ML20134C872

Draft Confirmatory Survey for Fsv Nuclear Station Public Svc Co of Co,Platteville,Co

ML20134C872
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/29/1997
From:	Abelquist E OAK RIDGE ASSOCIATED UNIVERSITIES
To:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML20134C851	List: ML20134C848 ML20134C872
References
NUDOCS 9702040103
Download: ML20134C872 (52)

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DRAFT REPORT 1  :

CONFIRMATORY SURVEY 1 l

FOR THE FORT ST.VRAIN NUCLEAR STATION I 1

PUBLIC SERVICE COMPANY OF COLORADO PLAITEVILLE, COLORADO l

[ DOCKET NO. 50-267]

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E.W. ABELQUIST 1

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U.S. Nuclear Regulatory Commission .

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i CONFIRMATORY SURVEY FOR THE FORT ST. VRAIN NUCLEAR STATION PUBLIC SERVICE COMPANY OF COLORADO PLATTEVILLE, COLORADO Prepared by l

E. W. Abelquist i

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l Environmental Survey and Site Assessment Program i Environmental and Health Sciences Division Oak Ridge Institute for Science and Education

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Oak Ridge, Tennessee 37831-0117 1 1 l l

l Prepared for the U.S. Nuclear Regulatory Commission Division of Waste Management DRAFT REPORT l l

i JANUARY 1997 l I

l This report is based on work performed under an Interagency Agreement (NRC Fin. No. A-9093) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy. Oak Ridge Institute for Science and Education performs complementary work under contract number l DE-AC05-760R00033 with the U.S. Department of Energy.

This draft report has not been given full review and patent clearance, and the dissemination of its information is only for official use. No release to the public shall be made without the approval of the Communications, Printing, and Design Department, Oak Ridge Institute for Science and Education.

Toit St. Vram Nuclear Suuan (621) January 29, :907 h%ssapveport:4wam4wam003

l l

< l CONFIRMATORY SURVEY

! FOR THE I FORT ST. VRAIN NUCLEAR STATION PUBLIC SERVICE COMPANY OF COLORADO l PLATTEVILLE, COLORADO i i

Prepared by: Date:

E. W. Abelquist, Assistant Program Director i Environmental Survey and Site Assessment Program l l

Reviewed by: Date:

T. J. Vitkus, Survey Projects Manager Environmental Survey and Site Assessment Program l

Reviewed by: Date:

R. D. Condra, Technical Resources Manager

{

Environmental Survey and Site Assessment Program i Reviewed by: Date:

A. T. Payne, Administrative Services Manager  !

Quality Assurance / Health & Safety Manager Environmental Survey and Site Assessment Program Reviewed by: Date: _

W. L. Bea, Program Director Environmental Survey and Site Assessment Program 4

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l ACKNOWLEDGMENTS l The author would like to acknowledge the significant contributions of the following staff l members:

FIELD STAFF J. S. Cox T. D. Herrera A. L. Mashburn LABORATORY STAFF R. D. Condra J. S. Cox M. J. Laudeman ,

S. T. Shipley j CLERICAL STAFF -

D. K. Ash T. S. Fox K. E. Waters ILLUSTRATOR T. D. Herrera ,

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r TABLE OF CONTENTS l PAGE 1

ist o f Fi gu re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii j

i List of Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 1

Abbreviations and Acronyms .......................................iv j I

i n trod u ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 )

! l l Site De scription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Objectives ...................................................5 Document Review ..............................................5 Proce du re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 j Fi n di n gs an d Re su l ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 1

l Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 l

S u m m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Re fe re nce s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 l Appendices:

Appendix A: Major instrumentation I

i Appendix B: Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86, Ter.nination of Operating Licenses for Nuclear

[ Reactors 1 l

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LIST OF FIGURES PAGE FIGURE 1: Location of the Fort St. Vrain Site-Platteville, Colorado . . . . . . . . . . . 14 FIGURE 2: Plot Plan of the Fort St. Vrain Nuclear Station . . . . . . . . . . . . . . . . . . 15 FIGURE 3: A0007, Visitor Center #107-Measurement and Sampling Locations . . . . 16 FIGURE 4: B0006, Equipment Building #17-Measurement and Sampling Locations . 17 l,

FIGURE 5: B0006, Machine Shop #15-Measurement and Sampling Locations . . . . . 18 FIGURE 6: B0006, Paint Shed #1003-Measurement and Sampling Locations . . . . . . 19 FIGURE 7: B0006, Weld Shop #16-Measurement and Sampling Locations . . . . . . . 20 FIGURE 8: B0012, Building 1, Helium Storage-Measurement and Sampling Locations 21 FIGURE 9: C0004, Turbine Building, Northwest Office Areas / Electrical Shop-- ,

Measurement and Sampling Imations . . . . . . . . . . . . . . . . . . . . . . . 22 i l FIGURE 10: C0009, Turbine Lube Oil Storage Tank Room-l Measurement and Sampling Locations . . . . . .. .. .. ......... 23 l

FIGURE 11: C0030, Turbine Deck, Northwest General Area-Measurement and Sampling Locations . . . . .. .. ...............24 FIGURE 12: C0031, Turbine Deck, Southwest General Area-Measurement and Sampling Locations . . . . . . . . . . . .... . .. ... . 25 FIGURE 13: D4800, Alternate Cooling System-Measurement and Sampling Locations . . 26 FIGURE 14: F0077, Reactor Building, Level 5 East, SE General Area-Measurement and Sampling Locations . .. ..... .... .... ..... 27 FIGURE 15: F0084, Reactor Building, Level 7 West, SW General Area-l Measurement and Sampling Locations .. ... . . . ... . . . 28 1

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LIST OF TABLES l

PAGE TABLE 1: Summary of Surface Activity Levels . . . . . . . . . . . . . . . . . . . . . . . . 29  ;

TABLE 2: Comparison of Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 TABLE 3: Statistical Test for Surface Activity Measurements . . . . . . . . . . . . . . . . 32 l

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l ABBREVIATIONS AND ACRONYMS R/h microroentgens per hour ASME

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American Society of Mechanical Engineers em centimeter l cm 2 square centimeter 1 i

cpm counts per minute  !

DOE Department of Energy dpm/100 cm2 disintegrations per minute /100 square centimeters  ;

EML Environmental Measurements Laboratory l

. EPA Environmental Protection Agency l

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ESSAP Environmental Survey and Site Assessment Program l l FSV Fort St. Vrain GM_ Geiger-Muller HTDN hard-to-detect nuclide HTGR High Temperature Gas-Cooled Reactor m meter l m2 square meter l mm millimeter  ;

MDC minimum detectable concentration MeV million electron volts MWe Megawatts electric-Nal sodium iodide i- NIST National Institute of Standards and Technology NRC Nuclear Regulatory Commission

, OR!SE Oak Ridge Institute for Science and Education l

pCi/g picoeuries per gram PCRV prestressed concrete reactor vessel

PIC pressurized ionization chamber l PSC Public Service Company of Colorado l

A l

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l CONFIRMATORY SURVEY l

FOR THE FORT ST. VRAIN NUCLEAR STATION PUBLIC SERVICE COMPANY OF COLORADO PLATTEVILLE, COLORADO INTRODUCTION The Public Service Company of Colorado (PSC) operated a 330 MWe High Temperature Gas Cooled Reactor (HTGR) from July 1979 until August 1989. The plant, designated as the Fort St.

Vrain Nuclear Station (FSV), was authorized for construction on September 17,1968 when the U.S. Nuclear Regulatory Commission (NRC) issued a provisional construction permit.

Construction was completed in December 1973 and a facility operating license, License No.

DPR-34, Docket No. 50-267, was granted on December 21, 1973. Initial fuel loading commenced on December 26,1973 and initial criticality was achieved January 31,1974. After a prolonged period of startup testing, low-power operation and plant modifications, the plant was committed for commercial operation on July 1,1979. Full power was achieved November 6, 1981 (PSC 1995a).

In the nuclear steam supply system for FSV, heat was produced by fission in the HTGR utilizing a uranium-thorium fuel cycle. Graphite was used for the moderator, core structure, and reflector.

High temperature helium was used as the primary coolant to produce superheated and reheated steam at a temperature of 1,000 F to match conventional thermal station conditions. The entire nuclear steam supply system, including the reactor core, graphite moderator and reflector, ste:im generators and helium circulators, was contained within a Prestressed Concrete Reactor Vessel (PCRV).

During the operational period, FSV operated for approximately 890 effective full-power days; FSV was shut down on August 18,1989. The PSC Board of Directors reviewed and confirmed the Executive Management decision that FSV would not be restarted, and that PSC would pursue decommissicning of FSV. The decision to permanently shut down and decommission FSV was based on related technical and financial considerations. Problems were identified with the control I

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rod drive assemblies and the steam generator steam ring headers that presented significant technical obstacles which could be overcome, but at a significant financial cost and time

! commitment. In addition, due to the uniqueness of the HTGR fuel cycle, the cost to purchase new fuel was prohibitive. This, in conjunction with low plant availability and correspondingly high l operating costs, made continued operation of FSV impractical.

PSC's objective is the dismantlement and decommissioning of FSV to release all site areas for i unrestricted use. To accomplish this, a portion of the PCRV structure and the radioactive balance-of-plant equipment that exceed the limits for unrestricted use was decontaminated or l

removed as described in the Fort St. Vrain Decommissioning Plan. In May 1991, the NRC granted a 10 CFR 50 Possession Only License. On November 23,1992, the NRC issued the Order to Authorize Decommissioning of Fort St. Vrain and Amendment No. 85 to Possession Only License No. DPR-34 (PSC 1995b).

4 The FSV facility was largely left intact following decommissioning; dismantlement of structures was confined to the PCRV, and portions of the Reactor Building, Turbine Building, and Liquid Waste System.

Following defueling, the PCRV contained the majority of the remaining radioactive material inventory. Portions of the PCRV concrete were activated due to direct irradiation from the reactor core, and concrete exceeding unrestricted use limits was removed prior to final survey and disposed of as radioactive waste at a licensed radioactive waste disposal facility. Thus, the radoactive source term at FSV was primarily a result of neutron activation of both metallic and concrete components of the PCRV and neutron activation of impurities contained in graphite components of the PCRV. These activation products included beta-gamma emitters such as Co-60, Eu-152, and Eu-154, and low-energy beta and x-ray emitters such as H-3, C-14, and Fe-55.

It should be noted that H-3 and Fe-55 were the largest contributors to the total radionuclide inventory (PSC 1995a).

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FSV's final survey has included all pertinent structures, surfaces, systems and components, concentrating on those previously identified as contaminated or potentially contaminated during the dismantlement / decommissioning phases. The FSV final status survey included:

• Sampling of soil, pavement, water, and liquid effluent ditch and pond sediment for

, radionuclide analysis and measurement of gamma exposure rate outside the restricted area l

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1 l- of PSC property, Sampling of soil, basin sediment, pavement and water for radionuclide analysis and j measurement of gamma exposure rate inside the restricted area of PSC property, l

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• Radiological surveys of the PCRV and Reactor Building, and Radiolo . cal surveys of the Turbine Building, Radwaste Compactor Building, New Fuel j Storage Building, Radiochemistry Laboratory, Helium Transfer and Storage System, and l

Liquid Radwaste System.

At the request of the NRC's Division of Waste Management, the Environmental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) performed an independent confirmatory suvey of the repower area in March 1995 at the Fort St.

I Vrain site in Platteville, Colorado (ORISE 1995a). Subsequent independent survey activities at FSV included licensee survey package reviews, afirmatory surface scans, and comparison surface activity measurements (e.g., side-by-side measurements) performed from September 25 through 27,1995 (ORISE 1996a). During the peiiod January 22 through 25,1996, ESSAP l performed instrument comparison activities-including side-by-side surface activity measurements 1 and surface scans-and reviews of the licensee's embedded piping program and use of in situ gamma spectrometry for determining the licensed material contribution to exposure rate (ORISE I

1996b). Most recently, ESSAP performed independent confirmatory surveys during the period l September 30 through October 3,1996. Specifically, ESSAP performed surface scans, direct measurements of surface activity, and exposure rate measurements; and reviewed the licensee's l

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hard-to-detect nuclide (HTDN) assessment program. A preliminary report describing these

activities was prepared by ESSAP and submitted to the NRC in a letter dated October 22,1996 (ORISE 1996c).

SITE DESCRIPTION The FSV facility is located approximately 56 kilometers (35 miles) north of Denver and 5.6 kilometers northwest of the town of Platteville, in Weld Cour.ty, Colorado (Figure 1). The site is located in an agricultural area with gently rolling hills. Grade elevation at the plant is 1,460 i

meters (4,790 feet) above sea level. The site consists of 6995 hectares (17,300 acres) owned by PSC, identified as the Owner-Controlled Area, of which approximately one square mile was designated as the exclusion area during plant operation. Farming has been continued on Owner-Controlled areas of the site, but there are no farming operations or permanent residences located within the Restricted Area. The Restricted Area is surrounded by a security fence, and access is I controlled for purposes of protection of individuals from exposure to radiation.

The station is located approximately two miles south of the confluence of the South Platte River and the St. Vrain Creek. Neither of these two streams are considered navigable. Cooling for the plant is provided by mechanical draft cooling towers. Make-up to the cooling towers is obtained i from the two streams, and is supplemented by shallow well water. Nineteen shallow wells are located on the site. The licensee also owns surface water rights in four irrigation ditches which traverse portions of the site.

The major structures within the Restricted Area include the Reactor Building which contains the PCRV, Turbine Building, Radwaste Compactor Building, New Fuel Storage Building, Technical Support Building which contains the Radiochemistry Laboratory, Mechanical Draft Cooling Towers, Warehouse and Construction Workshops, Evaporation Ponds, and the Electrical Switchyard (Figure 2). The ground surface covering within the Restricted Area is composed primarily of gravel and vegetation, with smaller portions devoted to concrete or asphalt roadways and laydown areas.

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l OBJECTIVES i

l The objectives of the confirmatory survey were to provide independent contractor field data L reviews and radiological data for use by the NRC in evaluating the adequacy and accuracy of the licensee's procedures and final status survey results.

L DOCUMENT REVIEW ESSAP reviewed the licensee's final status survey documentation for those survey units contained within Volumes 1 through 5 (PSC 1996). Additional documentation related to the licensee's HTDN program were reviewed during the on-site visit. Subsequent to the confirmatory survey, ESSAP reviewed the licensee's final status survey documentation contained in Volumes 6 through 11, including any revisions made to Volumes 1 through 5. Documents were reviewed for adequacy, accuracy, completeness, and consistency. Data were reviewed for appropriateness of ,

calculations and interpretations relative to the guidelines.

PROCEDURES During the period 9eptember 30 through October 3,1996, ESSAP performed confirmatory survey activities at the F st. Vrain site in Platteville, Colorado. Survey activities included technical review of the HTDN program and independent confirmatory survey activities. Eight survey units were i selected by the NRC, with input from ESSAP, for confirmatory survey activities-including surface scans, surface activity measurements, and exposure rate measurements. The selected survey units were (convention used-survey group / survey unit): A0007/BZ003, B0006/FZ001, B0012/FZ002, C0004/WZ001, C0009/BZ001, C0030/FZ001, C0031/SZ001, and D4800. In addition, surface scans >

were performed in NRC-selected. survey units F0015, F0039, F0077, F0084 (and a ponion of the adjacent F0115), F0126, and the refuel floor (Level 11 of the Reactor Building). The survey was in accordance with a plan dated September 16,1996 (ORISE 1996d) submitted to and approved l by the NRC's Division of Waste Management. Survey procedures were performed in accordance with the ORISE/ESSAP Survey Procedures and Quality Assurance Manuals (ORISE 1995b and i c). This report summarizes the procedures and results of the survey.

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SURVEY PROCEDURES The following procedures apply to survey units selected for independent confirmatory activities.

Reference System The reference systems established by FSV were used by ESSAP for referencing measurement and sampling locations. Measurement and sampling locations on ungridded surfaces were referenced to prominent building features or the existing grid. All measurement and sampling locations were recorded on maps prepared by the licensee for each survey unit.

Surface Scans 1

Surface scans for beta and gamma activity were performed over 50 to 100% of floor and lower wall surfaces and up to 50% ofequipment surfaces within each of the eight building surface and structure survey units selected for confirmatory survey activities. In addition, surface scans for alpha and beta activity were perfonned over accessible floor, lower wall, and equipment surfaces in NRC-selected survey units from Group F. Scans were performed using gas proportional, GM, and/or Nal scintillation detectors coupled to ratemeters or ratemeter-scalers with audible indicators. Locations of elevated direct radiation identified by scans were marked for further investigation.

Surface Activity Measurements Construction material specific backgrounds for red brick, brick, carpet, concrete, metal, painted concrete, and tile brick were determined in areas of similar construction but without a history of radioactive material use. ESSAP used the licensee's plastic shield to obtain three local area background measurements in each survey unit. The total background level at each direct i measurement location was obtained by adding the construction material background and the local l area background.

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l For each building surface and plant system survey unit selected, ESSAP performed 10 to 30 direct measurements for total beta surface activity-resulting in a total of 171 direct measurements for surface activity. Additional direct measurements were performed at two locations of elevated l j direct radiation detected by surface scans. ESSAP also performed 20 direct measurements on  ;

miscellaneous concrete, metal, and wood debris from the demolition of the New Fuel Storage Duilding and Building 28, staged on the east side of the Fort St. Vrain site. Direct measurements l:

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! were performed using GM or gas proportional detectors coupled to ratemeter-scalers.

A smear sample for determining the removable activity level was collected from each direct

' 1 measurement location. Measurement and sampling locations are shown on Figures 3 through 15.

i Exposure Rate Measurements l

l A total of 28 exposure rate measurements were performed within survey groups A0007, B0006, B0012, C0004, C0009, and C0030 (Figures 3 through 11). All exposure rates were measured at 1 meter above surfaces using a pressurized ionization chamber (PIC).

SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data were returned to ORISE's ESSAP laboratory in Oak Ridge, Tennessee for analysis and interpretation. Sample analyses were performed in accordance with the ORISE/ESSAP Laboratory Procedures Manual (ORISE 1995d). Smears were analyzed for gross alpha and gross beta activity using a low background gas proportional counter, and the results were converted to j units of disintegrations per minute per 100 square centimeters (dpm/100 cm 2), )

i Direct measurements for surface activity were converted to units of dpm/100 cm2 and the surface i activity results for each survey unit were statistically evaluated. The goal of the test was to i determine, with a given confidence level, that the FSV surface activity levels were not biased low l compared to ESSAP. The null hypothesis was stated that, in a survey unit, surface activities as calculated by FSV are greater than or equal to those determined by ESSAP, i.e., HJ PrSv 2 PESSAP-l For surface activity comparisons, it should be noted that side-by-side comparisons performed during Fort St Vram Nuclear Statmn (621). Jarmry 29.1997 7 h \essapueportsw_

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a a previous ESSAP site visit (ORISE 1996b) have detennined that ESSAP's measurements are biased low as compared to FSV's measurements. Therefore, the statistical test may be considered as a measure for demonstrating that FSV's surface activity measurements are in statistical control, relative to the previously established relationship between ESSAP and FSV. This hypothesis was tested at the 95% confidence level (0.05 level of significance). If the hypothesis was rejected at that confidence lee, the alternative hypothesis was to be accepted i.e., Hi Prsv < Ptsse. The test J statistic, t, was calculated using the following equation:

Xg - X, (ng -1pg* + (n,-l& ' ng + n ,'

) ng + n, - 2 n,n, ,

where:

X~r is the FSV mean for a survey unit Xg is the ESSAP mean for the same survey unit ng is the number of FSV measurement / sampling data points ne is the number of ESSAP measurement / sampling data points Sr, Se are the standard deviations for the FSV and ESSAP measurement data, respectively.

The calculated I was compared to the critical value of Student's t-distribution (one-tailed) for the appropriate degrees of freedom at the 95% confidence level (0.05 level of significance). If the H,:

pysy a pess,was rejected, then ESSAP conferred with the NRC as to the recommended approach.

Exposure rates were reponed in microroentgens per hour (pR/h). Additional information regarding major instrumentation, sampling equipment, and analytical procedures is provided in Appendices A and B. The data generated were compared with the licensee's documentation and NRC guidelines established for release for unrestricted use (Appendix C).

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FINDINGS AND RESULTS l

DOCUMENT REVIEW 1

ESSAP reviewed the licensee's final status survey documentation and performed an on-site record review of the HTDN program. A preliminary report on the review of the licensee's HTDN program and a comment letter documenting the review of Volumes 1 through 5 were submitted to the NRC (ORISE 1996c and e). The licensee's documentation provides an adequate description of the radiological status of the surveyed areas and sufficient information on the HTDN program.

SURVEY RESULTS Surface Senns Surface scans performed within the eight survey u o selected for confirmatory survey activities did not identify any locations of elevated direct radiation. l Surface scans performed within the Group F survey units identified two locations of elevated i direct radiation. One location of elevated direct radiation was identified in survey unit F0077,  !

an affected arer located on Level 5 of the Reactor Building. The elevated direct radiation was located in a relatively inaccessible area behind an electrical raceway, and was identified with a GM detector.

The second location of elevated direct radiation was identified while performing scans in survey unit F0084, an affected area on Level 7 of the Reactor Building. The location of elevated direct radiation was identified on the exterior wall of the PCRV using a GM detector. It was subsequently determined by the licensee that this location of elevated direct radiation was actually l within F0ll5, the survey unit immediately adjacent to, and above, survey unit F0084.

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Surface Activity Levels Surface activity levels for total beta activity and removable alpha and beta activity within the eight survey units selected for confirmatory survey activities are summarized in Table 1. Total beta 2

activity in surveyed areas ranged from -890 to 1,200 dpm/100 cm . Removable activity was less 2

than 9 dpm/100 cm for gross alpha and ranged from < 15 to 18 dpm/100 cm2 for gross beta.

Surface activity levels for total beta activity and removable alpha and beta activity for the debris l removed from Building 28 ranged from -370 to 1,400 dpm/100 cm2 . Removable activity was less 2

than 9 dpm/100 cm for gross alpha and less than 15 dpm/100 cm2 for gross beta.

The surface activity level at the location of elevated direct radiation within F0077, as measured 2

by the GM detector, was 19,600 dpm/100 cm . The surface activity level measured in survey unit F0115 with a 126-cm2 gas proportional detector resulted in 9,210 dpm/100 cm2 . It should be noted that the licensee had also identified this elevated area with their scan survey and reported the surface activity level as 8,716 dpm/100 cm2, Exposure Rates l l

Exposure rates in surveyed areas are tabulated in Table 2 and ranged from 9 to 21 R/h. i COMPARISON OF RESULTS WITH GUIDELINES The primary contaminants of concern for this site are beta-gamma emitters resulting from the operation of the FSV facility. The applicable NRC guidelines for beta-gamma emitters in unaffected areas are (NRC 1974):

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l Total Activity 5,000 dpm/100 cm2 averaged over a 1 m2 area 15,000 dpm/100 cm2 maximum in a 100 cm2 area i

Removable Activity

.1,000 dpm/100 cm2 l

The NRC has approved site-specific allowable surface contamination guidelines for H-3 and Fe-55, particularly in activated concrete and steel (NRC 1994). These viidelines are:

Total Activity 200,000 dpm/100 cm 2, averaged over 1 m 2 600,000 dpm/100 cm2 , maximum in a 100 cm2 area Removable Activity 40,000 dpm/100 cm 2 l

FSV has assessed the radionuclide composition at the site and determined the site-specific surface activity gisielines for affected areas (PSC 1996). The effective beta-gamma surface contamination limits are: )

Total activity 4,000 dpm/100 cm2 , averaged over a 1 m2 area 12,000 dpm/100 cm2 , maximum in a 100 cm2 area Removable activity 750 dpm/100 cm 2 The two locations of elevated direct radiation identified by scans that exceeded the surface activity l

l guideline were remediated by the licensee. No other direct measurements exceeded the average or maximum total surface activity guideline, All removable activity levels were below the

! appropriate guidelines.

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A comparison of the ESSAP mean surface activity levels to the FSV mean surface activity levels using the Student's t-test is provided in Table 3. The results indicated that the ESS AP mean was statistically less than or equal to the respective mean determined by FSV in all 8 of the confirmatory survey units. That is, the null hypothesis (H,: Prsv 2 PESSAP) Could not be rejected because the test ?atistic was less than the critical value of the t-test for each survey unit.

l The exposure rates guideline, measured at 1 meter above the surface, is 5 R/h above background (NRC 1988). Due to the' variable exposure rate background, ESSAP elected to compare the total exposure rate (i.e., background was not subtracted) measured during the confirmatory survey to the licensee's total exposure rate. Comparison of total exposure rates from the six survey units evaluated showed general agreement between the licensee and ESSAP, with the exception of 1 Survey Group B0012, survey unit FZ002 (Table 2). .

SUMMARY

)

During the period September 30 through October 3,1996, the Environmental Survey and Site Assessment Program of ORISE performed confirmatory survey activities at the Fort St. Vrain site in Platteville, Colorado. Survey activities included technical review of the hard-to-detect nuclide program and independent confirmatory survey activities, including surface scans, surface activity measurements, and exposure rate measurements.

l The licensee's documentation provides an adequate description of the radiological status of the  !

surveyed areas and sufficient information on the HTDN program. Confirmatory survey activities in eight survey units identified no locations of elevated direct radiation. Surface activity levels ,

were below those reported by the licensee and therefore, demonstrated compliance with the J surface activity guidelines. Statistical tests of data further support the conclusion that the licensee's surface activity levels were not biased low as compared to ESSAP's measurement results.

Fort St Vram Nudear Staten (621) . January 29.1997 12 63 aaaps,,,casasai_vrainsai_vrain003

i l Comparison of scan results from the five survey units evaluated showed general agreement between the licensee and ESSAP. Overall, the confirmatory survey results for the surveyed areas are consistent with those of the licensee and support the conclusion that surface activity levels and exposure rates satisfy the guidelines for release for unrestricted use.

l l

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Measurement and Sampling Locations Fort SL hn knew stauon (621) . Januny 29,1997 28

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l l l

TABLE I

SUMMARY

OF SURFACE ACTIVITY LEVELS l

FORT ST. VRAIN NUCLEAR STATION PLATTEVILLE, COLORADO l

Range of Total I Number of ange of RemovaW Activity j l Group / Survey Unit Measurement (dpm/li,J cm') Activity (dpm/100 cm2) I N r(s)

Locations -

Beta Alpha Beta A0007/BZ003 Floor and Lower Walls 3 30 -770 to 310 <9 < 15 B0006/FZ001 Floors 4 to 7 24 -300 to 1,100 <9 < 15 B0012/FZ002 Floors 8 20 -370 to -60 <9 < 15

)

C0004/WZ001 1

12)wer Walls 9 25 -890 to 490 <9 < 15 C0009/BZ801 Floor and Lower Walls 10 20 -460 to 210 <9 < 15 C0030/FZ001 Floors 11 30 -650 to 150 <9 < 15 C0031/SZ001 Structures and Equipment 12 10 -380 to 180 <9 < 15 l D4800 Alternate Cooling System 13 12 -480 to 1,200 <9 < 15 to 18 DEBRIS FROM BUILDING 28 l

Miscellaneous . NA 20 -370 to 1,400 <9 < 15 F0077 LEVEL 5 EAST- SE GENERAL AREA (REACTOR BUILDING)

Wall 14 1 20,000 <9 < 15 l F0084 LEVEL 7 WEST- SW GENERAL AREA (REACTOR BUILDING)

' Wall (Location in F0115) 15 1 9,200 <9 < 15 Fort St Vrain Nuclear Station (621) January 29,1997 29 %vgat,aine

l l

l TABLE 2 COMPARISON OF EXPOSURE RATES l FORT ST. VRAIN NUCLEAR STATION PLATTEVILLE, COIERADO Group / Survey Unit I4 cation. Total Exposure Rate (pR/h) at 1 in Above Surface" l l A0007/BZ003 ESSAP FSV 1 21 2 21 3 19 19 to 23 4 20 l 5 19 B0006/FZ001 ESSAP FSV l

1 14 2 13 i 3 16 About 13.5 4 16 5 15 B0012/FZ802 ESSAP FSV 1 10 9 to l0 2 9 8t010 3 10 S to 7 4 11 4 to 6 5 10 5 to 6 C0004/WZ001 ESSAP FSV 1 17 2 14 l

I i 3 14 14 to 18 l ._

4 14 5 16 Fort St. Vram Nuclear stata (621) . January 29,1997 30 6 t ,, ,usi_v, soo3

l l

TABLE 2 (Continued)

COMPARISON EXPOSURE RATES FORT ST. VRAIN NUCLEAR STATION PLATI"dVILLE, COLORADO Group / Survey Unit Location. Total Exposure Rate (pR/h) at 1 m Above Surface6 C0009/BZ001 ESSAP FSV 1 12 2 12 About 12 3 10 C0030/FZ001 ESSAP FSV i 10 1

2 13 i 3 11 11 to l4 1

4 14 '

5 j4

' Refer to Figures 3 through 11.

Total exposure rates provided; background has not been subtracted.

I 1

l Fort St Vram Nuclear Station (621)- Janusy 29, t997 31 nu p,,pon,ww.m 003 l

1 i TABLE 3 p

h STATISTICAL TEST FOR SURFACE ACI'IVITY MEASUREMENTS 5

7 i FSV ESSAP 5 Test g Group / Survey Standard Standard Statistic Critical Value Unit 7 No. Meas. Deviation ' No. Meas.

g g,",2) Deviation oft-test j j d 1 2)

(dpm/100 cm2) (dpm/100 cm2)

(t)" '

-tg A0007/BZ003 67 31.5 365 3 30 - -124.4 241.6 -2.14 ~ 1.66 3 .

.t B0006/FZ001 36 325.3 349.8 24 26.3 305.0 -3.41 1.67 B0012/FZ002 78 182.9 155.8 20 -203.0 91.3 -10.6 1.66 C0004iWZ001 110 72.9 368 25 -10.0 '252.8 -1.07 1.66

, C0009/BZ001 87 -27.3 257.9 20 -123.8 220.5 -1.55 1.66 w

C0030/FZ001 314 -99.6 233.8 30 - 144.2 240.3 -1.00 1.65 C0031/SZ001 32 397.1 291 10 -151.4 171.1 -5.63 1.68

~

D4800 57 I80 303.7 12 167.7 449.2 -0.12 1.67 ,

' Null hypothesis is stated as prsv2 PEsse The null hypothesis is x@ at the 95% confidence level if the test statistic (t) is less than the entical value.

i h I J.

t

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f l'

I l

l REFERENCES Oak Ridge Institute for Science and Education (ORISE). Confirmatory Survey for the Repower Area, Fort St. Vrain, Platteville, Colorado. Oak Ridge, TN; June 1995a.

Oak Ridge Institute for Science and Education. Survey Procedures Manual for the

! Energy / Environment Systems Division, Environmental Survey and Site Assessment Program, t

Revision 9. Oak Ridge, TN; April 30,1995b.

! Oak Ridge Institute for Science and Education. Quality Assurance Manual for the Energy / Environment Systems Division, Environmental Survey and Site Assessment Program, Revision 9. Oak Ridge, TN; January 31,1995c.

Oak Ridge Institute for Science and Education. Laboratory Procedures Manual for the Energy / Environment Systems Division, Environmental Survey and Site Assessment Program, Revision 9. Oak Ridge, TN; January 31,1995d.

Oak Ridge Institute for Science and Education. Final Report-ORISE Support of NRC License Inspection at Fort St. Vrain on September 25 to 27,1995. Oak Ridge, TN; March 1996a.

Oak Ridge Institute for Science and Education. Final Report-ORISE Support of NRC License Inspection at Fort St. Vrain on January 22 to 25,1996. Oak Ridge, TN; March 1996b.

Oak Ridge Institute for Science and Education. Preliminary Report-ORiSE Support of NRC License Inspection at Fort St. Vrain on September 30 to October 3,1996 (Docket No. 50-267, RFTA No. 96-05). Oak Ridge, TN; October 22,1996c.

Oak Ridge Institute for Science and Education. Confirmatory Survey Plan for the Fort St. Vrain Nuclear Station, Public Service Company of Colorado, Platteville, Colorado (Docket No. 50-267, RFTA 96-5). Oak Ridge, TN; September 16,199Cd.

Oak Ridge Institute for Science and Education. Document Review- Fort St. Vrain Nuclear Station Decommissioning Project Final Survey Report (Volumes 1 through 5), Fort St. Vrain, Platteville, Colorado (Docket No. 50-267, RFTA No. 96-05). Oak Ridge, TN; August 14,1996e.

Public Service Company of Colorado (PSC). Final Survey Plan for Site Release (revision 1). Fort St. Vrain Nuclear Station Decommissioning Project. May 25,1995a.

Public Service Company of Colorado. Final Survey Report for Release of Repower Area. Fort St.

Vrain Nuclear Station Decommissioning Project. March 2,1995b.

Public Service Company of Colorado. Fort St. Vrain Nuclear Station Decommissioning Project, l Initial Final Survey Report (Volumes 1 through 5). May 31,1996.

l 1

Fort St Vrnm Nuclear Stauan (621) Janumy 29.1997 33 h*,,apw.po,i,(w.m oo3

l l

l REFERENCES (Continued)

U.S. Nuclear Regulatory Commission (NRC). Termination of Operating Licenses for Nuclear Reactors. Regulatory Guide 1.86. Washington, D.C.; June 1974. ,

U.S. Nuclear Regulatory Commission. Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities. NUREG-0586; August 1988.

U.S. Nuclear Regulatory Commission (letter from C.L. Pittiglio), to A.J. Bortz, Long Island Power Authority, subject " Approval of a Modification of Facility Release Criteria for Tritium and Iron-55 Surface Contamination at Shoreham Nuclear Power Station, Unit 1," June 7,1994.

j Fort St Vram Nuclear Station (621) . huary 29.1997 34 6 w ,,,,,po,,,w, _

,,,,too3

1 i

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l I

APPENDIX A MAJOR INSTRUMENTATION .

1 1

)

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Fort St. Vrain Nuclear Statmn (621). January 27 Ic77 hhap%W_vram 003

l l APPENDIX A MAJOR INSTRUMENTATION l

The display of a specific product is not to be construed as an endorsement of tie product or its l manufacturer by the author or his employer.

l DIREcr RADIATION MEASUREMENT Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM)

Ludlum Floor Monitor )

Model 239-1  !

(Ludlum Measurements, Inc.,

Sweetwater, TX)

Ludlum Ratemeter-Scaler Model 2221 ,

(Ludlum Measurements, Inc.,

Sweetwater, TX)

Detectors ,

Eberline GM Detector Model HP-260 Effective Area,20 cm2 (Eberline, Santa Fe, NM) ,

l Ludlum Gas Proportional Detector l Model 43-37  :

Effective Area,550 cm2 l

(Ludlum Measurements, Inc., j Sweetwater, TX)

, Ludlum Gas Proportional Detector l Model 43-68 l l Effective Area,126 cm2 (Ludlum Measurements, Inc.,

Sweetwater, TX) l l

Fort St Vram Nuclear Station (621)- January 29,1997 A-1 hs p..s wvonan l

l l

Reuter-Stokes Pressurized Ionization Chamber Model RSS-111 (Reuter-Stokes, Cleveland, OH)

Victoreen NaI Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal (Victoreen, Cleveland, OH)

LABORATORY ANALYTICAL INSTRUMENTATION Low Background Gas Proportional Counter Model LB-5100-W (Oxford, Oak Ridge, TN)

I Fo,t St Vram Nuclew Statmn (621)- Janumy 29.1997 A-2 6 w pwpo,1,s., .oo3

l l

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l d

APPENDIX B SURVEY AND ANALYTICAL PROCEDURES i

l I

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Fort St Vram Nuclear Statum(621). January 29,1997 h \cssapVepats\st_vnun 001

(

1 1

APPENDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES Surface Scans Surface scans were performed by passing the detectors slowly over the surface; the distance between the detector and the surface was msintained at a minimum - nominally about I cm. A large surface area, gas proportional floor monitor was used to scan the floors of the surveyed areas. Other surfaces 2 2 were scanned using small area (20 cm or 126 cm ) hand-held detectors. Identification of elevated levels was based on increases in the audible signal from the recording and/or indicating instrument.

Combinations of detectors and instruments used for the scans were:

Alpha-Beta - gas proportional detector with ratemeter-scaler Beta - gas proportional detector with ratemeter-scaler

- GM detector with ratemeter-scaler Gamma - Nal scintillation detector with ratemeter Surface Activity Measurements Measurements of total beta activity levels were primarily performed using gas proportional and GM detectors with portable ratemeter-scalers.

Count rates (cpm), which were integrated over 1 minute in a static position, were converted to 2

activity levels (dpm/100 cm ) by dividing the net rate by the 4 n efficiency and correcting for the active area of the detector. The total background level at each direct measurement location was obtained by adding the construction material background and the local area background. The local s s, pv p.t.s.i_vr o03 Fort St Vram Nuclear Statmn (621) . January 29, IM B-1

L area background was obtained in each surveyed area by placing the licensee's plastic shield between the surface being assessed and the detector. The detector's response with the plastic shield in place l was assumed to be only from the local area gamma radiatinn levels, and not from the surface material. '

Because different building construction materials (poured concrete, concrete block, metal, wood, 7

l. etc.) can have very different background levels, average background counts were determined for each l material encountered in the surveyed area at a location of similar construction and having no known  !

L l radiological history. The construction material background count rates for the proportional detectors .

, averaged 268 cpm for red brick,228 cpm for b.-ick,114 cpm for painted concrete,594 cpm for tile brick, 264 cpm for concrete floor, 73 cpm for carpet, and 31 cpm on metal. The local area  ;

l background count rates ranged from 334 to 594 cpm. Net count rates were determined by subtracting the appropriate material backgrc,und and local area background from the gross count rate for each measurement location. Beta efficiency factors ranged from 0.25 to 0.26 for the gas proportional detectors and from 0.18 to 0.19 for the GM detectors calibrated to Tc-99. The physical window areas for the gas proportional and the GM detectors were 126 cm 2 and 20 cm2 , respectively.

Removable Activity Measurements Removable activity levels were determined using numbered filter paper disks,47 mm in diameter. .

Moderate pressure was applied to the smear and approximately 100 cm2 of the surface was wiped.

Smears were placed in labeled envelopes with the location and other pertinent information recorded.

Exposure Rate Measurements Measurements of gamma exposure rates were performed at 1 m above the surface, using a pressurized ionization chamber (PIC). '

l l.

Fort SL Vraan Nuclear 8tation (621) January 29,1997 B-2 %w,ona(mm m3

Analytical Procedures GIpss Alpha / Beta Smears were counted on a low background gas proportional system for gross alpha, and gross beta activity.

Uncertainties and Detection Limits The uncertainties associated with the analytical data presented in the tables of this report represent the 95% confidence level for that data. These uncertainties were calculated based on both the gross sample count levels and the associated background count levels. Additional uncertainties, associated with sampling and measurement procedures, have not been propagated into the data presented in this report. 1 Detection limits, referred to as minimum detectable concentration (MDC), were based on 2.71 plus 4.65 times the standard deviation of the background count [2.71 + 4.65/BKG]. When the activity )

was determined to be less than the MDC of the measurement procedure, the result was reported as

]

less than MDC. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides in samples, the detection limits differ from sample to sample and instrument to instrument.

Calibration and Quality Assurance Calibration of all field and laboratory instrumentation was based on standards / sources, traceable to NIST, when such standard / sources were available. In cases where they were not available, standards of an industry recogmzed organization were used. Calibration of pressurized ionization chambers was performed by the manufacturer.

l.

Fort St Vrmn Nuclear Stauon (621) . January 29.1997 B-3 w ai, w o m m3

Analytical and field survey activities were conducted in accordance with procedures from the following documents of the Environmental Survey and Site Assessment Program:

Survey Procedures Manual, Revision 9 (April 1995)

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Laboratory Procedures Manual, Revisioa 9 (January 1995)

Quality Assurance Manual, Revision 7 (January 1995)

The procedures contained in these manuals were developed to meet the requirements of DOE Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to assess pr.ocesses during their performance.

Quality control procedures include:

Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.

Participation in EPA and EML laboratory Quality Assurance Programs.

. Training and certification of all individuals performing procedures.

. Periodic internal and external audits.

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I APPENDIX C l

REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS l

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Fat SL Vram Nuclear Stataan (621) . Jamry 29,1997 hMuapveports4wam003

U.S. ATOMIC ENERGY COMMISSION June 1974 REGULATORY GUIDE DIRECTORATE OF REGULATORY STANDARDS REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR WUCLEAR REACTORS A. INTRODUCTION facility requirements are imposed by the Part 50 license and the tecimical specifications to assure that proper Sectica 50.51, " Duration of license, renewal," of 10 surveillance is performed and that the reactor facility is CFR Part 50, " Licensing of Production and Utilization maintain ~1 in a safe condition and not operated.

Facilities," requires that each license to operate a production arxl utilization facility be issued for a specified A possessionenly license pennits various options and duration. Upon expiration of the specified period, the procedures for decommissioning, such as mothballing, license may be either renewed or terminated by the erdombment, or dismantling. The requirements imposed Commission. Section 50.82, " Applications for depend on the option selected.

termination of licenses," specifies the requirements that must be satisfied to terminate an operating license, Section 50.82 providos that the licensee may dismantle including the requirement that the dismantlement of the and dispose of the component parts of a nuclear reacter in facility and disposal of the component parts not be accordance with existing regulations. For research inimical to the common defense and security or to the reactors and critical facilities, this has usually meant the health and safety of the public. This guide describes disassembly of a reactor and its shipment organi.ation for methods arel procedures considered acceptable by the further use. The site from which a reactor has been Regulatory staff for the termirA of operating licenses removed must be decontaminated, as necessary, and for ruclear reactors. The advisory Committee on Reactor inspected by the Commission to determine whether Safeguards has been consulted concerning this guide and unrestricial access can be approved. In the case of has concurred in the regulatory position. nuclear power reactors, dismantling has usually been accomplished by shipping fuel offsite, making tly. reactor B. DISCUSSION inoperable, and disposing of some of the radioactive components.

When a licensee decides to terminate his nuclear reactor operating license, he may, as a first step in the Radioactive components may be either shipped off-site process, request that his operadng license be amended a for burial at an authorized burial ground or secured on the restrict him to possess but not operate the facility. The site. Those radioactive materials remaining on the site advantage to the licensee of converting to such a must be isolated from the public by physical barriers or possession-only license is reduced surveillance other means to prevent public access to hazardous levels requirements in that periodic surveillance of equipment of radiation. Surveillance is necessary to assure the long  ;

important to the safety of reactor cperation is no longer tenn integrity of the barriers. The amount of surveillance i required. Once this possession-only license is issued, required depermis upon (1) the potential hazard to the reactor operation is not permitted. Other activities from health and safety of the public from radioactive material the reactor arul placing it in storage (either onsite or remammg on the site and (2) the integrity of the physical offsite) may be contin.ued. barriers. Before areas may be released for unrestricted use, they must have been decontanunated or the A licensee having a possession-only license must radioactivity must have decayed to less than prescribed retain, with the Part 50 license, authorization for special limits (Table 1).

nuclear material (10 CFR Part, 70, "Special Nuclear Material"), byproduct material (10 CFR Part 30, " Rules The hazard associated with the returned facility is of General Applicability to Licensing of Byproduct evaluated by considering the amount and type of Material"), and source material (10 CFR Part 40, remaining contamination, the degree of confinement of the

" Licensing of Source Material"), until the fuel, remainim radioactive materials, the physical security radioactive components, and sources are removed from provida!',y the confinement, the susceptibility to release the facility. Appropriate administrative controls arxl llfjh^ 3(TR

of radiation as a result of natural phenomena, and the entombment. An appropriate and continuing duration of required surveillance. surveillance pmgram shoukt be established under a possession-only license.

C. REGULATORY "OSITION l c. Removal of Radioactin. Components and

1. APPLICATION FOR A LICENSE TO POSSESS Dismantling. All fuel assemblies, radioactive fluids BUT NOT OPERATE (POSSESSION-ONLY and war.:e, and other materials having activities above i LICENSE) accepted umestricted activity levels (Table 1) shoukt be removed froro the s;te. The facility owner may A request to amend an operating license to a then have unrestricted use of the site with no possession-only license should be made to the Director of requiremen: for a license. If the facility owner so Licensing, U.S. Atomic Energy Commission, desires, the remainder of the reactor facility may be Washington, D.C. 20545. The request shouki include dismantled and all vestiges removed arxl disposed of.

the following information:

d. Conversion to a New Nuclear System er a Fossil

a. A description of the current status of the facility. Fuel System. This alternative, wnich applies only to nuclear power plants, utilizes the existing turbine

b. A description of measures that will be taken to system with a new steam supply syvem. The original prevent criticality or reactivity changes and to ranclear steam supply system shouki be separated fmm nummize releases of radioactivity from the facility, the electric generating system arul disposed n 6 accordance with one of the previous three retirement

c. Any proposed changes to the technical alternatives.

specifications that reflect the possession-only facility status and the necessary disassembly / retirement 3. SURVEILLANCE AND SECURITY FOR TIIE activities to be performed. RETIREMENT ALTERNATIVES WIIOSE FINAL STATUS REQUIRES A POSSESSION-ONLY

d. A safety analysis of both the activities to be LICENSE accomplished and the proposed changes to the technical specifications. A facility which has been licensed under a possession <mly license may contain a significant amount

e. An inventory of activated materials and their of radioactivity in the form of activated and contaminated location in the facility. hardware and stmetural materials. Surveillance and commensurate security shoukt be provided to assure that

2. ALTERNATIVES FOR REACTOR RETIREMENT the public health and safety are not endangered.

a. Physical security to prevent inalvertent exposure Four alternatives for retirement of nuclear reactor of personnel shoukt be provided by multiple locked facilities are considered acceptable by the Regulatory barriers. The presence of these barriers shoukt make it staff. These are: extremely difficult for an unauthorized person to gain access to areas where radiation or contammation levels

a. Mothballing. Mothballing of a nuclear reactor exceed those specified in Regulatory Position C.4. To facility consists of putting the facility in a state of prevent inadvertent exposure, radiation areas above protective storage in general, the facility may be left 5 mR/hr, such as near the activated primary system of a intact except that all fuel assemblies and the power plant,6ouki be appmpriately marked and should radioactive fluids and waste shouki be removed from not be accessible except by cutting of wekled closures or the site. Adequate radiation monitoring, the disassembly and removal of substantial stmetures environmental surveillance, arxl appropriate security and/or shiekling material. Means such as a procedures shoukt be established under a renxte-readout intrusion alarm system shouki be provided possession-only license to ensure that the health and to indicate to designated personnel when a physical barrier safety of the public is not endangered. is penetrated. Security personnel that provide access control to the facility may be used instead of the physical

b. In-Place Entomhmeat. In-place entombment barriers and the intrusion alarm systems.

consists of seahng all the remaining highly radioactive or contamineted components (e.g., the pressure vessel b. The physical barriers to unauthorized entrance into and reactor internals) within a structure integral with the facility, e.g., fences, buiklings, wekled doors, arxl the biological shield after having all fuel assemblies, access openings, shoukt be inspected at least quarterly to radioactive fluids and wastes, and certain selectcxl assure that these barriers have not deteriorated and that components shipped offsite. The structure shouki locks and locking appasatus are intact.

provide integrity over the period of time in which significant quantities (greater tinn Table 1 levels) of c. A facility radiation survey shoukt be performed at radioactivity remain with the material in the least quarterly to verify that no radioactive material is

! Note: Section electronicaHy reproduced from photocopy. C-2 I

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escaping or being transpo ted through the containment health and safety. After the decontammation is barriers in the facility. Sampling shouki be done along satisfactorily accomplished anl the site inspected by the the most probable path by which radioactive material such Comrnission, the Commission may authorize the license as that stored in the inner containment regions could be to be termmated and the facility abandoned or released for l transported to the outer regions of the facility and unrestricted use. The licensee shouki perfomi the ultimately to the environs. <temiwion using the fallowing guidelines:

d. An environmental radiation survey shouki be a. The licensee shou'd make a reasonable effort to performed at least semiannually to verify that no elimmate residual contanitation.

significant amounts of radiation have been released to the environment from the facility. Samples such as soil, b. No covering shoukt be applied to radioactive vegetation, and water shoukt be taken at locations for surfaces of equipment of structures by paint, plating, or which statistical data has been established during reactor other covering material until it is known that operations. contamination levels (determmed by a survey and

<l-m-ted) are below the limits specified in Table 1. In

e. A site representative shouki be designated to be addition, a reasonable effort should be made (and responsible for controlling authorized access into and documented) to further muumize contamination prior to movement within the facility. any such covering.

f. Adnumstrative procedures shock! be established for c. The radioactivity of the interior surfaces of pipes, the notification and reporting of abnormal occurrences drain lines, or ductwork should be determined by making such as (1) the entrance of an unauthorized person or measurements at all traps and other appropriate access persons into the facility and (2) a significant change in the points, provided contanunation at these locations is likely radiation or contamination levels in the facility or the to be representative of contamination on the interior of the offsite environment, pipes, drain lines, or ductwork. Surfaces of premises, equipment, or scrap which are likely to be contaminated

g. The following reports should be made: but are of such size, constmetion, or location as to make the surface inaccessible for purposes of measurement (1) An annual report to the Director of Licensing, should be assumed to be contaminated in excess of the U.S. Atomic Energy Commission, Washington, D.C. permissible radiation limits.

20545, describing the results of the environmental and facility radiation surveys, the status of the facility, and an d. Upon request, the Commission may authorize a evaluation of the performance of security and surveillance licensee to relinquish possession or control of premises, measures. equipment, or scrap having surfaces contaminated in excess of the limits specified. This may include, but is (2) An abnormal occurrence report to the llegulatory not limited to, special circumstances such as the transfer Ope =: ions Regional Office by telephone within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of premises to another licesed organization that will of discoveiy of an abnormal occurrence. The abnormal continue to work with radioactive materials. Requests for occurrence will also be reported in the annual report such authorization should provide:

descded in the preceding item.

(1) Detailed, specific information describing the

h. Records or logs relative to the following items prmuses, equipmerd, scrap, and radioactive contammants shouki be kept and rdamed until the license is terminated, and the nature, extent, and degree of residual surface after which they must be stored with other plant records: contamination.

(1) Environmental surveys, (2) A ddaikd health arxl safety analysis indicating that the residual amounts of materials on surface areas, (2) Facility radiation surveys, togdher with other consklerations such as the prospective use of the premises, equipment, or scrap, are unlikely to (3) Inspections of the physical barriers, and result in an unreasonable risk to the health and safety of the public.

(4) Abnormal occurrences.

e. . 'rior to release of the premises for unrestricted use, the licensee shouki make a comprehensive radiation

4. DECONTAMINATION FOR RELEASE FOR survey establishing that contamination is within the limits UNRESTRICTED USE specified in Table 1. A survey report should be filed with the Director of Licensing, U.S. Atomic Energy If it is desired to terminate a license and to eliminate Commission, Washington, D.C. 20545, with a copy to the any further surveillance requirements, the facility should Director of the Regulatory Operations regional Office I be sufficiently decontaminated to prevent risk to the public havingjurisdiction. The report should be filed at least 30 Note: Section electronically reproduced from photocopy. C-3

days prior to the p'anned date of abandonment. The Regional Office inspects the facility aml verifies survey report shouki: completion in accordance with the dismantlement plan. If residual radiation levels do not exceed the values in Table (1) Identify the premises; 1, the Commission may terminate the license. If posanionenly license under which the dismantling (2) Show that reasonable chott has been made to activities have been conducted or, as an alternative, may reduce residual contamination to as low as practicable make application to the State (if an Agreement State) for levels; a byproduct materials license.

(3) Describe the scope of the survey and the general procedures followed; and (4) State the fimling of the survey in units specified in Table 1.

After review of the report, the Conunission may inspect the facilities to confirm the survey prior to granting approval for abandonment.

5. REACTOR RETIREMENT PROCEDURES As indicated in Regulatory Position C.2, several ahematives are acceptable for reactor facility retirement.

If minor disassembly or "mothballing" is planned, t' tis could be done by the existing operating and maintenance procedures under the license in effect. Any planned actions involving an unreviewed safety question or a change in the technical specifications shouki be reviewed and approved in accordance with the requirements of 10 CFR 5 50.59.

If major structural changes to radioactive components of the facility are planned, such as removal of the pressure vessel or major components of the prn.w; system, a dismantlement plan including the information required by 5 50.82 shouki be submitted to the Commission. A dismantlement plan should be submitted for all the alternatives of Regulatory Position C.2 exces mothbalhng. However, minor disassembly activities may i still be performed in the absence of such a plan, provided l they are permitted by existing operating and maintenance

procedures. A dismantlement plan should include the

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following:

a. A description of the ultimate status of the facility

b. A description of the dismantling activities and the precautions to be taken.

c. A safety analysis of the dismantling activities including any effluents which may be released.

d. A safety analysis of the facility in its ultimate status.

Upon satisfactory review and approval of the dismantling plan, a dismantling order is issued by the Commission in accordance with i 50.82. When dismantling is completed and the Commission has been notified by letter, the appropriate Regulatory Operations Note: Section electronict.lly reproduced from photocopy. C-4

TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS '

NucUde' Average Maximum'd Removable

U-nat, U-235, U-238, and associated decay products 5,000 dpm a/100 cm 2 15,000 dpm a/100 cm2 1,000 dpm a/100 cm 2 Transuranics,Ra-226 Ra 228, Th-230 Th-228, Pa-231, Ac-227, I-125, I-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 l Th-nat, Th-232, Sr-90, Ra-223, Ra-224, U-232, I.126, I-l 31, I-l 33 1,000 dpm/100 cm2 3,000 dpm/100 cm 2 200 dpm/100 cm2 Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above. 5,000 dpm Dy/100 cm2 15,000 dpm Dy/100 cm2 1,000 dpm Dy/100 cm2

'Where surface contamination by both alpha- and beta-gamma-emitting nuclides exists, the limits estsblished for alpha- and beta- gamma-emitting nuclides should apply independently.

"As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instmmentation.

'Measulements of average contaminant should not be averaged over more than 1 square meter. For objects ofless surface area, the average should be derived for each such object.

dThe maximum contaminetion level applies to an area of not more than 100 cm 2, 2

"The amount of removable radioactive material per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbcnt paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efliciency. When removable contamination on objects ofless surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.

Fort st vrmn Nuclear Stahon (621) . January 29, im C-5 a wm_mme

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