ML022750212
| ML022750212 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 09/30/2002 |
| From: | GPU Nuclear |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| -nr | |
| Download: ML022750212 (137) | |
Text
SNEC FACILITY LICENSE TERMINATIONPLAN REVISION I 3.0 IDENTIFICATION OF REMAINING DISMANTLEMENT ACTIVITIES
3.1 INTRODUCTION
As discussed previously, this final phase 'of the&SNEC Facility Decommissioning Project commenced in April 1998 following NRC approval of a License Amendment authorizing decommissioning. "Since that time, the 'greater-,art of 1998 w-as devoted:to removing and shipping the SNEC Facility Large Components (i.e. Reactor Vessel, Pressurizer and Steam Generator).
Following removal of the larg'ecomponents, the latter-part of 1998"and the beginning of 1999 were devoted to removing the bulk of the remainder of the systems from the SNEC Facility., Following that, concrete decontamination commen6ed and'continued into2000 when it was determined that complete removal of the concrete from the Containment 'Vessel (CV) would be required in order to releafse'the site in ac'oraarice iwith'NRC Regulati6ns (10 CFR Part 20.1402, Radiological Criteria for Unrestricted Use). Concrete removal is ongoing.
As described in Section 2.2.4 the site is divided into eight areas in describing radiological
"")riditionis. 'These 'same area desighations will 'be' used in describing "remainirng site dismantlement activities.
FIGURE 3-1 Photograph of SNECFailiity CV,& DSF Buildings 3-1 REVISION 1
,SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 3.2 REMAINING TASKS 3.2.1 Area I (CV Basement) Remaining Site Dismantlement Activities Area 1 will be completely removed to the steel CV liner.
3.2.2 Area 2 (Primary Compartment) Remaining Site Dismantlement Tasks Area 2 will be completely removed to the steel CV liner.
3.2.3 Area 3 (Auxiliary Compartment) Remaining Site Dismantlement Activities Area 3 will be completely removed to the steel CV liner.
3.2.4 Area 4 (Operating Floor) Remaining Site Dismantlemient Activities Remaining tasks in Area 4 include removal of miscellaneous steel, removal of the polar crane and its girders and disconnection of the Containment Ventilation System. Additionally, some volumetrically contaminated concrete remains to be removed.
The polar crane is the last large component to be removed from the SNEC Facility. Plans for Polar Crane removal have not been finalized. It can be removed using conventional cutting and rigging techniques and will most likely be dispositioned as radioactive waste.
The Containment Vessel and the Containment Vessel Ventilation System are special cases.
The SNEC Facility License' requires' the interior of the Containment Vessel to be maintained as an exclusion area and the ventilation system to be operating during activities that have the potential to cause a measurable release. Thus, prior to removal of the upper dome of the Containment Vessel and the Containment Vessel Ventilation System the SNEC Technical Specifications will need to be revised. A change request to the SNEC Facility Technical Specifications (Reference 3-1) has been submitted to permit this activity. Additionally, prior to the removal of the upper dome of the Containment Vessel, the Final Status Survey (FSS) for the bottom head of the Containment Vessel will be completed and the area backfilled prior to removal of the upper dome. During removal of the upper dome, the backfilled portion of the lower Containment Vessel will be protected by an impermeable membrane and the area resurveyed after completion of the removal of the upper dome.
3.2.5 Area 5 (Pipe Tunnel) Remaining Site Dismantlement Activities As described in Section 2.2.2, Area 5 consisted of a concrete tunnel, which extended 235 degrees around the CV. The tunnel has been removed except for the portion under the Material Handling Bay (MHB) section of the Decommissioning Support Facility.
The portion of the tunnel under the MHB will be removed and the area surveyed and remediated as appropriate to meet site, release levels.
3.2.6 Area 6 (Reactor Compartment and Storage Well) Remaining Site Dismantlement Activities Area 6 will be completely removed to the steel CV liner.
3-2
-1
lNIV I M&1l1l 'T I ICFM l I TERMINATION PiLAN iREVISION 1
3.2.7 Area'7 (Exterior of Containment Vessel Dome) Remaining Site Dismantlement Activities
,Remaining tasks in Area 7 consist of removing the CV Dome to at least three feet below grade and completing the requisite Final Status Survey on the remaining below grade portions of the CV. Remediation, if needed, will be performed as appropriate.
3.2.8 Area 8 (SNEC Facility Yard Areas)'Remaining Site Dismantlement Activities Remaining tasks in Area 8, with the exception of the Saxton Steam Generating Station (SSGS) area will consist of some soil removal and'removal of utilities located outside the CV.
"The SSGS was used by the'SNEC Facility to support its operation, generate electricity and to provide dilution.flow for radioactive effluent liquid discharges. Surveys in this area indicate there was some residual contamination that needed to be removed.,Thisincluded water and sediment removal, some concrete decontamination, and extensive embedded, piping -surveying and removal. These activities were completed in 2001. The small amount of remaining embedded piping will be addressed during the FSS design process.
3.3-ACCESS CONTROL MEASURES Since all decommissioning activities may not be completed prior to thestart of final status survey work,' measures as described in Section 5.3,will be implemented to protect survey areas from recontamination during-and after the final status survey.' In all cases, decommissioning activities creating a potential for the spread of contamination will be completed within each survey area prior to starting a final status survey in the area. Additionally,"de6ommissioning activities that create a potential for the spread of contamination to adjacent survey areas will be evaluated and controlled using 'barriers,' covers; or restricting or rescheduling activitieS.
3.4 10 CFR 50.59 REVIEW 10 CFR 50.59 "Changes, Tests and Experiments" permits-licensees to'-rhake chianges to the facility as described in the Safety Analysis Report (as updated), make changes in procedures as described in-the-Safety Analysis Report (as 'updated), and co6duct' tests-or experiments not described in the Safety Analysis Report (as updated) without obtaining a license amendment under certain conditions.
With respect to the SNEC Facility this License Termination Plan serves to support a License Amendment to perform the Final Status Survey and terminate the license. In addition as indicated in Section 3.2.4 above a change request to the SNEC Facility Technical Specifications (Reference 3-1) has been submitted to allow removal of the upper Containment Vessel dome and the Containment Vessel Ventilation System. The remaining tasks described in Section 3.2 are bounded by the Definition for Decommissioning Activities in the SNEC Facility Technical Specifications (Reference 3-3) and are thus permitted by the Technical Specifications.
Additionally these tasks have been reviewed against the criteria of 10 CFR 50.59 and are bounded by the SNEC Facility USAR (Reference 3-4).
Thus, although some changes will be made to the facility as described in the SNEC Facility USAR, e.g. Removal of the Polar Crane, these activities were anticipated to be performed.
Further, greater than 99% of the curies present at the SNEC Facility at the commencement of K>
decommissioning in 1998 have been safely removed from the site and shipped for proper 3-3
SNEC FACILITY LICENSE TERMINATION PLAN PF'JIQ~!AM I disposition.
Finally, each activity requiring a procedure as defined by the SNEC Facility Technical Specifications is reviewed to ensure it satisfies the criteria of 10 CFR 50.59. Thus, with the exception of the removal of the Containment Vessel upper dome and the Containment Vessel Ventilation System,, none of the remaining tasks identified in Section 3.2 requires prior NRC approval in accordance with 10 CFR 50.59.
3.5 DECOMMISSIONING TASKS REQUIRING COORDINATION WITH OTHER FEDERAL OR STATE REGULATORY AGENCY Prior to License Termination, tasks requiring coordination with any other Federal or State Regulatory Agency include disposal of the water in Area 8. Additionally, some water treatment will be necessary for water collecting in excavated areas e.g., the yard area North of the CV.
Discharge of this water will be controlled radiologically in accordance with the requirements of the ODCM (Reference 3-2). Non-radioactive sampling and analysis will be in accordance with a program accepted by the Commonwealth of Pennsylvania. Additionally, the Army Corps of Engineers gave its consent for locating temporary on-site excavations and infiltration galleries.
3.6
SUMMARY
The tasks remaining for site dismantlement are permitted by the SNEC Facility License Technical Specifications and USAR. Based on current estimates, these tasks will be completed in the second quarter of 2003 for an estimated exposure of approximately 37 person-rem (Table 3.1). Of this total approximately 35 person-rem has been expended to date. Estimated remaining low level radioactive waste generation is provided in Table 3.2.
3.7 REFERENCES
3-1 SNEC Facility Technical Specification Change Request No. 62, April 2002.
3-2 SNEC Procedure E900-PLN-4542.08 "Off-Site Dose Calculation Manual" 3-3 SNEC Facility Technical Specifications 3-4 Updated Safety Analysis Report for Decommissioning the SNEC Facility 3-4 R l:Vlq II*'I H 4
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I Table 3.1 SNEC Facility Decommissioning Person-Rem Estimate Task Person-Rem Asbestos Remediation 2.97 System Dismantlement 12.83 Large Component Removal 7.38 Structure D&D 4.17 Waste Management 2.13 Miscellaneous Support Activities 2.41 Scaffold & Shielding 5.32 Characterization Activities 0.63 Total 37.84 Table 3.2 SNEC Facility Low Level Radioactive Waste Projection Type Quantity Metal 50,000 lbs.
Soil 500 ft3 Water 1600 Gal Sediment 50 ft3 Concrete 5,750,000 lbs.
Dry Active Waste (DAW) 2000 ft3 Debris 15,000 lbs.
3-5 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
.SNEC FACILITY LICENSE TERMINATION PLAN RVSO 4.0 REMEDIATION PLANS 4.1
-INTRODUCTION
- The ultimate goal of the decommissioning project at the SNEC Facility is to release the site in
,accordance with NRC 'regulations (10 CFR Part 20.1402, Radiological Criteriafor Unrestricted Use). Release,of the-site requires assurance that any future use of the sitewill not result in exposing individuals to unacceptable levels of radiation. The site-release criteria require the Total Effective Dose Equivalent (TEDE) to an average-member of the critical population group from residual contamination to be less than 25 mrem/year. In addition, plant-related contamination in groundwater and surface water with the potential to be used as a source of drinking,vwater will be 'evaluated against the 4 mrem/-year dose criteria in the National Primary Drinking Water Standards contained in 40 CFR 141. Remediation of someareas of the site will be necessary in order to meet these release criteria.'Chapter 2 of the LTP provides information on the radionucllides present at the SNEC Facility as well as their location, activity and radiation levels. Section 5.4.4 of the LTP describes the data investigation process used to evaluate the "effectiveness -'of r6emediatibn 'to determine wheth& further investigation -or"remediation is
'required to'ensure that an area 'meets the site release criteria: Section,4.3 a_16d_4.4 identify the remediation methods that may be used and describe the areas on site 'that may be" subject to remediation 4.2 RADIOLOGICAL CONTROL PROGRAM CHANGES The Radiological Controls Program -will continue, to be fully integrated Into all aspects of operations at the SNEC Facility throughout the'license termination -process..The program will ensure that operations at the,SNEC Facility are performed in accordance with the ALARA philosophy of protecting plant personnel and the general public.
The SNEC Radiation'Protection Plan establi'shes b'a-sic l~olicies, philosophies,aian d objectives to maintain doses.to'workers and members [of the "public 'as ldw 'a's'reasonably achievable (ALARA). Specific details as to how the plan is' implemented are inbC6ip'orated into Radiological Controls procedures. These procedures are established and will be maintained in accordance with the SNEC Facility Technical Specifications. GPU Nuciear does 'not anticipate significant changes to Radiological Controls procedures during the license termination process. However, any changes.will be performed in accordan'c6,with the SNEC Facility Technical Specifications "and 10 CFR'50.59 process.',..
4.3 REMEDIATION METHODS' In conjunction with the near completion of Site' Dismniitlemernt Activities as'd escribed in Chapter 3, remediation will be performed in accordance with the gerleral deeontarnination and dismantlement considerations of Section 5.0 of the SNEC Updated Safety Analysis Report (Reference 4-1). Any area of the site that does not meet the release criteria will be remediated as appropriate..In keeping with,the principles of ALARA, additional residual -levels of radioactivity may be reduced cormmensurate wviththe mirnimization of total'risk. As an example, becau.e of the effectivenes's 6f some remediation m ethods, some areas may be remediated to residual dose levels well below'25 mrem/y'earuat little'6r'no'additional cost.
4-1 REVISION 1
SNEC FACILITY LICENSE TERMINATION PLAN REVISIONI 4.3.1 Building and Structure Surfaces Many decontamination methods exist to decontaminate surfaces of buildings and structures.
Several factors determine the choice of decontamination method for a given application, including the extent of the contaminated area, surface' material, depth of contamination, arid access considerations:'Typical technologies that may be used are' listed bel6w. All methods will be utilized with proper, regard to the potential for airborne contamination and industrial safety, and may be used in conjunction with containment enclosures or HEPA filtration units in order to minimize airborne radioactivity (as appropriate).
Abrasive Vacuum Blastina with Recyclable Blast Media - This method is highly effective for' surface contamination up toga depth of 6rmm. Unlike conventional abrasive blasting,' this method uses a recyclable blast media such as steel shot to minimize the'volume of waste that is generated.
Concrete Sectioninq - This method can be used to -remove large pieces of contaminated concrete where other methods would be too slow.
Cutting may be performed with an abrasive blade or a diamond wire saw.
Needle Gunning - This method is used for removing surface contamination in areas that are too small or inaccessible for other methods.
Manual Removal of Building/Structural Materials' - Contaminated building materials such as steel and loose concrete will be cut and manually removed and disposed of at an appropriate low level radioactive waste burial or treatment facility.
Concrete Scabblinq - Contaminated concrete structures are routinely scabbled to remove surface' activity deposits. Scabbling' involves m6chanical pounding of the surface to rem6oe surface deposits and some subsurface contamination.
4.3.2 Surface Soils, Gravel and Asphalt Soil remediation Will involve the removal of soil, gravel or asphalt as necessary to meet tiie site release criteria. Asphalt'and gravel will be treated' in the 'sarie'e way as soil. Soil, gravel and asphalt will be removed with excavation equipment, and due care will be taken to prevent the spread of contamination during excavation and handling, including controls, to minimize the creation of fugitive dust materials.
All radiologically contaminated'soil, gi'avel and asphalt determined to be above release criteria will be disposed of at a low-level radioactive material processing or disposal facility.
4.3.3 Sediment As discussed in Chapter'3, sediment in the SSGS was removed as part of site dismantleme'nt activitiesý.
No additional 'seddimfient contamination has beený found that exceeds the' release criteria, including sediment impacted by lice~rised discharge's monitored as part of the SNEC Radiological Environmental' Monit6ring Pr6gram (REMP). Any additional sediment found to be contaminated from plant operations will be removed if the levels of contamination exceed the release criteria. Contaminated sediment may be removed by dredging or by other manual excavation techniques.
4-2 I-
, SNEC FACILITY LICENSE TERMINATION PLAN
- 4.3.4
,Sub-floor Soil
_'Normaly,' when radionuclide,'6oncentrations in sub-floor soil or beneath structures are found that exceed the applicable release criteria,,the condrete floors or structures will be cut or broken up and the underlying soil excavaied and removed along with any contaminated structural material.
However, becads'e 6f thW difficulty in excavatingb6eneath some existing structure s, remediation of sub-floor soil may take place after the structure has been'demolished. Reimediation may also be performed with the structure intact. Several factors influence the approach that is used, including structural stability issues, groundwater level, the extent and -depth of the contamination,*-the accessibility of the work area to excavation equipment, and the need to protect the work area from the elements as well as the safety of the workers. If it is necessary to first demolish a structure before accessing the below grade soil or other contaminated materials, the part that is above grade and the accessible parts below grade (interior floors, walls, ceilings) will be decontaminated first (as deemed appropriate). A final status survey will then be performed on these areas to verify that release criteria have been met. NRC approval of the results will release the surveyed parts of the structure for conventional demolition and disposal.
After the debris is removed, radiological controls will be re-established and the radiologically contaminated sub-floor soil will be excavated as necessary.
4.4 AREAS TO BE REMEDIATED Site characterization to determine remaining areas that need remediation is ongoing and will continue throughout decommissioning. The site release criteria and the principles of ALARA will be applied on a case by case basis to identify the extent of the remediation necessary. The following areas have either already been remediated, identified for remediation, or may undergo further analysis as deemed appropriate:
All Outdoor Radioactive Material Storage Areas - These areas may change as site land use needs change. Initially, all Potentially Contaminated Areas (PCA) will be surveyed and be shown to meet site release criteria or be remediated and then surveyed prior to site release. In addition, any subsurface area previously used for underground liquid radioactive waste storage will be sampled and surveyed as appropriate.
Soil and/or Gravel Under or Near Existing Site Facilities - Some contaminated materials have already been removed from beneath site structures during the soil remediation project. Additional soil and/or gravel will be remediated as necessary if future site characterization finds contamination in excess of the site release criteria.
In addition, soil deposits around or under existing structures, such as the old steam tunnel and concrete slabs exterior to the CV, will be sampled, analyzed, and remediated as necessary during dismantlement operations.
Structural Decontamination - Buildings and structures which have been identified for remediation include the Containment Vessel, Steam Tunnel, Saxton Steam Generating Station and surrounding concrete slabs. If additional structures are found to be contaminated in excess of the site release criteria, they will be decontaminated when they are no longer required to support site decommissioning operations.
4-3 REVISION 1
Storm Drain Out-fall - Section 2.0 in the SNEC Facility Site Characterization Report describes various contamination events that have occurred over the course of the plant's life that may' have resulted in discharges of low levels of radioactivity.
However, no event is listed that describes above permissible release of liquid to the storm drain and weir system. In addition, the weir and 10" drain line emptying to the river have been removed.
4.5 REFERENCES
4-1 Updated Safety Analysis Report for Decommissioning the SNEC Facility Revision 2, February 1998.
4-4 SNEC FACILITY LICENSE TERMINATION PLAN RI*VIRIt3 N t
SNEC FACILITY LICENSE TERMINATION PLAN 5.0 SNEC FACILITY FINAL STATUS SURVEY PLAN
5.1 INTRODUCTION
The SNEC Facility Final Status:Survey Plan (FSSP) has been prepared using the guidance provided -inapplicable regulatory guidance documents -described in -Section 5.1.1 below.
Ultimately, this plan will. be used to develop lowertier procedures and/or work instructions to accomplish the Final Status Survey for the SNEC Facility.
5.1.1 Purpose The FSSP describes the final survey process that will be used to demonstrate that the SNEC Facility and all additional near site impacted areas-meet -radiological criteria for. license termination'. 10 CFR 50.82(a)(9)(ii)(D) (Reference 5-l), and Regulatory. Guide 1.179 (Reference 5-2) have been used as guides in the preparation of this plan., This plan incorporates.the site release criteria provided in 10 CFR 20.1402 (Reference 5-3) and addresses concerns of NUREG-1727, the NMSS Decommissioning Standard Review-Plan,- (Reference 5-4), and NUREG-1505 (Reference 5-6). Other documents, such as Draft NUREG-1549 (Reference 5-9),
were also reviewed in the process of preparing this plan.
5.1.2 Scope The final site survey will encompass structures, land areas, and any remaining facility systems which, because of licensed activities, were originally contaminated or had the potential to be contaminated. Areas that exhibited the highest contamination levels were located, within the SNEC Containment Vessel (CV), as illustrated in Chapter 2 of this License Termination Plan (LTP). As ofthe date of the SNEC Facility LTP submittal, the majority, of-all-contaminated systems, components, and soils will have been removed from the site.-: Continued remediation in selected areas will ensure these areas satisfy unrestricted release criteria.before the Final Status Survey_(FSS) process begins.
5.1.3,
Summary The SNEC Facility FSSP describes the final survey process and the methodology used to develop guideline values against which residual radioactivity levels, remaining, at the SNEC Facility at the time of the FSS will be compared. The final survey process is described as a series of steps - survey preparation, survey design, data collection, -data assessment, and final survey report preparation.
However, in practice, this is an iterative process in that once the results from one step are known they may. promptrepeating one or ;more previous steps. In addition, the process is designed to be flexible in that modifications to the survey process Tmay be made as more information is collected.
FSS activities begin when dismantlement and decontamination activities are believed to be
.complete. Each survey area,is divided intosurvey-units thatare L classified, according to their potential for retaining residual radioactivity, or in-accordance with known contamination levels.
Survey data, collected from each survey unit are, collected raccording to-data collection requirements and frequencies established for each classification. When residual radioactivity is measured above pre-set levels, an investigation is performed.
Based on the results -of the investigation, the survey unit may be remediated, reclassified, resurveyed or determined to meet the release criteria I :
I 5-1 REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 There are three principal types of survey results collecte~d during the FSS effort. They are 1) scan measurement data, 2) fixed-point measurement data, and 3) sampling of volumetric materials for laboratory analysis. In-situ gamma-ray spectrometry niay also be included in the release survey process as well as the results of any special measurements or analysis.
Statistical testing criteria for special measurements will be -applicable to the survey methods used., All collected data are first verified to be of adequate quantity and quality, capable of supporting underlying assumptions necessary fort statistical testing. Wh'ere necessary, previous survey steps are re-evaluated. Each survey unit Will then be tested'and compared to the release criteria. To meet the release criteria, the survey data must pass the statistical tests applied.
When the data set fails statistical testing criteria, the survey unit is not suitable for unrestricted release without further actions.
Upon completion of FSS activities, a final survey report will be prepared which summarizes the data. The report will docume'nt the conclusion that the SNEC'Facility and near site areas meet the 10 CFR 20.1402 release criteria and may be released for unrestricted use.
5.2 SURVEY OVERVIEW This section describes the scope and! methodology of the final survey process. It includes quality assurance measures and access control procedures.
It also describes how implementation of this plan will demonstrate that the remaining structures and site areas meet the 10 CFR 20.1402 criteria for unrestricted release. Also described herein, are the methods used to develop guideline values against which residual radioactivity levels will be compared.
5.2.1 Identity of Radiological Contaminants The radionuclide inventory at the SNEC-Facility was 6stimated during the initial Site characterization process, which was conducted between 1995 and 1996. Those data are corfipiled in the SNEC Facility Site Characterization Report (Reference 5-7). Station Work Instructions, site procedures, and Survey Requests have since been-used to collect additional site characterization data. This more recently collected information is summarized in Chapter 2 of this plan. All of the data were reviewed and a final radionuclide listing was developed. Refer to Chapter 6, Section 6.2.2.3.
5.2.2 Site Release Criteria 5.2.2.1 Radiological Criteria for Unrestricted Use These site release criteria correspond to the radiological criteria for unrestricted use given in 10 CFR 20.1402, which are:
Dose Standard Residual radioactivity, distinguishable from background radiation and resulting in a Total Effective, Dose Equivalent (TEDE) to an average member of the critical giroup will not exceed 25 mrem/y, including that from groundwater sources of drinking water.
ALARA Standard Residual radioactivity will be reduced to levels that are As Low As Reasonably t*
Achievable (ALARA), as addressed in Section 6.4.
5-2
5.2.2.2 Conditions Satisfying the Site Release Criteria Levels of residual radioactivity that correspond to the allowable radiation dose and ALARA levels are calculated (derived) by analysis of various scenarios and pathways (e.g., direct radiation, inhalation, ingestion, etc.)..
These derived levels are referred 'to as "derived concentration guideline levels" (DCGLs), and form the basis for-the following conditions which if met, satisfy the site release criteria (DCGL is defined in Section 5.8):
All measurements of residual radioactivity above background from all survey units are equal to or below the DCGLw*. No further testing is required, or Individual measurements from small areas within a selected survey unit that exceed the DCGLw, do not exceed the DCGLEMC*. - In addition, the sumof the fractions'for both DCGLw and DCGLEMC in applicable sections of a survey unit are less than unity for all site survey units, and
. All survey data pass applicable statistical testing criteria.
In all cases, remediation is performed (as applicable) to reduce-the levels of residual radioactivity concentrations to ALARA values.
NOTE: *The DCGLw is the permitted average concentration in the survey unit. The DCGLEMC is the elevated measurement comparison DCGL.
5.2.3 Development of Derived Concentration Guideline Levels 5.2.3.1 Dose Modeling Dose-models based on NUREG/CR-5512, Volume 1 (Reference 5-8) and RESRAD'(Reference 5-10) were used to calculate site DCGLs. GPU Nuclear, Inc. also employed URS Corporation to assist in developing-suitable subsurface dose'models for the SNEC site. These dose models translate residual radioactivity levels into potential radiation doses to,the,public and are defined by three factors::1) the scenario, 2) exposure pathways, and 3) the critical group. The 'scenarios described in NUREG/CR-5512 address the major,'exposure pathways,of direct exposure,to penetrating -radiation and inhalation and ingestion of radioactive materials., -These scenarios also identify the critical group. The critical group is the group of individuals reasonably expected to receive the greatest exposure to residual radioactivity within the assumptions of the particular scenario.
Two representative scenarios were selected for the SNEC Facility. They are 1) building occupancy, and 2) residential farming.,These scenarios; described below, represent reasonable and plausible human activities and future uses of the SNEC Facility site.
5-3
,. ý_.- "...
REVISION 1 SNEC FACILITY LICENSE TERMINATION PLAN
5.2.3.1.1 Building Occupancy Scenario Because surface decontamination operations may not completely remove surface radioactivity, a scenario describing surface contamination' is -considered.
This scenario accounts for exposure to both fixed and removable thin-layer radioactivity for a structure. This scenario also assumes that individuals occupy the building in a passive mannerwithout deliberately disturbing the residual radioactivity on building surfaces. Occupancy of the building is assumed to begin immediately after license termination. The exposure duration is assumed for a full work year (2000, hours) continuing for seventy (70) years.
The critical group consists of the building occupants, who are the people who work in the building following license termination.
This scenario is described in Chapter 6, Section 6.2.1.
5.2.3.1.2 Residential Farming Scenario Soil at the site is contaminated from licensed operations, accidental spills, and long-term accumulation of material in the soil from effluent releases.
It is also contaminated from intentional disposal or burial of concrete or other structural debris such as pavement, masonry or structural steel. The residual radioactivity is assumed to be distributed in a surface soil layer covering the site on property that is used for residential and light farming activities, or is buried beneath this surface.
The scenario assumes continuous exposure via multiple exposure pathways to the critical group. The critical group is the resident farming family who lives on the site following site remediation, grows some portion of their diet on the site and drinks water from a source at the site. This scenario is described in Chapter 6, Section 6.2.2.
5.2.3.2 Derived Concentration Guideline Level (DCGL)
The surface contamination and radionuclide concentration levels of structures, land areas, and facility systems remaining at the time of the FSS are compared to DCGLs calculated using the dose models. A DCGL is defined as "that concentration of ýesidLial, radioactivity distinguishable from background radiation which if distributed uniformly throughout a survey unit, would result in a Total Effective Dose Equivalent (TEDE) of 25 mrem/y to an, average member of the critical group." The average member of the critical group is the individual who is assumed to represent the most likely exposure situation based on the assumptions and parameter Values used in the dose model calculation. *SNEC Facility volumetric DCGLs are calculated using a deterministic approach that incorporates any input parameter that were probabilistically determined to be sensitive during the analysis. This'approach ensures that the annual TEDE dose to the average member of the critical group (expected within the first 10,000 years after license termination) will be at or below 25 mrem. DCGLs are presented in terms of surface or volumetric radioactivity concentrations and are expressed in units of dpm/1 00 cm 2 or pCi/g.
SNEC Facility personnel have proposed a comprehensive set of volumetric DCGLw limits that were developed from the lowest calculated results from both-the surface and subsurface regions within or near the SNEC site. SNEC management believes that this all-inclusive set of DCGLw values considers all viable pathways of exposure. These proposed DCGLw values consider both the surface and subsurface conditions of the site, modeled for locations that exhibit differing site dependent geological and hydrological characteristics. Soils and other materials were modeled down to bedrock and take into account the former Spray Pond area, Saxton Steam Generating Station footprint and the SNEC CV area. Surface DCGLw values are modeled down to one meter but the lowest values determined from the surface or subsurface models were then selected to represent the SNEC site 25 mrem/y DCGLs. This method ensures that relocated site soils will never deliver more than the previously described dose limit.
5-4 W_
I* I:::*/I
- II'*M 4
Modeling was also intended to provide DCGL-values to meet the administrative goal of 4 mrem/y drinking water. These values,in many.cases, are more restrictive than the 25 mrem/y DCGLw values developed using -all' pathways* modeling previously described. It is SNEC management's intent-to consistently-acknowledged this lower -limit, (4 mrem/y DW) when designing Final, Status Surveys for site property. To accomplish this, both sets of values are "listed irn the SNEC LTP. The 25 mrem/y TEDE limits develop ed using all pathways modeling will be used to meet license termination' requirements ahd therefore are6 site specific NRC reviewed and approved'DCGLs. The second set of values relating t6 the 4mr'em/y drinking water goal, is provided a's a SNEC Facility commitment to the general public.
A deselection process was used to reduce the applicable site radionuclide listing -to eleven radionuclides which can be seen in Table 5-1. The:,dose from 'all deselected radionuclides will be accounted for-in the final site survey design process by, adjusting down-thie-applicable DCGL as per section 6.2.2.3.
5.2.3.2.1 NRC Screening DCGLs' The NRC screening DCGLs are intended to be used as the principal means of releasing structures at the SNEC Facility. However, since th&esite is radiologically complex,(i.e.; multiple radionuclides are found in various distributions throughout the site) screening:DCGLs are seldom, if ever applied directly to determine compliance With the site release do'se-criteria, Rather, they are used,to develop surrogate ratio, gross activity, or elevated measurement comparison DCGLs (DCGLEMC), or are applied using the unity rule.
The gross activity or surrogate ratio DCGLsare termed "effective" DCGLs in the sections that follow. Where the broad conservatism inherent in the screening DCGLs prove unrealistically restrictive, 'a set of site specific DCGLs may be developed and used in their place (with regulatory approval).
NRC screening DCGLs for surface contamination, are presented in the second column of Table 5-1, and were calculated by entering standard default conditions i6to the NRC's dos"e iriolelirdg]
computer software "DandD". Alpha emitting radionuclides were,calculated, using, the same default parameters pr6eided in the DandD conimluter software (uýface coni6entrations only).
5-5 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
SNI=C FACILITY LICENSIE TE*RMINATION P1 AN Table 5-1 SNEC Facility DCGL Values (a) 25 mremly Limit 4 mremly Goal 25 nmreml/y Limit (All Pathways)
(Drinking Water)
Radionuclide Surface Area Open Land Areas Open Land Areas (
(dpmllOOcm 2)
(Surface & Subsurface)
(Surface & Subsurface)
(pCi/g)
(pCifg)
Am-241 2.7E+01 9 9 2.3 C-14 3.7E+06 2
5.4 Co-60 7.1E+03 3.5 67 Cs-1 37 2.8E+04 6.6 397 Eu-152 1.3E+04 10.1 1440 H-3 1.2E+08 132 31.1 Ni-63 1.8E+06 747 1.9E+04 Pu-238 3.OE+01 1.8 0.41 PU-239 2.8E+01 1.6 037 Pu-241 8.8E+02 86 19.8 Sr-90 8.7E+03 1.2 061 NOTES:
(a) While dnnking water, DCGLs will be used by SNEC to meet the dnnking water 4 mrem/y goal, only the DCGL values that constitute the 25 mrenmy regulatory limit will be controlled under this LTP and the NRC's approving license amendment (b) Listed values are from the subsurface model. These values are the most conservative values between the two models (I e, surface & subsurface).
5.2.3.2.2 Site-Specific DCGLs Where DandD cannot be easily modified to incorporate the'use'of site-specific data,, other computer dose modeling software, such as the US Department of Energy's RESRAD (Reference 5-10) and RESRAD-BUILD (Reference 5-11) software, may be used to generate site-specific DCGL values. The best approximation of real site conditions was obtained by replacing some default-input parameters in the RESRAD computer code, with site-specific parameters. Site-specific parameters, representing actual site conditions, were obtained from contracted geological services. Site-specific Kd data were empirically derived by Argonne National Laboratories (ANL) from samples of SNEC site soils and construction debris. These empirical values reduce overlapping conservatism and produce a more realistic estimate of site specific conditions.
Site-specific values used in developing site-specific DCGLs are documented in Chapter 6. Site-specific DCGLs are listed in the third column of Table 5-1.
5.2.3.2.3 Surrogate Ratio DCGLs Surrogate ratio DCGL values may be established for areas where ratios between radionuclide concentrations are reasonably consistent.
Establishing ratios between radionuclide concentrations allows other radionuclide concentrations to be predicted and therefore assumed present in the mix at a fixed fraction of another radionuclide. Likewise, a surrogate ratio DCGL allows the DCGLs specific to "hard-to detect" (HTD) radionuclides in a mix to be expressed in terms of a single radionuclide which is more easily measured. The measured radionuclide is called the surrogate radionuclide.
5-6 I-1:7 I=*11* It*kl 4
To obtain the'best approximation of the actual ratio between'radionuclides present'in any area, a sufficient number of measurements or samples, spatially separated throughout the area must be collected. The number of m6asurements or samples'needed to determine'the ratio is based on the chemical, physical,'and radiological characteristics of the radionuclides'-expected to be present, in the:area. -- The surrogate,ratio may-also be determined'l'6y u~ing the most
,conservative ratio present in the 'data 'set'(i.e.,'the ratio'that will producetfhe'highest calculated dose per'unit concentration). One may also consider the variability in the 'sampling results by determining the upper 2 sigma bound for each radionuclide present in applicable sam'ple data.
These 2 sigma values would then be used with the mean value of the surr6gate"radionuclide--to form a conservative, mix for a survey area.- In cohclusion, s'electing a `conservativ&6'surrogate ratio ensures that potential exposures from individual radionuclides 'are not underestimated.
The surrogate method will be used only when similar physical and geological characteristics are present in a survey unit.
The general equation used for a simple surrogate situation involving two isotopes such as Cs-137 and Sr-90 (as an example), is shown below:
DCGLcs, Modf,ed = DCGLcs x DCGLsr'I(((Csr I Cc) x'DCGLcs) +DCGLsr)
Where:
Csr/Cc,=
surrogate.ratio of Sr-90 to Cs-137 (NUREG-1575, Equation 4-1 (Reference 5-5))
Post-remediation sampling may be perfor'med.to re-establish surrogate ratios when there is reason to believe that the remediation process may, have significantly changed the ratios of the original radionuclide concentratlon mix. Hovwever, thlesurrogate ratio will usually be'established using -data 'collected prior to remediation. Additi6nal -'sample data may' be collect6d or an evaluation performed when there is reason tolbelieve a surrogate ýatio is n6n-representative of the survey area. In all cases, SNEC will use the best representative data for the-survey area: A radionuclide mix justification will be documented in each survey package.
5.2.3.2.4 Gross Activity DCGLs Where multiple radionuclides are present, a gross activity DCGL may be devel6ped. The gross activity DCGL enables-field measurements of gross activity rather, than'the determination -of individual radi6nuclide activity, for comparison-to the radionuclide-specific DCGL.,The 'g~oss activity DCGL, or I'DCGLGA, for surfaces _or volumrs' with multiple radionuclides is calculated using the following equation (NUREG-1575,"Ref6rence'5-5, Equati6n 4-4):
Gross Activity DCGL =1 I((f, I DCGL,) + (f2 I DCGL2) +...(f, I DCGL,)),
Where:
f,
=
relative fraction of the total activity contributed by radionuclide 1 f2
=
relative fraction of the total activity contributed by radionuclide 2 f,
=
relative fraction of the total activity contributed by the nth radionuclide in the mix 5-7
'_REVISIO)N I
'SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 In situations where an area has unknown or a highly variable concentration of radionuclides throughout, it is acceptable to select the most conservative mix of radionuclides present when developing a gross, DCGL.
Remediation efforts at the SNEC Facility will seek to remove contamination from concrete surfaces by removing significant near surface volumes of concrete.
On other structural surfaces (e.g., support steel), abrasive decontamination techniques may be used to remove surface deposited contamination. Additionally, by aggressively chasing cracks and other structural anomalies, little (if any) surface deposited radionuclides should remain.
Those that do remain will be,,at_ a significantly reduced fraction of their previous level.
Characterization data and/or post-remediation sampling data will be used to adjust gross DCGL values before performing-, the FSS. In either case, a radionuclide mix justification will be documented in each survey package.
Gross activity measurements will be performed in both reference areas (background areas) and areas to be surveyed.
Open land areas will be scanned by semi-automated, contractor supplied measurement equipment and/or by hand-held Nal detector instrumentation. Samples will also be obtained.
When necessary, spectra of the components of a radiation field may be used to differentiate between site related radioactive materials and natural occurring radioactivity.
5.2.3.2.5 Elevated Measurement Comparison (EMC) DCGLs Elevated Measurement Comparison (EMC) DCGLs are DCGL values that have been modified by reducing the dose model area size in the dose modelingcomputer code. The effect is to allow higher levels of contamination to exist in smaller well-defined on-site areas. Assuming the residual radioactivitV is concentrated in a much smaller area rather than uniformly over the entire survey unit, is the Ibasis for developing the DCGLEMc. The, methodology used to calculate the DCGLEMC is given in Appendix 5-1.
Area fact6rs for individual radionuclides have been calculated and are presented in Tables 5-15 and 5-15A.
5.2.3.2.6 Unity Rule Typically, each radionuclide specific DCGL corresponds to the release criteria (e.g., regulatory limit in terms of dose or risk). However, in the presence of Multiple radionuclides, the total of the DCGLs for all radionuclides would exceed the release criteria. In this case, the individual DCGLs need to be adjusted to' account for the presence of multiple'radionuclides contributing to the total dose. One method for adjusting the DCGLs is to modify the assumptions made during exposure pathway modeling to account for multiple radionuclides. The surrogate measurement method discussed previously', describes one method for adjusting the DCGL to account for multiple radionuclides. Additionally, another method includes the use of the unity rule.
The unity rule, represented in the expression below, is satisfied when radionuclide mixtures yield a combined fractional concentration limit that is less than or equal to one (1) (NUREG 1575, Equation 4-3):
(CIDCGL,) + (C2IDCGL 2) +...(C,,IDCGL,)
- 1 Where:
C = concentration of radionuclide (1, 2,...,n)
DCGL = guideline value for each individual radionuclide (1, 2,....,n) 5-8
A higher sensitivity will be needed in these measurement methods, as the values of C become smaller.
In addition, this may influence statistical testing considerations by increasing the number of data points necessary for application of a specific statistical test.
5.2.3.2.7 Handling of Multiple Source Terms When determining DCGLs in areas where there are multiple source terms, Equation 6-1 will be used.
5.2.4 Facility and Site Classification Not all areas of the site have the same potential for residual radioactivity and, accordingly, do not need the same level of survey effort to demonstrate compliance with the site release criteria.
Using the criteria.given belowv, different sections of the site are grouped into impacted and non impacted areas based on thepotential for residual radioactivity to~be present. Classification of site areas is based on professional judgment, operational history (Historical Site Assessment (HSA) information, Reference 5-19), site characterization data, operational surveys performed in support of decommissioning, and routine surveillance. See the site facility diagrams Chapter 2, and the SNEC site map (Figure 5-1), which is located at the end of this chapter.
5.2.4.1 Non-Impacted Areas Non-impacted areas have no reasonable potential for the presence of residual radioactivity from licensed activities. These areas do not need any level of survey coverage since there was no radiological impact from site operations. No surveys are performed in these areas other than those used to determine a reference area (background).
5.2.4.2.
Impacted Area Impacted areas are areas that have a reasonable potential for the presence of residual radioactivity from licensed activities.
Impacted areas are subdivided into-three classes described below.
55.2.4.2.1 Class 1 Areas Class 1 areas -are areas that have or have'had (prior to remediation), a potential for, radioactive contamnination (basedon 'site o6perating history), orkno~vn contamination (based on previous radiological surveys). r Examples of Class 1 areas are:
"* Areas previously subjected to remedial actions Locations where leaks or spills are known to have occurred Former burial or disposal sites
"* Waste storage sites Areas with contaminants in discrete solid pieces of material at high specific activity Areas containing contamination more than the DCGLw before remediation 5-9
°SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 5.2.4.2.2 Class 2 Areas Class 2 areas are those that have or have had, a potential for radioactive contamination but' are not expected to contain radioactive material greater than the DCGLw. Examples of Class 2 areas are:
Locations where radioactive materials were present in an unsealed form, Potentially contaminated transport routes, Areas downwind of stack release points, Upper walls and ceilings of some buildings or rooms subject to airborne radioactivity, Areas where low concentrations of radioactive materials were handled, and Areas on the perimeter of radioactive material control areas.
5.2.4.2.3 Class 3 Areas Class 3 areas are any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fraction of the DCGLw This would again be based on site operating history and previous radiological survey information. Examples of Class 3 areas are:
Buffer zones around Class 1 or Class 2 areas, Areas with a very low potential for residual contamination, but where insufficient information exists to justify a non-impacted classification.
5.2.4.3 Initial Classification The initial classifications of the SNEC Facility are given in Table 5-2. They are based on site characterization data, the results of the Historical Site Assessment, and recommendations and concerns of SNEC Facility personnel knowledgeable of site conditions. Site characterization data and radiological history information on Table 5-2 survey areas are summarized in Chapter
- 2. When there was an uncerta'inty regarding the preliminary classification of a SNEC Facility impacted area, the area was initially assumed a Class 1 area until determined otherwise.
5-10
-I-
SN(_
,CILITY LICENSE TERMINATION PLAN C
LICENSE.F 1-.
v IPIAI'N Table 5-2 Initial Classifications of Site Areas Survey Unit Designations of the SNEC Facility and Surrounding Impacted Areas Survey Unit Classificatin j
Survey Unit Area (m2) ()
Number of IType of DCGL N u m b e r (
')
~ ~ ~~De s c r ip t io n N m e fT p
f D G Number D
1 1
2 1
3 Floor I Walls Ceiling Other Survey Units m Applied Ic)
MISCELLANEOUS AREAS & ITEMS MAI Airborne Monitoring Stations X
<10 1
1 MA2 SSGS Discharge Tunnel Outfall (Land Area)
X 600 1
2 MA3 Weir Outfall X
25 1
2 MA4 Weir Outfall Buffer X
200 1
2 MA5 Northeast Dump Site X
7000 1
2 MA6 Northwest Open Land Area X
4100 1
2 MA7 Northwest Open Land Area X
100 1
2 MA8 Miscellaneous Concrete Slabs (Around Site)
X
<100 1 each 1
CONTAINMENT VESSEL (CV) - INTERIOR & EXTERIOR STEEL SHELL CVI-X Interior Vertical Wall of CV Shell <-804 5' El X
392 4
1 (C)
CV2-X,-'
Internal Support Ring Areas X
,65 22 (d) 1 (
CV3-X Interior Curved Bottom of CV Shell X
255 3
1(e)
'CV4-X Exterior Wall - 802 6'El up to Cut-off X
16 1 (e)
CV5,ý Exterior Wall 1 Meter Below Class 1 Area (Down to 797.6' El)
X 10 11
()
CV6 External Rock Anchor Support Ring Assembly Area X
- 66
-1ji) e)
., MATERIAL HANDLING BAY (MHB) - SNEC AREA MH1 7
Floors & Walls Up to 2 Meters (Interior)
X 20 1,
1 MH2
,Upper Walls & Ceiling (Interior)
X 63 22 11 MH3,
Roof
,X 24 11 MH4,
Exterior Walls.
56I 1,
NOTES:
(a) *X" designates a sequential number starting with 1, and defines a survey unit withih a-survey area (b) This data was estimated with best available information No survey unit, regardless of its classification will exceed 10,000 square meters.
(c), NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs = 2 -.....
(d) Survey units were established a's'the ring reas became available to field personnel doing the survey work........
(e) Activation of CV steel liner to be addressed when region is accessible (f) This facility may be removed prior to performing Final Status Survey.
(g) Based on projected cut-off at 804 5' El 5-11
Table 5-2 (continued)
Initial Classifications of Site Areas Survey Unit Designations of the SNEC Facility and Surrounding Impacted Areas Survey UniDecptoI Classification Survey Unit Area (in2) (b Number of IType of DCGL Number ()i esrpto 1 112 13 Floor IWalls ICeiling] Other Survey Units Applied(c id) PERSONNEL ACCESS FACILITY (PAF) - SNEC AREA
"" PF1 Floors & Walls Up to 2 Meters (Interior)
X 36 49 1
1 PF2 Upper Walls & Ceiling (Interior)
X I
116 36 1
1 PF3 Roof X
40 1
1 PF4 Exterior Walls X
133 1
1 "Id) DECOMMISSIONING SUPPORT BUILDING (DSB) - SNEC AREA DB1-X-Floors & Walls Up to 2 Meters (Interior)
X 212 121 5
1 DB2 Upper Walls & Ceiling (Interior)
X 290 212 1
1 D03 Roof X
225 1
1 DB4 Exterior Walls X
325 1
1 DB5 DSB Carport Slab X
62 1
1 DB6 DSB Carport Roof/Ceiling X
124 1
1 NOTES:
(a)
"X" designates a sequential number starting with 1, and defines a survey unit within a survey area (b)
This data was estimated with best available information. No survey unit, regardless of its classification will exceed 10,000 square meters (c)
NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs = 2 (d) This facility may be removed prior to performing Final Status Survey Q
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
SC. FACILITY LICENSE TERMINATION PLAN Table 5-2 (continued)
Initial Classifications of Site Areas 5-13 Survev Unit Desianations of the SNEC Facility and Surroundina Impacted Areas Survey Unit Descrptio classification Survey Unit Area (m)
Number of Type of DCGL Survy Unt, Description Number 1
1 2
1 3
Floor I Walls I ceiing Other Survey Units (
Applied (c)
SAXTON STEAM GENERATING STATION (SSGS), INTAKE & DISCHARGE TUNNELS SS1 Floor of Discharge Tunnel (first-150')
X 120 1
1 SS2 Floor of Discharge Tunnel (next -235')
X 175 1
1 SS3 Floor of Discharge Tunnel (last -315')
X 234 -
1 -
1 SS4 Ceiling of Discharge Tunnel (first -150')
X 120 1
1 SS5 Ceiling of Discharge Tunnel (last -550')
X 400 1
1 SS6-X Walls of Discharge Tunnel (first.150')
X 290 3
1 SS7 -
Walls of Discharge Tunnel (last -550')
X_
600 1
-1 SS8-X In DT-Seal Chambers (1, 2, & 3) -
X 230-3 1
SS9 Spray Pump Pit Floor -
X 120 1-1 SS10....
Spray Pump Pit Walls Below 795' El.
X 20 1-1 SS11-Spray Pump.Pit Walls Above 795' El.--
X 100 1
1.
SS12.
SSGSBoilerPad(811EI)
-X
-1800.
1.
SS13-.
SSGS Firing Aisle (806' El.)
X
-560-80 1
1 SS14-X SSGS Basement Area Floor (790' El)
X-360 4
1 SS15 SSGS Basement Walls _- East End
-X 100 1
SS16 SSGS Basement Walls Up to 2 Meters X
240 1
1 SS17.
SSGS Basement Walls > 2 Meters'.
X 350 1
1 SS18__
TopofSealChambers.
70 1
1 SS19-X--....
Section of SSGS Intake Tunnel Floor X
493 2
1 SS20-X -.
Section of IntakeTunnelWalls.-.
, X1 2150
-1 3
1 SS21
-.Section of Intake Tunnel Ceiling.
X-493 2
1 NOTES:
(a) -X" dgsignatei arisiquential ndrber starting with 1, and defines"a survey unit within a survey area.
(b) This data was estimated with best available information: No iurvey unit, regardlefss of itfsclassrfication will exceed 10, 00d square metjers (c)
NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs=2 2
(
FCILTY ICESE ERMNATON LANREVIS C%. I
Table 5-2 (continued)
Initial Classifications of Site Areas Survey Unit Designations of the SNEC Facility and Surrounding Impacted Areas Survey Unit I
Classification Survey Unit Area (m) (b)
Number of Type of DCGL Number (a)
Description 1
2 3
Floor Wails j Celng I Other Survey Units M Applied (c)
SAXTON STEAM GENERATING STATION (SSGS) SPRAY POND AREA SPi Open Land Area X
6600 1
2 SNEC FACILITY SITE OPEN LAND AREA "OL1-X-
'SNEC Facility Site & Near Site Area-X" 11000 11 2
GPU ENERGY (PENELEC) SITE OPEN LAND AREA:
OL2-X Westinghouse and Adjacent Areas X
5700 6
2 OL3 Warehouse Burn Area -X_
200 1
2 OL4-X BufferZones-X 5600 4
2 REMAINING IMPACTED OPEN LAND AREA OL5-X Site Road Access Areas X
20500 9
2 OL6-X Stack Release Area (NNE)
X 14600 3
2' OL7-X Stack Release Area (SSM)
X 12700 2
2 OL8-X Buffer Zones X
47900 5
2 (d' WAREHOUSE (LARGE GARAGE - SOUTH) - PENELEC AREA WA1-X Floors & Walls Up to 2 Meters (Interior)
X 450 -
290 2
1 WA2 Upper Walls & Ceiling (Interior)
X -
292
- 450 1
1 WA3 Exterior Walls X
374 1
1 WA4 Roof..
X 418 1
-1 NOTES:
(a)
"X" designates a sequential number starting with 1, and defines a survey unit within a survey area.
(b)
This data was estimated with best available information. No survey unit, regardless of its classification will exceed 10,000 square meters.
(c)
NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs = 2 (d)
This facility may be removed prior to performing Final Status Survey (e)
Includes substation yard drainage area.
SNEC FACILITY LICENSE TERMINATION PLAN RFVIRIt3M 4
(
REVIsrb 1 I SN."-
iACILITY LICENSE TERMINATION PLAN Table 5-2 (continued)
Initial Classifications of Site Areas Survey Unit Designations of the SNEC Facility and Surrounding Impacted Areas.,
Suvynt I,-
1 W)
Surveyr escrption Classification
.Survey Unit Area
) °b)
Number of Type of DCGL Numbr 1
3 1 Floor Walls Ceiling Other Survey Units T Applied
-di GARAGE (SMALL GARAGE - SOUTHWEST) - PENELEC AREA GAI-X Floors & Walls Up to 2 Meters (Interior)
ý,X 109 122 4
.1 GA2-X UpperWalls& Ceiling (Interior)
X 297 109 2
1 GA3 Exterior Walls X
180 1
1 GA4
- Roof, X
116 1
1 LINE SHACK - PENELEC AREA LS1-X Floors & Walls Up to 2 Meters (Interior)
X 290 177 5
1 LS2-X Upper Walls & Ceiling (Interior)
-X 191 412 7
1 LS3 Exterior Walls X
343 1
1 LS4
ý Roof X
324 1
-1 LS5 Roof Drainage System X
<10, 1
1,2
- PENELEC SWITCHYARD BUILDING & YARD STRUCTURES PS1
,Interior "X
55 89 55 1
1 PS2 Exterior Walls and Roof 151 68 1
PS3 Switchyard Units - Base Pads X
<500 1
1 NOTES:
(a)
"X" designates a'sequential number starting with 1, and defines a survey unit within a survey area.
(b)
This data was estimated with Vest available Information No survey unit, regardless of its classification will exceed 10,000 square meters.
(c)
NRC Default'Surface DCGLs1= 1, Site Specific Volumetric DCGLs = 2 (d)
This facility riiay be removed prior to performing Final Status Survey.'
5-15
5.2.4.4 Changes in Classification Changes in classification are based on survey data and other relevant information that indicates a different area classification is more appropriate. Changes in area classifications which decrease an area classification will be in accordance with License Condition 2.E.(i).
5.2.5 Final Survey Process In general, FSS activities do not commence in the area to be surveyed until decontamination activities are believed to be complete and radioactive waste materials are removed. The FSS process begins with survey area preparation activities such as'gridding and review of final remediation support survey information, as well as survey area walk-downs.
Survey design calculations and the issuance of Survey Requests to field survey teams follow this phase. Field survey teams then collect the data and assemble the survey results in an organized and understandable format in accordance with site procedures.
Data assessment and documentation concludes this process.
5.2.5.1 Survey Design Overview Survey design, as described in Section 5.4, identifies relevant components of the FSS process and establishes the assumptions, methods, and performance criteria to be used. Areas ready for FSS are classified as Class 1, Class 2 or Class 3 and are divided into survey units.
Systematic scan and static measurements are prescribed according to a pattern and frequency established for each classification.
Investigation levels are established which, if exceeded, initiate an investigation of the survey data. A measurement' froni the survey unit that exceeds an investigation level may indicate a localized area of elevated residual radioactivity.
Such locations are marked and investigated to determine the area and the level of the residual radioactivity present. Depending on the results of the investigation, the survey unit may require remediation, and/or re-survey or re-classification.
When necessary, a two-stage, sampling process may be, used lAW Reference 5-20. This sampling approach allows a second set of samples to be taken to meet the requirements of the statistical design of the survey. When used, this process will be incorporated as an option in the original survey design for the area.
Quality Control (QC) measurements are prescribed to identify and control measurement error and uncertainty attributable to measurement methods or analytical procedures used in the data collection process.
QC measurements provide qualitative and quantitative information to demonstrate that measurement results are sufficiently free of error and accurately represent the radiological condition of the SNEC Facility.
5.2.5.2 Survey Data Collection As deemed appropriate, a final post-remediation survey is performed using similar instrumentation, quality control-and survey techniques to be used in the FSS process. The review of the final post-remediatidn survey data is then carried out to verify that residual radioactivity levels are acceptable 'and that no additional remediation will be needed in the survey unit.
If an area of elevated residual radioactivity is identified, and remediation is determined to be ALARA, the area is remediated and re-surveyed to ensure meeting FSS requirements. The data collected during the final post-remediation survey (when performed),
5-16 SNEC FACILITY LICENSE TERMINATION PLAN RMVSION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I provide a sound basis for interpreting radiological conditions that may be encountered.during the FSS process.
Following the collection'of acceptable post-remediation survey results (as app licable), the FSS is performed. It ensures that any remaining residual radioactivity meets thef 25-mrem/y TEDE site release dose criteria. Measurement results stored as FSS data constitute the FSS of record and are included in the'data set used to -determine c0mpliance with the site release criteria.
5.2.5.3 Survey Data Assessment Survey data'assess~ment, described'in Section 5.6, is perfornmed to verify that the final survey data are of adequate -quantity ýanid ýUality Graphical repre-sentations and statistical comparisons of the data are made, which provide both qualitative and quantitati(,e information about thebdata?. Asses'smrints are performed to:verify that thec data supp 6
rt the underlying assumoptions' necessary-for the statistical tests. If thee quality', cquantity, or one 'or'more of the assuLiptioirs a&e called into quiestiori, previous survey steps are reevaluatecd. Statistical tests are then applied andfconclusions are drawn fiom the-data as-to whether the'survey unit meets the site release criteria.
5.2.5.4 Survey Results Survey results-are documented in history files, survey'unit release records, and in the'FSS report. A FSS report is prepared that summarizes the data and states the conclusions of t~ie survey process.
5.2.6 Project Management The planning and implernentation of the FSS-1r6ces's is perforriied by SNEC Fa'cility personnei supplemented by the Decontamination and Decommissioning Engineering group of GPU Nuclear, Inc. Aspects of the final survey project are outlined below.
5.2.6.1 Final Survey Organization The Program Director, SNEC Facility - serves as the primary decision-maker (often called the Project Coordinator/Manager (PCIM)) iinid i 'responsible'for the overall implementation of the FSS process.
The SNEC Facility Radiation Safety Officer (RSO) - is the lead individual on the FSS team. He is responsible for managing the FSS processrcand will 'ct as" a technical cohnultant to'the PC/M.
The SNEC-Site Supervisor- -'leds 'the Dismantlerri't Orgazation at"'the'SNEC site. -In addition, he provides craft support to "the ES
-~n O..ai. '
ra the' SNEC site jn" addtio, h prvids caftsuportto heFSS team egsca'ffold erection, to permit surveys-of difficult to access areas.
S ta e g.,- s
'ei The Manager Decontamination and Decommissioning Engineering provides engineering support to the FSS Team and Dismantlement Organization.
Oversight of the process to assure compliance with the Quality Assurance/Quality Control (QA/QC) Program is provided under contract with qualified third party contractors.
5-17
5.2.7 Quality Assurance and Quality Control (QAIQC)
QA/QC is an integral part of all FSS activities. The objective of QA/QC, as applied to the FSS program, is to ensure the survey data collected, are of the type, and quality needed to demonstrate that the site is suitable for unrestricted release.
Proper application of QA/QC activities will ensure that:
- 1. The elements of this plan are correctly implemented as prescribed,
- 2.
The quality of the data collected is adequate, and
- 3.
QA/QC reviews will identify deficiencies so that corrective actions, when needed, are implemented in a timely manner and confirmed to be effective.
The basis of the SNEC Facility FSS QAQC program with respect to on-site sampling and analysis, has been 'derived from the applicable sections of Regulatory Guide 4.15,,,"Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment" (Referenice 5-12). The program is implemented at the SNEC Facility through procedure E900-QAP-4220.02, "SNEC Count Room Quality Assurance Program".
The SNEC Facility Decommissioning Quality Assurance Plan (DQAP) (Reference 5-13) is discussed in Section 3.0 of the SNEC Technical Specifications, and is applicable to all final status survey activities.
Examples of the QA program application are described in the subsections that follow.
5.2.7.1 Training Training is conducted to achieve initial proficiency and to maintain that proficiency throughout the final survey process.
Personnel performing surveys receive training to qualify in the procedures being performed. Training includes:
- 1.
Overview and objectives of the FSS process
- 2.
Procedures governing the conduct of the final survey
- 3.
Operation of the appropriate field and laboratory instrumentation
- 4.
Collection of final survey measurements and samples
- 5.
Survey data evaluation and documentation The extent of training and qualifications is commensurate with the education, experience, and proficiency of the individual and the scope, complexity, and nature of the activity. Records of training are maintained in accordance with approved facility procedures (see Table 5-3).
5-18 SNEC FACILITY LICENSE TERMINATION PLAN RE*VIRION 1
5.2.7.2 Written Procedures All FSS tasks, that are essential to survey data quality, shall be implemented and controlled by approved procedures -or-work instructions.
Final Status -Surveys will, be performed in a controlled, deliberate manner, providing assurance 6f accurate results. Applicable provisions of the SNEC Facility Decommissioning Quality Assurance Plan (1000-PLN-3000.05, Reference 5 13),, and the-SNEC Radiation Protection Planr (E900-PLN-4542.01, Reference 5-14), apply to all FSS activities. Other implementing procedures have beeri or are under development to support FSS activities.' The following matrix (Table 5-3) 'provides procedure titles -relating to' SNEC decommissioning and FSS issues, including FSS design and report generation.
N-,
5-19 FIEVISION *1
-_SNI=C FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LIC=NRF* TFRMINIATIrAN Pl AM Table 5-3 SNEC Procedure Matrix Listing Area Of Applicability Procedure Title Number Quality Assurance Plan -
Saxton Nuclear Experimental Corporation Facility 1000-PLN-3000.05 Decommissioning Quality Assurance Plan Embedded Pipe and Treatment of Embedded Piping and Components E900-ADM-4500.58 Components Off-Site Dose SNEC Facility Off-site Dose Calculation Manual E900-PLN-4542.08 Radiation Protection SNEC Radiation Protection Plan E900-PLN-4542.01 ALARA ALARA Review Program E900-ADM-4500.35 Sample Analysis Operation of the Packard Tri-Carb 2550 Liquid E900-OPS-4524.46 Scintillation Analyzer Gamma-ray Spectroscopy Operation of the Portable Gamma Spectroscopy E900-OPS-4524.43 System Instrumentation SNEC Rad Con Instrument Operations Manual E900-OPS-4524.42 Gamma Scans of Samples Operation of the SNEC Gamma Spectroscopy E900-OPS-4524 33 System FSS Survey/Sampling Survey Methodology to Support SNEC License E900-IMP-4520.04 Termination Gridding Areas Establishing a Reference Coordinate Grid System E900-IMP-4520.03 Sample Preparation Preparation of Sample Materials for Analysis E900-IMP-4520.02 Isolation of Areas Post Remediation Isolation E900-ADM-4500.54 Source Accountability SNEC Radioactive Source Accountability and E900-ADM-500 53 Control Program Decommissioning Safety SNEC Facility Regulatory Review Process E900-ADM-4500.52 ALARA SNEC ALARA Program E900-ADM-4500.47 Calculations SNEC Facility Calculations E900-ADM-4500.44 Training Training Requirements for SNEC Facility Workers E900-ADM-4500.42 Chain of Custody Chain of Custody for Samples E900-ADM-4500.39 Air Sampling Environmental Air Sampling E900-ADM-4500.30 Environ mental Measurements Environmental Monitoring E900-ADM-4500.22 Radiological Surveys Radiological Surveys: Requirements &
E900-ADMA500.12 Documentation Procedures & Work SNEC Facility Procedures and Work Instructions E900-ADM-4500.07 Instructions FSS Design Final Site Survey Planning E900-ADM-4500.59 Records Records Retention E900-ADM-4500.04 Quality Control SNEC Count Room Quality Assurance Program E900-QAP-4220.02 Radiological Quality Assurance Program for Radiological E900QAP4220.01 Instrumentation Instruments Contamination Control Alpha Control Program 6575-ADM-4500.51 Soil Erosion SNEC Soil Erosion and Sedimentation Control 6575-PLN4542.02 Plan 6575-PLN-4542.02 Process Control SNEC Facility Process Control Program 6575-PLN-4542.09 NOTE: This procedure listing is subject to change 5-20 I____________
t.
D I:*11* It't kl 4
5.2.7.3 Instrumentation Selection, Calibration and Operation Proper selection and use of instrumentation will ensure that sensitivities are sufficient to detect radionuclides at the minimum detection requirements specified by the survey design which will assure the validity of the survey data. Instrument calibrations will be performed in accordance with approved GPU Nuclear, Inc. and/or vendor supplied procedures,,using calibration sources traceable to the National Instit6te of Standards and'-Technology (NIST), or by qualified contractors providing results traceable to NIST. Operation'of all survey instrumentation will be established by approved procedures.
5.2.7.4 Sample Chain of Custody One of the most important aspects of sample management is to ensure that the integrity of the sample is maintained; that is, that there is an accurate record of sample collection, transport, analysis, and disposal. This ensures that samples are neither lost nor tampered with and that the sample analyzed in the laboratory is actually and verifiably the sample taken from a specific location in the' field.-' SNEC has developed procedures for sample chain-of-custody busing NUREG-1575 guidelines,(Reference 5-5).
5.2.7.5 Quality Control Surveys Quality 'control surveys vwill be made for all structures, 'facilities arid. oen land areas.
These measurements 'vill be performed 'in accordance with Section 5.4.5.
Quality conitrol surface contamination measurements will be compared to the original FSS measurernients. If the same conclusions are reached without any exceptions, the,original.FSS results for these survey units will be accepted 'as satisfactory. 2 All discrepancies between FSS' surVey,results and those obtained from quality control re-survey efforts shall be resolved with documented resolutions. If quality.control measurements fall outside of -their acceptance criteria, a documented investigation wiIl be performed which may result in re-s'urvey, re-sampling or other management actions.
5.2.7.6 SNEC Facility Sample Analysis High-resolution gamma spectroscopy is used to identify gamma-emitting isotopes, in addition to other appiropriate measurement methods'used to determine the total radionjuclide composition of sample materials Sample data results will be reviewed in accordance with existing requirements to ensure a reasonable interpretation of results.
Based on characterization,data describedinChapter2, of this plan, samples obtained at the SNEC'Fa~ility -pmay contain one or more of the-radionýUclide, li sted in, Table-5-4.
Minimum Detectable Concentrations' (MDC) determined -during'afialysis'will be performed using count times and instrumentation sensitive 'enough to detect typical enviroinmental levels for 'these radionuclides (as'aporbpriate).
5-21 REVISION 1 SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LICFWN*I TERMINATInlNl Pl AM SNE FACILIT A,
i'rVlIljld I
Table 5-4 SNEC Facility Radionuclides of Concern Open Land Areas & Structures Am-241 Ni-63 C-14 Pu-238 Co-60 Pu-239 Cs-1 37 Pu-241 Eu-1 52 Sr-90 H-3 NOTE: This data was denved from existing information (i e, Historical Site Assessment, characterization and scoping surveys).
SNEC Facility samples will be analyzed in accordance with the following requirements:
- 1.
All samples will be analyzed using a Gamma Spectroscopy System.
- 2.
If consistent radionuclide ratios have not been determined from existing data (i.e.,
scoping or characterization results), at least 5% of samples from the final status survey will include analysis for all radionuclides of concern, including Transuranic and Hard-To-Detect (HTD) radionuclides as listed in Table 5-4.
- 3.
Inter and intra laboratory cross-checks will be performed on at least 5% of those FSS samples which were analyzed onsite.
- 4.
QC samples will be taken per Section 5.4.5.1 and analyzed per SNEC procedure E900-QAP-4220.02, "SNEC Count Room Quality Assurance Program".
5.2.7.7 Access Control of Surveyed Areas and Systems Administrative and physical controls for access to surveyed areas will be established to preclude the possibility of re-contamination subsequent to completion of the final status survey in that area.
5.2.7.8 Control of Vendor Supplied Services Quality-related services, such, as'instrument calibration and laboratory analysis, are procured from qualified vendors whose irntemal QA program is subject to approval in accordance with the SNEC Facility DQAP. Vendors or contractors that are not listed on an approved vendor listing shall perform all necessary wdrk in accordance with SNEC procedures and work instructions.. If vendor services include survey or sampling data to be used as FSS results, these services will be reviewed and approved in accordance with the SNEC procedure (E900-ADM-4500.44, Reference 5-15).
5.2.7.9 Audits and Independent Reviews Periodic audits are performed to verify that survey activities comply with established procedures and other aspects of the SNEC Facility DQAP and to evaluate the overall effectiveness of the program. The audits are conducted in accordance with approved procedures and performed by individuals who are independent of the activities being audited. Audit results are reported to 5-22
SNEC FACILITY LICENSE TERMINATION PLAN Pl:,.qVIfM 4 Sresponsible mahageme'ht inr writing, and a'ctions to resolve identified deficiencies 'are tracked and appropriately documented. Qualified personnel will perform an independent review of the Final Status Survey Report.
This review will ensure that FSS results are performed and documented in accordance with appropriate methodology, and that all conclusions reported are accurate and correctly presented.
5.2.8 Survey Records and Documentation Generation, handling, and storage of FSS design information and survey data are-controlled by approved procedures. Survey records and documentation' are maintained as'quality records and decommissioning records in accordance with approved facility procedures.
Where possible, they are also maintained as electronic media.
At a minimum, each final status survey record will include:
- 1.
Date and time survey was performed
- 2.
Instrumentation used and calibration due date(s)
- 3.
Survey location (grid location or other reference markings)
- 4.
Type of measurement performed (scan survey, fixed-point measurements etc.)
- 5.
Survey team member(s) involved
- 6.
Name of field supervisor(s) responsible for reviewing survey-data 7.. Survey and Sample Request numbers Generation, handling and storage of the original final status survey design and data' packages shall be in accordance with the SNEC Records Retention procedure (E900-ADM-4500.04, Reference 5-16) and Radiological Surveys: Requirements & Documentation procedure (E900 ADM-4500.12, Reference 5-17).
5.2.9, Calculations Formal calculations that support License Termination activities are prepared in'accordance-with the SNEC" Facility Calculations Procedure (E900-ADM-4500.44,,'Referrence 5-15).
'These calculations provide sufficient details with "respect'to'purpose, method; assumptions, design input, references and units such that a person technically qualified in the subject 6an review and understand the analysis as well as verify the 'adequacy 'of the results without 'frequently consulting the originator. Calculations may be used for activities such as survey design, dose modeling, and computer code verification.
5.2.10 Schedule Final status surveys are planned, scheduled, and tracked as 'a' part-of the overall decommissioning planning process.
The schedule is dependent upon the progress and completion of several decommissioning activities and' review and "approval of the License Termination Plan. Presently, survey data collection is expected to begin in the fourth quarter of 5-23 I* I:VI_* I 1"1N 4
2002. Final survey activities are planned and will be discussed with the NRC in advance to allow scheduling of the required public meeting on the License Termination Plan.
5.2.11 Stakeholders The stakeholders for the SNEC decommissioning project include those organizations and concerned individuals listed below:
Citizens Task Force (CTF)
Concerned Citizens for Saxton Safety (CCFSS)
Liberty Township Huntington and Bedford Counties The Commonwealth of Pennsylvania FirstEnergy Companies Applicable Contractors US Army Corps of Engineers 5.3 FINAL POST REMEDIATION SURVEYS The professional judgment of the SNEC Facility staff will be used to implement final post remediation surveys in areas where former contamination levels required extensive remediation or in other areas as deemed appropriate. Properly designed, post remediation surveys can facilitate the transfer and control of areas, and minimize the impact of planned or ongoing dismantlement activities in adjacent areas.
5.3.1 Walk-down A walk-down of the survey unit is performed prior to isolation. The principle objective of the walk-down is to assess the physical state of the survey unit and the scope of work necessary to prepare it for final survey. During the walk-down, requirements are identified for accessing, isolating, and controlling the survey unit.
Support activities necessary to conduct the, final survey, such as scaffolding, interference removal, and electrical tag-outs, are identified.' Safety concerns such as confined space entry, high walls, and ceilings are identified. For systems, the walk-down includes a review of system flow diagrams and piping drawings. The walk-down is performed when the final configuration is known, usually near or after the completion of dismantlement activities.
5.3.2 Isolation Criteria The following criteria will be satisfied prior to acceptance of a survey unit by the FSS team. The physical aspects of these criteria are verified during the walk-down.
- 1.
Planned dismantlement activities within the post remediation survey unit are completed.
5-24 SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1
-qMFt'- PAr'll ITV I I1'=MQC: T=DMIMKATIAM1 01 AKI
~~~~~~~r
- 1. aff*If
[UN.
.~l*
- 2.
Planned dismantlement activities affecting or adjacent to the post remediation survey unit are completed, or are evaluated and determined to not have a reasonable
.potential to introduce radioactive material into the post remediation survey unit.
- 3.
An operational radiation protection survey of the post remediation survey unit is completed and all outstanding items are addressed.
-4.
Planned physical work in, on, or around a'post remediation survey unit,',other than routine surveillance or maintenance, is complete.
- 5.
Tools, non-permanent equipment, and material not needed for survey datacollection are removed.
-6.
Housekeeping, clean up, and remediation of the survey unit are completed.
- 7.
Scaffolding, temporary electrical and.ventilation equipment, and components, and other material or equipment needed for survey data collection is radiologically-clean and left in place.
- 8.
Transit paths to/through the post remediation survey unit are eliminated or re-routed.
- 9.
Appropriate measures are instituted to prevent the re-introduction of radioactive material into isolated area from ventilation systems, drain lines, system vents, and other potential airborne and liquid contamination pathways.
- 10. Measures are instituted to control access and egress. and otherwise restrict radioactive material from entering the survey unit.
5.3.3 Transfer of Control Once a walk-down has beern, performed and the isolation criteria are met, control of activities within the post remediation survey unit is transferred from the dismantlement organization to the FSS team. The need for localized remediation within the isolated area may, be identified after transfer of control.
Localized remediation may be) performed under the control of the FSS organization. However, if large areas require remediation, the isolated area may be transferred back to the dismantlement organization forfurther decontamination.
5.3.4
,Is olatioinand Control Measures Prior to performing the FSS, the post remediation survey unit is isolated and controlled. Routine access, equipment removal, material storage,-and worker and material transit through the area without proper controls are no longer, allowed. One or more ofthefollowing administrative and physical controls will be established to minimize the possibility of introducing radioactive material from ongoing decommissioning activities in adjacent or nearby areas.
- 1.
Personnel training
- 2.
-Installation of barriers to control access to the area(s)-
- 3.
Installation of postings with access/egress requirements
- 4.
Locking or otherwise securing entrances to the area 5-25
- 5.
Installation of tamper-evident seals or labels Isolation and control measures are implemented through approved facility procedures and remain in place through the FSS data collection process until license termination.
5.4 SURVEY DESIGN The survey design identifies relevant components of the FSS process, and establishes the assumptions, methods, and performance criteria to be used. The methodology for planning a FSS, including a FSS in the subsurface region is identified in the applicable site procedure.
Survey design is summarized in Table 5-5.
The application of survey design criteria to structures and land areas will vary based on the type of survey media and the relative potential for elevated residual radioactivity.
For facility systems, many of the survey design criteria applicable to structures and land areas do not apply or are dictated by the physical system layout and the accessibility to the system piping and components. To accommodate these factors, the survey design integrates both non-systematic (random) and judgmental (biased) methods to data collection to achieve the overall objective of the final survey process. Survey design will be performed in accordance with SNEC procedures E900-ADM-4500.59, "Final Site Survey Planning" and E900-ADM-4500.58, "Treatment of Embedded Piping and Components". When necessary, a two-stage sampling process may also be used lAW Reference 5-20.
Each survey design package will address the following areas of interest:
- 1.
A brief overview describing the final status survey design;
- 2.
A description and map or drawing of impacted areas of the site, area, or building classified by residual radioactivity levels (Class 1, Class 2, or Class 3) and divided into survey units, with an explanation of the basis for division into survey units. Maps should have compass headings indicated;
- 3.
A description of the background reference areas and materials, if they will be used, and a justification for their selection;
- 4.
A summary of the statistical tests that will be used to evaluate the survey results, including the elevated measurement comparison, if Class 1 survey units are present, a justification for any test methods not included in MARSSIM, and the values for the decision errors ( and ) with a justification for values greater than 0.05;
- 5.
A description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide
- 6.
For in-situ sample measurements made by field instruments, a description of the instruments, calibration, operational checks, sensitivity, and sampling methods, with a demonstration that the instruments, and methods, have adequate sensitivity; 7
A description of the analytical instruments for measuring samples in the laboratory, including the calibration, sensitivity, and methodology for evaluation, with a demonstration that the instruments and methods have adequate sensitivity; 5-26 SNEC FACILITY LICENSE TERMINATION PLAN RF*VISION 1
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1
- 8.
A description of how the samples will be collected, controlled, and handled;
- 9.
A description of the final status survey investigation levels and how they were determined;
- 10. A summary-of any significant additional residual radioactivity that was not accounted for during site characterization;
- 11. A summary of direct measurement results and/or soil concentration levels in units that are comparable to the DCGL and, if data is used to'estimate or Update the survey unit;
- 12. A summary of direct measurements or sample data used 'to both evaluate the success of remediation and to estimate the survey unit variance.
5-27
SNEC FACILITY LICENSE TERMINATION PLAN IIr,11AM
,1 Table 5-5 Survey Design Summary Class I Class 2 Class 3 Plant Specification Pln Structures T Land Areas Structures Land Areas Structures Land Areas Systems SUGGESTED SURVEY UNIT & REFERENCE GRID SIZES Up to Up to 100 to 2,000 to Size Range=>
100m 2 2,000 m2 1,000 m2 10,000 m2
< 10,000 M2
< 10,000 m2 N/A Reference Coordinate 1 to 2 m 10 to 20 m 1 to 2 m 10 to 20 m 5 to 10 m 20 to 50 m N/A Grida=:
SCAN MEASUREMENTS Scan 10 to 1 0 0%b for Floors &
Coverage=>
100%
10 to 50% for Upper Judgmental, up to 10%
Variablec Walls & Ceilings Scan Area Judgmental, systematic Judgmental; Selection=>
Accessible surface areas along transects or of Judgmental, random surface randomly selected grids surea area STATIC MEASUREMENTS Number of Measurement Calculated using the methodology of Appendix 5-2 asu (Default Value is 30e for Small Survey Area Sizes, see Section 5 4 3 2)
Locat=on Accessible Selection=
Random starting point, systematic spacingg Random Points Spacing (L)=>
L = (A/n)11 2 is for a square grid (see Section 5.4 3 2)
N/A N/A A = total survey unit area; n = # of measurements Type of Surface So Mea Surface Soil MeaS h Surface Soil Meas Surface SurveypS Cont.
So M Cont Cont T Cont' NOTES:
a)
A square or triangular gnd system pattern is used and multiple grid patterns are employed as necessary unless survey needs dictate otherwise.
b) Where scanning coverage greater than 50% is judged appropriate, the survey unit may be reclassified as a Class 1 survey unit.
c)
Performed according to the scan coverage for the class of survey unit (where possible).
The amount of accessible surface area dictates the percentage of surface area scanned d)
Includes health and safety considerations.
e)
This number is sufficient for structural survey units less than -10 square meters and land areas less than -100 square meters in area, and may be used for embedments such as brackets, unistrut or sections of piping.
f)
As allowed by plant system size and accessibility to system interior surfaces.
g)
Except when statistical tests are not applied (i e, post remediation survey data for a survey unit are all less than DCGL) When systematic spacing is not practical, a random survey/sampling pattern will be used.
h)
Subsurface samples will be obtained from randomly selected locations as well as biased locations (see FSS Design procedure (Table 5-3) i)
Scale and sediment samples will be collected from embedments (e g., piping, unistrut) as appropriate 5-28 X-I* I:::*rl
- I I*/'d 4
5.4.1 Survey Units As reflected in Table 5-5, impacted areas are divided into survey units to -facilitate -survey design. A survey unit isa physical area of specified size and sfhape with rsirila"r charactdristics and potential for residual radioactivity for which data analysis and statistical analysis are performed. A separate decision is made for each survey iunit'as' to its ace'ptability for relea'se.'
5.4.1.1 Survey Unit Size Professional judgment is used to'divide facilities and areas into iappropriately sized survey nits.
Survey unit sizing is sufficient to assure'that the total ýnurfiber of data points,- based -6n' the measurement,frequency, will'allow statistical evaluation of -the data." Considerati6rAs-folr establishing survey units are physical -characteristics, concentration levels, isotopic2 ratios, and previous remediation efforts, as Well asspatialand logistical considerationsi '"
Survey units are sized to ensure data points are relatively uniformly 'distributed 'among areas 7of similar potentialfor residual contamination.
Small survey units 'are develdped to ensure a conservatively -established coverage of an area, 'location or residdal syster'ipipinrg or embedment.
Survey units conform to site physical characteristics to the extent practical. SLurvey, units-have relatively compact shapes unless an unusual shape is appropriate because of site operational history or site topography.
Where 'possible,' existing -facility or site 6haeacteristics such as horizontal and vertical structural supports (beams, concrete p our s~eams,'-or piping runs) are used to define the boundaries of the study. Suggested survey unit sizes are given in Table 5-5.
Survey"6nit sizes may b6 adjusted as necessary as lonIg as assilmpti6nh used to dedvelop area dose models remain valid.
5.4.1.2 Reference Coordinate System A reference coordinate system is used to' facilitate'selection of mea'su'rerment location's, arid provide a method for locating the same points in the future: A reference coordinate system 'isa set of intersecting lines referenced to a fixed site location or benchmark." Ty'pically; grid lines are arranged in a perpendicular pattern, dividing the survey unit into squares of equal area.
However, other types-of grid patterns maybe used. Examples of ai reference coordinate system are shown in Figures 5-4 and 5-5, located at th& enid of this cha ite'r.,
Scale drawings, maps,'or photographs -of eachl sirvey'unit 'are preparec, along 'with an ove-rlay of the reference coordinate system. 'While some~of theI pspre'sehted with this'plan 'have a grid overlay, the actual or final lay6ut may be soinewh~t differeritthan that provided herein. The reference coordinate system provided on'.attache*d'drawuinrgs are intendad* primarily'-for illustration purposes and do not necessarily dictate final grid layouts.
Physical gridding is used only-where it ii useful and 16bst effective.: Wh6ere Class1 anid Class 2 survey units'are gridded, the basic -grid'patterns-are* *at-l,to',2vmeter intervals pon striL'ctuial' surfaces, and at 10 td-20 meter intervals over land areas. For practical pur'pos es, Cla'ss 3 areas' may typically be gridded at larger intervals,' fore.ýample,'5 to 10 rnieteirs-fo-lairge ur faces' nd 20 to 50 meters for land areas. -
The physical grid layout on a structural surface is marked-by chalk line 0r-other appropriate" means along the entire reference line or only at line' intersectionsi. Forilarid areas, the reference 5-29
'SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I coordinate system is marked by wooden stakes or metal pins driven into the soil at line intersections, or by other appropriate surface markers. The selection of an appropriate marker depends on the physical characteristics and routine use of the survey area. Examples of gridded structures are shown at the end of this chapter (Figures 5-4 & 5-5).
5.4.1.3 Background Reference Areas The residual radioactivity of a survey unit may be compared directly to the DCGL. However, the residual radioactivity may contain radionuclides, which occur in background.
To identify and evaluate those contributions attributable to licensed activities, representative background radionuclide concentrations are established using background reference areas. Background reference areas have similar physical, chemical, radiological, and biological characteristics as the areas to be surveyed. They are usually selected from near site non-impacted areas, but are not limited to natural areas undisturbed by human activities. Surveys will be conducted of one or more background reference areas (where appropriate), to determine background levels for comparison with specific survey units. However, environmental background concentrations of Cs-137 will not be subtracted from nuclide specific measurement results (e.g., gamma-ray spectroscopy measurements of sample materials etc.) for any open land area. Appendix 5-3 provides additional discussions in selecting and applying background reference areas.
Background reference areas are not necessary where:
- 1.
The residual radioactivity does not contain radionuclides occurring in background and the detection method is radionuclide-specific.
- 2.
The background levels are known to be an acceptably small fraction of the DCGLw 5.4.2 Scan Measurements Scan measurements are performed to locate elevated areas of residual radioactivity that will require further investigation., They are performed according to a preset pattern established for each classification.
The level of scanning effort is proportional to the potential for finding elevated measurement results.
Scan measurements of Class 1 survey units are performed over 100 percent of the accessible surface area. However, personnel. health and safety issues are taken into consideration when determining whether an area is accessible. Scan surveys are designed to detect small areas of elevated radioactivity that would not be detected by a select-number of static measurements, using a systematic-measurement pattern.
If the sensitivity of the scanning method is not sufficient to detect levels of residual radioactivity below the DCGL, the number 'of static measurements may be adjusted appropriately.
Appendix 5-2 describes how this is accomplished.
Scan measurements of Class 2 survey units are typically performed over 10 to 100 percent of the'surface area. Class 2 survey units have a lower probability of elevated residual radioactivity than Class 1 survey units.
Those areas with the highest potential for elevated residual radioactivity (e.g., corners, ditches, and drains) are included in the survey based on professional judg'ment. If the entire survey unit has an equal probability of having areas of elevated residual radioactivity, systematic scans are performed along grid intersections of randomly selected grid blocks. A 10 percent scanning coverage is appropriate if it is unlikely that any area would exceed the DCGL. Coverage of 25 to 50 percent is appropriate when there may be locations 5-30 a-
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I t above the DCGL. Where scanning coverage of greaterithan 50 percent is judged appropriate,
- the survey unit maybe 'reclassified as a Class 1 'survey dnit.
Scan measurements of'Class 3 survey units,are performed typically over approximately -10 "percent Of the surface area.'Corners, ditches, 'and drainage' or collection:areas are included in the s'urvey base'd 6n professional judgment. Class 3 survey uInits have 'the lowest probability of containing elevated residual radioactivity. Those areas with the highest potential for elevated residual radioactivity, based on professional judgment, are selected for scanning in a Class 3 area.
Scan measurements of facility systems 'are performed acco'rding' to req'uired-scan coverage for thle clissification" of the survey unit.- The -am6unt of accessible surface. area dictat'esthe perceritage of the surface area scanned. However, the majority of'all SNEC Facility sistems and *components have been removed from the site. When scanning'is, not'possible,'sbcl-as in water-covered areas, outfalls etc., a sufficient -number of randomly selected samples will be colle cted as appropriate'.
5.4.3 Static Measutrements Static measurements provide a quantitative measure of the radioactivity present at the location measured. Static measurements are performed at a'frequericy-and location, throughout each survey unit,' such that a statistically souInd conclusion can be developed. Static measurements may be,performed. at locations of :elevated.residual radioactivity identified,by scan measurements. These types of static.measurements may include direct surface contamination measurements, and soil and bulk material measurements.
There are several vendors that can supply semi-automated, large area, -position sensitive, radiation measurement equipment. The use of this type of instrumentation is applicable in areas where relatively flat surfaces exist (either for structures or for surface soils). Scanning results using this edquipment could be acceptabl, substituied for static measurements when the'scan MDC is'well below'the 'requirements for releasing the area (6.g., 1 10% of the applicable DCGL).
This"type of equipment ca'n provide greater ýconfidence in the survey results in that-surveyor, error is greatly reduced and-ty'ic'al dlete-ctidn-sensitivities are usually higher than that obtained using hand held survey equipment. In addition, an entire area is more appropriately scanned at 100% coverage making statistical testing of survey areas unnecessary.
GPU Nuclear, Inc. is currently evaluating these types of equipment and their capabilities in an effort to expedite and possibly improve overall scanning capabilities. If instrumentation of, this type is used for scan measurements and the measurements are capable of providing data of sufficient quality as that provided by static measurements, theycould be used in place of static measurements. The same logic may be applied for using in-situ gamma' sp-ectrometry in place of sampling and analysis for soil and other volume.contaminated materials in concert with appropriate surrogate radionuclides. However, GPU Nuclear,-Inci.
has ag-reed that s-oil samples will still be collected in open land areas additional tothese semi-automated scan survey-or-in situ gamma spectrom try special 'measu-remenrt techniques.
5-31
5.4.3.1 Number of Measurements As described in Appendix 5-2, the MARSSIM process incorporates design constraints that ensure that an adequate number of sample measurements are taken per survey unit. However, a minimum number of 30 measurement points per survey unit may be collected in areas or locations where there is difficulty in applying MARSSIM fixed-point requirements, because of survey unit size, access problems or obstructions. Survey units of this type will be investigated thoroughly prior to planning the FSS.
5.4.3.2 Measurement Locations Measurements in Class 3 survey units and background reference areas are taken in random locations. Random means that each measurement location in the survey unit has an equal probability of being selected. The random selection process uses random numbers that correspond to a survey unit's 'reference coordinate system to establish the measurement locations within the survey unit. The random numbers are generated using a random number generator or other acceptable random selection process. Measurement locations selected that do not fall within the survey unit area, or that cannot be surveyed due to site conditions, including health and safety considerations, are replaced. A technical evaluation will be written to justify not surveying an area due to safety and health considerations.
Remaining sections of facility system piping, and embedments of all types with limited accessibility, may not allow the'full number of measurement points. When this is the case, an assessment of the embedment is performed to determine if the embedment can easily be removed or can be prepared for the FSS survey. The history of the'area where the embedment resides is first reviewed along with the expected radionuclide& concentration present on/in the embedment. In these special cases, the same MARSSIM approach encompassing historical assessment, characterization, remediation, and post remediation survey data is used as a basis for biased scanning and sampling to the extent practicable to ensure that the release criteria are met.
For Class 1 and Class 2 survey units, a random-start systematic pattern is used in place of a random pattern. This is done to meet survey design objectives and to locate small areas of elevated residual radioactivity that may exist within the survey unit. A random selection process determines the starting point. For a survey point placement using a square grid, the physical spacing of the measurement locations, L, is determined as follows (NUREG-1575, Equation 5-6):
L = -2A/n Where:
A
= area of the survey unit nEA = number of measurement points The calculated value of L is rounded down to the nearest distance easily measured in the field.
Using the reference coordinates, the measurement locations are identified around the starting point in a perpendicular manner at intervals of L. This process is repeated to identify the pattern of measurement locations throughout the survey unit. Where other than a square grid system is used, the physical spacing of the measurement locations is determined such that they are 5-32 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
distributed around the starting point in a systematic, equidistant manner across the survey unit area.
Equation 5-5 of NUREG-1 575 is used when a triangular grid sampling point layout is used.
L =.IA / 0.866nEA Measurement locations selected using a random selection process or a systematic pattern that do not fall within the survey unit area, or that cannot be surveyed due "to`'ite conditions, including health and safety considerations, may be replaced. A technical evaluation will be written to justify not surveying an area due to safety and health considerations. Replacements are made using other measurement locations determined by the randomh,selection process.
Supplemental measurement locations are also determined using the random selection process but may require previous survey design considerations such as those listed in Reference 5-20.
Measurement locations selected based on professional judgment violates the assumption of unbiased measurements used to develop the statistical tests and are not used in the statistical evaluation.
However, special considerations are necessary for structural survey units with surface areas less than 10 M 2, land areas less than 100 M2, and remaining site system piping and other embedments. "The data generated from these smaller survey units may be obtained using professional judgment, rather than systematic or random sampling design: Once again, the number of 30 measurement points for each of,.these smaller survey units may be used whenever practicable. Additionally, when'the random-start systematic spacing pattern cannot be applied correctly because of a survey areas limited size or accessibility, then-a completely random selection process will be used for purposes of locating survey and,sampling points within the survey area.
5.4.3.3 Location Identification Measurement locations within the survey unit-are-clearly identified and documented to ensure that they can be found again as necessary. Actual measurement locations are -marked with tags, self-adhesive labels, bar codes, permanent markings, stakes, notations on survey maps, or equivalent methods. A unique identification" code or number identifies each measurement location.
The number convention allows survey data to be referenced back to a specific measurement point that is identified in photographs, drawings, diagrams or maps of the survey unit.
5.4.4 Data Investigation The data collection, investigation and evaluation, process checklist is presented in Figure 5-2.
Use of this checklist or a similar organizer-pr0ovide's a high degree of confidenIce that all data requirements have been met.
i I I
- 1. I 5-33
'SNEC FACILITY LICENSE TERMINATION PLAN
ý.
- VREVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I Figure 5-2 Example of a Data Interpretation Checklist 1
Collect Data Structural Open land areas Embedments
- 2.
Convert Data to Standard Units Structural activity in dpm/1 00 cm 2 Solid media (soil, etc.) activity in pCi/g
- 3.
Evaluate Elevated Measurements Identify elevated data/mark location Compare data with derived elevated area criteria Determine need to remediate and/or reinvestigate elevated condition Compare data with survey unit classification criteria Determine need to investigate and/or reclassify
- 4.
Assess Survey Data Review survey planning and design data Verify that data of adequate quality and quantity were obtained Perform preliminary assessments (graphical methods) for unusual or suspicious trends or results - investigate further if appropriate
- 5.
Perform Statistical Tests Select appropriate tests for category of contaminant Conduct tests Compare test results against hypotheses Confirm power level of tests
- 6.
Compare Results to Guidelines Determine average or median concentrations Confirm that residual activity satisfies guidelines
- 7.
Compare Results with survey design information (ALARA may be included in the survey design)
Determine if all planning and design requirements are satisfied Explain/describe deviations from design-basis survey requirements 5-34
- I-
SNECFACIITY ICENE TEMINAION LANRF=Vl~I(nm 5.4.4.1 Investigation Levels Examples of typical investigation levels are shown in Table 5-6, taken from the MARSSIM manual (NUREG-1575, Table 5.8). Investigation levels are radioactivity levels that are based on the site release criteria, which if exceeded,, initiate an investigation of the survey data.
Investigation levels are typically established for each class of survey unit.
Table 5-6
- Typical Investigation Levels (from NUREG-1575)"
Survey Unit-Flag Direct Measurements or Flag Scan Classification Sample Result When:
Measurements When:
Class I
>DCGLEMC or >DCGLw and > a statistical
>DCGLEMC parameter based value Class 2
>DCGLw
>DCGLw or > MDC Class 3
>fraction of DCGLw
>DCGLw or*> MDC '
The principal purpose of an.investigation level is to guard against the possible misclassification
,of the survey unit. They also serve as a QC check during the final survey process. A survey 0measurement that exceeds an investigation level may indicate that the survey unit has been improperly classified. It-may also indicate a failirj surv6y-instrument-or,a localized area of
,elevated residual radioactivity where there was a failure in the'remedi tion process.
Large
.variations in background may also result in investigative surveys.' -mniestigative surveys of this
,type may be performed using a variety of measurement tools-including' in-situ gamma-ray spectroscopy instrumentation.
!Depending upon the results -of the investigation, survey units may-require no action, may be, iremediated, 'or may be reclassified and/or re-surveyed.
Initial adrmnirstrative, acti6n 'or,
-investigation level guidelines may be found in'Tablle 5-7.
For a Class.1 survey :unit,: while'
!measurements above the DCGL are not necessarily unexpected, any measurement exceeding' the DCGL is investigated. The site release criteria allows individual measurements representing Ismall areas of residual radioactivity to exceed the DCGL. ý.,However, any, measurement that exceeds the DCGL is subject to the elevated measurement comparison "(EMC), described in
'Appendix 5-1. For a Class 2 survey unit, any measurement above the DCGL is unexpected and is investigated. As there is a low expectation for residu'al radioactivity in a Class 3 survey unit,_
,any-above,background static measurement,- ekceeding-a small-fraction of -a -DCGL isý investigated. If the scanning MDC exceeds the 'DCGL, any'indication' of residual radioactivity during the scan is also investigated.
' I I
If a background reference area is to be applied to the survey unit, the meanof the background reference area measurements may be added to the appropriate investigation level to which the survey measurements are compared. Where an excessive, number of measurements exceed the investig ation level, the results are reviewed to'ensui-e thiat the applied background reference area is appropriate. Additionially, rml'ultiple'referenrce are'as-maybe used if.eferencearea', have significantly', differeht backgrounrdlevels be~aiusý- sf the variabilit~i in background between areas.
Furthermore,' when different -material types are Present, in-the survey,,area, the lowest background material average may be used along with the highest material sample variability to produce a conservative assessment'of residual activity in the survey area.
5-35 SNEC FACILITY LICENSE TERMINATION PLAN DPVI.qlt3N 1
SNEC FACILITY LICENSE TERMINATION PLAN RF:VIR~if'lN I Table 5-7 Summary of SNEC Investigation/Action Levels SURVEY AREA TYPE CLASS I AREA CLASS 2 AREA CLASS 3 AREA Flag any discrete Flag any discrete Flag any discrete measurement > DCGLw measurement > 50%
measurement > 10% of Investigation or or > mean + 3 sigma of Action Level the survey unit. Flag any of DCGLw. Flag any DCGLw. Flag any scan scan measurement measurement > DCGLw or>
DCGLw.
DCGLw or > MDC.
MDC.
If scan or discrete measurement values are >
If scan or discrete DCGLw, area should be measurement values reclassified as Class 1. If Reclassification N/A are > the DCGLw, scan or discrete Level area should be measurement values reclassified as Class exceed 50% of the DCGLw,
- 1.
but are < the DCGLw Area should be reclassified as Class 2 area Consider remediation when residual activity exceeds 3 times the DCGLw for any scan or discrete measurement averaged over 1m00 cm2, Not indicated for Class 2 or Class 3 survey areas Remediation or when the mean for any unless above limits specified for Class 1 survey area.
contiguous -M2 area is Re-classify and Re-survey, as necessary.
Guideline more than the DCGLw for discrete measurements.
Soil mean is > the DCGLw for 100 m2 area or DCGLEMc is greater than 3 times the DCGLw NOTE 1: The above values are initial values that will not be changed without applicable reviews in accordance with License Condition 2.E.(d).
All investigation/action levels assume background values have been subtracted (as appropriate).
5.4.4.2 Investigation Locations identified by scan or static measurements with residual radioactivity, which exceed an investigation/action level, are'ma'rked and investigated.
The elevated measurement is then confirmed to exceed the investigation/action level. The area around the elevated measurement is investigated to determine the extent of the elevated residual radioactivity, and to provide reasonable assurance that other undiscovered areas of elevated radioactivity do not exist.
Scan coverage of the area being investigated is increased to 100 percent (if not already at that 5-36 I________________
level). Static measurements are also taken if scan measuremrents are-not capable'of providing sufficient data to characterize the elevated area. A posting plot, described in Section 5.6.2.1, is generated to 'document the tarea investigated and the levels of-residual ýadioactivity 'found.
Depending 'on the results of the investigation, the survey unit,may -require remediation, reclassification, 'and/or re-survey. 'Possible outcomes' of the! data inrestigation process aire shown in Table 5-8 below.
Table 5-8 Possible Actions Resulting From Data Analysis No.'
Data Results Class I Class 2 Class 3 One or more data Perform statistical 1
points > DCGLEMC or testing, remediate and Re-classify & re-survey Re-classify & re DCGLw re-survey as necessary survey' Survey Unit passes 2
All data points <
applicable elevated N/A N/A DCGLEMC measurement comparisons
.. Determine if re All data points:5 Determine if re-claIssification'is 3
All Survey Unit passes classification is required DCGLWrequired as as follows below: -
f61I6ws below:'
One or more points >
Increase survey.
6 overa-ge or review &
Re-61assify' &'re'
- ---J/
4 50% of DCGLw but <
Survey Unit passes,'
'6vrg1 o eiw&
R-l'si oDCG Lbw re-classify & re-survey,,
survey as necessary One or more points >
5 10% of DCGLW but <
Survey Unit passes Survey Re-classify & re-'
50% of DCGLw survey..
6-All data points_ 10%
Survey Unit passes Survey Unit -asses SurveyUnit 6of DCGL.
passes Static measurements above the investigation/action level that should have been, but were not identified by scan measurements may indicate that the scanning method is inadequate. In that case, the scanning method is evaluated and appropriate corrective actions are taken.
Corrective actions may include re-*scanning affected survey units using other survey -protocol or survey instrumentation.
5.4.4.3 Remediation
'C I
Areas of, elevated -residual -radioactivity -above., the DCGLEMC+-are -remediated-to acceptable levels. Based on the survey data, it may be -necessary to remediate all or a portion -of a survey unit. Remediation activities are addressed in Chapter 4.0.
5-37
'-REVISION 1
SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 5.4.4.4 Subdividing SurveyUnits Due to size restrictions and other considerations, a survey unit may need to be divided into two or more smaller survey units. Survey unit sizes may be adjusted as necessary as long as assumptions used to develop area dose models remain valid. Suggested survey unit sizes are provided in Table 5-5.
5.4.4.5 Resurvey If a survey unit is reclassified or if remediation activities are performed, then a re-survey using the methods and frequency applicable to the new survey unit classification is performed. This includes the case where only a small fraction of the area (< 10%) of a Class 1-survey unit is remediated.
In the case where a new survey unit is separated out from an existing survey unit or an existing survey unit is subdivided, Class 3 survey units need only additional randomly located measurements to complete the survey data set. Class 1 and Class 2 survey units require a new survey design based on random-start systematic measurement locations.
When a new survey unit is separated out from an existing survey unit or is subdivided, the new survey unit will include a buffer zone that adequately bounds the area of identified contamination when it borders a non-impacted area.
5.4.5 Quality Control (QC) Measurements QC measurements are a component of the survey quality assurance process, and include quality checking and repeat measurements.
Quality checking and repeat measurements are performed to identify, assess, and monitor measurement error and uncertainty attributable to measurement methods or analytical procedures used in the data collection process. Quality checking includes direct observations of survey data and sample collections, and sample preparation and analyses.
Repjeat measurements are mLiltiple measurements at the same location or from the same survey unit. Repeat measurements provide quantitative information to demonstrate that measurement results are sufficiently free of error to accurately represent the radiological condition of the SNEC Facility. Results of QC measurements are documented in accordance with approved site procedures.
5.4.5.1 Type, Number, and Scheduling QC checks will typically be performed by randomly re-sampling and/or re-surveying 5% of all sampling and/or survey points. For a low number of points (10 or less), the number of re-survey or re-sample locations will not be less than one (1). The type, number, and scheduling of QC measurements may also be determined by a performance-based method as described in Section 4 9.2 of NUREG-1575. This method is based on the potential sources of error and uncertainty, the likelihood of occurrence, and the consequences in the context of final survey data accuracy.
The primary factors considered here are 1) the number of persons or organizations involved in the data collection, 2) the number of measurement types or analytical methods used, and 3) the time interval over which the data are collected. Other factors include:
- 1.
Number of survey measurements collected,
- 2.
Experience of personnel involved, 5-38
SNEC FACILITY LICI=N5I= TlFiMlNZATIAMk DI AM
- 3.
Types of measurement methods or sampling and analytical procedures used,
- 4.
Variability of survey instruments used,
- 5.
Level of radioactivity in the survey unit, and
- 6. "How close the measurement level isto the detection limit.
QC measurements will be collected throughoiit the data collection process to verify'that sources of error and uncertainty are minimized and controlled.
5.4.5.1.1 Scan Measurements Quality checking'of surface.'scanning surveys"is performed to evaluate the effectiveness of scanning methods for identifying areas of elevated residual radioactivity.- The frequency,.of quality checking for scan surveys is dependent on:
- 1.
The number of surveyors,
- 2.
The number of scanning methods employed,
- 3.
The time interval over which scanning data are collected,
- 4.
The scan MDC, and
,5.-
Professional judgment.
The ability of surveyors to identify areas of elevated residuai radioactivity by scanning will be periodically tested and documented in accordance with approved SNEC Facility lirocedure E900-ADM-4500.42, "Training Requirements for SNEC Facility Workers".'
5.4.5.1.2 Static Surface Contamination Measurements Repeat measurements of static surface -contamination are performried to assess error and uncertainty associated with field measurement~methods. Measu'rement locations are selected based on measurement results and are representative of tthe entire dynamic range of residual radioactivity found. The usable range'of radioactivity'includes thehighest measurement result and the lowest measurement result with an acceptable measurement unc6rtaiinti,.compared to, the desired level of accuracy. Repeat measbrirrmints with results at or neaý.the detection limit are not used because the'measurement uncertain'ty-is'usually greater tkan'the' d6sired level of,'
accuracy.-
5.4.5.1.3 Soil and Bulk Material Measurements Selected soil and bulk samples are analyzed and resubmitted for analysis as to determine,.
(measure) precision.' 'Replicate,samples 'may-be analyzed in their'entirety or split to,performo multiple analyses.(split sample). The original analytica'l result's are then cbpar'e dwith the replicate analysis results for agreement: When possible,`repllicate samples are analy'zed using a second (different) instrument and are" submitted 'to the laboratory(s) 'a4.unknowns.*
.,This, process,is implemented at the SNEC Facility through procedu'e :E900-QAP-4220.02, "SNEC Count Room Quality Assurance Program".
5-39 S...
v,., *.-*,
HI:VIbIUN I
I____________
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 5.4.5.2 Measurement Accuracy Measurement accuracy may be estimated using the results of QC repeat measurements and comparing the results to the original measurements. For laboratory analysis, the results of the split samples are compared to one another. The accuracy estimates based on two or more surveyors (or laboratories) refer to the agreement expected when different surveyors or laboratories perform the same measurement using the same method.
Acceptance criteria for measurement accuracy are established during the survey design process or in approved SNEC procedures. Where the acceptance criteria are not met, a documented investigation of the data collection and/or sample analysis process is initiated to assess and identify the extent of error or uncertainty and to determine corrective actions.
5.5 SURVEY DATA COLLECTION Survey data are collected after the survey unit has been isolated and controlled to ensure that radioactive contamination has not been introduced from ongoing decommissioning activities in adjacent or nearby areas.
5.5.1 Survey Performance Survey data are collected from the post remediation survey, the final survey, and any investigation surveys performed. The final survey uses both random and biased data collection methods and is performed using the methodology, techniques, and quality control requirements prescribed in this plan. The post remediation and investigation surveys are biased surveys performed using similar methodology, techniques, and quality control requirements as the final survey. A post remediation survey, when performed, precedes the final survey. Investigation surveys may be performed at any time.
5.5.1.1 Post Remediation Survey A special post remediation survey is performed where extensive dismantlement activities have occurred.
These surveys are based on the professional judgment of the staff, and are performed when operational radiation surveys do not provide sufficient confidence that the' survey unit is ready for the FSS. These special post remediation surveys are designed to verify that residual radioactivity levels are acceptable and that no additional remediation is necessary.
It is conducted using similar methodology, techniques, and quality control procedures, as those required for the final survey.
Professional judgment is used to: 1) identify measurement' locations most likely to have elevated levels of residual radioactivity, and 2) establish the scanning coverage and the number of static measurements to be taken.
5.5.1.2 Final Survey The objective of the final survey is to collect data of sufficient type, quantity, and quality, supporting conclusions'regarding the radiological condition of the site. The final survey uses both random and biased data collection methods. Scan measurements are taken from both biased and prescribed measurement locations.
Static measurement results used in the statistical tests are obtained from randomly derived measurement locations. The FSS of each survey unit is performed to demonstrate that residual radioactivity in each survey unit satisfies the predetermined criteria for release for unrestricted use 5-40
I 5,
-fKIVISIUN 1
5.5.1.3 Investigation Survey An investigation survey is performed.when one -or more survey measurements -exceed an
'investigatiofr{le-vel as described in Section 5.4.4., The purpose of the investigation survey is to
,define the area and level of the elevated residual -radioactivity., The'data collected during
-investigation surveys are used to characterize the area being investigated and to provide the basis for any further actions.
5.5.2 instrumentation Commercially, available portable and laboratory instruments and detectors are used to perform
-three types-of measurements: 1) surface scanning, 2) direct surface -contamination measidremients, and 3) laboratory analysis of soil and bulk materials. Other instrurientation is also used to perform other types of measurernents' as dictated by data collectibn requirements.,
The issuance, control,, calibration and operation of-ýsurvey instrumentation is' controlled by approved procedures.
5.5.2.1, Instrument Selection Radiation detection and measurement instrumentation is selected based on reliable operation, detection sensitivity, operating characteristics, and expected pe'rformancedin the field. Typically, instruments used for static measurements are capable of detecting the radiation of concern at an MDC that is less than 50-percent of the DCGLw,
Generally,-instruments used for sdarn mriea~uiFern-ts "are capable of detecting the radiation of concern to a MDC less than the DCGLw. Typical instrumentation that may be used is identified' in Table 5-9. The detectors used for direct surface contamination measurements are usually operated with data logging survey meters.
t r
541
, SNF:C FACIILITY I l;lM:I=*: T1IDRJVIMATI*%K Of A^K
Table 5-9 Typical Survey Instrumentation Characteristics Measurement Effective Detector Area and Instrument and Detector Type Wetndow Densit Type Window Density Model Alpha Scan Gas-flow 126Ludlum 43-68 proportional 0.8 mg/cm2 Aluminized Ludlum 2350-1 (or equivalent)
Mylar Beta-Gamma Gas-flow 126 cm 2 Ludlum 43-68 Scan proportional 0.8 mg/cm2 Aluminized Ludlum 2350-1 (or equi4ent)
Mylar Gamma Scan Nal Scintillator 1" D x 1" L, also Ludlum 2350-1 Ludlum 44-2 or 44-10 2" D x 2" L (or equivalent)
Dose Equivalent Plastic Bicron Micro Instrument Scintillator 11 D x 11 Length Rem Meter N/A Static Surface Gas-flow 126 cm2 Contaminatio pro oasflow 0 8 mg/cm2 Aluminized Ludlum 2350-1 Ludlum 43-68 Contamination proportional Mylar (or equivalent)
Soil and Bulk High-punty Material Germanium 1.60" x 1 94", 2.16" x 2.32" Ortec/Canbenra N/A Exposure Rate Pressurized Ion 8 Liter Sphere Reuter-Stokes NIA Instrument Chamber (PIC)
RSS-131 (or eq) 5.5.2.2 Calibration and Maintenance Instruments and detectors are calibrated for the radiation types and energies of interest.
Anticipated radionuclide mixture ratios and varying energies are accounted for during calibration by using a calibration source with a conservative and/or representative average energy as compared to the weighted average energy of the anticipated nuclide mixture. For calibration of beta detectors, detector efficiencies are determined with calibration sources consisting of Tc-99, which emits an average energy of approximately 85 keV. This average energy is conservative when compared to the expected weighted average beta emission energy of 157 keV (the predominant detectable beta emitter is Cs-137). Similarly, the use of Th-230 or Pu-239 calibration sources having weighted average alpha energies of 4.654 MeV and 5.128 MeV, respectively, is representative as compared to the alpha emission energies (approximately 5.1 MeV to 6.1 MeV) of the expected surface nuclide mix for alpha emitters (Pu-238, Pu-239 & Am 241).
Instrument calibration and maintenance are performed in accordance with approved procedures.
If vendor services are used, these services are conducted in accordance with approved procedures and a vendor QA program that is subject to approval in accordance with the Quality Assurance Program For Radiological Instruments and the SNEC Decommissioning Quality Assurance Plan (DQAP).
Radioactive sources used for calibration purposes are traceable to the National Institute of Standards and Technology (NIST) or equivalent standards.
Calibration source efficiency variations resulting from source back scatter is usually small compared with efficiency loss due to uneven, wetted, corroded and/or semi-dusty surfaces typically encountered during the final survey process. Additionally, the distance between the source and the detector is a primary contributor to detection efficiency variations. To account for 5-42 I-r'l i*l iI p i/,*
i I.dl
SNEC FACILITY LICENSE TERMINATION PLAN survey surface to detector variations in the field such as surveys over scabbled concrete, the average variation in distance from the surveyed surface to the detector is determined based on the unevenness of the survey surface. A correction or distance factor is then applied to the calibration factor of the instrument. The distance factor'wkill 'be dete~rmined -emlpirically at the SNEC Facility using an appropriate large area (150-cm 2) calibration 'source and an approved procedure.
Efficiency correction factors for wetted, corroded or semi-dusty surfaces will be determined in'accordance with NUREG-1507 (Reference 5-18) guidance. If these survey area characteristics are shown to have an impact on detection efficiency, appropriate compensatory corrections will be made in the survey design, as well as the resulting calculated survey results.
5.5.2.3 Response Checks Instrument response checks are conducted to assure consistency in instrument response, to verify the detector is operating properly, and to"demonstrate Ithat measurement res ults are not the result of detector contamination or other disturbances. At a minimum, instrument response is checked before instrument use each' da,. -'Poitable instruments' are also checked after instrument use each day. A check source is used that emits the same type of radiation (i.e.,
alpha, beta, and/or gamma) as the radiation being measured and thatgiv'es a similar instrument response. However, the check source does not necessarily use the same radionuclide as the radionuclide being measured.
The response'check is performed U-sing' a specified source detector alignment that can be easily repeated. If the instrument fails its response check, it is not used until the problem is resolved." Measurements made betweeh the-last acceptable response check and the failed check are evaluated and either'retained or discarded, as appropriate.
"5.5.2.4'A: Minimum Detectable Concentration (MDC)
The MDC is determined for the instruments and techniques that are used for survey, data collection. ',The-MDC is the concentration th'at a specific iristrrmierit"and technique can be expected to dete*t 95 per-cent of the time 'under actual field 'conditions.'
5.5.2.4.1 Beta-Gamma Scan MDC For Structural Surfaces Scanning methodology implemented at the SNEC Facility will follow an approved training process. -,A check source assembly will be used to train persbnrnel in let'ctinb area' exhibiting' two :to three,times the background count rate-.' The' sotrce is-adjusted t6 a height below the source assembly surface such that two to three times the b'ackground co'unt-rate is ýIesen'ted to the detector while siting directly over the source. "cPersnrhiel will then be asked to riioýve Lthe detector'across the source'exposure area at a'fixdd ratelof speed, ývhile-listening'for elevated' counting rates.,.Scan speeds are adjusted abcording to"the ability f thesurveyor'to disce'rn'a specified counting rate abovethe existing'survey area-backgrounnd lev-el.
Thfeiecheck source assembly is also used to train personnel in proper scanrsUrey techniques.'
The scan MDC, or MDCSCa, for scanning structural surfaces for beta and gamma emitters is determined from NUREG-1727,- Equation E-2:
S-.
- A4X:
4 (,.,c r1l270,000 138B,
~Fp6gs At 5-43 REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 Where:
MDCSCAN
=
minimum detectable concentration for scanning building surfaces (pCi/m 2) 270,000
=
conversion factor to convert to pCi/m 2 (other conversion factors may be used as well to convert to other applicable units) 1.38
=
index of sensitivity d' B
=
number of background counts in time interval t p
=
surveyor efficiency, (assumed to be 0.5, Reference 5-4)
E
=
instrument efficiency for emitted radiation (cpm/dpm) es
=
source efficiency for emissions/disintegration A
=
probe's sensitive area (cm 2) t
=
time interval of the observation while the probe passes over the source In seconds NOTE:
The correction factor (A) as defined above, is not used in the "Compass" computer program developed for purposes of survey planning by the Oak Ridge Institute for Science and Education (ORISE).
The value of p represents a mean value for normal field conditions and is discussed in Section 6.6 of NUREG-1507, "Minimum Detectable Concentrations With Typical Radiation Survey Instruments for Various Contaminants and Field Conditions" (Reference 5-18).
5.5.2.4.2 Alpha Scan MDC for Structural Surfaces For most conditions found at the SNEC Facility, the presence of alpha emitting radionuclides may be predicted by use of a surrogate beta-gamma emitting radionuclide such as Cs-137.
Scanning for the surrogate radionuclide is significantly easier than scanning for alpha activity.
During the scanning operation, when levels of the surrogate radionuclide are found to exceed a predetermined level, both a static alpha and beta-gamma measurement may be employed to verify the elevated scan reading.- Relative ratios of radionuclides will-be verified using sample and analysis techniques. This method of indirectly surveying for alpha surface activity provides additional assurance that these radionuclides are at acceptable levels on structural surfaces, and is superior to scanning only for the alpha radionuclide by itself.
When only alpha scans are employed, scanning for alpha emitters will differ significantly from scanning for beta and gamma emitters, in that the expected background response of most alpha detectors is close to zero.
Since the time the probe is in the area of elevated residual radioactivity varies and the background count rate may be less than 1 cpm, it is not practical to determine a fixed MDC for scanning. Instead, another approach described in Section 6.7.2.2 of NUREG-1 575 (Reference 5-5) is used. Given the DCGLw and a known scan rate, the probability of detecting an area of elevated residual radioactivity is calculated. The probability of detecting given levels of alpha surface contamination can be calculated by using Poisson summation 5-44 I-
SNEC FACILITY LICENSE TERMI...ATI..
P1 AM I
I v.....
I statistics (see Section 6.7.2.2 of NUREG-1575 (Reference 5-5) for a more complete description of thiS me6thod):
'For alpha survey instrumentation with backgrounds<* 3 cpm, a2 single cou'nt provides a surveyor
ýsufficient, cause'to Stop and investigate further. When'one or more c6dunts are registered, the surveyor pauses scanning*6perations and waits for a predetermined-timet6' determine if the counts are from elevatel residual radioactivity. The time interval of the "paue corresponds to a 90 percent probability of detecting counts associated with elevatedresidual radioactivity. This time interval may b6e calculated in accordance with Equation 6-13 of NUREG-1575 (Reference 5-5).
5.5.2.4.3 Gamma Scan MDC for Land Areas TheaMDCSCN values for the'Sodium Iodide detectors and radionuliides,(show*
in Table 6.7 of NUREG-1 575 (Reference 5-5)), are examples of typical MDCSCAN values that can be calculated assuming specific background levels are present in the 'survey, area. The *method given in NUREG-1507 (Reference 5-18), provides a more detailed example of how the 'Scan MDC for gamma emitters can be determined. This is the method that will be used by.the SNEC Facility when this survey approach is used. Site specific'MDCs for all -survey instrumentation will be derived incd incorp'0orated into survey packages.
=
5.5.2.4.4 Static MDC for Structural Surfaces For static'measbrements of'surfaces* the MDCstabc may be calculated usinrg' NUREG-1727, Equation E-3 (Reference 5-4). More specific values for the calibration constant K shotvnin that equation are shown below in numbers 1 through 3:
- 1.
The area of the detector (A)
- 2.
The source efficiency factor (6s), and
- 3.
The instrument efficiency for the emitted radiation(s) (6i)
MDCm
=
3 +4.65-B Where:
MDCstat,c = minimum detectable concentration for static counting (dpm/100 cm2)
B
=
background counts during measurement time interval t (counts) t measurement counting time interval (minutes)
- i
=
instrument efficiency for emitted radiation (counts/emission) s=
source efficiency for emitted radiation (emissions/disintegration)
A
=
area of detector (cm 2) 5-45 r=J i*,l / i 4* i.,*, IL i.4
SNEC FACILITY [ICENSE TERMINATIflN PH Am DC11lIOENIkg.
- 4.
The total efficiency (FT) is the product of the instrument (Ei) and source (6s) efficiencies. These values will be determined during the calibration process for the specific radionuclide mix expected in each survey area/unit (as appropriate). Actual instrument efficiencies are continuously monitored by site personnel. Any information or calculations used to establish instrument efficiencies for final status survey work will be available at the site for NRC on-site inspection purposes.
- 5.
Other correction factors may be applied to the above equation as deemed appropriate.
5.5.2.5 Detection Sensitivity The detection sensitivity of typical detectors for surface contamination measurements is estimated and the results summarized in Table 5-10. The results are shown for the principal instruments that are 'expected to be used for alpha and beta-gamma direct surface contamination measurements.
Count times are selected to ensure that the measurements are sufficiently sensitive with respect to the DCGLw.
For example,' the count times associated with measurements for surface contamination and gamma spectral analysis (soil and bulk materials) are normally set to ensure an MDCstatc is equal to or less than 50 percent of the DCGL. The scan rate associated with surface scans is normally set to ensure an MDCSCAN of no more than 75 percent of the DCGL. If the MDCSCAN exceeds the DCGL, additional static measurements may be required, as discussed in Appendix 5-1.
5-46 Dr"ILII* I/*kl 4
,SNEC FACILITY LICENSE TERMINATION Pl AN
!Table 5-10 Typical Detectio'n Sensitivities BKGND
- instrum ment C.ount MDC Instrument and Radiatio Count BKGND Efficiency" Time dpmS nMDC Detector n
Time (cpm)
(cpm/dpm)
(min) cm2 dpm/lO0 cm2 (min)
Ludlum Model
- - 2350-1, Alpha 5
2 0.155 1
49
- 500c 43-68 Prbbe
-Ludlum Model Beta-.
2350-1,. _
eGamma 5
-243, 0.275
-1
'220 511 43-68 Probe 2"x2" Nal Gamma 1
10k-900 cpmfuR/h 1 sec
-6 4 pCi/g 20k. ]
(weighted)
(scan)
Bicron Micro-Gamma N/A prem/h Read Out in Rem Meter r (varies) irem/h N/A N/A N/A, Reuter-Stokesd Pressuriz'd Ion Gamma N/A pPRh Read Out in N/A N/A NIA Chamber (varies)
PR/h N
NOTES:
a Actual calibration sources may be Cs-1 37, Tc-99, Am-241 or Pu-239 The efficiency is determined by counting the source with the detector in a fixed position from the source (reproducible geometry) bMDCs.n is calculated by assuming a scan rate of 5 cin/sec (unless otherwise marked), which is equivalent to a count time of 0.03 mi, assuming an 8 9 cm detector width c The alpha scan MDC is determined by the approach described in Section 6 7.2.2 of NUREG-1 575 (Reference 5-5) It assumes > 1 cpm is necessary for the surveyor to pause dThe pressurized Ion Chamber jPlC) is used for comparison only -No release survey data are collected using this instrument "The total efficiency (E?) is the product of the instrument (Q aind source (rzs) efficiencies These values will be determined during the calibration-process for the specific radionuclide mix expected in" each "stirvey irealunrt (as appropriate) Actual instrument efficiencies are continuously monitored by site personnel Any information or calculations used to establish instrument efficiencies for final status survey work will be available at the site for NRC on-site inspection purposes 5.5.3 Survey Methods Survey measurements are performed in accordanCe 'with sl`e6ific instructiorns corntain 6d in approved site procedures. Measurements include surface'scans, static surfacecbntamination*
measurements,- and laboratory analysis of soil and bulk 'mat~rials.*- Other easuremenrts' such as removable surface contamination and exposure rate measur-merets may also be obtained'as required.
Y 5.5.3.1 Scan Measurements Scanning is performed to locate small areasof residual radioactivity above the inVeStigation level. If an area of elevated residual radioactivity is'ide'ntified during the'scanr of a survey unit the area is marked for investigation.
4 Structure and site system surfaces are scanned for beta-gamma emitting radionuclides. Beta scintillation or thin window gas-flow proportional detectofs'ýre typically used. 'In g'neral',' the detector is held less than 2 cm from the -surface and move~d at'aboutl -detector widthi pW second. The scan rate is adjusted such that residual radioactivity canrbe 'detected at or below the irnvestigation level.
I 4
I 5 D I:::1.11
- f t"l kl 4
SNEC FACILITY LICENSE TERMINATION PLAN Scanning for alpha emitters and low-energy beta emitters (<100 keV) are limited to structural surfaces. Where scanning is performed, alpha scintillation or thin window gas-flow proportional detectors are typically used. The detector is kept close to the surface, usually less than 1 cm, and moved at a rate such that there is a high probability of detecting elevated residual radioactivity.
Land areas are scanned for gamma-emitting radionuclides.
Sodium iodide scintillation detectors are normally used because of their high detection efficiency.
The detector is held close to the ground surface (usually less than 6 cm), and moved in a serpentine (S-shaped) pattern. This is done while walking at a speed that allows the surveyor to detect the residual radioactivity at or below the investigation level. A scan rate of approximately 0.5 meters/second "is typically used to start. This rate is adjusted depending on the required detector response time and the general background level encountered in the field.
5.5.3.2 Static Surface Contamination Measurements Static measurements are taken to detect surface contamination.
Static measurements are generally performed by placing the detector on, or near the surface measured. Measurements are made over a discrete measurement time interval. Results are recorded. A one minute integrated count is a practical time interval for most field survey instrumentation and provides detection sensitivities that are usually below the DCGLw. Other time intervals may be used as warranted.
Static measurements are taken with 100 cm 2 detectors or are corrected to reflect a 100 cm 2 area. In the event that contamination is more than what would be acceptable for an area of 100 cm2, an evaluation is performed to ascertain compliance with the DCGLw.
Static measurements are typically restricted to relatively smooth impermeable surfaces where the radioactivity is present as surface contamination.
Because the detector is used in close proximity to the potentially contaminated surface, contamination of the detector or damage to the detector caused by irregular surfaces is considered before performing direct measurements.
5.5.3.3 Soil And Bulk Material Samples Soil and bulk material samples are routinely collected and analyzed. Soil samples are generally collected down to a depth of 15 cm at static measurement locations.
Sampling at greater depths is done in areas where site characterization or other information indicates the potential exists for contamination below 15 cm. Sample preparation may include, removing extraneous material, homogenizing, splitting, drying, and compositing sample materials for counting. For QC repeat measurements, the samples are obtained from the selected measurement locations as indicated in the survey/sample design package.
Samples of paint chips, tank sediment, sewage sludge, roofing material, concrete, pavement, and other bulk materials are collected for laboratory analysis as part of the biased sampling effort. Such samples may be collected in drain receptacles, sumps, and other locations in impacted areas. Selected storm drains are sampled in accessible locations. These samples are quantitatively analyzed by gamma spectroscopy for principal gamma-emitting radionuclides and the results compared to the DCGLw. If residual radioactivity can be measured at DCGLw levels by in-situ gamma-ray spectroscopy techniques, in-situ techniques may be used in place of, or supplement, sampling and analysis methods. For gamma emitting radionuclides, the above data may also be supplemented by dose rate measurements.
.[NA.ION.PLA P1=11MInmIN
5.5.3.4 ISpecial Measu rernents The historical site assessment and site characterization surveys are used to-indicate if residual radioactivity is'pres6ert in'areas or locations 'described below. Any of thlsee,ty'pes of.'areas may
,require'the application of special measurement techniques.
5.5.3.4A1 Cracks, Cre-vices, And Small Holes Surface contamination on non-planar or irregular structure surfaces, such as cracks, crevices, and small holes, may be difficultto" measure directly Lising standard -field 'suirey detebtor's ajnd established techniques-. Where no remediation h'as occurred, and residual,radioactivity is'not expected ab6ve background levels,' cracks,' 'crevices, and, small holes'are' assumed' to have the same level of residual radioactivity as thait foLund on adjacent surfaces. The accessible 'sirfacei are surveyed the 'same as other structural surfaces and no-special measurement methods are applied.
Where remediation hasoccurred on surrounding surfaces, or where residual radioactivity above background levels is expected, a representative' sarmple of surface contamination -within the crack,'crevice, or small hole is obtained. 'The level of r"esidual'radioactivity is measured and detection sensitivities can be adjusted such that reasonable"'app'rqoximations may" be midezusing indirect measurement techniques. For the most'lart,'the areas irequiring special attention With respect to this type of survey'and remediati6n'work are located in the Saxton Steam Generating Station area and its' associaited Discharge -and*
Intake Tunnels. At the SNEC 'Fa6ility, contaminated regions of this type are remediated when possible. Additional discussion 'on this subject is presented in Appendix E, Section 10 of NUREG7 1727, NMSS Decommissioning Standard Review Plan (Reference 5-4).
r, I
5.5.314.2 Paint Covered Surfaces J,
Surfaces painted-to-fix loose contamination are remediat d before' FSS'°activiti'es begin. For other surfaces painted after site start-up, representative samples in areas where it is suspected that -elevated levels' of residual radioactivity'coiild 'have "been 'covered 6ver,"'are 'taken and analyzed. Detection-sensitivities are adjusted 'or remediation is performed as dictated'by the sample analysis results.
5.5.3.4.3 Facility Systems, Floor Drains, Embedded Piping,Uriistrut' "
Surface contamination on internal surfaces of site'systemn's,' floor 'drains, embedded piping andd embedded unistrut sections may be inaccessible or difficult tormneasLir'e:directly'using typical field survey instrumentation. For surface mounted supports, fixtures, unistrut and other similar, hardware, no, special measurement methods are' generally neededl where' no remediation has' occurred in the area where these'items re'side. When-thi&isi 'the 'cas,, internal surfaces of embedments are assumed to have the same level 'of residual "radioactivity" as that found on, accessible surfaces.' Areas where no-remediatir6"'hasb'ee-n'erided ieterally exlibit at frouendr background levels of contamination.
- ed g ei a or
,e, On the other hand,,contaminated or potentially,'contamrinated'haFrware'ite'mrs iiiay be evaluated' in several ways., Where remediation has occurred," or where' ýsidual radioactivity above background levels is expected, representative samples of these embedments may beobtained
<>.--./
using a random sampling approach. The ievels-bf residual ia'dioa*tivitV are'then measured and evaluated on the removed embedment sections, and the results are "extrapolated to the 5-49 SNEC FACILITY LICENSE TERMINATION PLAN I* *:VI.* ll* N 4
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 remaining population. As an example, several unistrut sections of known length are first removed from an area, and then are assayed in a low background area to determine the residual activity per unit length (or unit surface area). From these assay results, a mean activity per unit length (or unit' surface area) is then estimated as well as a postulated standard deviation and variance for the remaining population. The required number of additional samples are then calculated for the remaining lengths of unistrut. After unit conversion, these assay results may be compared directly to the applicable DCGL values using the appropriate statistical test criteria.
For other types of facility hardware such as embedded piping, scale and sediment, samples may be obtained (as allowed by system size and accessibility), at locations along a length of pipe system. The sample' results are then assayed individually and a mean and variance are established for the locations sampled. This information is then used along with the fraction of the system sampled to estirmate, how many additional samples are needed to bound the total activity of the piping system. In addition, calibrated detectors may be used to measure the internal activity of piping and ductwork. Appropriately sized sodium iodide detection systems attached to multi-channel analyzers may be inserted into piping systems to perform in-situ analysis of hard gamma emitters (gamma-ray emitters will be used as the surrogate radionuclide in this case). Other types of detection systems that can measure internal surface beta and/or alpha emissions directly (in properly prepared piping sections), may also be used when appropriate. These types of measurements are considered special measurements and as such will be performed in accordance with a fully approved site procedure. Documentation of this type of survey process will be performed using the SNEC Calculations procedure or other approved site procedure.
GPU Nuclear, Inc. has' performed many internal video inspections of piping at the SNEC Facility to assist in determining the viability of internal measurements, and also to map out or diagram the extent of some system piping.
This process was necessary to determine the interconnections and level of sediment in pipe sections. GPU Nuclear, Inc. will continue to use internal video inspections as necessary to bound system piping on an as needed basis.
As a final resolution, when no reasonable method exists, for evaluating the residual activity remaining in/on the embedment, the embedment is conservatively assumed to be above limits and is remediated. The embedment evaluation, measurement, and assessment program will be performed concurrently with other on-going remediation efforts at the SNEC Facility in accordance with approved site procedures.
5.5.3.4.4 Volumetrically Contaminated Structural and Re-Fill Materials (Soils & Crushed Construction Debris)
It is assumed that remediation efforts will be successful in removing residual activity to meet the established limits for structural surfaces. However, continued evaluations and sampling will be used (as necessary)' to determine the depth of residual activity remaining in concrete surfaces before final status survey work begins. Additionally, indirect measurement techniques such as evaluating exposure rates, gross gamma count rates and/or in-situ gamma-ray spectrometry measurements may be performed as needed to derive residual contamination levels on or within structural materials. Detection sensitivities will be established such that reasonable but conservative approximations of the quantity of remaining volumetric activity can be made.
To aid in simplifying dose modeling requirements for this kind of material, samples of site soils and construction debris have been sent to an off-site laboratory where Kd (distribution 5-50
Scoefficient) values have been developed for relevant'site radionuclides. These Kd values have then been used to de-velop final site DCGLw Values for all volumetric material types at the SNEC site. By selecting the most conservative DCGLw developed from these vari6us material types, a universally applicable DCGLw may then be used for all SNEC Facility volumetric materials. As a result of this modeling and pathways analysis technique, SNEC site DCGLw values may be used for both surface and subsurface soil and construction debris (re-fill'or otherwise). Any residual activity allowed to'remain in SNEC site structures or in -soil materials will meet the site dose criteria for unrestricted release based on these DCGLw values.
A sampling and measurement program will-be implemented to monitor-and control residual contamination levels in re-fill materials. The sampling program will be statistically based and be applied through the implementation of fully reviewed SNEC site procedures and/or work instructions. Sampling and analysis will meet requirements stated in Section 5.2.7.6 of this plan.
5.5.3.4.5 Paved Parking Lots, Roads, Sidewalks, And Other Paved Areas Paved parking lots, roadways, concrete slabs,-and other paved areas are treated -as structure surfaces. Scan and static measurements are taken as prescribed by the survey design Where remediation has occurred or where residual radioactivity above background levels is suspected, direct surface contamination,;-measurements-are taken and a representative -number of subsurface samples (below 15 cm) will be collected and analyzed. Depending on the size of the paved area and the distribution of the residual radioactivity, the paved area may be'a separate survey unit or be included as part of a larger survey unit. Sampling of these areas is also appropriate where there is reason to believe that contamination resides in, on, or below these structures.
5.5.3.4.6 Trailers And Temporary Facilities Trailers or other temporary facilities used to support FSS 6rdecommissi6oring w6rk are not included in this study, but instead will be released in accordance with current SNEC Facility Radiological Controls work practices and procedures. Any temporary facilities remaining at the time of FSS activities shall be classified and surveyed in accordance with the applicable area or use classification.
5.5.3.4.7 Subsurface Soil Contamination Survey The subsurface sampling/measurement program will be controlled by site procedures and will follow a systematic process for collecting subsurface information. In this methodology, each zone (surface, subsurface and buffer zone below the potentially contaminated region) will represent a sample population. The buffer layer will extend below the depth of any formerly buried components and the suspected depth of the contamination zone. The buffer layer depth and starting point will also be adjusted as indicated by sampling. The number of cores to be taken within each zone is the number N required for the applicable statistical test applied. The core samples will be homogenized over each 1 meter of depth during the sample preparation process. The appropriate test (WRS or Sign) will be applied to the results, as applicable. If the test indicates that the layer being assessed fails, the layer or the volume will be considered for remediation. Additionally, in-situ measurements may be considered when any layer exhibits results approaching 50% of the release criteria.
Areas where subsurface contamination may be present at the SNEC site are identified and sampled through the following process:
5-51 SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 Characterization and Historical Site Assessment (HSA) information were reviewed and used to determine the appropriate area classification. The area classification chosen considers both surface and subsurface volumes below structures as well as any previous remediation or survey efforts.
A review of any existing measurement and/or sample results in the subsurface volume is then performed to determine if sufficient sampling results are available for planning a FSS.
"* These areas are then made accessible; i.e. obstacles to sampling and survey work are removed (where possible), including any structural impediments.
"* Where sampling below structures is prohibitively difficult or expensive; sampling through floor/slab structures or road coverings may be the appropriate choice rather than removing the entire structure to access the subsurface volume.
The final state subsurface regions are identified including the depth and thickness of the buffer zone.
"* Each subsurface layer is sampled and surveyed lAW a survey and sampling plan.
When any sample or survey result suggests or necessitates remediation of a volume, the remediation is performed before a final round FSS design is planned.
Identified locations where subsurface sampling/measurements will be planned include:
- 1.
The Spray Pond area (-5500 square meters)
- 2.
The 1.148 acre SNEC Facility site. To date, a significant portion of this area has been remediated.
- 3.
Any suspect subsurface areas identified by site management that have shown contamination levels approaching the DCGLw.
5-52 9-
Ul Ar*.. II
/-
tVl.Ii I i--I'*MI,.ATI.,
D.lIlll'alvl AMr**
DE*...i.Cl lM 4 5.5.3.4.8 CV Steel Shell Activation Area Survey The activated section of the CV. steel-liner is currently assumed to be a region of the CV-shell that extends from about the 790'- El,(operational water line inthe reactor-cavity) up-tothe proposed cut off region at about the 805' El.(-15 feet).- Additionally, the region.is assumed to extend for.a full quadrant of the CV or about 39'.of the circumference of this building (centered horizontally at the former location of the reactor).
When the interior.surface of the CV shell is thoroughly decontaminated,'from residual surface contamination, samples of the steel shell will becollected within the activation zone previously described. Theanalysis of these samples.will provide the best average concentration, for the
'st6el shell in the activation region. Additionally,r a gamma measurement of the-shell in-this region may be used to augment the sampling efforts. These types of gamma, measurements are special measurements and are described in more detail in Section 5.5.3.4.9. The direct and indirect dose contribution will be added to the.-dose contribution, from. -residual surface contamination. The sum from these two sources will be maintained below 25 mrem/y TEDE.
5.5.3.4.9 CV Steel Support Ring Surveys During 2002, SNECwas tasked with surveying 'and releasing several steel surface areas of the SNEC Containment Vessel (CV) steel shell in support of installation of steel i-beams, which were designed to stabilize the shell during removal of concrete. Survey areas were-first aggressively cleaned using methods such as surface grinding which removed surface oxides, paint and any residual concrete that had adhered tothe SNEC CV steel surface, as well as a thickness of the steel itself. This cleaning process removed contaminants to essentially the base metal, thus ensuring that the vast majority of surface contamination -had beenremoved before the surveys -began. Pre and post cleaning surveys were performed to verify that the cleaning effort was successful.
The survey was designed using NRC screening DCGLs for surface contamination as described in Table 5-1. A conservative scanning speed was set to locate elevated areas within the survey units 'which when detected, were re-measured for, a full one minute of count time. Elevated measurement locations were re-cleaned ard -re-surveyed as necessary. Randomly-located static measurement points were also counted for one minute.
These areas have been surveyed "at risk" in that they have been surveyed before NRC approval of the SNEC License Termination Plan (LTP). Conservative survey, planning -and remediation efforts have been used to ensure that all ring installation areas were decontaminated thoroughly below potential site release limits. In addition, radiological controls remained in place throughout the survey process to prevent survey area re-contamination.
This survey information will be included in the Final Status Survey Report.
5.5.3.5 Investigation Measurements Removable activity, dose r'ate, and in-situgamma spectrometry measurements may beused as diagnostic tools to further characterize the radiological conditions in selected areas, and to evaluate potential response actions. Sodium iodide detectors can also be used, both for hard to reach' areas e.g., embedmrents, piping and duct work,:as wellas for subsurface monitoring efforts su-ch as gamma-logging.
Sodium iodide detectors become especially useful when employed in conjunction with multi-channel analyzers that are capable of discerning between 5-53 I* r"*ll* I/*lkl 4
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 natural occurring and site-specific radionuclides.
Gamma-logging using a multi-channel analyzer is useful in both screening surveys (to determine depth and average concentration of contamination) and in final status surveys (to provide an upper limit of the average radionuclide concentration). If no significant counts are obtained in the detection system's region of interest (ROI), within a bore hole or piping system, then a "less than" value, or minimum detectable concentration (MDC), can be quoted for the soil around the bore hole or for a measured section of system piping at a given confidence level (95%). By ensuring that the MDC is less than the release criteria, the surveyor can designate the soil in the vicinity of the detector (or section of pipe) to be below the release criteria.
Additionally, this type of measurement system is sensitive to elevated materials in adjacent buried piping or elevated po*kets of contamination outside of the immediate sampling zone.
Therefore, GPU Nuclear, Inc. will consider using gamma-logging in conjunction with sampling in areas where volumetrically contaminated materials approach the release criteria or when contamination is thought to be present in piping systems within a survey area.
5.5.3.6 Hard-To-Detect (HTD) Radionuclides Many radionuclides are comparatively simple to detect in the field at environmental levels using routine gamma-ray spectroscopy analysis techniques. In contrast, the "Hard-To-Detect" (HTD) radionuclides are not easily identified using any routinely applied field measurement practices.
SNEC has identified H-3, C-14 and Ni-63 as being the only HTD type nuclides of significance at the SNEC Facility. A summary of the radionuclide selection process can be found in Section 6.2.2.3.
5.5.4 Sample Handling and Analysis When sample custody is transferred (e.g., when samples are sent off-site to another lab for analysis), a chain-of-custody record accompanies the sample for tracking purposes.
The sample chain of custody record documents the custody of samples from the point of measurement or collection until final results are obtained. These tracking records are controlled and maintained in accordance with approved site procedures. On-site laboratory capabilities are used to perform gamma spectroscopy of bulk sample materials,' gross beta-gamma and alpha counting of smears and Tritium analysis in liquid samples. Off-site laboratory services are procured as needed for Sr-90, TRU and other Hard-To-Detect (HTD) radionuclides. Laboratory analytical methods are generally capable of measuring levels at 10 to 50 percent (or less) of applicable DCGLw values.
5.5.5 Data Management Final survey data may be collected from post remediation surveys, final surveys, investigation surveys or special measurement evaluations such as those made to determine embedment or sub-surface activity levels.
5.5.5.1 Other Scan Measurements When 100% of any area is scanned at a high detection efficiency, capable of discerning low levels of residual activity (well below established DCGLw levels), collected results have a greater assurance that survey areas meet the site release criteria.
Therefore, the need to measure a finite number of randomly selected survey points are reduced or eliminated.
Consequently, some scan survey measurement efforts performed for initial phase and/or 5-54 9-
investigative purposes, may be accepted as final survey data provided the following conditions are met, "1. The MDA for the scan is a small fraction of the required DCGLw for thesurvey area, and there, is sufficient sensitivity present in the survey design-at-anacceptable confidence level..*
- 2.
All applicable survey data collection requirements as prescribed in Section 5.5 and 5.6.1 are followed.
- 3.
The area was isolated after the survey activity.
5.5.5.2, Other Static Measurements Other static measurements performed during post remediation and investigation surveys are based on professional judgment. Since they are biasedand;not random, they may not be used in the statistical tests. -However, this does not necessarily preclude their acceptance as final survey data. These measurements may be accepted as final survey data provided:
_1.
Allapplicable survey datacollection requirements as prescribed in Section 5.5 and 5.6.1 are followed.
- 2.
Thirty or more data points are collected within the survey unit. For piping and other embedments, accessibility to interior -surfaces may -not allow this number of measurements. In these cases, similar survey methodology encompassing historical assessment, characterization, remediation, and post remediation survey data will be
-,,-used as-a basis for biased measurements and sampling, to ensure that the release criteria are met.
- 3.
None of the data points exceeds the DCGLw.
- 4.
The area was isolated after the survey activity.
5.5.5.3 Data Recording Survey measurements -will be recorded in units appropriate for comparison to the DCGLw by correcting for material, specific background, efficiency,-- geometry, detector area, and measurement size as applicable.
The recording _units ;are dpm/100. cm2 -for surface contamination and pCi/g for volumetric radionuclide concentrations.
Records of survey data are maintained in accordance with approved site procedures. Survey data records-include the iderntification of the surveyor, type of measurement, location, instrumentation used, results,-time and date measurement was performed and the instrument calibr'ation information.
5-55 SNEC FACILITY LICENSE TEiRMINATION PLAN REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 5.6 SURVEY DATA ASSESSMENT The data assessment process checklist is illustrated in Figure 5-2. Final survey data, described in Section 5.5, are reviewed to verify they are of adequate quantity and quality.
Graphical representations and statistical comparisons of the data can be made which may provide both quantitative and qualitative information about the data. An assessment is performed to verify the data. If the quantity or quality of the data is called into question, previous survey steps are re-evaluated. The statistical tests are applied and conclusions are drawn from the data as to whether the survey unit meets the site release criteria.
5.6.1 Data Verification and Validation The final survey data will be reviewed to verify they are authentic, appropriately documented, and technically defensible. The review criteria for data acceptability are:
- 1.
The instruments used to collect the data are capable of detecting the radiation of interest at or below the investigation level. If not, acceptable compensatory measures have been taken.
- 2.
The calibration of the instruments used to collect the data is current and radioactive sources used for calibration are traceable to recognized standards or calibration organizations.
- 3.
Instrument response is checked before and, where required, after instrument use each day data are collected.
- 4.
Survey team personnel are properly trained in the applicable survey techniques, and this training is adequately documented.
- 5.
The MDCs and the assumptions used to develop them are appropriate for the instruments and the survey methods used to collect the data.
- 6.
The survey methods used to collect the data are appropriate for the media and types of radiation being measured
- 7.
Special measurement methods used to collect data are applied as warranted by survey conditions, and are properly documented in accordance with an approved site procedure or Station Work Instruction.
- 8.
The custody of samples that are to be sent for off-site laboratory analysis, are tracked from the point of collection until the final results have been obtained, and
- 9.
The final survey data set consists of qualified measurement results representative of current facility status are collected as prescribed by the survey design package.
If a discrepancy exists where one or more criteria are not met, the discrepancy will be reviewed and corrective actions taken (as appropriate) in accordance with site procedures.
5-56 9-
5.6.2 Graphical Data Review Survey data may be graphed or plotted to identify patterns and relationships in the data that might g6&'_hnoticed'6sing purely' numerical methods.- When-neede-d, a posting plot 'and/or a
,frequency plot may' be used.
Other graphical data' represe6tati6n tools can also be used as appropriate.
5.6.2.1 Posting Plot Posting plots, generated during investigation surveys, may be used to identify spatialpatterns in the data.- A 'posting plot is simply a map of the -survey unit with -the.data.values' entered at the measurement locations. The posting plot can reveal non-homogeneous spatial characteristics in the survey unit such as patches of elevated residual radioactivity-r'" 'drioupings" of measurements that exceed the DCGLw. Even in a background reference area, a posting plot may reveal spatial tiends in backgrou n-d data that might'affect theresult* of the statistical tests.
In somecases, the trends could be du6 to residual radioactivity, but may also irndicate nonr homogeneous cha_ýalcteristics in the background 'eferen'ce 'are'a..
5.6.2.2 Frequency'Plot A frequency plot'is used to examine the general shape'of the data distribution. A frequency plot
'is'a bar chart of the number of data points Within a certainrange"of values. 'The frequency plot may reveal any obvious departures from-symmietry,' 'sich as skewness'*or bi-mbdality (two peaks), in th6,data"distributions for'the survey unit or background reference area.- Whben the data distribution is highly skewed, it is often because there are a few elevated areas of residual radioactivity. The presence of two peaks in the data may indicate the existencb of isolated areas of residual radioactivity or a mixture of background concentration distributions due to differe nt'soil types, construction -materials, etc'- The gr:eater variability/ in-the data-due to'the "presence of such a mixtui-e will reduce the power of the statistical'tests to detect an inadecliiately remediated survey unit. 'These situations'maj indicate the need to mi-ore carefully match-background reference areas to the survey-unit, or to divideT tlie'siivey unit into-survey units with more homogeneous backgrounds.
5.6.3 Basic Statistical Comparisons Statistical quantities (range, median, mean, and standard deviation) are calculated for the final
-survey data set Whgr'one' or more data points exceed thi DCGL'
'.The'calculated quantities are compared to the Values shown in-T6ble'5-'1 1. The statistical conparison values represenrt assumptions underlying the statistical test to be used. Where the statistical quantity fails the comparison, the data set and/or survey design as'suriipfiinsare e'x'amined.
I 5-57 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
Table 5-11 Basic Statistical Comparisons Statistical Quantity Acceptable Value Failure Response Range (R)
R _ 3 a Examine data for outliers Median (u )
I(u - k ) / al*- 0.5 Examine data for outliers and anomalies Mean (X)
X _ DCGLw Apply reference area, remediate or reclassify Sigma (a) a < 0 3 x DCGLw Review initial survey design parameters 5.6.3.1 Outliers Where the range is greater than 3 standard deviations, the-data are examined for outliers, Outliers are measurements that are extremely large or small relative to the rest of the data set and, therefore, are suspected of misrepresenting the population from which they were collected.
Outliers may result from measurement collection and recording errors.
Outliers may also represent true extreme values of a distribution, such as areas of elevated residual radioactivity, and indicate more variability in the population than was expected. Not removing true outliers and removing false outliers both lead to a distortion of estimates of population parameters.
Tests developed to detect outliers in a data set may be used to identify data points that require further examination.
A test alone cannot determine whether statistical outliers should be discarded or corrected. This decision is generally based on professional judgment.
5.6.4 Statistical Testing The Sign or the Wilcoxon Rank Sum (WRS) statistical test, also known as the Mann-Whitney test, may be applied to the final survey data set where one or more measurements exceed the DCGLw. The statistical test is based on the hypothesis that the level of residual radioactivity in the survey unit exceeds the,,DCGLw.
There must be sufficient survey data with levels of residual radioactivity at or below the DCGLw to reject this statistical hypothesis and to conclude the survey unit meets the site release criteria.
5.6.4.1 Application of Statistical Tests The statistical test does not need to be performed when the survey data clearly show that the survey unit meets the site release criteria. The survey unit clearly meets the criterion if:
- 1.
Every measurement in the survey unit is less than or equal to the DCGLw, or
- 2.
A background reference area is used and the difference between the maximum survey unit measurement and the minimum background reference area measurement is less than or equal to the DCGL.
In these instances, the statistical test is not applied.
The statistical test is applied where one or more measurements exceed the DCGLw. Similarly, for a survey unit where a background reference area is used, the statistical test is applied where the difference between any survey unit measurement and any background reference area measurement is greater than the DCGL. Survey results and the corresponding conclusions for when a background reference area is and is not used are shown in Tables 5-12 and 5-13, respectively.
5-58 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
Table 5-12 Initial SurveyResults and Conclusions When A Backgi-und Refere'nce Area Is Not Used
'I
==
Conclusion:==
Survey' unit mneets Survey Result (Class 1'Areas)'
- site release dose criteria Yes No All measurements less than or equal to DCGLw" Mean greater than DCGLw Any measurement greater than DCGL with mean less than or equal to DCGL which passes Sign test Any measurement greater than DCGL with mean less than or equal to DCGL which fails Sign test Table 5-13 Initial Survey Results and Conclusions When a Background Reference Area is Used;
==
Conclusion:==
Survey unit meets "Survey Result (Class I Areas) site release dose criteria "Yes.
N Difference between maximum survey unit measureme nt and minimum background referencearea measurement'less'than or equal to DCGLw Difference between survey unit mean and background reference area mean greater than DCGLw
,Difference between any survey unit measurement and any background reference area measurement greater than DCGLw, "ahd difference between survey unit mean,and-background reference area mean less than-or equal to,.DC-GLw which I
passes WRS test e
tD w
Difference between any survey unit measurement and any background reference area measurement greater than DCGL and difference between survey unit mean and background reference area mean less than or equ6l to DCGL~~wmhiclh fails WRS test
-'4 5:59 -1
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 5.6.4.2 Sign Test The one-sample Sign statistical test is used if the radionuclide of concern is not present in background and radionuclide-specific measurements'are made. The Sign test may also be used if one or more radionuclides are present in background at such small fractions of the DCGLw as to be considered' insignificant.
In this case, background concentrations of the radionuclides are included with the residual radioactivity (in other words, the entire amount is attributed to facility operations). Thus, the total concentration of the radionuclides is compared to the site release criteria. This option is only used if it is expected that ignoring the background concentration does not affect the outcome of the statistical test. The advantage of ignoring a small background concentration is that no background reference area is necessary.
The Sign test is applied as follows:
- 1.
List the survey unit measurements, X, i = 1, 2, 3,...,n; where: n = the number of measurements.
- 2.
Subtract X, from the DCGLw to obtain the difference (DCGLw - X,, i=1, 2, 3...n).
- 3.
Discard differences where the value is exactly zero and reduce n by the number of zero measurements.
- 4.
Count the number of positive differences. The result is the test statistic S+. Note that a positive difference corresponds to a measurement below the DCGLw and contributes evidence that the survey unit meets the site release criteria.
Compare the value of S+ to the critical values in Appendix I, Section 1.3 of NUREG-1575 (Reference 5-5). The Table columns equate to the false positive decision error rate, a. The value of a is the probability of passing a survey unit which actually fails to meet the site release criteria, which is obtained from the survey design (the initial value is 0.05 - see Appendix 5-2).
If S+ is greater than the critical value for the false positive decision error rate given in the Table, the survey unit meets the site release criteria. If S+ is less than the critical value, the survey unit fails to meet the criterion.
5.6.4.3 Wilcoxon Rank Sum (WRS) Test The two-sample WRS statistical test is used when the radionuclide of concern appears in background, or when a measurement method is used that is not radionuclide-specific. Because gross activity measurements are not radionuclide-specific, they must be performed for both the survey unit being evaluated by the WRS test and for the corresponding background reference area.
The WRS test is applied as follows:
- 1.
Adjust the background reference area measurements by adding the DCGLw to each background reference area measurement, Xi (X, + DCGLw).
- 2.
Sum the number of adjusted background reference area measurements, m, and the number of survey unit measurements, n, to obtain N (N = m + n).
- 3.
Pool and rank the measurements in order of increasing size from 1 to N.
If several 5-60
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I measurements have the same value, they are all assigned the average rank of that group of measurements.
- 4.
Sum the ranks of the adjusted background reference area'mniasurer-ient's to obtain Wr.
- 5.
Calculate the critical value using equation 1.1, NUREG-1575 1(Reference 5-5). This equation is used when there are several measuremrient's that haveý the-eame value.
Critical Value m((mnn +mr+1))/2) + (z nm(n+m4-1)112)
Where:
z = The (1 -a) percentile of a standard normal distribution, which can be found in the Table 5-14 below.
Table 5-14 Values For a and z a
z 0.001 3.090 0.005 2.575 0.01' 2.326 0.025
- 1.960 0.05 1.645 0.1 1.282 NOTE: The value of a is obtained from the survey design (initial value is 0 05 - see Appendix 5-2).,NRC approval is Z
req*ired to increase the a (type 1 decision error) >0 05 In accordance with License Condition 2 E.(h) Where m and n
""are less than 20, the critical value is given in Table 14 of NUREG-1575 (Reference 5-5)
- 6.
Compare the value of Wr with the critical value calculated'aboveý- If Wr is greater than the critical value, the survey unit meets the site release criteria. If Wr is less than the critical value, the survey unit fails to meet the criterion.'
5.6.5 Data Conclusions The results of the statistical test allow one of two conclusions to be drawn.' The first conclusion is the survey unit meets the site release criteria. The data have provided statistically significant evidence that the level of residual radioactivity in the survey-Ldnit'does' not 'exceed the site release criteria. The decision that the survey unit is acceptable for unrestricted release can be made with sufficient confidence and without further analysis."'
The second conclusion that is that the survey unit fails tormeet the 'Site -eleasbecriteria. -The data does not provide sufficient statistically significant evidence that the level of residual radioactivity in the survey unit does not exceed the site releas& criteria: ;Th6 data is analyzed further to determine why the statistical test result led to this conclusion.
5-61
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I Possible reasons the survey unit fails to meet the site release criteria are:
- 1.
It is in fact true,
- 2.
It is a random statistical fluctuation, or
- 3.
The test did not have sufficient power to detect that it is not true. The power of the test is primarily based on the actual number of measurements obtained and their standard deviation. A retrospective power analysis for the test may be performed as described in Appendices 1.9 and 1.10 of NUREG-1575 (Reference 5-5). If the power of the test is insufficient due to the number of measurements, additional data may be collected.
If it appears that the failure may be due to statistical fluctuations, the survey unit may be resurveyed and another set of discrete measurements collected for statistical analysis. A larger number of measurements increases the probability of passing if the survey unit actually meets the site release criteria. If it appears that the failure was caused by the presence of residual radioactivity in excess of the site release criteria, the survey unit is remediated and resurveyed.
5.7 SURVEY RESULTS Survey results are documented in history files, survey unit release records, and are summarized in the final survey report.
Other detailed and summary data reports may be generated as requested by the NRC or SNEC Management.
5.7.1 Survey Unit Release Record The survey unit release record is the complete release record in a standardized format prepared for each survey unit or group of survey units with similar histories. The survey unit release record is a collection of information necessary to demonstrate compliance with the site release criteria. This record includes:
- 1.
A history file checklist:
The history file checklist references relevant operational, and decommissioning data.
The purpose of this checklist is to provide a basis for the survey unit classification.
The history file will reference relevant sections of the Historical Site Assessment (Reference 5-19) and other compiled records including:
History of remediation The survey unit operating history affecting radiological status Scoping, site characterization and post remediation survey data Other relevant information.
- 2.
Description of the survey unit
- 3.
Survey design information for the survey unit
- 4.
Survey unit ALARA analysis, if performed 5-62
ISNEC FACILITY LICENSE TERMINATION PLAN
- 5.
Survey measurement locations and corresponding survey data
- 6.
Survey unit investigations performed with documented results, as applicable
- 7.
Any survey unit data assessment results
- 8.
Results of any special measurements performed for the survey unit 5.7.2 Final Survey Report A final survey report will be prepared and submitted 'to the.NRC. Thereport will provide a summary.,of. any ALARA analysis, survey data results, and overall,conclusions, which demonstrate that the SNEC Facility and site meet the radiological criteria.for unrestricted use.
Information such as the number and type of measurements, basic statistical quantities, and statistical test results will be included in the report.
The following outline illustrates a general format that may be used for the final status survey report. The outline below may be adjusted to provide a clearer presentation of the information.
The level of detail will be sufficient to clearly describe the final status survey program and certify the results.
Information to be submitted (Reference 5-4, Section 14.5):
- 1.
A summary of the results of the final status survey.
- 2.
A discussion of any changes that were made in the final status survey from what was proposed in the LTP or other prior submittals.
3.1, A description of the method by which the number of samples were determined for each survey unit (see Reference 5-5, Section 5.5.2).
- 4.
A summary of the values used to determine the numbers of samples and a justification for these values (see Reference 5-5, Section 5.5.2).
- 5.
Survey results for each survey unit including:
Number of samples taken for the survey unit.
A map or drawing of the survey unit showing the reference system and random start systematic sample locations for Class 1 and 2 survey units, and random locations shown for Class 5 survey units and reference areas.
Measured sample concentrations.
Statistical evaluation of the measured concentrations (see Reference 5-5, Section 8.3, 8.4 and 8.5).
Judgmental and miscellaneous sample data sets reported separately from those samples collected for performing the statistical evaluation.
5-63
SNEC FACILITY LICENSE TERMINATION PLAN V...
.M Discussion of anomalous data including any areas of elevated direct radiation detected during scanning that exceeded the investigation level or measurement locations in excess of the DCGLw.
A statement that a given survey unit satisfied the DCGLw and the elevated measurement comparison if any sample points exceeded the DCGLw.
- 6.
A description of any changes in initial survey unit assumptions relative to the extent of residual radioactivity.
- 7.
When a survey unit failed, a description of the investigation conducted to ascertain the reason for the failure and a discussion of the impact that the failure has on the conclusion that the facility was ready for final radiological surveys.
- 8.
If a survey unit failed, a description of the impact that the reason for the failure has on other survey unit information.
5.7.3 Other Reports If requested by the NRC, computer-generated and/or summary data reports will be provided in hard copy or electronic form.
Survey data include date, instrument, location, type of measurement, and mode of instrument operation.
Other data, such as conversion factors, background reference areas, and the MDCs used, are available which will allow independent verification of the results. Measurement results will also be presented graphically. The FSS report will be independently reviewed.
Any independent verification survey performed will be performed by an organization outside the SNEC Facility staff and management organization.
Reports generated as a result of any independent verification survey process initiated by the SNEC Facility, will be available upon request.
5-64 I*PVI*InN 4
5.8-DEFINITIONS
- 1.
Accessible Surface Area -An area available to a radiation detector for direct
-scanning or fixed-point measurements.
2.'
Area Factor (AEMc) - A factor, used to adjust the.DCGLw to estimate DCGLEMC and the minimum detectable -concentration for scanning surveys in Class 1 survey, units, (DCGLEMC
-=DCGLw x AEMC.
The area factor (AEMc) is the magnitude by which the residual radioactivity in a small area of elevated activity can exceed the DCGLw, while maintaining compliance with the release criterion. SNEC Facility area factors are listed in Table 5-15 of Appendix 5-1.
- 3.
Background Radiation. - Naturally_ occurring radiation which may include cosmic, terrestrial (radiation from the naturally radioactive elements) and man made radiation from global fallout.
- 4.
Characterization Survey - A radiological survey and -its supporting evaluations performed to establish the SNEC Facility radiological condition for planning decommissioning activities.
- 5.
Confidence Level: The probability associated with a confidence interval which expresses the probability that the confidence interval% contains the population parameter value being estimated.
- 6.
Derived Concentration Guideline Level (DCGL) - Residual radioactivity levels that equate to the site release criteria - for that, particular pathway or measurement. The two (2),basic DCGLs defined in this plan are 1) the DCGLw and, 2) the DCGLEMC. The DCGLw is the average concentration limit for the standard size survey-area. The DCGLEMC is the elevated measurement area DCGL, which is-used for small areas of elevated activity'(above the DCGLw).
When not defined, DCGL refers to the DCGLw.- Other.DCGLs discussed in this plan (e.g, DCGLGA etc.) are derived from these two basicdefinitions and are sometimes referred to as an "effective DCGL".
.7.
Elevated Area - Areas of residual contamination exceeding the'guideline value.
- 8.
Final Status Survey (FSS) - Radiological -measurements, evaluations and supporting activities undertaken to demonstrate that the SNEC Facility satisfies the criteria for unrestricted use,.
- 9.
Hard-to-Detect Nuclide (HTD) - A radionuclide,emitting-radiation(s) that are difficult to detect with field or laboratory based instrumentation.
10.,
History,File - A compilation of -information used to justify the classification and
";.,survey design for the suryey unit..-It should reference sections of the Historical Site Assessment, characterization survey, data, remediation surveys and other information used to establish the basis for the design of the final status survey.
11..
Independent Verification Survey-An information only radiological survey, performed by an organization independent of the-SNEC Facility staff and 5-65 SNEC FACILITY LICENSE TERMINATION PLAN RFVIRIt3N 4
management, which will provide SNEC Facility management with an additional level of confidence concerning the validity of the Final Survey results.
- 12.
Minimum Detectable Activity (MDA) - The minimum level of radiation or radioactivity that can be measured by a specific instrument and technique. The MDA is usually established on the basis of assuring false positive and false negative rates of less than 5%.
- 13.
Minimum Detectable Concentration (MDC)
The minimum activity concentration on a surface or material volume that can be statistically detected above background. This is usually set at the 95 % confidence level.
- 14.
Multiple Source Terms - Generic term used when more then one source term element is encountered (e.g., a remaining site structure with surface contamination and embedments).
- 15.
Operational Survey - A radiological survey performed in accordance with SNEC procedures in support of routine site operations.
- 16.
Quality Control Survey - A survey that consists of repeat measurements on a specified fraction of the survey areas. The survey areas are usually selected at random to provide an additional check of final status survey measurements.
- 17.
Release Criteria - A term used to identify the radiological requirements for release of the SNEC Facility for unrestricted use.
- 18.
Remediation Survey - Any survey performed that is used to determine the effectiveness of'remediation activities. The final post remediation survey is a special remediation effectiveness survey performed with instrumentation similar to the type used for the FSS. The survey methodology is also similar to actual FSS methodology.
- 19.
Scan Survey - A qualitative radiological monitoring technique that is performed by moving a detector over a surface at a specified speed and distance to detect elevated activity areas or locations. Also called a "Surface Scan".
- 20.
Scoping Surveys - Surveys such as investigative surveys used to provide a quick look at conditions before or during FSS work. These surveys are not necessarily documented.
- 21.
Structures - All SNEC Facility site buildings and their surfaces.
In addition, platforms, restraints and supports, and external surfaces of piping systems, heating and ventilation systems, tanks, stacks, etc., are also treated as structures in the final status survey if they exist beyond remediation efforts.
- 22.
Surface Contamination - The total of both fixed and removable contamination.
For the purposes of this plan, this would also include any remaining neutron activated material near the surface. Also called total surface contamination.
- 23.
Survey Area - The basic survey entity for the management of the Final Status Survey. It is comprised of one or more survey units, the bounds of which are 5-66 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
S Y
LN E
AREVISION I
defined by existing facility physical features, such as -a room, intersection of walls, column-and-row layout of a floor elevation, or structural I-beams.
- 24.
Survey Location - In a structural or open land survey area, a-survey location is usually represented by a single grid block. In a system survey area, a specified length of piping or, a component su-ch as b valve,or tank'is 'referred to as a survey location. A survey location can contain one or more survey points. Also referred to as measurement locations.'
- 25.
Survey Unit Release Record - A collection of information in a standardized format for controlling and documentinb fi~ld measurements taken for the Final Status Survey. A survey unit release record is prepared for each survey area.
The survey unit release record may include the survey instructions, a control form, grid map(s), survey measuremnent data sheets' and'survey mnps. It may also be called a survey package.
- 26.
Survey -Point - A smaller 'subdivision within a 'stirvey-location (grid block, system, component) where local measurements 'are takeni-For structures and systems, a survey point generally refers to an area covered by~a detector, or an area of 100 cm2 when a smear is taken. For open land areas, a survey point refers to the area covered by a detector (for paved surfaces), the point at which a dose'rate measurement is taken: 6r the point'atiwhich'a'soil or'pavement sample is collected.
- 27.
Survey Unit - A geographical area consisting of structures -or land' areas of specified size and shape at a remediated site for-which'a sep*rate decision will be made whether the unit attains the site-specific reference-based cleanup standard for the designated pollution parameter.-Survey, units aire generally formed by grouping ýcontiguoi's sit& areas with a'sirii!lr use fhistory and the same classification of contamination potential.) Suirvey units are established to facilitate the survey process and the statistical analysis of~survey data.
-28.-
Total Effective Dose Equivalent (TEDE) -The sum of the deep dose equivalent (for, external exposures) and -the "committed ffeective 'dose-equivalent (for internal exposures).
- 29. - Unity Rule -Where more thano6ne radionuclide is 'present,'the'sum of the ratios of each'radionuclide concentration to 'itsires~e~6tive DCGL-'should not exceed unity. When this method is used, the effective DCGL is equal to one (1).
5-67 SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LIC:FN'JI TI:RMIMATItAM DI ALd r.,,
I V I S I O N 1
5.9 REFERENCES
5-1 Code of Federal Regulations, Title 10, Part 50.82, "Termination of License" 5-2 Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors," January 1999 5-3 Code of Federal Regulations, Title 10, Part 20.1402, "Radiological Criteria for Unrestricted Use" 5-4 NUREG-1727, "NMSS Decommissioning Standard Review Plan", September 2000.
5-5 NUREG-1575/EPA 402-R-97-016, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)," August 2000 5-6 NUREG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys" 5-7 SNEC Facility Site Characterization Report, May, 1996 5-8 NUREG/CR-5512, "Residual Radioactive Contamination From Decommissioning, Final Report," Volume 1, October 1992 5-9 Draft NUREG-1549, "Using Decision Methods for Dose Assessment to Comply With Radiological Criteria for License Termination," July 1998 5-10 Yu, C. F. et al., Manual for Implementing Residual Radioactivity Materials Guidelines Using RESRAD, Environmental Assessment Division, Argonne National Laboratory 5-11 Yu, C. F. et al., RESRAD-Build, A Computer Model for Analyzing the Radiological Doses Resulting from the Remediation and Occupancy of Buildings Contaminated with Radioactive Material. Environmental Assessment Division, Argonne National Laboratory 5-12 Regulatory Guide 4.15, "Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment" 5-13 SNEC Procedure, 1000-PLN-3000.05, "SNEC Facility Decommissioning Quality Assurance Plan" 5-14 SNEC Procedure, E900-PLN-4542.01, "SNEC Radiation Protection Plan" 5-15 SNEC Procedure, E900-ADM-4500.44, "SNEC Facility Calculations" 5-16 SNEC Procedure, E900-ADM-4500.04, "SNEC Records Retention Procedure" 5-17 SNEC Procedure, E900-ADM-4500.12, "Radiological Surveys: Requirements &
Documentation Procedure" 5-68
5-18 NUREG-1507, "Minimum Detectable Concentrations With Typical Radiation Survey Instruments for Various Contaminants and Field Conditions," June 1998 5-19
- '5-20 SNEC Facility Historical Site Assessment Report, January 2000
,Letter Report, "Use of Two-Stage or Double Saripling in Fin'al Statis Decommissioning Surveys", Prepared by C.-Gogolak for NRC, Febrbary, 2000.
5-69 SNEC FACILITY LICENSE TERMINATION PLAN
-REVISION 1
9___________
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I APPENDIX 5-1 ELEVATED MEASUREMENT COMPARISON (EMC)
The EMC, sometimes called a "hot spot test," is a simple comparison of measured values against a limit. There are two applications of this comparison in the final survey process. It is used when the sensitivity of the scanning technique is not sufficient to detect levels of residual radioactivity below the DCGL (i.e., where the MDCscan is greater than the DCGL).
In this application, the number of static measurements may need to be adjusted.
Appendix 5-2 describes how this is done. The second application in this appendix, is when one or more scan or static measurement data points exceed the DCGL. The use of the EMC for measurements above the DCGL provides assurance that unusually large measurements receive the proper attention and that any area having the potential for significant dose contributions is identified.
The EMC is intended to flag potential failures in the remediation process.
Locations, identified by scan or static measurements, with levels of residual radioactivity, which exceed the DCGL, are investigated (see Section 5.4.4).
The size of the area where the elevated residual radioactivity exceeds the DCGL and the level of the residual radioactivity within the area are determined. The average level of residual radioactivity is then compared to the DCGLEMC. If a background reference area is to be applied to the survey unit, the mean of the background reference area measurements may be added to the DCGL or the DCGLEMc to which the average level of residual radioactivity is compared.
The DCGLEMC is calculated using the following equation (NUREG-1575, Equation 8-1):
DCGLEMc = Area Factorx DCGL The area factor is the multiple of the DCGL that is permitted in the area of elevated residual radioactivity without requiring remediation. The area factor is related to the size of the area over which the elevated residual radioactivity is distributed. That area, denoted AEMC, is generally bordered by levels of residual radioactivity below the DCGL, and is determined by the investigation. The area factor is the ratio of dose per unit area or volume for the default surface area for the applicable dose modeling scenario to that generated using the area of elevated residual radioactivity, AEMc. It is calculated based on the methodology given in chapter 8 of NUREG-1505 (Reference 5-6).
If the average level of the elevated residual radioactivity is less than the DCGLEMc, there is reasonable assurance the site release criteria is still satisfied and the area does not require remediation. Radioactivity at the DCGLEMc distributed over the area AEMC delivers the same calculated dose as does residual radioactivity at the DCGL distributed over the default surface area. If the DCGLEMc is exceeded, the area is remediated and resurveyed. Area factors for open land areas at the SNEC Facility are provided in Table 5-15. Area factors for surface area DCGLs supplied by the NRC are provided in Table 5-15A 5-70
SiLC. ACILITY LICENSE TERMINATION PLAN C
REVISIQ. i Table 5-15 Area Factors (AF) For Open Land Areas Based on 25 mremly TEDE and Upper 1 Meter Volumetric Surface Modeling File Names =
NEW XXXXX.RAD*
NEW XXXXXA.RAD NEW XXXXXB.RAD NEW XXXXXC.RAD NEW XXXXXD.RAD NEW XXXXXE.RAD AREA=
10000 m, 2500 m2 400 m2 100 m2
-.-25m 2 -
I m2 Radionuclides Base DCGL AF Implied DCGL AF Implied DCGL AF Implied DCGL AF Implied DCGL AF Implied DCGL AF
_____EMVC EMC EMOI EMVC 1
1ýEMVC Am-241 25.7 1.0 47.7 1.9 110.1 4.3 321.7 12.5 699.1 27.2 3005 116.9 C-14 26.8 1.0 151.1 5.6 984.8 36.7 2 69E+03 100.2 7206 268.9, 1.79E+05 6682.8 Co-60 3.5 1.0 4.4 1.3 4.9 1.4 5.4 1.6 7.0 2.0 43.4 12.4 Cs-137 6.6 1.0 14.9 2.3 19.9 3.0 238 3.6 31.1 4.7, 189.3
'28.7 Eu-152 10.1 1.0 10.5 1.0 11.1 1.1 12.1 1.2 15.5 1.5-,
94.3
-9.3 H-3 645 1.0 1.47E+03 2.3 3.23E+03 5.0 7 87E+03 12.2 1.78E+04 27.6 3.55E+05 '
550.2 NI-63 747 1.0 3.66E+03 4.9 1.29E+04 17.2 5.14E+04 68.8 2.05E+05 275 5.07E+06 6789.8 Pu-238 30.1 1.0 57.7 1.9 142.9 4.7 408.2 13.6:
694.4 23.1,,
'1.08E+04" -
358.8 Pu-239 6.8 1.0 11.9 1.7 269 4.0 56.4 8.3 114.8 16.9 1374 202.1 Pu-241 866 1.0 1607 1.9 3713 4.3 1.09E+04 12.6 2.39E+04, 27.6 1.02E+05 118.1*
Sr-90 1.2 1.0 36 3.0 9.8 8.1 38.5
-.32.1
-.146.7.
122.3. -
2826 2355 Where "XXXXX" is the radlonuclide computer file name, as an example "Am241".
NOTE 1: Base case DCGLs (in pCl/g) are for 10,000 square meter surface model only.
NOTE 2: The above set of DCGL values are used only to determine the Area Factors (AF) that will then be applied to the values listed in Table 5-1 (surface materials only).
NOTE 3: When AF values are calculated In the RESRAD computer code, the settings for contaminated fractions for plant food, meat and milk must be re-set to their default condition (41) in order to allow the computer code to scale the food supply for the size of the areas appropriately.
5-71
SEFCLTLC STE IAI L
RVISIN 1 Table 5-15A Area Factors For Structural Surfaces (Based on NRC Screening Values - see Table 5-1)
Nuclide 36 m2 25 m 2 16 m 2 9 m 2 4 m 2 I m 2 Am-241 1
1.5 2.3 4.1 9.2 36.2 C-14 1
1.4 2.2 4.0 8.9 35.9 Co-60 1
1.2 1.5 2.0 3.4 10.1 Cs-137 1
1.2 1.5 2.2 3.7 11.2 Eu-152 1
1.2 1.5 2.1 3.5 10.7 H-3 1
1.4 2.2 4.0 8.9 35.8 Ni-63 1
1.4 2.2 4.0 9.0 35.3 Pu-238 1
1.4 2.3 4.0 9.1 36.9 Pu-239 1
1.4 2.2 4.0 9.0 35.4 Pu-241 1
1.4 2.2 4.0 9.0 34.8 Sr-90 1
1.4 2.2 1.5 8.8 34.7 NOTE: DCGL is in dpm/100 cm 2
<-I K>
5-72 SNEC FACILITY LICENSE TERMINATION PLAN RFVIRION t
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 "APPENDIX 5-2 STATISTICAL INFORMATION The method described in NUREG-1575 (Reference 5-5) and NUREG-1505 (Reference 5-6) for determining the number of survey measurements necessary to assure a population set sufficient forstatistical analysis is summarized here in the'manner it is applied at the SNEC
,Facility.-% An effective survey design slightly -overestimates both the number -of survey measurements and the standard deviation to ensure adequate power of the statistical test. -This ensures that a survey unit is not subjected to additional remediation simply because the final survey is not sensitive enough to detect that residual radioactivity is below the DCGL.
TERMS AND STATISTICAL PARAMETERS A minimum number of measurements are needed to obtain sufficient statistical confidence that the conclusions drawn from the survey data are correct.
Several terms 'and statistical parameters, are described in this chapter that, are used, in determining the number-of measurements needed to apply the statisticaltests..
Lov~er Bound f Gray Region (LBGR)
The LBGR is the concentration to which the'survey unit rnust' be remediated in order to have an acceptable' probability'of pa6sirig 'the statistical test'for' meeting the site'Felebse citria. 'It represents the lower bound ofthe area of uncertainty regarding.the concentration of-residual radioactivity Iin the'su rvey unit. 'The' DCGL represents the upper bound. The widthof the gray 4
region is equal to the difference between the DCGL and the LBGR, as illustrated in'the Figure 5 3 below.
Gray Region Diagram ei v '
ried Concentration Guidline Level (DCGL):.
92 Lower Bound of Gray Region (LBGR) 0 Initially, the LBGR is arbitrarily set at 0.5 times the DCGL. The LBGR is adjusted as needed to result in a relative shift of 1 to_3.(see NUREG-1575 and-NUREG-1505(documents 'for full description of terms (Reference 5-5 and 5-6)).*
1 The survey design goal is to achieve an LBGR between,0.5.and 1 times-the DCGL. Since the LBGR serves as the remediation goal for'de'c6rmmissi6oi*ng activities, generally the smaller the LBGR, the more rigorous are the requirements for dismantlement, decontamination, and post 5-73
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I remediation surveys.
As the LBGR approaches the value of the DCGL and the band of uncertainty narrows, the number of samples needed to demonstrate compliance with the site release criterion rises dramatically.
Standard Deviation (a)
The standard deviation, denoted by a, represents the spatial variability in the concentration of the residual radioactivity in the survey unit. The mean and standard deviation are calculated using standard equations.
An estimated value, of a may be calculated either from existing data or by taking limited preliminary measurements of the concentration of the residual radioactivity in the survey unit (sometimes called a pilot or scoping study). It is also acceptable to assume that a equals 0.3 times the mean concentration. Alternatively, a reasonable estimate based on available site knowledge may be used.
The value selected as an estimate of a for the survey unit may be based on data collected only from within that survey unit or from data collected from a much large" area of the site, since there may be some difficulty in determining which individual measurements from a preliminary survey may later represent a particular survey unit. The most practical solution may be to estimate a for each area classification (Class 1, Class 2 and Class 3). If there are multiple types of materials within an area classification, additional estimates of a may be required.
A separate estimate of a may be obtained for every background reference area. If the a in the background reference area is larger than the a in the survey unit, the larger value may be used.
The survey design goal is to avoid an estimated a that is overly optimistic or conservative. If the value is grossly underestimated, the number of measurements will be too few and may result in unnecessary remediation or re-survey. If the value is grossly overestimated, the number of measurements will be unnecessarily large.
Relative Shift (Ala)
The number of measurements needed depends on a ratio involving the concentration to be measured relative to the variability in the concentration. The ratio is called the relative shift, denoted by A/a. It is defined as (NUREG-1727, Equation E-4) (Reference 5-4):
A/c = DCGL - LBGR Where:
The variables have been previously defined.
The survey design goal is to achieve A/a values between 1 and 3.
The number of measurements needed rises dramatically when A/a is smaller than one. Conversely, little is usually gained by making Ala larger than about three. If Ala is greater than three, the LBGR should be increased until the relative shift is equal to three.
5-74 K-
SNEC FACILITY LICENSE TERMINATION PLAN DECISION ERRORS The principal study question or statement is,- "are the levels of residual -radioactivity in all survey-units below applicable release criterion and can the site be released?" t-Results from, surveys and other environmental testing will be used to determine the answer tothis question.
A decision error is the probability of making an error in the decision on a survey unit, either passing a survey unit that should fail or failing a survey unit that should pass. The first decision error, passing a survey unit that should fail, is referred to as a false positive or TYPE I decision error. The probability of makihg this error is denoted by ax: Setting high value for a results in a higher risk of passing a survey unit that should fail. --Setting low value of a lowers the risk of passing a survey unit that should fail.
The second decision error, failing a survey unit that should pass, is referred to as a false negative or TYPE II decision error and is denoted by P. Sele6ting a high value for 13 "resiilts in a higher-risk of failing a survey unit that should pass and subjecting it to further investigation.
Selecting a low value for 13 lowers the risk and minimizes these investigations. The cost of setting a low value for either a or 13 is a higher valuefor.the other or an increased number of measurements to demonstrate compliance with the release criteria..
When-using the statistical testing procedures as described in NUREG-1575 and NUREG-1505 (Reference 5-5 ahd 5-6) documents i.e., the Sign' Test or the Wilcoxon Rank Sum (WRS), larger decision errors may be unavoidable when encountering difficult or adverse conditions.. This is particularly true'when trying to measure residual radioactivity concentrations close to the variability in the'concentration of those materials in natural background.,-ln'order to avoid an unreasonable number of samples when A/a is very small, larger values of cc may be crisidered as shown in Table 5-16 below.
Table 5-16 Acceptable Decision Error (a as a Function of DCGL DCGL/a a
>3 0.05 1.2 to 3 0.10 0.6 to 1.2 0.25
<0.6 0.30 Table 5-16 values are based on the assumption that the LBGR should not have to be set to less than 0.5 times the DCGL, and that if a is allowed to increase, 13 will also be allowed to increase.
There are no constraints on the value of 13. However, decreasing 13 increases the number of samples needed, making vary small value's of 13 urnattractive.
The survey design ol6jective is then to establish the'value of a equal to or less than 0.05 and to minimize the value of 13 while maintaining the minimum number of measurements at an optimal number.
NRC approval is required to increase the a (type 1 decision error) >0.05, in accordance with License Condition 2.E.(h).
,5-75 REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I NUMBER OF MEASUREMENTS The statistical parameters c, 13 and A/a are used to estimate the number of measurements that will produce the desired values of cc and 13. The number of measurements are based on the statistical test which is applied to the survey unit. The two statistical tests used in the final survey data analysis process are the Sign Test and the Wilcoxon Rank Sum (WRS) Test. The criteria for using these testing procedures are summarized in Table 5-17.
Table 5-17 Statistical Tests and Criteria For Their Use Statistical Test Criteria for Use Radionuclide of concern appears in background, or measurements are used that are not radionuclide-specific.
Radionuclide of concern is not present in background and radionuclide Sign Test specific measurements are made, or radionuclides are present in background at such small fractions of the DCGL as to be considered insignificant.
NOTE: For specific information on statistical testing procedures, see Table 2 3 of NUREG-1 505 (Reference 5-6).
The number of measurements is determined by rounding up the number calculated using the appropriate statistical' test and adding 20% more measurements. Additional measurements are added to protect against the possibility of lost or unusable data.
Wilcoxon Rank Sum (WRS) Test The two-sample WRS test is used when the radionuclide of concern appears in background or if measurements are used that are not radionuclide specific.
Because gross activity measurements are not radionuclide specific, they must be performed for both the survey unit(s) being evaluated by the WRS test and for corresponding reference area(s).
The number of measurements needed for the WRS test is determined from the following equation (NUREG 1727, Equation E-5) (Reference 5-4):
n = (1/2) (Z 1a Z 1 ') 2 (31(Pr -
0.5)2 Where:
n
=
number of measurements in survey unit Zi.
=
percentile represented by decision error a (NUREG-1 575, Table 5.2)
Zj
=
percentile represented by decision error 13 (NUREG-1575, Table 5.2)
Pr
=
probability that a random measurement from survey unit exceeds random measurement from background reference area by less than DCGL when 5-76 I-
,SNEC FACILITY LICENSE TERMINATION PLAN survey-unitmedian is equal'to 'LBGR concentration above background
-(NUREG-1575,-Table 5.1)
Y
=
factor included in Equation 5-1 of NUREG-1575 to define n as the number of measurements in a survey unit Additional n measurements are also needed in the background reference area.
Sign Test The 06ne sample'Sign test'is used if the radionuclide, of.,concern is not present'in background and radionuclide specific measurements are made. The' Sign test may als obe used if one or more radionuclides are present in background at such smallfractions of the
-DCGL as to be considered insignificant.,
In this case, backgro'u'rd concentrations. ofthe radionuclides are included With'the' residual radioactivity (in other'w6rds, the entir6 amorunt is attributed to facility operations,). IThus,-thejtotal concentration of radionuclides are compared to.the release criteria.
This option ;is only used if it is expected that ignoring the background concentration will not "affect'the outcom6 of-the statistical tests. The 'advantage of ignoring, a, small background concentration is'that'no background reference area is heeded, which 'simplifies the final survey considerably.
"C Since'SNEC, Facility radionuclides, outside the' SNEC CV are largely composed of Cs-137 -and Co-60, Which are relatively easy to' measu're,'the" Sign test "is the most likely statistical testing procedure to be employed. Inclusion of Cs-137 and Co-60 background values int6 the DCGLs should not significantly effect meeting site release criteria.
The number of measurements needed for' the 'Sign test is determiried from the 'following equation (NUREG-1727, Equation E-6) (Reference 5-4):
n z a +
-,-f
)
=4(Sin-0 5)2 Where:
Sign p, estimated probability that random measurement for a survey. unit is less
"-than the'DCGL whbn'the survey united*liari conb6ntratibh is actually'at the LBGR (Table 5.4 of NUREG-1575 (Reference 5-5)).
All other variables have been previoi:sly d fine'd.
Elevated MeasuirementCompa'rson (EMC)
The EMC is used to determine if additional measurements may be needed when thelevel of residual radioactivity that can be detected by 'scanning (MDCs seeSection 5.5.2.5) 0slarger than the DCGL. The WRS and Sign tests evaluate whether or,not theresidual radioactivity exceeds the DCGL for.contaminiation -that is approximately'unif6rim acrosithe surivey unit.'
These tests: may,not successfully' detect small areas of 'elevatbc 'conma-irn`ti n. _Instead, systematic 'measurements, in conjunction'with'surfa°e scanning, are 'sed*
to 'dbtairi.ad6quate assurance that srmiall areas of elevated radibactMity will rm6et' the, DCGL, Wh*re
'the MDCsCan 5-77' REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I exceeds the DCGL, the EMC provides the reasonable level of assurance that any small areas of elevated residual radioactivity that could be significant are not missed'during the final survey.
The number of measurements needed to detect an elevated concentration, nEMC, in a survey unit area, A, is (NUREG-1727, Equation E-8) (Reference 5-4):
Area Factor = MDC.
DCGL Once the area factor is determined, the corresponding value for AEMC can be obtained (see Appendix 5-1) and the value of nEMC calculated.
If nEMC is larger than n, additional samples (up to a total in the survey unit of as many as nEMC) may be needed to demonstrate that areas of elevated residual radioactivity meet the site release criteria.
In cases'where nEMC is larger than n, the site characterization should be considered and, based on what is known about the site, it may be possible to estimate a concentration that is unlikely to be exceeded. This maximum concentration may be converted into an area factor (multiple of the DCGL), and then the corresponding AEMC value obtained (see Appendix 5-1) and used in the above equation. Similarly, based on knowledge of the site it may be possible to estimate the smallest area likely to have elevated levels of residual radioactivity.
If this is so, that area can be used in the above equation.
Likewise, knowledge of how the residual radioactivity would be likely to spread or diffuse could determine an area to be used in this equation.
The EMC is only applied to Class 1 survey units, since areas of elevated residual radioactivity should not be present in Class 2 or Class 3 survey units.
Optimizing the Number of Measurements Once the acceptable design values have been established, survey design constraints may be changed to evaluate how these changes affect the number of measurements for several basic measurement designs. The survey design constraints are:
Limits on the decision error probabilities ca and P; Width of the band of uncertainty or gray region (by adjusting the LBGR); and Survey unit boundaries (it may be possible to reduce the number of measurements by changing or eliminating survey units that may require different decisions).
The process may be iterative in that the initial values selected must be modified to allow dependent variables to fall within the survey design constraints.
Selecting a Minimum Number of Measurements As discussed above, the MARSSIM process incorporates design constraints that ensure that'an adequate number of sample measurements are taken per survey unit. However, to simplify the final survey process and to ernsure conservatism without an associated unreasonable expenditure of resources, a minimum number of 30 sample measurements per survey unit may be collected for survey units less than 10 m2 whenever possible.
5-78
11.*(l
/A*t/
ITV I l.l.i.Q.
Tll Mil
'ATIlnI DI A/'
IRIlN 1 GRID SIZE DETERMINATION AND DESIGN,'
The sample size, determined in'previous sections, provides'the necessary input on setting sampling locations and survey patterns., The method is outlined in section 5.5.2.4,of NUREG 1575_ (Referel~ce, 5-5).
The refe'rence grid system on,which rsampling ".p'atte'rns,are superimposed, is a 'basic alpha, numeric ox-y, layout unless otherw'Aise determin6d to -be impractical because 6f actual field 6onditions. Table 5-16 presents a summarized presentati-on of typical survey'design parameters.
p
r I I' 5-79 I* I=Vl.q I*N 1 t
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 APPENDIX 5-3 BACKGROUND REFERENCE AREAS Background reference areas are used if: (1) the residual radioactivity contains a radionuclide that occurs in background, or (2) the measurements to be made are not radionuclide-specific.
Background reference areas are not needed when radionuclide-specific measurements are used to measure concentrations of a radionuclide that is not present in background. They also are not needed when one or more radionuclides are present in background at such small fractions of the DCGL as to be considered insignificant. Surveys are conducted in one or more background reference areas to determine background levels for comparison with the conditions determined in specific survey units.
SELECTION Background reference areas are selected which have similar physical, chemical, radiological, and biological characteristics as the survey unit being evaluated. They are usually selected from non-impacted areas, but are not limited to natural areas undisturbed by human activities.
Generally, background reference areas are not part of the survey unit being evaluated.
However, where necessary, they may be associated with the survey unit being evaluated, but cannot be potentially contaminated by site activities. For example, background measurements may be taken from core samples of concrete, pavement, or other types of surface materials.
Occasionally, the survey unit itself may serve as the background reference area when a surrogate radionuclide in the survey unit can be used to determine the background.
For example, it may be possible to use the measured alpha-or beta-emitting fraction of an established radionuclide distribution in embedded piping to calculate a net activity and subtract it from a gross activity to determine background levels.
For materials present on-site, either in buildings or as non-soil materials present in outdoor survey units (e.g., concrete, brick, drywall, fly ash, petroleum product wastes, etc ), background reference areas of non-impacted materials that are as similar as possible to the materials on site are used. Sometimes such materials are not available. In those situations, a good faith effort is made to find the most similar materials readily available or use appropriate published estimates.
Each background reference area should have an area at least as large as the survey unit, if practical, in order to include the full potential spatial variability in background concentrations.
MARKEDLY DIFFERENT SURVEY UNIT MATERIAL BACKGROUNDS Survey units may sometimes contain a variety of materials with markedly different backgrounds.
An example might be a room with drywall walls, concrete floor, glass windows, metal doors, wood trim, and plastic fixtures. A separate survey unit is not made for each material.
If one material is predominant or if there is not too great a variation in background among materials, a background reference area comprised of the predominant material may be appropriate.
For example, a room may be mostly concrete but with some metal beams, and the residual radioactivity may be mostly on the concrete.
In this situation where the presence of concrete predominates, the background for concrete is used for the room. If a measurement location on the random-start grid falls on the metal beam, the measurement location may be moved to be on the nearest piece of concrete. Alternately, the measurement could be taken on the metal, but this is then noted on the survey record so that an unusual reading can be explained.
K J 5-80
When there are different materials with substantially different backgrounds in a survey unit, a non-impacted room with roughly the same mix-of materials may-be used as,a background reference area. Alternately, measured backgrounds for the diff&rentmaterials or for groups-of
-similar materialsmay be used. In this case, it is acceptable to perform the'Sign test on-the difference between the paired measurements from the survey unit and from -the,appropriate background material.
,MARKEDLY DIFFERENT BACKGROUND REFERENCE AREAS
-If significant differences in backgrounds among background reference areas are-found, a value of three times the standard deviation,of the mean between the background reference'areas is added to the mean background for all background reference areas-to define a' background concentration.
The value of three times the standard deviation of the mean is chosen to minimize the likelihood that a survey unit that contains only background would fail the statistical test for release.
The-WRS test-is then used-to test whether the, survey unit meets the radiological criteria for license termination.
BACKGROUND LEVEL DETERMINATIONS In general, background levels will be established (as needed) for each type of instrument-used for total surface contamination measurements, removable contamination measurements, and gamma exposure rate measurements. - Inj addition, backgrounds will be -determined -for specialized detectors and detector systems on an-as needed basis.- These, background measurements may include large area detector backgrounds for -floor monitoring instrumentation, background determinations for detectors used for surveying piping interiors and background values for gamma-ray spectroscopysystems.
For soil and sediment samples, background levels will be determined for -man-mrade radionuclides not resulting from plant activities.
The objectives,of, background determinations,, for -the--SNEC Facility final status -survey measurements will be to:
Establish reference backgroundvalues for each type of detector system,used in the SNEC Facility FSS survey.
Assess the variability in background responses for principal detectors under different applications and conditions of use.
Determine the need for correction factors or special measurements to establish the background for final status survey measurements in specific locations.
"* Account for man-made radioactivity not resulting from plant operations.
Measure reference values at different directions and distances from the center of the SNEC Facility outward, including down wind directions.
Background determinations will be performed in accordance with approved procedures and survey request work package guidance. Background values will be evaluated in accordance with NUREG-1575 (Reference 5-5) guidance. Methods used to determine background values for each type of final status survey measurement method are summaýri.ied belo.1w.
5-81 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1 Background measurements will be performed on surfaces unaffected by licensed activities. The principal criteria for, selection, of measurement locations for building surfaces will[ be the location's similarity, to SNEC Facility construction and freedom from radioactivity attributable to SNEC Facility operations. Instrumentation will be of the same type to be used for the SNEC Facility FSS process.
Soil and sediment samples will be collected from areas unaffected by licensed operations in order to establish the background levels of man-made radionuclides not resulting from plant operations. The background samples will be collected at locations similar to their respective on site sampling locations whenever possible. Gamma-ray spectroscopy will be used to determine individual concentrations of radionuclides.
Additionally, laboratory analysis for other non gamma emitting radionuclides will be performed as necessary.
A pressurized ion chamber (PIC) will be used to establish the background gamma exposure rate for the site.
The pressurized ion chamber will be used as the reference instrument for establishing gamma-ray exposure rate background and the true pR/h equivalent response of portable survey meters.
Additionally, a Micro-Rema Meter (or equivalent) will be used to evaluate site dose rates. Measurements obtained with this instrument will be compared to PIC measurements at selected locations. The Micro-Rem Meter will then be used for the bulk of the final status survey gamma-ray exposure rate measurements at the site.
Compiled background information will be used as the reference areas for most survey units as necessary. Where actual background conditions are shown to vary significantly, area-specific background values will be 'used whenever possible.
Alternatively, in-situ gamma spectroscopy may be used to identify those components of the background gamma radioactivity that are possibly due to plant contamination.
Background values of direct beta-gamma measurement instruments are affected by conditions in the immediate vicinity of the detector. Significant variations from background reference values may be observed. These variations are caused by the natural radioactivity composition in surrounding materials and by shielding effects in some cases.
Consequently, background measurements for special conditions will be compiled for use in calculating values (net dpm/100 cm2, etc.) to reduce the bias in survey unit population statistics. Special condition backgrounds may be compiled for poured concrete, steel and iron, aluminum and any other material as deemed appropriate.
a Micro-Rem Meter is product of the Bicron Instrument Company.
5-82
7 I I I I I I I I
p rt-it-rt-iiri-i-rrsriirrirT D I CLASS -
I IMPACTED Ak6A U,
I 4
4, r'
1 r7tNý L iflttl a
2 I I I
I I I
11.11 I I I liii III J II
[liii 11 I [
I 11111 I~
~ ~~~~~1 4
tW ~
Q f
A 1/2-
_______________I H d i t I~---
I II I clS MPATE 0.E 22
-I ii I HI IiE ""
LAEI]5EHE2EE
- 11LIIIII--------
TFTTT-S E
Ar II I
I I I II I I T
I IIII
.4 17 Z4 U mký i:
A
-t A
L A
7
£
'I,
/
I I.fl-flW-4-AS1 I4...I.
~
T I'll ICr I I11 It I I I Itt111 I 111 1U 11
'F*
7771l
!Fn mug flu-
.
-. t
n.,
Wif#' 't a
~~~7 I
32I I
U H
F E
I L11i OkAI
stAlk I.-
v 1
1 5
3 1101-m-ME-ON IN
I i-I ýý I
0 j
1 6
5NmRM-fl¶3 1Tfl I
1 4
3 D
C B
A 17 18 19 20 21 22 23 24 25 25 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 40 INSIDE ELEVATION l I I
lI I-I--I-
!I
--. I_...&.
I
",z FLQOQORLAN NOLLVA313 3JISNI
- 1. SEE SNECRM-013 SH. 3 FOR ELEVATIONS DENOTED BY 0.
FIGURE 5-5 (IGRO -- I1scI. mater)
THIS IS A COMPUTE Am!
IGIENERATED DRAWIN%
O NOT REVISE IT MANUALLY R.S0/15,XSNEC FACILITY ft.
/
PENELEC WAREHOUSE FLOOR PLAN AND EXTERIOR WAML ELEVATIONS SNCIRM-013 ISCAM.L1/8-1-0 Ih I*"*
NO.1 Do.NO.
IITLE INTERFAQCOCUO43EBA NO.
A I
I I
a' I
K 5
A 2
I 2
I
(
D C
B A
5
-I.
-l-I-
-l-l-JI-l--L I
-II--
I l
I 1r 1- ' " ' "
" ' "'-4"--1 F-TTLWLLLLLL 1L'44 S------1-1.--I I
-L
'+
4 -
-I- -
I I+III-I-1 Il V
T-I:
I I
I LI:
I
-- 1:ITIF:
I I
I T-I1I-II:ITI I
2 3
4 5
6 1
8 9
1 i72 5 1
5 1 7
8&
0 1 2 324 2 8a7 2 9 3 Fw-'l.NE.*NCRM--013 I
6
-PC - --
I I
Ft"Sms IF
- I*FFw*IrJF*
II K
I I
AK
.q I
4 I
I I
F71 v N
'~MF FA~ll TY IICF5~F FRMIATIN P1AN-REVISIOfN I
6.0 COMPLIANCE WITH RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION 6.1 DISCUSSION
-The SNEC decommissioning objective is to release the site for unrestricted use perl 0 CFR 50 and Subpart E of10 CFR Part 20.1402.' This section of the SNEC LTP will briefly'discuss the methods and assumptions used to demonstrate that the total effective dose-equivalent -(TEDE) from residual radioactivity is less than 25 mrem per year (0.25'mSv) to an average member of the critical group. References 6-5, 6-6 and 6-7 have been used to determine the appropriate methodologies for meeting site release criteria.
These methodologies are also described in Chapter 5 of this plan.
6.2 DOSE MODELING The distribution and variability of radioactivity in the environment, and dose rates from natural sources, provide the framework for developing models in' determining compliance with criteria for site cleanup and restoration.
The dose limit of 25 mrem TEDE in any year is of approximately the same magnitude as the: geographic variability-- of, doses 'from,1atural background. It is comparable to the difference in annual doses likely to be experienced by a person who moves from one location to another. The TEDE, adopted in1991 by the'NRC, is the sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures). For site'cleanup'and restoration'stahdards,'the'dose' limit is applied to all site-specific sources, including natural radionuclides whose concentrations have been enh-anced by site activities.
"Dose modeling is used to estimate the TEDE to an average'member of the'critical group*from residual radioactivity. The critical group means the group of individuals reasbriably'expected to receive the greatest exposure to residual radioactivity for any applicable set of circ-um-staniesi.
The surface contamination and radionuclide concentrations levels of structures, land areas, and plant 'systems remaining 'at the time of the final-'site survey-are c6mlpared 'to derived concentration guideline levels (DCGL) calculated using applicable dose"models.-A DCGL is defined as the concentration of residual radioactivity distinguishable from background radiation which, if distributed uniformly throughout a survey unit, would result in d, TEDE of 25 mrem/yr to an average member of the critical group.' DCGL's are presenrted'ifnterms' of surface and volumetric radioactivity concentrations and are expressed in units of dpm/1 00 cm 2 or pCi/g.
Data and information 'related totypes Iand amoiunts-of radio6ti6'imagteriail-'on,the SNEC site have been documented in the Site CharacterizatioW-'RepdrtF(R6ference 6:18),ah-d Chapter 2 of this plan.
Prior to site release consideration, all radioactive waste will have been disposed, sources Will have been disposed or scheduled to be taflnsfer?6d to arnothle" licerFieea*ll 'bliildirigs rand -open land'-areas remediated, and mifibr'amounts of contaiination 'ill hav7 ben detected and documented via surveys.
Dose models'are based, 6nitwo types of conceptualizations, -i.e. two-6limen*i6rial surface area geometry and a three-dimensional volumetric geometry. In certain cases, e.g. embedded pipes, a combination of dose models will 'be employed -to calculate"th-e 'radiatio0n dosse* fromsurfk&c contamination nrd volumetric sources.' Dose model'sceriari6s ae'dis6us6ed below.
6-1
ý Qm1NI:=PCI1LIIITY LICE.FNSE TER:IMINATIO'N PL AN
. RFVIRICIN t
SNEC FACILITY LICENSE TERMINATION PLAN RFUI5lIOIJ I 6.2.1 Building.Occupancy Scenario For the two-dimensional surface area model, the building occupancy scenario is used as, the best representation. This model applies to the SNEC Containment Vessel (CV), Saxton Steam Generating Station (SSGS), tunnels and/or other impacted site structures.
Since these structures will either, be-demolished or left behind, the-building occupancy scenario is most representative.
Other buildings (garages and support, structures) left behind will also be surveyed and released using the building occupancy scenario.- DandD Version 1.0 has been used as the preferred modeling software for the building occupancy scenario.
The exposure pathways selected for analysis in the building occupancy scenario. include external exposure to penetrating radiation from surface sources, inhalation of resuspended surface contamination, and inadvertent ingestion of surface contamination.
The selection of these pathways provides a balanced analysis for:
a)
External exposure to penetrating radiation from surface sources; b)
Inhalation of resuspended surface contamination; c)
Inadvertent ingestion of surface contamination; and d)
External exposure to penetrating radiation from embedded sources.
This scenario accounts for exposure to both fixed and removable thin-layer or surface radioactivity within a structure. This scenario assumes individuals (critical group) occupy the building in a passive, mode without deliberately disturbing the residual radioactivity on building surfaces. Occupancy of the building, is assumed to begin immediately after license termination.
The exposure duration is assumed to be a full work year (approximately 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />). The selection of this scenario assumes the following site conditions will exist to qualify:
- 1. The contamination on, building surfaces (e.g. walls, floors, and ceilings) should be surficial and non-volumetric.
- 2.
Contamination on surfaces is mostly fixed (not loose),with the fraction of loose contamination not to exceed 10% of the total surface activity.
- 3.
This screening criteria will not apply to contaminated surfaces on buried components or conduit (e.g. drainage or sewer pipes). Such component and conduit surfaces will be treated on a case by case basis Surface contamination DCGLs welre developed through the use of the DandD computer code using the standard,default parameters. The results from the DandD runs for each of the 11, site related radionuclides are in units of mrem TEDE per dpm/100 cm 2. Default surface area DCGLs (Reference 6-8) are listed in Table 6-2. These default values were verified using DandD. The DandD output for the building occupancy scenario model is contained in Appendix 6-1.
Building and component surfaces will be decontaminated to as low as reasonably achievable to meet the release criteria building occupancy screening levels. For building debris that cannot be assessed using the surface screening criteria, a volumetric assessment of the materials will be made. These materials will be characterized (if not already known) as to radionuclide contaminant(s), type of material, corrosivity rates (metals), and leach rates (concrete/debris) for 6-2 RI::VI.q It3 N '!
-.input. This switch to volumetric consideration brings the resident farmer scenario back as the
-,release scenario.
Since some of the material -will.be, buried 3 feet below grade, the contamination zone may be in the saturated zone. A subsurface volumetric dose model has been developed to evaluate this condition.
Exposure pathway (d) listed above applies to areas where there is penetrating radiation from
-embedded sources, of, radioactivity, such as. embedded piping.. -To the extent practical embedded sources will be filled with grout or concrete. For modeling this scenario a bounding calculation using sum of the fractions method will be employed.
This method will combine applicable surface and volumetric DCGLs along with a shielding code,(e.g. MicroShield) to calculate the respective dose from residual activity remaining on structural surfaces, within residual piping, walls and floors or within activated metal (e.g. CV steel'liner).,Use of Equation 6-1 will ensure the combined exposure is bounded for all applicable sources and result in less than the 25 mrem/yr limit.
Equation 6-1
~~(D&~LS
+
G)
+ [Direct r Dose]*
- DCL, ÷DCGL,,,
25 Where: C, = Surface contamination of radionuclide i (dpm/1 00 cm2).
C,, = Specific volume concentration of radionuclide i (pCilg).
DCGLs, = Surface contamination DCGL of radionuclide i from Table 6-2.
DCGLvI = Volumetric DCGL (25 mrem/yr) of radionuclide i from Table 6-2.
Direct y Dose = Microshield (or equivalent) shielding code calculation (mrem/yr).
6.2.1.1 Surface Area Factors Surface area factors have been developed using comparative analyses between DandD'1.0
,and RESRAD-BUILD, 3.0: -Derivation of these area-factors has been documented in Reference 6-10. These area factors have been used to deVelop'DCGLEMc-screening ývalues for residual radioactivity on building surfaces. Default surface area screening values (Reference 6-8) were
,used as inputs into-the RESRAD-BUILD, 3.0 program to determine the-annual default dose-at
-36 m2. This dose was then used to ratio agairnst-doses calculated for 25' 16, 9, 4,'and 1-rn2 areas. The calculated ratio is equal to the area factor'value for the-respective area sizes. The surface area 'DCGL."can be multiplied by the derived area'fact6r-to d6termine the DCGL-'*c.
Surface area factors for SNEC are listed in Chapter 5, Table 5-15A.
6.2.2
'Resident Farmer Scenario For this scenario the assumption is that residual radioactivity is distributed in a surface soil layer covering the plant site (surface model) or in subsurface fill materials (subsurface model). The receptor is considered to reside in a home in or near any of the areas of concern. Use of the site is for residential or light farming activities. The scenario assumes continuous exposure via multiple exposure pathways to the critical group. The critical group is the resident farming family who lives on the plant site following site remediation, grows some portion of their diet on 6-3
, SNEC FACILITY LICENSE TERMINATION PLAN I REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I the site, and drinks water, from a source at the site. The most conservative parameters are selected from.each of the areas of concern to identify a site-wide residential scenario, which results in the highestexposure.. This site-wide exposure is' then used to determine nuclide specific DCGLs for each surface and subsurface layer.
The pathways that apply to the residential farming scenario include:
a)
External exposure (while indoors and outdoors) to penetrating radiation from volume sources in the contamination layer; b)
Inhalation of resuspended surface sources through wind erosion while indoors or outdoors, tracked indoors, while excavating and spreading contaminated overburden material during home construction and yard leveling; c)
Ingestion of drinking water from a groundwater source (e.g. bedrock well);
d)
Ingestion of plant products grown in contaminated soil and/or irrigated with contaminated groundwater; e)
Ingestion of animal products (e.g. beef and milk from cattle raised onsite that ingested contaminated drinking water, plant products and soil);
f)
Direct soil ingestion; g)
Ingestion of fish from a contaminated surface water source; and h)
Direct exposure from re-excavated volume sources.
At SNEC, the shallow water table and boulders in the overburden layer discourage construction of a basement for an on-site residence. However, excavation and, spreading of fill material from beneath the top meter and into the upper overburden layer could occur in leveling sloped areas for a home site. This scenario was analyzed as part of the subsurface modeling.
Two models have been developed covering surface (Reference 6-9) and sub-surface (Reference 6-11) open, land -areas for the Resident Farmer, scenario.
Both models were developed using the RESRAD Version 6.1 computer code using the deterministic-and probabilistic options. GPU Nuclear, Inc. developed the surface model while URS Corporation developed a sub-surface model, incorporating many of the same input parameters used inmthe surface model. Due to the voluminous nature of the dose modeling results documentation has been included in electronic media (CD-ROM) and submitted to the NRC for review (Reference 6-12). The dose modeling approach and input parameter selection are illustrated in Figure 6-1.
General approaches and selection of key input parameters are discussed in the following sub sections.
6-4
"SNEC FACILITY LICENSE TERMINATION PLAN DCGL results were compared between the two models. The' most conservative DCGL values were combined to form a single list for the 25 mremlyr release limit.
The most conservative'DCGLs to implement SNEC's 4mremlyr drinking' water dose goal were similarly derived. These DCGL values are listed in Table 6-2.-
6.2.2.1 Probabilistic Approach For each radionuclide RESRAD 6.1 (in the' probabilistic -mode) was used to perform uncertainties analyses and determine the sensitive parameters.
The. appropriate input file containing all physical, behavior and metabolic parameters was generated. This file included
-Haley & Aldrich hydrogeology values (Reference 6-17),'Kds developed by Argonne National Lab (Reference 6-15), and contaminated'zone :dimensions.
DandD default values were used for metabolic and behavior inputs. RESRAD default values and distributions were used for physical parameters that could not be empirically tested or where no site-specific data existed.
A random seed of 1000 was used for uncertainty sampling. The Latin Hypercube Sample (LHS)
-method was used to generate samples of input values for the probabilistic analysis. -Uncertainty correlations were established between density and total porosity, -density and effective porosity, and total porosity and effective porosity with a correlation value specified as 0.99 for all three zones (i.e. contaminated,- saturated and unsaturated)..
The first 6 correlation tables (coefficients for;'peak of mean cdose time dose' and 'peak all pathways dose') of the MCSUMMAR.REP computer file.were extracted. Within these tables, the higher. correlation coefficient (r2 value) between the PRCC and PCC columns was selected.
,These:values determine the sensitive nature of the parameter. -Sensitive parameters were identified with correlation values greater than or equal to 0.25 or less than or equal to -0.25.,
A default case of RESRAD was run in the probabilistic mode with only the sensitive parameters varying. An LHSBIN.DAT report was then generated and imported into an EXCEL spreadsheet to identify the means and 25th and 75t" percentile values for the sensitive parameter distributions.; Applicable values were then used as base deterministic inputs.
With the exception of-C-14 and H-3, Kd values were developed -for-each SNEC related
-radionuclide by-Argonne National,Laboratories (ANL) from analysis of a-.group-of samples
,collected at the SNEC site that included materials such as soils and fly ash,-and building construction materials such as pulverized concrete, :brick and block, etc. These values were
,then reviewed~to determine their impact on dose. In all cases the lowestKd developed for each
,radionuclide from each sample type producedthe highest site dose. GPU Nuclear then selected the most conservative Kd value for each radionuclide,to represent all material types at the site, thus site soils and re-fill materials may be placed in any location at the site without exceeding site dose limits.
For C-14 and H-3, ANL recommended a value near,1 as the appropriate Kd'totbe used at the site based on the type of volumetric materials present. Since these values were recommended and not empirically derived, a review of the impact on dose at Kd values within a range of possible Kd values near..1 was conducted by GPU Nuclear, Inc. The results indicatedthat a default value of 0.25 for H-3 and 1 for C-14.would provide'the greaterimpact on dose and therefore these values were selected for use when the probabilistic analysis indicated Kd was a non-sensitive parameter. When sensitive, the approach previously described using the 2 5 th or 75h percentile of the RESRAD Kd default parameter set was selected.
6-5 REVISION 1
SNEC FACILITY LICENSE TERMINATION PLAN 6.2.2.2 Deterministic Approach Prior to running RESRAD in the deterministic mode, a new input file containing information from probabilistic mode runs, was created as follows:
Suppression of the uncertainty analysis.
The 75th percentile value-was used to replace the base-deterministic input value, for those sensitive parameters with sensitivity coefficients greater than or equal to 0.25.
c I
The 25h, percentile value was used to replace the base-deterministic input value for those sensitive parameters with sensitivity coefficients less than or equal to -0.25:
The mean value was used to replace the base-deterministic input value for those sensitive parameters not bounded by the 2 5 th and 75th percentile values.
Except when the coefficients of sensitivity for the distribution coefficients (Kd) are greater than or equal to 0.25, the minimum Argonne developed Kd was used.
To determine the applicable DCGL values for each radionuclide, RESRAD was run in the deterministic mode with the revised input file. The summary report provided the peak dose, year of occurrence and pathway breakdown for each peak-dose.-, The 25 mrem/yr dose limit was divided by the peak dose to determine a DCGL reprdentihg exposure from all pathways.
This process was used foF each radionuclide, soil region and SNEC area of concern.
For`4 mrem/yr drinking water dose goal, the above process was repeated with all pathways turned off except for the drinking water pathway. Files generated for drinking water dose analysis were appended with DW.
6.2.2.3 Radionuclide Selection To date, eleven (11) radionuclides have been identified as being significant dose contributors for the SNEC site with Cs-137 being identified as the most predominant. Reference 6-13 provides the analysis for determining site-related, radionuclides. These'radionuclides have been loaded into both RESRAD and DandD software codes to determine applicable DCGLs for each respective model. Guidance from --NUREG/CR-3474 and NUREG/CR-0130 was used to first develop a comprehensive' list 'of radionuclides that could potentially be found in media at the SNEC site, during its operation and post shutdown periods: From this list various criteria was used to deselect radionuclides. Information on site-specific radionuclides was also determined using results of,characterization' surveys, waste stream' analyses and historical site assessments that are appropriatefor each medium.
Once a list was developed a 4-step process was used to deselect radionuclides that are not applicable to SNEC.
Step 1 - SNEC has been shut'down for almost 30 years. All radionuclides with half lives less than 3 years have been deselected since they have decayed 10 half lives.
Step 2 - Over 500' samples in various media have been analyzed as part of the characterization process. Radionuclide results below minimum detectable activity (MDA) levels were deselected.
6-6
,a-REVISION I
Step 3 - Radionuclides in media that were <.1% of the total mix activity and < 10% of the
- -dose limit were also deselected. -Per Appendix E, of NUREG-1727 (Reference 6.5),
radionuclides contributing < 10% of the dose limit can be screened out.
Step 4 - Evaluate which sample media contain certain radionuclides.
From this -analysis,, seven (7) nuclides were-deselected for meeting the <1% of-the mix and
<1 0% of the dose limit criteria. Together, all these nuclides contributed 3.45% of the total dose limit (25 mrem/yr). DCGLs will be adjusted in the final site designprocess to take into account this small fraction of the dose limit. As a result of the-deselection, process and most recent characterization data, Table 6-1 has been developed listing.radionuclides present at the SNEC site. This table represents the list of radionuclides potentially found:in yolumetric media and ýon structural surface areas.
Table 6-1, SNEC Radionuclide List H-3 Eu-152 C-14 Pu-238, "
Co-60 Pu-239 Ni-63 Pu-241 Sr-90 Am-241 Cs-1 37 To date the riesults of samllple analyses -t t'he SNEC site have proViided no vali8corifirmation for the presence of Np-237 above minimum detectable activity (MDA). -Since thisý radiohunclide'is a daughter of Am-241 there is a minimal possibility'of it showihg up as a positively identified radionuclide. In the DandD and RESRAD codes the,computer analysis takesinto accourit'the dose of the parent and all the daughters in the decay chain. -Therefore, Np'237 is accounted for in the dose analyses for Am-241 and not included in the list of radionuclides of concern for the SNEC site. This is similar to how Cs'-137 (paient) and its d6ughter,',Ba-1 37mi, are treated in the dose analysis. Lab6ratory analyses are reviewed to ensure radionuclides in Table 6-1 continue to be representative of the Site: ' Should-'ar'radionuclide-appear'which-is, not onITable:6-1,a technical analysis will be performed to determine its validity*'
6.2.2.4- 'Cntam natd Zone Descriptio6n, The soil guideline (DCGL) is'defined as the radiological concentration in soil that is acceptable'if the site'is to be used without radiological Irestri6tlons. TheSNEC nsurfacetmodelt isbacsepdn a maximum sized 10,000.m 2-contaminated area, one meter thick with no cover material. The concentration of a radionuclide is considered to exceed background concentrations if'it is greater than the mean background plus twice the standard deviation of the background measurements.' Based on years of radiological'surveys at theisite the_10,000 m _contaminated area'dirnensi6n was' seiected 'as adose model"default parambeter aan6d is'cbnriidered.bunding.'
The one-meter thickness was selected based on remediation work coniducted in* 1994 at the site 6-7
,SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
5SNyC FACII ITY I TI E TlRMINATION PiLAN PI iVIlIf I
(Reference 6-14) and the average below grade groundwater level. For areas less than 10,000 m2, area factors havei been developed and listed in Chapter 5, Table 5-15. Soil at the SNEC site is defined as unconsolidated earth materials, including concrete and other structural debris that might be present.
The subsurface model calculates the dose from contaminants that may be in the saturated zone as a result of reuse of fill and-debris materials. Subsurface materials for the Spray Pond and general site areas'are very similar, consisting of approximately two meters of overburden and a greater thickness of underlying bedrock. The subsurface material in the SSGS consists of crushed, homogenized site-construction debris that is covered with one meter of clean fill.
Because of these differences, DCGLs were developed for only one material (homogenized debris) in the SSGS and for two materials (overburden and bedrock) in the Spray Pond and general areas.
6.2.2.5 Dose Calculation Times (years)
Radiation doses, health risks, soil guidelines and media concentrations are calculated over user-specified time intervals. The-source is adjusted over time to account for radioactive decay and ingrowth, leaching, erosion and mixing. Although the regulatory recommendation is to use a 1000-year period, a 10,000-year period (more conservative assumption) was used to account for in-growth and decay of long lived transuranic nuclides'that have a potential impact on the ground water pathway dose.
RESRAD uses a one-dimensional groundwater model that accounts for different transport of parent and daughter radionuclides with different distribution coefficients (Kd).
6.2.2.6 Site Geology and Hydrology Subsurface investigations have been conducted at the SNEC Facility since 1981. The purpose of the investigations was to define the geologic and hydrogeologic characteristics at-the site.
Several df the'early investigations-focused on monitor well installations at key plant locations.
Recent investigations examine'groundwater trends beyond the immediate plant area at more distant locations in - order 'to characterize a broader aspect of the geologic conditions, groundwater flow and hydraulic conductivity.
There is reportedly approximately 7 to 18 feet of overburden material overlying bedrock, (a fractured siltstone).' 'The-overburden materials generally consist'of a fill overlying a natural boulder 'layer in a dense sandy, silty, clay matrix.
Groundwater occurs in both the overburden/bedrock interface and bedrock.
Groundwater flow is toward the northwest from the Facility in both the overburden/bedrock interface and bedrock. The direction of flow is not effected by seasonal water level changes.
The groundwater data indicates that the Raystown Branch of the Juniata River is a groundwater discharge feature.- A' subsurface'discharge tunnel of a formiter, coaI fired generating station affects groundwater flow at the overburden/bedrock interfaze,, acting as both a barrier and a drain. Groundwater flow in bedrock is controlled by northwest trending fractures.
Site-specific geometry, (cross-section view) and hydrology data were used for input into the RESRAD code. This input'data was' based'on studies conducted by a contracted geology firm (Refere n'ce 6-17) or default parameters determined by the RESRAD code, whichever was more conservative.
6-8' R*VIqI*M 4
6.2.2.7 Chemical Form and KdS The chemical form of the SNEC residual radioactivity is bounded by the use of the default dose conversion factors (DCFs) in the RESRAD 6.1 code. -These DCF values are based on chemical form information in Federal Guidance Report # 11 that give the individual the highest dose per unit intake.
Distribution coefficient (Kd) values are used in the RESRAD 6.1 code to predict the behavior of radionuclides in soil.
Argonne National Laboratory has conducted tests and provided Kd measurements on SNEC soils and-fill materials:.- Results of these tests are contained in Reference 6-15.
6.2.2.8 - Water Transport Parameters '
The well from which water is withdrawn for domestic use or irrigation is conservatively assumed to be located either in the center of the contamination zone (in the mass balance, MB, model) or at the downgradient edge of the contaminated zofie (in the'-n ondispersion, ND, model). Fo'r either-location, radionuclides are assumed to enter the well as~soon as they reach the-wate'r table.- Usually, the MB model is used for smaller contaminated areas (e.g. 1000 m2 or less) and the ND model is used for larger areas. For the SNEC surface model the ND input was used as the RESRAD input. For the SNEC subsurface model the MB input was used.
6.2.2.9 Volumetric Area Factors Volumetric area factors were developed using RESRAD 6.1 and SNEC inputs for the surface modeling parameters (Reference 6-9).
In the base-case surface model the ýcdntaminated fraction of plant, meat and milk products was assumed "to -equal one (1) using jthe resident farmer scenario.
Default values of were substituted for these three input parameters for areas less than 10,000 m2. This was done so RESRAD could also scale s-maller conrtarininated areas (2500 m2,-400 M2, 100 M2, 25 M2, and 1 M2). The threedefault parameter values (-1) appropriately size the contaminated fractions of plants, meat and milk obtained from the site wheri smaller'and smalle'r area sizes are input'int& the RESRAD conmputer code.' Volumetric area factors for SNEC are listed in Table 5-15.
6-9 SNEC FACILITY LICENSE TERMINATION PLAN
, - I I
REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I 6.3 DCGL
SUMMARY
& DOSE ASSESSMENT The DandD and RESRAD codes were run to determine compliance with 10CFR20.1402. DCGL results are listed in Table 6-2.
Detailed information from'dose modeling computer runs is contained on electronic media (CD-ROM) that has been submitted to the NRC (Reference 6 12).
Table 6-2 SNEC Facility DCGL Values a 25 mremly Limit 4 mrem/y Goal 25 mremly Limit (All Pathways) '
(Drinking Water),
Radionuclide Surface Area Open Land Areas Open Land Areas b (dpm/oO0cm2)
(Surface & Subsurface)
(Surface & Subsurface)
(pCilg)
(pCilg)
Am-241 2.7E+01 9.9 2.3 C-14 3.7E+06 2
54 Co-60 7.1E+03 3.5 67 Cs-i 37 2.8E+04 6 6 397 Eu-152 1.3E+04 101 1440 H-3 1.2E+08 132 31.1 NI-63 1.8E+06 747 1.9E+04 Pu-238 3.OE+01 1 8 0.41 Pu-239 2 8E+01 1.6 0.37 Pu-241 8.8E+02 86 19.8 Sr-90
'8.7E+03 1.2 061 Footnotes:
a) While drinking water DCGLs will be used by SNEC to meet the drinking water 4 mrem/yr goal, only the DCGL values that constitute the 25 mrem/yr regulatory limit will be controlled under this LTP and the NRC's approving license amendment.
b) Listed values are from the subsurface model. These values are most conservative between the two models (i e.
surface & subsurface).
The dose assessment using these values indicates that the dose will be below 25 mrem/year TEDE release limit and the 4 mrem/year groundwater dose goal. Therefore, there is a high degree of confidence that additional refinement of the source terms and modeling assumptions are unnecessary and the site can be released for unrestricted use.
6-10
<2 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
6.4 ALARA ANALYSIS Subpart'E of 10 CFR 20 contains specific requirements for a demonstration that residual radioactivity has been reduced to a level that is ALARA (10 CFR 20.1402). Both Section 7 and Appendix D of NUREG-1727 (Reference 6-5) describe a iriethod for determining whether levels of residual radioactivity are ALARA. The method presented is used to estimate when a remedial action provides a net benefit of dose reduction when compared to the cost of performing that action. It also provides exaniple calculations of vari6us rdmedial'actiofis, sbrne of which can be applied at the SNEC Facility to reach the DCGLs for surface and soil contarrmination. As stated in Chapter 4 of this plan, comprehensive remediation will be performed at the SNEC Facility Site.' NUREG-1727:states that the ALARA requirement is met by performing the-remediated action where appropriate. The guide'further states that if a reniedial action is jerformed there is no need to evaluate whether it is necessr'g to meet the'AALARA requirerrenrt.
The SNEC'Facility rembdiation goal for all striuctfjre&su'rfaces is to reduce contamination levels to or below the screening levels (Reference Table 6-2) established 'by Reference 6-8 and the DandD code. These values were determined using highly conservative parameters to ensure that any 'residual radioactivity remaining at the site would not result in any significant impact to public health and safety. Therefore, further demonstration need not be performed.
In the case of volumretric contamination, the SNEC Facility remediati6nl"fpans are to remove residual contamination to below those levels established in Table 6-2. As stated in NUREG 1727, for sites that select the unrestricted release criteria, a mathematical ALARA analysis for removing residual'radioactivity from soil at these sites is not necessary, largely because of the high costs of waste disposal. Thus in, the-case 'of 'the SNEC Facility, no further ALARA evaluation is required after the removal of soil contamination to reach the DCGLs.
6.5 REFERENbES 6-1 Code of Federal regulations, Title 10, Part 1402, "Radiological Criteria for Unrestricted Use."
6-2 Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors," January 1999.'
6-3 RESRAD, Version 6.1, United States Department of Energy and Argonne National Laboratory, April 1998.
6-4 DandD, Version 3.1 for Windows, United States Nuclear Regulatory Commission and Sandia National Laboratory, July 1996.
6-5 NUREG-1727, "NMSS Decommissioning Standard Review Plan," September 2000.
6-6 NUREG/CR-5512 Volume 1, "Residual Radioactive Contamination from Decommissioning," October 1992.
6-7 NUREG-1549, "Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination," July 1998.
6-11
ýSNEC FACILITY LICENSE TERMINATION PLAN
- REVISION 1
SNEC FACILITY LICENSE TERMINATION PLAN 6-8 Federal Register: Wednesday, November 18, 1998 (Volume 63, Number 222)
Notices, pg 64133, Table 1 -
"Acceptable License Termination Screening Values of Common Radionuclides for Building Surface Contamination."
6-9 SNEC Calculation 6900-02-008, Revision.I, "SNEC Facility DCGL Values for Volumetric Contamination," September 2002.
6-10 SNEC Calculation E900-01-005, "Determination of Surface Area Factors", April 5, 2002.
6-11 Rogers & Associates Engineering Branch, URS Corporation, "Calculation of the Sub-Surface, DCGLs, for the Saxton Nuclear Experimental Corporation Site",
RAE-42613-003-4R1, September 20, 2002.
6-12 GPU Nuclear Letter E910-02-039 dated August 20, 2002.
"Submittal of Information to NRC in July 31, 2002 Meeting."
6-13 SNEC Calculation E900-01-030 Revision 3, "SNEC Radionuclide List", July 29, 2002.
6-14 GPU Nuclear Report, "1994 Saxton Soil Remediation Project Report", May 1995.
6-15 Argonne National Laboratory Report, "Kd Study of Site Soils and Construction Debris from the SNEC Decommissioning Project", February 2002.
6-16 CoPhysics Corporation Report, "Embedded Pipe Radiation Survey Report, GPU Nuclear Corp. Saxton Experimental Nuclear Co., Saxton, Pa.", January 2002.
6-17 Haley & Aldrich Report, "Summary for RESRAD Parameters Inputs (Revision 1)", October 18, 2001.
6-18 GPU Nuclear -Report, "Saxton Experimental Corporation Facility Site Characterization Report", 1996.
6-12 REVISION I
SNEC FACILITY LICENSE TERMINATION PLAN Figure 6-1, (Resident Farmer Scenario)
Dose Modeling Logic Chart Identify inpit parat6eiters for Which the 'absolutevalue of the sensitivit 'coefficie'nt Is'>0.25 fron 6 tables.
t file
-I&A, F Print sensitive variables faults.
Begin to run RESRAD-base case in probabilistic mode Aters with only sensitive parameters varying.
in Halt RESRAD
/
analysis after SLatin-H~yp ercube° /
sampling is I
completed.
Import LHSBIN.DAT table
'int6 Excel and ideniify means, 25% and 75%
no values for sensitive
,parameter distributions.
Examine PCC sensitivities Print mean, 25% and
] 75%
values.
6-13 REVISION 1
SNECFACIITY ICENE TEMINAION LANREVISM!fl I
Figure 6.1 (cont.)
Dose Modeling Logic Chart (cont.)
Edit RESRAD file to include 75% value as ba6e deterministic value input for each sensitive parameter that is positively correlate'd (>_0.25).
Replace base deterministic value for sensitive parameters with statistical mean value for those parameters for which mean is not bounded by 0.25
& 0.75 values..
resul
>0.5 in es tables?
Replace base deterministic KD with no75% sample statistic value.
Replace base deterministic KjD with lowest Argonne value.
Ct_
K-I 6-14 Edit RESRAD input file to in clude 25% value as base deterministic input for each sensitive parameter that is negatively correlated (<ý-0.25)
SINEC FACILITY LICENSE TERMINATION PLAN R l:Vl.*ll"* KI 1
SNEC FACILITY LICENSE TERMINATION PLAN REVISION I APPENDIX 6.1 DandD DCGL CALCULATIONS FOR BUILDING OCCUPANCY SURFACE AREA MODEL 6-15 SNEC FACILITY LICENSE TERMINATION PLAN REVISION I
Anv-i4l Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Am-241 Description :
Single Run Am-241 Executed
- 09/28/99 at 15:57:59 NRC Report Occupancy Input Section Execution Options History fil'e will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 24lAm 1.00 Code-Generated Radionuclide Activities Chain dpm/100cmA2 241Am 1.OOOOE+000 237Np 0.0000E+000 233Pa 0.0000E+000 233U 0.OOOOE+000 229Th 0.OOOOE+000 225Ra 0.0000E+000 225Ac 0.0000E+000 221Fr 0.0000E+000 217At 0.0000E+000 213Bi 0.0000E+000 213Po 0.OOOOE+000 209T1 0.0000E+000 209Pb 0.OOOOE+000 Basic Parameters Name Value Units Default Page 1
'End Time' 25560.0000 Occupancy Output Section Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 9.34E-001 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose TEDE (mrem)
Percentage External Inhalation Ingestion Total
- 3. 85E-005 9.29E-001 4.26E-003
- 9. 34E-001 0.00 99.54 0.46 100.00 "Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 9.34E-001
- 1. 84E-007 3.91E-011
- 4. 96E-014
- 1. 71E-017
- 5. 02E-020
- 5. 64E-020 1.03E-022
- 1. 05E-024 days 365.2500 Pathway 241Am 237Np 233Pa 233U 229Th 225Ra 225Ac 221Fr 217At 100.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Page 2
Am-241 213Bi 5.49E-022 0.00 213Po 0.00E+000 0.00 209T1 1.42E-022 0.00 209Pb 2.15E-024 0.00 Total 9.34E-001 100.00 Page 3 0.
C______14____
Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run C-14 Description :
Single Run C-14 Executed
- 09/28/99 at 15:22:50 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 14C 1.0000 Code-Generated Radionuclide Activities dpm/100cm^2 1.OOOOE+000 Basic Parameters Name Value Units Default
'End Time' 25560.0000 Occupancy Output Section I,,
Paae 1 Chain 14C days 365.2500
C-14 Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 6.83E-006 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose TEDE (mrem)
Percentage External 2.26E-008 0.33 Inhalation 4.37E-006 63.96 Ingestion 2.44E-006 35.71 Total 6.83E-006 100.00 Radionuclide Component of Maximum Annual Dose' Radionuclide TEDE (mrem)
Percentage 14C 6.83E-006 100.00 Total 6.83E-006 100.00
- 3.?(
joc
2
Page 2 Pathway
I Co-60 Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Co-60 Description :
Single Run Co-60 Executed
- 09/28/99 at 15:27:54 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 60Co 1.00 Code-Generated Radionuclide Activities Chain dpm/100cm^2 60Co 1.0000E+000 Basic Parameters Name Value Units Default
'End Time' 25560.0000 Occupancy Output Section Page 1 days 365.2500
Co-60 Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of license termination.
3.55E-003 TEDE (mrem) occurred 1.00 year(s) after Pathway Component of Maximum Annual Dose TEDE (mrem)
Percentage External 3.09E-003 87.07 Inhalation 4.29E-004 12.10 Ingestion 2.95E-005 0.83 Total 3.55E-003 100.00 Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 60Co 3.55E-003 100.00 Total 3.55E-003 100.00.
3.5 '-e..
e
/oCL CJ-o ree3=t. I )-: 3 Ae2-2 Page 2 4)
Pathway
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Cs-137 Description :
Single Run Cs-137 Executed
- 09/28/99 at 15:40:59 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 137Cs 1.00 Code-Generated Radionuclide Activities Chain dpm/100cmA2 137Cs 137mBa 1.0000E+000 0.OOOOE+000 Basic Parameters Name Value Units Default
'End Time' 25560.0000 Occupancy Output Section Page 1 days 365.2500
Cs-137 Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 8.93E-004 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose TEDE (mrem)
Percentage External 7.69E-004 86.13 Inhalation 6.61E-005 7.41 Ingestion 5.78E-005 6.47 Total 8.93E-004 100.00 Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 137Cs 1.24E-004 13.92 137mBa 7.69E-004 86.08 Total 8.93E-004 100.00 lo L,
)
Page 2
)
Pathway
Eu-152 Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Eu-152 Description :
Single Run Eu-152 Executed
- 09/28/99 at 15:42:47 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 152Eu 1.00 Code-Generated Radionuclide Activities Chain dpm/100cm^2 152Eu 1.0000E+000 152Gd 0.0000E+000 Basic Parameters Name Value Units Default
'End Time' 25560.0000 Occupancy Output Section Page 1 K-,ý days 365.2500 I-
Eu-152 Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 1.97E-003 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose TEDE (mrem)
Percentage External 1.51E-003 76.70 Inhalation 4.51E-004 22.93 Ingestion 7.38E-006 0.38 Total 1.97E-003 100.00 Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 152Eu 1.97E-003 100.00 152Gd 7.07E-018 0.00 Total 1.97E-003 10"0.00
)
Page 2 I
Pathway
H-3 Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run H-3 Description :
Single Run H-3 Executed
- 09/28/99 at 15:19:52 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 3H 1.00 Code-Generated Radionuclide Activities Chain dpm/100cm^2 3H 1.OOOOE+000 Basic Parameters Name Value Units Default
'End Time' 25560.0000 Occupancy Output Section Page 1 days 365.2500
H-3 Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 2.03E-007 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose TEDE (mrem)
- Percentage External 0.OOE+000 0.00 Inhalation 1.30E-007 64.17 Ingestion 7.28E-008 35.83 Total 2.03E-007 100.00 Radionuclide component of t*aximum Annual Dose Radionuclide TEDE (mrem)
- Percentage 3H 2.03E-007 100.00 Total 2.03E-007 100.00 0..%
7 PAO't ý, 3 Ot p:00oC
)
Page 2 Pathway
Ni-63 Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Ni-63 Description :
Single Run Ni-63 Executed 09/28/99 at 15:31:04 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 63Ni 1.00 Code-Generated Radionuclide Activities Chain dpm/100cmA2 63Ni 1.OOOOE+000 Basic Parameters Name Value Units Default
'End Time' 25560.0000 days 365.2500 Occupancy Output Section Page 1
Ni-63 Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of license termination.
1.38E-005 TEDE (mrem) occurred 1.00 year(s) after Pathway Component of Maximum Annual Dose Pathway -
TEDE (mrem)
Percentage External 0.00E+000 0.00 Inhalation 1.31E-005 95.13 Ingestion 6.73E-007 4.87 Total 1.38E-005 100.00 RadionuclideComponent of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 63Ni 1.38E-005 100.00 Total 1.38E-005 100.00 1.3 AA fg~~
)
Page 2
)
- DandD Version 1.0 Build 1.00.02
- Single Run Pu-238 Description :
Single Run Pu-238 Executed
- 09/28/99 at 16:03:46 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 238Pu 1.00 Code-Generated Radionuclide Activities Chain dpm/100cm^2 238Pu 1.0000E+000 234U 0.OOOOE+000 230Th 0.OOOOE+000 226Ra O.OOOOE+000 222Rn 0.OOOOE+000 218Po 0.OOOOE+000' 214Pb 0.OOOOE+000 218At 0.OOOOE+000 214Bi 0.OOO0E+000 214Po 0.OOOOE+000 21OPb 0.OOOE+000 21OBi 0.OOOOE+000 21OPo 0.OOOOE+000 Basic Parameters Name Program Session Value Units Default Page 1
'End Time' 25560.0000 days 365.2500
)
Occupancy Output Section Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 8.22E-001 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose Pathway TEDE (mrem)
Percentage External 1.17E-006 0.00 Inhalation 8.18E-001 99.55 Ingestion 3.73E-003 0.45 Total 8.22E-001 100.00 Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 238Pu 8.22E-001 100.00 234U 3.93E-007 0.00 230Th 2.90E-012 0.00 226Ra 9.OOE-018 0.00 222Rn 2.40E-022 0.00 218Po 5.41E-024 0.00 214Pb 1.55E-019 0.00 218At 0.00E+000 0.00 214Bi 8.66E-019 0.00 Page 2
Pu-238 214Po 4.95E-023 0.00 210Pb 8.85E-020 0.00 21OBi 9.56E-022 0.00 210Po 9.54E-021 0.00 Total 8.22E-001 100.00 3.o61 Page 3
Pu-239 Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Pu-239 Description :
Single Run Pu-239 Executed
- 09/28/99 at 15:52:17 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentrati'on data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 239Pu 1.00 Code-Generated Radionuclide Activities dpm/100cm^2 1.OOOOE+000 0.OOOOE+000 0.000OE+000 0.OOOOE+000 0.0000E+000 0.OOOOE+000 0.OOOOE+000 0.OOOOE+000 0.OOOOE+000 0.OOOOE+000 0.OOOOE+000 0.OOOOE+000 0.OOOOE+000 0.OOOOE+000 Basic Parameters Page 1 J
Chain 239Pu 235U 231Th 231Pa 227Ac 223Fr 227Th 223Ra 219Rn 215Po 211Pb 21lBi 211Po 207TI
Pu-239 Name Value Units Default
'End Time' 25560.0000 days 365.2500 Occupancy Output Section Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 9.03E-001 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose Pathway TEDE (mrem)
Percentage External 5.15E-007 0.00 Inhalation 8.99E-001 99.54 Ingestion 4.14E-003 0.46 Total 9.03E-001 100.00 Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 239Pu 9.03E-001 100.00 235U 1.27E-010 0.00 231Th 1.44E-014 0.00 231Pa 9.24E-015 0.00 227Ac 3.78E-016 0.00 223Fr 3.80E-023 0.00 227Th 6.88E-019 0.00 223Ra 2.98E-019 0.00 Page 2
- 1. 32E-021 4.20E-024
- 1. 54E-021
- 1. 10E-021
- 5. 12E-025 9.02E-023 9.03E-001 0.00 0.00 0.00 0.00 0.00 0.00 100.00 o.1o -5
-V-;
)
Page 3 219Rn 215Po 211Pb 211Bi 21lPo 207T1 Total
)
26 V
Pu-241 Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Pu-241 Description :
Single Run Pu-241 Executed 09/28/99 at 15:54:54 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 241Pu 1.00 Code-Generated Radionuclide Activities Chain dpm/100cm^2 241Pu 1.0000E+000 241Am 0.0000E+000 237U 0.OOOOE+000 237Np 0.OOOOE+000 233Pa 0.OOOOE+000 233U 0.OOOOE+000 229Th 0.0000E+000 225Ra 0.OOOOE+000 225Ac 0.OOOOE+000 221Fr 0.OOOOE+000 217At 0.OOOOE+000 213Bi 0.OOOOE+000 213Po 0.OOOOE+000 209T1 0.OOOOE+000 209Pb 0.OOOOE+000 Basic Parameters Page 1
Pu-241 Name Value Units Default
'End Time' 25560.0000 Occupancy Output Section Maximum'Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 2.84E-002 TEDE (mrem) occurred 57.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose TEDE (mrem)
Percentage External Inhalation Ingestion Total
- 1. 12E-006
- 2. 83E-002 1.30E-004
- 2. 84E-002 0.00 99.54 0.46 100.00 Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 1.14E-003 2.73E-002 3.27E-010 4.39E-007 1.15E-010 1.07E-011
- 2. 52E-013 days 365.2500 Pathway 24lPu 241Am 237U 237Np 233Pa 233U 229Th 4.03 95.97 0.00 0.00 0.00 0.00 0.00 Page 2
Pu-241
- 9. 34E-016
- 1. 27E-015 2.32E-018
- 2. 36E-020
- 1. 23E-017
- 0. OOE+000
- 3. 20E-018
- 4. 84E-020
- 2. 84E-002 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 225Ra 225Ac 221Fr 217At 213Bi 213Po 209T1 209Pb Total 0 0 CPA 47A>K 0,
Page 3 100.00
Sr-90 Program
- DandD Version 1.0 Build 1.00.02 Session
- Single Run Sr-90 Description :
Single Run Sr-90 Executed
- 09/28/99 at 15:32:48 NRC Report Occupancy Input Section Execution Options History file will be generated.
Implicit progeny doses will not be included with explicit parent.
Concentration data will be calculated.
Initial Radionuclide Activities Chain dpm/100cm^2 90Sr 1.00 Code-Generated Radionuclide Activities Chain dpm/100cm^2 90Sr 90Y 1.OOOOE+000 0.0000E+000 Basic Parameters Name Value Units Default
'End Time' 25560.0000 Occupancy Output Section
)
Page 1 days 365.2500
Sr-90 Maximum Annual TEDE This scenario started 0.00 year(s) from now and ran for 69.98 year(s).
The peak dose of 2.89E-003 TEDE (mrem) occurred 1.00 year(s) after license termination.
Pathway Component of Maximum Annual Dose TEDE (mrem)
Percentage External 7.70E-006 0.27 Inhalation 2.71E-003 93.61 Ingestion 1.77E-004 6.12 Total 2.89E-003 100.00 Radionuclide Component of Maximum Annual Dose Radionuclide TEDE (mrem)
Percentage 90Sr 2.85E-003 98.72 90Y 3.69E-005 1.28 Total 2.89E-003 100.00
,i(ewi Page 2 Pathway
7.0 UPDATE OF THE SITE-SPECIFIC DECOMMISSIONING COSTS NRC's request for additional information dated November 8, 2000 requested additional information with respect to the site specific decommissioning cost information provided in Revision 0 of the SNEC License Termination Plan. GPU Nuclear's response to this request was reviewed and accepted by the NRC in conjunction with their review of the merger between FirstEnergy Corp. and GPU, Inc. The adequacy of decommissioning funding assurance for the SNEC Facility was documented by the Nuclear Regulatory Commission in the "Order Approving Application Regarding Proposed Merger of GPU, Inc. and FirstEnergy Corp. - Saxton Nuclear Experimental Facility (TAC NO. MB0215)" dated March 7, 2001.
Since that time the cost and schedule associated with the current Containment Vessel (CV) concrete removal project has exceeded what was assumed in this response. This has resulted in an overall $7 million increase in the remaining project cost beyond the $19.8 million estimate provided in GPU Nuclear letter E910-01-002 dated February 14, 2001, "Partial Response to Request for Additional Information, RE: License Termination Plan, (TAC NO. MA8076) dated November 8, 2000). Thus the current overall project cost estimate is approximately $63 million.
As of July 31, 2002 approximately $50 Million has been spent on the SNEC Decommissioning Project. Thus the remaining cost to complete the project is approximately $13 Million.
GPU Nuclear Letter E910-01-004, dated February 19, 2001, "Parent Guarantee for Decommissioning Funding" committed the SNEC Owners to carry out the required activities or setup a trust fund in favor of the NRC in the event GPU Nuclear failed to perform the required decommissioning activities. The amount of this guarantee is $20 million, which exceeds the remaining cost estimate of $13 million. Thus adequate funding exists to complete the project.
7-1 REVISION I SNEC FACILITY LICENSE TERMINATION PLAN
SNEC FACILITY LICENSE TERMINATION PLAN 8.0 SUPPLEMENT TO THE ENVIRONMENTAL REPORT The SNEC Facility Environmental Report was originally submitted in April 1996 to support SNEC Facility decommissioning. Three supplements to the report, in the form of responses to NRC requests for additional information, were provided on July 18, 1996, March 3, 1998 and March 31, 1998. In support of the SNEC Facility License Termination Plan, these documents were reviewed and revisions were made to reflect updates to information. Revision 1 of the SNEC Decommissioning Environmental Report was submitted under separate cover as GPU Letter 1920-00-20025, dated February 2, 2000.
Revision 2 of the Updated SNEC Environmental Report is being submitted under a separate cover letter as GPU Nuclear Inc.
E910-02-046.
The most significant changes in this document are:
- 1.
Updated information for the SNEC site to reflect its current configuration. Selected plant structures such as the septic system, weir pipe and other underground piping have been removed. A description of the SSGS Intake Tunnel and various Penelec buildings has been added to reflect the impact of decommissioning activities.
- 2.
Updated the estimated occupational exposure from approximately 37 person-rem to 37.84 person-rem based on changes in work scope for CV concrete removal and experience gained on the SNEC Facility Decommissioning Project. This occupational exposure is still well within the GElS (NUREG 0586) estimate of 344 person-rem,
- 3.
Updated geology, hydrology, endangered species and groundwater flow studies conducted in 2001. In addition population and population trends have been updated to reflect current year 2000 census data.
- 4.
Updated current status of plant dismantlement activities, sources of waste generation (radioactive, hazardous and asbestos), and current decommissioning schedule.
- 5.
Updated accident and exposure analyses for total public dose from radwaste shipping and re-estimated land area occupied by disposal of radwaste.
- 6.
Updated historical and future radwaste volumes.
Analyzed these volumes in comparison to NUREG-0586 criteria for a reference test reactor.
- 7.
Included table summarizing SNEC vs NUREG-0586 comparative analyses which includes updated and previously submitted information.
These changes are still within the bounds of the GElS and so it can be concluded that there are no significant environmental changes at the SNEC Facility associated with License Termination.
8-1 REVISION I