ML20076M992

Disposal of Low Level Radioactively Contaminated Secondary- Side Clean-up Resins in Onsite Settling Basins at Davis-Besse Nuclear Power Station

ML20076M992
Person / Time
Site:	Davis Besse
Issue date:	05/31/1983
From:	J. STEWART BLAND CONSULTING
To:
Shared Package
ML20076M968	List: ML20076M966 ML20076M992
References
TAC-60875, NUDOCS 8307210173
Download: ML20076M992 (24)

Text

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Hay 1983 DISPOSAL OF LOW-LEVEL RADI0 ACTIVELY CONTAMINATED SECONDARY-SIDE CLEAN-UP RESIES i

IN THE ON-SITE SETLING BASINS AT THE DAVIS-BESSE NUCLEAR POWER STATION I

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l J. Stewart Bland Consulting P.O. Box 4154 I,

Annapolis, MD 21403 301-261-8205 r

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DISPOSAL OF LOW-LEYEL RADIDACTIYELY CONTAMINATED SECONDARY-SIDE CLEAN-UP RESINS IN THE ON-SITE SETTLING BASINS AT THE DAVIS-BESSE NUCLEAR POWER STATION INTRODUCTION The identification and control of radioactive material is dictated in a I

practical manner by the capabilities to identify or measure its radioactive With technological advances in instrumentation and the increasing content.

1 attention being dictated by the required control of even the slightest contamination, much effort is being expended on the identification, control and disposal of waste materials at nuclear power plants - materials that may only be slightly radioactively contaminated not justifying the costly disporal as The NRC has recently recognized that the absolute control of these radwaste.

very low-level radioactively contaminated materials for disposal at a licensed radioactive waste burial site may 'not be the best method of treatment: control as radioactive material may not only be impractical from an operational standpoint but unjustified from a regulatory posture considering the costs of disposal and the negligible environmental radiation exposure associated with disposal as non-radioactive waste.

By IE Information Notice Number 83-05: " Obtaining Approval For Disposing of the NRC Very-Low-Level Radioactive Waste - 10 CFR Section 20 302" (NRCB3),

called attention to a little-used section of the NRC regulations (10 CFR 20.302(a)) that provides a method for obtaining NRC approval for disposing of radioactive material in a manner other than as licensed material at a radioactive waste burial site.

As specifically stated in the information notice:

The purpose of this information notice is to bring the provisions of 10 CFR The NRC staff believes that 20 302(a) to the attention of licensees.

submittals and approvals in accordance with 10 CFR 20 302(a) can provide a reasonable alternative to high cost disposals by shallow land burial at waste repositories of large volumes of material contaminated at low levels.

Such submittals could also provide a data base for further development of regulatory provisions for disposing of specific wastes below some activity level without regard to their radioactivity similar to the provisions of 10 CFR 20.306 for disposing of certain licensed materials containing low levels of carbon 14 and tritium.

There are several waste streams at nuclear power plants that have the potential of generating significant volumes of waste - waste that may only be potentially 1

It is the contaminated or contain very low-levels of radicoctivo actorial.

cvaluation of these waste streams that the NRC is calling for in IE Information Notice Number 83-05.

During periods of primary-to-secondary leakage (steam generator tube leaks) at a PW R, the condensate demineralizer clean-up resins on the norm ally non-radioactive secondary-side will become contaminated with radioactive material.

And, for a finite period of time af ter an identified leak has been isolated, the clean-up resins will continue to collect residual radioactive material in diminishing levels until the secondary-side has been recycled sufficiently to remove essentially all of the radioactivity.

Also, during routine operations, very low-levels of radioactive material may accumulate on the clean-up resins due to unidentifiable leaks (leaks thought in-part to be attributable to thermal expansions and contractions within the system).

During periods of active leaks, contamination levels can be sufficiently high to justify treatment as radioactive waste, requiring shipment to a licensed radioactive waste burial site.

However, it is the very low-level radioactive contamination of the clean-up resins af ter isolation of a leak or from unidentifiable leakage that treatment and disposal as radioactive waste is unjustified considering the activity levels, volumes, burial costs, and environmental radiation doses. The purpose of this report is to evaluate and analyze the use of the on-site settling basins at Davis-Besse for the receipt and disposal of these very low-level radioactively contaminated secondary-side clean-up resins.

SETTLING BASIN DESCRIPTION Two on-site earthen settling basins are used at Davis-Besse as the recipients of discharges from the Water Treatment Building and the Condensate Demineralizer Backwash Receiving Tank. All discharges (except during periods of Basin #1 cleaning) are directed to Basin #1. Basin #1 flows to Basin #2 via an outlet From Basin #2 flow is directed via a sump discharge to the Collection Box w ei r.

with the Collection Box being the recipient of all liquid releases from the site prior to discharge to Lake Erie.

An emergency overflow from the Basin #2 sump i

discharges to the Toussaint River.

l The basins are approximately 200 feet long, 70 feet wide, and 10 feet deep and The basins are are clay lined to minimize any potential ground water seepage.

located approximately 500 feet SE of the on-site office building that is adjacent the turbine building. Sludge depth varies depending on location and J

proximity to discharge point in the basin.

Currently, the sludge depth in Basin

1. 1 on an average is around five (5) feet. Basin #2 has an average sludge depth i

of approximately one (1) foot.

2

l The larger fraction of the discharges to Basin #1 is from the Wotor Troctaont Building which onstitutes over 90% of the total volume. Discharges to Basin #1 the Condensate Demineralizer Backwash Receiving Tank are condacted from approximately once (1) per week.

Total volume of' powdered resin per discharge is twenty (20) cubic feet in a volume of 10,000 gallons (water-resin alurry).

Basin #1 has a water volume of 875,000 gallons; Basin #2's volume is 1.376,000 With a weighted average input flow rate of 130 gallons per minute from gallons.

the Water Treatment Building and the Condensate Demineralizer Backwash Receiving Tank to the basins, conservative estimates of detention times are 2 3 days for Basin #1 and 3 7 days for Basin #2.

The basins are located approximately 3/4 mile from Lake Erie and 1/2 mile from Toussaint River.

There are no mechanisms available that could feasibly result in the release of the basin bottoms to the lake or the river.

Any basin embankment failure would result in release to the Davis-Besse site: migration of any basin bottoms via either surface run-off or ground water migration to off-site locations is extremely remote.

Figures 1 and 2 provide the settling basin configuration, receiving and discharge arrangements, and system inputs and outputs.

DOSE CRITERIA The benefical use of nuclear power in the public interest carries with it the generation of radioactive material - material that ultimately must be disposed of.

Whether disposal is at a licensed radioactive waste burial site or in some other manner, there exists a finite potential of increased radiation exposure to members of the public.

In keeping with the philosphy of the ICRP (ICRP77), it is important that for any activity that may involve the, exposure of individuals to radioactivity the following criteria should be considered:

a) no practice shall be adopted unless its introduction provides a positive net benefit; economic b) all exposures shall be kept as low as reasonably achievable, and social factors being taken into account; and c)the dose equivalent to individuals shall not exceed the limits recommended for the appropriate circu= stances by the (ICRP) Commission.

l The overall benefits of alternative disposal methods for very low-level radioactive material have already been recognized (NRC83).

However, for determining acceptable alternative disposal methods, it is necessary that any radiation exposures should not only be kept ALARA but shall not exceed a maximum 3

4 l

that is appropriate for the circumstance - the circumstance of alternntivo disposal. In keeping with this philosophy for evaluating the use of the on-site cettling basins for the disposal of slightly radioactively contaminated clean-up resins, an annual dose limit of 1 arem total body dose equivalent to a maximum cxposed individual has been used.

In determining the overall acceptability of i

this limiting dose criteria for the disposal of very low-level radioactively contaminated material in a manner other that at a licensed radioactive waste burial site, the following current NRC and ICRP philosophies and criteria on Jig, i

miniris or trifle radiation exposures were considered.

NRC - 10 CFR 20.906 Rule i

The NRC regulation 10 CFR 20 306 allows for the disposal of liquid scintillation medium and animal tissue containing 0.05 microcuries or less The of H-3 or C-14 per gram of medium without regard to its radioactivity.

NRC's dose analyses supporting this rule concluded that any actual doses to individuals of the public from the disposal of these materials as non-radioactiv e ma terial (i.e.,

not requiring disposal at a licensed radioactive waste burial site) would be less than 1 mrem /yr.

c NRC - 10 CFR 20 Draft ProDosed Revision In the draft proposed revision to 10 CFR 20 (January 13, 1983 version) the NRC has defined.dg minimis levels as being "... levels where calculated a trifle with respect to other risks risks are negligibly small, e.g.,

encountered daily." A numerical dose level of 0.001 rem (1 mrem) has been determined to represent this negligibly small or trifle risk.

This annual dose of 1 mrem is also considered to be conclusively ALARA.

ICRP - Publication 26 (ICRP77)

The International Commission on Radiological Protection has judged a risk in the range of 10-6 to 10-5 per year as likely being acceptable to any individual member of the public. Based on the risk coefficients of ICRF-26 (10-4/ rem), a risk of 10-6 per year corresponds to an annual dose of 0.01 rem (10 mrem).

l NRC - 10 CFR 20.902 Acerovals Recently the NRC has approved alternative methods of disposing of very low-These level radioactively contaminated waste at two nuclear power plants.

authorizations have been accomplished under the provisions of 10 CFR 20 302. For these two cases, the NRC approved:

4

(1) r:locatien cf cpprcximetsly 17,000 cubic fcst cf cantcainatsd geil with average concentrations of Spci/g of co-60 and 3pCi/g of Cs-137; and (2) relocation of approximately 50,000 cubic feet of contaminated settling basin bottoms with average concentration of about 12 pCi/cm3 of Co-60.

For each case, the NRC's analysis to support the approvals concluded that any doses to individuals of the public would be less than 5 mrem /yr.

EEVIROEMENTAL DOSE ASSESSMENT An evaluation of feasible release scenarios and environmental transport and exposure pathways has been performed to assess the radiological limits of disposing of low-level contaminated secondary-side clean-up resins in the on-site settling basins.

With the planned retention of the dredged basin bottoms on the Davis-Besse site (i.e.,

no off-site disposal), actual doses to any individual of the public, if any, will be exceedingly small. However, to assure a negligible potential environmental impact in the unlikely event of an accidental release or disposal off-site, release scenarios, environmental transport, and maximum individual exposures have been conservatively evaluated.

Dose pathway models used to evaluate potential off-site dose consequences from either an accidental release or from the ultimate disposal upon dredging the basin bottoms are based on the models as expressed in Regulatory Guide 1.109, with minor modifications necessary to account for the extended use of the settling basin over a five (5) year period prior to dredging.

Doses were evaluated based on the concept of committed total body dose equivalent as expressed by the ICRP-in Publications 26 and 30 (ICRP77, ICRP78).

Based on the Annual. Limits of Intake (ALI) as presented in Publication 30 (and supplements thereto), effective total body dose factors (mrem /uCi, ingested or inhaled) were derived and are presented in Table A-1. For all radionuclides evaluated the most j

limiting ALI was used: no differentiation was made between stochastic and non-I stochastic limits.

The radionuclides included in the evaluation were those presented in (ANSI 76) for a reference PWR with the very short half-life radionuclides excluded because of the operational and environmental transport and decay times.

Refer to Appendix A for the details of the dose models.

Routine Release The analyses of potential radiation exposures considered the dredging of the basins at the end of a five (5) year period and the relocation of the bottoms 5

to an on-site location.

The basin bottoms were assumed to consist of 10%

contaminated clean-up resins.

For the potential exposure to individuals of the public, it was further assumed that a portion of the dredgings was diverted, contrary to intended on-site disposal, for personal use as a food crop fertilizer.

The dredged bottoms were conservatively assumed to constitute 50% of the cultivated soil with an individual having sufficient crop yield to provide the maximum individual consumption of vegetables per Regulatory Guide 1.109 (i.e., 54% of 630 or 340 kg/yr). For the direct exposure and the inhalation pathways, it was assumed that the same individual is exposed for a total of 210 hours0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> per year (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per week throughout a 6 month growing season).

The State of Ohio (Environmental Protection Agency) requires a permit application for disposal of any dredged settling basin bottoms.

According to State officials, a preferred and desirable disposal of settling basin bottoms of this nature, if conducted off-site, would be as a fill and/or cover mix with other clay soils at a sanitary landfill. Therefore, even if the basin bottoms were released off-site, disposal would be to a sanitary landfill, where additional mixing and restricted access would further limit any potential exposures.

It is important to note that the above described dose pa thw ay analyses conservatively assumed that all secondary-side resins are contaminated at the maximum allowable concentration that yields a maximum exposed individual dose commitment of 1 arem.

As an upper bound, contaminated resins are expected to constitute 20% of all secondary-side resin discharges. Therefore, any actual doses to individuals of the public, if any, will be well within the dose criterion of 1 cres.

Accidental Release Assessment of doses to individual members of the public due to accidental release of contaminated resins considered the potential of an accidental release Since the basins are located approximatley 3/4 mile and the release pathways.

from Lake Erie and 1/2 mile from the Toussaint River with no real pathways existing for either surf ace or ground water flow to an off-site receptor, the accidental release of the basin contents is not considered. Breach of basin integrity would release the contents to the Davis-Besse site: release off-site is not feasible.

There is a remote probability of an inadvertent discharge of sulfuric acid (H SO ) to the basins that has the potential of slightly regencrating a portion 2 4 of the resins.

Actual regeneration would be very limited due to the 6

characteristica of the resins and the nntura cf the settling bnain (i.e.,

powdered resins on the basin bottom providing only limited, surface contact with the basin liquid flow).

An analysis of this release scenario proved not to be a controlling pathway.

Therefore, the controlling potential pathway for an accidental release is from the inadvertent discharge of a batch of resins from the Condensate Demineralizer Backwash Receiving Tank directly to Lake Erie instead of to the on-site settling basins - an unlikely event due to the fact that this release pathway is controlled through redundant closed valves that are only operated for cleaning of the basins.

However this pathway represents the only potential inadvertent release mechanism to the off-site environment.

For the inadvertent release, it was assumed that the discharge was to Lake Erie via the Collection Box.

A dilution factor of 57 was included which is the minimum dilution to an off-site receptor (beach wells located approximately 0.6 miles NW of discharge).

No additional radioactive decay was included in the dose analysis.

It was conservatively assumed that an individual drinks the diluted resin / water slurry discharge at the maximum individual consumption rate of Regulatory Guide 1.109 for a one (1) day period (2.0 liters).

As discussed later, in addition to evaluating doses for an inadvertant release, for establishing limiting radionuclide concentrations, consideration was also given so as not to exceed the 10 CFR 20, Appendix B.

Table 2, Column II concentrations at the beach wells.

Limiting Radionuclide Concentrations Based on the accidental and routine release pathways and dose analyses, limiting radionuclide concentrations have been calculated corresponding to a maximum

' exposed individual dose of 1 mrem, committed total body dose equivalent. Table 1 presents individual radionuclide concentrations corresponding to 1 mrem per Table 2 year effective total body dose via the identified pathways of exposure.

presents the limiting concentration for each radionuclide based on the most controlling pathway (accidental release or exposure to dredged basin bottoms).

For exposure to the dredged basin bottoms, the pathways of direct exposure, inhalation, and f ood crops have been evaluated for collective exposure to a j

single individual.

Where limiting, a single radionuclide concentration corresponding to this collective exposure has been included in Table 2.

l In addition to considering the radionuclide concentration corresponding to a maximum individual dose of 1 mrem from either dredging of the basin botto=s or an inadvertent release, a co=parison with the MFC values of 10 CFR 20, Appendix B, Table II, Column 2 at the nearest drinking water supply for an inadvertent release was also perf ormed. In a f ew cases (for the radiciodines, Ba-La-140, 7

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-.,-y

and Pr-143), the MPC value at the off-site baach wells proved to be more limiting than the 1 arem total body dose equivalent criterion.

In keeping with the guidance of Standard Review Plan 15.7 3, the limiting concentrations for these radionuclides have been established so as not to exceed the MPC values at the nearest drinking water supply.

The evaluation of limiting concentrations for Sr-89, Sr-90, and Fe-55 have not been delineated within the analyses.

Use of these radionuclides for practical operational criteria is unsuitable because of the needed radiochemical'anaylses to determine concentration.

For the identification and control of potentially radioactive material it is necessary to base evaluation on radionuclides that are readily identifiable.

Also, based on past operating data, the radionuclides, Sr-89 and Sr-90, will never be limitins.

The more abundant radionuclides of Cs-134 and Cs-137 in practicality will be the dominant dose contributors and will provide the limiting criteria for evaluating appropriate treatment and disposal of any contaminated resins.

Fe-55 is an activation, corrosion product with concentrations in the primary coolant being a fraction of the fission products as reflected in (ANSI 76).

As with Sr-89,-90, the cesiums (Cs-134,-137) will be more limiting than Fe-55; therefore, Fe-55 can also be disregarded in establishing practical, operational criteria.

(Similar evaluations could be performed to exclude essentially all radionuclides except a selected few (e.g.,

Mn-54, Cs-134 and Cs-137), which are radionuclides that will be most prominant and dominate the dose analysis. However, for completeness other readily identifiable radionuclides have been included.)

It is again important to note that any actual doses to individuals of the public from disposal of slightly contaminated resins in the on-site settling basin will be small fractions of the dose criterion of 1 mrem per year total body dose s

equivalent. It was conservatively assumed that,the basin bottoms were available for exposure to individuals of the public: actual disposition of the dredged J

basin bottoms will be on-site further limiting any potential exposure.

For an accidental release, it was assumed that the activity was available for uptake by o

the beach wells; no radioactive decay during transit to an off-site recipient

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was assumed; no additional dilution by the normal plant liquid discharges was included.

OCGIPATIONAL EXPOSURE I

y By allowing the disposal of low-level radioactively contaminated secondary-side clean-up resins in the on-site settling basins, a reduction in occupational exposure would be anticipated.

The activity on the resins would be of such low levels that any direct contribution to occupational exposures (either for p

treating the materials as radwaste or for discharge to the basin) would be negligible. However, the advantages of not having to control, process, package, 8

and ship the resins as radwaste would minimize time and effort that would be required of plant personnel in radiation areas, thereby reducing the overall occupational exposure.

COST ANALYSIS OF DISPOSAL AS RADIOACTIVE MATERIAL The option to the disposal of very low-level conataminated resins in the on-site settling basin is disposal at a licensed radioactive waste burial site. As recognized in IE Information Notice Number 83-05, for such negligible levels of contamination there are alternatives to required disposal as radwaste that are overall more desirable - not only from an operational viewpoint but also from a regulatory posture.

A major factor that has entered into the rationalization of alternative disposal is the recognized negligible or so-called " trifle" associated radiation exposures.

Another factor that is a major justification is the costs - costs of radwaste control, treament, packaging, shipment and disposal - costs that are unjustified considering the trifle radiation exposures.

Having to treat all slightly contaminated secondary-side clean-up resins as radwaste imposes an unjustified cost.

Treatment of a single 20 cubic foot batch of resins as radwaste carries with it an associated disposal cost of over $1200.

This cost does not include plant personnel time required for transfer, storage, treatment and packaging of the waste for shipment, which can be expected to et least double the shipment and disposal cost. For an estimated 10 batches of resins per year that may contain 'very low-level contamination, annual disposal costs alone could exceed $12,000.

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Table 1 Radionuclide Concentrations Corresponding to 1 ares per Year Total Body Dose Equivalent Accidental Release Exposure Pathways for Radionuclide Resin Activity

• Avg. Activityee Dredged Basin Bottoms in Discharge Tank (uCi/cm3)

(uCi/cm3) Direct Exp. Inhalation Food Crops (uCi/cm3) (uCi/cm3) (uCi/cm3) Cr-51 1.6 x 101 2.4 x 10-1 3 4 x 10-1 17 x 101 4.0 x 102-3 Mn-54 7.4 x 10-1 1.1 x 10-2 7,o x 10-4 3 2 x 10-2 6.6 x 10 Fe-59 3 1 x 10-1 4.6 x 10-3 5 4 x 10-3 1.1 x 10-1 1.3 Co-58 3.8 x 10-1 5.7 x 10-3 3 2 x 10-3 1.4 x 10-1 5.9 x 10-2 co-60 7.4 x 10-2 1.1 X 10-3 7 2 x 10-5 3 2 x 10-4 6.9 x 10-4 Y-91 1.9 x 10-1 2.9 x 10-3 1.1 2.6 x 10-2 1 3 x 10-1 Zr-95 3 8 x 10-1 5 7 x 10-3 5 0 X 10-3 2.4 x 10-2 36 Nb-95 6.7 x 10-1 1.0 x 10-2 1,1 x 1o-2 5 4 x 10-1 3 3 x 10-1 Mo-99 3.5 x 10-1 5 2 x 10-3 ese Ru-103 7.4 x 10-1 1.1 x 10-2 1,4 x jo-2 3 0 x 10-1 5.1 x 10-2 Ru-106 7.4 x 10-2 1.1 x 10-3 2.4 x 10-3 3 4 x 10-4 3.5 x 10-4 Ag-110m 1.9 x 10-1 2.9 x 10-3 2.6 x 10-4 4.4 x 10-3 4.4 x 10-4 Te-125m 3.8 x 10-1 5 7 x 10-3 1,7 x 10-1 1,1 x 10-1 5.5 x 10-4 Te-127m 2.3 x 10-1 3 4 x 10-3 1.5 x 10-1 3 6 x 10-2 1.5 x 10 Te-129m 1.9 x 10-1 2.9 x 10-3 9.6 x 10-2 1.2 x 10-1 6.3 x 10-4 Te-131m 1.1 x 10-1 1.7 x 10-3 Te-132 7.4 x 10-2 1,1 x 10-3 I-131 1.1 x 10-2 1,7 x 10-4 I-133 3.8 x 10-2 5.7 x 10-4 I-135 3.1 x 10-1 4.6 x 10-3 Cs-134 2.7 x 10-2 4,1 x 10-4 1,8 x 10-4 1.8 x 10-3 3 5 x 10-4 Cs-136 1.6 x 10-1 2.4 x 10-3 2.8 x 10-2 2.8 4.4 x 10-1 Cs-137 3 8 x 10-2 5.7 x 10-4 2.0 x 10-4 1.8 x 10-3 2.4 x 10-4 Ba-140 1.9 x 10-1 2.9 x 10-3 3 4 x 10-1 4.4 1.2 La-140 2.3 x 10-1 3.4 x 10-3 Ce-141 7.4 x 10-1 1.1 x 10-2 7 2 x 10-2 4.2 x 10-1 13 Ce-144 7.4 x 10-2 1.1 x 10-3 1.7 x 10-2 4,4 x 30-4 8.8 x 10-3 1 2 2.4 1.9 x 10 Pr-143 1.9 2.9 x 10-2 6.0 x 10 3 Considers only the volume of resins in a discharge (20 ft ) Includes the dilution of water of the Condensate Demineralizer Backwash Receiving Tank Pathways with no entry reflect the short half-life of the radionuclide relative ese to basin accumulation and environmental transport times, thereby effectively providinE no pathway because of radioactive decay. 12

TABLE 2 Limiting Radionuclide Concentrations In Secondary-side Clean-up Resins for Allowable Discharge to On-Site Ssttling Basin Radionuclide Limiting Exposure Concentration e Pathway (uCi/cm3) Cr-51 3 3 X 10-1 dredged basin bottoms Mn-54 6.2 X 10-4 dredged basin bottocs Fe-59 5.1 X 10-3 dredged basin bottoms Co-58 3 0 X 10-3 dredged basin bottoes Co-60 5.4 X 10-5 dredged basin bottoes Y-91 2.1 X 10-2 dredged basin bottoes Zr-95 4.1 X 10-3 dredged basin bottoms Nb-95 1.0 X 10-2 dredged basin bottoms Mo-99 3 5 X 10-1 accidental release Ru-103 1.0 X 10-2 dredged basin botto=s Ru-106 1.6 X 10-4 dredged basin botto=s Ag-110c 1.6 X 10-4 dredged basin bottoms Te-125m 5.4 X 10-4 dredged basin bottoms Te-127m 1.5 X 10-4 dredged basin bottoms 6.2 X 10-4 dredged basin bottoms Te-131m 1.1 X 10-1 accidental release Te-132 7.4 X 10-2 accidental release I-131 1.1 X 10-3 accidental release-MPC I-133 3.8 X 10-3 accidental release-MPC 12 accidental release-MPC I-135 1.5 x 10 Cs-134 1.1 X 10-4 dredged basin botto=s Cs-136 2.6 X 10-2 dredged basin bottoes Cs-137 1.0 X 10-4 dredged basin botto=s Ba-140 1.1 X 10-1 accidental release-MPC La-140 7 4 X 10-2 accidental release-MPC Ce-141 5.8 X 10-2 dredged basin bottocs Ce-144 4.1 X 10-4 dredged basin botto=s Pr-143 1.9 X 10-1 accidental release-MPC 'With more than one radionuclide identified in a resin batch, the evaluation for acceptable discharge to the on-site settling basin shall be based on the "su= of the fractions" rule as follows: Determine for each identified radionuclide the ratio between the concentration present and the liciting concentration; the sum of such ratios for all radionuclides may not exceed unity (1). 13 l

REFERENCES ANSI 76, American National Standards Institute,

1976, "S o ur c e Term Specifications," ANSI Standard N237-1976, (published by American Nuclear Society)

BuT7, Burson Z. G.,1977," Structure Shielding in Reactor Accidents", Health Physics 33., 287

Fo78, Foderaro A.,

" Photon Shielding Manual, Second Edition", 1978, (available from the Penn State Bookstore) ICRP78, International Commission on Radiological Protection,1978, " Limits for Intakes of Radionuclides by Workers," ICRP Publication 30, (New York: Pergamon Press) ICRP77, International Commission on Radiological Protection,

1977,

" Recommendations of the International Commission on Radiological Protection," ICRP Publication 26, (New York: Pergamon Press)

NRC83, Nuclear Regulatory Commission, 1983, IE Information Notice No. 83-05:

" Obtaining Approval for Disposing of Very-Low-Level Radioactive Waste - 10 CFR Section 20.302," (available from NRC Public Document Room)

Ko79, Kocher D. C., " Dose-Rate Conversion Factors for External Exposure to Photon and Electron Radiation from Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," ORNL/NUREG/TH-283, (available from NTIS) i L

14 9

1 APPENDIX A DOSE MODELS The dose models used for the analyses of maximum potential off-site radiation doses for the disposal of contaminated resins in the on-site settling basins are based on the models as expressed in Regulatory Guide 1.109 with necessary modifications to reflect basin operation and release pathways. For routine ~' releases (i.e., dredging of the basin bottoms at the end of a five (5) year accumulating period), the exposure pathways include direct exposure, inhalation of resuspended radioactive material, and ingestion of food crops grown on the i bottoms. For accidental releases, the most controlling pathway was the inadvertent discharge of a resin batch directly to Lake Erie and not to the settling basin. The dose models incorporate the ICRP-26 concept of a committed total body dose equivalent (ICRP77). Based on the Annual Limits of Intake as presented in ICRP Publication 30 (and supplements thereto) (ICRP78), effective total body dose f actors (arem/uci, ingested or inhaled) have been derived and I are presented in Table A-1. The radionuclides that have been included in the evaluation are those as presented in ANSI N237-1976 (ANSI 76)for a reference PWR (with the short half-life radionuclides excluded because of the operational and j environmental transport and decay times). To allow for the use of the ground plane dose factors for the exposure to a volume source (such as would be experienced for the dredged basin bottoms), two factors must be included: (1) a soil depth to determine an effective surface l contamination level; and (2) a correction to account for ground attenuation. For the dose modeling, a contaminated soil depth of twelve (12) inches has been assumed and a correction of 0.2 has been applied. The details and derivation of these values are addressed in Appendix B. RADI0 ACTIVE MATERIAL ACCUMULATION AND DECAY Because the life of the settling basin between cleaning (dredging of the basin bottoms) is expected to be 5 years, it is necessary to account for the accumulation and decay of radioactive material over this period. By solving the differential equation for the change in radioactivity as a function of rate of I productin (R) and decay ( AN) (i.e.,aN/At = R-lN), it can be shown that for a constant input rate and concentration the average concentration at any time is expressed by the following equation: ayg C 1 - e-At C At ayg = average concentration of radionuclide i in the basin at the where: C end of the time period t 15 l

C a initial conc 3ntrntica cf rtdienuclida i ocatinually cddad to the basin over the time period t radioactive decay constant for radionuclide i 1 t = time period over which the : basin continually receives radioactive material at concentration C The addition of radioactive material to the basins will not be continuous in Most of the additions will be non-radioactive - an anticipated > 905 of nature. all additions will be non-radioactive. How ev er, the simplified assumption of accounting for radioactive material accumulation and decay at a constant rate does not jeopardize the conservatism of the dose analyses. For the dose pathways associated with the dredged bottoms, the above equation has been included in the dose analyses to account for accumulation and decay. The following equations have been used to evaluate maximum potential doses to individuals from the disposal of slightly contaminated resins in the on-site settling basins. Inhalation 7 -exp(- At ) 1 b D = K1

• U ' BR e R
• P
• FR e F3 e Ci
• DFHi '(

Atb ' 6p(- ).t5) EQ2d Croos (1-exp(- lt ) b exp(- At ) Atb D = K2

• U ' FR
• FS fSd* Ci eBi
• DFIi*

h Direct ExDosure 1-exp(- Xt ) b / ' exp(- lt ) l D = K3

• U ' T
• F
• FR
• FS
• Ci
• DFGi *g Atb h

Potable Water I D = K4 8 Ueyage C e ppIf 1 where: i total body dose equivalent due to exposure via identified pathway (crec/yr) D = environmental usage or exposure time (hr/yr, kg/yr, or 1/yr; see U = Table A-1) i 16 --,----aw--

BR = 0.2, breathing rate (m3/ min) 1 X 10-6, resuspension factor (m-1) R = 18, plow depth (cm) P = FR = 0.1, fraction of settling basin bottoms that is radioactively contaminated resins FS = 0.5, fraction of the cultivated soil that is settling basin bottoms i = concentration of radionuclide i in the resins (uCi/cm3) C DFI = ingestion ef fective total body dose f actor for radionuclide 1 (mrem /uC1, 1 ingested) DFH = inhalation effective total body dose factor for radionuclide 1 (mrem /uC1, i inhaled) DFG = total body dose f actor for radionuclide i due to direct exposure i from an infinite plane source (surface deposits) (mrem /yr per uC1/cm2) S = 2000, soil density (kg/m3) d B = concentration factor for uptake of radionuclide i from soil by y edible parts of crops (pCi/kg, wet weight per pCi/kg, dry soil) T = 0.2, reduction f actor to account for differences between dose rate from volumetric contamination and assumed surface contamination (see Appendix B) i i F = 30, soil depth assumed for exposure as a surface conta=ination (cm) M = 1.7 5 X 10-2, dilution factor (minimum dilution corresponding to beach wells located approximately 0.6 miles NW of discharge) 1.50 X 10-2 dilution factor considering the liquid volume of the V = Condensate Deminerilizer Backwash Receiving Tank (10,000 gal) i 11 = radioactive decay constant for radionuclide 1 (day -1) t = 1825, time period of accumulation of bottoms prior to dredging (days) b i i 17

1 1 t = 30, decay time from dredging to ralecca to envircament and sxposure h of off-site individuals (days) = 6.0 X 10, conversion constant (cm2 m2 e min /hr) 5 / K1 = 1.0 X 10, conversion constant (cm3/m3) 6 K? K3 = 1.14 X 10-4, conversion constant (yr/hr) K4 = 1.0 X 103, conversion constant (ml/1) 1 18

l TABLE A-1 Effective Total Body Dose Factors Radionuclide Half-life Ground Plane Ingestiori Dose Inhalation Dose (days) Dose Factor' Factor" Factor ** (mrem /yr per (arem/uC1, ingested) (arem/uci, inhaled) 2 uCi/cm ) Cr-51 27.7 3 81 x 104 1.2 x 10-1 2.5 x 10-1 Mn-54 312.7 8.69 x 105 2.5 6.2 1 Fe-59 44.6 1.20 x 106 6.2 1 7 x 10 Co-58 70.8 1.03 x 106 5.0 7.1 2 Co-60 1924 2.54 x 106 2.5 x 101 1 7 x 10 1 Y-91 58.5 3 66 x 103 1.0 x 101 5 0 x 10 1 Zr-95 64.0 7.84 x 105 5.0 5 0 x 10 Nb-95 35.1 8.09 x 105 2.5 5.0 Mo-99 2.75 1.75 x 105 5.0 50 Ru-103 39.4 5 32 x 105 2.5 8.3 1 2 Ru-106 368.2 2.25 x 105 2.5 x 10 5 0 x 10 1 1 5.6 x 10 Ag-110m

• 249.8 2.88 x 106 1.0 x 10 4

1 Te-125m 58 2.50 x 10 5.0 1.2 x 10 4 1 Te-127m 109 1.29 x 10 8.3 1.7 x 10 1 1 Te-129m 33.6 1.02 x 105 3,o x 10 2.5 x 10 6 3,7 x 3o1 1.2 x 101 Te-131m 2.08 1.98 x 10 1 1 2.5 x 10 Te-132 3 26 2.83 x 105 2.5 x 10 I-131 8.04 4.49 x 105 1,7 x go 3,o x 102 2 l 1 17 x 1o 1-133 0.87 6.75 x 105 5.0 x 10 I-135 0.28 1.58 x 106 6.2 2.5 1 1 5.0 x 10 Cs-134 752.6 1.67 x 106 7.1 x 10 Cs-136 13 2 2.26 x 106 1.2 x 101 71 1 Cs-137 1.10 x 104 7.13 x 105 5.0 x 101 2.5 x 10 Ba-140 12.8 2.04 x 105 1.0 x 101 5.0 La-140 1.68 2.34 x 106 8.3 50 ce-141 32.5 1.44 x 105 2.5 8.3 2 Ce-144 284 3 3.86 x 104 2.5 x 101 5 0 x 10 k Pr-143 13.6 1.02 x 102 1.0 8.3

• Values taken from (Ko-79) se Based on the Annual Limits of Intake from ICRP 30 (ICRP78) 19

TABLE A-2 Environmental Usage and Exposure Times for Dose Assessment Pathway Usage or Accumulating Period Decay Time Exposure Time in Settling Basin (t ) from Release to b (U) (days) Exposure (t ) h (days) Drinking water 2.0 1/ day N/A 0 Inhalation 210 hr/yr 1825 (5 years) 30 Food Crops 340 ks/yr 1825 30 Direct exposure 210 hr/yr 1825 30 Drinking water pathway assumes inadvertant discharge directly to Lake Erie e { I l t s. l t 1, l l-20

i APPENDII B COMPARISON OF DOSE RATES FROM VOLUMETRIC CONTAMINATION AND SURFACE CONTAMINATION The analysis of direct exposure to contaminated soils nas tradtionally relied on the use of dose factors based upon exposure to a contaminated ground surf ace. Most actual situations involve exposure to a contaminated volume, not a smooth plane surface upon which the derivation of the ground surface dose factors are Previous evaluations have shown that natural groun'd roughness alone can based. reduce dose rates by up to 505 (BuT7). It is, therefore, necessary to consider the actual source geometry when evaluating the appropriate use of the ground [ surface dose factors. I In order to evaluate direct exposure. to a volumetric source, it is necessary to perform a more entailed dose analysis that considers the size and shape of the source and the distance to receptor. Source attenuation and build-up, which for the case of ground contamination are funtions of source depth, must be considered. To evaluate the most conservative and appropriate correction factor to apply ~ when using a ground surface exposure dose factor for evaluating volumetric contamination, a comparison of calculated doses for a 0.5 Mev gamma source and a 1.0 Hev gamma source was performed. Results of the comparison are presented in Table B-1. For the comparison dose rates were calculated using the dose factors for a ground plane exposure and assuming all the activity to the defined soil depth was contained on the surface - no consideration was given to soil attenuation and build-up. Dose rates were also calculated by the radiation transport approximation method as expressed by Foderaro (Fo78) for a truncated volumetric (Cone dimensions were selected so as to approximate a ground, cone. l source without imposing any real error.) l As demonstrated by Figure B-1, the maximum dose rate is achieved at a I contaminated soil depth of about twelve (12) inches. Further increasing the contaminated soil depth provides negligible increase in dose rate due to soil attenuation. At the depth of 12 inches and by assuming all contamination to be surface deposited, actual dose rates would be over estimated by about a factor of 6 (see Table B-1). Therefore, for the analyses of doses due to direct I exposure to contaminated ground, a correction factor of 0.2 has been included in the equation for use with an assumed contaminated soil depth of 12 inches. By applying the 0.2 correction factor, actual exposures can be reasonably approximated by using the ground plane exposure dose factors (Table A-1) and assuming all contamination within a 12 inch depth is located on the ground surface. i 21 1

Figure B-1 Dose Rate at 1 Meter. Above Contaminated Soil 107 _g ... s g..t_4mmg.

.) '.

H. =s _~d - ^ ~ ~ ~~ ^ _.:_.V. - : -1 E:) :;--- :. ; = ;-- } - i. __^_~.^ ~^ '^~ ~ ~ W.: - l 4. .__:_.==_ =.z =_: j _.._ _= = = y :...:.= = = = q - 4; M E v ~~ _: :._ = 1 e. -..: - -- :. -- _ 4 . 5 -- 3 - - - ~ - - - i - - - - - - ~ ~ ~ ~ s._.. t~ j .._..=.___.._.;....__ __._a ..._...._.s.. P

s.._. _

_i___._._._.______ g / j- - -. qc} l_.

i-j l-- - - 2 1 :

.5 M EV ~ m en v .. _ _. _ __ c i-4...____ s _q- .._.__..q__..._. 4. _..., _ _ a _.4_.._. w a ____.__.._~___.._y_.__..._._ 1.__._.._.__ 4. __q u 8. i . ___ __i. _ 1___ ._._t.. u x 'g 106 -4 j .. _. $ __ _ y_.3 n !( =._ ::. _ = =.. _..- =:. : - _== - i : :. :i.- .___..._d_.. w

21 U

ac .t.........__.. c_d_.: _3 y _1 .s- - _ :__ :._4.:

:. - :_ _ = _ : :.. :.: : :.. u j --

t ._ I {, __ .} .I _. J..: . c.._ ) j ... _ _... g. _4 j _ _ _.. _ q..._ _. _ _.. g. __.J._ __ _ _ _. .____l___._.._.._. 105 ( 0 1 3 6 12 24 Contaminated Soil Depth (inches) 22 l

- s TABLE B-1 Comparison of Calculated, Dose Rates from a Volumetric Source and an Equivalent Surface Contamination Source Energy Contaminated Dose Rate' Ratio (Mev) Soil Depth (arem/yr per uCi/cm2) Surface (inches) Equivalent Surface Volumetric Volumetricj Contamina tion *

• Sourceeen 0.5 1

1.4 X 106 1,1 X 106 1.3 1.0 2.5 X 106 1,9 X 106 1.3 0.5 3 4.2 X 106 1.8 X 106 23 1.0 7.6 X 106 3 9 X 106 1,9 0.5 6 8.4 X 106 2.3 X 106 37 1.0 1.5 X 107 4 7 X 106 32 0.5 12 1.7 X 107 2.4 X 106 7,1 1.0 (1 ft.) 3 0 X 107 5.0 X 106 6.0 0.5 24 3.4 X 107 2.4 X 106 ja 1.0 (2 ft.) 6.1 X 107 5.1 X 106 12 At one (1) meter above ground surface s i

• • Assumes all contamination within the designated depth is located on ground surface
    • Calculations based on radiation transport approximation method of Foderaro (F076) 23}}