ML100550053

ISFSI, Submittal of Annual Radiological Environmental Monitoring Report

ML100550053
Person / Time
Site:	Three Mile Island, 07200020, 07200009
Issue date:	02/16/2010
From:	Bradley Davis US Dept of Energy, Idaho Operations Office
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
Download: ML100550053 (18)

Text

Department of Energy Idaho Operations Office 1955 Fremont Avenue Idaho Falls, ID 83415 February 16, 2010 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Submittal of the Annual Radiological Environmental Monitoring Report per 10 CFR 72.44(d) (3), for the Three Mile Island Unit 2 Independent Spent Fuel Storage Installation, Docket 72-20), and for the Ft. St. Vrain Independent Spent Fuel Storage Installation, Docket 72-09 (EM-FMDP-10-018)

Dear Sir or Madam:

The Department of Energy, Idaho Operations Office hereby submits the Annual Radiological Environmental Monitoring Report per 10 CFR 72.44(d) (3) for the Three Mile Island Unit 2 (TMI-2) Independent Spent Fuel Storage Installation (ISFSI) (Docket 72-20), and the Ft. St.

Vrain (FSV) ISFSI (Docket 72-09). These reports cover operations at both ISFSIs for calendar year 2009.

If you have any questions please call me at (208) 526-5381.

Sincerely, Bradley J. D vis FSV/TMI-2 ISFSI Facility Director Enclosures cc: U.S. NRC Region IV (TMI-2 and FSV ISFSI Reports)

D. A. Butcher, Colorado Dept. of Public Health (FSV ISFSI Report)

U.S. EPA - Region 8, Denver Co (FSV ISFSI Report) k1~

DOEIID-10739 (2009)

Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2 Independent Spent Fuel Storage Installation D. F. Barker G. G. Hall, CHP Published February 2010 Idaho National Laboratory Idaho Nuclear Technology and Engineering Center Idaho Falls, Idaho 83415 Prepared for the U. S. Department of Energy Assistant Secretary for Environmental Management Under DOE Idaho Operations Office Contract DE-AC07-051D14516

ABSTRACT This report presents the results of the 2009 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

SUMMARY

The purpose of this report is to present the results of the Radiological Environmental Monitoring Program (REMP) conducted during 2009 for the Three Mile Island, Unit 2, (TMI-2), Independent Spent Fuel Storage Installation (ISFSI). TMI-2 core debris was transferred to the ISFSI between March 1999 and April 2001 and remains in interim storage at the ISFSI.

The REMP was implemented from January through December 2009. Results of the loose surface radioactive contamination surveys indicated no increase in either beta or Cs-I137 radioactivity attributed to the facility operation. The results of the airborne radioactivity sampling did not indicate releases of' airborne particulate radioactivity from the loaded Horizontal Storage Modules (HSM) that would contribute to an increase in the estimated maximum potential dose commitment to the general public. The results of the thermoluminescent dosimetry network did not indicate an increase in radiation levels above pre-operational background attributed to the facility operation.

The monitoring program results support the conclusion reached in the Final Environmental Impact Statement that operation of the facility will not result in a significant dose commitment to the Maximum Exposed Individual.

CONTENTS ABSTRACT 2

SUMMARY

3 INTRODUCTION 5 PROGRAM DESCRIPTION 5 RESULTS 6 DISCUSSION . 8 CONCLUSION 10 REFERENCES. 10 FIGURES I. TMI-2 ISFSI TLD Station Locations 6 TABLES I. Highest Radiation Level Summary Inside HSM Rear Panel Doors (rnrem/h) 7

2. TMI-2 ISFSI Air Sample Results (pCi/m 3 ) 7

3. TMI-2 ISFSI TLD Results (mrem/d) 8

4. TMI-2 ISFSI Estimated Airborne Radioactive Material Releases (Ci/y) 9

5. Gamma Spectroscopy Itiercomparison Results for GEL (Bq/sarnple) 9 4

Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation INTRODUCTION The Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation (TMI-2 ISFSI) is a spent fuel dry storage facility designed for interim storage of the TMI-2 core debris. The TMI-2 ISFSI, located within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Laboratory (INL), is operated by CH2M - WG Idaho, LLC for the Department of Energy (DOE). The TMI-2 ISFSI was licensed on March 19, 1999 by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 72 for authorization to receive, possess, store, and transfer spent fuel and fuel debris, resulting from the 1979 TM i-2 accident, for a twenty-year term."'

The TMI-2 ISFSI is a modified NUI-HOMS spent fuel storage system, designated NUHOMS-l 2T.

Each of the thirty NUIIOMS-12T modules within the facility provide for the horizontal dry storage of up to twelve TMI-2 stainless steel canisters inside a dry shielded canister (DSC) which is placed inside a concrete horizontal storage module (HSM). The NUHOMS-12T, modification includes venting of the DSC through high efficiency particulate air (HEPA) grade filters during storage. The vent system allows for release of hydrogen gas, generated due to radiolysis, and monitoring and/or purging of the system during operation.

The TMI-2 core debris which had been stored in stainless steel canisters in a fuel pool at the Test Area North (TAN) site within the INL has been transferred to the TMI-2 ISFSI for interim storage. A Settlement Agreement entered into by the State of Idaho, the Department of Energy, and the Department of the Navy in October 1995 established a schedule for commencing core debris transfers by March 31, 1999, and completing such transfers by June 1, 2001 .- The first core debris transfer was completed on March 31, 1999. Nine additional transfers were completed during 2000. The remaining nineteen transfers were completed during 2001, with the last one completed on April 20, 2001.

A Radiological Environmental Monitoring P.rogram (REMP) was developed for the TMI-2 ISFSI and implemented in accordance with 10 CFR 72.44. This report presents the REMP results during the TMI-2 ISFSI operation in 2009.

PROGRAM DESCRIPTION The REM P is designed to monitor the two predominant radiation exposure pathways inherent with the facility design: potential airborne radioactivity releases and direct radiation. The airborne radioactivity release pathway is monitored using a combination of loose surface radioactive contamination surveys and periodic airborne radioactivity sampling. The direct radiation exposure pathway is monitored using thermoluminescent dosimetry (TID) located along the outer perimeter fence of the TMI-2 IS:SI. Contact radiation levels on the HSM rear panel doors and DSC purge and vent port filter housings are also measured during regularly scheduled surveillances performed in accordance with the Technical Specifications.

Loose surface radioactive contamination surveys are performed at the vent and purge ports of each DSC as well as the drain port of each loaded HISM. The survey frequency was monthly during the 5

first year, quarterly during the second through fifth years, and is now annually. The frequency coincides with the radiation monitoring surveillance schedule required by the TMI-2 ISFSI Technical Specifications. ' Sample media is analyzed for beta radioactivity. Depending on the amount of beta radioactivity detected, gamma isotopic analysis is either perfbrmed for each sample or for an annual sample composite. The presence ofCs-137 is determined and quantified during the gamma isotopic analysis with a required Lower Limit of Detection (LLD) no greater than 5 nCi/sample.

Twenty-two TLD stations are located and maintained along the outer perimeter fence of the TM(-

2 ISFSI. The TLD station locations are noted in Figure 1. Dosimetry igchanged out on a quarterly frequency. The minimum detectable dose is no greater than 10 mrem.

Figure 1. TMl-2 ISFSI TLD Station Locations.

1 40 41 42 43 44 45 46 47 A

- ~ 4N

) II I I I II I I I I I I I II

/

60 \ TLD Location W Air Sampler J

411 15 14 13 112 1 11 9 1 7 1615 41312 1 50 158 57 565 53 5Z 51 A low-volume air sampler is used to collect air through a particulate filter during a seven-day period each month. The air sampler is located between the two rows of HSMs inside the TMI-2 ISFSI.

Each air particulate sample is analyzed for beta radioactivity with an LLD no greater than 0.01 pCi/m 3.

Depending on the amount of beta radioactivity detected, gamma isotopic analysis is either performed for each air particulate sample or for an annual sample composite. The presence of Cs-137 is determined and quantified during the gamma isotopic analysis with a required ILLD no greater than 0.01 pCi/m3.

RESULTS The radiation levels measured on the [-ISM rear panel doors during 2009 were all less than 5 mrem/h;well below the Technical Specification limit of 100 mrem/h. The radiation levels measured on contact with the DSC purge and vent port filter housings during 2009 were all less than or equal to 17 mremi/h; well below the Technical Specification limit of 1,200 mrem/h. Radiation levels for I lSMs 4 and 22 include I to 5 mren/h neutron radiation attributed to either spontaneous fission of Pu-240 or AmBeCrn neutron startup source material. The highest gamma radiation levels measured in the purge and vent port filter housing access areas during 2009 are summarized in Table 1.

6

Table 1. Highest Radiation Level Summary InsideHSM Rear Panel Doors (torem/h)*..

I ISM/I)SC Dose Rate HSM/I)S(C l)ose Rate lISM/I)SC Dose Rate 1/29 27 11/12 25 21/9 21 2/28 20 12/19 36 22/5 105 3/26 27 13/I5 24 23/16 10 4/I 17 14/17 37 24/11 22 5/24 8 15 Empty 25/13 35 6/23 29 16/2 <1 26/7 26 7/22 14 17/3 37 27/8 14 8/21 20 18/20 40 28/10 4 9/25 30 19/18 90 29/27 14 10/6 23 20/4 80 30/14 100 The loose surface contamination survey results for the purge, vent, and drain ports were less than the Minimum Detectable Activity (MDA), 99 dpm/l 00 cm 2 beta/gamma and 17 dpm/l 100 cm 2 alpha, calculated in accordance with NUREG/CR-1507.5 The gamma isotopic results for the purge, vent, and drain port contamination survey composite samples from the HSMs indicated no fission product radioactivity. Cs-137 radioactivity was less than MDA which ranged from 2 to 3 pCi/sample); well below the required LLD of 5 nCi/sample (5,000 pCi/sample).

Monthly air sampling beta radioactivity results for the TMI-2 ISFSI are presented in Table 2.

Beta radioactivity was not detected above the established threshold of 4E-14 paCi/cc (0.04 pCi/m 3) on the monthly samples. Gamma spectroscopy results of the composited air samples collected throughout the 3 year did not indicate the presence of fission or activation product activity. The MDA of 2.82E-4 pCi/m for Cs- 137 (the quotient MDA of 1.98 pCi/sample divided by a composite air sample volume of 7.02E+3 3

in ) was well below the required LLD of 0.01 pCi/mn.

Table 2. TMI-2 ISFSI Air Sample Results (pCi/m 3).

Sample Date Bela Sample Date Beta January 0.039 July 0.013 February 0.024 August 0.025 March 0.012 September 0.018 April 0.016 October 0.010 May 0.019 November 0.017 June 0.009 December 0.037 TLD results are presented in Table 3 in units of mretn/d. TLD results include an artificial phantom backscatter correction of 3% to express the results in dose equivalent units. Quarterly standard deviations were generally <0.04 mrem/d. Analysis of variance results indicated quarterly variances were in all cases different than the pre-operational baseline variance measured in March 1999. T-test results indicated quarterly mean TLD responses were in all cases significantly lower than the pre-operational baseline mean measured in March 1999 due to introduction of a new environmental dosimeter and processing system in June 1999.6 Mean TLD responses were 0.4 to 0.5 mrem/d. Radiation monitoring at

. Gamma and neutron 7

other locations within the 100 meter perimeter was not performed due to extremely low building occupancy factors.

Table 3. TMI-2 ISFSI TLD Results (mrem/d).

LOCATION MAR JUN SIP DElC MEAN 40 0.5 0.4 0.5 0.4 0.5 0.5 0.4 0.5 41 0.5 0.4 42 0.5 0.4 0.4 0.4 0.4 43 0.5 0.4 0.5 0.4 0.4 44 0.5 0.4 0.5 0.4 0.4 45 0.5 0.4 0.5 0.4 0.5 46 0.5 0.4 0.4 0.4 0.4 47 0.5 0.4 0.4 0.4 0.4 48 0.5 0.4 0.4 0.4 0.4 49 0.5 0.4 0.4 0.4 0.4 50 0.5 0.4 0.4 0.4 0.4 51 0.5 0.4 0.5 0.5 0.5 52 0.5 0.5 0.5 0.5 0.5 53 0.5 0.5 0.5 0.5 0.5.

54 0.5 0.5 0.5 0.5 0.5 55 0.5 0.5 0.5 0.5 0.5 56 0.5 0.5 0.5 0.5 0.5 57 0.5 0.5 0.5 0.5 0.5 58 0.5 0.5 0.5 0.5 0.5 59 0.5 0.5 0.5 0.5 0.5 60 0.5 0.4 0.4 0.4 0.5 61 0.5 0.4 0.5 0.5 0.5 MEAN 0.5 0.4 0.5 0.5 0.5 DISCUSSION The TMI-2 ISFSI REMP was conducted in accordance with established procedures. Analytical results ofcomposited loose surface contamination surveymedia indicated no fission or activation product activity. There were no changes made to the TMI-2 ISFSI REMP during 2009.

The loose surface radioactive contamination survey and vent port radiation survey results (stable trends) neither indicate a build up of radioactivity in the vent port HEPA filters, nor a breach of DSC containment. The loose surface radioactive contamination surveying and airborne radioactivity sampling results indicate there has been no measurable release of radioactive material from the DSCs stored in the HSMs at the ISFS! above and beyond that projected in the Final Environmental Impact Statement (EIS),

estimated for 40 CFR 61 reporting purposes, and summarized in Table 4.- 8.*9 Radioanalytical results are not significantly different from pre-operational results as well as those projected in the EIS and reported in accordance with 40 CFR 61.

a

Table 4. TMI-2 ISFSI Estimated Airborne Radioactive Material Releases (Ci/y).

Radionuclide Release Radionuclide Release Radionuclide Release Cs-1 37 1.91: 2 Co-60 11-3 2.71: -2 Sr-90 1.5E-2 Pu-239 3.91:4 Eu- 155 2.41t:-5 Pu-24 I 8.2 1` 3 Sm- 151 2.91,-4 Pu-238 9.2E-5 Kr-85 2.01>33 Pu-240 2.0 1-4 Sb- 125 4.2E-6 Prm-147 4.7E -5 Ni-63 1.81-4 Cs-134 6.7E-7 Arn-241 4.11:-5' Eu-154 6.31-5 1-129 3.7F-2 The radiation dosimetry results indicate there has been no measurable increase in ambient background radiation levels outside the TMI-2 ISFSI perimeter fence attributed to storage of the TMI-2 core debris. The absence of any significant increase in radiation levels outside the TMI-2 ISFSI perimeter fence also supports conclusions reached in the EIS.

The radioanalytical laboratory that provides gamma spectroscopy services for composite sample analysis participated in the Mixed Analyte Performance Evaluation Program (MAPEP) conducted by the DOE Radiological and Environmental Sciences Laboratory (RESL) in 2009. The intercomparison results for the sample geometry used for composite samples of surface contamination survey and air sample media conducted during 2009 are summarized in Tables 5. The evaluation criteria are described at the MAPEP website (www.inl.gov/resl/mapep). The evaluation results for Cs-137 identification indicate the General Engineering Laboratory, LLC (GEL) had an acceptable average negative reporting bias of 1.2%.

Table 5. Gamma Spectroscopy Intercomparison Results for GEL (Bq/sample).

Sample (Date) Radionuclide Gll. Value RESI. Value Bias (%) Evaluation MAPEP-09-RdF20 Cs-134 2.763 2.93 -5.7 Acceptable (February 2009) Cs-137 1.487 1.52 -2.2 Acceptable Co-57 1.347 1.30 +3.6 Acceptable Co-60 1.413 1.22 +15.8 Acceptable Mn-54 2.403 2.2709 +5.8 Acceptable Zn-65 1.613 1.36 +18.6 Acceptable MAPIFP-09-RdF2 I Cs-134 0.034 Acceptable (July 2009) Cs- 137 1.397 1.40 -0.2 Acceptable Co-57 6.73 6.48 +3.9 Acceptable Co-60 1.127 1.03 +9.4 Acceptable Mn-54 5.697 5.49 +3.8 Acceptable Zn-65 4.350 3.93 +10.7 Acceptable Calibration and quality control of instrumentation used for beta analysis of surface contamination and airborne radioactivity sample media is maintained in accordance with procedures used by the Idaho Cleanup Project (ICP) Radiological Control Program.") Radioactive sources used for instrumentation calibration and quality control are traceable to the National Institute of Standards and Technology (NIST).

This release value from the TMI-2 ISFSI SAR, Table 7.2-3 accounts for Ani-241 in-growth.

9

CONCLUSION Airborne radioactivity releases and direct radiation exposure from the facility during 2009 did not contribute to any increase in the estimate of maximum potential dose commitment to the general public; characterized as 2.7E-3 mrem/y to the Maximum Exposed Individual reported in the EIS. There were no radioactive liquid effluents released from the facility, hence no radionuclides to report.

REFERENCES I. Materials License SNM-2508 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation (TAC No's L22283 and L[22800), March 19, 1999, Docket No. 72-20.

2. 10 CFR 72, "Licensing Requirements flor the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste". Code of FederalRegulations, Office of the Federal Register.

October 2004.

3. Settlement Agreement between the State of Idaho, Department of the Navy, and the Department of Energy, October 16, 1995.

4. Technical Specifications and Bases for the INL TMI-2 Independent Spent Fuel Storage Installation.

5. NUREG/CR-1507, "Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions", December 1997.

6. P. E. Ruhter, New Environmental Dosimeter Respwonse, letter PER-I 7-99, July 29, 1999.

7. NUREG-1626, "Final Environmental Impact Statement for the Construction and Operation of an Independent Spent Fuel Storage Installation to Store the Three Mile Island Unit 2 Spent Fuel at the Idaho National Engineering and Environmental Laboratory", Docket No. 72-20, March 1998.

8. 40 CFR 61, "National Emission Standards for Hazardous Air Pollutants", Subpart 14, "National Emission Standards for Emissions of Radionuclides Other Than Radon from Department of Energy Facilities", Code oJFederalRegulations, Office of the Federal Register, October 2002.

9. G. G. Hall, ProjectedRadionuclideEmissionsfi'omn the TMI-2 ISFSI, Engineering Design File 3420, February 25, 2003.

I 0. ICP, Radiological Control Manuals 15B and 15(.

t0

DOE/ID-10742 (2009)

Annual Radiological Environmental Monitoring Program Report for the Fort St. Vrain Independent Spent Fuel Storage Installation J. R. Newkirk F. J. Borst, CHP Published February 2010 Idaho National Laboratory Idaho Nuclear Technology and Engineering Center Idaho Falls, Idaho 83415 Prepared for the U. S. Department of Energy Assistant Secretary for Environmental Management Under DOE Idaho Operations Office Contract DE-AC07-051D14516

ABSTRACT This report presents the results of the 2009 Radiological Environmental Monitoring Program conducted in accordance with I 0 CFR 72.44 for the Fort St. Vrain Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the predominant radiation exposure pathway, direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

ii

SUMMARY

The purpose of this report is to present the results of the Radiological Environmental Monitoring Program (REMP) conducted during 2009 for the Fort St. Vrain (FSV)

Independent Spent Fuel Storage Installation (ISFSI). The results of the thermoluminescent dosimetry network did not indicate an increase in radiation levels above post-loading ambient background attributed to the facility operation. The monitoring program results support the conclusion reached in the Safety Analysis Report that operation of the facility will not result in a significant dose commitment greater than 0.1 5 mrem/y to the nearest resident.

CONTENTS ABSTRACT ii

SUMMARY

iii INTRODUCTION I PROGRAM DESCRIPTION I RESULTS 2 DISCUSSION 2 CONCLUSION 3 REFERENCES 3 FIGURES I. FSV ISFSI Radiological Environmental Monitoring Locations I TABLES

1. FSV ISFSI Exposure Rates (mR/d) 2 iv

Annual Radiological Environmental Monitoring Program Report for the Fort St. Vrain Independent Spent Fuel Storage Installation INTRODUCTION The Fort St. Vrain (FSV) Independent Spent Fuel Storage Installation (ISFSI) is a spent fuel dry storage facility located near Platteville, Colorado. The FSV ISFSI is operated by CH2M - WG Idaho, LLC'(CWI) for the Department of Energy (DOE). The FSV ISFSI is licensed (SNM-2504) by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 72 for authorization to store spent nuclear fuel from the Fort St. Vrain Nuclear Generating Station.' Spent fuel from the FSV reactor was transferred to the FSV ISFSI between December 26, 1991 and June 10, 1992. The FSV ISFSI license was transferred from Public Service Company of Colorado (PSCo) to the U.S. Department of Energy, Idaho Operations Office (DOE-ID) on June 4, 1999. A Radiological Environmental Monitoring Program (REMP) has been implemented for the FSV ISFSI in accordance with 10 CFR 72.44. This report presents the REMP results for 2009.

PROGRAM DESCRIPTION The REMP is designed to monitor the predominant radiation exposure pathway inherent with the facility design: direct radiation. The direct radiation exposure pathway is monitored using thermoluminescent dosimetry (TLD) located along the 100 meter perimeter fence of the FSV ISFSI.

Monitoring locations are identified in Figure I. A control station is located at the Weld County Sheriff Office in Greeley, Colorado, approximately 17 miles NNE from the FSV ISFSI. Twenty TLDs are located around the 100 meter perimeter fence to monitor direct radiation from the FSV ISFSI. One third of the perimeter fence TLDs are changed out and processed each month. The control station TLD is changed out and processed each month. TLD processing services are provided by the Idaho National Laboratory (INL).

Figure 1. FSV ISFSI Radiological Environmental Monitoring Locations 1.13

,I r- t

• FLU h,-fatl,,I cor',UI1110 IS LOCattd l*t V,&F=O tautlty $Ihe,,l OfficeŽ 1-16#

/

/

Ill '

I-~8

• ~. I-I -~

RESULTS TLD results for the FSV ISFSI are presented in Table I in units of mR/d. The mean exposure rate of 0.40 +/- 0.08 mR/d measured at the ISFSI perimeter fence is not statistically different from the pre-operational background exposure rate of 0.34 +/- 0.03 mR/d and is consistent with the five-year historical operation mean exposure rate of 0.39 +/- 0.05 mR/d last reported by Colorado State University (CSU). 2 Additionally, the control station TLD responses (0.37 +/- 0.10 mR/d) are consistent with historical values associated with the control station location. Therefore, both the perimeter fence and control TLD responses are consistent with historical values.

Table 1. FSV ISFSI Exposure Rates (mR/d)

ILocation JAN F113 MAR APR MAY JUN JUL AUG SEP OCT NOV DEC Mean I-I 0.43 0.31 0.38 0.39 0.38 1-2 0.40 0.42 0.37 0.39 0.40 1-3 0.67 0.43 0.38 0.39 0.47 1-4 0.42 0.27 0.38 0.40 0.37 1-5 0.41 0.41 0.39. 0.41 0.41 1-6 0.65 0.40 0.36 0.38 0.45 1-7 0.41 0.33 0.36 0.37 0.37 1-8 0.41 0.41 0.37 0.40 0.40

'-9 0.66 0.42 0.38 0.40 0.47 1-1(0 0.42 0.41 0.36 0.39 0.40 0.42 0.43 0.36 0.40 0.40 1-12 0.65 0.41 0.37 0.39 0)46 1-13 0.42 0.24 0.36 0.37 0.35 1-14 0.39 0.40 0.36 0.39 0.39 1-15 0.65 0.38 0.37 0.40 0.45 1-16 0.42 0.30 0.36 0.38 0.37 1-17 0.40 0.41 0.36 0.39 0.39 1-18 0.65 0.38 0.37 0.38 0.45 1-19 0.42 0.28 0.36 0.38 0.36 1-20 0.40 0.39 0.36 0.37 0.38 Mean 0.42 0.40 0.66 0.31 0.41 0.40 0.37 0.37 0.37 0.38 0.39 0.39 0.40 Control 0.38 0.38 0.65 0.24 0.38 0.34 0.33 0.33 0.33 0.34 0.36 0.36 0.37 DISCUSSION The FSV ISFSI REMP was successfully implemented during 2009. There was no loss of radiological monitoring data. There were no sampling location changes. There were no deviations from the established sampling schedule.. The results of the monitoring period ending March 2009 were unusually high while the results of the monitoring period ending April 2009 were unusually low. This was determined to be a result of exposure while in transit based on the results of the associated control dosimeters used to check for exposure during shipment between the INL and FSV. The average of the results for these two monitoring periods (0.47 mR/day) is within one standard deviation of the annual average. Therefore, the radiation dosimetry results indicate there has been no measurable increase in ambient background radiation levels beyond the FSV ISFSI perimeter fence attributed to storage of the FSV fuel. There were no radioactive liquid effluents released from the facility, hence no radionuclides to report. There are no sources of radioactive material that may become airborne during normal operations, hence no radionuclides to report.

CONCLUSION Direct radiation exposure from the facility during 2009 did not contribute to any increase in the maximum potential dose commitment (0.15 mrero/y) to the nearest resident (located 797 meters from the ISFSI) projected in the FSV ISFSI Safety Analysis Report.4 REFERENCES 10 CFR 72, "Licensing Requirements flor the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste", Code 0/FederalRegulations, Office of the Federal Register, August 1988.

2. Results of ISFSI Site Background Radiation Study, Department of Radiology and Radiation Biology, Colorado State University, November 2, 1990.

3. Fort St. Vrain Independent Spent Fuel Storage Installation (ISFSI) Radiological Environmental Monitoring Program (IREMP), Summary Report for the Period January 1 to December 31, 1997, Department of Radiological Health Sciences, Colorado State University, February 26, 1998.

4. Fort St. Vrain Independent Spent Fuel Storage Installation Safety Analysis Report, Section 7.5, Estimated Offsite Collective Dose Assessment.