ML20071Q249

Rev 3 to Shoreham Decommissioning Project Termination Survey Plan

ML20071Q249
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	07/31/1994
From:	LONG ISLAND POWER AUTHORITY
To:
Shared Package
ML19304C479	List: ML20071Q249 SNRC-2177, Forwards Rev 3 to Shoreham Decommissioning Project Termination Survey Plan. Portions of Rept Withheld
References
PROC-940731, NUDOCS 9408110144
Download: ML20071Q249 (91)

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LONG lSLAND P OWER l AUTHORITY I

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Shoreham i Decommissioning l Project l s

Termination Survey Plan l l

l Revision 3

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Termination Survey Section Radiological Controls Division Operations and Maintenance Department ,

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July,1994 __

9408110144 940804 i PDR ADDCK 05000322 l l

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i Effective Date: 7-2s-44 Long Island Power Authority Shoreham Decommissioning Project  :

i Termination Survey Plan  !

Revision 3 i

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Signature D6te /

LIPA Site Review Committee Approved by: W / 8472 #/9f Signature // Date LIPA SNPS Resident Manager a

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Termination Survey Plan Rev. 3 Preface This document describes the methods used by the Long Island Power Authority (LIPA) to demonstrate that radiation and radioactive contamination levels of the Shoreham Nuclear Power Station have been reduced to levels below criteria established for release for unrestricted use. It supplements and updates the description of the proposed final radiation survey presented in the Shoreham Decommissioning Plan (LIPA90).

This Plan describes the technical methods to be used and provides guidance for planning and design of the Termination Survey. It is prepared and controlled under the LIPA Nuclear Management Control Manual, Termination Survey Program Description (LIPA92).' The methods described are derived from regulatory guidance, specifically Regulatory Guide 1.86 (USAEC74) and draft NUREG/CR-5849 (BE92); and from recent U.S. reactor facility decommissioning experience (Pathfinder, Saxton, Shippingport, UC Berkeley), taking into account conditions at the Shoreham facility.

Revision 0 of this Plan was issued November 20,1992 and was used for early survey work beginning in January,1993. Revision 1, issued in April 1993, incorporated changes necessitated by initial survey experience as well as commitments made to the Nuclear Regulatory Commission (NRC) resulting from staff review of the Plan. Revision 2, issued in December,1993 corrected typographical errors, and provided updated information on: instruments used, the number of survey units, the final report outline and other minor editing changes. The current revision, Revision 3 has been issued to incorporate modifications to the Shoreham release criteria which account for the potential presence of Tritium and Iron-55 in areas where residual contamination from neutron actvated materials of construction may be present (USNRC94). The revision describes the use of adjustment factors for surface activity measurements to account for Tritium and Iron-55 for comparison of measurement results to release criteria guideline values.

It also updates descriptions ofinstrumentation to add new detector assemblies developed for survey of embedded piping and updates the Shoreham Termination Survey Classification Summary to account for changes in the survey unit listing.

8 This Plan is a companion document to the Shoreham Decommissioning Project Termination Survey Program Description (LIPA92). The Program Description describes the organization and management responsibilities for the Termination Survey Program.

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Termination survzy Plan Ray. 3 SHOREHAM DECOMMISSIONING PROJECT l l

TERMINATION SURVEY PLAN l 1

l Table of Contents Section Iille Page No.

1.0 Historical Background Information 1-1 2.0 Site Information 2-1 3.0 Termination Survey Overview 3-1 4.0 Survey Plan and Procedures 4-1 5.0 Data Interpretation 5-1 6.0 Final Repon 6-1 7.0 References 7-1 8.0 Glossary 8-1 Appendices:

A. Survey Design Guidelines A-1

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Termination Surv2y Plan Rev. 3 List of Tables - l i

i Number Title Page No.

i 2.1 Shoreham Termination Survey 2-3  ;

Classification Summary i

i 3.1 Acceptable Surface Contamination Levels 3-1 3.2 Radioactivity Composition for Surface l Activity Measurements 3-3  !

4.1 Termination Survey Instrument Summary 4-2 j i

4.2 Measurement Detection Sensitivities 4-4 4.3 Embedded Piping Detector Assembly Sensitivities - 4-5  ;

I 5.1 Adjustment Factors for Surface Activity Measurements 5-3 i

5.2 Gross Activity Guideline Values 5-9 {

6.1 Termination Survey Detail Data Report 6-5 l

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Termination Survey Plan Rev. 3 List of Fivures ,

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l 2.1 Shoreham Site (Panial) Showing Termination 2-4 Survey Area i 4.1 View of Stmetural Survey Unit Showing 4 - 12 l Grid Placement r

4.2 Shoreham Decommissioning Termination 4 - 13 i Survey Grid Map 5.1 Data Review Flowpath For 5 - 10 i Total Surface Activity -  :

.Affected Survey Unit .

5.2 Data Review Flowpath For 5 - 11 ;

Total Sortace Activity j Unaffected Survey Unit ,

5.3 Data Review Flowpath For 5 - 12 Removable Surface Activity 5.4 Data Review Flowpath For 5 - 13  ;

Gamma Exposure Rate Measurements j 5.5 Data Review Flowpath For 5 - 14 Soil Samples l

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Termination Survey Plan R:v. 3 1.0 Historical Background Information The Shoreham Nuclear Power Station (SNPS) power generation plant consists of a boiling water reactor (BWR) nuclear steam supply system (NSSS) and a turbine generator both furnished by General Electric Company. The balance of the plant was designed by Stone

& Webster Engineering Corporation. The plant was designed to provide a gross electrical output of 849 Megawatts (LILCO90).

The SNPS achieved initial criticality in February 1985 following receipt of the initial operating license from the Nuclear Regulatory Commission (NRC). A license to operate at power levels not to exceed 5% of full power was granted and low power testing commenced in July 1985. The plant was operated intermittently at power levels not exceeding 5% of full power until the final critical operation in January 1989. This operating history corresponds to 2.03 effective full power days (EFPD) of average fuel exposure (LIPA90).

Pursuant to the 1989 agreement among The Long Island Lighting Company (LILCO), the State of New York and the Long Island Power Authority (LIPA), power generating operations at (SNPS) were terminated. The irradiated fuel was removed from the reactor vessel pressure in August 1989 and placed in the spent fuel pool.

The DECON alternative was selected for decommissioning SNPS and an order approving the LIPA Decommissioning Plan was issued by the US Nuclear Regulatory Commission (NRC) in June,1992 (USNRC92). The objective of the approved DECON alternative was to decontaminate the SNPS facility and site and release them for unrestricted use. To accomplish this, the majority of the radioactive portions of the reactor pressure vessel and pressure vessel internals were disassembled, segmented and removed. Contaminated and activated portions of plant piping systems and equipment were decontaminated or removed as described in the LIPA Decommissioning Plan (LIPA90), including subsequently approved changes thereto.

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1 Termination Survey Plin Rev. 3 l

2.0 Site Information i 2.1 Site Description ]

The Shoreham Nuclear Power Station site is located in the Town of Brookhaven, Suffolk County, New York on the north shore of Long Island. The site is 50 miles east of the confluence of the East River and Long Island Sound, near la Guardia Airport. -

The developed portion of the she comprises 80 acres, and is located within a larger parcel of 499 acres owned by the long Ialand Lighting Company (LILCO).8 The 499 ,

acre parcel is bounded on the nonh by long Island Sound and on the east by the Wading River marshland. It is bounded on the west by a parcel of approximately 429 acres  !

known as the Shoreham West propeny, also owned by LILCO, and on the south by highway Route 25A. The 499 acre SNPS site property is divided across its midsection ,

in the east-west direction by North Country Road which branches off Route 25A about  ;

three miles west of the site and rejoins 25A about three miles east of the site. North Country Road is about 1,500 ft. south of the Reactor Building at its closest point to the developed area of the site. Figure 2.1 shows the SNPS site plan and the location of major buildings on the developed portion of the site. r The site elevation varies from sea level at Long Island sound (the northern boundary of the site) to elevation 200 feet midway between North Country Road and the southern border of the site. Except for the developed area, the site is wooded with wetlands along the east and west boundaries extending as much as 1,300 feet inland from the Sound. The developed portion of the site is fairly level with the exception of several graded slopes, the largest of which is a terraced slope about 30 feet in height which traverses the site in an east-west direction immediately to the south of the Reactor Building. The ground surface covering in the developed area is mostly gravel with smaller portions devoted to lawn and paved areas (sidewalks, loading areas and roadways). The site soil cover in unpaved or undisturbed areas is a mixture of sand and glacial till (gravel). Vegetation cover in undisturbed areas is a mixture of grass and weeds with a few shrubs and small ,

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2.2 Site Conditions for Termination Survey The Shoreham facility remains largely intact following decommissioning. Dismantlement of structures was confined to small ponions within the Reactor, Radwaste and Turbine Buildings. Removal of such structures was performed to provide paths for removal of contaminated piping and equipment. As described in the LIPA Decommissioning Plan (LIPA90), most radioactive piping and equipment was dismantled, removed from the facility and disposed of as radioactive waste at a licensed radioactive waste disposal

' Under the Asset Transfer Agreement, approximately 11 acres of the site which include ,

i the power block, adjacent office and suppon buildings and connecting roadways have been transferred to LIPA for conduct of the decommissioning (LIPA90).

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e-l Termination Surv2y Plan Rzv. 3 l

facility. This includes the reactor pressure vessel, reactor pressure vessel internals and major portions of 14 plant systems. Decontamination of plant piping and equipment was performed to a much lesser extent. Equipment items which remain and were decontaminated in place include the reactor vessel bottom head, a portion of the main steam lines in the Reactor Building, the Condensate Storage Tank (CST), and Residual Heat Removal (RHR) system heat exchangers (shell only). It is estimated that greater than 75 percent of the piping and equipment on site during the time of reactor operation will remain after decommissioning is complete.

2.3 Site Areas Covered - Scope of Survey The Shoreham facility and environs have been evaluated to identify the areas to be covered in the termination survey. The termination survey focuses on the area within the Secured Area fence as shown in Figure 2.1. The area (approximately 20 acres) contains the Reactor, Radwaste and Turbine buildings, and other buildings, facilities, and grounds within the Secured Area fence. The Secured Area is described in the Shoreham Nuclear Power Station Updated Safety Analysis Report (LILCO90). The area within the Secured Area fence coincides with the Restricted Area defined in Shoreham radiological control procedures as the area where access has been controlled and radioactive materials controlled for purposes of protection ofindividuals from exposure to radiation. Following completion of Shoreham decommissioning and removal of the irradiated fuel from the site, areas outside the Restricted Area may be added to the survey if used for temporary storage or handling of irradiated fuel or other radioactive materials.

Major attention in the termination survey is given to the areas most affected by reactor operations and by decommissioning activities. The area covered by the termination survey has been divided into approximately 390 individual " survey units" for management of the survey. Each survey unit is classified as "affected" or " unaffected" for survey implementation.2 Affected areas are largely confined to the Reactor and Radwaste buildings, and portions of the Turbine building. Table 2.1, Shoreham Termination Survey Classification Summary, summarizes the breakdown of the facility into affected and unaffected areas. This table is updated from time to time, as needed, and maintained via an approved procedure.

The environs of the facility beyond the area of the site encompassed by the termination survey have been demonstrated to be free of detectable radioactivity from Shoreham operations. This is well documented by the Shoreham Radiological Environmental Monitoring Program (REMP).

2 An "affected" area as defm' ed in draft NUREG/CR-5849 (BE92) and used in this Plan ,

is a designation used to indicate that an area (survey unit) has a potential for containing residual radioactive contamination. An " unaffected" area is one which is not expected  !

to contain residual radioactivity based upon the operating history and previous j radiological surveys (see Glossary).

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Termination Surysy Pl:.n Rev. 3 Table 2.1 -

Shoreham Termination Survey Classification Summary l Total No. No. of Affected No. of Unaffected ;

DESCRIPTION CODE of Survey Units Survey Units Survey Units STRUCTURES ,

Reactor Bldg. RB 85 84 1 Drywell PC 14 14 Suppression Pool SP 5 5 i

Turbine Bldg. TB 106 8 98 Radwaste Bldg RW 50 48 2 i

Control Bldg CB 4 4 l

O & S Bldg OB 4 4 i

l O & S Bldg Annex AB 4 4 Other Site Bldgs OS 2 _1 _8 l

Structure Totals 281 160 121 OUTSIDE AREAS  !

I Site Grounds SG 7 7  !

i Soil Samples SS 1 1 Structure Exteriors SE M M .

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Outside Area Totals 22 0 22 PLANT SYSTEMS SU 82 38 44 i TOTALS 385 198 187 i

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Termmation Survey Pian Rev. 3 3.0 Termination Survey Overview 3.1 Survey Objectives The termination surveyisdesigned to demonstrate that licensed radioactive materials have been removed such that residual levels of radioactive contamination are below applicable Regulatory Guide 1.86 limits (USAEC74). Radiation detection instrumentation requirements for the survey are based upon the conclusion that the controlling radioactive species in determining compliance with Regulatory Guide 1.86 release limits are activation products dominated by Co-60. These are beta-gamma emitters as defined in Regulatory Guide 1.86(USAEC74), hence the limits for beta-gamma emitters shown in Table 3.1, below, apply. Special limits have been established for the hard-to-measure radionuclides Fe-55 and Tritium in areas where residual contamination from neutron activated materials of construction may be present (USNRC94). The release criteria limits for Shoreham are shown in Table 3.1.

The applicable release limits for alpha emitters are also shown in Table 3.1.8 These are for Natural Uranium, U-235, U-238 and associated decay products. Instruments and methods are being incorporated into the survey which are adequate to measure alpha surface activity at levels below the limits in Table 3.1.

Table 3.1 Accentable Surface Contamination Levels

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(dpm per 100 cm')

Fixed Plus Removable Removable -

Activity Averace Maximum Beta-Gamma 5,000 15,000 1,000 Alpha 5,000 15,000 1,000 Fe-55, Tritium 200,000 600,000 1,000 The release criteria for Shoreham also include guideline values for gamma exposure rate and soil radioactivity concentration. The average guideline value for gamma exposure rate is 5 pR/hr above background, measured at one meter from accessible surfaces in the facility buildings and outdoor areas (LIPA90). The average is calculated over an area not to exceed 10 m 2. In addition, any individual gamma exposure rate measurement shall not exceed 10 R/hr above background, the elevated area guideline. The guideline value for soil and bulk materials radioactivity concentration at Shoreham is 8 pCi/gm. Bulk materials include activated concrete, sewage sludge, tank bottoms and sediments, radwaste treatment media (e.g. charcoal

• The various release limits are also referred to as guideline values (Be92). See Section 8.0, Glossary, for definitions of guideline value and elevated area  ;

guideline. (

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Termination Surv:y Plan Rev. 3 beds) and any other materials not readily evaluated by direct measurements. No elevated area guideline has been established for soil and bulk materials.  !

3.2 Identity of Contaminants The Shoreham Characterization Study (LILCO90a) reported a total facility radioactivity inventory due to Shoreham operations of about 600 Curies (Ci) (not including the irradiated feel, control blades and readily removable reactor asssembly items). All but a very small fraction (less than one Ci) of this inventory was contained in activated materials of the reactor pressure vessel and vessel internals l which have been removed from the site. The calculated radioactivity composition of l the activated components (as of July,1990) is: Fe-55,69%; Co-60,28%; and Ni-63, i 2%. Minor amounts of other species are calculated to be present, including: H-3, C-14, and Ni-59 (LILC090a). Laboratory analysis of activated reactor pressure vessel and vessel internals samples indicates the presence of low levels of Mn-54,2n-65, and Ni-63,in addition to Co-60. Analysis of bioshield wall samples show low levels of Co-60, Mn-54 and Eu-152 (TU92).

In most areas of the facility, the most likely source of residual contamination is activated corrosion products (crud) deposited in piping systems which could have been transported into the facility via leaks or handling of contaminated equipment.

Detectable radionuclides in surface corrosion deposits are confined to a small number of radionuclides. An analysis of piping system corrosion products in nine (9) discrete samples shows that these are comprised largely of Co-60 and Fe-55, wit a "best fit"(by least squares regression) Fe-55:Co-60 ratio of approximately D , fa L1 i0.13)(TU92). The " average"value of the individual Fe-55:Co-60 ratios frc;

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same 9 samples is 0.30. Accounting for isotopic decay from the time of con wn until July 1994, the average value becomes 0.19. This last number is used to adjust individual measurements, as described in Section 5.0,to account for the Fe-55 not detected by survey instruments. Laboratory analysis of plant corrosion product deposits has not identified the presence of reactor-produced alpha emitters above lower limits of detection.

Several areas have been identified where residual contamination sources exhibit composition characteristics which differ from surface corrosion product deposits described above. The Reactor Vessel Bottom Head, the Spent Fuel Pool, the Reactor Biological Shield wall remnants and areas where Reactor Vessel and Vessel internals were cut up and dismantled, may contain significant fractions of hard-to-measure (HTM) radionuclides such as Fe-55 which decays by electron capture, and Tritium which is a low energy beta emitter.

The radioactivity compositien for surface contamination measurements in all areas in the termination survey is given in Table 3.2.These results (except for the Spent Fuel Pool) are derived from neutron activitation calculations reported in the Shoreham Site Characterization report (LILCOO9a) adjusted for radioactive decay to July 1,1994. The Spent Fuel Pool composition is obtained from radiochemical analysis of a spent fuel pool cleanup system filter (SCI 94). Table 3.2also contains the adjustment factors, f which are applied to direct and removable surfac activity measurements. Application of adjustment factors is described in Section 5.0.

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Tennination Survey Plan Rev. 3 Table 3.2 Radioactivity Composition for Surface Activity Measurements Area - Description Nuclides Fraction of Adjustment Total Activity Factor Reactor Bioshield Concrete Co-60 0.0278 35.9 Fe-55 0.3159 H-3 0.6559 Reactor Bioshield Concrete Co-60 0.0809 12.4 removable activity Fe-55 0.9181 Reactor Bioshield Steel Co-60 0.0413 24.2 Fe-55 0.9487 Reactor Bioshield Steel Co-60 0.830 1.2 removable activity Fe-55 0.170 Dry Cutting Station Co-60 0.3840 2.6 (total and removable activity) Fe-55 0.5635 Ni-63 0.0517 Spent Fuel Pool Co-60 0.4324 2.3 (total and removable activity) Fe-55 0.4764 Ni-63 0.0868 Reactor Vessel Bottom Head Co-60 0.3858 2.6 Fe-55 0.5617 Ni-63 0.0521 All Other Areas Co-60 0.830 1.2 Fe-55 0.170 3.3 Organization and Responsibilities An organization, identified as the Termination Survey Section of the Radiological Controls Division, has been created within the LIPA - Shoreham Decommissioning Project organization for planning and implementation of the termination survey.The organization and responsibilities of the Termination Survey Section and the interfaces and responsibilities for all other elements of the Decommissioning Project Organization for the termination survey are described in the Termination Survey Program Description (LIPA92).

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Termination Surv:y Plan Rev. 3 3.4 Training 3.4.1 Technician Qualification A job qualification category for termination survey technicians will be established under the existing Shoreham Station " Health Physics Technician Selection, Training and Qualification Program" procedure. The training and qualification acceptance criteria from this procedure will be adopted. The training will include classroom and functional job performance training on termination survey procedures and specialized instrumentation. The training and qualification process for individual technicians willrequire from three to five days to complete. Termination survey technicians will generally be selected from the pool of technicians who have previously been qualified to perform HP responsibilities on the Shoreham Decommissioning Project. When new technicians are hired directly to support the survey, their training will include in addition to termination survey training, those portions of the overall site Health Physics Technician training necessary to ensure proper job performance.

3.4.2 Classroom Training Classroom training includes: an overview of the Termination Survey Program, instrumentation, and procedures. The overview will cover termination survey objectives, survey methods, the role and responsibilities of termination survey technicians, the importance of personnel safety, termination survey quality assurance, the Termination Survey Plan and program implementation.

3.4.3 Functional Training Functional training will involve hands-on performance of principal HP technician termination survey tasks. A survey of a typical structural and system survey unit will be performed by each technician under the surveillance of a qualified instructor.

3.5 12boratory Services Laboratory radioanalytical services of the Shoreham Nuclear Power Station Radiochemistry Section and/or Health Physics Section, both within the Radiological Controls Division will be used in support of the termination survey. The on-site i capabilities include gamma spectroscopy (GeLi) of filters, smears and bulk samples; liquid scintillation; gas proportional counting; gross beta-gamma counting of smears; and gross alpha counting of smears. Both Sections operate under approved QA programs and procedures. Additionally, a contract is in place with a qualified vendor for specialized radiological analysis of samples on an as-needed basis. Vendors are selected in accordance with the requirements of the LIPA QA Manual, Appendix N (LIPA92a).

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Termination Survry Plan Rev. 3 3.6 General Survey Plan The termination survey is implemented at the individaal survey unit level. Three l categories or types of survey units have been established: 1) structures, which include i building interiors, 2) plant systems and 3) outdoor areas. These categories combine survey units into groups with similar physical characteristics. The survey is planned for measurements to be taken for each survey unit independently. The measurement intensity of each survey unit is based upon its classification as affected or unaffected.

Due to the large scope of the termination survey and the requirement that some survey activities be conducted in parallel with decommissioning work, a systematic approach is necessary. Further, it is essential that key interfaces between survey activities and other decommissioning work activities be identified. l The termination survey planning and implementation process for each survey unit involves the following steps: 1) initial classification; 2) history file preparation and classification review; 3) turnover for termination survey; 4) walkdown; 5) survey design; 6) preparation of modification packages (primarily for plant systems surveys);

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7) preparation of work requests and scheduling; 8) preparation of final survey instructions; 9) physical support activities such as erection of scaffolding, system 1 tagout and system breaches for system surveys; 10) performance of the survey; and

11) post survey restcration and control of surveyed areas (isolation of systems after survey completion). These are described in the following paragraphs.

3.6.1 Initial Classification The classification of the facility into "affected"and " unaffected" areas provides the planning basis for the termination survey. It was conducted using results from the Shoreham Site Characterization Program (LILCO90a) and the recommendations of experienced Shoreham personnel using the classification criteria contained in Appendix A, Survey Design Guidelines. The classification status of all survey units is maintained in the Termination Survey Classification Description, and is controlled by a procedure of the same title.

3.6.2 History File Preparation The history file is a compilation, in a standardized format, which summarizes the operational and radiological history of each survey unit included in the termination survey. Preparation of the history file involves review of tl system description (for system survey units), plant operating records, the Shoreham Characterization Report (LILCO90a), radiological surveys and other relevant information. Specific operating history which could affect the 1

radiological status is sought in this review. The purpose of this process is to provide a substantive basis for the survey unit classification, and hence the level ofintensity of the termination survey.

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Termination Survey Plan R:v. 3 The history file contains a summary description of the survey unit and, for system history files, summarizes relevant operational data. Relevant operational data includes operating lineups to radioactive systems, and other events which could affect the radiological status. Similarly, for structures and outdoor areas, the extent of radioactive materials involvement in the area (if any) is summarized.

Possible outcomes of the review are: the initial classification is verified, or it is modified to reflect the more thorough evaluation. Additionally, the review may suggest specific areas of a structure or components of a system which should be highlighted in the final survey. The review and conclusions are documented in a checklist. The history file also identifies the sources of information used. It may contain summaries, excerpts or complete documents which are useful for survey design. The system history file and its preparation are described in a Termination Survey procedure.

3.6.3 Turnover for Termination Survey Prior to acceptance of a survey unit (structure, system or outdoor area) for the termination survey, a number of conditions must be satisfied.

Decommissioning activities are completed, all tools are removed, housekeeping and area cleanup is completed, decontamination of affected structural areas and system residual components is completed and verified by operational radiological surveys,and scaffolding needed to be left in place for termination surveyisidentified. Radiological surveysverifyingthe status of the area, if a structure, and remaining system components, if a system, are provided to the Termination Survey Section. Turnover and control of systems, structures and outside areas is controlled by a Termination Survey procedure.

3.6.4 Walkdown The walkdown is a key activity in the preparation of the survey design. For systems,it includes review of system flow diagrams and piping drawings, and physical walkdown of the system. Structures and outdoor areas are also physically walked down. A principal objective is to assess the physical scope of the survey unit and to identify potential breakdown into subunits. Special access needs are identified. Potential support requirements for conduct of surveys are identified, such as scaffolding, component disassembly, interference removal, engineering modifications, electrical tagout and system alignment to provide access for surveys. Safety concerns, such as access to confined spaces, high walls, and ceilings, are idendfied and resolved. It is noted that for survey units involved with decommissioning activities, the walkdown is best completed when the final configuration is known, usually near or after the completion of decommissioning work. Early information is available through Decommissioning work packages and material takeoff lists and drawings.

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Termination Surv:y Plan Rtv. 3 3.6.5 Survey Design The survey design results in the identification of the locations (grid blocks, system components) to be surveyed and the type of radiological measurement or sample to be collected at each location. The design is based upon the guidelines in Appendix A.The design of the servey for individual survey units is prepared in accordance with a Termination Survey procedure.

3.6.6 Engineering Once the survey design is prepared, engineering review and support requirements are developed. In the rare instances where engineering modifications are required for surveys,the modification packages are prepared and the design reviewimplemented through approved engineering procedures.

3.6.7 Work Planning and Scheduling Upon completion of any required engineering review, the physical modifications are specified. Field work is implemented via the Maintenance Work Request (MWR) process. The MWRs identify all components which require opening, identify all modifications, indicate restoration requirements and indicate whether a system is to be isolated or returned to service. The MWR process is also used to initiate support work and tagouts necessary for surveys of structural and outdoor survey units. The survey unit support work is then placed upon the Project work schedule for performance.

3.6.8 Survey Instructions The survey instructions are provided to the Lead HP technician assigned responsibility for the specified survey unit. They specify the number and type of radiological measurements to be taken at each location or component identified in the survey design. The instructions identify smear samples and other samples to be collected. The survey instructions identify those survey points (components or other specified locations) where QC verification surveys are required. The survey instmetions are prepared by the designated Termination Survey Radiological Engineer in accordance with a Termination Survey procedure.

3.6.9 Field Support The MWR identifies each component or survey location requiring support work and tagouts. In cases where special surveys are required such as components, embedded piping, or large tanks which are classified as affected, other preparation work may be required. This may include gridding oflarge tanks once access is provided and safety precautions have been satisfied. j 1

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Termination Survey Plan Rev. 3 3.6.10 Survey Measurements Termination survey measurements are conducted in accordance with Termination Survey procedures and the specific survey instructions for the survey unit. The measurements include surface scans, direct measurements of surface contamination, smear samples for removable surface contamination and gamma exposure rate measurements. The measurements are discussed in Section 4.0. Prior to conducting the survey, a walkdown is conducted by the cognizant Termination Survey Radiological Engineer and the Irad HP technician to verify the survey locations and the details of the instructions.

3.6.I1 Data Management and Evaluation Upon completion of field measurements and sample collection (smears, soil and sediment as applicable), all measurement data and sample counting results for each survey unit are reviewed for completeness. The survey measurement data is then entered into a custom database designed to store measurement raw data, perform calculations necessary to convert individual measurements to reporting units, calculate summary statistics and generate data reports for each survey unit. The calculations and evaluations performed are described in detail in Section 5.0, Data Interpretation.

3.6.12 Reporting Results - Release Records The results of the survey of each survey unit are reported individually via a document called a " release record". Each release record consists of a written summary which describes the survey unit and presents the comparison of the survey results to release criteria guideline values. Attached to the text are data reports which include each measurement result as well as the calculated summary statistics for the survey unit. Maps are also included with each release record to identify the location of survey measurements. In rare cases, a release record may cover a partial survey unit. e.g. those portions of the Radioactive Waste system within the Turbine Building. Also, where several survey units are of similar composition or are very closely related functionally, they may be combined into a single release record for simplification of the reporting process.

3.6.13 Restoration and Isolation

a. Systems After survey measurements have been taken, reviewed and approved, and QC verification survey measurements have been completed in system survey units, the system is restored and components are replaced as specified in the MWR, If indicated in the MWR, the system is isolated to protect against recontamination. Isolation and 3-8 i

Termination Surv:y Plan Rev. 3 control of plant systems after termination survey is performed under a specific approved procedure.

It is noted that many plant support and service systems will be returned to service after completion of termination survey ,

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measurements. Examples are: compressed air, heating and cooling, ventilation and fire protection. When a system is required to remain in service, administrative controls will be used to minimize the possibility of system contamination. These include, but are not limited to, surveillance activities to ensure that the system is not aligned or operated in a manner which could compromise termination survey results.

b. Structures and Outdoor Areas After measurements are completed in structures and outdoor areas, appropriate controls are used to prevent or minimize possible contamination. This is controlled by an approved station procedure.

Numerous structural survey units have been defined within the Radiological Controlled Area (RCA) of the Shoreham facility. These include all of the structural survey units which comprise the Reactor Building, the Turbine Building and the Radwaste Building, representing the majority of the areas within the power block. As sufficient numbers of surveys are completed in contiguous RCA survey units (with completion of the surveys being indicated by complete, approved release records) to allow manageable reduction of the RCA boundaries, these reductions will be made. Removal of such areas from the RCA provides additional assurance that material containing radioactivity is not used or transported through previously surveyed areas.

3.7 Quality Assurance l

3.7.1 General Provisions .

As indicated in the LIPA Decommissioning Plan and in the Termination l Survey Program, quality assurance for the termination survey is subject to the provisions of the Decommissioning Program Quality Assurance Manual (LIPA92a). In addition, the Termination Survey Program itself has established quality control measures as an integral part of the program. Principal measures established to meet quality objectives are:  !

a. Selection and Training of Personnel  !

Qualification requirements and responsibilities are established for key personnel performing termination survey tasks. A technician training l l

3-9

1 Tenmnation Surv y Plan Rev. 3 i and qualification program has been established which includes classroom training and job-functional training. Training and qualification records are maintained on all technicians selected for the termination survey.

b. Instnimentation Selection, Calibration and Operation An evaluation and testing program was conducted to select radiation detection instrumentation for the suney. Instrument calibration is performed either under approved SNPS calibration procedures using calibration sources traceable to the National Institute of Standards and Technology (NIST), or by qualified vendors with the results traceable to NIST. Measurements are performed using approved written procedures for each instrument. Control ofinstruments is established by an instrument control procedure.

c. Survey Documentation Each termination survey measurement isidentified bydate, instrument, technician, location, type of measurement, and mode of instrument operation.

d. Quality Control - Verification Replicate measurements are performed independently on a selected sample of survey measurements on an ongoing basis,

e. Written Procedures All termination survey tasks which are essential to survey data quality are controlled by procedures reviewed and approved by the LIPA Site Review Committee.

f. Mockup of Procedures and Processes Dry nms and mockups are performed to test principal procedures and methods prior to implementation in the field.

g. Chain of Custody Written procedures establish responsibility for custody of samples and survey data between the point of measurement or collection until final results are obtained.

3 - 10

Termination Survey Plan Rev. 3

h. Records Management Generation, handling and storage of termination survey design and i data packages is controlled by an approved procedure.

i. Data Management Software I

Computer programs generated for processing of survey measurement data shall be tested and verified.

j. Independent Review of Survey Results Th- release record of each survey unit is given independent review prior to acceptance for final management approval.

k. Control of Surveyed Areas and Systems Administrative (i.e., procedural) and physical controls are established on areas and systems to minimize the possibility of contamination subsequent to the survey.

1. Control of Vendor Supplied Services Essential services, such as instrument calibration and laboratory sample analysis, will be procured only from qualified vendors, in accordance with an approved procedure whose internal QA programs are subject to LIPA audit.

3.7.2 Termination Survey Quality Control Procedure A Termination Survey Quality Control procedure controls essential quality assurance activities not addressed in other procedures. These include:

a. conduct of QC replicate sampling measurements,

b. routine verification of survey measurement data accuracy,

c. control charts for individual instruments,

d. testing of computer data calcula* ion programs,

e. validation of operational survey data used as final survey data,

f. documentation of surveys, and g, custody ofinstruments, samples and measurement data.

3 - 11

Tcrmination Survzy Plan Rev. 3 3.8 Schedule The termination survey detailed schedule is maintained wnhin the Project Schedule i by the Project Controls Division of the Finance and Administration Department and the Work Planning Section of the Operations and Maintenance Department.

The termination survey is scheduled to be completed in several major phases which encompass distinct portions of the facility.The initial phase included only the survey of the SNPS main turbine, and was reported in February,1993.The first major phase which included the outside area survey units, structural survey units outside the power block, and most systems and structural survey units within the Turbine Building, was reported in September 1993.The second phase includes systems and structural survey units within the Reactor Building Primary Containment and Suppression Pool, and was reported in February 1994. The third phase includes systems and structural survey units primarily within the Radwaste Building, and was reported in June 1994. The final phase includes all remaining survey units, the majority of which are those areas impacted by the removal of irradiated fuel from the facility,primarily the Reactor Building refuel floor and other areas not reported earlier. The final phase is scheduled to be reported in October 1994. Upon completion of each phase the release records will be compiled and the survey units covered will be available for NRC verification surveys.

3.9 Survey Report The Final Report will be prepared for submission to the Nuclear Regulatory Commission to meet the intent of Regulatory Guide 1.86(USAEC74) for final survey reporting. The Final Report will follow the guidaace of Draft NUREG/CR-5849 (BE92) regarding content. The Final Report will be submitted in stages with an updated report prepared for each of the major survey phases discussed in Section 3.8.

Each edition of the Final Report will include the survey measurements obtained during the appropriate phase and evaluation of the results. An update to the overall survey status report as well as other documents which comprise the Final Report will also be provided. The full outline of the Final Report is described in Section 6.0.

3 - 12

I Termination Survey Plan Rev. 3 l

l 4.0 Survey Plan and Procedures 4.1 General The design approach of the Shoreham termination survey is considerably affected by the final configuration of the facility, which is largely intact with the majority of equipment left in place. The Reactcr, Turbine and Radwaste Buildings contain over 200 equipment rooms. These rooms contain approximately 80 plant piping systems and the majority of the systems occupy multiple rocms.

The majority of the survey effort is confined to the areas contained within the Reactor and Radwaste Buildings and those areas in the Turbine Euing where radioactive materials were handled. These are classified as affected areas. The remainder of the areas within the scope of the survey are classified as unaffected. All radioactive material handling, movement and storage on the site has been controlled under approved procedures. No detectable activity of SNPS origin has been detected on the site grounds or environs following extensive measurements of site soil and outdoor surfaces in the Site Characterization Program (LILCO90a) and in the REMP program prior to, during, and subsequent to Shoreham operation. 8 The survey plan and procedures are designed accordingly, to focus primarily on remaining plant structures and systems in the affected areas. Instrumentation has been selected and measurement procedures developed to detect and measure surface contamination levels (primarily Co-60) and gamma exposure levels in these affected areas.

4.2 Instrumentation Radiation detection and measurement instrumentation for the termination survey has been selected to provide reliable operation with adequate sensitivity to demonstrate attainment of the release criteria. An evaluation has been conducted of instruments and detectors produced by several manufacturers. Detectors have been selected based upon detection sensitivity, operating characteristics and expected performance in the field under conditions of use. The detectors selected and their detection characteristics are summarized in this Section. Recording instruments (survey meters) for use with these detectors have also been evaluated. Instrumentation to be used for gamma exposure rate measurements and special purpose measurements is also described.

4.2.1 Instrument Description The principal instruments selected for termination survey measurements are identified in Table 4.1, Termination Survey Instrument Summary. The detectors used for total surface contamination monitoring are for the most part operated with data logging survey meters.

8 Trace amounts of Co-60, on the order of 0.1 pCi/gm have been detected in a sanitary sewage septic field distribution tank located on the owner controlled area of the site in 1993. This is above the Shoreham Station Radiological Environmental Monitoring Program (REMP) lower limit of detectability (LLD) for this isotope .

4-1

Table 4.1 i

Termination Survey Instrument Summary

• Information Withheld Per

$2.790(a)(4) 1 4-2 1

Tsrmination Survzy Plan Rev. 3 4.2.2 Detection Sensitivity The detection sensitivity of the detectors selected for termination survey measurements has been evaluated. These results are summarized in Table 4.2, Measurement Detection Sensitivities. Special detector configurations have been developed for termination survey of piping which is not readily accessible using conventional survey techniques. Table 4.3 summarizes the detector assemblies developed for survey of various sized piping. Most of these are multiple-GM detector assemblies.

For count rate measurements, the minimum detectable activities (MDA) shown in Tables 4.2 and 4.3 are calculated using the following equation:

^

(kfk ,) _5+B (4.1) MDA ccta = '

j '

E(100) where: k, = critical value at the upper, one-sided (100 - a)% confidence level (C.L.) (for normal statistics, k, = 1.645 at 95% C.L.)

k, = critical value at the lower, one-sided (100 - b)% confidence level $

(C.L.) (for normal statistics, k. = 1.645 at 95% C.L.)

S= Sample (plus background) count rate (cpm),

B= Background count rate (cpm),

l t, = Sample count time (minutes),

i t, = Background count time (minutes), and E = Instrument detection efficiency, counts per disintegration, and - .

A = Detector sensitive area (cm2).  ;

9 The MDAs are calculated using the above Equation 4.1 with the indicated values of k for the 95% confidence level and assuming the probability of Type 1 and Type 2 errors to be equal. These MDAs represent detection sensitivities under static conditions, one minute counts with fixed geometries. Detector efficiencies are empirically determined using sources traceable to the National Institute of Standards ,

and Technology (NIST). Beta-Gamma detectors are calibrated with Co-60 and alpha detectors with Pu-239. Nominal background values are used for the calculation, with i

30 minute background count times. During scanning type surveys, the detection sensitivity mav H. reduced. Therefore threshold or practical detection  ;

4-3

i Table 4.2 Measurement Detection Sensitivities Information Withheld Per

$2.790(a)(4) i l

l 1

4-4

l Termination Survey Plan Rev. 3 1 Table 4.3 l

Embedded Piping Detector Assembly Sensitivities l Piping Detector Bkgnd2 Efficiency Detection Dia. nominal Assembly 4x 2 Sensitivity (ID, in) (eff. area) 8 (area adjusted)' (dpm/100 cm2) 12 4 - FT126 GMs 350 cpm 0.058 210 2

(504 cm ) (0.295) 10 8 - HP260 GMs 165 cpm 0.166 210 2

(124 cm ) (0.206) 8 6 - HP260 GMs 130 cpm 0.152 270 2

(93 cm ) (0.141) 6 9 - HP260 GMs 210 cpm 0.148 230 2

(140 cm ) (0.207) 4 6 - HP260 GMs 165 cpm 0.134 350 2

(93 cm ) (0.124) 3 4 - HP260 GMs 80 cpm 0.156 310 2

(62 cm ) (0.097) 2 4 - TGM N1003 19 0.168 520 GMs (0.046) 2 (27.2 cm )

1.5 1 - HP190A 17 cpm 0.0055 2800 (end window)

Notes.1. Effective sensitive area is the total sensitive area of all detectors in the assembly.

Survey measurements are normally performed by summing the counts from all detectors. The assemblies can also be operated whereby the counts from each detector are recorded separately.

2. Nominal background values; 30 minute counts.

3. The 4 x efficiency is obtained from the combined detector response to a Co-60 flexible mylar source in an annular sleeve placed inside a piping spool piece. The l source active area is such that it subtends the active area of all detectors in the I detec:or assembly.

4. The area adjusted efficiency is the product of the 4 x efficiency and the ratio A/100, where A is the effective sensitive area of the detector assembly.

5. Efficiency obtained by detector response to an annular sleeve source with an effective area of 100 cm2.

l 4-5

Termination Survey Plan R:v. 3 limits have been empirically determined for each detector used in the scanning mode.

Scanning mode efficiencies were obtained via a series of measurements whereby the detector was passed over calibrated sources of various dimensions at an established scanning speed to determine an effective scanning " efficiency". The scanning MDAs were then calculated using equation 4.1 with the scanning efficiency substituted for E, the efficiency term. These results are shown in Table 4.2 for the various detectors used for scan surveys.

4.2.3 Calibration and Maintenance Instruments and detectors used in the Termination Survey are calibrated and maintained at Shoreham Nuclear Power Station according to approved procedures.

Detectors for surface beta-gamma measurements are calibrated using NIST traceable Co-60 sources. Specialized instruments, for example, the pressurized ion chamber, are calibrated by the vendors.

4.3 Survey Plan 4.3.1 Classification Each survey unit is classified into one of two categories which identifies each unit as "affected" or " unaffected". Units identified as affected have a possibility of containing residual contamination and those identified as unaffected have a very low probability of residual contamination. Classification of individual survey units is based on the site characterization study (LILC090a) and the history of radioactive materials involvement or potential for contamination of the survey unit. Criteria for classification of survey units are given in Appendix A.

4.3.2 Reference Grids  :

i Gridding consists of dividing areas to be surveyed into regular subdivisions for the j purposes of identifying survey locations and for use as guides for scanning surveys. j For surveys designed on the basis of prescribed sampling plans, whether random or  ;

systematic, gridding provides a means of selecting individual measurement locations.

Mapping is used to document measurement locations. Detailed guidance for gridding is contained in Appendix A. Figure 4.1 shows grid placement in a structural survey  !

unit classified as affected (an equipment room). Placement of grids in structures is  !

directed by a specific work instruction.

4.3.3 Grid Maps Grid maps are used for survey design and to document the measurement locations.

Figure 4.2, Shoreham Decommissioning Termination Survey Grid Map, shows a grid map of a structural survey unit. It shows the floor and walls up to two meters.

This map is identified as the base map for the survey unit. Additional grid maps may be prepared for subunits as needed to plan and document surveys of subunits.

4-6

Termination Surv:y Plan R:v. 3 4.3.4 Grid Numbering Grids are uniquely identified by an ID code or number. The numbering convention is to start at the reference location, usually the southwest corner, and proceed sequentially west to east numbering each row on the floor from south to north in a continuing sequence. Grids are numbered in sequence for each subunit similarly as described 4- the floor numbering. An individual grid block has a unique identifich. code as determined by its survey unit (or subunit) ID and the number of the grid block within that unit. The numbering convention is illustrated in Figure 4.2.

4.3.5 Survey Maps Survey maps are prepared to document the details of survey measurements in circumstances where grid maps are not practical. Survey maps are typically prepared to document detailed surveys of piping and system components.2 These may be prepared by the technician who performs the survey, or may be specially prepared by direction of Termination Survey Radiological Engineers for detailed surveys of complex components.

4.3.6 Surface Scans Surfaces are scanned according to the prescriptions in Appendix A. Scanning surveys are performed to screen large areas efficiently to search for areas above the average total surface contamination release criteria and to detect " hot spots", i.e.,

localized areas above the maximum total surface contamination release criteria. The scanning methods utilized (instrument and survey technique) are capable of detecting 75% of the average total surface contamination release criteria, e.g.,3750 dpm/100 cm2 for total surface beta-gamma contamination, as shown in Table 4.2. When scanning surveys indicate that contamination levels above the average total surface contamination release criteria may be present, appropriate followup investigation and/or measurements will be performed.

4.3.7 Surface Activity Measurements Surface activity measurements are taken at measurement locations selected in accorcance with the survey design guidelines in Appendix A. The general set of measurements is direct beta-gamma and removable beta-gamma. In areas and systems identified as alpha affected, direct surface and removable surface alpha measurements are taken.

It should be recalled that system component and equipment exterior surfaces are included in the survey of the structural survey unit in which they reside. The component interior surfaces are surveyed in the survey of the system survey unit to which the component belongs. Survey maps ofindividual components may thus be prepared for each type of survey. i 4-7 l 1

I Termination Surny Plan R;v. 3 j 4.3.8 Exposure Rate Measurements Gamma exposure rate measurements, when directed by survey design guidelines, are taken at one meter from surfaces at all measurement locations in stmetures and outdoor areas. In locations where it is not physically possible to locate an instrument one meter from the surface, gamma exposure rate measurements are not taken. The methods for gamma exposure rate measurements are discussed in Section 4.4.3 below.

4.3.9 Soil Sampling Soil samples will be collected in accordance with Section 7.2.5 of Appendix A.

Additional soil samples will be collected in the Termination Survey if a contamination event or spill occurs, or survey measurements indicate outdoor areas of elevated activity above applicable release criteria limits.

4.3.10 Special Sampling and Measurements

a. Sampling of Sediment and loose Material Samples of loose paint, dust or other sediment, tank bottoms, sewage sludge, radwaste media, concrete and other bulk materials are collected for laboratory analysis .a part of biased sampling and measurements. Such samples may be collected in drain receptacles, sumps, and other catchments in affected areas. Selected storm drain catchments may be sampled in accessible locations on the site. These samples are analyzed by gamma spectroscopy for Co-60. Those samples with detectable activity are quantitatively analyzed and the results compared to the applicable guideline value for soil and bulk materials.

b. Embedded Piping Surveys Measurements are taken to demonstrate that normally inaccessible piping, e.g., embedded piping, or runs of piping in confined pipe chases, is below the release limits for surface contamination. This technique involves the use of GM detectors, specially calibrated for direct surface measurements within piping interiors. The detectors are mounted in multiple-detector assemblies called " pipe crawlers" (see Section 8.0 Glossary). These assemblies are inserted into piping runs in a controlled manner and measurements of total surface activity are taken systematically over the length of piping.

4.4 Background Ievel Determination 4.4.1 General Requirements Backgrounds are established for each type of instrument to be used for surface contamination and gamma exposure rate measurements. Surface contamination 1

4-8 i

l Termination Surv2y Plan Rsv. 3 measurements include total surface beta-gamma and alpha, and removable surface beta-gamma and alpha. Gamma exposure rate measurements require determination of the gamma background response of detectors at one meter from surfaces. The  ;

background responses of the pressurized ion chamber and microrem instruments i must be determined. In addition, backgrounds are determined for specialized detectors and detector systems. These include: large area detectors for floor i monitoring and detectors for surveying piping interiors (multiple GM detectors).

4.4.2 Objectives of Background Determinations The objectives of background determinations for Shoreham Decommissioning l Termination Survey measurements are to:  ;

a. establish the reference background mean values for each type of detector -

used in the Survey;

b. assess the variability in background responses for principal detectors under different applications and conditions of use; and i

c. determine the need for correction factors or special measurements to i establish the background for Termination Survey measurements in specific .

locations. l 4.4.3 Background Measurements Several locations have been used to obtain measurement data for establishing reference backgrounds for each type of measurement. Principal criteria for i background measurement locations are: similarity to Shoreham facility construction l and free of non-natural radioactivity. One on-site building, the Colt Diesel Generator <

building, was selected due to its similarity in construction to the Reactor, Radwaste and Turbine Buildings. An off-site building, the Shoreham-Wading River fire house, a multistory reinforced concrete building was also selected. Collection of measurements for background determination is performed in accordance with an ,

approved station procedure. The methods for background determinations for each type of Termination Survey measurement are summarizal below.

a. Direct Surface Beta-gamma Measurements )

i To determine background for direct surface beta-gamma measurements, a series of counts of at least one minute duration are taken in sequence. The counts are accumulated by a scaler in the preset timc accumulation mode.

Outdoor background measurements are performed on types of surfaces where beta-gamma surface measurements may be taken in the Termination Survey.

These include concrete pads, loading docks, pavement, and roofs.

4-9 I

Termination Survey Plan R v. 3

b. Direct Surface Alpha Measurements The background response of alpha survey instruments is small in comparison with the response when contamination is present. Backgrounds for direct alpha measurements are collected as needed according to an approved procedure.

c. Removable Surface Beta-gamma Measurements Background determinations of beta-gamma smear counters are made by taking a count of a blank smear, usually on each day of operation.

Background determination is performed in accordance with an approved procedure.

d. Removable Surface Alpha Measurements The background count rate in smear counter alpha detectors is also very low.

Background for each counter is determined by counting a blank smear, in accordance with an approved procedure.

e. Gamma Exposure Rate Measurements A pressurized ion chamber is used to establish gamma exposure rate background at the Shoreham site. The background is used as the baseline for demonstrating that residual gamma exposure rate levels are below 5 R/hr above background (measured at one meter). The pressurized ion chamber is used as the reference instrument for establishing the gamma exposure rate background and the R equivalent response of portable microrem meters.

The latter are used for the bulk of the Termination Survey gamma exposure rate measurements. A series of paired measurements has been taken with the two instruments to develop the correlation.

A reference value for the gamma exposure rate has been established for the Shoreham site. It is based upon a series of pressurized ion chamber measurements taken one meter above the ground surface in outdoor areas and inside buildings on the site in areas demonstrated to be free of residual contamination from Shoreham operations. The value is 6 R/hr (5.8 i 1.2 R/hr - one standard deviation). This reference value is used for all survey units, except where actual background conditions are shown to vary significantly, such that area-specific background values are warranted to reduce the bias in reported population average 'above background" values.

f. Specialized Measurements It has been observed that detector background for direct beta-gamma l measurements is affected by conditions in the immediate vicinity of the '

4 - 10 l

i Termination Surv2y Plan Rev. 3 detector. Significant variations from background reference values have been <

observed. These variations are caused by the natural radioactivity composition of materials and by shielding effects in some cases. As a result, background measurements for special conditions have been compiled for use in calculating measurement net dpm/100 cm2 values to reduce the bias in survey unit population statistics. Special condition backgrounds have been compiled for example for: ceramic and clay tile materials, porcelain fixtures, poured concrete materials, roofimg materials, large bore piping, embedded piping and others. Guidance for application of special condition backgrounds i is provided in an approved procedure.

g. Verification of Background Measurement Population Each population of background measurements is analyzed using Equation 8-22 of draft NUREG/CR-5849 (BE92) to ensure that the number of measurements in the data set is adequate to characterize the background mean value to within i 20% at the 95% confidence level.

4.4.4 Documentation And Control of Background Measurements Background measurements are collected and recorded in accordance with a Termination Survey procedure. Background results for reference values and special condition backgrounds are compiled in a memorandum which is attached to the  ;

survey Final Report.

4.5 Sample Analysis As indicated in Section 3.5, an in-depth sample analysis capability is available for the Termination Survey. Routine samples of sediment, paint chips and debris will be qualitatively evaluated for the presence of Co-60 via gamma spectroscopy. The need for additional sampling and analysis will be determined on the basis of this initial evaluation.

1 4 - 11

Figure 4.1 View of Structural Survey dnit Showing Grid Placement

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Figure 4.2 Shoreham Decommissioning Termination Survey Grid Map 5

1 Information Withheld Per

$2.790(a)(4) i 4 - 13

Termination Sumy Plan Rev. 3 5.0 Data Interpretation All measurements are converted to the appropriate units for comparison with release criteria limit values. Surface activity measurements are converted to units of dpm per 100 cm2 . Gamma dose rate measurements are converted to exposure rate units of R/hr and the background value subtracted to obtain the net exposure rate. These calculations are performed using equations given in Section 8.00f draft NUREG/CR-5849 (BE92), indicated below. Additional calculations are made to determine a " critical value" for both total and removable surface activity measurements, to adjust these total and removable surface activity measurements for the presence of the undetected Fe-55 and other hard-to-measure radionuclides where appropriate, and to determine the measurement specific " action level" which indicates the need for additional investigations and/or measurements. Average values for each survey unit (and/or subunit) are compared with the release criteria values.

Confidence intervals are estimated for mean values of each survey unit (and/or subunit) at the 95% confidence level. Figures 5.1 through 5.5 describe the flowpaths followed to convert, calculate, analyze and interpret the data for the four (4) basic types of survey measurements made for the termination survey.

5.1 Conversion of Measurements to Reporting Units 5.1.1 Direct Measurements - Total Surface Activity Measurements of total surface contamination activity are converted from observed gross counts per minute to net activity concentration by subtracting the background counting rate for the instrument and correcting the net count 2

rate for geometry and efficiency to obtain results in dpm/100 cm units. Total surface activity measurement results are reviewed to ensure that the applied background values are appropriate, i.e.,not "too high" so as not to potentially

" mask" areas of contamination.

The following calculations and conversions are applicable to measurements of total surface activity:

a. Unit Conversion RCPm - bcpm dpm ad ,

(5.1) A E (100) where:

dpmo,, = total surface activity dpm/100 cm2, gepm = detector reading in gross counts per minute (assumed to be entirely due to Co-60 activity),

bepm = detector background in counts per minute, 5-1

Termination Survey Plan Rev. 3 E = detector efficiency in counts per disintegration, and A = area of detector sensitive region (cm2),

b. Calculation of "Criticallevel",L (in units of dpm/100 cm )

(5.2)

L, = l % }s, + s j

E (100) where:

Le = Critical Level, defining measurements above normal background distribution, and s, = counting error in sample measurement, or s#

• =E g2 where:

c = measurement total counts (sample plus background), and t = measurement count time.

sb

counting error in background measurement, or s>

tg where:

B = total background counts, and t3 = background count time.

E = detector efficiency in counts per disintegration, A = area of detector sensitive region (cm2),and 5-2

Termination Survey Plan Rev. 3 1.96 = 97* percentile value of a one-tailed normal distribution.

c. Adjustment to Account for Hard-To-Measure Radionuclides Measurements converted using Equation 5.1 which exceed the " critical level" as determined by Equation 5.2 are reported as having surface activity above background levels. This activity is assumed to result from residual Co-60 contamination. Therefore, adjustment for the undetected presence of Fe-55, Tritium and other hard-to-measure (HTM) nuclides is made as follows:

(5.3) dpmg - f (dpm,)

I where:

dpm 4 = surface activity concentration adjusted to include Fe-55 and other hard-to-measure species (in dpm/100 cm 2), and dpmo,, = surface activity concentration from Equation 5.1 (in dpm/100 cm 2), and f = numerical factor; (see Table 5.1 for factors).

Table 5.1 Adjustment Factors for Surface Activity Measurements Area - Description Adjustment Factor, f Reactor Bioshield Concrete 35.9 Reactor Bioshield Concrete Removable Activity 12.4 Reactor Bioshield Steel 24.2 Reactor Bioshield Steel Removable Activity 1.2 Dry Cutting Station 2.6 Spent Fuel Pool 2.3 Reactor Vessel Bottom Head 2.6 All Other Areas 1.2 _

The numerical factors, fin Table 5.1 are described in Section 3.2.

5-3

Termmation Surv;y Plan Rzv. 3

d. Calculation of " Action Level",LR(in units of dpm/100 cmf)

(5.4) La = 3 lsl j+ sl  ;

E (100) where:

La = Action Level, defining those measurements considered significantly above background (and assumed to include Fe-55 and possibly other HTM radionuclides). In survey units (or subunits) classified as " Unaffected Areas", values above this level require investigation, including additional measurements and possible reclassification of the survey unit (or subunit).

s,, sb , E, and A are as defined for Equation 5.2,and 3 = 99.7* percentile value of the normal distribution. 3 5.1.2 Removable Contamination Measurements Measurements of removable surface activity are converted from gross count rate to units of net dpm/100 cm 2by subtracting the background count rate of the smear counting detector and correcting the net count rate for detector geometry and efficiency.

The following calculations and conversions are appropriate to measurements of removable surface activity:

a. Unit Conversion - determined by using Equation 5.1.

b. Calculation of " Critical Level",L -c determined by using Equation 5.2.

c. Adjustment for Fe-55 and other Hard-to-Measure Radionuclides l Similar to the treatment of total surface activity measurements, ,

measurements of removable surface activity which exceed the " critical I

The three (3) sigma coefficient represents the 99.7* percentile of the standard normal distribution (two-tailed), taken from NUREG/CR-2082, " Monitoring for -

Compliance with Decommissioning Termination Survey Criteria", p.132. 1 i

5-4  ;

Termination Survey Plan Rev. 3 level"are adjusted to include the undetected Fe-55,and possibly other HTM activity, using Equation 5.3.

d. Calculation of " Action Level", LR For removable surface activity measurements, the " action level",LR ,is established at 250 dpm/100 cm 2for all measurements. This value corresponds to 25 percent of the removable residual contamination release criterion as specified in Table 3.1. In survey units (or subunits) classified as " Unaffected Areas", measurements above this level require additional investigation, including additional measurements and possible reclassification of the survey unit (or subunit).

5.1.3 Gamma Exposure Rate Measurements Most gamma exposure rate measurements are taken with microrem meters.

A correction factor is applied to each reading to convert to units of pR/hr, as discussed in paragrapn 4.4.3.e. The gamma exposure rate background is subtracted from the gross pR/hr value to obtain net gamma exposure rate in units of pR/hr. Gamma exposure rate measurement results are reviewed to ensure that the applied background values are appropriate, i.e.,not "too high" so as not to potentially " mask" areas of elevated gamma exposure rates.

The following calculations and conversion are applicable to gamma exposure rate measurements:

a. Unit Conversion

• E (5.5) pR hr - 3.06 + 1.07-( Ehr ) -hr( E )g where:

pR/hr = net gamma exposure rate in micro R per hour, pRem/hr = measured gross micro rem per hour, l

(pR/hr) Ms = background exposure rate in R/hr, 3.06 = constant term of the correlation factor, and  ;

1.07 = proportional term of the correlation factor determined as described in paragraph 4.4.3.e.

l 5-5 l

Termination Survey Plan Rev. 3

b. Calculation of " Action Ixvel", LR For gamma exposure rate measurements, the " action level", LR , is established at 5 pR/hr above background for all measurements.

Measurements above this level require additional investigation, including additional measurements, and a calculation to demonstrate the exposure rate averaged over a 10 m2 area centered on the measurement location exceeding LR does not exceed 5 R/hr above background.

5.2 Comparison With Release Criteria Limits The method outlined below will be used to demonstrate attainment of the release criteria limits.

5.2.1 Attainment of Release Criteria for Surface Contamination

a. Total Surface Activity (fixed plus removable contamination)

Individual measurements: Do not exceed elevated area gross activity guideline value (contaminated area not to exceed 2

100 cm ). See Table 5.2 for values.

Iocal area average: Does not exceed gross activity guideline value averaged over an area of one m 2. See Table 5.2 for values.

2 Population - random sampling: Upper limit of confidence interval (Equation 5.6) for the mean value is below gross activity guideline value. See Table 5.2 for values.

2 Population - biased sampling: Upper limit of confidence interval (Equation 5.6) for the mean value is below gross activity guideline value. See Table 5.2 for values.

2 A population for Termination Survey purposes, refers to a survey unit (or subunit if so specified). More precisely, a population represents the collection of all possible values of a parameter, e.g., total surface contamination, being measured through a sample ofits members.

5-6

Temanation Survzy Plan Rev. 3

b. Removable Surface Contamination Individual measurements: Do not exceed 1000 dpm/100 cm2, Population - random sampling: Upper limit of confidence interval (Equation 5.6) for the mean value 2

is below 1000 dpm/100 cm ,

Population - biased sampling: Upper limit of confidence interval (Equation 5.6) for the mean value 2

is below 1000 dpm/ 100 cm ,

c. Calculation of Upper Limit of Confidence Interval (5.6) u, = x + t where:

u, = upper confidence limit of population mean, and x = population mean value t 1 -a, df = upper confidence level value (as obtained from Appendix B, Table B-1,of draft NUREG/CR-5849); df (degrees of freedom) is equal to n - I and "a"is the false positive probability.

sx = population standard deviation n = number of measurements in the population 5.2.2 Attainment of 5 microR per hour Criterion Individual measurements: Net exposure rate does not exceed 10 R/hr.

Local area average: The the net exposure rate does not exceed 5 gR/hr when averaged over an area of 10 m

Population - random sampling: Upper limit of the confidence interval Equation 5.6) for the mean net exposure rate does not exceed 5 pR/hr.

5-7

Termination Survey Plan R;v. 3 Population - biased sampling: Upper limit of the confidence interval (Equation 5.6) for the mean net exposure rate does not exceed 5 pR/hr, 5.2.3 Evaluation of Soil Sample Results Since soil contamination from Shoreham operations has not been detected in the Site Characterization Study (LILCO90a), or in the REMP Program, it is believed to be unlikely that detectable contamination will be found in the soil column on the site in the Termination Survey. As there is no limit or criterion for residual contamination in soil for release of the site for unrestricted use, soil radioactivity concentration measurements will be compared to REMP MDAs and historical REMP measurement result ranges for individual isotopes. For gamma emitters where no REMP data exists, as in the case of Co-60 in soil, the current NRC criterion is adopted and results below 8 pCi/gm average concentration in the top 15 cm of the soil column will be considered to be acceptable. 3 5.3. Calculation of Gross Activity Guideline Values for Total Surface Activity When mixtures ofradionuclides are involved whose members have different guideline values, the gross activity guideline value (GAG) is substituted for the guideline value.

The gross activity guideline values are obtained from the " sum of fractions rule" in Appendix A of NUREG/CR-5849 (BE92), whereby the sum of the ratios of each radionuclide to its guideline value must be s 1.The resulting formula for the GAG is:

GAG =

I (5.7) Fi F2 F,

- + - +.... -

G, G2 G, where:

GAG = Gross Activity Guideline Value in dpm/100 cm2, F; = fraction of total activity due to the ith radionuclide, G i= guideline value for the ith radionuclide, in dpm/100 cm 2.The guideline values for principal radionuclides at Shoreham are: Co-60 = 5000; Fe-55 = 200,000; H-3

= 200,000dpm/100 cm 2(USNRC94).

3 The value of 8 pCi/gm in soil is applied to the total concentration of Shoreham-produced gamma-emitting radionuclides.

5-8

Termination Surv:y Plan Rev. 3 The elevated area gross activity guideline value is obtained as the product: 3 times the gross activity guideline value. The calculated gross activity guideline values applicable to the various areas are given in Table 5.2.

Table 5.2 Gross Activity Guideline Values Area Gross Activity Elevated Area Guideline Value Gross Activity Guideline Value Bioshield Concrete 95,900 287,700 Bioshield Steel 76,900 230,700 Dry Cutting Station 11,100 33,300 Spent Fuel Pool 9,400 28,200 Vessel Bottom Head 11,100 33,300 All Other Areas (1) 5,000 15,000 Table 5.2 Note: (1) The gross activity guideline and elevated area gross activity guideline values for all other areas are calculated to be 5,900 and 17,700dpm/100 cm 2,respectively when the special limits for Fe-55 and Tritium are applied. However, the standard Regulatory Guide 1.86 guideline values of 5,000 and 15,000 dpm/100 cm are used to maintain consistency with earlier results and to provide an additional degree of conservatism. ,

l l

l i

l l

5-9

Figure 5-1 DATA REVIEW FLOWPATH FOR TOTAL SURFACE ACTIVITY-AFFECTED SURVEY UNITS

+

Convert date to dpre /100 em' l

Coloulate ertueel level, Lo, for eneh mesourement Perform NO L.

ed..on.1 ,

surveye A VES y

Adjust for Fe 55 1 Memediate h Compare to Devated

^ Aree guidelines

( 15000 d /100ont ,

Adjust for Fe45 NO 1 and Tetsum se Compare to guideline

( 5000 dpm /100 om' )

Ust all 2 measuremente above La YES 5000 dpm /100pm' NO O2 >

y

, y C.i.ui..on of i-

- -ero.e invoe8pelo bee %round; , ppg , g, Colleet addisonef date 9%,

(ifnoseeeery)

Ust all meneuremente Coloulets looel eroe 4 above guideline value everage over 1 m' 3 Avs.

s YES  ?

5000 dpm /100em' NO b4 >

T y Celeutete summary statio5ee, S.D., C.L SU NO sritoria Colleet addl6onal  ?

date ; reeeleulate YES SU NO moete YES

( release >

ertterte Y 7 Propero Final Data Report ,

5 - 10

" O##

eATA. AAu M

Figure 5 -2 DATA REVIEW FLOWPATH FOR TOTAL SURFACE ACTIVITY-UNAFFECTED SURVEY UNIT Convert data to dpm /100 em' t

Coloulate ortueel level, La, ter eesh meneurement

~

NO messi,,, rom r- e , e,,d r to I, ury.,

y se Affected survey unit .

A YE8 V

Adjust for Fe.55 g Denne extent of eree .

h Estabilah new outenit eh I I realeselRoegon A estion level La Y

measurement to L e YES Le NO b t e NOTES:

y Adjust for Fe46 1 and Tridum se Cosmpers to appropriate i guideline value YES NO

~

UM dl

( ED e 2 measuromonte 100 om above Le I

V Evoluete beolypround and meneurements; sollect addidonal date (if necessary);

reseloulete YES NO s t y Coloulete stede5es :

Avg. , S.D. , C.L SU NO

( release ortlerie

?

YES V

Propero Final Data Floport 5 - 11 em o,e,e, ear.uw me

- _ . . .- ~ - . .-

Figure 5 -3 i DATA REVIEW FLOWPATH FOR l REMOVABLE SURFACE ACTIVITY  !

+  !

Convert dets to  :

dpse /100 em s j

Celeutete artesel level. Lo, ter eneh  !

mesaurement  !

fleele8844. remoanets ,

(If nemenery), no roeurvey as AfInsted Le survey unit  ? +

t E  !

ir osane entent et erse, s-.m ,, i estabiloh near sub unit l * ~ * * ,

If nemoneery n -e .

NO Affected YES l

S. U. {

7 i k

if if A Compem to Compere to guideline realeseineation l estion level f

i

~ ~

NO YES 380 elpm /100em' 1000 alpm /100em'

?

{

?  ;

if ir NO inveengets, obtain

"** ae' de*

inves.gets, eensei l edemonal dem >

(if namesary)

{

YES y 1000 dpee /100am' >  ;

?  !

I

• "O if h/

( ss0 M 100eme y< NO  !

?

if f statsees. UCL f

Su  ;

seesto i g

telenes > E N  !

T j NOTE:

y YE8 Must ter M '

and Trtsum ao Prepare final data report 1

1 5 - 12  !

nm ene<= , , , . . ,

i

i Figure 5 -4 1 DATA REVIEW FLOWPATH FOR GAMMA EXPOSURE RATE MEASUREMENTS v

Convert siste to not M/hr Perform addisonet l

-- Core- .e A elevated eron guideline Remedlete b 1 YES to e /hr

?

NO Compare to guideline value e

1 YES A NO 1 s m /hr

?

NOTES:

V investigate ' 'ng List all Colleet addluonal date 1 measuremente above guidelines h 9 Celeulets everage exposure vote over 10 m :

Avg.

1 NO m

< YES s M /hr '

? V Celeutete ,

Avg.,UCL UCL Colloot addluonal '

YES 1 anste, reenleulate s M/hr

?

UCL -

, YES 1 NO t

^ r s A /hr

?

V Release criteria satisfied ,

Prepare final report l

5 - 13 l e., ea,

w. - -

Figure 5 -5 DATA REVIEW FLOWPATH FOR SOIL SAMPLES

+ t

~

• t Individual Soil Samples LLD's ~

l so ne uns,

@ idenefy non-bookground isotop > un go Above YEs Collect additional LLD samples

?

A "O

Co-80 above Remediate if necessary l YES 3r A Adjust for Fe 65,

, Add Fe-55 amount i

h( '

k Sum all isotopes and compare to j 8 pCi/ gm t

hf s

t um YES beloW NO t  !

8 pCI/gm r if 7 Calculate Avg, UCL for all soll samples l

UCL below NO SpClIgm 7 7

e un. ns.

> vEs *- a * * ***a hI @ cosnpees own ne e pcygm ressessenaeuen essen nevei.

Prepare data report O **'""*" e.e*so, uci.,--u r som enmpung owver unne, other son seenpies evekssted 5 -14 mww mad e,.ad w indevidual s.U.

End 07107/94 gg

Termination Surv2y Plan Rev. 3 6.0 Final Report i Upon completion of each major phase of the Termination Survey,an updated Final Report f will be prepared for submission to the Nuclear Regulatory Commission. This report will meet the intent of Regulatory Guide 1.86(USAEC74) for final survey reporting. The report l' will follow the guidance of draft NUREG/CR-5849 (BE92) regarding content.

6.1 Topical Outline  ;

i The Final Report willaddress the following topics. The report will provide adequate l

data and discussion of each topic to meet the intent of NUREG/CR-5849 (BE92). 4 The following describes the format for the Final Report with regards to document volumes, topical outline and content: i Volume 1 l

1.0 Background Information 1.1 Reason for Decommissioning t 1.2 Management Approach and Organization  !

[

2.0 Site Description  ;

2.1 Type and Lxx:ation of Facility  ;

2.2 Ownership i 2.3 Facility Grounds - Survey Scope 2.4 Facility Structures  !

2.5 Plant Systems ,

2.6 Outdoor Areas  !

l 3.0 Operating History f 3.1 Licensing and Operation 3.2 Processes Performed  ;

3.3 Waste Disposal History and Practices j 4.0 Decommissioning and Supporting Activities >

4.1 Decommissioning Objectives  :

4.2 Site Characterization  !

4.3 Radiological Environmental Monitioring Program  !

f (REMP) 4.4 Radiological Effluent Reports l 4.5 Decontamination and Dismantlement Activities  !

i 5.0 Termination Survey Methodology l 5.1 Sampling Parameters 5.2 Background Levels 5.3 Major Contaminants ,

5.4 Guidelines Established f

?

6-I

i i

I Termmation Survey Plan Rev. 3 j l

5.5 Equipment and Techniques Applied 5.6 Survey Process 5.7 Survey Controls 5.8 Data Analysis 6.0 Termination Survey Results 6.1 Survey Results

6.2 Findings

6.3 Final Configuration 7.0 References 8.0 Glossary Volume 2,(Multiple Books)

Survey Unit Release Records

1. Tabulated Results for Individual Survey Units

2. QC Replicate Surveys

3. Survey Maps Volume 3 Termination Survey Program Description Termination Survey Plan Implementing Procedures Technical Memoranda and Information Volume 4,(Multiple Books)

Supporting Documentation

1. Nuclear Quality Assurance Surveillance Reports

2. Nuclear Quality Assurance Audits

3. LIPA Deficiency Reports As discussed in Section 3.8, an updated Final Report will be submitted upon completion of the each phase of the termination survey. The first of these reports contains the four volumes, as described above, with books included for Volumes 2 and 4 to detail the surveys completed. The second and subsequent submittals of the Final Report consist of updates to Volumes 1 and 3, and additional books to be included in Volumes 2 and 4.

The Final Report provides information which substantiates the survey findings and conclusions. Such information includes, but is not limited to: Reports of Nuclear Quality Assurance Department (NQAD) audits and surveillances of the termination 6-2 i

Tenmaation Survsy Plan Rev. 3 survey, quality control (QC) survey results, and survey unit release records (including applicable survey maps).

I 6.2 Reporting of Survey Findings 6.2.1 Summary Measurement results are reponed at several levels of detail. An overall summary of the measurement results and conclusion that the facility meets the release criteria is provided. A tabular data summary shows the results hr each  :

i major category of survey unit: structures, outdoor areas and plant systems. '

This tabulation identifies the number of survey units in each category, the ' '

maximum individual measurement values and maximum survey unit UCLs for the measurements of each type: total surface beta-gamma, total surface alpha,  ;

removable surface beta-gamma and removable surface alpha activity concentration, and gamma exposure rate. A more detailed summary table, is provided which shows the following for each survey unit: number of  ;

measurements, maximum, and UCL for each type of measurement. _ Maximum ,

individual measurement values and upper limit of confidence intervals about ,

the mean (at the 95% confidence level) in units of dpm/100 cm2 are reported ,

for comparison to the release criteria surface activity limits in Table 3.1. A j summary of gamma exposure rate measurements is presented by similar  !

treatment, showing that maximum value is less than 10 pR/hr above -l background and the mean value (UCL) for each survey unit is less than 5  ;

microR/hr above background. The results for soil sampling are similarly l presented. l 6.2.2 Summary Data Reporting for Each Survey Unit Within the release record for each survey unit or subunit, the number of-measurements and the upper limit of the confidence interval about the mean  !

(at the 95 % confidence level) are reported in tabular form. These are 2  ;

reported in units of dpm/100 cm for each type of measurement: total' surface -

beta-gamma, total surface alpha, removable surface beta-gamma and 1 removable surface alpha activity concentration. Gamma exposure rate i measurement results are reported showing the number of measurements and l upper limit of the confidence interval about the mean (at the 95 % confidence

~

level) for each survey unit (and/or subunit). The release limit value for each- ,

type of measurement is provided in the release record. ,

i Results of sampling measurements, e.g., sediment, paint, concrete, other i debris, are reponed in the release record for each survey unit. 1 Any ponions of individual survey units found not to meet any of the various  !

release criteria willnecessarily be remediated. The Final Report willidentify  !

all areas within the survey which required reclassification and/or remediation. l l

6-3

Ternunation Survey Plan Rsv. 3 For each of these areas, the initiating survey results, investigation, corrective actions, and followup survey results will be provided.  ;

6.2.3 Detailed Data Reporting  !

The result of each measurement taken in the termination survey are tabulated in the individual release record for each. survey unit. Table 6.1 shows an example of a typical format for individual measurement tabulations for a  :

structural survey unit. Results of alpha activity measurements are reported in a similar format for those survey units where alpha measurements are taken. .

Tables of similar format are used to report data for systems and outdoor ,

survey units. Final data reports provide indication ofindividual measurements ,

which exceed the " critical level", calculated as described in Section 5.0,and -

the adjustment for Fe-55 activity, where applicable. Furthermore, the release records for survey units classified as " unaffected" provide a listing ofindividual measurements which exceed the " action level", calculated as described in Section 5.0,and the results of the investigation of these measurements.

a

.i I

f i

6-4 i

Termination Survey Plan Rev. 3 Table 6.1 Termination Survey Detail Data Report page No. 1 Date: 04/07/93 SHOREHAM DECOMMISSIONING PROJECT Termination Survey Data Report survey Unit ID. :SUO68 System Code: W23 Survey date 01/29/93 Name: CHLORINATION Building: MULTI BUILDING SYSTEM total surface activity (beta-gamma)

_____l____l_______; ______l__________l_____l___ ______- _____ _____

pointigcpmlbk_ cpm l eff ldpm/100cm2l Lc j>Lc +Fe55 Lr >Lr 1l 33 22 '

O.017 "

647 ll 855 1309 5l 23 22 0.017 58 '

773 1184 6l 26 i 22 0.017 n 235 798 1223 7l 31) 22 0.017 "

529 839 1285 11 22 22 0.017 0l 764 1171 ,

12 22 22 0.017 i 0l 764 1171 '

13 29. 22 0.017 411 823 l 1260 17 16l 22 0.017 -352 710 1088 18 29' 22 0.017 411 823 1260 19 21 22 1 0.017 -58 756 1157 23: 29 22 0.017 411 823 .1260 24' 15 22 0.017 -411 701 ,1073 25 27 22 0.017 294 807 1235 29 24 22 0.017 117 781 1197 30 '

17 , 22 0.017 -294 720 1102 31 22 22 0.017 0 , 764 1171 35 27 22 0.017 i 294 '

807 1235 36 41' 22 0.017 '

1117 915 X 1330 1401 37' 28 22 0.017 352 815 1248 41 15 22 0.017 -411 701 1073 42 24 22 0.017 117 781 1197 43- 20 22 0.017 -117 747 1144 47 21 22 0.017 -58 756 , 1157 48, 32 22 0.017 588 847 l l1297 6-5 i l

l l

2 Termination Survey Plan Rzv. 3 7.0 References BE92, J. Berger, " Manual for Conducting Radiological Surveys in Support of License Termination", NUREG/CR-5849 Draft, June,1992.

LILC090, long Island Lighting Company, " Updated Safety Analysis Report - Shoreham Nuclear Power Station," Docket No. 50-322, Rev. 4, December 1990.

LILCO90a, leng Island Lighting Company, "Shoreham Nuclear Power Station Site Characterization Program Final Report," May 1990 (Addendum 1, June 1990; Addendum 2, October 1990; Addendum 3, June 1992).

LIPA90, long Island Power Authority, "Shoreham Nuclear Power Station Decommissioning Plan," NRC Docket No. 50-322, December 1990, as supplemented.

LIPA92, Long Island Power Authority, LIPA Nuclear Management Control Manual,

" Nuclear Organization Management Control Program for Decommissioning Termination Survey Program Description," PDXOM-01, February 1992.

LIPA92a, Long Island Power Authority, Quality Assurance Manual, Appendix N,

" Decommissioning Activities," June, 15, 1992.

MA90, F.H.C Marriott, "A Dictionary of Statistical Terms", Longman Scientific and Technical,1990.

NCRP85, National Council on Radiation Protection and Measurements, "A Handbook of Radioactivity Measurement Procedures," NCRP Report No. 58, Feb.1985.

SCI 94, Scientech Inc, " Radioactive Sample Analysis Report - Shoreham SFSP Paper l Filter", May 27,1994.

TU92, M. Tucker, " Technical Report on Radiochemical Analytical Results, Radionuclides, and Analytical Methods Affecting Termination Survey Plan l Development," July 2,1992. i I

4 USAEC74, U. S. Atomic Energy Commission, Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors," June 1974. j l

USNRC92, " Order Approving the Decommissioning Plan and Authorizing l Decommissioning of Shoreham Nuclear Power Station, Unit 1," Docket No. 50-322, i June 11,1992.

USNRC94, U. S. Nuclear Regulatory Commission, Letter from C. Pittiglio to A. Bortz, )

" Approval of a Modification of Facility Release Criteria for Tritium and Iron-55 Surface i Contamination at Shoreham Nuclear Station Unit 1", Docket No. 50-322, June 7,1994.

1 1

7-1 I

I

Terminatio 3 Surv y Plan Rev. 3 8.0 Glossary Action Ixvel - A contamination level, denoted L, (in dpm/100 cm 2, adjusted for Fe-55),

used as the investigation threshold in unaffected survey units to evaluate the need for reclassification as affected. The action level is defined as:

Removable surface contamination measurements: 250 dpm/100 cm2, Total surface contamination measurements: the upper confidence limit at the 99.7

% confidence level, (3 sigma), of the net count distribution.

soil contamination: 6 pCi/gm.

Affected Area - A designation assigned to a survey unit which ino; cates that the survey unit has a potential for containing residual radioactive contamination.

Alpha Affected Area - A designation assigned to a survey unit which indicates that the potential exists for alpha contamination.

Biased Sample - A method of selecting survey measurement locations which incorporates a non-random error, i.e., a method which selectively chooses locations for measurements which have a higher probability of contamination than those locations not selected.

Characterization Survey - A radiological survey and supporting evaluations performed to establish the Shoreham Facility baseline radiological condition for planning decommissioning activities. The Characterization Survey activities are described in and controlled by the Site Characterization Program Description.

Component - An individual equipment item, e.g., a valve, pump, tank, motor, etc. which is identified in the Shoreham Composite Component List.

Composite Component List (CCD - A controlled listing of components installed at Shoreham. The CCL identifies all components other than cable tray and conduit supports in the SNPS design. The CCL also defines which of these comp <ments are safety related, which were formerly safety related, as stated in the USAR, an? 'hich are not.

Confidence Interval - A range of values derived from a sample such that there is a probability a, that a population parameter being estimated, e.g. a mean value, lies within the range.

Confidence 12 vel - The probability a, associated with a confidence interval which expresses the probability that the confidence interval contains the population parameter value being estimated (MA90).

l 8-1 l

l

Termination Surv:y Plan Rsv. 3 Critical Izvel - A calculated value, used as a decision level to determine when a surface contamination measurement is above background, i.e., is due to contamination. It is defined as the upper confidence limit, at the 95 % confidence level, (1.96 sigma) of the observed net count distribution whose mean value value is zero (converted to dpm/100 2

cm ).

DECON - The decommissioning alternative which involves prompt removal of radioactive materials to achieve residual contamination and radiation levels which are below limits estab?ished to permit the facility to be released for unrestricted use.

Direct Measurement - A radiological survey measurement performed by holding a detector stationary on or close to the surface and recording the response.

Eevated Area Guideline - The value (net value above background) which individual measurements may not exceed under any conditions. For total surface contamination the elevated area guideline is 15,000 dpm/100 cm 2. For gamma exposure rate measurements, it is 10 pR/hr. Elevated area guidelines have not been established for removable surface contamination.

Eevated Area Gross Activity Guideline - Equivalent to the elevated area guideline, but applied to cases where the gross activity guideline applies. It is obtained as the product:

3 times the gross acdvity guideline. i Fixed Point Measurement - A synonym for direct surface contamination measurement.

History File - A compilation ofinformation prepared for use in planning the termination survey of a survey unit. It summarizes the operational history, characterization survey data, operational surveys and other information to help establish the basis for the design of the termination survey.

Gross Activity Guideline - The guideline which applies when multiple radionuclides are present which have different guideline values. The gross activity guideline value is calculated using the sum of fractions rule in NUREG/CR-5849 Appendix A as shown in Equation 5.7, Section 5.3 of this document.

Guideline value - The principal numerical limits in the facility release criteria. The guideline values are:

total surface beta-gamma contamination: 5000 dpm/100 cm2 ,

removable surface beta-gamma contamination: 1000 dpm/100 cm2 ,

total surface alpha contamination: 5000 dpm/100 cm2 ,

removable surface alpha contamination: 1000 dpm/100 cm2, gamma exposure rate: 5 pR/hr, and 8-2

~

I Tzrmination Surv7y Plan R v. 3  !

soil contamination: 8 pCi/gm.  !

These limit values are expressed as net values, above the background values for the measuring instruments used.

NOTE: Under certain conditions, individual total surface contamination and gamma exposure rate measurements may exceed the guideline values, as long as they do not exceed the elevated area guideline values and local area average values do not exceed the guideline values. ,

Maintenance Work Request (MWR) - A form used at Shoreham Nuclear Power Station, '

controlled by a station procedure, to initiate and track work activities.

NRC - U. S. Nuclear Regulatory Commission Operational Radiological Survey - A radiological survey performed under Shoreham Health Physics procedures. Operational surveys are distinct from, and usually performed prior to, termination surveys.

Outdoor Area - A category of survey units which includes site grounds, outside surfaces of buildings and small structures located out of doors.

Pioe Crawler - A term used in the Shoreham Decommissioning Project to denote a mechanical device equipped with multiple-detector assemblies (up to nine GM detectors) used to take direct surface radioactivity measurements of piping interior surfaces. The number of detectors used depends on the pipe diameter. The crawlers are manually inserted and maneuvered through piping with flexible fibreglass push / pull rods.

Plant Structures - All Shoreham Nuclear Power Station site buildings and their surfaces (generally identified as civil structures). For purposes of the termination survey, all

-tructures such as platforms, restraints, supports and other physical items not identified in the system MFSK drawings are considered to be structures. External surfaces of piping systems, heating and ventilation systems, tanks, stacks, etc., are also treated as structures in the termination survey.

Pooulation - A colledon of all possible values of a radiological parameter being measured in the termination survey. A survey unit (or subunit) is considered to be a population for purposes of drawing inferences regarding the value of a parameter, such as contamination level mean value in the entire survey unit (or subunit), based upon a sample of measured values.

Power Block - The group of major buildings on the SNPS site directly associated with electrical power generation. This group consists of the Reactor, Turbine, Radwaste, and Control buildings.

Process System Index - A listing, controlled by a station procedure, which identifies and assigns a unique identification code to each plant system.

8-3

Tmnination Surv2y Plan Rev. 3 OC Reolicate Survey - A radiological survey which consists of repeat measurements at j a specified fraction of the survey measurement locations in a survey unit, usually selected l at random, to provide an independent check of termination survey measurements. l i

Random Samole - In survey design, a method for selection of measurement locations I whereby each of the individual locations defined in the sample space has an equal probability of being selected. Related terms are: random selection and randomly selected.

Release Criteria - A term used to identify the radiological requirements for release of the Shoreham facility for unrestricted use. These requirements, which consist of specified limits for residual contamination and radiation levels, are specified in the Shoreham Decommissioning Plan.

Release Record - A document compiled for each survey unit (structure, system or outdoor area) which demonstrates that it is suitable for unrestricted use. It contains evaluated survey data and supporting information to provide a concise record of the results and basis for the conclusion that the release criteria are satisfied.

Reportine Units - The units in which each type of survey measurement is expressed for comparison to release criteria limits. For surface contamination measurements the reporting units are dpm/100 cm 2and for gamma exposure rate measurements the units are R/hr.

Scan Survey - A qualitative radiological monitoring technique which is performed by moving a detector over a surface (typically within one cm of the surface) at a specified constant speed to detect elevated contamination or radiation levels. Similar terms applied to this technique are: Scan and Surface Scan.

Site Characterization Report - A report (including addenda) which documents the surveys, calculations and evaluations and presents the results of the SNPS Site Characterization Program.

Subunit - A subunit, as used in survey design, is a subdivision of a complex survey unit that incorporates a structure, on item of equipment, or some other feature in order to establish that an appropriate number of survey measurements be made within the subunit, as well as within the survey unit.

Survey Desien - The process of determining the type, location, number and frequency (or density) of radiological measurements to be taken in the termination survey.

Survey Desien Guidelines - Criteria established to provide the appropriate level of survey intensity for systems, structures and outdoor areas, based upon their classification.

Survey Instructions - Written directions which specify the type and number of measurements to be taken in a survey unit. The survey instructions are in a standard format on forms controlled by a Termination Survey procedure. Each survey package includes survey instructions.

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Termination Survzy Plan Rev. 3 Survey Location - A discrete area or subdivision of a survey unit that is smaller than a subunit but larger than a survey point. In survey design, a survey unit (or subunit) is divided into a collection of survey locations. Specific locations are selectal in accordance with the design guidelines based upon the type and classification of the survey unit. In a structural or outdoor survey unit, a location is usually represented by a single grid block. In a system survey unit, a specified length of piping or a component such as a valve is referred to as a survey location. A survey location can contain one or more survey points.

Survey Package - A collection ofinformation in a standardized format for controlling and documenting field measurements taken for the termination survey. A survey package is prepared for each Survey Unit. The survey package includes the survey instructions, a control form, grid map (s), survey measurement data sheets and survey maps.

Survey Point - A smaller subdivision within an area designated as a survey location (grid block, system component) where local measurements are taken, generally referring to an area covered by a detector, or an area of 100 cm2 when a smear is taken.

Survey Unit - A division of the facility consisting of a grouping of contiguous (usually) structural areas, outdoor areas, or functionally contiguous equipment items. The Survey Unit is the basic entity for management of the termination survey. Three categories of survey units have been established: plant systems, structures and outdoor areas.

Survey Unit Classification Descriotion - A listing of all survey units established for the termination survey which identifies the classification of each as "affected" or

" unaffected".

System Final Configuration: - The status of plant systems following completion of the termination survey. The final configuration establishes the nature of the controls required to maintain the integrity of survey results. It also determines the nature of configuration control and engineering review of access methods needed to obtain survey measurements of system component internals. Four categories are established:

(a) Operable - maintained to meet Technical Specifications.

(b) Functional - Essential support, not required per Technical Specifications however, necessary for minimal plant functions, habitability, and preservation concerns.

(c) Protected - not to be operated in the defueled mode. These systems will be left in a de-energized, safe state and laid up in accordance with System 12y-up Implementation Package (SLIP), which specify maintenance and custodial services necessary to protect them pending disposition the LIPA possession only license.

(d) Decommissioned - taken out of service, and completely or partially removed.

Remnants of removed systems are abandoned in place.

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Termination Surv:y Plan R2v. 3 Systematic Sample - A sample which is obtained by some systematic method as opposed to a random sample; for example, selection from a list using a specified interval for selection (MA90). In a structural survey unit which has been uniformly gridded, a systematic sample could be comprised of every fourth block, for example.

Termination Survey - Radiological measurements, evaluations and supporting activities undertaken to demonstrate that the Shoreham facility satisfies the criteria for unrestricted use. Termination survey field activities may be divided into phases called the Preliminary and Final surveys.

Termination Survey Section - A Section established within the Radiological Controls Division, Operation and Maintenance Department of the LIPA Decommissioning Project Organization to design and implement the termination survey.

Termination Survey Reoort - A report describing the methods, and results of the Termination Survey. It initiates the NRC review and final inspection of the facility for termination of the facility license. It is also called the Final Report. It will be issued in stages as phases of the Termination Survey are completed. The fm' al update of the Termination Survey Report will document the completed Termination Survey.

Type I Error - In the statistical theory of hypothesis testing, the error incurred by rejecting a hypothesis when it is actually true (MA90). In the termination survey application it is the probability of deciding that the facility (or a survey unit) meets the release criteria, when in fact, it does not. Also called the a error; the error probability is denoted by a.

Tvoc II Error - The probability of deciding that the facility (or a survey unit) does not meet the release criteria, when the true condition is that the facility does meet the criteria. The Type II error probability is denoted by #.

Unaffected Area - A designation used to identify a survey unit which is not expected to contain residual radioactivity from licensed operations based upon the operating history and radiological surveys.

Work Instruction - A document used to guide performance of a task. Work instructions are similar in format and content to a procedure and are issued and controlled under a station Health Physics procedure. A work instruction is approved at the Section Head level.

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l Termmation Survey Plan Rev. 3 j Appendix A Survey Design Guidelines 1.0 Introduction This Appendix provides guidance for preparation of the design of the Shoreham Decommissioning Project Termination Survey.Itincorporates the guidance in NUREG/CR-5849 (BE92) with adaptations to account for conditions at the Shoreham facility.

The objective of survey design is to defm' e the measurements necessary to demonstrate that the facility satisfies the release criteria. As it is not feasible to perform detailed radiological surveys of 100 percent of the facility, measurements are performed at selected locations.

Due to the physical complexity of the facility with the possibility of residual contamination limited to a relatively small portion of the facility,a stratified sampling approach is utilized.

This approach adjusts the intensity of measurements according to the likelihood of encountering residual radioactivity. This approach is implemented at the facility level by classification ofindividual survey units according to contamination potential. Survey units are farther broken down into subunits, if necessary, to provide the appropriate degree of survej intensity for complex structures and plant systems.

2.0 Survey Design Process 2.1 Facility Organization and Classification At the facility level, the design and organization for the survey consists of the following:

a. division of the facility into discrete entities (survey units) for management of the survey,

b. division of the facility into categories with similar physical characteristics, i.e., structures, systems and outdoor areas,

c. classification of survey units into two major strata based upon potential for ,

residual contamination, i.e.,affected and unaffected, and

d. establishment of a reference grid system for identification of measurement locations.

2.2 Survey Design for Individual Survey Units The Termination Survey is implemented in the field at the individual survey unit level.The design isdeveloped, measurement instructions prepared and measurements completed for each survey unit independently. The guidelines in this Appendix are A-1

Termination Survey Plan Rev. 3 applied to each survey tmit based upon its classification as affected or unaffected.

The guidelines establish ti.=: level of measurement intensity for each type of measurement needed to demonwate that the release criteria are satisfied.

3.0 Facility Breakdown into Survey Units 3.1 Definition of Survey Unit A survey unit is defined as a division of the facility consisting of like entities for purposes of management of the Termination Survey.Three categories or types have been established: structures, which include building interiors, plant systems, and outdoor areas. These categories allow the grouping oflike elements.

3.1.1 Structures For Termination Survey purposes, structures include the indoor portions of site buildings including the exterior surfaces of plant systems, equipment and furnishings located therein. Structural survey units are established by division of the facility buildings into discrete (usually contiguous) geographical areas taking advantage of existing structural boundaries where possible, i.e., rooms or major elevations of small buildings.

3.1.2 Outdoor Areas Outdoor areas include all site grounds determined to be within the scope of the Termination Survey, the outside surfaces of buildings and miscellaneous outdoor structures not identified as separate structures. These miscellaneous outdoor structures include equipment storage pads, switchyard transformers and storage tanks, for example.

3.1.3 Systems Systems included in the Termination Survey as survey units include all nuclear steam supply, reactor control, and process systems, and building service systems associated with the SNP5 design and described in the SNPS USAR (LIILO90). Each system identified in the Records Management File Code List is treated as an individual survey unit. A system survey unit encompasses the interior surfaces of fluid carrying piping and components of the system.

3.2 Independence of Structures and Systems The survey of each individual survey unit is conducted independently. Interior surfaces of piping and equipment which comprise the system survey unit are surveyed independently of the exterior surfaces. Exterior surfaces of equipment and piping are included in the survey of the structural survey unit in which they are located.

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Termination Sumy Plan Rev. 3 3.3 Size and Scope of Survey Units 3.3.1 Structures A structural survey unit includes the entire surface area of floors, walls, ceilings, and outside surfaces of equipment and furnishings. The size of structural survey units is established by natural boundaries such as rooms in buildings or by building elevation in small buildings. The optimal floor area of a structural survey unit classified as affected is an area of approximately 100 m'. Units with floor plans as small as 50 m'and as large as approximately 300 m2can be established if necessary, due to variations in room size or building floor plan size. Unaffected areas in large buildings may be combined to encompass multiple rooms of similar composition into a single survey unit for efficiency in survey administration as long as guidelines for measurement intensity are followed.

3.3.2 Outdoor Areas The size ofindividual outdoor area survey units which consist of site grounds is generally determined by features such as roadways, major building boundaries, fences, etc. Building exteriors (all surfaces) generally comprise a single survey unit.

3.3.3 Systems Each individual system within the scope of the Termination Survey is established as a survey unit. The boundaries of piping and other systems are established by the controlling drawings, usually the "MFSK" series of flow diagrams.

4.0 Classification of Survey Units by Contamination Potential After breakdown of the facility into survey units, each is classified into one of two strata which identifies each unit as "affected" or " unaffected". Units identified as affected have a possibility of containing residual contamination and those identified as unaffected have a very low probabihty of residual contamination. Classification ofindividual survey units is based on the history of radioactive materials involvement or potential for contamination of the overall survey unit.

4.1 Structures and Outdoor Areas 4.1.1 Affected Areas Structures and outdoor areas are classified as affected for survey design purposes when the following conditions apply:

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TGrmmation Survey Plan Rev. 3

a. areas with potential contamination based on operating history, or known contamination based on radiological surveys, or

b. areas where radioactive materials were used or stored, and where records indicate spills or occurrences which could have resulted in contamination.

4.1.2 Unaffected Areas Structures and outdoor area survey units are classified as unaffected if the following conditions apply:

a.not expected to contain residual radioactivity from licensed activities based upon the operating or utilization history and radiological survey results, and

b. not classified as affected.

4.2 Systems 4.2.1 Affected Systems Systems are classified as affected for survey design purposes if the following conditions apply:

a. the potential for contamination exists based on operating history, or there is a known history of contamination based on radiological surveys, or

b. the system circulated, stored or processed radioactive materials, inchding: primary coolant, radioactive process or treatment media which were associated with the operation or control of the Nuclear Steam Supply System (NSSS) such that they could become contaminated or experience neutron activation; or where records indicate spills or occurrences which could have resulted in contamination.

4.2.2 Unaffected Systems System survey units are classified as unaffected if the following conditions apply:

a. not expected to contain residual activity based upon the operating or utilization history and radiological survey results, and b, do not meet the criteria for classification as affected.

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Termmation Surv:y Plan Rev. 3 4.3 Reclassification Survey units may be reclassified subsequent to the initial classification according to the following criteria:

4.3.1 Upward Reclassification j A survey unit (or subunit) must be reclassified from unaffected to affected if j at any time in the survey planning, design, or during the actual survey, new J information is obtained which indicates that the criteria for an unaffected survey unit are no longer satisfied. An investigation, to determine whether upward reclassification of the survey unit (or subunit) is required, will be initiated if any total surface contamination measurement exceeds the " action level" calculated as described in Section 5.0 of this Plan, if any removable surface contamination measurement exceeds 25 percent of the removable surface contamination limit specified in Table 3.1, or if a single gamma exposure rate measurement exceeds 5 pR/hr. The investigation of a survey unit (or subunit), including additional measurements taken to determine the extent of residual contamination present, should either conclude that the survey unit (or subunit) continues to meet the criteria for an unaffected area or the survey unit (or subunit) will be reclassified as an affected area. If a survey unit is reclassified upward, sufficient measurements must be taken to meet survey design guidelines for affected areas. Only under conditions of such reclassification is it acceptable to classify a subunit as an affected area within a survey unit previously classified as unaffected.

4.3.2 Downward Reclassification A survey unit may be reclassified downward from affected to unaffected prior to the survey design if additional information is obtained such that the criteria for classification as an unaffected area are satisfied. For example, through the investigation for the history file preparation, additional survey results may become available subsequent to the initial classification which show that no contamination is detected and records indicate that there is no history of radioactive materials use.

4.4 Alpha Affected Areas Due to the lack of any history of alpha contamination from SNPS operations, alpha surveys are not routinely performed during the Termination Survey. However, in the event that alpha contamination is observed a mechanism is needed to expand the scope of the Survey to ensure that applicable release criteria limits for alpha contamination are satisfied. A survey unit (structural area, system or outdoor area) is classified as an alpha affected area if 2e following conditions apply:

a. alpha activity greater than 25 percent of release criteria limits has been detected, or A -5

i Tennination Surv;y Plan Rev. 3

b. the area or system is immediately involved with fuel handling or storage.

1 In alpha affected survey units, direct measurements and measurements of surface  !

alpha contamination are added to the list of measurements normally performed at the selected measurement locations. Surface scans for alpha activity are performed in the vicinity of the selected measurement locations.

i 5.0 Reference Grid System and Gridding of Survey Units The reference grid system for the termination survey establishes a discrete-uniform subdivision of areas covered by the survey.The principal objectives of grid marking are: to establish measurement locations for the survey design and to document where measurements were taken if the need arises to verify measurement results by repeat measurements, or to establish boundaries of areas needing remediation.

5.1 Gridding Affected Areas 5.1.1 Indoor Areas

a. Floors and Iewer Walls Floors and lower walls up to 2 meters from floor level are gridded, i.e.,

grids are marked on the surfaces. The grid size is one meter square.

b. Upper Walls and Ceilings Grids are marked on upper walls and ceilings of affected structural survey units if the areas are determined to be suspect as defined in Section 7.1.2.lf determined to be suspect, the areas are gridded in accordance with Paragraph a. above. If the area is not suspect, the guidance in Paragraph c. below is applied.

c. Other Surfaces Grids are marked on surfaces of equipment and other non-regular surfaces in affected structural survey units as needed to aid in the design of surveys and to control and document the location of measurements. If not possible to mark such items into one meter square grids, regularly spaced markings one meter apart may be used.

5.1.2 Outdoor Areas

a. Site Grounds Site grounds in affected areas are marked offin 10 meter square grids.

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Termmation Surv;y Plan Rev. 3

b. Roofs and Building Exterior Walls Roofs and exterior walls in affected areas are marked offinto grids not to exceed 5 meters square. Smaller grids (one to three meters square) may be used on small surfaces where the use of five meter grids would result in fewer than 10 grid blocks on a surface.

5.2 Gridding Unaffected Areas 5.2.1 Indoor Areas Grid marking of surfaces in unaffected indoor structural survey units is not required. However, grids may be marked for reference purposes. The grid size is optional (one to 10 meters square), commensurate with the size of the area.

Temporary grid markings may be used to control scan surveys.

5.2.2 Outdoor Areas Grid marking of surfaces in outdoor areas is not required. If desired for reference purposes, grids may be used. The size is optional, but shall not exceed 30 meters square on site grounds and 10 meters square on building exterior surfaces.

5.3 Gridding for Systems Surveys Gridding on surfaces of system interiors is not normally performed. Identification of measurement locations is shown on survey maps which contain drawings or diagrams.

Interiors oflarge tanks and vessels in affected systems may be gridded as needed to design the survey. In such cases, the surfaces are marked offinto one meter grids.

Square grids are used where possible, but tank ends and other non-rectangular surfaces are marked in regular shapes which approximate one meter in area.

5.4 Accuracy of Grid Marking An accuracy objective which is reasonable for the scale of the SNPS facility is to document a measurement location with an uncertainty of less than one meter. To ensure that this is attained, grid blocks will be marked to an accuracy within 15% of the specified dimension. This also allows for accommodation of areas of irregular shape within a grid block if needed. Remnant areas at the end of rows can be incorporated into the last full size grid block as long as the area of the resultant block is not greater than approximately 25% above the area of the specified grid size. ,

6.0 Stratification-Breakdown of Survey Units l

6.1 Structures In affected areas, structures are generally divided into subunits for survey design and control. The floor and lower walls comprise a subunit and the upper walls and ceiling a separate subunit. Due to additional structural features and the presence of piping i

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i Termination Survey Plan Rev. 3 l and equipment, additional subunits may be needed. Separate subunits are usually created to establish " populations" where biased surveys are indicated. For example, piping penetrations in floors and walls where contaminated piping systems were  !

removed are usually grouped into a subunit. A large sump or catchment basin I location in a structural survey unit is treated as a separate subunit. Grating, decking, .

stairs and ladders are usually grouped into a subunit. Equipment skids are usually identified as subunits for survey of the exterior surfaces. Unaffected areas in '

structures need not be divided into subunits, but in some cases subunits may be used if needed to group elements in complex structures.

6.2 Outdoor Areas Outdoor site ground areas may be divided into subunits based upon the nature of the surface, e.g., to separate paved and unpaved areas which have different survey requirements.

6.3 Systems Systems with multiple and separate components are usually divided into subunits for survey design purposes. A typical subunit comprises a large bore component 'and the adjacent piping. Usually tanks, heat exchangers and other vessels are treated as subunits.

6.4 Classification of Subunits Subunits may be classified differendy from the survey unit in which it resides under certain conditions. A subunit within an affected survey unit may be classified as unaffected ifit is a physically separate entity with no likely mechanism for transfer of contamination from the affected portions to the unaffected portions. The typical application of this classification option is the classification of upper walls, ceilings and overhead areas within an affected survey unit as unaffected. The converse is not allowed, however, i.e., classification of an affected subunit within a unit classified as unaffected, except as provided for in Section 4.3.1 for reclassified subunits.

7.0 Measurement Location and Frequency of Measurements 7.1 Structures This Section provides guidance for determining the frequency of measurements, i.e.,

the number of measurement locations in structural survey units (and/or subunits).

The frequency of measurements in affected areas is determined by:

Large bore at SNPS refers to piping and the piping system components associated with piping having inside diameters of 3 inches or greater.

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i Termination Survey Plan R:v. 3 l l

a. the sensitivity of scanning surveys for floor and lower walls', and  !

i

b. the expected level of contamination relative to release criteria limits for upper walls and ceiling areas.

The measurement frequency in unaffected areas is not controlled by scan sensitivity.

Iocations for fixed measurements are at or near the center of the one meter grid blocks. Unless specified otherwise in the following paragraphs, the measurements taken at fixed locations are: direct surface beta-gamma, removable surface beta-gamma (smear) and gamma exposure (or microrem) rate measurement at one meter from the surface.

In survey units classified as  % affected, direct surface and removable surface alpha measurements are tak; ~tdition to the measurements identified above.

7.1.1 Affected Areas - Hoors and Lower Walls Floors and lower walls of affected areas are surface scanned over 100% of the area. Measurements are taken at fixed locations. Measurements are taken systematically in every second grid block. At least 30 measurement locations are selected.

7.1.2 Affected Areas - Upper Walls and Ceilings In large structural survey units, and those with complex structures and piping in the overhead areas, the survey design is usually implemented by establishing individual subunits for vertical and horizontal surfaces and perhaps additional subunits for equipment and furnishings. Survey intensity of upper walls and ceilings classified as affected areas is determined based upon the suspected level of residual contamination.

a. Not Suspect The rule for determining than an area is not suspect is: expected contamination levels must be below 50 percent of release criterion for total surface contamination and below 25 percent of the release criterion for removable surface contamination, i.e.,below the following

~

The scanning sensitivity value which triggers an increased frequency of fixed measurements is 75 percent of the surface contamination release criteria limit for average total surface contamination: 5000 dpm/100 cm2.That is,the scanning method must be capable of detecting levels below 3750 dpm/100 cm2 averaged over an area 2

not to exceed one m . As described in Section 4.2 of the Termination Survey Plan, the Termination Survey instrumentation scanning sensitivities are below this value, so the conditional measurement frequencies in NUREG/CR-5849 are not invoked.

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Tenmnation Surv:y Plan Rev. 3 values: 2500 dpm/100 cm' direct total surface contamination and 250 dpm/100 cm 2removable surface contamination. If the area is not suspect, a minimum of 30 measurement locations are selected on horizontal surfaces, and 30 on vertical surfaces. The measurement density shall be at least one measurement every 20 m' averaged over the area of the subunit. A scan survey is performed in the immediate area of each selected measurement location. The scan survey covers an area of approximately one m2 or greater. The measurement point locations are usually selected in a biased manner, i.e., focusing on locations where demolition, piping removal or other remediation work occurred and on surfaces where contamination potential is highest.

b. Suspect.

If any area is suspected to be or is above the levels prescribed in Paragraph a. above, the survey protocol is the same as for the floors and lower walls of an affected area. That is, a 100 percent scan with the fixed location measurement frequency of one measurement every second grid block, for a minimum of 30.The area is gridded in 1 m' grids.

7.1.3 Unaffected Areas

a. Inside The Power Block Floors and lower walls of unaffected indoor areas (structures) within the Power Block are scan surveyed over a minimum of ten percent of the area, focusing on walkways and commonly traveled areas.

Measurement locations are randomly selected. At least 30 locations or 2

an average of one measurement per 50 m (of floor plus lower wall surface area) are required whichever is greater.

Thirty locations or an average of one measurement location per 50 m',

whichever is greater, are selected at random from upper walls and ceilings. No scan survey is performed.

b. Outside The Power Block Floors of unaffected areas outside the Power Block are scan surveyed over a minimum of ten percent of the area, concentrating on traffic areas. At least 30 locations are selected at random, or an average of one per 50 m 2, whichever is greater. Lower walls are not included in the survey of floor areas.

Thirty locations are selected at random from the walls (including lower walls) and ceiling areas, or an average of one per 50 m 2, whichever is greater. No scan survey is performed.

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Termmation Surv:y Plan Rev. 3 7.2 Outdoor Areas 1

7.2.1 Affected Areas - Site Grounds l Affected site ground areas are surveyed according to the nature of the surface l covering.

a. Paved Areas Paved areas are surface scan surveyed over 100 percent of the area with a beta-gamma detector. Measurement locations are selected systematically to achieve an average of at least one per 10 m2 grid such that a minimum of 30 locations are surveyed. Direct and removable surface beta-gamma measurements and gamma exposure rate measurements (at one meter from the surface) are taken at each measurement location.

b. Unpaved Areas Unpaved areas are scan surveyed over 100 percent of the area using a gamma detector. Gamma exposure rate measurements at one meter from the surface are performed with at least one location per 25 m',

including at each location where a soil sample is collected, such that at least 30 locations are surveyed.

7.2.2 Unaffected Areas - Site Grounds Unaffected ground areas are surveyed according to the type of ground surface,

a. General - All Surfaces A minimum of 30 survey locations are selected at random. At each location (at the approximate center of the grid), a gamma exposure rate measurement is taken at one meter from the surface. Gamma exposure rate measurements are also obtained at each of the randomly selected locations where soil samples are obtained.

b. Paved Areas The paved areas within each survey unit are evaluated to identify traffic areas and areas with any potential for residual contamination.

These areas are scanned for beta-gamma contamination. All such areas are cleared of material as necessary to provide access for the survey. A minimum of 10 percent of paved areas will be scanned. A direct surface beta-gamma measurement is taken at or near the center of each survey location (grid block) identified for survey as described in paragraph 7.2.2.a,above.

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Termmation Survey Plan Rzv. 3

c. Unpaved Areas Unpaved areas are surface scanned over at least 10 percent of the area with a gamma detector (NaI or microrem). Emphasis is placed on traffic areas and roadways.

7.2.3 Affected Areas - Building Exteriors No building exteriors have been identified which are classified as affected. If classified as affected, they are scan surveyed over 100 percent of the area. A minimum of 30 measurement locations are selected on each of wall surfaces and roofs or an average of one per 20 m 2, whichever is greater. The survey design should focus on likely areas for deposition, e.g., downwind of stacks, vents and drains.

7.2.4 Unaffected Areas - Building Exteriors The survey of exterior surfaces of buildings and structures which are classified as unaffected is as follows:

a. Exterior Walls Surfaces The survey of building exterior walls focuses on the area within two meters of the ground surface. The areas are surface scanned over a minimum of 10 percent of the area. Measurement locations are randomly selected for a minimum total of 30 or an average of one per 50 m 2, whichever is greater.

b. Roofs The survey of building roofs which are classified as unaffected is limited to the roofs of the Power Block and immediately adjacent buildings. If any of these surfaces are classified as affected, they are surveyed in accordance with the guidelines in paragraph 7.2.3above.

Otherwise the roof areas are surveyed at a minimum of 30 location or one per 50 m2,whichever is greater. Roofs of other buildings in outside areas which are classified as unaffected are not surveyed.

7.2.5 Soil Sampling Soil samples will be collected from the site grounds for the Termination Survey. Thirty locations will be selected at random from the site. The soil samples will be collected from the top 15 cm of the soil column and will be analyzed as REMP samples. The analysis willinclude gamma spectroscopy for Co-60.

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l Terminatica Surv:y Plan Rev. 3 i l

Additional soil samples are collected as needed to support the Termination Survey of plant systems, some of which extend beyond the Secured Area of the site. Soil samples may be collected, for example, from leach fields, soil adjacent to or below catchments, sumps and building foundations.

Any soil sample containing detectable concentrations of Co-60, or any other non-naturally occurring radionuclides above the REMP detection limit (LLD),

will be investigated to determine the source of the contamination, as well as the areal extent and depth of the contamination.

7.3 Plant Systems 7.3.1 Selection of Measurement locations Surveys of systems are biased sampling surveys. Selection of measurement locations in both affected and unaffected systems focuses on the locations with the highest probability for contamination considering the system design and operational characteristics. Typically, the inlets, outlets, system interfaces, system crud traps and flow impingement areas are selected. A " location"in a system survey design is generally identified as an area which includes a surface defined by the interior of a component or of a discrete-specified length of piping in piping systems. A component includes any discrete element which is distinguishable from piping. When a component is removed or opened for access for Termination Survey measurements, typically three measurement locations are established: the component interior surfaces and the two piping spool pieces immediately adjacent. In non-piping systems such as ventilation systems, locations are identified as local areas where air impinges, such as inlet and outlet baffles and ducting interior surfaces at bends and cleanouts.

The survey design should take advantage oflow cost access methods including destructive cutting of piping (where permitted).

7.3.2 Breakdown Into Subunits Systems, including mechanical systems, such as the main steam turbine, are l broken down into subunits by functional areas, components or locations.

large vessels and heat exchangers are usually considered to be subunits of the system in which they are located. In piping systems, each fluid carrying interface to other systems is usually established as a subunit.

7.3.3 Survey Measurements The interior surfaces of components selected for survey measurements are usually scanned over 100 percent of the accessible area. When possible, locations are selected such that both direct and removable surface contamination measurements can be taken. Gamma exposure rate  ;

i measurements are not usually taken within systems,except where interiors are large enough for human occupancy, i.e.,large tanks and spaces such as the A - 13

Termination Survzy Plan Rev. 3 l Condenser Hotwell. At least one direct surface and removable surface contamination measurement is taken at each location.

7.3.4 Measurement Frequency - Affected Systems In general, the measurement frequency is strongly influenced by the size and  ;

complexity of a system. In affected systems, the design guideline is a minimum of 30 measurement locations. Exceptions may arise where the system is  ;

limited in physical extent such that 30 discrete locations cannot be identified.  :

7.3.5 Measurement Frequency - Unaffected Systems In unaffected systems, a minimum of 10 measurement locations will be ,

selected. Similarly,in unaffected systems, the number of measurements may ,

be limited by the physical extent of the system, so exceptions may be made to  !

the minimum number guideline. l l

7.3.6 Embedded Piping j Several piping systems, primarily Liquid Radioactive Waste (G11), and Fuel Pool Cooling and cleanup (G41) (P71), have substantial lengths of piping l which are embedded in reinforced concrete, i.e., floors and reactor cavity  ;

areas. For management of the survey,each system with embedded piping may  ;

be divided into subunits for survey design and control purposes. Detailed l surveys of embedded piping are performed with multiple GM detector t assemblies, called pipe crawlers, inserted into embedded piping runs. All' openings and normal access points (cleanouts) willalso be surveyed. All runs [

of piping will be systematically surveyed over at least 25 percent of their l length.

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LSNRC -2177 ENCLOSURE 2 ADJUSTMENT FACTORS for TERMINATION SURVEY MEASUREMENTS i

. Adjustment Factors for Termination Survey Measurements Summary The principal release criteria guideline values for surface contamination from beta-gamma emitting 2

radionuclides at Shoreham are the Regulatory Guide 1.86 limits of 5,000 dpm/100 cm for average total surface activity and 1000 dpm/100 cm' for removable surface activity. For areas where certain other hard-to-measure radionuclides may be present, such as Fe-55 which decays by electron capture, special guideline values apply. When mixtures of radionuclides are involved whose members have different guideline values, gross activity guideline values and elevated area guideline values are substituted for guideline values. Direct beta-gamma measurements above the critical level ' are multiplied by a " factor" (f) to account for the presence of the hard-to-measure species which may be present. Table 1 identifies the areas where special guideline values apply, ,

and presents the values of gross activity guideline values and the adjustment factor values. ,

Table 1 Summary of Gross Activity Guideline Values and Adjustment Factors Gross Activity Guideline Values Adjustment Factor, f Description Total Surface Activity (dpm/100 cm')(1)

Guideline Value Elevated Area Total Removable Guideline Value B-y B-y Bioshield Concrete 95,900 287,700 35.9 12.4(2)

Bioshield Steel 76,900 230,700 24.2 1.2 (3)

D-S Storage Pool 11,100 33,300 2.6 2.6 (Cutting Station)

Spent Fuel Storage Pool 9,400 28,200 2.3 2.3 Reactor Assembly 11,100 33,300 2.6 2.6 (bottom head)

All Other (4) 5,000 15,000 1.2 1.2 (balance-ofplant)

Table 1 Notes:

(1) Special guideline values have not been established for removable surface contamination.

The guideline value of 1000 dpm/100 cm2 applies.

(2) Residual activity from Tritium is not present in removable form in activated concrete.

(3) Activated steelis not considered to be removable. Any removable surface contamination ,

measured on the Bioshield steelliners is assumed to be similar to balance-of-plant deposits  !

(4) Gross activity guideline values are not applied to other areas, as special guideline values were not in effect for the majority of measurements in these areas.

' The criticallevel L., is used to identify measurements which are "above" background.

It is defined in the Shoreham Termination Survey Plan as the upper limit of the 95 %

confidence interval of the mean value of the net count distribution. For this use the critical level is expressed in unit:, of dpm/100 cm 2, Page 1 of 5 7/20/94 )

l l

. Adjustment Factors for Termination Survey Measurements Radionuclide Composition To compute gross activity guideline values and adjustment factors for direct surface activity measurements, radionuclide compositions were determined for the areas where special guidelines apply. For each of the areas, the most likely sources of residual contamination were identified.

These are shown in column two of Table 2. The radionuclides which comprise residual radioactivity sources for each of the areas are shown in column 3 and the sources of data evaluated to estimate activity ratios are shown in column 4 of Table 2.

Table 2 Determination of Radionuclide Composition Area Description Source of Residual Principal Data Sources (2)

Contamination Radionuclides (1) .

Bioshield Concrete neutron activated Fe-55, Co-60, activation analysis, concrete H-3 core sample analysis Bioshield Steel Liner neutron activated steel Fe-55, Co-60 activation analysis, core sample analysis Dryer-Separator Pit neutron activated vessel Fe-55, Co-60, activation analysis, Cutting Station and vesselintemals Ni-63 Engineering records which were cut up in of materials the cutting station processed Spent Fuel Storage sediments derived from Fe-55, Co-60, fuel pool cleanup Pool fuel, control rods and Ni-63 system filter sample associated hardware analysis stored in pool, primarily activated corrosion products Reactor Vessel neutron activation of Fe-55, Co-60, activation analysis, Bottom Head stainless steel cladding Ni-63 vessel cladding on inner surface, normal sample analysis plant corrosion products on outside surface Table 2 Notes:

(1) Radionuclides listed include Co-60 and hard-to-measure species which comprise 2: 5 % of the total activity. Eu-152 has also been reported at concentrations above 5 % of the total activity in Bioshield concrete samples. It is not accounted for separately, as it is detected by survey instrumentation. Any residual activity due to Eu-152 will be accounted for along with Co-60 and other detectable radionuclides against the guideline values for B-y emitters.

(2) The principal source of information used to calculate activity compositions is highlighted in bold type.

Page 2 of 5 7/20/94 l

- \

Adjustment Factors for Termination Survey Measurements The principal source of information on radionuclide composition of neutron activated materials of construction is neutron activation analysis reported in the Shoreham Site Characterization Report (LILC090). However, additional information f or Bioshield and Reactor Vessel materials is available from radiochemistry analysis of samples.These include samples of Bioshield and Reactor Pressure Vessel material collected during May 1994. Four samples of Bioshield steel, one of Reactor Pressure Vessel bottom head cladding material and two of Bioshield concrete were collected.

Samples of Bioshield steelliner and concrete were sent to three separate laboratories for analysis.

Each sample was analyzed for gamma emitters and Fe-55. Activation analysis results are the only source of information for the characterization of materials handled in the Dryer-Separator Pit Cutting Station. Spent Fuel Storage Pool residue was characterized using results from radiochemistry analysis of a Pool Cleanup System filter.

An evaluation was performed to determine the best data source for use in characterization of Bioshield steel and concrete and the Reactor Vessel Bottom Head. This evaluation is summarized in Attachment 1. The attachment summarizes the radiochemistry sample results and shows ratios of Fe-55:Co-60 activity derived from the Bioshield radiochemistry samplo data. It is seen that Fe-55:Co-60 ratios for individual samples vary widely, ranging from 11 to 1,500,000 for Bioshield steel and from 1 to 2353 for Bioshield concrete. As an alternative method to estimate Fe-55:Co-I 60 ratios using radiochemistry results, ratios were obtained using calculated averages of Fe-55 and Co-60 concentrations. These composite Fe-55:Co-60 ratios were compared to Fe-55:Co-60 ratios obtained from neutron activation results. Good agreement is obtained for concrete (7.1 vs 11.4) and poor agreement is obtained for steel (577 vs 23) for the radiochemistry and neutron activation results, respectively. The principal limitation in the radiochemistry data is the high degree of uncertainty in the Fe-55 results. Two of the three laboratories reported no positive confirmation of Fe-55. In these cases, reported MDA values were used to calculate Fe-55:Co-60 ratios.

After a review of all the available data on Bioshield and Reactor Vessel Bottom Head radionuclide composition, it is concluded that the activation analysis results provide the most consistent basis for the characterization. This is also consistent with other radionuclide characterizations performed for Decommissioning purposes such as for Radioactive Waste shipment.

Calculation of Gross Activity Guideline Values The gross activity guideline values are obtained from the " sum of fractions rule" in Appendix A of NUREG/CR-5849 (BE92), whereby the sum of the ratios of each radionuclide to its guideline value must be s 1. The resulting formula for the GAG is:

GAG =

F, F, F,

- + - + . . . . -

G, G2 G, i

where: GAG = Gross Activity Guideline Value in dpm/100 cm , 2 F, = fraction of total activity due to the ith radionuclide, G, = guideline value for the ith radionuclide, in dpm/100 cm 2. The guideline values for each nuclide identified in Table 2 are: Co-60 = 5000; Fe-55 = 200000; H-3

= 200000; and Ni-63 = 5000 dpm/100 cm .2 Page 3 of 5 7/20/94 i

}

Adjustment Factors for Termination Survey Measurements The calculations are summarized in Table 3. The radionuclide composition for each area is shown in columns 2 and 3. The factor F, in column 3,is the fraction of the total activity represented by each nuclide. The adjustment factor f, whose values are shown in column 4, is obtained as the quotient 1/Fco, where Fe, is the fraction of total activity due to Co-60 in each area. It is applied to direct activity measurements to account for the total activity which may be present. Column 5 shows the individual fractions, F,/G,, which appear in the denominator of the GAG equation. It also shows the sums of the fractions. The GAGS shown in column 5 are obtained as the reciprocal of the sum of the fractions.

Table 3 Gross Activity Guideline Value Calculations Area Nuclides F f F/G, GAG (dpm/100 cm9 Bioshield Concreto (1) Co-60 0.0278 35.9 0.00000557 95923 t Fe-5 5 0.3159 0.00000158 (95900)

H-3 0.6559 0.00000328 sum 0.00001043 Bioshield Concrete Co-60 0.0809 12.4 0.00008090 1001 Removable (1) Fe-55 0.9181 0.00091810 (1000) sum 0.0009990 Bioshield steel (1) Co-60 0.0413 24.2 0.00000826 76902 '

Fe-55 0.9487 0.00000474 (76900) sum 0.00001300 Dry Cutting Station (1) Co-60 0.3840 2.6 0.00007680 11116 Fe-55 0.5635 0.00000282 (11100)

Ni-63 0.0517 0.00001034 sum 0.00008996 spent Fuel Pool (2) Co-60 0.4324 2.3 0.00008648 9414 Fe-55 0.4764 0.00000238 (9400)

Ni-63 0.0868 0.00001736 sum 0.00010622 Vessel Bottom Head (1) Co-60 0.3858 2.6 0.00007716 11063 Fe-55 0.5617 0.00000281 (11100)

Ni-63 0.0521 0.00001042 sum 0.00009039 All Other Co-60 0.830 1.2 0.00016600 5993 Fe-55 0.170 0.00000085 (5900) sum 0.00016685 Table 3 Notes:

(1) Data from Attachment 2.

(2) Data from Attachment 3. >

(3) Gross activity Guideline values rounded down to nearest 100 dpm/100 cm2 .

(4) Minor constituents not included (F, < 5 %). j l

l Page 4 of 5 7/20/94 l

i Adjustment Factors for Termination Survey Measurements ,

1 References BE92, J. D. Berger, " Manual for Conducting Radiological Surveys in Support of License Termination", prepared for U. S. Nuclear Regulatory Commission, NUREG/CR-5849 Draft, June,1992.

LILCO90, Long Island Lightin0 Company, "Shoreham Nuclear Power Station Site Characterization Program Final Report", May 1990 (Addendum 1, June 1990; Addendum 2, October 1990; Addendum 3, June 1992).

List of Attachments

1. Evaluation of Fe-55:Co-60 Ratios in Bioshield and RPV Materials.

2. Activity Ratio Calculations for Termination Survey Special Areas

3. Scientech, Inc. " Radioactive Sample Analysis Report: Shoreham SFSP Paper Filter",

5/27/94.

Page 5 of 5 7/20/94

. 1 Attachment 1 l

154ul-94 Evaluation of Fe-55:Co-60 Ratios in Bioshield and RPV Materials ,

i

1. Vendor Sample Radiochemistry Analysis Results  !

Location Scientech Teledyne INEL Fe-55 pCi/gm l D8 inner Liner 757 100

• 600 '

D8 Outer Liner 110

• 200

• 1500 ' I D8 Concrete 30 10

• 30 " l D11 Inner Liner 12300 300

• 1000 **

D11 Outer Liner 353 500

• 1000 "

D11 Concrete 45 8* 12 "

Vessel Bowl Shavings 5960 300

• l l

Co-60 pCi/gm I D8 inner Liner 1.37 8.83 1.26 D8 Outer Liner 0.16 1.85 0.001 1 D8 Concrete 0.39 5.33 0.05 D11 Inner Liner 1.87 1.05 1.58 D11 Outer Liner 0.06

• 0.001 0.005 D11 Concrete 14.3 5.85 0.0051 Vessel Bowl Shavings 1.85 2.92 II. Fe-55:Co-60 Ratios Obtained From Radiochemistry Results l Location Scientech Teledyne INEL l D8 inner Liner 553 11 476 I D8 Outer Liner 688 108 1500000 D8 Concrete 77 2 600 D11 inner Liner 6578 286 633 D11 Outer Liner 5883 500000 200000 D11 Concrete 3 1 2353 Vessel Bowl Shavings 3222 Notes:

• Results less than reported MDA, MDA values used to calculate Fe-55:Co-60 ratios.

" Uncertainty > t 100 % and not reported as isotope found.

The MDA values shown are used to calculate Fe-55:Co-60 ratios.

File:FE55_ANLWkt Page 1 of 2

Attachment 1 15-Jul-94 Evaluation of Fe-55:Co-60 Ratios in Bioshield and RPV Materials 111. Comparison Of Fe:Co Ratios From Radiochemistry and Activation Analysis of Bioshield Materials Fe-55 pCi/gm Co-60 pCi/gm Source Steel Concrete Steel Concrete Scientech D8-IL 757 1.37 Scientech D8-OL 110 0.1600 Scientech D8-Conc 30.0 0.39 Scientech D11-IL 12300 1.87 Scientech D11-OL 353 0.06 Scientech D11-Conc 45.0 14.3 Teledyne D8-IL 100 8.83 Teledyne D8-OL 200 1.85 Teledyne D8-Conc 10 5.33 Teledyne D11-IL 300 1.05 Teledyne D11-OL 500 0.001 Teledyne D11-Conc 8 5.85 INEL D8-IL 600 1.26 INEL D8-OL 1500 0.001 INEL D8-Conc 30 0.05 INEL D11-IL 1000 1.58 INEL D11-OL 1000 0.005 INEL D11-Conc 12 0.0051 Trimmed Mean

• 632 20.5 1.022 2.905 Std Dev 435 9.5 0.76 3.11 Activation Analysis Results pCi/gm Calculated Avg 1467.5 27.6 63.9 2.43 Estimated Error " 1467 27.6 63.9 2.43 Fe-55:Co-60 Ratios Steel Concrete Mean Std Dev Mean Std Dev Radiochemistry 618.4 i 459.9 7.1 i 7.6 Activation Analysis 23.0 1 23.0 11.4 i 11.4 Notes: ' Trimmed mean is obtained from data set with high and low values eliminated.

" Error of activation analysis concentrations estimated to be 100 %.

File:FE55_ANLWkt Page 2 of 2

14-Jul-94 Activity Ratio Calculations for Termination Survey Special Areas t

Part 1: Biosh' eld A. Total actisity - in Curies on July 1,1990 H-3 C-14 Fe-55 Co-60 Ni-59 Ni-63 Total Concrete 0.00968 0.0000019 0.0108 0.000555 0.000000 0.000002 0.02103 Steel 0.00098 0.0000018 0.371 0.00942 0.000004 0.000578 0.38198 B. Activity decayed to July 1,1994 Concrete 0.007721 0.0000018 0.00371876 0.000327 0.000000 0.000002 0.01177 Steel 0.000781 0.0000017 0.12774654 0.005561 0.000004 0.000560 0.13465 C. July 1,1994 activity concentration in pCi/gm Mass gm g Concrete 134650668 57.342 0.014 27.618 2.433 0.000 0.018 87.4248 iii Steel 87049350 8.980 0.021 1467.519 63.887 0.048 6.443 1546.89 S b

D. Activity ratios as of July 1994 (individual isotope pci/gm over total pCi/gm) 3 Concrete 0.655899 0.0001596 0.31590419 0.027834 0.000001 0.000201 1 M

Steel 0.005805 0.0000132 0.94868534 0.041300 0.000031 0.004165 1 E. Activity ratio for concrete removable activity only (assumes H-3 not available)

July 1994 activity 0.0139552 27.6178909 2.433396 0.000131 0.017583 30.0829 Activity ratios 0.000464 0.918058 0.080890 0.000004 0.000584 1 Notes:

1. Activity data from Shoreham Site Characterization report; TLG report LO1-22-001, April,1990.

2. Eu-512 not included in concrete activity data - calculated to be < 0.002 of total activity.

3. Tritium in activiated concrete is bound in water of hydration and not considered to be reinovable.

I File: ACT_ RATS.WK1 Page 1 of 3

14-Jul-94 Activity Ratio Calculations for Termination Survey Special Areas Part II: Cutting Station A. Total Activity -in Curies on July 1,1990 Component Weight-gm H-3 C-14 Fe-55 Co-60 Ni-59 Ni-63 Total Core shroud 22888316 0.0381 0.0043 118.662 47.3917 0.0283 3.9 170.024 t Jet Pumps 7409731 0.0018 0.0002 5.5189 2.2043 0.0013 0.1815 7.908 Core Spray Sparg 380065 0.000000 0.0000000 0.00102 0.000408 0.000000 0.000033 0.001 Top Guide Plate 4976868 0.0744 0.0084 232.1502 93.6201 0.0553 7.6298 333.538 Core Support 5891167 0.0017 0.0002 5.2119 2.0817 0.0012 0.1714 7.468 Vessel Clad 4411866 0.000029 0.0000033 0.0921 0.0369 0.000021 0.00303 0.132 Vessel Wall 111216810 0.000165 0.0000016 0.328 0.0115 0.000003 0.000428 0.340 Steam Dryer 25909090 0.0125 0.0096 0.022 Moist Separator 49895454 0.0521 0.0401 0.092 l Total 232979367 519.526

?

B. Activity decayed to July 1,1994 multiplied by weight fraction of component. G Component Wt_ fraction H-3 C-14 Fe-55 Co-60 Ni-59 Ni-63 Total Er

' Core shroud 0.0982 0.00299 0.00042 4.01405 2.74870 0.00278 0.37177 7.141 3 Jet Pumps 0.0318 0.00005 0.00001 0.06044 0.04139 0.00004 0.00560 0.108 $

Core Spray Sparg 0.0016 0.00000 0.00000 0.000001 0.000000 0.00000 0.00000 0.000 [

Top Guide Plate 0.0214 0.00127 0.00018 1.70759 1.18069 0.00118 0.15815 3.049 Core Support 0.0253 0.00003 0.00001 0.04538 0.03108 0.00003 0.00421 0.081 Vessel Clad 0.0189 0.00000 0.00000 0.00060 0.00041 0.00000 0.00006 0.001 Vessel Wall 0.4774 0.00006 0.00000 0.05391 0.00324 0.00000 0.00020 0.057 Steam Dryer 0.1112 0.00000 0.00000 0.00048 0.00063 0.00000 0.00000 0.001 Moist Separator 0.2142 0.00000 0.00000 0.00384 0.00507 0.00000 0.00000 0.009 Totals 1 0.00440 0.00061 5.88630 4.01120 0.00404 0.53998 10.447 C. Activity Ratios as of July 1994 Fraction of total 0.000421 0.000059 0.563469 0.383975 0.000386 0.051690 1 Notes ; 1. Incluoes allitems cut up in the oryer separator pli except UHt$ AY rack. I ne ContODullon of the rack is negligible because it's weight fraction is 0.03 and the Fe:Co ratio is approxately 1.

2. Activity data from Shoreham Site Characterization Report, primarily April 1990 TLG report LO1-22-001.

3. Moisture Separator & Steam Dryer activity all due to surface contamination; Fe-55 estimated as 1.3 x Co-60. i

4. Surface activity included in the activity shown for each component if data avalieble from Site Char Report. l

5. Mn-54 not included due to limited data and short half-life (312 d).

l l

File: ACT_ RATS.WK1 Page 2 of 3

14-Jul-94 Activity Ratio Calculations for Termination Survey Special Areas Part lil: Vessel Cladding H-3 C-14 Fe-55 Co-60 Ni-59 Ni-63 Total July 1990 Activity (Ci) 0.000029 0.0000033 0.0921 0.0369 0.000021 0.00303 0.132 July,1994 Activity (Ci) 0.00002 0.000003 0.0317 0.0218 0.00002 0.0029 0.0565 Fraction of total 0.00042 0.00006 0.56169 0.38585 0.00039 0.05207 1 Note: Activity data from Shoreham Site Characterization Report; April 1990 TLG report LO1-22-001.

Part IV: Fuel Pool Cs-137 Mn-54 Fe-55 Co-60 Co-57 Ni-63 Total May 1994 Activity ( Cilgm) 0.000055 0.0000779 0.0325 0.0295 0.000168 0.00592 0.06822 Fraction of total 0.00081 0.00114 0.47639 0.43242 0.00246 0.08678 1 Note: Activity data from Scientech Interim Radioactive Sample Analysis Report,5/01/94; not corrected for decay. g E

Part V: Decay Factors H-3 C-14 Fe-55 Co-60 Ni-59 Ni-63 0-Half-life (yr) 12.26 5720 2.6 5.26 80000 92 h Decay Factor 0.798 1.000 0.344 0.590 1.000 0.970 3 M

4 Note: Decay period is four years - from July 1,1990 to July 1,1994.

File: ACT_ RATS.WK1 Page 3 of 3 1

i o e 20$ PEIRY PARKWAY,810 m GAITHERSBt'RG, MD 20877 a PH0hfa 301-977-4480 m FAX. 301-840-2182 Radioactive Sample Analysis Report Plant Name  : Shoreham Sample Type  : SFSP PAPER FILTER Reference Date/ Time  : 05/01/94 12:00 Receipt Date  : 05/06/94 Reporting Date  : 05/27/94 Sample Number  : 24017 Plant Sample ID  : 94A-05-014 Volume / Weight  : 2.6730E+02 gram Purchase Order Number : C93L1099 Project Number  : 1-011-92-214 Measured Concentration (uCi/ gram)

Nuclide Value  % Unc. Nuclide Value  % Unc.

H-3 <6.6E-05 Ru/Rh-106 <1.1E-04 ,

C-14 <1.1E-06 Ag-108m <1.9E-05 Cr-51 <1.7E-04 Ag-110m <3.1E-05  ;

Mn-54 7.79E-05 33 Sn-113 <2.6E-05 Fe-55 3.25E-02 15 Sn-117m <1.4E-05 i

<2.8E-05  ;

Co-57 1.68E-05 32 Sb-124 Co-58 <4.3E-05 Sb-125 <6.0E-05  ;

Fe-59 <5.9E-05 I-129 <1.6E-05 Ni-59 <1.3E-03 I-131 <3.8E-05 .

Co-60 2.95E-02 15 Cs-134 <2.7E-05 t Ni-63 5.92E-03 10 Cs-136 <4.6E-05 Zn-65 <1.0E-04 Cs-137 5.52E-05 38 Se-75 <1.3E-05 Ce-139 <1.0E-05 i

~

Sr-85 <2.4E-05 Ba/La-140 <2.1E-05 Y-88 <1.3E-05 Ce-141 <1.9E-05 Sr-89 <2.4E-06 Ce/Pr-144 <7.0E-05  !

Sr-90 <2.3E-06 Eu-152 <3.3E-05 Nb-94 <3.1E-05 Eu-154 <1.7E-05  :

I Nb-95 <4.0E-05 Eu-155 <2.7E-05 Zr-95 <6.7E-05 Np-237/Pu-242 1.04E-07 66 Tc-99 6.24E-06 25 Pu-238 4.37E-07 60 Ru-103 <2.7E-05 Pu-239/240 2.743-07 33 Am-241 <1.4E-06 Pu-241 <2.7E-05 Gross Alpha <4.0E-06 Cm-242 <3.7E-07 Cm-243/244 <1.1E-06 Radiometrics Manager ',$k)

Laboratory Manager I /

/ /]

The data contained in this report were produced and d6cumented in accordance with approved quality control and quality assurance procedures. All of the results are decay corrected to the reference date listed above. Indicated errors are two (2) standard deviations based on counting statistics only.

C/ANY ycsxm -4