Text

Final Status Survey Report for Concrete Rubble in Sub-Area F
	ML19363A016
Person / Time
Site:	07000925; (SNM-0928)
Issue date:	03/10/1998
From:	Jim Larsen; Cimarron Corp
To:	Kenneth Kalman; Office of Nuclear Material Safety and Safeguards
Shared Package
ML19365A023	List: ML19361A179; ML19363A001; ML19363A002; ML19363A003; ML19363A004; ML19363A005; ML19363A006; ML19363A007; ML19363A008; ML19363A009; ML19363A010; ML19363A011; ML19363A013; ML19363A014; ML19363A015; ML19363A016; ML19364A126; ML19364A231; ML19364A232; ML19364A233; ML19364A234; ML19364A235; ML19364A236; ML19364A237; ML19364A238; ML19364A239;
References
	Download: ML19363A016 (42)
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S. JESS Lt\\RSE1"1 VIC'E PlU:lSlDENT March 10, 1998 CIMARRON CORPORATION P.O. BOX 2~S6,

OKLAHOMA CffY, O~LAHOMA 1~125 Mr. Kenneth L. Kalman, Project Manager Facilities Decommissioning Section Low Level Waste & Decommissioning Projects Branch Division of Waste Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Conunission Washington, D_C. 20555-0001 Ref:

Docket No.70-92S; License No.. SNM-928 Final Status Survey Report for Concrete Rubble in Sub-Area "F"

Dear Mr. Kalman:

Cimarron Co:tporation has recently completed Final Status Suivey work on the Concrete Rubble

.~

located in Sub-Area ~-'F". This Final Status Survey work was performed in order to demonstrate that the Concrete Rubble can be unconditionally re]eased under the BTP Option #1 criteria (i.e., 30 pCi/g total uranium;, excluding background) for soils ru:id debris.

The putpose of this letter is to provide the above referenced report to the NR.C staff for review and approval. Please find enclosed three (3) copies of this report for your review and approval. Two (2) copies of this report are for you and your staff and one (1) copy is for Mr. David Fauver. In addition5 one (1) copy ofthis report has also been submitted to both the NRC Docket and to Mr. Louis Carson atNRC Region IV.

Please feel free to contact me if there are any questions or concerns.

. Sincerely,

/)

_J _//

~-~<)/.~.~~~

/1ess Larsen Vice-President Enclosures

)

A SUBS!OIAfW OF KiRH*.MCGEECORPORATlDN

FINAL STATUS SURVEY REPORT FOR CONCRETE RUBBLE

. IN SUB-AREA F for Cimarron Corporation's Former Nuclear Fuel Fabrication Facility Crescent, Oklahoma License Number: SNM-928 Docket No: 70-0925 Prepared for:

Cimarron Corporation Oklahoma City, Oklahoma March, 1998

TABLE OF CONTENTS

TABLE OF CONTENTS 1.. -0 lN'TR.ODUCTION.............................................................. *...................... "'...... -.............,.. "'................... 1

.2.0 BACKGROlJNI) ***...*.**..**.***.**.**,.*. ~~*************~*************~~*-***********~**~***********~*2

2. 1 Phase I Area.........,..........................................,.....

********************************* 3 2.2 Phase II Area............,..,................................. ~..........,................................. 3

2. 3 Phase Ill Area *..............................................,.........,.......... _..................,.... 4 3.. 0 SITE DESCRIPTION........ *****~**"'**** **~-.......... ~...............................................,,.............. 4 4.. 0 FACIL-ITY DESCRIPTION.....,.....................................................,,.............................. 4 5.0 IIISTORY OF SITE OPERATIONS *...*....** ~*********<<********~*-******************************6 6.0 CONCRETE RUBBLE DECO:rv.rMISSIONING ACTIVITIES............. u.-ohhhu*+..***6

6. 1 Identification of Conta:rni.nants..*...... ~..*... ~..................... ~......................,..**..... 6

6. 2 Site BackgrotlD.d Levels..........,........................... ~.......................................... 6 6.2.I Natural Background Radioactivity of Concrete................................................ 6 6.2.2 Establishment of a Background Value for Gross Alpha and Gross Beta Surface Acti v*ity in Concrete............................. _....... _......................................... 7 6.2.3 Soils.......................... -....................................................................... 8 6.2.4 Exposm*e Rates.?,................................................................................... 9 6.3 Characterization Data.........................................................,.........,........... 10

6. 4 Environmental Monitoring Data.........,................................... ~..,................. 10 7.0 FINAL STATUS SURVEY PROCEDURE.... ~.............................................................. 11 7.1 Survey Procedure......,...........,...................,.....,...... ~..............,................. 11 7.2 Exposu:re Rate Measurements.................................................,.,................,... 13 7.3 Environmental Exposure Rate Measurements................................................,.... 13 7.4 Surface Water Sampling...............,***~*.,.................................,..................... 13 7.5 Soil/Sediment Sampling.......................................................,.........,............. 13 7.6 Relationship Between Surface Activity and Concentration............,............,........... 14

7. 7 Guidelin.e Values....................................................................................... 15

7. 7.1 Concentration Guidelines for Concrete....,................................................. 15

7. 7. 2 Exposure Rate Guidelines (External Dose).......................,........ _................ 15 7.7.3 Volumetric Activity of Soils and Sediments............................................... 16 7.8 Equipment Selection....................,.......................... _.................................. 16

7. 8.1 Equipment and Instrumentation.............................................................. 16 7.8.1.1 Micro~R Survey Meter...............................,..................................... I 7 7.8.1.2 Soil Counter (Gamma Spectroscopy).*................. -~........ -~-..............,..... 17

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i

.FINAL STATUS SUR VEY REPORT PHASE U SUB-AREA F CONCRETE RUSBLE

7.9 Procedures/Plans....................................... *-........................................... 18 r----......

7.9.1 Orga:n.ization.......................,.~--*~*******~,................................................ 18

7. 9. 2 Training..........,.........................................,................ _....................... 19 7.9.3 Radiation Protection Program........................................ *********~***........... 19 7.9.4 Cimarron Quality Assurance Program (QAP}............................................ 20 8.0 St.JR.VEY FINDIN'GS..................................................................... 111..................... <<.......... ".....,. 21 8.1 Thermoluminescent Dosimeter (TLD) Exposure Rate Data.................................. 21 8.2 Soil/Sediment Sa.II1ples...... ~.........,,................... ~.. ~.......,................. ~....,.. ~..,. 23

8. 3 Surface Water Samples...,.....................................,............,.......,................... 23 8.4 Micro-R Measmements..........................'................................................... 24 8. 5 Gross Alpha and Gross Beta-Gamma Surface Activity Data......,..... ~................... 25 8.5.1 Gross Beta-Gamma Data...................,...................................................... 25 8.5.2 Gross Alpha Data............................................ ~************************************ 26

8. 6 Calculations........................... ~........................................................... ~...... 27 8.6. l Calculation of Average Concrete Thickness.............. h

............................... 27 8.6.2 Volumetric Concentration Conversion Factor.................

H

............. A............. 27 8.6.3 Volumetric Concentration Calculations.........,........................................... 30 8.6.4 Source Tenn Calculation............................,..,.....,.........,...................... 31

8. 6.5 Pathway Analysis.................... _................... ~.... -*.................................. 31 Appendix I Drawings 95MOST-RF3 95SITE 95M0ST-RF1 98FCONC-O 98_TLD 98FCRSS 98FCRER 98FCRA 98FCRB 98FCRC Appendix II Data Tables Table 1 Exposure Rate-Surface and 1m (Random Sample)

Table 2 Exposure Rate-Surface and lm (All Sampled Grids)

Table 3 Gross Beta~Gamma (Random Sample)

Table 4 Gross Beta-Gamma (All Sampled Grids)

Table 5 Gross Alpha (Random Sample)

Table 6 Gross Alpha (All Sampled Grids)

Appendix III Average Concrete Thickness Calculations Appendix IV '.RESRAD Output FINAL. STATUS SURVEY l<EPCRT ii PHASE IJ SUB-All.EA F CO'NCltETE lt'UBBLE

REFERENCE.

REFERENCES

1.

Chase Environmental Group, Inc. "Radiological Characterization Report for Cimarron Corporation 1s Fonner Nuclear Fuel Fabrication Facility, Crescent, Oklahoma"~ October 1994.

2.

USNRC, Comments dated July 11, 1996 on the Decommissioning Plan for Cimarron Corporation.

3.

USNRC, 11Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Tennination of License for By-Product, Source, or Special Nuclear Material", August 1987.

4.

USNRC, "Branch Technical Position on Disposal or On-site Storage of Residual Thorium and Uranium from Past Operations\\ FR. Vol. 46: No. 205, Page 52061~ October 23, 1981.

5.

Cimarron Corporation Nuclear Materials License, SNM-928 Docket No. 070-00925, issued for possession only March 31, 1982; Amendment No. 10, issued November 4~ 1994.

6.

Cimarron Corporation Nuclear Materials License, S1\\1M~1174, Docket No. 070-1193, terminated February 5~ 1993.

7.

Cimarron Corporation Letter to USNRC, August 20, 1990.

8..

USNRC Letter from Mr. Richard E. Cunningham, Director, Division of Industrial and Medical Nuclear Safety to Dr. John Stauter, Director of Environmental Services, Cimarron Corporation7 dated February 5, 1993.

9.

Chase Environmental Group~ Inc. unecommissioning Plan for Cimarron Corporation's Former Nuclear Fuel Fabrication Facility, Crescent> Oklahoma'\\ April 1995.

10.

Chase Environmental Group, Inc. 'tpinal Status Survey Plan for Unaffected Areas for Cimarron Corporation~s Fonner Nuclear Fuel Fabrication Facility, Crescent, Oklahoma,

October 1994.

11. USNRC Letter from Mr. Michael F. W ebet\\ Chief Low-Level Waste and Decommissioning Project Branch~ Division of Waste Managementt to Mr. Jess Larsen) Vice President Kerr-McGee Corporation, dated May 1) 1995.

12.

Cimarron Corporation, {'Final Status Survey Report, Phase I Areas at the Cimarron Facility, License No_ SNM~928", July 1995.

FINAL STATUS SURVEY REPORT iii PHASE I[ SUB-AREA F CONCRETE RUBBLE

13.

US1\\1RC Letter from Mr. R. A. Nelson,, Acting Chief Low-Level Waste and Decommissioning Project Branch, Division of Waste Management, to Mr. Jess Larsen, Vice President, Cimarron Corporation~ dated April 23, 1996.

14. Chase Environmental Group, Inc., Final Status Survey Plan for Phase II Areas for Cimairon Corporation~s Fonner Nuclear Fuel Fabrication Facility 1

\\ Crescent, Oklahoma, July 1995.

15. USNRC Letter from Mr. Kenneth L. Kalman~. Project Manager, Low~Level Waste and Decommissioning Projects Branch, to Mr. Jess Lars~ Vice President~ Cimarron Corporation, Dated March 14, 1997.

16. US NRC Letter from Mr. George M. McCann, Chief,, Materials Licensing Section to Dr.

John Stauter, Vice President, Kerr-McGee Corporation, dated December 30, 1992.

17.

Chase Environmental Group, Inc. "Final Status Survey Plan for Phase III Area for Cimarron Corporation~s Fonner Nuclear Fuel Fabrication Facility"', Crescent~ Oklahoma, June 1997.

18.

USNRC~ HBackground as a Residual Radioactivity Criterion, for Decommissionint\\

NDR.EG-1501) Draft Report, August, 1994.

/'~ 19. Eichholz1 G. G., et al., RadiatiQn Exposure from..BuUding M_aterials, in "Natural Radiation Environment Ilr" ~ US DOE CONF'-780422, 1980.

20.

Ingersoll, J. G., "A Survey of Radionuclide Contents and Radon Emanation Rates in Building Materials Used in the United States'\\ University of California Lawrence Berkeley Laboratory Report LBL-11771~ 198L

21.

E.W. Abelquist, "Confinnatory Survey for the South Uranium Yard Remediation, Kerr-McGee Corporation, Cimarron Facility, Crescent, Oklahoma/\\ Oak Ridge Institute for Science and Educ.a.tion, November 1995.

22. J.D. Berger, nManual for Conducting Radiological Surveys in Support of License Tenninationu; Draft Report for Comment, 0.ak Ridge Associated Universities, NUREG/CR-5849, June 1992.

23.

American National Standards Institute, '

4Radiation Protection Instrumentation Test and Calibration~ ANSI N323-1978.

24.

USNRC letter from Mr. Ross A. Scarano, Director Division of Nuclear Materials Safety to Mr. S. Jess Larsen! Vice President, Cimarron Corpora.tic~ dated July 31) 1997.

~---------------""-~------------*------

FINAL ST.ATIJS SURVEY REPORT iv PHASE II SUB-AR.EA F CONCRETE RUBBLE

25. USNRC Letter from Mr. Michael F. Weber, Chie£ Low-Level Waste and Decommissioning Project Branch, Division of Waste Management to Mr. Jess Larsen; Vice President, Kerr~McGee Corporation!) dated May 31~ 1995.

26.

USNRC, 4(Scenarios for Assessing Potential; Doses Associated with Residual Activiti'\\

Policy Guidance PG 8-08.

27.

USNRC., Environmental Assessment Associated with the BTP Option #2 Onsite Disposal Cell at Cimarron'\\ 1994.

PINAL STATUS SURVEY REPORT V

PHASE II SUB*ARl;A F CONCRETE RUBBLE

TEXT

FINAL STATUS SUR\\lEY REPORT FOR DECOMMISSIONffiG

~..

CIMARRON FACILITY CONCRETE RUBBLE IN SUB-AREA "F

1..0 Introduction This Final Status Survey Report is submitted by Cimarron Corporation to the Nuclear Regulatory Commission {NRC) to release for unrestricted access concrete rubble located in and near the drainage areas and the discharge area of Reservoir #2, which is located in Sub-Area "F' of the Cimarron site. Sub-Area "F' is one of the five Sub~Areas within Phase IL Sub-Area "F" is shown on Drawing No. 95M0ST-RF3 {Appendix I)) and includes affected and unaffected areas that have been surveyed as part *of the ongoing site decommissioning process. This report provides the infonnation and justification for leaving the concrete located in Sub-Area '

4F 1 in-place. The concrete was placed as rip-rap to correct erosion problems associated with the Sub-Area "F drainage ways. A drawing showing the surface topography for the site is provided in Appendix I (95SITE).

This report documents the survey and sampling data. obtained for the concrete and establishes the basis for unconditional release of the concrete rubble.

Section 2.3 of the Cimarron Decommissioning Plan1 provides the proposed criteria for leaving the concrete in-place.

In addition, there have been several NRC comments and Cimarron responses to comments regarding concrete in drainage areas.

In this report, Cimarron has incorporated NRC recommendations (see NRC Comment #18 of the ~~c Comments dated July 1 t 1996 on the Decommissioning Plan for Cimarron Corporation"2) to

consider volumetric concentration averaging as a method for unconditional release of the concrete.

The concrete rubble met all of the applicable surface contamination criteria for unconditional release when it was relocated to Sub-Area "F drafoage ways for erosion control. The concrete rubble was placed into Sub-Area "F~' since decommissioning activities commenced in 1976, and was subject to various release criteria in effect which depended on the time of release. The concrete rubble originated in on-site buildings and structures undergoing decommissioning and was surveyed for alpha contamination, and in some cases for beta-gamma contamination, before it was used for erosion control in drainage areas north of Reservoir #2 and northeast of Burial Ground #1 in Sub-Area "F"~. However, the surface contamination release criteria in effect during the early phases of facility decommissioning were not as restrictive as those currently in place and ranged as high as 25,,000 (maximum) dpm/100 cm 2 gross alpha (per Annex A to License Sl\\1M-928, Section 3.4, Revision dated Au.gust 30~ 1976). In addition, practices which were approved by the NRC and in effect during the early phase of facility decommissioning did not entail surveys for beta.:.gamma activity prior to release when the contaminant was known or believed to be pure enriched uranium. Consequently, surveys performed more recently have identified levels of gross beta-gamma activity and gross alpha activity which exceeds the 1987 NRC unconditional release criteria contained in "Guidelines of Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Tennination of Licenses for Byproduct,

. FINAL sr A ros SURVEY REPORT PHASE 11 SUB-AREA "'F" CONCRETE RUBBLE

Soill'.ce, or Special Nuclear Materiais

3 The co11erete rubble in Sub-Area c~F' is representative

,,--..... '" of concrete from all areas of the buildings and structures at the facility and also contains concrete from all facility areas.

The proposed release criteria for the concrete is equivalent to the 1981 BTP4 Option #1 criteria for soils, which is 30 pCi/g for enriched uranium. This criteria is appropriate for concrete rubble and debris due to the similarities in dose pathways. Concrete rubble is~ in many respects, less likely to contribute dose to members of the public due to limited accessibility and a decreased probability of mechanical or physical dispersion.

This report also demonstrates that the estimated doses due to the pathways of concern for the concrete rubble are significantly less than the 1981 BTP Option #1 limit concentrations for enriched uranium, and are insignificant in comparison to exposures resulting from natural background.

Cimarron Corporation intends to leave the concrete in place by demonstrating that the risks from

the concrete remaining in place are insignificant. In addition, the safety hazards and costs associated with removal of the concrete from the drainage areas are significant. This is due to the location of the concrete in drainage areas) the random manner in which it was placed into the drainage are~ and the physical hazards from rebar protrusions and unstable~ irregular surfaces which could result in falling or tripping. Finally, the concrete continues to serve the intended useful purpose of preventing unnecessary erosion in the drainage and spillway areas~

This report includes a discussion of the final status survey performed to more precisely define the

-~' extent and magnitude of residual contamination present in the concrete located within Sub-Area

F". The final status survey was conducted in order to demonstrate that the guideline values for the Cimarron site have been met. TI1e results of the Sub-Area "F" concrete rubble Final Status Survey are presented in this Report, and indicate that the estimated activity of total uranium contained in the concrete rubble is 4.6 millicuries. The maximum projected dose rate to the hypothetical resident from residual activity contained in the concrete rubble was calculated to be approximately one millirem per year based upon the RESRAD computer code.

2.. 0

Background

Cimarron Corporationl a subsidiary of Kerr... McGee Corporation, operated nvo plants near Crescent7 Oldahoma, for the manufacture of enriched uranium and mixed oxide reactor fuels.

The 840--acre Cimarron site was originally licensed under two separate SJ\\cTM Licenses. License SNM.,.928 5 was issued in 1965 for the Uranium Plant (U-Plant) and License SNM-1174{i was issued in 1970 for the Mixed Oxide Fue! Fabrication (MOFF) Facility. Both facilities operated through 197 5, at which time they were shut down and decommissioning work was initiated.

Decommissio11ing efforts at the MOFF Facility were completed in 1990 and Cimarron Corporation applied to the NRC on August 20, 1990\\ to terminate License SNM-1174. After confirmatory surveys, the NRC terminated the MOFF Facility License, SNM.. 1174~ on

,.~ February 5~ 19938*

FtNALSTATUSSURVtYREPORT

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PHASG 11 SUB"AREA ".r* CONCRETE RUBBLE.

~

Decommissioning efforts involving characterization:- decontamination and decommissioning for the 840-acres, licensed under SNM-928, were initiated in 1976 and are still ongoing. The final objective of the decommissioning effort is to release the entire 840-acre site for unrestricted use.

Based upon historic knowledge of site operations and the characterization work completed to date, Cimarron Corporation completed and submitted in October 1994. the Cimarron Radiological Characterization Report.9 As discussed in this report, the site has been divided into affected and unaffected areas. The affected and unaffected areas are shown on Drawing No.

95MOST-RF3, included in Appendix I. For the Final Status Survey Pfan> the entire 840-acre site has been diVJded into three major areas which contain both affected and unaffected areas. Each of these three major areas.are also shown on Drawing No. 95M0ST--RF3 and are designated by Roman Numerals I, II, and III (herein referenced as Phases I1 II, and ill). These three major areas are then further subdivided into smaller Sub-Areas (i.e. A, B~ C, D, etc.).

2.1 2.2 Phase I Area As presented in the Cimarron Decommissioning Plan, 1 the Final Status Survey Plan (Phases l, ll and III) was discussed in general tenns, with the understanding that each of the three phases would be submitted to the NRC under separate cover for approval. The Final Status Survey Plan for the first of these three phases (Phase 1 1°) was approved by the NRC via letter dated May 1, 1995.11 The Final Status Survey Report1:i for Phase I was submitted to the NRC and confinnatory sampling for the Phase I areas has been completed by the Oak Ridge Institute for Science and Education (OR1SE). Cimarron Corporation received license Amendment #13 from the NRC to release this area from Sh'M... 928; the amendment was forwarded by letter dated April 23, 199613*

This amendment reduced the licensed facility acreage from 840 to 152 acres.

Phase II Area The area designated as Phase II on Drawing No. 95M0ST-RF3 (Appendix I) contains both affected and some contiguous unaffected areas, and represents approximately 122 of the remaining licensed 152 acres.

The Final Status Survey Plan for Phase. II was submitted to the l\\1RC in July 1995 14 and approved by the NRC on March 14, 1997'5*

Phase II includes Sub-Areas F> G, H, I and J. Included within Phase II are Burial Area #1 which was released in December 1992 by the NRC16, backfilled with clean soil> and seeded. Also included in Phase II are the East and West Sanitary Lagoons, the MOFF Plant Building exterior and yard area, the Emergency Building, the Warehouse Building

{Building #4) and surrounding yard, and numerous drainage areas.

Cimarron has substantially completed the remediation of each Sub~Area and final status surveys are currently underway. In general, Sub-Area "F is located north of Reservoir #2 and includes the roadway from the northern end of Reservoir #3 to the northern end of Resetvoir #2. The concrete rubble within Sub.... Area UF is located on the berm area and in the drainage area to the north of Reservoir #2 and also alongside a drainage to the

-.RNALST ATUS SURVEY REPORT 3

PffASE Ii SUB~AkEA "P' CONCRETE RUBBLE

2.3 northeast of Burial Area #1. The concrete rubble addressed in this report includes an area of approximately 0.3 acres of the entire 17 acre area within Sub--Area,;

4F". The final status survey for the remaining acreage has been completed and is being assembled for submission under a separate report.

Phase III Area The Phase III area survey is the last phase for completing the final status survey for the entire Cimarron site1 and represents approximately 30 acresr This area is designated as Phase III on Drawing No. 95M0ST-RF3. The Final Status Survey Plan for release of this area from the site license, has been submitted to the NRC17 for approval. The Phase III area includes the Uranium Processing buildings and yard are~. Burial Areas #2 and #3, the New Sanitary Lagoon, the New On-site Disposal Cell (Burial Area #4}, and the Five Former Waste Water Po.nds. These five fonner waste water ponds consist of Uranium Waste Ponds #1 and #2, the Plutonium Waste Pond~ the Uranium Emergency Pond~ and the Plutonium Emergency Pond.

3.0 Site Description The Cimarron Facility is located in Logan County~ State of Oklahoma~ on the south side of the Cimarron River approximately 05 miles north of the intersection of Oklahoma State Highways

,-.,. #33 and #74. Figure 3.1 shows the site location. The 840-acre site (the licensed portion was recently reduced to 152 acres) is located in an area of low, rolling hills and incised drainages.

Local elevations range from about 940 foet along the river to 1,010 feet Mean Sea Level at the plant. The county is primarily rural with an economy primarily based upon agriculture and ranching. The entire site is owned by Cimarron Co!J)oration, a wholly owned subsidiary ofKerr-McGee Corporation.

In general~ Sub~.Area Hpt, is located north of Reservoir #2 and includes the roadway from the northern end of Reservoir #3 to the northern end of Reservoir #2. The concrete rubble v'.-1thin Sub-Area HF'i is located on the benn area and in the drainage area to the north of Reservoir #2 and also alongside the drainage to the northeast of Burial Area #1.

4.0 Facility Description License SNM-928 was originally issued in 1965 to Kerr-McGee Nuclear Corporation for the manufacture of enriched uranium reactor fuels. The Uranium Plant (U-Plant) was constructed to be a complete nuclear fuel service facility. Initial equipment provided for the production ofU02, UF.i, uranimn metal and the recovery of scrap materials from facility operations. In 1968 the plant was expanded by increasing the U02 and Pellet facilities through the installation of another complete production lin.e for the fabrication of fuel pellets. In 1969 fabrication facilities were

,,.--...___ added for the production of fuel pins. In 1970 facilities were added for the production of the fuel FINAL STATUS SURVEY REPORT 4

PHASE !I SUB-AREA "F' CONCRETE RUBBLE

/"~-.

1.

l I

L 1 ro ~~rr:.;~

,._.... --- ~

................... ~

FlNAl. STATUS SURVEY ltEPORT PHASE 11 S'OB*AR.E.A 'T-" CONCRETE RUBFJLE 11

I Neri ro SCAl.Z N

i t~ T Cimarron Corporation Crescent Oklahoma Facility Lccaticn Map FIGURE3.1 5

elements. Equipment initially installed for the recovery of enriched scrap material was not used after work performed under a scrap recovery contract was completed in 1970. AH equipment

'~ utilized in fuel production activities has been either decontaminated and removed from the site for salvage or packaged and transported offsite for disposal at a commercial LLRW facility.

5.0 History of Site Operations The Cimarron Facility was originally licensed under two separate licenses. License SN1Yf-928 was issued for the U-Plant Facility and License SNM.. J 174 was issued for fue MOFF Facility, Both facilities tenninated production operations in 1975. Decontamination and decommissioning of the MOFF Facility was completed by 1990, and the license was terminated by the N°'R.C in 199:3. The U-Plant Facility decommissioning is in progress and nearing completion. A complete history of site operations can be found in both the Characterization Report9 and the Decommissioning Plan 1

6.0 Concrete Rubble Decommissioning Activities The purpose of this section is to discuss the status of the on-going site decommissioning activities related to the concrete rubble located in Sub-Area ~'F'} and to present the radiological criteria and guideline values utilized throughout the remediation and final status survey.

,..,...--.*, 6 l Identification of Contaminants Based upon the knowledge of past site operations, the results of numerous characterization efforts to date, and other independent characterization efforts by regulatory agencies and their respective subcontractors1 the possible radiological contaminants within this Sub-Area have been detennined to consist of U-234, U-235 and U.-238. The uranium is comprised of enriched forms, with an average emichment above the naturally occurring level.

Toe average U~235 enrichment at Cimarron has been previously established as approximately 2.7 weight percent.

Thorium:> although not considered a contaminant of concern for the concrete rubble, has been included in the soil and sediment analysis for comparison with background levels found in other areas of the facility. Thorium was not used or processed at the facility.

6.2 Site Background Levels 6.2.1 Natural Background Radioactivity of Concrete Concrete contains naturally occurring elements from the earth which emit radiation.

NUREG-1501, "Background as a

Residual Radioactivity Criterion for Decommissioning" 1\\

gives the typical radionuclide content for concrete used in the

~iATUSSU_R_V-EY_RE_PO_R_T--*~-----*---*-----*~-_.._---.

6 PHASE I! SUB~AREA **r* CONCRETE. RUBBLE

United States. Table 23 of NUREG-1501 gives the range of U-238 as 19-89 Bq/kg

/~

(0.5-2A pCi/g), as reported by Eichholz19. Assuming radiological equilibrium between the Uraniumu238 and it's radioactive daughter products through Ra-226, as suggested in NUREG-1501, the total uranium concentration (Le.Cf U-234 + U-235 + U-238) would range from 38~178 Bq/kg (LO to 4.8 pCi/g). In addition to uranium)' Table 2.3 of NUREG-1501 states that concrete contains Th-232 (15 to 118 Bq/kg or OA to 3.2 pCi/ g) and K-40 (262 to 1I47 Bq/kg or 7.1 to 31 pCi/ g).

Ingersol120 collected ordinary concrete samples from around the country. These data are presented in Table 2.7 of NUREG-1501 and show a range for U-238 of 8 to 38 Bq/kg (0.2 to 1.0 pCi/g). Assuming radiological equilibrium between the U-238 and it's radioactive daughter products through Ra-226~ the total uranium concentration (i.e., U-234 + U-235 + U-238) would range from 16-*76 Bq/kg (0.4 to 2.1 pCi/g).

Ingersoll also measured similar variations in concentrations of Th-232 and K 40.

A representative sample of <<background" concrete was collected from the unaffected floor area under the former facility manager~s office. This sample is representative of

~~background" concrete due to the fact that it: 1) Was poured at about the same time as the concrete rubble in Sub-Area "F"; 2) Was located in an office area of the facility that was never contaminated; and"t 3) Had been covered with linoleum tile prior to use of the facility for uranium processing. This background sample* was analyzed by an outside independent laboratory and was reported as containing 1.1 +/- 0.4 pCi/g U-234,

.-~-

<0.2 pCi/g U-235, and 0.4 +/- 0.2 pCi/g U-238. The sample was also analyzed using the on-site soil counter which produced a total uranium concentration of 5. 7 +/- 1 _ 3 pCi/g.

Conversion of the onsite counter results to natural uranium, using the conversion factor of 0.67/1.5 as described in Section 6.2~3, results in a total. uranium concentration of 2.:5 pCi/g~

This confirms that tlris concrete sample is similar in concentration to background soils and is within the range of activities listed in NUREG-1501 (after consideration of measurement uncertainty).

Concrete, such as that in the Sub-Area "P; drainage areas, contains naturally occurring radioactivity such as Uranium1 Thorium and Potassium-40 which affect instrument measurements and laboratory analysis. The natural background concentrations from activity due to total uranium in concrete are therefore subtracted from counting results to obtain the contribution of activity which resulted from facility operations. The total uranium background for concrete which will be used in this report is LS pCi/g, based upon the sample collected at the facility and tlle literature which has been cited.

6.2.2 Establishment of a Background Value for Gross Alpha and Gross Beta~Gamma Surface Activity in Concrete The average background for gross beta-gamma surface activity in concrete was determined using data gathered during the survey of the concrete rubble as well as data from the representative sample of ~.. backgroundn concrete as described in Section 6.2.1.

F'lNALSTA'fl:S :-.iL:RVEY REJ>ORT 7

PHASE H SUl3-ARBA "F" CONCRETE RUBBLE

In order for data to qualify as representative of background, data from each selected 5m x

.~*-.

5m grid area had to have gross beta-gamma levels (average) of less than two times background (for concrete). Grid areas were initially selected by utilizing the data froin areas with gross alpha survey results of less than 100 dpm/100cm2 (maximum). In addition, the. reported exposure rate at both the surface and at a distance of one meter for the grid area could not exceed 10 µR/h and could not be greater than 1 µR/h above the "one meter measurement location at any point within the grid area. Data utilized in the determination of gross alpha and gross beta-gamma surface activity background are summarized in Tab le 6, 1.

Table 6.1 Gross Alpha and Gross Beta Surface Activity Background Data Ave. Alpha Max.Alpha Ave. Beta t Max.Beta Surface f lm Location Dpm/100cm2 dpm/100cni1 dpm/1 O()cm:).

dpm/100cm1

µR/h"*

µRlh....

,r----.

Grid #6 10 40 1045 1650 7

8

~Grid#?

---10*

278 -~ '--*""-

6 40 1S84 7

~

Grid#8 10 40 924 1397 7

7 c-Grid #9 10 80 I

557 935 6

6 Grid.#21 50 80 1199 3212 6

6

.--Grid #33 5

20 889 1386*1 7 7

Grid#103 10 40 860 1.353 7

7

~

Representative "ba~kground" 60" 60" 649.

649" 5"

s*

Aver.age 21 50 800 1520 6.5 "Core sample-measurements taken at the surface of the area to be cored prior to sampling~ As this area is within a building over a concrete slab, the ambient background exposure rate* was lower.

0 Exposure rate measurements include background contribution. All other measurements are net (instrument background subtracted).

6.5 Based upon Table 6.1, the average background for gross alpha surface activity was established as 21 dpro/100 cm2. The gross alpha background activity is presented for

reference purposes only.

The data and evaluations presented in this report did not utilize the gross alpha background for subtraction as the value was insignificant to the overall conclusions made in the report. Thus, the data tables and drawings reflect the gross alpha activity, including background.

The average background for gross beta surface activity was established at 800 dpm/100 cm2* The gross beta background was subtracted from the surface activity data prior to calculation of the volumetric activity concentration of total uranium..

6_2_3 Soils Natural backgroundlevels for uranium in soil have been established through numerous measurements by Cimarron personnel utilizing the on~site soil counter and through fINAL STATUS SUH'EY REPORT PHASE ll SUD*I-.R.EA "F" CONCRETE RUBBLE I

I

independent laboratory analysis.

Analytical results from Cimarron Corporation 1s environmental sampling program are reported to the 1{RC annually in Environmental Reports. These reports provide sample analysis results for soil samples collected from numerous off... site locations which are representative of background in surrounding soils.

Cimarron personnel collected and analyzed 30 surface soil samples from the perimeter of the Cimarron site during the first quarter of 1995 to further validate background levels.

Total uranium ranged from 2.3 pCi/g to 6.6 pCi/g, with the average being 4.0 +/- 2.6 (2o) pCi/ g. These values were obtained as a result of using the Cimarron on-site soil counter.

The on-site soil counter is calibrated to assume an enrichment of 2. 7 weight percent as this is the average enrichment found throughout the site.

\\Vhen a correction factor (0.67/1.5) is applied to these results to convert the values from. an assumed 2.7 weight percent enrichment to a 11atural enrichment, the converted results ranged from 1.0 pCi/ g to 2.9 pCi/g with an average of 1.8 +/- 1.0 (2cr) pCi/g total uranium. Based upon these results, the average value of 4 pCi/g total uranium for background was used when the soil sample analytical results were compared to guideline values.

6.2.4 Exposure Rates Background exposure rates have been established at the Cimarron site by taking micro-R readings and pressurized ion chamber (PIC) readings at off-site sample locations in addition to Cimarron site areas which are unaffected by past operations. Site background exposure rates of approximately 7-10 µR/11 have been observed in background areas by Cimarron personnel utilizing a Ludlum Micro-R survey meter, and have been used in past reports to the NRC, Site background exposure rates of approximately 7-10 µR/h have also been determined by ORISE personnel utilizing similar instrumentation. In addition, site background exposure rates were measured by ORAU (now ORJSE) personnel utilizing a PIC21, and were determined to be 9 to 10 µ.Rlh.

Thus, depending upon instrumentation utilized, the background exposure rate at the Cimarron site ranges from 7 to 10 µRlh.

Cimarron personnel performed exposure rates measurements at background. locations in Sub-Area "F') in 1995 using a Micro-R meter.

Confirmatory measurements were obtained at the same locations in 1997 using a Reuter~Stokes PIC. These data are tabulated below in Table 6.1. The average background as measured using the Micro... R meter was 7.6 +/- 1.3 (2o') 11.Rlh, and is about 15 percent less than the average for the PIC measurements of 9.0 +/- 1.1 (2o) µR/h. These differences are not significant and indicate good agreement between the Micro-R measurements and the PIC measurements.

Cimarron will continue the use of 7-10 µR/h as representative of background exposure rates for Micro-R measurements in accordance with past reports.

FINAL ;;/-\\ITS Sl'J,\\'EY REPORT 9

PBASE l1 SUB-AREA *r CONCRETE RUBBLE

6.3 6.4 TABLE6.2

~

l Grid Location Sample ID No..

UAF-BKG-1

---"s19W-81N UAF.. BKG-7 1600E-120N UAF-BKG-11 840W-700S

-* UAF-BKG-13 840W-288S

-UAF.. BKG-16 i

808W~282S UAF-BKG-19 640W-700S UAF-BKG-23 1610E-300S UAF-BKG-25 1610E-69N 1610E-469N I UAF-BKG~27

- ul\\F-BKG-28 --*

1610E-634N I

AVERAGE Characterization Data Micro-R Reading

(µR/h) 9 I

-7 I

8

~*

8 9

5 6 ~--*--

7 eading R/h)

PICR

(µ

9. 8 7.6 I

9.

2_-~-

I

.8

-i 9

9 1

7 0~ l

.8

__.......;--.~--

7.6

7.

~

r"__...........,_ ________

8

.6 8

9 7.6 +/- 2.7 (2cr) 1 9.. 0+/-2

.3 (2cr)

As noted earlier, the Cimarron site has been subdivided into survey units. These units are natu.rally distinguishable or have a common history of use, characterization and decommissioning activities. Throughout most of the decommissioning process at the Cimarron site} a unit was characterized, remediated (if required)~ and resurveyed. The description of the decommissioning activities and the final status survey data were then submitted to the NRC for review and approval. After review of the submittal, the NRC either released the unit and/or contracted with ORISE (previously ORAU) to perform a confirmatory survey. Based upon the ORISE confirmatory survey (if requested by the NRC}, the NRC either would release the unit or require additional remediation. The concrete rubble within Sub-Area F" is one such unit.

Cimarron persormel have completed the characterization and final status survey of the concrete rubble.

Environmental Monitoring Data The concrete rubble in Sub-Area "F" is near one location where envirorunental surface water monitoring has been performed. in accordance with the Cimarron Environmental Sampling Program. All analyses for samples collected from this location are performed by an independent off-site laboratory.

The surface water monitoring at this location (location #1205) consists of an annual sample from Reservoir #2, which is upstream from the concrete rubble. The location of surface water sampling location #1205 is shown in Drawing No. 95M0ST-RF1 (Appendix I). Since 1986, Gross a concentrations at this location have been reported as less than 10 pCi/1,, and _gross '3 concentrations have been less than 20 pCi/L Total U has FINAL STATUS SURV.EY RE.PORT PHASE lI SUB~AREA "F CONCRETE RUBBLE

been reported as less than 0.005 mg/1 over the same period. The laboratory reported 0.5

-~-

pCi/1 U-234, <0.1 pCi/1 U-235, and 0.2 pCi/1 U-238 for surface water location #1205 in 1997 ~ Based upon the historical environmental data, Reservoir #2 has not oeen affected by facility operations or decommissioning activities.

The final status survey for the concrete rubble also included the collection and analysis of additional water samples from the stream originating from Reservoir #2. These results are discussed in Section 8.3 ohhis Report.

7.0 Final Status Survey Procedure The purpose of this section is to describe the methodology utilized for the collection of the Final Status Survey data. The final status survey data will be used to demonstrate that the applicable radiological parameters (i.e~., guideline values) have been met and that the concrete rubble in Sub-Area 1-,pn can be released from License SNM-928. Due to the nature and location of the material and the physical limitations involved with the performance of surveys, Cimarron Corporation developed innovative calculational techniques to demonstrate compliance with the release criteria+

7.1 Survey Procedure A 5m x 5m grid system was established for the concrete rubble. The grid system was established such that each grid area could be easily relocated in the future for additional survey work and/or confirmatory surveys. The concrete grid system can be described as consisting of three areas, consisting of: 1) the main body of concrete in the drainage area north of the reservoir {Grid #'s 20 through 122); 2) the concrete along the west bank of tlle spillway area (Grid #'s 1 through 19); and, 3) the concrete west of the main drainage area, northeast of Burial Ground #1 (Grid ti's 123 through 134).

Grid locations are shown on Drawing No. 98FCONC-O.

The concrete was then further subdivided into those grids containing z85 % concrete surface area and grid areas containing significantly less concrete.

A computer generated random sampling plan representing 32 of the grids containing significant areas of concrete was developed.

The random sample p]an also contained those accessible areas known or suspected to have the highest gross beta-ganuna surface contamination. The random sample plan thus should represent a worst case average for the concrete. The random sample represents 24% (32 of the 134) 5m x 5m grid areas containing concrete rubble. However, since the random sample was selected from grid areas containing > 85% concrete, the random sample should. generate data which represents even a higher percentage of the surface area for the concrete rubble as a whole. The grid.areas representing the random sample are also shown in Drawing No..

98FCONC-O {Appendix I).

J*lNAL STATUS SURV£Y REl10RT 11 PHASE ti SUB-AREA "F" CONCRETE RUBBLE.

Each randomly selected 5m x 5rn grid area was then 100 % scanned, to the extent practicable, for gross beta-gamma activity to identify any elevated areas of activity.

"'~-.

Any location within each grid that exceeded 5 t 000 dpm/ 100cm2 was documented on survey forms. The 5j000 dpm/100cni2 cut-off value was chosen because it represents the unconditional release surface activity criteria (average) for uranium. -Supplemental gross beta-gamma measurements were also collected for each of the Sm x 5m grid areas containing concrete along the west bank of the spillway t as well as at other locations, as shown in Drawing 98CONC-O (Appendix I).

Gross alpha scans to identify elevated locations were not performed due to the fact that the concrete had been previously surveyed and released (all gross alpha was assumed to be due to uranium). Additionally, gross alpha scanning over rough surfaces has not proven to be an effective method due to the effects of geometry (source to detector distance) and attenuation (shields the alpha particles before they are able to reach the detector). The survey data for areas containing elevated measurements of gross beta-gamma activity versus the gross alpha measurement l'esults provides evidence of this (See Tables 3 through 6 in Appendix II). It was therefore determined that the *most effective method for locating the elevated areas was through the use of gross beta-gamma scans.

A lm x 1m grid was established surrounding any Jocation within a selected 5m x Sm grid area which exceeded 5,000 dpm/100cm2 beta-gamma (i.e.1 hot-spot areas).

Surveys or scans were then performed in each lm x lm grid area as described below.

The average gross beta-gamma dpm/100cm2 was measured.

The maximum gross beta-gamma dpm/100cm2 was measured.

The average gross alpha dpm/lOOcm.2 was measured.

The maximum gross alpha dpm/100cm2 was measured~

A 1-1R/hour reading at the surface and at one meter above the surface directly above the location with the highest beta-gamma activity was measured.

In addition, a "representative" lm x lm survey area was selected from each of the selected 5m x 5m grid areas. The purpose for the representative area was to establish survey data that represented the surface areas within each 5m x 5m grid that did not contain flllY locations with elevated surface activity (i.eT, above the 5 1000 dpm/100cm2 i~hot... spof' criteria). The methodology utilized i11 locating these representative areas was to locate them in areas that were representative of the remainder of the concrete surfaces within the 5m x 5m grid area, preferably at the center of the grid area, but not in an area previously determined to have elevated readings above 5,000 dpm/100cm2 beta-gamma.. Measurements were performed in each of the representative grid areas as described above for the "hot-spot areas exceeding 5,000 dpm/100 cm2*

In addition to the 32 random survey measurements~ Cimarron persormel performed final status surveys on 34 grid areas to supplement the random sample data and to provide assurance that the random sample survey data were representative.

The FINAL ST A TUS SURVEY REPORT 12 PHASE H SUB*AREA "F CONCRETE RUBBLE

7.2 7.3 supplemental surveys were performed in exactly the same manner as the random sample surveys.

Surface activity measurements for gross beta-gamma were obtained using a Ludlum Model 2221 with a Model 43-68 gas proportional probe, or equivalent (e.g., Ludlum Model 43-89).

Gross alpha surface scans were obtained using an Eberline Model PRM-6 with an Eberline Model AS-15 probe~ MicrowR measurements were obtained utilizing a Ludlum Model 19.

The top surface and any other accessible concrete surfaces were surveyed. This intent of the survey was to be as representative of the concrete rubble as a whole, through the use of statistically based sampling techniques and the application of reasonable calculational assumptions.

Removable contamination measurements were not obtained due to the fact that the concrete has been in the drainage area for over 10 years. It was determined that any potentially removable contamination remaining would have been removed by environmental interactions.

Monitoring of soils and sediments downstream of the concrete provided confirmation that removable contamination has not contributed to natural concentrations of uranium in downstream soils and sediment.

Exposure Rate Measurements Exposure rate measurements were obtained at the surface of the concrete and at one meter above the concrete surface utilizing a sodium iodide based Ifiicro-R meter, or equivalent~ These :rne.asurements support the position that the concrete rubble does not pose any significant external dose hazard and that exposures are similar to background.

Environmental Exposure Rate Measurements Three environmental therruoluminescent dosimeters (TLDs) were placed near and above the concrete rubble to support the results of previous exposure rate measurements and to demonstrate that there are not any seasonal or long-term upward trends in exposure rates. Data from these TLDs are compared to site background area TLD measurements.

7.4 Surface.Water Sampling 7.5 Two surface water samples were collected from areas where surface water is normally available. These. surface water samples were collected up-gradient and down-gradient from the concrete rubble area. The water samples were analyzed for isotopic uranium and thorium at an independent off-site analytical laboratory.

Soil/Sediment Sampling Thirteen soil/sediment samples were collected from the concrete rubble area to demonstrate that the concrete is not contributing to levels of uranium and thorium HNAL STATUS SURVEY RE.FORT 13 PHASE II SUB-AREA "?" CONCRETE RUBBLE

.~-.

7.6 present in the sediment. One soil/sediment sample was collected up-gradient of the concrete, two samples inunediately down-gradient in locations likely to collect sediments, and ten samples from soils and sediments beneath or within the concrete rubble. All samples were analyzed for total uranium and thorium activity utilizing the Cimarron soil counter. One sample collected down-gradient from the concrete rubble was also sent to an independent off-site laboratory for analysis.

Relationship Between Surface Activity and Concentration A special study was conducted to determine the depth profile of the contamination in the concrete and for the purpose of establishing a relationship between gross beta-gamma surface activity (measured in dpm/100cm2) and uranium concentrations (measured m pCi/g). A brief outline of this study is provided below:

1.

Concrete rubble slabs with the following approximate gross beta-gamma surface activity were selected:

2.

Background (~ l, 000 dpm/ 100cm2 average beta-gamma)

~ 5,000 dpm/ 100cm2 average beta-gamma

~ 20,000 dpm/100cm2 average beta-gamma The following measurements were then performed on each piece of concrete rubble selected for the s~dy~

Exposure rates at the surface and at 1 meter above the surface utilizing micro-R meter;

Gross alpha dpm/lOOcm 2

- one measurement for each probe sized area located inside the area to be scabbled;

Gross bet.a-gamma dpm/100cm2 - one measurement for each probe sized area located inside the area to be scabbled; Hot~spot - all of the measurements described jn #2 performed at the hot-spot.

3.

The thickness of the concrete slab was measured.

4.

The elevated area on the concrete rubble was scabbled to remove a layer of residual contamination.

5.

Concrete scabble dust was collected and mixed thoroughly.

A sample was obtained and analyzed using the on-site Soil Counter and/or an independent off-site laboratory.

6.

All measurements were recorded, including the estimated thickness of the scabbled area.

FI.NALSTATUS SURVEY REPORT 14 PHASE 11 SUB~AR.EA F" CONCRETE RUBBLE

1/.,,,--...,.,.....,.

7.

Steps #2 through #6 were repeated until measurements of the scabbled area indicated that gross beta-gamma activity was reduced to less than twice background.

7. 7 Guideline Values The radiological guideline values discussed in this section are utilized for comparison to the final survey data.to verify that the concrete rubble in Sub-Area "F can be released from License SNM-928.

7. 7.1 Concentl'ation Guidelines for Concrete Draft NUREG/CR-584922, Section 242, states that "Volume concentration guideline values, which apply to soil, induced activity, and debris, are expressed in terms of activity per unit mass [typically picocuries per gram (pCi/ g). J,.,. Cimarron Corporation has established the release guideline values for concrete in Sub-Area <<p" in accordance with the Branch Technical Position24 (BTP) Option #1.

The BTP Option #1 criteria is 30 pCi/g total uranium (enriched), above background.

Due to the physical obstacles and potential safety hazards associated with monitoring the concrete rubblei it was necessary to develop new calculational methods for determining volumetric concenn*ations. The overall objective of the survey effort was to demonstrate that the release guideline values were complied with.

Due to the nature of the random sampling performedt the methods of draft NUREG/CR-5849 for "hot-spot" averaging were not directly utilized. Rather, justification that the overall average concentration meets the

-release criteria (Le., BTP Option #1 guideline value) is provided herein. This was determined to be appropriate based upon the characteristics of the concrete rubble and the low probability that a portion of the concrete rubble would be extracted and used in a manner that could contribute significantly to inadvertent exposure.

7. 7.2 Exposure Rate Guidelines (External Dose)

The exposure rate ( external dose) guideline value was established as 1 O micro-Roentgens {µR) per hour (average) above background at one meter above the surface in accordance with the BTJ>4. Option #1 criteria. As stated in the BTP; this is compatible with proposed EPA cleanup standards for inactive uranium processing sites.

Exposure rates may be averaged over a 100 m2 grid area as described in draft NUREGiCR-584922*

Draft NUREG/CR-5849 also states that the maximum FINAL ST ATVS SUR VEY REPORT 15 PHASE H SUB~AREA "F" CONCRETE RUSBLE

exposure rate at any discrete location within a 100 square meter area cannot exceed 20 µR/h above background. Cimarron Corporation utilizes 7 to 10 µR/h as the avei+ag~ background exposure rate.

7.7.3 Volumetric Activity of Soils and Sediments The guideline value for residual concentrations of total uranium which tnay remain in soils or sediments is specified as BTP Option #1 material for Sub-Area F. For enriched uranium, as specified in Table 2 of the BTP 1

\\ the Option #1 limit is 30 pCi/g total uranium above background. The amount of soil and sediment sampling performed during this fmal status survey was confined to selected locations which were utilized to demonstrate that the concrete had not resulted in significant impact to site soils and/or sediments (i.e.~ the residual contamination is fixed),

7.8 Equipment Selection Special Work Permjts (SWPs) and Work Plans (WPa) were written and approved for the field work required during the conduct of this fina1 status survey. The SWPs and/or WPs for this project specified the type of instnunentation to be utilized in performing the site surveys. The instrumentation utilized by site personnel is discussed below, 7 Jt 1 Equipment and. Instrumentation The instrumentation utilized to generate the characterization and final status survey data is calibrated and maintained in accordance with the Radiation Protection Program procedures. These procedures utilize the guidance contained in ANSI N323*1978, IIRadiation Protection Instrumentation Test and Calibration 1123* Specific requirements for instrumentation include traceability to NIST standards, field checks for operability~ background radioactivity checks, operation of instruments "'i.thin established environmental bounds (i.e.

temperature and pressure), training of individuals, scheduled perfonnance checks!'

calibration with isotopes with energies similar to those to be measured, quality assurance tests, data review) and recordkeeping.

Portable survey instruments utilized during the survey (micro-R survey meters, a./~ survey meters~ scalers/ratemeters, etc.), are calibrated on a quarterly basis.

All instrumentation is calibrated with NIST traceable standards.

Where applicable, activities of sources utilized for calibration are also corrected for decay. In addition to the quarterly calibration requirements, source checks are perfonned on a daily basis for all instruments being utilized for characterization and final status surveys. A calibrated electronic pulse generator is *utilized for instrument scale linearity checks. All calibration and source check records are FINAL ST ATVS StJRVEY REPORT 16 PHASE 11 SUB.AREA 'T' CONCRETE RUBBLE

completed, reviewed, signed off ar1.d retained in accordance with Cimarron Quality Assurance Program requirements.

An SV1.P was 'Written and approved prior to commencement of field work covered under this Final Status Survey Plan.

The SWP specified the type of instrumentation to be utilized in performing the site surveys, The instrumentation utilized by site persom1el is discussed below.

7.8.1.1 Micro-R Survey Meter The Micro-R meter utilizes a 1 ~~ :x l"' NaI/Tl crystal ganuna detector and measures exposure rates between O and 5,000 µR/h. Background readings are obtained daily at a defined location prior to placing each instrument into service. This instrument was utilized, in general,. for dete:n:nination of exposure rates at both systematic and random locations., and at locations of elevated radiation as identified by gross beta-gamma scans.

Quarterly comparisons and/or confinnatory measurements are obtained routinely to provide information concerning any measurement bias. These comparisons or confirrnatory measurements are made using a pressurized ion chamber.

7~8.12 Soil Counter (Gamma Spectroscopy)

The Cimarron Soil Counter consists of a 4 11 x 4 n x 16" sodium iodide crystal housed in a shielded chamber which is computer linked to a multi-channel analyzer (MCA). The soil counter is programmed to detennine the total uranium concentration present in the soil sample by calculating the U-234 concentration present from the U-235 concentration which is measured in the soil. These two isotopic values are then summed with the measured U~238 value to determine the concentration of total U.

Calibration of this counting system is performed annually and is traceable to NIST standards through contractor labo.ratory evaluations of the on-site standards.

ORlSE has been used by the l\\TRC for verification of a majority of the decommissioning work completed to date at the Cimarron site.

ORJSE has conducted an evaluation of the Cimarron Soil Counting system 1s ability to accurately measure total uranium concentrations in soil samples. This was done by comparing ORISE sample analysis results obtained by alpha pulse height analysis and gamma spectroscopy with the results obtained from the use of the Cimarron Soil Counter.

OR1SB ru1d Cimarron analysis results compared favorably at levels above background as demonstrated by the most recent confirmatory analysis perfonned for the On-Site Disposal Cell, Pit #3 (NRC cover letter dated July 31, 1997?

4 NRC inspection Report #70-925/97~02, which accompanied this letter, states that "no significant bias or statistical errors between FINAL STATUS SURVEY REPORT 17 PEASE U SUB-AREA F" CONCRETE RUBBLE

7.9 the license~s soil results and the NRC's results were identified". Additionally, the confirmatory analysis performed on select soil samples collected during ORISE,s site visit to investigate the South U-Yard21, and D1\\P-3 stockpileis verified previously that Cimarron' s on-site counter results are statistically identical to ORISE's results.

Established quality assurance measures for the soil counter include Cesium-13 7

-cent.roid checks, Chi-square tests1 background determinations~ and the counting of soil st2ndards. All of these quality assurance controls are recorded on control charts and are trended on a continuing basis.

Standards used for calibration and quality assurance checks for the soil counter have been analyzed by outside laboratories and are :NIST traceable through these analyses. Comparisons have been made benveen. the standards as counted using the soil counter and two off-site independent laboratories. The assigned values for the standards are the average of the results obtained from the off-site independent laboratories) when the standards were analyzed by more than one laboratory. The standards used at Cimarron range in concentration from 4.5 pCi/g total uranium to 292 pCi/g total uranium.

Cimarron personnel determine uranium and tht;uium activities based upon the evaluation of net counts from the soil counter. Activities are calculated through the use of efficiency and i)Orrection factors obtained using appropriate standards.

Soil concentrations are calculated by dividing the net activity by the soil mass*

Soil masses are determined on a laboratory scale which is checked on a daily basis (when in use) utilizing NIST traceable standards. Corrections for soil moisture content are also made as necessary.

Procedures/Plans As discussed in Section 7.8; SWPs and \\VPs were written and approved for the field work required for this final status survey. These SWPs and vVP' s are an integral part of this site>s radiation protection and quality assurance programs.

Project organization and responsibilities, which are a part of the site's quality assurance program~ are discussed in this section.

7.9.1 Organization The final status survey of concrete rubble in Sub-Area HF" was perfom1ed by a survey team consisting of qualified personnel from the Cimarron site. The final status survey team operated under the general direction of a Decommissioning Supervisor who served as the Project manager and reported directly to the Site Manager of the Cimarron Facility.

f!NAL ST.A ms SUfWEY REPORT PHASE 11 SUB-AREA '

1F" CONCRETE RUBBLE

The selection of field measurement equipment and sample collection techniques was performed under the direction of the RSO/Health Physics Supervisor who also reports to the Cimarron Site Manager. Actual field measurements and sample collection were performed under the direction of the Decommissioning Supervisor. The Deconnnissioning Supervisor was responsible for developing the SWP and WP for this sub-Area with input from the RSO/Health Physics Supervisor and the Cimarron Site Manager. The SWP and \\VP were reviewed and approved by the Cimarron Site Manager and the RSO/Health Physics Supervisor.

7.9.2 Training Cimarron Corporation provides continuing training to Cimarron personnel and any other personnel (i.e., contractors, visitors, etc.) who are allowed access to the site. All members of the final status survey team attended an in-house training session on the S\\VP and WP for the work performed under the final status survey plan. All survey procedures and quality assurance requirements were reviewed during this training session.

7.9.3 Radiation Protection Program Cimarron Corporation maintains a radiation protection program which meets and/or exceeds all of the applicable regulatory requirements associated with activities co;ndiicted under Special NJciear Materials Lic'~nse SNM-928. The C1marron Radiation Protection Prograxn currently in place for all deconunissioning activities is administered through the use of the following

documents:

Cimarron Quality Assurance Plan and Procedures Cimarron Radiation Protection Procedures Cimarron Site Health and Safety Plan Cimarron Emergency Plan It is the policy of Cimarron Corporation to perform all work in strict compliance with applicable regulatozy and internal requirements. The goal of the Cimarron decommissioning effort is to conduct all operations at a level of excellence which exceeds regulatory requirements.

Cimarron staff will continue to exercise appropriate radiation protection precautions throughout the remaining decommissioning work and :final survey process.

Independent Kerr-McGee Corporate audits for regulatory and internal requirements are conducted on a periodic basis and include the review of the Cimarron Radiation Protection Program and associated programs. Assessments of program effectiveness also are perfonned periodically by the Cimarron FINAL STATUS SURVF.Y REPOR'i PHASE II SUB-ARf:A "f" CONCRETE'. RUBBLE 19 ii

~ I

~ il jl II

RSO/Health Physics Supervisor. Additionally, the Cimarron Radiation Protection Program is inspected for compliance with applicable rules and regulations by NRC Region IV and NRC Headquarters staff.

7.9.4 Cimarron Quality Assurance Program. (QAP)

The Cimarron Corporation QAP is an integral part of the Cimarron Radiation Protection Program. A principal c~mponent of the QAP is the confirmation of the quality of project work perfonned during decommissioning by assuring that all tasks are performed in a quality manner by qualified personneL The Program ensures that samples are collected, controlled, and analyzed in accordance with applicable quality controls to provide adequate confidence that the resulting data accuracy and validity are verifiable.

Such quality controls allow for the independent verification of analysis results by a third party review.

The Cimarron QAP is implemented and maintained in accordance with written policies, procedures, and instructions. This Program is administered under the direction of the Quality Assurance Manager. Periodic audits and reviews are conducted to ensure that all aspects of the Program are addressed.

"Written procedures~ designated as SWPs and V!P's, are prepared1 reviewed and approved for activities involved in carrying out the decommissioning process.

The Sub-Area uF' concrete rubble survey SWP and WP were ~ntten in accordance with the Cimarron QAP. These documents designated the type of surveys to be performed, samples to be collected, frequency of sample collection:>

number of samples to be split with an off-site independent laboratory, and the type of field instrumentation required for the tasks required.

The facility performs its own radiological soil analysis in accordance with \\vritten procedures and QA/QC protocols. Field data are gathered and maintained in logs for all samples in accordance with the Cimarron QAP.

Necessary data are tra11sferred to the on-site laboratory sample log when the sample is brought to the on-site laboratory for analysis.

The sample logs provide a record of sample collection and transport ( chain of custody) and are incorporated into the facility quality assurance records.

In addition., off~site independent radiological analysis of split samples (srunples are :first counted on-site and then sent to an off-site independent laboratory) is an integral part of the Cimarron QAP. Samples sent to an off-site independent laboratory for analysis are accompanied by a chain of custody fonn in accordance with the Cimarron QAP. These forms provide documentation for all aspects of sample control and are maintained by the Quality Assurance Manager as pennanent records.

FI'NAL STA'fL!S SURVEY REPORT PHASE II SUBwAREA "f" CONCRETE RUBBLE 20

Sample and survey data are reviewed by the Health Physics Department for accuracy and consistency and are compared to the guideline values. Reviews are performed on a regular basis. When identified, corrections to recognized deficiencies are performed.

Planned and periodic audits of Cimarron 1s Quality Assurance Program are performed by individuals who do not have direct responsibilities for the areas being audited. Audit results are documenteo. for review by management 8.0 Survey Findings Final Status Survey data was generated for the concrete rubble located in Sub... Area HP' in order to demonstrate that this concrete rubble could be unconditionally released. The survey findings, including the methodology employed to evaluate the data, are described in this section.

8.1 Thennoluminescent Dosimeter (TLD) Exposure Rate Data Thermoluminescent dosimeters were placed at three locations (#AM015, #AM016, and

AM017) near or above the concrete rubble.

TLD data for 1996 and 1997 are provided in Tables 8.1 and 8.2, along with data for TLD location #AM014, which is located approximately one half mile south of the facility near the junction of Highways

~.

33 and #74. Location #AM014 represents background. All TLDs were placed at a height of approximately one meter above the ground or concrete surface, and were oriented to face the area to be monitored. Drawing No. 98 _ TLD depicts these TLD locations.

Table 8.1 TLD Exposure Rate Measurements-1996 jTLD#

/ Description I lQ96 2Q96 1 3Q96 4Q96 96 Ave~

j µR/h

µR/h

µRib

___ µRib ____

AM014 Junction of Highways 33 & 74 7.6 8.3 5.6 I

7.0 I

7.1 i

I (Background)

I I

AM015 Res. #21 SE of Rubble 8.2 7.0 5.7 6.5 6.9 I

I

~~016--l Res. ~2, Middle of Rubble 9.3 10.2 5.3 I 6.7 7.9 I

\\

i 1 AM017 I Res. #2, Below Rubble I ~ts I 7.6 I

5.0 5.7 6.7 I

j rTNAL STATUS SURY.EY REPORT 21 PHASI;. lJ SUB~A!'-t.eA **r+' CONCRETE RUBBLE

Table 8.. 2 TLD Exposure Rate Measurements-1997 I Description 1Q97 2Q97 3Q97 97 Ave....

TLD#

µR/h

µR/h AM014 Junction of Highways 33 & 74 9.2 6.9 i 8.8 8.3

(Background)

I AM015 Res. #2, SE of Rubble 6.3 I 5.9 5.7 6.0 I

I AM016 Res.. #2, Middle of Rubble 6.7 6.0 6.4 6.4 A.M017 Res. #2, Below Rubble I

4.6 4.6 6.1 I 5.1 I

I I

j

Only the first three quarters of 1997 data were available at the time this report was published.

During 1996 1 the exposure rate near the concrete averaged 7.2 µR/h at the three indicator locations (#At'\\1015~ #AM0161 and #AJv.l:017)y and averaged 7.1 µR/h at the background location (#AM014). The exposure rates during the first three quarters of 1997 were similar~ averaging 5.8 µR/h at the three indicator locations, and 8.3 µR/h at the background location.

Data for the fourth quarter of 1997 have not yet been received from the contractor and are therefore not available.

The TLD exposure data does not indicate any elevated exposures occurring as a result of the elevated concrete surface contamination. This would be expected considering the low activity levels present in the concrete~ and the attenuating characteristics of the concrete for any low energy gammas present due to the low residual uranium. ln addition, the concrete provides shielding for naturally occurring gamma emitters within surface soils, and thus exposure rates are expected to be lower over concrete than in other locations. This observation is supported by the survey data presented in this report.

The TLD data also supports the measurements obtained with micro-R meters.

The TLD data indicates that the guideline value of 10 micro-R above background is met at each TLD location during 1996 and 1997 to date.

FINAL STATUS SURVEY REPORT 22 PHASE II SUB-AREA '

1'F" CONCRETE RUBBLE

8.2 Soil/Sediment Samples In accordance with the Work Plan, one soil/sediment surface sample (6" depth) was collected from area upgradient of the main body of concrete rubble, ten surface samples from areas within and beneath the concrete~ and two surface samples from areas downgradient of the concrete. The sampling locations and sample results are summarized in Table 8.3.

The locations where each soil/sediment sample was collected are shown in Drawing No. 98FCRSS (Appendix I). Sample results do not indicate any samples above the BTP Option #1 guideline concentration of 30 pCi/ g, above background. Concentrations of total uranium ranged from 2.9 to 8.6 pCi/g, while total thorium ranged from 0.2 to 1.0 pCi/g. The concentrations in these soil and sediment samples are similar to those found in unaffected areas, and do not indicate that there is any significant contribution occurring as a result of the concrete in the immediate vicinity.

Table 83 Soil/Sediment Samples Collected Around Concrete Sample Number Grid Location Total u* (pCi/g)

Total Th. (pCi/ g)

FA-535 (upgradient) 1451E-820N 4.2 1.2 FA~536 1419E-839N 2.9 0.9 FA-537 1419E-845N I

5.9 0.8 FA-538 1420B--848N 8.6 0.2 FA-539 1412E-855N 7.0 0.8 FA~540 1404E-860N 4.2 0~7 FA-541 i'34 IE-865N 4.6

0.7 FA-543 (dmvngradient) 1358E-870N 3.9 I

0.3 j FA.. 544 1387E-870N 4.5 0.7 jFA:545 1412E-870N 5.3 1.0 I

I FA-546.(downgradient) 1365E-873N 4.3 0.7 I FA-547 1378E-875N 4.4 0.6 FA-548 1370E~878N 3.6 0.6

Reported measurements include the contribution from natural background.

8.3 Surface Water Samples Surface water samples were collected upstream and downstream of the concrete on September 30~ 1997. Sample results are summarized below in Table 8.4.

The upstream and downstream surface water samples show low levels of naturally occurring concentrations of uranium and thorium_ There was no *significant difference

.~

between the upstream and downstream sample. In addition, the sample data do not indicate any significant differences from the historical data reported for environmental FINAL STATUS SURVEY REPORT 23 PHASE H SUB-Al~ ui:** CONCRETE RUBBLE

monitoring location #1205 (See Section 6.4). This indicates that the concrete is not contributing to the concentrations of these naturally occurring contaminants.

Table 8.4 Laboratory Data for Surface Water Samples (Results in pCi/1)

I I Location Th-228 Th-232 U-234 U-235 U-238 l

I I

l Upstream

! (FA-WAT~315) 0.4+/-0.3

<0.4 0.5+/-0.3 i

<0.1

<0.3***--*i

[ Downstream I

I (FA-WAT-316) 0.4+/-0.4

<0.5 I

0.8+/-0.4

<0.2 0.3+/-0.2 8.4 Micro-R Measurements A tabular summary of all micro-R meter surveys for the random grid samples and for the all sampled grids is provided in Tables 1 and 2 of Appendix n. Drawing No.

98FCRER (Appendix I) also presents the average exposure rate measurements for each 5m x 5m grid area for which measurements were collected. For the random sample, the average exposure rate over any 25 m2 grid area ranged from 6 µR/h to 10 µR/h at one meter from the surface. The overall average exposure rate for the random sample was 7 µRfh at one meter from the surface. At one meter above the surface! the maximum exposure rate was 10 µR/h, including background.

This maximum was measured at the location that had the highest exposure rate at the surface, which was 25

µR/h, including background (grid# 52). Assuming a background of 7 µR/h, the net annual exposure rate at this surface location would be 18 µR/h.

This location was evaluated to determine any significance with respect to exposure of the general public.

Under normal circumstances, it is unlikely that any additional exposure would occur to members of the public as this piece of concrete is within a drainage area, and is on land owned by Cimarron Corporation. The possible exposure scenarios evaluated included hunting the land or an intruder inadvertently remaining in the area for a period of time.

Assuming ten hours per year exposure, the hypothetical individual could receive an annual dose of 180 µrem to the portion of the skin of the whole body or to any organs situated directly in contact with the concrete with high residual activity. A more likely scenario is from a person standing in the area for a period of several hours per year.

The net annual dose rate above background from this hypothetical activity would be approximately 9 µrem/y to the whole body, based upon the net measured exposure rate at a height of one meter. Both of the above dose scenarios are unlikely, in that the concrete rubble is not in an area where it would be desirable to spe:11d any amount of time.

In comparison to the exposure that an individual receives from natural background radiation { ~300 millirem/y) t. the calculated hypothetical doses of 180

µrem/y and 9 µrem/y are insignificant.

FTNAL STATUS SURVEY REPORT 24 PHASE H SUwAREA "f CONCRETE RUBBLE.

I I

.~

8.5 The data summary for exposure rate measurements of all sample grids (Le.~ random sample plus supplemental grid area data) is also included in Table 2 of Appendix IL This data is similar to the random grid sample data described above; and also indicates that there are no significant exposure risks from leaving the concrete in place.

The data summarized in Tables land 2 of Appendix II, and presented in Drawing No.

98FCRER, include the exposure due to natural background radiation~ which has been previously determined to range from 7~ l O µRib at the Cimarron site. Therefore, the estimated dose at one meter is essentially equal to that which would be received due to natural background.

Gross Alpha and Gross Beta-Gamma Surface Activity Data Gross beta-gamma scans were perfonned over the entire surface of selected random grids containing concrete *rubble to determine the nature and extent of the activity. In addition~ the random sample data were supplemented with additional measurements to ensure that the random sample was representative. Gross alpha and gross beta~gamma surface activity data are summarized in Tables 3 through 6. in Appendix II. Drawings No. 98FCRA and 98FCRB (Appendix I) also present the average gross alpha and gross beta-gamma surface activity data for each 5m x Sm grid area. The data tables pre.sent the measured activity for each representative area and hot spots within the sampled 5m x 5m grids. Where applicable, background subtraction was performed on the gross beta-ganuna hot spot data to obtain the expected increase that is due to residual activity.

The average and maximum activity ( dpm per 100 cm2) is calculated for each of the 5m x 5m *grids. The average volumetric concentration over the 5m x 5m grid area was then calculated using the relationship between gross beta-gamma surface activity and volumetric concentration that is presented in Section 8.6.2.

The volumetric concentration calculation assumes that the average gross beta-gamma surface activity is representative of the grid area as a whole, and that there is equal probability of measurement of the residual activity on the most elevated side of the concrete as there is for measurement on the least elevated side~ This assumption is reasonable based upon the random manner in which the concrete was placed, and the random _manner in which the sampling of the grid areas was performed. In order to account for the probability for residual activity to exist on both the top and bottom sides of each concrete slab, the average thickness for the concrete was divided by two.

Therefore, since the average thickness of the concrete was found to be one foot, the volumetric concentration was calculated over a thickness of six inches.

8.5.1 Gross Beta-Gamma Dat.a Overall, the random grid sample data contained more elevated locations per grid and was found to be more limiting in that the data indicated higher overall concentrations than the data which included both the random sample and the FlNAL STA TCJS SURVEY REPORT 25 PHASE Tl SUB*AREA 'T' CONCRETE RUBBLE

supplemental survey data. The maximum gross beta-gamma surface activity

( concrete background subtracted:, and averaged over 1m2) was found within grid

43 (hot spot #1).* This hot spot location measured 261075 dpm/100 cm1. The highest average gross beta-gamma surface activity over any 5m :x 5m grid was found in grid #5lt which averaged 4,867 dpm./100 cm\\ with background subtracted. Using the conversion to volumetric concentration, this equates to

7. 4 pCi/ g average total uranium concentration over the 25nr1 area.

This concentration is well below the BTP Option #1 guideline value.

The overall average volumetric concentration calculated for the random sample was 2.9 pCi/g~ For all sampled grids (i.e., random sample and supplemental samples), the average volumetric concentration was found to be 1.8 pCi/g.

Although above typical background levels for concrete 1 these concentrations are similar to those found in nature and are well below the BTP Option #1 criteria.

Therefore, the health and safety significance of leaving the concrete in place is similar to the health and safety considerations for natural soils. The potential

. uses of the concrete are limited by its portability and by the difficulty that would be experienced through attempts to remove it from the drainage areas.

Therefore, it is anticipated that the any exposures to the concrete would be from casual contact or from its gradual disintegration over time due to environmental forces.

8.5.2 Gross Alpha Data The gross alpha survey data are presented in Tables 5 and 6 in Appendix II and are also summarized in Drawing No. 98FCRA (Appendix I). Data are summarized for the random sample as well as for all sampled grids.

Data generally indicate that the concrete rubble would meet the* current gross alpha guideline criteria of 5)000 dpmflOO cm1 (average) and 15,000 dpm/100 cm2 (maximum) for unconditional release.

As previously discussed. in Section 1.0, the concrete was placed into Sub-Area np>> after gross alpha surveys were performed. The criteria in effect at the time the concrete was placed into Sub-Area "P' varied from the criteria currently in effect (as discussed above) up to 25,000 dpm/100 cm\\ which was allowable under the NRC criteria which was in effect at the time, Review of the data indicate that two of the sampled grids contained hot spots exceeding 5,000 {average) or 15,000 (maximum) dpm/100 cm2* Grid #52 contained a hot spot with 9,000 dpm/100 cm.2 (average) and 16,000 dpm/100 cm2 (maximum). In addition, Grid #56 contained a hot s'pot with 15~600 dpm/100 crn2 (maximum).

The criteria proposed for release of the concrete is based upon volumetric concentration. The concrete rubble does not have any smearable contamination and the activity would have to be removed through mechanical or physical forces. While it is probable that environmental forces v.rill eventually act to remove the radioactivity; FINAL STATUS SURVEY REPORT 26 PBASE II SUB-ARE.A "F" CONCRETE RUBBLE

8.6 normal environmental dispersion will result in insignificant quantities available for ingestion or inhalation.

Calculations This section describes calculations and evaluations that were required in order to evaluate the concrete rubble and perform comparisons with the proposed release criteria. The calculations included determinations of tbe average thickness of the concrete} which was used to determine the appropriate volume of concrete for averaging the residual activity, The calculations performed to determine the relationship between gross beta-gamma surface activity and volumetric concentration of total uranium are also described in this section. Finally, source term was estimated based upon the random sample data..

The source term allows for evaluation of the acceptability of the proposed action, which is to leave the concrete in place.

The source term can be utilized for input into computer models or for use in comparing the residual activity with that present in the natural environment. A RESRAD computer model was run to calculate hypothetical dose to individuals over a period of 1000 years.

8..6.1 Calculation of Average Concrete Thickness The volume estimate for the concrete was previously presented in the Decommissioning Plan. The data was evaluated and a weighted average thickness for the concrete was calculated.

The weighted average thickness accounts for. the presence of different volumes of concrete rubble that also have different thickness. The weighted average is calculated as follows; Weighted Average = LI(thickness) x (volume)J/I;[volume};

The above formula can be explained as follows. For each area of the facility, the concrete thickness in the area is multiplied by the volume of the concrete that came from the area~ These individual volume thickness calculations are summed and divided by the overall volume of the concrete (i.e., all concrete that was placed into Sub-Area (tF") to obtain the weighted average.

Using the above calculation'f the weighted average thickness of the concrete in Sub-Area "F" was estimated to be 1 foot. The data tables showing the dimensions of the concrete present in Sub-Area <<p~ are presented in Appendix Ill.

8.6.2 Volumetric Concentration Conversion Factor A special stUdy was performed as described in Section 7. 6 to determine a relationship between gross beta-gamma surface activity on the concrete (measured in dpm/100crn2) and the volumetric concentration (measured in pCi/g) of total uranium. The solution to

-~,

this problem is confounded due to the natural presence of beta and gamma emitters in concrete, including uranium.

In addition., the low energy of the beta emitters FfNAL ST A TVS SUR VEY REPORT 27 PHASE U SUB-AREA "'F" CONCRETE RUBS.LE

associated with enriched uranium and the variability in depth of the contaminated layer hinder the determination of a single conversion factor for this purpose. However, the data presented for the two slabs studied indicate a reasonable agreement between the data.

For this study, two slabs were selected. Slab #1 had an initial average gross beta-gamma surface activity of 4496 dpm/100cm2 (concrete background subtracted), while slab #2 had an initial average gross beta-gamma surface activity of 17,697 dpm/100cm 2

(concrete background subtracted).

Measurements were performed on each slab as described in Section 7.6,. followed by scabbling of another layer (average depth approximately 1 /8 inch), resurvey, and additional scabblin_g until the majority of the contamination was removed.

The scabble dust and particles were collected and analyzed for total uranium content i1sing either the onsite soil counter or using an off-site independent laboratory. Survey and laboratory data for the two slabs is presented in Table 8.5.

The special study data revealed that tbe maximum depth of the contamination was approximately 3/8 inch. The contamination was found to be highest, as expected, at the layer closest to the surface, and decreased substantially as additional layers of concrete were sca.bbled and removed.

The concentration of total uranium was assumed to be at background (i.e., 1.5 pCi/g) for an remaining concrete when the gross beta-gamma measurements indicated that the residual activity had been essentially removed.

Complete removal of the residual activity was achieved after two scabbling operations on Slab #1,. and after three scabbling operations on Slab #2.

It was determined that each scabbling operation removed approximately 1/8 inch from the surface.

The volumetric concentration conversion factor was determined using the average thickness of six inches as follows:

Slab #1 Volumetric average conversion factor (dpm/100 cm2 gross beta-gamma per pCi/g total U) =

(17,697 dpm/100 cm2) + {[814.7 pCiJg + 102.7 pCi/g + (LS pCi/g x 46)] + 48} =

S6L2 dpm/100 w~ __ gros.s_bet.a-gamma P.eI:~ll.Ci/g total U.

FIN AL STATUS SURVEY REPORT PHASE ll SUB~AR.EA "F CONCRETE RUBBLE 28 *

)

Gross Alpha dpm/100cm2 (ave)

S1ab Ill Initial Measurement 1251 After 1st Scabbling 157

-* After znrt Scabbling 24 Slab #2 Initial Measurement 329 After 1st Scabbling ll.5 After 2nd Scabbling 116 After 3rd Scabbling

<350

)

.,*TABLE8.S Survey Data for Concrete Rubble Slabs Used ro Determine the Volumetric Concentration Conversion Factor Gross Alpha Gross Beta Gross Beta Slab Slab dpm/100cm2 dpm/100cm2 Dpm/100cm2 Surface lm (max)

(ave)

(max)

µR/h

~lR/h 4800 17,697 44,540 9

7 480 2411 8330 6

6 160 0

1310 6

6 1280 4496 24,390 6

6 480 935 13,370 5

6 400 0

4,950 6

6

<350 0

230 9

9 Hot Spot Surface lm

_µR/.h

µR/h 11 7

6 6

9 8

8 6 ~-

6 6

9 9

Notes: 1) Concrete gross beta-gamma background (800 dpm/100cm2) subtracted from beta-gamma measurements.

2) No background subtraction performed fur all other measurements.

3) Measurement~ less than O afte.r background subtr.a(..*tion were recorded as 0.

')

Total U Cone.

(pCi/g) 814.7 102.7 313 74.9 17.7

4) Total U concentration assumed as 1.5 pCi/g when all residual gross beta-gamma activity was determined to be removed by scabbling.

FIN"r1.l ST ATOS SURVEY RE.PORT PHASE fl SUOwAREA "FH CONCRETE RUBBLE 29

/..-..._,_

Slab #2 Volumetric average conversion factor (dpm/100 cm2 gross beta-gamma per pCi/g total U) ::...

(4496 dpm/100 cm2) + {[313 pCi/g + 74.9 pCi/g + 17.7 pCi/g + (1.5 pCi/g x45)] + 48} =

456.2 dpm/100 cmi gross beta.. gamma yer pCi/g total U.

In the above calculations, each scabbled layer is assigned a thickness of 1/8 inch, which corresponds to the measured concentration of the concrete. It follows that there are 48, 1/8 inch layers in a six inch slab of concrete. Each background layer is assumed to have a total uranium concentration of L5 pCi/g, as discussed in.Section 6.2.1. The numerator in the above equations is the measured gross beta... gamma surface activity on the top layer.

Tiris data is readily available and was obtained during the surveys of the concrete rubble.

The two slabs studied in this special project indicate that the conversion factor is in the range of 456 to 861 dpm/100 cm2 gross beta-gamma per pCi/g total U.

The two samples resulted in calculated conversion factors that were within a factor of two, which is good agreement considering the numerous areas of uncertainty.

Toe average of the two measurements, which is 661 dpm/100 cu1 gross beta-gamma per pCi/g total U, was utilized to estimate the average volumetric concentration of residual activity present in the concrete. For comparison, the volumetric conversion factor was also calculated for an assumed concrete thickness of three inches.

Tlris conversion factor was also used to calculate average total uranium concentrations. The data tables for gross beta-gamma surface activity in Appendix II (Tables 3 and 4) also present the average volumetric concentrations for each 5m x 5m grid area.

8.6.3 Volumetric Concentration Calculations Calculation of the estimated volumetric concentration (in pCi/g total U) was performed by multiplying the conversion factor described in Section 8.6.2 times the measured gross beta-gamma surface contamination measurement (with background subtracted).

These calculations, which are summarized in Appendix II (Tables 5 and 6}) resulted in average total uranium concentrations (after background subtraction) ranging from -

0.8 pCi(g to 7.4 pCi/g. The negative results indicate that the Sm x 5m grid average was less than 1.5.pCi/g, Subtraction of the 800 dpm/100cm2 concrete background thus resulted in a value that was less than average. Assuming a norm.al distribution of a background distribution, one half of the samples collected would be expected to be less than background.

The maximum 5m x 5m grid average total uranium concentration was 7.4 pCi/g (grid

51). This concentration is less than 25% of the guideline value of 30 pCi/g. The average total uranium concentration for the random sample was 2.9 pCi/ g~ which is less

~..

than 10% of the. guideline value for enriched uranium (30 pCi/g). For the random FINAL STATUS SURVEY REPORT 30 PHASE U SUB~AR.EA "F" CONCRETE RUBBLE

sample and the supplemental grid areas (66 grid areast the total uranium average

concentration was 1. 8 pCi/ g, which is equal to 6 percent of the guideline value.

8.6.4 Source Term Calculation The average volumetric concentration for the random sample was calculated to be 2.9 pCi/g. Since the volume of concrete is known! an estimate of the total activity of uranium present in the concrete can be calculated as follows:

Total activity ~ (2.9 pCi/g-concrete) x (1.8 g-concrete/cc) x 31 1985 ft3 concrete X 28,320cc/ft3

= 4.7 E+09 pCi = 4.7 E-03 Ci Total Uranium, 8.6.5 Pathway Analysis The RESRAD computer code was used to evaluate the potential dose due to leaving the concrete in place.

The RESRAD code considers direct radiation, inhalation of resuspended radioactivity, ingestion of groundwater and foodstuffs grown in contaminated soils 1 or in soils irrigated with contaminated surface or ground water, and all other credible pathways.

The RESRAD model generally will predict a more conservative dose (i.e., a higher dose) than that which could potentially be received, as it generally utilizes conservative.assumptions and includes scenarios for use of the land area that are generally not consistent with the expected uses for concrete rubble.

The input parameters for RESRAD include those defmed in NRC's Policy and Guidance Directive (PG) 8-08 1 ~'Scenarios for Assessing Potential Doses Associated with Residual Activity~26. The uranium isotopic ratios were chosen to be the same as those used by the NRC for the "Environmental Assessment Associated with the BTP Option #2 Onsite Disposal Cell at Cimarron"27

., which were U-234 (79%), U~235 (1.7%). and U-238 (20%). The selected density for the concrete was L8 glee. The calculated area of the contaminated zone is 2970 m2 [31 ~985 ft3 x (0.3048m/ft)3 +

0.3048m], while the calculated thickness is 0.3048m.

The RESRAD calculated maximum dose rate will occur at 900 vears and result in a maximum hypothetical aimual dose to the resident of approxim;tely 1 millirem per year.

A printout of the parameters used and results of the RESR.iu) calculation are provided i:n Appendix IV.

F1NAL ST A TUS SURVEY REPORT 31 PHASE 11 SUB-AR.EA "f CONCRETa RUBBLE

9.0 Summary A Final Status Survey was performed on the concrete in Sub-Area ~rp". The survey incorporated NRC guidance and suggestions for volumetric concentration averaging. This Report presents the results of the Final Status Survey. The survey data were evaluated and doses from leaving the concrete in place were calculated. The evaluations presented in this report indicate that the concrete should be left in place for the following reasons:

RESRAD dose evaluation indicates that the projected maximum dose would occur after 900 years, and that this dose would be approximately 1 millirem per year..

Random sample data indicate that the average total uranium concentration for the concrete rubble is 2.9 pCi/g. The overall average total uranium concentr~tion in the concrete is less than 10 percent of the BTP Option #1 guideline value (30 pCi/g)~

The average total uranium concentration, based upon all sample data, was calculated to be 1.. 8 pCi/g.

The maximum concentration averaged ove:r any Sm x 5m grid area was found to be 7..4 pCi!g., which is less than 25% of the BTP Option #1 guideline value, The overall average exposure rate at -one meter from the surface was 7 µR/h, which is equivalent to natural background. The maximum el..'}losure :rate at one meter from the surface was 10 µRib, which is within the :range of natural background..

The calculated total uranium source term is 4.7 millkuries within the total volume of concrete rubble estimated at 31 ~985 ft 3

No elevated measurements observed for exposure rate as measured by thernwluminescent dosimeters placed in the field adjacent to the concrete rubble..

Surface water samples collected upstream and downstream did not indicate any contribution of radioactivity from the concrete rubble..

Soil/sediment samples collected upstream, within, and downstream of the concrete rubble reflect radioactivity levels which are characteristic of samples collected in unaffected areas of the f~cility..

The concrete continu~s to serve the useful purpose 0£ on-site erosion control.

The calculations presented in this report utilized conservative assumptions. Therefore, it is unlikely that even the low dose which was calculated could be received by any member of the public. Based upon the evaluations in this report, Cimarron Corporation requests authorization from the l\\1RC for unconditional release of the concrete.

FINAL STATUS SURVEY REPORT 32 PHASE tl SUB-AREA "F" CONCRETE RUBBLE

Person / Time
ML19363A016
Site:	07000925 (SNM-0928)
Issue date:	03/10/1998
From:	Jim Larsen Cimarron Corp
To:	Kenneth Kalman Office of Nuclear Material Safety and Safeguards
Shared Package
ML19365A023	List: ML19361A179 ML19363A001 ML19363A002 ML19363A003 ML19363A004 ML19363A005 ML19363A006 ML19363A007 ML19363A008 ML19363A009 ML19363A010 ML19363A011 ML19363A013 ML19363A014 ML19363A015 ML19363A016 ML19364A126 ML19364A231 ML19364A232 ML19364A233 ML19364A234 ML19364A235 ML19364A236 ML19364A237 ML19364A238 ML19364A239
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ML19363A016

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Dear Mr. Kalman:

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