ML20059A348

Discusses Plans for Treatment of Hard to Detect Nuclides During Final Radiation Survey for Decommissioning Project

ML20059A348
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/23/1993
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Joseph Austin NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
RTR-REGGD-01.086, RTR-REGGD-1.086 P-93121, NUDOCS 9312300120
Download: ML20059A348 (15)

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"*A hPublic Servicei 16805.WCR 191/2; Platteville, Colorado 80651. Sen$fcIsm.m40-December 23,1993 Fort St. Vrain Unit No.1 P-93121 U. S. Nuclear Regulatory Commission ATIN

Document Control Desk Washington, D. C. 20555 i

ATrN: Mr. John H. Austin, Chief  !

Decommissioning and Regulatory Issues Branch ,

Docket No. 50-267

SUBJECT:

Final Radiation Survey Plan, Treatment of Hard To Detect Nuclides  !

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Dear Mr. Austin:

This letter describes Public Service Company of Colorado's (PSC) plans for treatment . j of Hard To Detect Nuclides (HTDN) during the final radiation survey for the Fort St.  ;

Vrain (FSV) decommissioning project. The present FSV Final Radiation Survey Plan -l does not discuss specific plans for treating HTDN such as iron-55, tritium, or carbon-14, j which are not gamma emitters and are not readily detectable with field'. survey -

instruments. j The Final Radiation Survey Plan is included in the FSV Decommissioning Plan and is i based on the guidance of NUREG/CR-2082, " Monitoring .for Compliance with l Decommissioning Termination Survey Criteria" and, in part, the final release' criteria -  ;

established in NRC Regulatory Guide 1.86, " Termination of Operating Licenses for .

Nuclear Reactors." ' Application of the Regulatory Guide 1.86 contamination limits has l historically involved measurement of only those readily-detected beta-gamma and alpha-D j emitting contaminants as may be present.

The NRC has provided more recent guidance in NUREG/CR-5849, " Manual.for Conducting Radiological Surveys in Support of License Termination" (Draft). This guidance does not specifically address HTDN, but does provide guidance for developing site specific release criteria.

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December 23,1993 i P-93121 Page 2 To the best of our knowledge, the only NRC approved final survey plan which has addressed the treatment of HTDN is for the Shoreham decommissioning project. PSC and our decommissioning contractor, the Westinghouse Team, have reviewed the ,

Shoreham HTDN methodology and have determined that, while reducing surface contamination limits to account for the presence of HTDN was appropriate for the  :

Shoreham project, the approach is not appropriate for Fort St. Vrain. The costs  ;

associated with adopting this methodology for Fort St. Vrain cannot be justified by ALARA considerations and would provide no measurable reduction in the risk to public  ;

health and safety. As an alternative, .PSC proposes to account for HTDN by demonstrating that the dose contribution from HTDN is maintained less than 2 mrem per year, as described in the attached discussion paper, i

PSC has evaluated the Fort St. Vrain site specific levels of various radionuclides,  :'

including HTDN, in typical concrete samples, in dry active waste, and in samples obtained during the initial site radiological characterization survey. Resultant doses from  !

HTDN to individual members of the public were calculated, using the methods and dose conversion factors provided in NUREG/CR-5512, " Residual Radioactive Contamination l from Decommissioning," Volume 1, October 1992. The attachment demonstrates that i the dose contribution due to HTDN is 0.23 mrem per year for the most conservative  !

scenario. This is about two percent of the total pathway dose equivalent limit of 10 mrem per year. Based on site specific conditions at Fort St. Vrain, PSC proposes the i following treatment of HTDN for the final release survey-i

• At the conclusion of decommissioning activities in each given area, a final l radiation survey will be performed. Cleanup will be performed as necessary to meet the approved final release criteria of:  ;

!) 5 R/hr above background at I meter from accessible surfaces  ;

direct exposure limit; ,

2) Regulatory Guide 1.86 loose and fixed surface contamination  !

limits; and f

3) 10 mrem /yr maximum individual total effective dose equivalent  !

due to radioactive material concentrations in soil and water, based .  !

on pathways analyses using the guidance in NUREG/CR-5512.

• In addition, representative samples will be taken from various surfaces and  ;

structures remaining after the conclusion of decommissioning activities and j analyzed to identify the average HTDN activity due to nuclides such as iron-55,  ;

i carbon-14, and tritium. Sample collection and analysis methods comparable to those typically used for 10 CFR 61 waste characterization will be used to identify ,

l and quantify HTDN.

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• The annual dose equivalent due to the presence of HTDN will be determined, using the methods provided in NUREG/CR-5512.

• In cases where HTDN dose equivalent is less than 2 mrem per year, PSC will not take any additional actions with respect to remediation. If the dose equivalent due +

to HTDN is 2 mrem per year or greater, further evaluation will be performed to determine the dose equivalent for HTDN plus all other contaminants. If the total dose equivalent for the scenario is less than 10 mrem per year, no additional remediation will be required.

Detal:ed justification for this approach and further explanation of the site specific l characteristics that support the above proposed methodology for treating HTDN is provided in the attached discussion paper.  ;

PSC requests NRC approval of the proposed treatment of Hard To Detect Nuclides outlined above and discussed in greater detail in the attachment. A Final Radiation Survey Implementation Plan, incorporating this HTDN treatment, will be submitted to  ;

the NRC no later than February 28,1994. PSC plans to include the methodology of this Implementation Plan as an enhancement to the Final Radiation Survey Plan in Section 4 l of the FSV Decommissioning Plan. To support final survey planning and developmental  ;

activities (e.g., procedure development, equipment procurement, survey package development, and worker training) in accordance with the FSV decommissioning schedule, we request NRC approval of the proposed HTDN treatment methodology for use at Fort St. Vrain by April 1,1994.

If you have any questions regarding this information, please contact Mr. M. H. Holmes at (303) 620-1701.

l Sincerely, 1 b 4 IN.l0 0 U 0 t 7W Di/ , )S( -

Don W. Warembourg Decommissioning Program Director

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DWW/SWC l Attachment i I

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.j December 23,1993 .l' P-93121 Page 4

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cc: Regional Administrator, Region IV Mr. Robert M. Quillin, Director l Radiation Control Division  !

Colorado Department of Health  !

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e Attachment to P-93121 .

PROPOSED TREATMENT OF IIARD TO DETECT NUCLIDES DURING TIIE FORT ST. VRAIN FINAL RADIATION SURVEY Backgrounti The Fort St. Vrain (FSV) Nuclear Station was permanently shutdown in August of 1989 and is currently in the process of decommissioning. Physical decontamination and dismantlement activities have been in progress since August 1992, in accordance with an NRC-approved Decommissioning Plan. PSC currently plans to re-power the facility as a natural gas-fired generating station, that will use some of the existing secondary steam system piping and components. The majority of the secondary systems were determined to have been radiologically unaffected during the Initial Radiological Site Characterization survey.

The Final Radiation Survey Plan is included as Section 4 of the FSV Decommissioning Plan, and is based on the guidance of NUREG/CR-2082, " Monitoring for Compliance with Decommissioning Termination Survey Criteria." FSV final release criteria are as follows:

• Direct external exposure is limited to 5 R/hr above background, as measured at a distance of one meter from accessible surfaces.

• Residual loose and fixed surface contamination levels are limited to the criteria in Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors."

• Maximum individual total effective dose equivalent is 10 mrem / year, due to residual radioactive material concentrations in soil and water, based on pathways analysis using the guidance in NUREG/CR-5512, " Residual Radioactive Contamination from Decommissioning," Volume 1, October 1992.

The FSV survey plan contained in Section 4 of the Decommissioning Plan does not discuss specific treatment of Hard to Detect Nuclides (HTDN) such as iron-55, carbon-14, or tritium, whidi are not gamma emitters and are not easily detected by field survey instrumentation. The most recent NRC guidance for radiological surveys is contained in NUREG/CR-5849, " Manual for Conducting Radiological Surveys in Support of License Termination" (Draft). This guidance does not specifically address HTDN, but does provide guidance for developing site specific release criteria.

Methodologies for treating HTDN during the FSV final radiation survey are the subject of this attachment. HTDN treatment methodologies used during decommissioning of the 1

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4 Shoreham Nuclear Power Station are addressed and justification is provided for the HTDN treatment method proposed for the site specific conditions at Fort St. Vrain. _;

Shoreham's Treatment of Hard To Detect Nuclides i To the best of our knowledge, the only NRC-approved final survey plan which has addressed the treatment of HTDN is the one in use for decommissioning the Shoreham Nuclear Power Station. At Shoreham, the acceptable surface contamination levels provided in Regulatory Guide 1.86 were reduced to account for the portion of residual contamination due to iron-55.  :

The mode of decay for iron-55 is by electron capture. Although the K X-rays produced in these transformations can be detected by gas-flow proportional detection equipment, the detection efficiency is typically very low which precludes direct measurement by field survey.

The Shoreham approach involves a reduction of Regulatory Guide 1.86 acceptable levels of residual surface activity concentration by an amount equivalent to the average activity contribution ofiron-55 in residual contamination. No adjustment was made to recognize the ,

relative dose contribution resulting from iron-55. The appropriate adjustment was determined from analyses of samples collected from the reactor water clean-up system. Due to the unusually short duration of power operation at Shoreham, approximately 2 effective. ,

full-power days, the amount ofiron-55 that was produced was relatively small and required only a minor adjustment to the residual surface contamination limits provided in Regulatory Guide 1.86. For example, where Shoreham's ratio of iron-55 to cobalt-60 as the <

predominant gamma emitter is 0.2 to 1, the adjusted contamination limits are as followsi j i

Regulatory Adjusted Guide 1.86 Limit Shoreham Limit Total (fixed) contamination 5000 dpm/100 cm 2 4167 dpm/100 cm 2 Removable contamination 1000 dpm/100 cm 2 833 dpm/100 cm2 ,

In this instance, the impact upon the ability to perform field survey measurements in support  ;

oflicense termination was minimal. The methodology used at Shoreham, though appropriate '

for Shoreham, is not feasible for FSV because of the greater amount of iron-55 and other [

HTDN production resulting from operation over an extended period of time. This method would also likely be inappropriate for other licensees wishing to terminate their heense l within a reasonable period of time after ceasing extended power operations. _

Aonlicability of the Shoreham Anoroach at Fort St Vrain E Use of the Shoreham method at Fort St. Vrain is impractical because of the significantly different ratios of HTDN. The reduction of acceptable surface activity concentration levels

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by the ratio of HTDN activity levels (e.g., iron-55, carbon-14, and tritium) to activity levels of easily-detected nuclides (e.g., cobalt-60 and europium-152) in the case of Fort St. Vrain results in extremely low residual surface contamination limits that cannot be reliably.

measured by normal field survey instrumentation. The costs associated with the development of alternative instrumentation, extended counting times, reduced scanning speed, increased staff b perform the final radiation survey, and potential additional cleanup actions cannot be justified by ALARA considerations and provide no measurable reduction in risk to health and safety of the public.

The difficulties of applying the Shoreham approach to HTDN at FSV can be illustrated by three examples: ,

1. The 199310 CFR Part 61 analysis for the dry active waste stream indicated a' ratio of iron-55 to cobalt-60 as the predominant gamma emitter of 8.5 to 1 (iron-55 = 78% of total activity; cobalt-60 = 9% of total activity). This analysis also indicated a ratio of tritium to cobalt-60 of 1.3 to 1 (tritium = 12% of total activity). Using these ratios in the manner used by Shoreham would result in the following FSV limits (based on total ratio of ~

iron-55 plus tritium to cobalt-60 as the dominant gamma emitter of 9.8 to 1):

Regulatory Adjusted Guide 1.86 Limit FSV LIMIT Total (fixed) contamination limit 5000 dpm/100 cm2 463 dpm/100 cm2 Removable contamination limit 1000 dpm/100 cm 2 43 dpm/100 cm 2 This total (fixed) contamination limit of 463 dpm/100 cm 2is- not measurable using commercially available field instrumentation and standard survey techniques.

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2. The 199310 CFR Part 61 analysis for activated concrete and concrete dust generated by decommissioning activities indicated a ratio of iron-55 to europium-152 as the predominant gamma emitter of 2.5 to I (iron-55 = 16% of total activity; europium-152 =

6.3% of total activity). The tritium to europium-152 ratio was 12.3 to 1 (tritium = 77% of tual activity). Using these ratios in the manner used by Shoicham would result in the following FSV limits:

Regulatory Adjusted Guide 1.86 Limit FSV LIMIT Total (fixed) contamination limit 5000 dpm/100 cm 2 316 dpm/100 cm2 Removable contamination limit 1000 dpm/100 cm2 63 dpm/100 cm 2 This total (fixed) contamination limit of 316 dpm/100 cm2 is not measurable using commercially available field instrumentation and standard survey techniques.

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3. The Initial Site Radiological Characterization analysis for removable' contamination on readily accessible facility surfaces prior to decommissioning activities indicated a ratio ofiron-55 to cobalt 60 as the predominant gamma emitter of approximately 1.3 to 1. This example is provided for reference only since it excludes the effects of decommissioning -

activities. - Using these ratios in the manner used by Shoreham, scaled to cobalt-60 as the dominant gamma-emitter, would result in the following FSV limits:

Regulatory Adjusted Guide 1.86 Limit FSV LIMIT Total (fixed) contamination limit 5000 dpm/100 cm2 2200 dpm/100 cm 2 Removable contamination limit 1000 dpm/100 cm2 440 dpm/100 cm2 This total (fixed) contamination limit of 2200 dpm/IO0 cm 2is not measurable. using commercially available field instrumentation and standard survey techniques.

The limited detection capability of the best available field instrumentation for measuring nuclides decaying by electron-capture or nuclides emitting low energy beta radiation, coupled with the irregular nature of facility surfaces which could affect cetector-to-source geometry and hence detection sensitivity, constitute a formidable obstacle to performing direct measurement of these HTDN using field survey techniques. ' Also, it is apparent from the above examples that reduction of residual contamination limits in the manner implemented at Shoreham could result in very low residual contamination limits and unreasonably large expenditures of resources, while providing little assurance of a significant reduction in risk to public health and safety.

Proposed FSV Treatment of HTDN PSC proposes to account for the presence of HTDN such as iron-55, tritium, and carbon-14 in post-decommissioning residual radioactive material in buildings by confirming that the levels of these HTDN, when compared to cobalt-60 and other nuclides normally detected -

during field surveys, result in insignificant dose equivalent to exposed individuals involved with future use of the facility.

PSC proposes the following treatment of HTDN for the final release survey:

1. A final radiation survey will be performed, using the final release criteria identified in Section 4 of the Decommissioning Plan.

2. In addition to this survey, additional samples will be taken which are representative of the remaining surfaces and structures of the facility to determine average ~HTDN 4

t levels. These samples will be collected and analyzed in accordance with methods i comparable to those typically used for 10 CFR 61 waste characterization, to identify

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HTDN that may be present in residual radioactive materials that would not be reliably detected during field surveys.

3. Dose equivalent due to the presence of HTDN will be determined using methods and dose conversion factors provided in NUREG/CR-5512. j

4. If die calculated dose due to HTDN is less than 2 mrem / year, remediation solely for the reduction of HTDN will not be required. If the dose equivalent due to HTDN  ;

is 2 mrem per year or greater, further evaluation will be performed to determine the l' dose equivalent for HTDN plus all contaminants. If the total dose equivalent for the scenario is less than 10 mrem per year, no additional remediation will be required.

i Analysis of Dose Contributions Due to HTDN l

PSC has performed an analysis of the HTDN activity levels described in the above examples, using the methodology and dose conversion factors provided in NUREG/CR-5512. The radiation exposure scenarios analyzed include both typical building occupancy and building ,

renovation, involving the major exposure pathways of direct exposure to penetrating radiation, and inhalation and ingestion of radioactive materials. This evaluation provides a i prudently conservative estimate of annual dose equivalent to an average individual in a. -

population group exposed to residual radioactive material after completion of decommissioning activities.

The following examples and the resultant potential doses from iron-55, tritium, and carbon- l 14 are based upon the radionuclide composition specific to Fort St. Vrain as determined from a variety of samples which are representative of the contamination currently present at the facility. Nickel-63 was evaluated and determined to be of such low concentration even .

relative to the above HTDN that no further consideration was warranted. In addition to a comparison of facility radiological conditions against the limits provided in Regulatory Guide 1.86 for contaminated surfaces and those in Draft NUREG/CR-5849 for external gamma exposure rate during the Final Survey, building scenario analyses will be performed using the laboratory analyses results from a limited number of additional samples collected from '

remaining facility surfaces and structures to update or confirm these initial estimates of potential doses from HTDN. We do not expect anything to occur during decommissioning to change our ratios of HTDN to predominant gamma-emitters; in fact, the ratios may improve due to the short half-life of iron-55 compared to cobalt-60 and europium-152.

f In order to adequately evaluate the potential doses from both actisated and contaminated ,

materials, two scenarios from NUREG/CR-5512 for residual radioactive materials in l I

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buildings will be used: 1) building renovation (subsurface or volume sources); and 2) normal occupancy (surface contamination sources).

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1. Calculated Doses From HTDN in Activated Concrete -- Building Renovation Scenario During power operations at FSV, the portions of the Prestressed Concrete Reactor Vessel (PCRV) nearest to the reactor core became activated to an extent which would exceed release criteria. The tophead concrete and approximately 31 inches of activated concrete will be .

removed from the upper interior circumference of the structure prior to the Final Survey of the remaining portion of the PCRV structure.

This analysis evaluates the doses due to future PCRV dismantlement activities. These t include doses from direct exposure and from inhalation and ingestion of HTDN released during these activities. It is noted that PSC has no plans to totally dismantle the PCRV, since the site will be maintained for industrial use. By the time any further PCRV dismantlement would be undertaken, a significant fraction of the iron-55 (with a half-life of 2.7 years) and tritium (with a half-life of 12.3 years) will have decayed away. A; auch, HTDN doses due to the building renovation scenario are not likely to occur.

The results of the following analyses illustrate the armual total effective dose equivalent (TEDE) to an average individual in a population group exposed exclusively to nuclides such as iron-55 and tritium in residual radioactive material after completion of decommissioning -

activities, during the building renovation scenario.

1.1 Building Renovation (Volume Sources) - Activated Concrete, Characterization Sampic #6-The following Building Renovation pathway analysis has been based upon the ,~

laboratory analyses of a core sample of activated concrete collected from the PCRV. '

This sample was collected from the core mid-plane region where the extent of the activation is typically greatest, and at a distance of 30 inches from the reactor vessel >

liner. Since this sample is representative of the removed concrete located nearest to the remaining concrete, this sample provides a conservative estimate of the P.ctivity contained in the concrete which is expected to remain after the completion of demolition activities. ,

Nuclide Crac_entration Annual TEDE Tritium 6.57E+01 pCi/g 2.11E-05 mrem Iron-55 1.00E+01 pCi/g 3.10E-05 mrem- .

Annual TEDE: 5.21E-05 mrem 6

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1.2 Building Renovation Scenario (Volume Sources) - 199310 CFR 61 Analysis for Activated Concrete:

The following analysis is based en HTDN concentrations measured in a 1993 sample  !

of activated concrete being prepared for shipment and disposal as radioactive waste, i analyzed in accordance with 10 CFR Part 61. .g k

Nuclide Concentration Annual TEDE Tritium 1.76E+01 pCi/g 5.66E-06 mrem J Iron-55 3.55E+00 pCi/g 1.10E-05 mrem Annual TEDE: 1.67E-05 mrem 1

2. Calculated Doses From HTDN on Contamina1ed Surfaces -- Buildine Renovation '

Scenario The following analyses of the Building Occupancy scenario were based on samples collected from waste generat~1 during the dismantlement of the activated concrete and plant system components. The radionuclide composition represented by these samples is most characteristic of the radioactive material currently present at the facility. After the completion of decommissioning activities a limited number of additional samples collected from groupings of similar facility surfaces will be analyzed to define the nuclide ratios in residual surface contamination. The results of these analyses will be used to confirm the annual total effective dose equivalent to an average individual in a population group exposed exclusively to nuclides such as iron-55 and tritium in residual radioactive material after completion of decommissioning activities, during the Building Occupancy scenario.

2.1 Building Occupancy Scenario (Surface Sources) - 199310 CFR 61 Analysis for Dry Active Waste Generated During Decommissioning:

The following analysis is based on HTDN concentrations measured in a 1993 sample of Dry Active Waste generated during decommissioning, analyzed in accordance with 10 CFR Part 61.

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The ratio of the concentrations for all nuclides were normalized to 5000 dpm/100 cm  !

I of cobalt-60 contamination since europium-152 was not identified in this waste

,- stream.

Nuclide Ratio Annual TEDE Tritium 1.28 3.63E-03 mrem  ;

Iron-55 8.46 2.15E-01 mrem Carbon-14 0.11 1.05E-02 mrem Annual TEDE: 2.29E-01 mrem ,

t 2.2 Building Occupancy Scenario (Surface Sources) - Activated Concrete,  ;

Characterizelon Sampic #6:

t The following analysis is based upon the nuclide ratios identified in activated concrete l due to the potential for activated concrete dust to be present on facility surfaces.  !

Since europium-152 is the abundant gamma-emitter for activated concrete, the ratio 2

of the concentrations for all nuclides were normalized to 5000 dpm/100 cm of europium-152 contamination. Also, iron-55 activity was estimated for this sample by scaling to tritium concentrations measured in other activated concrete samples.

Nuclide Ratio Annual TEDE Tritium 18.7 5.30E-02 mrem Iron-55 2.84 7.21E-02 mrem Annual TEDE: 1.25E-01 mrem j 2.3 Huilding Occupancy Scenario (Surface Sources) - 199310 CFR 61 Analysis for Activated Concrete:

The following analysis is based on HTDN concentrations measured in a 1993 sample of activated concrete being prepared for shipment and disposal as radioactive waste, analyzed in accordance with 10 CFR 61. As in the previous example, this analysis is based on the potential for activated concrete dust to be present on remaining surfaces.

Since europium-152 is the abundant gamma-emitter for activated concrete, the ratio -t 2

of the concentrations for all nuclides were normalized to 5000 dpm/100 cm or europium-152 contamination.

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Nuclide Ratio Annual TEDE 1 Tritium 12.3 3.49E-02 mrem i Iron-55. 2.5 6.35E-02 mrem J

Annual TEDE: 9.83E-02 mrem 2.4 Building Occupancy Scenario (Surface Sources) - Radiological Site l Characterization The following analysis is based on HTDN concentrations measured in a core sample ,

of activated concrete collected from the mid-plane region of the PCRV. This sample is most representative of the concrete that will remain after the completion of decommissioning dismantlement activities, and it provides a conservative estimate of activated concrete dust that could potentially be present on remaining surfaces during i the building occupancy scenario.

The ratio ofiron-55 concentration was normalized to 5000 dpm/100 cm2 of cobalt-60 1 since cobalt-60 was the only gamma-emitter identified.

Nuclide Ratio Annual TEDE .

Iron-55 1.3 3.30E-02 mrem Annual TEDE: 3.30E-02 mrem Justification for Prooosed Treatment of HTDN - ,

The ultimate goal of decommissioning is to assure that future uses of Fort St. Vrain will not result in individuals being exposed to unacceptable levels of radiation and/or radioactive materials. As Iow As Reasonably Achievable (ALARA) exposure principles have been given due consideration during the development of the site specific release criteria, and the -

dose limitations described above will ensure that individual exposures due to HTDN are less than 2 mrem / year. This value is low compared to general public dose limits such as the 4 mrem contained in the EPA National Primary Drinking Water Regulations, the 10 mrem  ;

contained in the EPA National Emission Standards for Hazardous Air Pollutants, the 25  ;

mrem contained in the NRC Licensing Requirements for Land Disposal of Radioactive Waste  !

and in the EPA Environmental Radiation Protection Standards for Nuclear Power Operations, and the 100 mrem contained in the NRC Standards for Pro'ection Against Radiation for -

general population dose equivalent. In addition, a dose equivalent of 2 mrem per year results in a corresponding estimated lifetime excess cancer risk of 0.00004, which is extremely low compared to the estimated lifetirae cancer risk of 0.02 from exposure to  ;

background radiation. (See Reference 1)  ;

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The NRC has stated that "... in general, intakes of the electron-capture nuclides are less  !

hazardous, by an order of magnitude or more, than comparable intakes of Co-60 or Cs-137", ,

and ". . external sources of K X-rays (from electron capture decay) provide no serious problem in practical radiation protection because this type of radiation will be of importance '

in relatively few cases, and the resulting radiation dose or dose rate can be measured in ways similar to those used to measure low-energy beta doses and dose rates in most cases." (See Reference 2). In this same reference, the NRC also stated that "From a general health and safety perspective...for a given level of contamination, an increase in the percentage of -

electron capture nuclides in nuclear power plant surface contamination represents a decrease, or at least no increase, in radiological hazard."

Draft NUREG/CR-5849 requires that sites with multiple radionuclides at the time oflicense termination account for those radionuclides which would contribute greater than 10% of the  ;

total effective dose equivalent from all contaminants or which are present at concentrations which exceed 10% of their respective guideline values. Considering the nuclide ratios specific to Fort St. Vrain which have been identified to date, it is expected that the nuclide ratios identified at the end of decommissioning will not significantly affect the contribution to annual total effective dose equivalent. It should be noted that our proposed treatment of HTDN does not affect the detectability of non-HTDN.

ALARA considerations are an essential component of the decision-making process during the identification of site specific release criteria. Consideration of the biological significance _

or dose consequences associated with various alternative actions is a common and accepted practice, and is appropriate for review as part of the Fort St. Vrain final radiation survey.

Considering the future use of the FSV facility for generation of electricity, it is reasonable to assume that any potential for exposure to radiation and/or radioactive materials will be limited to the population of the workforce. The potential exposures of the futum workforce are insignificant relative to the exposures from natural background radiation (greater than 300 mrem per year in Colorado) and from average medical exposures. Draft Regulatory Guide DG 8013, "ALARA Levels for Effluents from Materials Facilities", October 1992 mentions that 10 mrem maximum to a member of the public from effluents is ALARA. The implementation of a methodology which would reduce the residual contamination limits by the proportion of HTDN and without regard to the biological significance of the nuclide >

composition will require a detection sensitivity which cannot be achieved by conventional field instrumentation. The costs associated with the - development of alternative instrumentation, extended counting times, reduced scanning speed, increased staff to perform ,

the final radiation survey, and additional cleanup actions cannot be justified by ALARA considerations for the conditions specific to Fort St.'Vrain.

Conclusion  ;

The above analyses and examples indicate that the dose contribution due to Hard To Detect Nuclides that are reasonably expected to remain after the completion of decommissioning is 10  ;

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1 significantly less than the dose due to all contaminants. In the most conservative _ scenario example, HTDN contribute only 0.23 mrem per year. This dose contribution is sufficiently low that additional efforts to detect, evaluate, or remediate, beyond the actions proposed for use at Fort St. Vrain, are not warranted.

PSC considers that the proposed FSV treatment of HTDN described above is a reasonable and technically justifiable approach to ensuring that the FSV final release criteria ' arc demonstrated to have been satisfied upon completion of FSV decommissioning activities and that the health and safety of workers and the general public are protected.

i REFERENCES

1. " Risk Assessments Methodology, Environmental Impact Statement, NESHAPS for l Radionuclides, Background Information Document - Volume I", EPA /520/1-89-005, September 1989

2. NRC Memorandum, L. J. Cunningham to J. H. Joyner, W. E. ' Cline, C. D.

Pederson, L. J. Cellan, and G. P. Yuhas, dated May 28,1992.

Subject:

" Monitoring at Nuclear Power Plants for Contamination by Radionuclides That Decay by Electron Capture" i

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