ML20092G368
| ML20092G368 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/19/1995 |
| From: | Abelquist E OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| CON-FIN-A-9076 ORISE-95-F-80, NUDOCS 9509190266 | |
| Download: ML20092G368 (45) | |
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34
. 50-ac,7 CONFIRMATORY SURVEYFORTHE REPOWERAREA FORT ST.VRAIN PLAITEVILLE, COLORADO
[uOCKET NO.50-267]
EW. ABELQUIST Prepared for the Division of Waste Management Headquarters Office U.S. Nuclear Regulatory Commision
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I The Pak Ridge Institute for Science and Education (ORISE) was established by the U.S. Department of Energy to undertake national and international programs in science and engineering education, training and management systems, energy and environment systems, and medical sciences. ORISE and its programs are operated by Oak Ridge Associated Universities (ORAU) through a management and operating contract with the U.S. Department of Energy. Established in 1946, ORAU is a consortium of 88 colleges and universities.
NOTICES The opinions expressed herein do not necessarily reflect the opinions of the sponsoring institutions of Oak Ridge Associated Universities.
This report was prepared as an account of work sponsored by the United States Government. Neither the United States Government nor the U.S. Department of Energy, nor any of their employees, makes any warranty, expres.ed l E
or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe on privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, mark, g manufacturer, or otherwise, does not necessarily constitute or imply its endorsement or recommendation, or favor g by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.
I I
ORISE 95/F-80 i.
so-a &7 CONFIRMATORY SURVEY FOR THE REPOWER AREA
> FORT ST. VRAIN PLATTEVILLE, COLORADO Prepared by E. W. Abelquist Environmental Survey and Site Assessment Program Energy / Environment Systems Division Oak Ridge Institute for Science and Education Oak Ridge, TN 37831-0117 i Prepared for the Division of Waste Management Headquarters Office U.S. Nuclear Regulatory Commission I
FINAL REPORT JUNE 1995 i
This report is based on work performed under an Interagency Agreement (NRC Fin. No. A-9076) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy.
Oak Ridge Institute for Science and Education performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy.
Fort SL Vrmin-Plamevine, CO June 13. 1995
l, CONFIRMATORY SURVEY FOR THE
! REPOWER AREA I
FORT ST. VRAIN PLA'ITEVILLE, COLORADO l
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Prepared by: Mt_ j / Date: 6 7 fk E. W. AbelquiY, Projdid Leader l Environmental Survey and Site Assessment Program 4
Reviewed by: /Arkk #cd, % Date: 4//4/95-M. J. Laudeman, Radiochemistry Laboratory Supervisor Environmental Survey and Site Assessment Program i
Reviewed by: % 7 Oua Date:
d// 9)_4r s A. T. Payne, Administhtive Services Manager, Quality Assurance / Health & Safety Manager Environmental Survey and Site Assessment Program 4
Reviewed by:
Date: g!/9If
/ /
Wl L. Beck, Program Director Environmental Survey and Site Assessment Program 1
l Fort m, Vrain-Finervine, CO . June 13. 1995
i ACKNOWLEDGEMENTS The author would like to acknowledge the significant contributions of the following staff members:
FIELD STAFF ,
I G. R. Foltz LABORATORY STAFF R. D. Condra J. S. Cox M. J. Laudeman S. T. Shipley i
CLERICAL STAFF ,
D. A. Adams i R. D. Ellis
. K. E. Waters 4
ILLUSTRATOR
- T. D. Herrera l
i i
i Fort R. Vrame-Pimmevige, CO June 13,1995
1 l
e TABLE OF CONTENTS PAGE
, _ List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii l
1 List of Tables ............................. . . . . . . . . . . . . . . . iii Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Obj ectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Docu ment Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 ;
1 1
Procedures ................................................ 3 Findings and Results .......................................... 5 1 Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 S u m mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 l l
References ...............................................22 Appendices: l Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors Fat 8L Vraie Plamevine. CO - June 13,1995 i
I LIST OF FIGURES PAGE FIGURE 1: Imcation of the Fort St. Vrain Site-Platteville, Colorado . . . . . . . . . . 10 FIGURE 2: Plot Plan of the Fort St. Vrain Nuclear Station ................11 FIGURE 3: Fort St. Vrain-Repower Area . . . . . . . . . . . . . . . . . . . . . . . . . . 12 FIGURE 4: Repower Area, Miscellaneous Concrete and Metal Surfaces-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 13 FIGURE 5: Repower Area, Evaporative Cooler Building-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 14 FIGURE 6: Repower Area-Exposure Rate Measurement Locations . . . . . . . . . . . 15 1
FIGURE 7: Repower Area-Soil Sampling IAcations . . . . . . . . . . . . . . . . . . . . 16 FIGURE 8: Fort St. Vrain-Background Soil Sampling and i Exposure Rate Measurement Locations .....................17 l 1
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Fort St. Vrain-Flausville. CO . June 13,1995 ii t
A LIST OF TABLES PAGE TABLE 1: Summary of Surface Activity Measurements . . . . . . . . . . . . . . . . . . 18
- TABLE 2
- Ex posu re Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 1
TABLE 3: Background Exposure Rates and Radionuclide Concentrations in Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 TABLE 4: Radionuclide Concentrations in Soil Samples . . . . . . . . . . . . . . . . . . 21 1
e 4
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ras m. v,-n.neva, co - w is, tw5 iii
l i
i ABBREVIATIONS AND ACRONYMS !
R/h microroentgens per hour
- ASME American Society of Mechanical Engineers em centimeter cm2 square centimeter cpm counts per minute DOE Department of Energy l
dpm/100 cm2 disintegrations per minute /100 square centimeters EML Environmental Measurement Laboratory ;
EPA Environmental Protection Agency 1 ESSAP Environmental Survey and Site Assessment Program FSV Fort St. Vrain HTGR High Temperature Gas-Cooled Reactor kg kilogram m meter m2 square meter mm millimeter MeV million electron volts a mrem millirem l MWe Megawatts electric Nal sodium iodide NIST National Institute of Standards and Technology 3 NMSS Office of Nuclear Material Safety and Safeguards NRC Nuclear Regulatory Commission !
ORAU Oak Ridge Associated Universities ORISE Oak Ridge Institute for Science and Education l 2
picocuries per gram I pCi/g PSC Public Service Company of Colorado TEDE total effective dose equivalent a
4 Fort 5L Vrai>Platteville, CO June 13,1995 iV
. )
CONFIRMATORY SURVEY FOR THE REPOWER AREA I FORT ST. VRAIN
- PLA'ITEVILLE, COLORADO INTRODUCTION AND SITE HISTORY Fort St. Vrain (FSV) was a 330 MWe High Temperature Gas-Cooled Reactor (HTGR) owned and operated by Public Service Company (PSC) of Colorado. The site consists of 6995 hectares (2798 acres) owned by PSC, of which approximately one square mile was designated as the ,
l
] exclusion area during plant operation. The licensee maintained complete control over this area. l The basic installation included a reactor building, turbine building, cooling towers, and an I electrical switchyard.
FSV was permanently shutdown in August 1989, with the decision to decommission the facility !
1 made during December 1989. On November 23,1992, the Nuclear Regulatory Commission (NRC) issued the Order to Authorize Decommissioning of Fort St. Vrain and Amendment No.
! 85 to Possession Only License No. DPR-34. During the period 1989 to 1991, a radiological
! characterization of the FSV site was performed. Currently, the FSV decommissioning is approximately 70 % complete, with completion expected early in 1996 (excluding the final site l
survey).
t PSC has committed to the Colorado Public Utilities Commission to resume electrical generation
- at FSV through the installation of gas turbines. In order to perform this project, a small section ofland in the southwest area of the site has been cleared in preparation for the repower effort.
4 This area is referred to as the repower area, where PSC plans to install natural gas-fired combustion turbines and heat recovery boilers to repower the facility.
2 During plant operations, pre-fabricated steel buildings were located in the repower area. These buildings accommodated a construction workshop, a quality control facility that performed radiography, a small warehouse, and a flammable storage building.
Fast St. Vrain-Plaaevitic, CO . June 13,1995
A q
The repower area does not have a known history of radioactive contamination. This was reaffirmed by the evaluation of the characterization results for the repower area, which did not identify radioactive contamination due to licensed activities. The repower area was therefore classified as an unaffected area. However, elevated levels of Cs-137 were identified in surface j soil collected from a localized area outside, but adjacent to the repower area. This area had previously been used for temporary storage of spent fuel shipping casks.
- At the request of the NRC's Division of Waste Management, Headquarters' Office, the I Environmental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) performed an independent co'1firmatory radiological survey of the repower area at the Fort St. Vrain site in Platteville, Colorado.
SITE DESCRIPTION
)
The FSV facility is located approximately 56 kilometers (35 miles) north of Denver and 5.6 kilometers northwest of the town of Platteville, in Weld County, Colorado (Figure 1). The site f consists of 6995 hectares owned by PSC of which approximately one square mile was designated i as the exclusion area during plant operation.
The repower survey area is located within the restricted area of the FSV facility on the east side i of the turbine building, north of the electrical switchyard (Figure 2). This location is approximately 12,225 square meters in size. The area has been isolated from the balance of the I restricted area by a chain link fence and locking gates controlled by FSV Security. The boundaries of the repower area include portions of the original restricted area fence on the south -
- and east sides and newly erected fence on the west and north sides (Figure 3). The boundary also includes the east, and a portion of the south exterior walls of the evaporative cooler building.
I 1
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Port k. Vrain Flamevdle, CO - June 13.1995 2 i
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4 OBJECTIVES The objectives of the confirmatory survey were to provide independent document reviews and radiological data for use by the NRC in evaluating the adequacy and accuracy of the licensee's procedures and final status survey results, t
- DOCUMENT REVIEW ESSAP has reviewed the licensee's final survey report and radiological survey data. Procedures and methods utilized by the licensee were reviewed for adequacy and appropriateness. The data were reviewed for accuracy, completeness and compliance with guidelines.
PROCEDURES i
j During the period March 20 through 22,1995, ESSAP performed a confirmatory survey at the i
Fort St. Vrain site in Platteville, Colorado. The survey was conducted in accordance with a survey plan dated March 17,1995, submitted to and approved by the NRC's Division of Waste
. Management, Office of Nuclear Material Safety and Safeguards (NMSS).2 This report summarizes the procedures and results of the survey. Additionalinformation concerning major f instrumentation, sampling equipment, and analytical procedures is provided in Appendices A
- and B.
SURVEY PROCEDURES 4
i The licensee's final status survey of the repower area included two general categories of survey units: surfaces and structures, and open land areas. The area was further divided into survey units. The surface and structure survey units within the repower area included the Valve Pit, 4 Miscellaneous Metal Surfaces, Concrete Slab at the Security Fence, Miscellaneous Concrete Surfaces, Evaporative Cooler Building (east and south walls below 2 m), and Evaporative Cooler Building (east and south foundation walls). The open land area survey units included the general soil area within the repower area, leach field soil area, septic system, and monitoring wells.
- ron m. vem, co.s u. ms 3
Reference System ESSAP selected specific measurement and sampling locations from each of the survey units.
Because survey maps illustrating the measurement locations were not provided by the licensee, ESSAP requested that the licensee identify some of their measurement and sampling locations
- for confirmatory measurements. Measurement and sampling locations were referenced to prominent site features and recorded on survey maps.
Surface Scans Soil surfaces were scanned for gamma radiation using Nal scintillation detectors. Approximately 10% of the soil in the general and leach field soil areas was scanned. Structure surfaces were 1
also scanned with gas proportional deteciors over the 0.5 m2 area surrounding each direct measurement location. Particular attention was given to cracks and joints in the surfaces and ,
walls, ledges, drains, and other locations where material may have accumulated. All detectors were coupled to ratemeters or ratemeter-scalers with audible indicators. Locations of elevated cirect radiation detected by scans were marked for further investigation. l Surface Activity Measurements i
Direct measurements for total beta activity were performed at 38 locations, representing each of the survey units within the repower area. Measurements were performed using gas proportional detectors, coupled to portable ratemeter-scalers. Smear samples, for determining l removable activity levels, were collected from each direct measurement location. Measurement and sampling locations are shown on Figures 4 and 5.
Exposure Rate Measurements Exposure rate measurements were performed within the general soil area of the repower area.
Exposure rates were measured at I m above surfaces at 10 locations using a pressurized ionization chamber (PIC). Background exposure rate measurements were performed using a PIC ron m. vemaan, co.we u.1995 4 l
l
_____________________._._____________J
4 at five locations within a 0.5 to 10 km radius of the site. Measurement locations are shown on Figures 6 and 8.
Soil Sampling Background soil samples were collected from each of the external background exposure rate -
measurement locations.
A total of ten soil samples was collected randomly from the general soil area within the repower area Soil sampling locations are shown on Figures 7 and 8.
~ SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data were returned to ESSAP's laboratory in Oak Ridge, Tennessee for analysis and interpretation. Soil samples were analyzed by solid state gamma spectrometry. Spectra were reviewed for Co-60, Cs-137, and any other identifiable photopeaks. Soil sample results were reported in units of picocuries per gram (pCi/g). Smear samples were analyzed for gross alpha and gress beta activity using a low background gas proportional counter, and the results converted to dpm/100 cm2 . Direct measurements for surface activity were converted to units 2
of disintegrations per minute per 100 square centimeters (dpm/100 cm ). Exposure rates were reported in units of microroentgens per hour ( R/h). Results were compared with the licensee's documentation and NRC guidelines established for release to unrestricted use, which are provided in Appendix C.
' FINDINGS AND RESULTS i DOCUMENT REVIEW ESSAP reviewed the licensee's final status survey report, including the f'mal status survey data and provided comments to the NRC.) The survey instrumentation and procedures used, including the assessment of background contributions to surface activity measurements, were discussed at Fort St. Vrain-Plansville.' CO . June 13.1995 5
i length. Guidelines for surface contamination and exposure rates were clearly stated, however, 3
radionuclide concentrations in soil that correspond to the 10 millirem (mrem) per year site-specific soll guideline were not specified. The site operational history, decommissioning activities, and final survey results provided sufficient information on the radiological status of the repower area.
SURVEY RESULTS j i
I ,
Surface Scans 3
Surface scans for beta activity on structure surfaces did not identify any locations of elevated 8 direct radiation. Surface scans for gamma activity within the general soil area also did not result in the identification of any locations of elevated direct radiation. l l
1 Surface Activity Measurements Surface activity measurements for total beta activity are summarized in Table 1. Total beta 2
activity levels for all measurement locations ranged from <340 to 710 dpm/100 cm ,
Removable activity levels were all less than the minimum detectable activity of the procedure 2
which was 12 dpm/100 cm2 for gross alpha and 16 dpm/100 cm for gross beta.
Exposure Rates Site exposure rates are summarized in Table 2. Gross exposure rates in the repower area ranged from 16.3 to 25.7 pR/h, and averaged 19 pR/h, at 1 m above the surface. The net exposure rates ranged from 0 to 9.4 pR/h, and averaged 2.7 pR/h. Background exposure rates ranged from 15.4 to 16.9 pR/h, and averaged 16.3 pR/h (Table 3).
j 1
Fort St. Vrain-Planeville. CO June 13,1995 6
i Radionuclide Concentrations in Soll Samnies Radionuclide concentrations in background samples are summarized in Table 3 and were <0.2 pCi/g for Co-60, <0.1 to 0.2 pCi/g for Cs-137,1.4 to 1.8 pCi/g for Th-228,1.5 to 2.2 pCi/g for Th-232, <0.1 pCi/g for U-235, and <2.1 pCi/g for U-238.
l Concentrations of radionuclides in surface soil samples collected randomly from the repower area summarized in Table 4. Radionuclide concentration ranges are as follows: <0.2 pCi/g for Co-60, < 0.1 pCi/g for Cs-137,1.1 to 2.0 pCi/g for Th-228,1.1 to 1.8 pCi/g for Th-232, < 0.1 pCi/g for U-235, and <2.3 pCi/g for U-238.
COMPARISON OF RESULTS WITH GUIDELINES The primary contaminants of concern for this site are beta-gamma emitters resulting from the operation of the FSV facility. The applicable NRC guidelines for beta-gamma emitters in unaffected areas are provided in Regulatory Guide 1.86.dThe guidelines are:
Total Activity 2
5,000 dpm/100 cm2 , averaged over a 1 m area 2
15,000 dpm/100 cm2 , maximum in a 100 cm area l
1 Removable Activity 1,000 dpm/100 cm2 Surface activity measurements for total and removable activity were all within the surface contamination guidelines.
The guideline values for radionuclide concentrations in soil are the radionuclide-specific concentrations which could result in an average annual total effective dose equivalent (TEDE) of 10 mrem to an individual in a population group exposed to radioactive material following decommissioning. These values may be determined in accordance with the methodology r n m. vem, co.w. u ms 7
t contained in NUREG/CR-5512, Volume 1 and as presented in NUREG-1500.5.6 Concentrations of radionuclides in soil san.ples are comparable to the concentrations measured in background samples (Tables 3 and 4). Therefore, compliance is demonstrated by the fact that soil samples collected from the repower area are indistinguishable from background levels.
The guideline for exposure rates, measured at 1 m above the surface, is 5 pR/h above
, background.7 With the exception of one elevated exposure rates measured in the northwert corner of the repower area (Figure 6, #5), all exposure rate were within the guideline. This elevated exposure rate measurement was due to dismantlement activities to remove the core support floor from the reactor vessel and to place it in a segmenting area on the refueling floor of the reactor building. Because the segmenting area is only separated from the repower area
' by sheet metal walls, the core support floor has temporarily affected exposure rates in the
- repower area, most notably in areas closest to the reactor building.8 The licensee performed exposure rate measurements before and after the core support floor l
move. Licensee exposure rates prior to the core support floor move were consistent with their background measurements of exposure rate, and thus indicated compliance with the exposure rate ,
guideline.8 Their results indicate that the average exposure rate in the repower area increased by about 3.5 pR/h following the core support floor move. An instrument comparison between i
ESSAP and the licensee performed during the confirmatory survey indicated good agreement between exposure rate measurements in the repower area (Table 2). Specifically, based on a pair-wise comparison t-test, there are no statistically significant differences (p>0.2) between l ESSAP's and the licensee's corrected exposure rate data.
1 Furthermore, soil sampling in the areas affected by these increased exposure rates resulted in l no indication of soil activity in excess of background levels.
l I
I Fort SL vrain Planeville. CO - June 13,1995 8 l
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SUMMARY
During the period March 20 through 22,1995, at the request of the NRC's Division of Waste Management, NMSS, the Environmental Survey and Site Assessment Program of ORISE i performed a confirmatory survey at the Fort St. Vrain site in Platteville, Colorado. Survey activities included document reviews, surface scans, surface activity measurements, exposure rate measurements, and soil sampling.
The confirmatory survey identified one location within the repower area that exhibited an elevated exposure rate measurement. This location within the repower area was influenced by i elevated radiation from dismantling activities on the core support floor in the repower area. Soil sampling in this area confirmed that the elevated exposure rate measurement was not the result
' of elevated soil concentrations. The confirmatory survey results are consistent with those obtained by the licensee and support the licensee's conclusion that residual activity levels in the
! repower area satisfy the guidelines for release to unrestricted use.
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TABLE 1
SUMMARY
OF SURFACE ACTIVITY MEASUREMENTS REPOWER AREA FORT ST. VRAIN PLATTEVILLE, COLORADO f
Range of Number of Range of Total Removable Activity Location
- Measurement Beta Activity (dpm/100 cm 2)
Locations (dpm/100 cm2)
Concrete Slab < 420 < 12 < 16 6
(at security fence)
Miscellaneous Concrete 3 < 420 < 12 < 16 Valve Pit 3 < 420 - 710 < 12 < 16 Miscellaneous Metal 4 < 340 < 12 < 16 Evaporative Cooler Bldg. 11 < 340 < 12 < 16 (East and South Walls)
Evaporative Cooler Bldg. < 420 < 12 < 16 11 (East and South Foundation)
- Refer to Figures 4 and 5.
re st vr mma, co a is.1995 18
! l TABLE 2 l EXPOSURE RATES REPOWER AREA FORT ST. VRAIN PLATI'EVILLE, COLORADO Gross Exposure Rates ( R/h) ESSAP Net Exposure tion, at 1 m Above Surface Rates ( R/h) at 1 m Above Surface' ESSAP Licensee6 I 17.7 17.00 1.4 2 18.0 18.27 1.7 .
3 20.5 22.66 4.2 4 18.3 21.88 2.0 5 25.7 31.65 9.4 6 21.0 21.09 4.7 7 19.0 18.65 2.7 8 17.1 16.54 0.8 9 16.4 15.47 0.1 10 16.3 14.55 0 l
' Refer to Figure 6.
6 Licensee gross exposure rates in repower area were measured with a NaI detector. PIC correction factor of 0.64 was applied to these measurements.
' Net exposure rates were determined by subtracting the background exposure rate (16.3 R/h) from each gross exposure rate.
re st wwwmme, co . we u.1995 19
TABLE 3 BACKGROUND EXPOSURE RATES AND RADIONUCLIDE CONCENTRATIONS IN SOIL REPOWER AREA FORT ST. VRAIN PLATTEVILLE, COLORADO Radionuclide Concentration (pCl/g) R te h)
Location" at 1 m Th-228 Th-232 U-235 U-238 Above Co-60 Cs-137 Surface 1 < 0.1 < 0.1 1.8 i 0.2 b 1.7 0.4 < 0.1 0.9 1.4 16.9 2 < 0.2 < 0.1 1.5 0.2 1.6 0.4 < 0.1 < 2.1 16.3 3 < 0.1 < 0.1 1.4 0.1 1.8 0.4 < 0.1 < 1.5 15.4 4 < 0.1 < 0.1 1.6 0.1 1.5 i 0.5 < 0.1 < 1.9 16.7 5 < 0.1 0.2 0.1 1.7 0.2 2.2 0.6 < 0.1 <1.8 16.3
' Refer to Figure 8.
bUncertainties represent the 95% confidence level, based only on counting statistics.
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Fort St. Vrain Platteville, CO . June 13, ;995 20
~
TABLE 4 RADIONUCLIDE CONCENTRATIONS IN SOIL SAMPLES REPOWER AREA FORT ST. VRAIN PLA'ITEVILLE, COLORADO Radionuclide Concentration (pC1/g)
Co-60 Cs-137 Th-228 Th-232 U-235 U-238 1 < 0.1 < 0.1 1.6 0.26 1.5 9.5 < 0.1 <1.7 2 < 0.1 < 0.1 1.7 0.2 1.6 i 0.5 < 0.1 < 2.1 3 < 0.1 < 0.1 1.5 0.1 1.6 i 0.4 < 0.1 0.9 1.3 4 <0.2 < 0.1 2.0 0.2 1.8 0.4 < 0.1 < 2.3 5 <0.1 < 0.1 1.5 i 0.1 1.6 i 0.4 < 0.1 1.0 i 0.9 6 <0.1 <0.I 1.7 i 0.1 1.6 0.4 < 0.1 < 1.4 7 < 0.2 < 0.1 2.0 0.2 1.8 0.4 < 0.1 < 2.2 8 <0.1 < 0.1 1.9 0.2 1.8 0.6 < 0.1 < 1.7 9 <0.2 < 0.1 1.6 0.2 1.4 0.4 < 0.1 < 2.1 10 <0.1 < 0.1 1.1 i 0.1 1.1 0.4 <0.1 < 1.4
- Refer to Figure 7.
b Uncertainties represent the 95% confidence level, based only on counting statistics.
Fort St. Vram-Plauevilk CO June 13.1995 21
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4 REFERENCES l
- 1. Public Service Company of Colorado," Final Status Survey Plan and Report," Cintichem,
- Inc., December 5,1994.
- 2. Oak Ridge Institute for Science and Education " Confirmatory Survey Plan for the j Repower Area, Fort St. Vrain, Platteville, Colorado (Docket No. 50-267)," March 17,
. 1995.
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! 3. Oak Ridge Institute for Science and Education, letter from E. W. Abelquist to l D. Fauver, NRC/NMSS,' " Document Review - Final Survey Report for Release of the Repower Area, Fort St. Vrain, Platteville, Colorado (Docket No. 50-267),"
March 13,1995.
4 1 4. U.S. Nuclear Regulatory Commission, " Termination of Operating Licenses for Nuclear Reactors," Regulatory Guide 1.86, Washington, D.C., June 1974.
4 5. NUREG/CR-5512, " Residual Radioactive Contamination from Decommissioning,"
Volume 1, October 1992.
- 6. NUREG-1500, " Working Draft Regulatory Guide on Release Criteria for Decommissioning: NRC Staff's Draft for Comment," August 1994.
- 7. Public Service Company of Colorado, letter from D. Waremburg to J. Austin (NRC),
" Final Survey Plan for Site Release," February 17, 1994.
- 8. Public Service Company of Colorado, letter from M. Fisher to M. Weber (NRC),
" Transitory. Radiation Levels in Fort St. Vrain Repower Area", March 15, 1995.
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I APPENDIX A MAJOR INSTRUMENTATION i
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l Fort St. Vram Pinuevik, Co. June 13,1995
APPENDIX A l
MAJOR INSTRUMENTATION l 1
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The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the authors or their employers.
DIRECT RADIATION MEASUREMENT :
Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM)
Ludlum Ratemeter-Scaler Model 2221 l (Ludlum Measurements, Inc.,
Sweetwater, TX)
Detectors Ludlum Gas Proportional Detector Model 43-68 Effective Area,100 cm 2 (Ludlum Measurements, Inc.,
Sweetwater, TX)
' Reuter-Stokes Pressurized lon Chamber l
.Model RSS-111 (Reuter-Stokes, Cleveland, OH)
Victorcen NaI Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal (Victoreen, Cleveland, OH)
I,ABORATORY ANALYTICAL INSTRUMENTATION
- igh Purity Extended Range Intrinsic Detectors Model No
- ERVDS30-25195 (Tennelec, Oak Ridge, TN)
Used in conjunction with:
Lead Shield Model G-ll (Nuclear Lead, Oak Ridge, TN) and von m. ven m., co.w n, ms A-1
l Multichannel Analyzer 3.100 Vax Workstation (Canberra, Meriden, CT)
High-Purity Germanium Detector
- Model GMX-23195-S,23% Eff.
(EG&G ORTEC, Oak Ridge, TN)
Used in conjunction with:
Lead Shield Model G-16 (Gamma Products, Palos Hills, IL) and l Multichannel Analyzer I
3100 Vax Workstation (Canberra, Meriden, CT) f Low Background Gas Proportional Counter Model LB-5110-W (Oxford, Oak Ridge, TN) 1 i
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- Part R. Vrain-Finesvius, CO - June 13,1995 A-2 ,
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APPENDIX B SURVEY AND ANALYTICAL PROCEDURES ;
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APPENDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES Surface Scans Surface scans were performed by passing the probes slowly over the surface; the distance between the probe and the surface was maintained at a minimum-nominally about 1 cm. Hand-held gas proportional detectors were used to scan the structural surfaces. Identification of elevated levels was bar,ed on increases in the audible signal from the recording ar.d/or indicating instrument. Combinations of detectors and instruments used for the scans were:
Beta -
gas proportional detector with ratemeter-scaler ;
Gamma -
NaI scintillation detector with ratemeter l
l Surface Activity Measurements l
l Measurements of total beta activity levels were performed using gas proportional detectors with ratemeter-scalers. Count rates (cpm), which were integrated over 1 minute in a static position, were converted to activity levels (dpm/100 cm 2) by dividing the net rate by the 41r efficiency and correcting for the active area of the detector. The beta activity background count rate for the l
gas proportional detectors was 507 and 761 cpm on metal and concrete surfaces, respectively. l
! The beta efficiency factors ranged from 0.24 to 0.25 for the gas proportional detectors calibrated 2
i to Tc-99. The probe area for the gas proportional detectors is 126 cm , j Soil Sampling l
Approximately I kg of soil was collected at each sample location. Collected samples were placed in a plastic bag, sealed, and labeled in accordance with ESSAP survey procedures.
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a l ANALYTICAL PROCEDURES l
1 Gamma Spectrometry Samples of soil materials were dried, mixed, crushed, and/or homogenized as necessary, and a portion sealed in 0.5-liter Marinelli beaker or other appropriate container. The quantity placed in the beaker was chosen to reproduce the calibrated counting geometry. Net material weights were determined and the samples counted using intrinsic germanium detectors coupled to a pulse J l
height analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system.
All photopeaks associated with the radionuclides of concern were reviewed for consistency of activity. Energy peaks used for determining the activities of radionuclides of concern were:
l Co-60 1.173 MeV Cs-137 0.662 MeV Th-228 0.239 MeV from Pb-212* ,
1 Th-232 0.911 MeV from Ac-228*
U-235 0.186 MeV l
- U-238 0.063 MeV from Th-234*
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- Secular equilibrium assumed.
Spectra were also reviewed for other identifiable photopeaks.
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UNCERTAINTIES AND DETFETION LIMITS The uncertainties associated with the analytical data presented in the tables of this report represent the 95% confidence level for that data. These uncertainties were calculated based on both the gross sample count levels and the associated background count levels. Additional j uncertainties, associated with sampling and measurement procedures, have not been propagated into the data presented in this report.
Detection limits, referred to as minimum detectable activity (MDA), were based , on 2.71 plus 4.65 times the standard deviation of the background count [2.71 + (4.65v BKG)]. When the activity was determined to be less than the MDA of the measurement procedure, the result was reported as less than MDA. Because of variations in background levels, measurement !
efficiencies, and contributions from other radionuclide in samples, the detection limits differ from sample to sample and instrument to instrument.
CALIBRATION AND QUALITY ASSURANCE Calibration of all field and laboratory instrumentation was based on standards / sources, traceable to NIST, when such standards / sources were available. In cases where they were not available, i standards of an industry recognized organization were used. Calibration of pressurized ionization chambers was performed by the manufacturer.
Analytical and field survey activities were conducted in accordance with procedures from the following documents of the Environmental Survey and Site Assessment Program:
- Survey Procedures Manual, Revision 8 (December 1993) e Laboratory Procedures Manual, Revision 9 (January 1995)
- Quality Assurance Manual, Revision 7 (January 1995)
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The procedures contained in these manuals were developed to meet the requirements of DOE i
Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to assess j processes during their performance.
1 Quality control procedures include:
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- Daily instrument backgrcund and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.
i e Participation in EPA and EML laboratory Quality Assurance Pograms. l i
- Training and certification of all individuals performing procecures.
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- Periodic internal and external audits.
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Fort St. Vrain-Finaeville, CO - June 13, 1995 B-4
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APPENDIX C l REGULATORY GUIDE 1.86, TERMINATION OF OPERATING
! LICENSES FOR NUCLEAR REACTORS J
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Fost St. Vrain-Pisaevale, CO - June 13,1995 4
e U.S. ATOMIC ENERGY COMMISSION June 1974 REGULATORY DIRECTORATE OF REGULATORY STANDARDS GUIDE REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION important to the safety of reactor operation is no longer required. Once this possession-only license is issued, Section 50.51, " Duration oflicense, renewal," of 10 reactor operation is not permitted. Other activities CFR Part 50, " Licensing of Production and Utilization from the reactor and placing it in storage (either onsite Facilities," requires that each license to operate a or offsite) may be continued.
production and utilization facility be issued for a specified duration. Upon expiration of the specified A licensee having a possession-only license must period, the license may be either renewed or terminated retain, with the Part 50 license, authorization for by the Conunission. Section 50.82, " Applications for special nuclear material (10 CFR Part, 70, "Special termination oflicenses," specifies the requirements that Nuclear Material"), byproduct material (10 CFR Part must be satisfied to terminate an operating license, 30, " Rules of General Applicability to Licensing of including the requirement that the dismantlement of the Byproduct Material"), and source material (10 CFR facility and disposal of the component parts not be Part 40, " Licensing of Source Matericl"), until the inimical to the common defense and security or to the fuel, radioactive components, and sources are removed health and safety of the public. This guide describes from the facility. Appropriate administrative controls methods and procedures considered acceptable by the and facility requirements are imposed by the Part 50 Regulatory staff for the termination of operating license and the technical specifications to assure that licenses for nuclear reactors. The advisory Committee proper surveillance is performed and that the reactor on Reactor Safeguards has been consulted concerning facility is maintained in a safe condition and not this guide and has concurred in the regulatory position, operated.
B. DISCUSSION A possession-only license permits various options and procedures for decommissioning, such as When a licensee decides to terminate his nuclear mothballing, entombment, or dismantling. The reactor operating license, he may, as a first step in the requirements imposed depend on the option selected.
process, request that his operating license be amended to restrict him to possess but not operate the facility. Section 50.82 provides that the licensee may The advantage to the licensee of converting to such a dismantle and dispose of the component parts of a possession-only license is reduced surveillance nuclear reactor in accordance with existing regulations, requirements in that periodic surveillance of equipment For research reactors and critical facilities, this has USAEC REGULATORY GUIDES cepies of pimiste udes mem obtere w tw eeung m viem Regulatory Gudee are issued to describe end make availabie te the public * "" M" 8Y *' " " "'"8"" I
- methode sceeptable te the AEC esgulatery sta'f of unplemenung specific parte Anonuen: Dwector of Regulatory Stenderde. Co'mments and'suggestiene for
""**" '" *** 8*d** e enceweged and should be eent to W of the Commmeien's reguienene, to eregmate techswoues used by the staff m Se retary of the Cemrmes on. U.S. Atomic Energy commmmen, Weehangton.
evoluotwig erecific protisema or postulated accidente, er to provide gudence to O.C. 2o64 Anemien: Ch.sf. Pubsee Proceedinge Staff.
esplicante. Regdetery Guides are not substitutes for regulatione and campsience with them e not reeweed. Methode end celutione defferent from those set out in the guidee will be acceptable if they provide e besie for the f eewete 80 the isevence er sentmuence of a portat er heenes by the p
- 2. Reeeerch and teet Reacters 7. Trenopertation R h.d g e .. be rev d -r,0.sc ly. e. p e-ste. ,e ._ed.te t-~~~~ . o--o-s _ e .nd ,ere,w ,r.. m,.rm. tion .,s.pe. _ e.
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4 usually meent the disassembly of a reactor and its minimize releases of radioactivity from the freility, shipment organization for further use. The site from which a reactor has been removed must be c. Any proposed changes to the technical - ,
decontaminated, as necessary, and inspected by the specifications that reflect the possession-only facility Commission to determine whether unrestricted access status and the necessary disassembly /retirem.mt l can be approved. In the case of nuclear power activities to be performed.
reactors, dismantling has usually been accomplished by )
shipping fuel offsite, making the reactw inoperable, d. A safety analysis of both the activities to be i and disposing of some of the radioactive components. accomplished and the proposed changes to the technical specifications.
' Radioactive components may be either shipped off-site for burial at an authorized burial ground or e. An inventory of activated materials and their l secured on the site. Those radioactive materials location in the facility. ]
remaining on the site must be isolated from the public l by physical barriers or other means to prevent public 2. ALTERNATIVES FOR REACTOR I access to hazardous levels of radiation. Surveillance is RETIREMENT necessary to assure the long term integrity of the barriers. The amount of surveillance required depends Four alternatives for retirement of nuclear reactor upon (1) the potential hazard to the health and safety of facilities are considered acceptable by the the public from radioactive material remaining on the Regulatory staff. These are:
site and (2) the integrity of the physical barriers.
Before areas may be released for unrestricted use, they a. Mothballing. Mothballing of a nuclear reactor must have been decontaminated or the radioactivity facility consists of putting the facility in a state of must have decayed to less than prescribed limits protective storage. In general, the facility may be (Table 1). left intact except that all fuel assemblies and the radioactive fluids and waste should be removed The hazard associated with the returned facility is from the site. Adequate radiation monitoring, evaluated by considering the amount and type of environmentai surveillance. and appropriate security remaining contamination, the degree of confinement of procedures should be established under a the remaining radioactive materials, the physical possession-only license to ensure that the health and security provided by the confinement, the susceptibility safety of the public is not endangered. l to release of radiation as a result of natural phenomena, )
and the duration of required surveillance, b. In-Place Entomhment. In-place entombment consists of sealing all the remaining highly
! C. REGULATORY POSITION radioactive or contaminated components (e.g., the pressure vessel and reactor internais) within a
! 1. APPLICATION FOR A LICENSE TO POSSESS structure integral with the biological shield after l BUT NOT OPERATE (POSSESSION-ONLY having all fuel assemblies, radioactive fluids and LICENSE) wastes, and certain selected components shipped offsite. The structure should provide integrity over A request to amend an operating license to a the period of time in which significant quantities I possession-only license should be made to the Director (greater than Teble 1 levels) of radioactivity remain
[ of Licensing. U.S. Atomic Energy Commission, with the material in the entombment. An' Washington, D.C. 20545. The request should include appropriate and continuing surveillance program
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! the following information: should be established under a possession-only license,
- a. A description of the current status of the facility,
- c. Removal of Radioactive. Components and
- b. A description of measures that will be taken to Dismantling. All fuel assemblies, radioactive fluids prevent criticality or reactivity changes and to and waste, and other materials having activities NitM Section electronically reproduced from photocopy. C-2
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above accepted unrestricted cctivity levels (Teble 1) and access openings, should be inspected at le:st should be removed from the site. The facility quarterly to assure that these barriers have not owner may then have unrestricted use of the site deteriorated and that locks and locking apparatus are l with no requirement for a license. If the facility intact.
owner so desires, the remainder of the reactor facility may be dismantled and all vestiges removed c. A facility radiation survey should be performed and disposed of. at least quarterly to verify that no radioactive material is escaping or being transported through the I
- d. Conversion to a New Nuclear Systesn or a containment barriers in the facility. Sampling should I
Fossil Fuel System. This alternative, which applies be done along the most probable path by which only to nuclear power plants, utilizes the existing radioactive material such as that stored in the inner turbine system with a new steam supply system. containment regions could be transported to the outer The original nuclear steam supply system should be regions of the facility and ultimately to the environs.
separated from the electric generating system and disposed of in accordance with one of the previous d. An environmental radiation survey should be three retirement alternatives. performed at least semiannually to verify that no significant amounts of radiation have been released to
- 3. SURVEILLANCE AND SECURITY FOR Tile the environment from the facility. Samples such as RETIREMENT ALTERNATIVES WIIOSE soil, vegetation, and water should be taken at locations FINAL STATUS REQUIRES A for which statistical data has been established during POSSESSION-ONLY LICENSE reactor operations.
A facility which has been licensed under a e. A site representative should be designated to be possession-only license may contain a significant responsible for controlling authorized access into and amount of radioactivity in the form of activated and movement within the facility.
contaminated hardware and structural materials.
Surveillance and commensurate security should be f. A<lministrative procedures should be established provided to assure that the public health and safety are for the notification and reporting of abnormal not endangered. occurrences such as (1) the entrance of an unauthorized
- a. Physical security to prevent inadvertent exposure person or persons into the facility and (2) a significant of personnel should be provided by multiple locked change in the radiation or contamination levels in the barriers. The presence of these barriers should make facility or the offsite environment. i it extremely difncult for an unauthorized person to gain i access to areas where radiation or contamination levels g. The following reports should be made:
l exceed those specified in Regulatory Position C.4. To prevent inadvertent exposure, radiation areas above (1) An annual report to the Director of 5 mR/hr, such as near the activated primary system of Licensing, U.S. Atomic Energy Commission, a power plant, should be appropriately marked and Washington, D.C. 20545, describing the results of the should not be accessible except by cutting of welded environmental and facility radiation surveys, the status )
closures or the disassembly and removal of substantial of the facility, and an evaluation of the performance of structures and/or shielding material. Means such as a security and surveillance measures.
i remote readout intrusion alarm system should be provided to indicate to designated personnel when a (2) An abnormal occurrence report to the ,
physical barrier is penetrated. Security personnel that Regulatory Operations Regional Office by telephone j provide access control to the facility may be used within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery of an abnormal l instead of the physical barriers and the intrusion alarm occurrence. The abnormal occurrence will also be systems. reported in the annual report described in the preceding itera.
. b. The physical barriers to unauthorized entrance into the facility, e.g., fences, buildings, welded doors, b. Records or logs relative to the following items Nm: Section electronically reproduced from pnotocopy. C-3
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e should be kept and retained until the license is licensee to relinquish possession or control of premises, terminated, after which they must be stored with other equipment, or scrap having surfaces contaminated in plant records: excess of the limits specified. This may include, but is <
not limited to, special circumstances such as the (1) Environmental surveys, transfer of premises to another licensed organization that will continue to work with radioactive materials. l (2) Facility radiation surveys, Requests for such authorization should provide:
(3) Inspections of the physical barriers, and (4) Abnormal occurrences.
(1) Detailed, specific information describing the premises, equipment, scrap, and radioactive
- 4. DECONTAMINATION FOR RELEASE FOR contaminants and the nature, extent, and degree of UNRESTRICTED USE residual surface contamination.
If it is desired to terminate a license and to (2) A detailed health and safety analysis indicating eliminate any further surveillance requirements, the that the residual amounts of materials on surface areas, facility should be sufficiently decontaminated to prevent together with other considerations such as the risk to the public health and safety. After the prospective use of the premises, equipment, or scrap, decontamination is satisfactorily accomplished and the are unlikely to result in an unreasonable risk to the site inspected by the Commission, the Commission may health and safety of the public.
authorize the license to be terminated and the facility abandoned or released for unrestricted use. The e. Prior to release of the premises for unrestricted licensee should perform the decontamination using the use, the licensee should make a comprehensive following guidelines: radiation survey establishing that contamination is within the limits specified in Table 1. A survey report
- a. The licensee should make a reasonable effort to should be filed with the Director of Licensing, U.S.
eliminate residual contamination. Atomic Energy Commission, Washington, D.C. 20545, with a copy to the Director of the Regulatory
- b. No covering should be applied to radioactive Operations regional Office having jurisdiction. The surfaces of equipment of structures by paint, plating, or report should be filed at least 30 days prior to the other covering material until it is known that planned date of abandonment. The survey report contamination levels (determined by a survey and should:
documented) are below the limits specified in Table 1.
In addition, a reasonable effort should be made (and (1) identify the premises; documented) to further minimize contamination prior to any such covering. (2) Show that reasonable effort has been made to reduce residual contamination to as low as practicable
- c. The radioactivity of the interior surfaces of levels; pipes, drain lines, or ductwork should be determined by making measurements at all traps and other (3) Describe the scope of the survey and the general appropriate access points, provided contamination at procedures followed; and these locations is likely to be representative of contamination on the interior of the pipes, drain lines, (4) State the finding of the survey in units specified
. or ductwork. Surfaces of premises, equipment, or in Table 1.
j scrap which are likely to be contaminated but are of l
' After review of the report, the Commission may such size, construction, or location as to make the J surface inaccessible for purposes of measurement inspect the facilities to confirm the survey prior to I- ' should be assumed to be contaminated in excess of the granting approval for abandonment, permissible radiation limits.
- d. Upon request, the Commission may authorize a 5. REACTOR RETIREMENT PROCEDURES Nite: Section electronically reproduced from phott' copy. C-4 i
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) As indicated in Regulatory Position C.2, several alternatives are acceptable for reactor facility retirement.. If minor disassembly or "mothballing" is planned, this could be done by the existing operating ,
and maintenance procedures under the license in effect. l l Any planned actions involving an unreviewed safety l question of a change in the technical specificatior.s should be reviewed and approved in accordance with
- the requirements of 10 CFR i 50.59.
If major structural- changes - to radioactive components of the facility are planned, such as removal of the pressure vessel or major components of the primary system, a dismantlement plan including the j information required by I 50.82 should be submitted to the Commission. A dismantlement plan should be submitted for all the alternatives of Regulatory Position C.2 except mothballing. However, minor disassembly J activities may still be performed in'the absence of such s , plan,' provided they are permitted by existing operating and maintenance procedures. A i dismantlement plan should include the following:
- a. A description of the ultimate status of the facility
- b. ' A description of the dismantling activities and
._ the precautions to be taken.
- c. A. safety analysis of the dismantling activities i including any effluents which may be released. ,
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- d. A safety analysis of the facility in its ultimate status. .
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Upon satisfactory review and approval of the :
, dismantling plan, a dismantling order is issued by the )
Commission in accordance with i 50.82. When dismantling is completed and the Commission has been notified by letter, the appropriate Regulatory Operations Regional Office inspects the facility and verifies - completion in accordance with the dismantlement plan. If residual radiation levels do not cxceed the values in Table 1, the Commission may terminate the license. If possession-only license under which the dismantling activities have been conducted er, as an alternative, may make application to the State
' (if an : Agreement - State) for' a byproduct materials .
. license.
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TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS Nuclide* Average
- Maximum *d RemovableA '
U-nat, U 235, U 238, and essociated decay products 5,000 dpm a/100 cm 2 15,000 dpm a/100 cm2 1,000 dpm a/100 cm2 Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, Ac 227, I-125,1-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th-nat, Th-232, Sr-90, Ra-223, Ra 224, U-232, I-126, I 131, I-133 1,000 dpm/100 cm2 3,000 dpm/100 cm2 200 dpm/100 cm2 Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above. 5,000 dpm Sy/100 cm2 15,000 dpm Sy/100 cm2 1,000 dpm Sy/100 cm2
'Where surface contamination by both alpna- and beta-garmna-emitting nuclides exists, the limits established for alpha- and beta- gamma-emitting nuclides should apply independently.
'As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors essociated with the instrumentation.
'Meesurements of average contaminant should not be averaged over more than 1 square meter. For objects of less surface erez, the average should be derived for each such object.
- The maximum contamination level applies to an area of not more than 100 cm2 .
"The amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
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