residual fissionable isotopes.
The uranium-only recycle option includes the recovary of uranium from spent fuel and its re-enrichment for production of new fuel.
The remain-ing fission products would be packaged and disposed of in a federal repository.
Separated plutonium may be stored for possible future u.?e or recombined with the fission products and treated as waste [4].
l IV.
Once-Through Fuel Cycle Option.
I 11.
Under the once-through fuel cycle, the spent fuel stored at reactors or in interim facilities will be packaged for ultimate disposal in an underground repository in a geological formation expected to be stable for very long periods of time.
3 In the Commission S-3 rulemaking proceeding (Docket RM-50-3), the Fuel Cycle Rule Hearing Board considered the environmental impacts of the light water reactor fuel cycle.
The S-3 proceeding resulted in issuance of Table S-3 of 10 l
CFR Part 51.
The proceeding did not result in a value for Tc-99 releases associated with the uranium fuel cycle, but concluded that the omission of Tc-99 from Table S-3 was compensated by the conservative assumption of cotoplete release of iodine-129.
t l - _.
12.
Certain geologic media are known to have remained stable for tens and hundreds of millions of years.
However, over the time period of interest for Tc-99 (on the order of one million years), no repository can be guaranteed to provide perfect containment, although careful design and siting can provide reasonable assurances of long term isolation.
Thus, proposed NRC regulations (.'.0 CFR Part 60) on disposal of high level radioactive waste in geological repositories require containment of the spent fuel or other high level waste within the warte package for a minimum of 1,000 years, with a maximum release rate of one part in 100,000 per year thereafter [5].
In my opinion, the proposed repository stability period and release rate represent conservative estimates of the long range performance of waste danositories.
13.
The half life of Tc-99 being long compared to the repository stability period, it was further assumed that
\\
virtually all 500 Ci/RRY of Tc-99 would be released over a period of time to the postulated intruding groundwater.
The radiological consequences of the Tc-99 release would depend on the release rate and the fraction of the contaminated water which ends up being consumed, directly or indirectly, by humans during the period of Tc-99 activity.
14.
For this analysis, it is assumed that all the 500 Ci/RRY of Tc-99 in the repository is dissolved in ground-water over a period of 100,000 years.4 Since the fraction of 4
The leachability of Tc-99 in solid waste is a function of fractures and porosity of the host material matrix, its.
contaminated groundwater intercepted for human use is a func-tion of the release rate and such site-specific parameters as so,l/ groundwater geochemistry, distance to and dilution before human use, nature of usage, etc., any dose analysis must recog-nize the range of parameters likely to exist in any realistic situation.
V.
Uranium-Only Recycle Option.
15.
Because spent fuel is reprocessed in this option, Tc-99 can be released to the environment in additional fuel cycle steps.
After interim storage, spent fuel is sent to a reprocessing plant to produce a purified uranium hexafluoride (UF6) feed stock for a uranium enrichment facility.
The re-processing process consists of two major stages: dissolution of spent fuel in acid and separation of uranium from fission products in the acid solution.
16.
Dissolution of spent fuel material in hot nitric acid converts the elemental Tc-99 to pertechnetic acid (HTcO ),
4 a stable non-volatile form.
There have been no reports of Tc-99 loss by carryover to the off-gas treatment system and subse-quent release to the atmosphere, and no Tc-99 releases are expected at this stage (8].
l (continued) temperature and chemical composition, anC the characteristics of the leaching solution.
Further, the solubility of Tc-99 limits its rate of dissolution and transport.
Under oxida-tion-reduction conditions exist ng in natural groundwater, Tc-99 would exist in a relative.y insoluble form [6, 7].
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17.
The nitric acid solution is then processed through a series of solvent extraction cyc'es to separate the uranium from the fission products.
Luring this process, the Tc-99 is proportioned between the main uranium stream and the side stream which leads to the high level liquid waste ("HLLW")
treatment facility and, potentially, to environmental releases.
18.
The amount of partitioning of Tc-99 between the HLLW stream and the uranium product stream has been variously estimated, a realistic range for long-term operation of reprocessing plants is 8 to 25% of the Tc-99 remaining in the uranium product stream, with the rest going to the HLLW stream
[9].
19.
In the uranium-only recycle fuel cycle, a sep-arate plutonium waste stream is present which will contain some Tc-99.
Roberts [10] found less than 1 percent of the Tc-99 in i
this stream, whereas others have suggested higher values [11].
Since the eventual recovery of plutonium for recycle is uncertain, the conservative assumption is that the Tc-99 will be apportioned only batween the HLLW waste stream and the uranium stream.5 5
Should plutonium be recycled, removal of Tc-99 impurity would not be such a critical part of the fuel fabrication l
process as it is in the uranium-only recycle, because fuel manufacture would be accomplished remotely.
Assumption of l
a plutonium recycle would reduce the amount of Tc-99 in the HLLW and uranium streams.
Ultimately, however, the Tc-99 in the plutonium fuel would go to HLLW storage.
The plutonium recycle mode releases would be equivalent to those in the uranium-only cycle.
I -
- - - =
20.
In summary, of the 500 Ci/RRY entering the fuel recovery process, 8 to 25 percent (40 to 125 Ci/RRY) is
-retained in the uranium stream, to be removed at a later step.
The remaining 75 to 92 percent (175 to 460 Ci/RRY) goes with the HLLW to long-term storage.
21.
The HLLW contains essentially all the non-vola-tile fission products and transuranic elements.
Five-year old spent fuel contains approximately 1.5 million Ci/RRY of fission products, less than 0.05% of which (375 to 460 Ci) is Tc-99
[12].
22.
The reference HLLW system used as basis for Table S-3 consists of storage for five years in stainless-steel tanks, solidification using a spray calciner/in-can melting process to form a borosilicate glass, packaging this waste form in stainless-steel waste cylinders, and storage in water basins for up to 10 years [13].
Gaseous releases from the solidifica-tion process will contain volatile and semi-volatile elements and compounds, nitrogen oxide, and some particulate matter.
An off-gas treatment system consisting of condensers, scrubbers, fractionators, filters and concentrators removes the vast majority of the contaminants befora release of the gases to the atmosp. sere.
It has been estimated that between 0.2 and 1.4% of i
the Tc-99 will volatilize and enter the off-gas treatment j
system (14].
This small portion will be removed by scrubbing 6
by the off-gas syster and recycled back to the solidification l
6 The decontamination factor for Tcy99 in the off-gas system has been estimated to be 1 x 10 (15].
This, combined i [
l
process, so that essentially all the Tc-99 entering the HLLW processing system (375 to 460 Ci/RRY) is contained in the packaged waste product that is sent for burial in the waste depository.
As in the no-recycle (once-through) case, the quantity of Tc-99 released from the repository will be deter-mined by the amount, if any, of groundwater tnat intrudes into the repository at some future time, as well as the site-speci-fic features affecting transport and dose.
The same analysis utilized in the once-through case would be applicable here to determine the range of dose commitment.
23.
Meanwhile, the uranium in the uranium product stream (which is in the form of uranyl nitrate) is converted to U0 and then to UF suitable for feed to a uranium enrichment 3
6 plant.
The Tc-99 impurity contained in the uranyl nitrate accompanies the conversion process.
However, a small amount is concentrated in the non-volatile metallic fluoride residues.
The remaining Tc-99 is separated to meet the maximum permissi-ble requirement for gross beta activity in the feed to the enrichment plant; this requirement is equivalent to leaving 4 parts per million of Tc-99 in the feed [16].
24.
For the proposed Exxon Nuclear Fuel Recovery and Recycling Center [17], the Tc-99 was to be removed by passing l
(continued) l with the Tc-99 volatilization fragtion, gives an overall decon-g l
tamination factor of between 7x10 and 5 x 10 NRCStafg have estimated the overall decontamination factor at 1x10.
These figures result in an annual atmgspheric emissgon rate from the treatment process of 7 x 10 to 4.6 x 10 Ci/RRY.
l l
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the contaminated UF6 over a MgF2 pellet solid absorber.
This
~
process was designed to remove at least 99% of the Tc-99.
The Tc-99 removed from the UF w uld have been purged from the MgF 6
2 with fluorine and then treated in the off-gas system.
Tc-99 would have been removed by gas scrubbing so that virtually all the ariginal 40-125 Ci/RRY (including the amount in the non-volatile fluoride residues) would have left the system as low-level solid waste ("LLW").
The system proposed for the Exxon facility is representative of the techniques available for removing Ta-99 at a conversion plant.7 The LLW generated at the conversion stage is buried in a low-level, near-surface burial facility.
If the buried waste is reached at some future time by intruding groundwater and conveyed to the human environment, some fraction of the 40-125 Ci/RRY of Tc-99 will be available for intake by humans.
25.
Finally, a very small amount of Tc-99 (0.2 to 1.0 Ci/RRY) remains in the reprocessed uranium, which is now in the form of UF The reprocessed uranium has to be reenriched 6
prior to fabrication of new fuel.
At the enrichment plant, the Tc-99 is removed from the product stream by way of the top purge of the cascade which removes light gas contaminants.
The purge flow is passed through alumina traps and discharged to the atmosphere.
Liquid wastes containing Tc-99 are also 7
The gases vented from the conversion facility wou tain a negligible amount of Tc-99, on the order of 10 }d con-Ci/RRY. t
generated from general maintenance and equipment decontamination operations.
26.
Direct emissions of Tc-99 at the enrichment
~
plant have been estimated as 6.6 x 10 Ci/RRY to the atmos-
-2 phere and 8.5 x 10 Ci/RRY to surface water [18].
These values have been verified by reports on Tc-99 discharges at the Oak Ridge gaseous diffusion plant (19).
VI.
Summary of Tc-99 Source Terms.
27.
The 500 Ci/RRY of Tc-99 produced in a power reactor are distributed as follows:
ONCE-THROUGH FUEL CYCLE OPTION Facility Amount Released Medium Remarks (Ci/RRY)
Interim Fuel Storage 0.0 Long Term Storage Up to 500 Groundwater Release at indeterminate future time from waste depository l
l URANIUM-ONLY RECYCLE OPTION Facility Amount Released Medium Remarks (Ci/RRY)
Fuel Reprocessing 0.0 High Level Waste Processing 7.5(-8)8 to Air 4.6(-6)
-8 8
7.5(-8) = 7.5 x 10,.,._~_,_
. ~.,
Up to 375-Ground-Release at 460 water indeterminate future time from waste depository UF Conversion 1(-7)
Air 6
Up to 40-Ground-Release at 125 water indeterminate future time from low level waste burial site c
Enfichment 6.6(-3)
Air GESMO [18) 8.5(-2)
Surface GESMO Water Up to 0.2-Ground-From low 1.0 water level waste burial VII.
Environmental Pathways.
A.
Atmospheric Releases 28.
As described above and summarized in para. 27, small atmospheric releases of Tc-99 may occur at three points in the uranium-only recycle fuel cycle:
during HLLW processing (7.5 x 10-8 to 4.6 x 10-6 Ci/RRY), during UF conversion (1 x 6
-7
-3 10 Ci/RRY) and at the enrichment plant (6.6 x 10 Ci/RRY).
The predominant chemical forms of these releases have not been established, but the most stable chemical form of Tc-99 in aqueous solution is the pertechnetate ion (TcO ), and this is 4
the chemical form most likely to enter surface soil [20].
The predominant dose pathway for atmocpheric releases of Tc-99 is soil deposition, root uptake, and human ingestion [20].
29.
The pertechnetate ion is very weakly retained in non-organic soils, but is fairly strongly retained by organic l._
soils [21].
Therefore, the movement through the soil of Tc-99 from atmospheric releases is site-dependent.
In sandy and/or non-organic soils, movement would be through ground deposition, percolation through soil to the water table after repeated periods of rain, and then groundwater transport.
In organic soils and where there is a continuous source for deposition, Tc-99 content would build up over time to a level at which leaching losses equaled deposition rate.
Then, while near the surface, Tc-99 would be taken up by vegetation through the root systems and eventually reach man directly through consumption of vegetables, or indirectly through consumption of meat and milk products from grazing animals.
30.
Parameters important to the assessment of doses through this pathway are deposition velocity, a weathering factor to account for soil depletion, and the soil-to-plant transfer factor.
The deposition velocity has been estimated at 1.1 cm/sec for the composite of all deposition processes [22].
The depletion of Tc-99 in the soil through rainfall depends on the sorption rate of Tc-99 in the particular soil type, itself a site-dependent parameter.
For inorganic soils, a conserva-tively high residence time for Tc-99 is one year.
For organic soils, the residence time might be much longer.
A conservative average residence time is 15 years [22].
A conservatively high value of the soil-to-plant transfer factor for Tc-99 is 50 pCi/g fresh vegetation weight per pCi/g dry soil weight [22].
31.
Based on the values for these three parameters, it is possible to calculate the population doses due to the
__ =-
atmospheric releases of Tc 99 attributable to the fuel cycle
~
for a reference reactor.
As discussed above, only the uranium recycle option results in Tc-99 releases to the atmosphere; these releases total 0.0066 Ci/RRi, essentially all of which is released in conjunction with the enrichment process.
The most significant exposure pathway is via intake of food rather than inhalation; Roddy et al. [23] have indicated total body and organ doses (other than lung) due to inhalation of Tc-99 releases to be more than four orders of magnitude less than those due to food ingestion.9 32.
Population doses have been estimated based on the models and calculations of Roddy, adjusted for the higher soil-to-plant transfer factor and the source term of 0.0066 C1/RRY.
With these modifications, the annual population doses from the release to atmosphere of Tc-99 are: total body: 4.6 x
~4 10 man-rem /RRY; bone: 0.0011 man-rem /RRY; kidney: 0.021 man-rem /RRY; and gastrointestinal ("GI") tract: 0.091 man-rem /RRY.
Although no thyroid doses are presented by Roddy, ratios of dose conversion factors as presented by Killough et al. [24]
would indicate annual population thyroid dose values of less than 0.1 man-rem /RRY.
9 The value of soil-to-plant transfer factor for Tc-99 used by Roddy [23] was 0.25 pCi/g fresh vegetation weight per pCi/g dry soil weight.
This value probably underestimates the root uptake rate of Tc-99.
Use of a soil-to-plant transfer factor of 50 results in annual intake of Tc-99 in food (vegetables, milk and meat) about 95 times greater than that calculated for a soil / plant transfer of 0.25.
Roddy's calculations were scaled up by a factor of 95 in this testimony to account for the dif-ference in transfer factors. _
B.
Surface Water Releases and Resulting Population Doses 33.
Except for very small releases from the poten-tial migration of near-surface groundwater to surface springs and streams, there are no expected releases of Tc-99 in scrface
-2 water discharges other than the 8.5x10 Ci/RRY released luring the enrichment process.
This release results in the following
-3 doses:
8.2x10 (total body), 0.12 (GI tract), and 0.52 (thy-roid) man-rem /RFY.
All other potential Tc-99 releases in water would come from groundwater contamination.
C.
Near-Surface Groundwater Releases and Doses From Shallow Waste Burial 34.
Siting and appropriate design of LLW burial sites should prevent releases of Tc-99 to the groundwater for long periods of time.
For the purposes of this Affidavit, it is assumed that there will be eventual releases of Tc-99 to groundwater due to water penetration and leaching of materials i
I from shallow waste burial sites associated with UF conversion 6
l (40 - 125 Ci/RRY) and enrichment (0.2 - 1,0 Ci/RRY).
To arrive l
at a conservatively high estimate of groundwater releases, Tc-99 is assumed to be leached from the waste by rainwater and to migrate to an aquifer below the waste site, in which it is transported to the surface stream into which the aquifer l
discharges.
l i
35.
The medel used in this testimony for computation of groundwater releases from shallow burial sites is that
(
developed by Adam and Rogers [25] for the Maxey Flats commer-l cial low-level waste disposal facility.
The model assumes a l
i '
s.
i groundwater transport distance of 800 meters to a surface stream, wf.th a groundwater velocity of 0.1 ft/ day.
Other parameters used by the model are a sorption coefficient of 1.0
-4 (i.e., no delay), and a 10 fraction of the Tc-99 inventory being leached per year, so that the entire inventory is released over 10,000 years.
I 36.
Topulaticn doses result from water use, pri-i marily consumption of surface water between the location of l
drainage of the aquifer to the stream and the point at which fresh water discharges into the ocean.
In the case of Maxey Flats, Rock Lick Creek drains to Licking River, then to the j
Chio River, and finally to the Mississippi and the Gulf of Mexico.
Population doses are calculated based on downstream populations using the mater and the dilution to those points of intake.
l l
37.
The Maxey Flats pathway constitutes one of the longest potential fresh water paths of any LLW site in the I
United States.
It provides, therefore, a conservative basis for evaluation of potential Tc-99 doses from this environmental 6
pathway.
A population of 5.7 x 10 people is assumed to be exposed via the downstream surface waters extending along the Ohio and Mississippi Rivers [25).
This population is assumed to drink f:om the respective surface streams and to obtain fish l
and shellfish from them.
Intake values and bioaccumulation t
factors for Tc-99 were taken from Regulatory Guide 1.109.
Dose conversions factors for Tc-99, based on an uptake in blood from i l
- _. -. _. _, ~,, _...,. -, - -........ - - -,,. -.., _ _. -,
. - - - _ _. _. - ~.. _ -...,.. -.
the small intestine of 80% of the ingested materials, are 2.14
-4
-3 x 10 3.2 x 10 and 1.41 x 10-2 rem /uci, to the total body, GI tract and thyro', respectively [26].
38.
For a shallow land burial of 125 Ci/RRY, and the other parameter ~ described above, yearly population doses of 0.0012, 0.018 and 0.077 man-rem /RRY to the total body, GI tract and thyroid, respectively, would result and continue over toe assumed duration of 10,000 years.
D.
Groundwater Releases and Doses from High Level Waste Repositories 39.
For the case of Tc-99 emplaced in deep geologi-cal repositories an a constituent of high level waste from recycle fuel management practices, or in spent fuel fec+ the once-through fuel cycle, the degree of isolation is expected to be much greater than that of wastes in shallow land burial sites.
As noted above, the NRC has proposed technical criteria for disposal of high level wastes which would restrict the release rate from those wastes in any year (after 1,000 years
-5 of isolation) to 1 x 10 of the inventory (or 0.005 Ci/RRY from an inventory of 500 Ci/RRY).
40.
Assuming, very conservatively, that the liquid pathway for the deep geological repository followed that of the shallow land burial site,10 the expected annual population dose 10 Since siting studies for high level waste repositories are being conducted so as to provide great confidence in a site that l
is not subject to groundwater intrusion, and since the buriad HLLW will be placed much deeper than the LLW, its environmental avail-ability is expected to be much lower than that of LLW.,
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(after 1,000 years of isolation) would be four-tenths that of the shallow site, or a maximum of 0.00048 whole-body, 0.0072 GI tract, and 0.0308 thyroid man-rem /RRY.
VIII.
Summary of Population Doses 41.
The major potential for population dose from release of Tc-99 would result if this material were to be released from waste burial sites or repositories for either spent fuel or reprc:ensed wastes to groundwater.
It would be
-4 expected that such releases would not exceed 10 of the
-5 inventory per year for LLW sites, or 10 for HLLW sites.
Application of these bounding releases to the Susquehanna 12 units would yield, for example, approximately.031 man-rem whole body,.46 man-rem GI tract, and 1.97 man-rem thyroid, as the yearly doses that would be attrilatable to the fuel cycle for Susqueharina due to the release of the Tc-99 in buried high level waste.
42.
A summary of population doses due to 's-99 releases attributable to the Susquehanna facility wou
,e as follows:
11 For the purpose of these estimates, no difference is assumed l
between long term storage of spent fuel and long term storage of solidified nigh level waste for the two fuel cycles considered.
The difference in the total amount of Tc-99 available for potential release (up to 500 Ci/RRY for the once through cycle and 375-460 Ci/RRY for uranium recycle) is small compared to the uncertainties of an analysis dealing with events occurring tens and hundreds of thousands of years in the future; therefore, the higher value (500 Ci/RRY) is assumed.
12 Each of the Susquehanna units is rated at 1085 MWe (gross).
Unit 1 is expected to operate for 30 years (1983-2013) and Unit 2 is expected to operate for 29 years (1984-2013) before their operating licenses expire.
Together, they account for 64 RRY's.
Therefore the above population doses have to be multiplied by a factor of 64 to a ive at Susquehanna's total contribution.
18 -
ONCE-THROUGH lt'EL CYCLE Facility Population Dose Remarks Interim Fuel Storage 0.0 High Level Waste 0.031 man-rem / year Release to ground Storage (whole body) water over 100,000 0.46 man-rem / year years (GI tract) 1.97 man-rem / year (thyroid)
URANIUM-ONLY RECYCLE OPTION Facility Population Dose Remarks Fuel Reprocessing 0.0 High Level Waste Negligible Atmospheric releases Processing UF Conversion Negligible Atmospheric releases 6
Enrichr.ent 0.029 man-rem Atmospheric releases (whole body) during lifetime of 0.070 man-rem plant (30 years)
( bone) 1.34 man-rem (kidney) 5.82 man-rem (GI tract) 6.4 man-rem (thyroid) 0.52 man-rem Surface water releases (whole body) during lifetime of the 7.7 man-rem plant (30 years)
(GI tract) 33.2 man-rem (thyroid)
Low-Level Waste 0.077 man-rem / year Release to groundwater Storage (whole body) over 10,000 years 1.15 man-rera/ year (GI tract) 4.93 man-rem / year (thyroid)
High Level Waste Same as for once-See n.
11 Storage through fuel cycle. _ _ _ _ _ _ _ _ _ -
IX.
Significance of Tc-99 releases.
43.
The significance of the Tc-99 releases that could be attributed to operation of the Susquehanna facility can be best appreciated by estimating the increase in radioac-tive dose that the average person would receive as a result of the operation of the facility over the dose that this person would receive from natural sources.
44.
The population dose from natural sources such as background radiation in a single year to the affected popula-tion (the population of 5.7 million downstream from the dispo-sal site) would be 570,000 man-rem assuming 100 millirem per person per year.
Thus, the yearly releases of the Tc-99 inventory attributable to Susquehanna from shallow land burial would increase the average thyroid dose of each member of that
-4 population ragment by 8.6 x 10 mrem, or less than one-thousandth of a percent.
The whole-body dose increase would be
-5 even smaller, 1.3 x 10 mrem.
Similarly, the yearly releases of Tc-99 attributable to Susquehanna from a high level waste repository would yield an individual dose on the order of 3.5 x
-4 10 mrem, again less than one-chousandth of a percent of the annual dose due to natural background radiation; the whole-body
-6 dose would increase by only 5 x 10 mrem.
Thus, the releases of Tc-99 attributable to the Susquehanna facility would regra-t--nh an insignificant increment to the natural radiation background dose to the affected population.
21 -
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l 22 -
s
+ - - _, -,, -,,,,
.,,-.-..,..,..,,,,,y
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~
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