ML20081A945
| ML20081A945 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 02/16/1995 |
| From: | Laudeman M, Payne A, Vitkus T OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC |
| Shared Package | |
| ML20081A936 | List: |
| References | |
| CON-FIN-A-9076 ORISE-95-B-81, NUDOCS 9503150354 | |
| Download: ML20081A945 (94) | |
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l I I I I n I I I I! The Oak Ridge Institute for Science and Education (ORISE) was established by the U.S. Department of Energy i to undertake national and international programs in science and engineering education, training and management systems, energy and environment systems, and medical sciences. ORISE and its programs are operated by Oak E Ridge Associated Universities (ORAU) through a management and operating contract with the U.S. Department of Energy. Established in 1946, ORAU is a consonium of 88 colleges and universities. g'; 1 i l NOTICES The opinions expressed herein do not necessarily reflect the opinions of the sponsoring institutions of Oak Ridge Associated Universities. l E j This report was prepared as an account of work sponsored by the United States Government. Neither the United g i States Government nor the U.S. Department of Energy, nor any of their employees, makes any warranty, expressed g or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe on privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement or recommendation, or favor by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. I I.
b~ ORISE 95/B-81 CONFIRMATORY SURVEY OF THE REACTOR BUILDING AND PHASE 4 SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK { Prepared by _ T. J. Vitkus Environmental Survey and Site Assessant Program Energy / Environment Systems Division Oak Ridge Institute for Science and Education Oak Ridge, Tennessee 37831-0117 Prepared for the U.S. Nuclear Regulatory Commission Headquarters Office Sponsored by the Division of Waste Management FINAL REPORT t FEBRUARY 1995 This repon is based on work performed under an Interagency Agreement (NRC Fin. No. A-9076) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy. u Oak Ridge Institute for Science and Education performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy. r Shoreham Rea: tor Buildmg and Phase 4 - February 16,1995
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e CONFIRMATORY SURVEY OF THE REACTOR BUILDING AND PHASE 4 SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Prepared by: T. J. Vitkus, gr/ ject I.eader Date: 8!/'/!77
/ /
Environmental Survey and Site Assessment Program Reviewed by: fA p k L u> h w Date: ill'if W M. J. Laudeman, Radiochemistry Laboratory Supervisor Environmental Survey and Site Assessment Program Reviewed by: om Y. k(m A. T. Payne, Quality Assuranc6 Officer Date: d!//pfII Environmental Survey and Site Assessment Program Reviewed by: Date: e2//7/95 W. L. Beck, Program Director ' ' Environmental Survey and Site Assessment Program Shoreharn Reactor Buildmg and Phase 4 - February 16.1995 '
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I I ACKNOWLEDGEMENTS The author would like to acknowledge the significant contributions of the following staff members: i FIELD STAFF T. L. Bright T. D. Herrera A. L. Mashburn ' E. H. Montalvo J. L. Payne l LABORATORY STAFF R. D. Condra J. S. Cox M. J. Laudeman CLERICAL STAFF D. A. Cox R. D. Ellis K. E. Waters l ILLUSTRATOR T. D. Herrera l l l l l l l Shoreham Reactor Buildmg mi Phase 4 - February 16,1995 l
i d TABLE OF CONTENTS ; 1 PAGE i
- List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv 4
il List of Tables ...................... .......................vii > d . Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . viii t Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ! Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 :
- Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 ;
3 Document Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 . Procedures ................................................ 4 1 Findings and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 8
- Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
- Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 )1 References ...................................... . . . . . . . . 68 4
Appendices: i Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86 Termination of Operating Licenses for Nuclear Reactors I j i 4 I 1 I i i l Shoreham Reactor Buildmg and Phase 4. February 16. 1995 iii l
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LIST OF FIGURES PAGE MGURE 1: location of the Shoreham Nuclear Power Station . .............18 MGURE 2: Plot Plan of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 19 MGURE 3: Reactor Building Floor Plans of 8'-0" and 40'-0" Elevations-Structural Units S>. tveyed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 MGURE 4: Reactor Building Floor Plans of 63'-0" and 78'-0" Elevations-Structural Units Surveyed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ 21 - MGURE 5: Reactor Building Floor Plans of 112'-0" and 126'-0" Elevations-H Structural Units Surveyed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 - MGURE 6: Reactor Building Floor Plans of 150'-9" and 175'-9" Elevationa-Structural Units Surveyed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 MGURE 7: Reactor Building, Primary Containment 63' NW G/A (PC004)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . 24 FIGURE 8: Reactor Building, Primary ",ontainment Sub-Pile Room (PC005)- ) Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 25 MGURE 9: Reactor Building, Primary Conuinment 78' NE G/A (PC007)- Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . 26 FIGURE 10: Reactor Building, Primary Containment 78' SE G/A (PC008)- Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . 27 FIGURE 11: Reactor Building, Primary Containment 103' SE G/A (PC012)- Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . 28 kiGURE 12: Reactor Building, 8' SE G/A (RB004)-Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 FIGURE 13: Reactor Building, 8' SW G/A (RB006)-Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 FIGURE 14: Reactor Building,40' SW G/A (RB014)-Measurement and Sampling locaticas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . 31 FIGURE 15: Reactor Building, 78' SE G/A (RB037)-Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 Shortham Reaciar Buildang and Phase 4 - Fetuuary 16.1995 iv
LIST OF FIGURES (Continued) P.MiE FIGURE 16: Reactor Building, West Accumulator Aisle (RB038)-Measurement and Sampling Imcations ..............................33 FIGURE 17: Reactor Building, Fuel Pool Pipe Tunnel (RB057)-Measurement and Sampling Locations ..............................34 FIGURE 18: Reactor Building, RWCU Regen /Non-Regen HTX's Room (RB061)- l Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 35 i i FIGURE 19: Reactor Building, Spent Fuel Storage Pool, Floor and North j and West Walls (RB068)-Measurement and Sampling Locations . . . . . . 36 ! FIGURE 20: Reactor Building, Spent Fuel Storage Pool, South and East , Walls (RB068)-Measurement and Sampling Locations . . . . . . . . . . . . 37 ; I FIGURE 21: Reactor Building, Reactor Cavity Floor (RB071 x01)-Measurement and Sampling Locations ..............................38 FIGURE 22: Reactor Building, Reactor Cavity Walls (R3071 x01),-Measurement l and Sampling Locations. .............................39 FIGURE 23: Reactor Building, Dryer / Separator Storage Pool (RB072 x01),- Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . 40 l FIGURE 24: Reactor Building,175' South G/A (RB103)-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 FIGURE 25: Reactor Building,175' NW G/A (RB106)-Measurement and Sampling Imcations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 FIGURE 26: Reactor Building, Polar Crane (RB109)-Measurement and , I Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 ! FIGURE 27: Reactor Building, Polar Crane Rail Area (RB109)-Measurement I and Sampling Imcations ..............................44 FIGURE 28: Radwaste Building,52' SE G/A Shield Blocks (RWO73 x02 and x03)- ; Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 45 j FIGURE 29: Radwaste Building, 52' SE G/A Shield Blocks (RWO73 x02 and x03)- Measurement and Sampling Imcations . . . . . . . . . . . . . . . . . . . . . . 46 l l Shoreham Reactor Buildmg and Phase 4 - February 16.1995 V
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LIST OF FIGURES (Continued) PAGE ,
)
FIGURE 30: Radwaste Building,52' SE G/A Shield Blocks (RWO73 x02 and x03)- 1 Measurement Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 , 1 FIGURE 31: Reactor Building, Reactor Assembly Lower Bowl (SU001)- , Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 48 FIGURE 32: Reactor Building, Nuclear Boiler Main Steam Relief Valve (SU002)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 49 FIGURE 33: Reactor Building, CRD Hydraulic Control Components (SU004)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 50 FIGURE 34: Reactor Building, Porous Concrete Sump Components (SU014)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 51 FIGURE 35: Outdoor Building Storin Drain Manways and Trenches Components . (SUO23)-Measurement and Sampling Imcations ................ 52 ! FIGURE 36: Reactor Building, Reactor Primary Containment Components (FUO58)- Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . 53 FIGURE 37: Reactor Building, Reactor Prunary Containment Components (SUO58)- l Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 54 t FIGURE 38: Reactor Building, Ventilation Components (SUO60)-Measurement and Sampling Locations ..............................55 FIGURE 39: Reactor Building, Ventilation Components (SUO61)-Measurement ! and Sampling locations ..............................56 l l l l l Shoreham Reactor Buildmg and Phase 4. February 16.1995 Vi i l
l i I LIST OF TABLES i PAGE TABE 1: Reactor Building Confirmatory Survey Units . . . . . . . . . . . . . . . . . . 57 TABE 2: Summary of Surface Activity Levels . . . . . . . . . . . . . . . . . . . . . . . 59 TABE 3: Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 . TABE 4: Confirmatory Radiological Status Summary-Structures . . . . . . . . . . . 62 TABE 5: Confirmatory Radiological Status Summary-Systems . . . . . . . . . . . . 66 - l. , e i ; 4 1 J l 1 { l l l i j i i i i Shoreham Reactor Building and Phase 4. February 16,199$ vii i
l i 1 ABBREVIATIONS AND ACRONYMS l 1 i ~ ASME American Society of Mechanical Engineers
- cm2 eg::n centimeter
[ - Co-60 cobalt-60
- cpm coums per minute 4 f DOE Department of Energy l 2
j dpm/IOD cm disintegrations per minute per 100 square centimeters i j EML Environmental Measurements laboratory- , ! EPA Environmental Protection Agency l [ ESSAP Environmental Survey and Site haemament Program ~ i Fe-55 iron-55 , :l l ft2 . square feet i i H-3 tritium i ha: hectare ! GAG- gross activity guideline i GM Geiger-Mueller .; km kilometer . I, critical level
- LILCO Long Island Lighting Company I LIPA Iong Island Power Authority '
- m meter i m2 square meter -
l MDA minimum detectable activity j; Nal sodium iodide Ni-63 nickel-63 NIST National Institute of Standards anxi Technology ; l NRC Nuclear Regulatory Communion i 1 ORISE Oak Ridge Institute for Science and Education ! i PC# primary containment structural survey unit designation ! .' pCi/l picoeuries per liter : 3 pCi/g picoeuries per gram l 3 QA Quality Assurance , l QC Quality Control RB# Reactor Building structural survey unit designation RW# .Radwaste Building structural survey unit designation j SNPS ' Shoreham Nuclear Power Station l SU# system r.:r,ey unit designation UCL upper confidence level R/h microroentgens per hour i ZnS zine sulfide A~ Shonham Reactor Buildmg and Phase 4. February 16,1995 Viii 1 i
t CONFIRMATORY SURVEY .! OF THE , REACTOR BUILDING AND PHASE 4 SYSTEMS SHOREHAM NUCLEAR POWER STATION l BROOKHAVEN, NEW YORK ; l INTRODUCTION AND SITE HISTORY ! i The Long Island Lighting Company (ULCO) constmeted a boiling water reactor known as the Shoreham Nuclear Power Station (SNPS). The plant was designed to provide a gross electrical f, output of 849 Megawatts and achieved initial criticality in Febmary 1985. The U.S. Nuclear ; Regulatory Commission (NRC)I ~~- No. NPF-82 (NRC Docket File No. 50-322) issued for l the facility allowed reactor operations at power levels not to exceed 5% of full power. Low i power testing in accordance with the license then commenced in July,1985'and continued i intermittently until January,1989, at which time power generating operations were terminated. The total reactor operating history was equivalent to 2.03 effective full power days of fuel ~ ! exposure. The irradiated fuel, which was a standard low enrichment (2 to 3% uranium-235) i uranium fuel, was removed Irom the reactor vessel and placed into the spent fuel pool in August ! 1989. Subsequently, the 61 was shipped off site during 1993 and 1994. ! l i Various reactor components, piping systems, and other equipment be- radiologically l contaminated as a result of reactor operation. The primary contaminants identified during characterization studies were iron-55 (Fe-55) and cobalt-60 (Co-60). Smaller quantities of nickel-63 (Ni-63), tritium (H-3), carbon-14, nickel-59, manganese-54, zinc-65, and europium-152 were also identified (ULCO 1990). The Long Island Power Authority (UPA) was established to decommission the facility and . release the site for unrestricted use. UPA developed a decommissioning plan, approved for l implementation by the NRC in June 1992, which included decontamination or . removal of. contaminated portions of the reactor and other plant systems and equipment. A major l consideration of the decommissioning plan was to maintain the integrity, when possible, of plant ; structures and systems.
. Shoreham Reactor Building and Phase 4 February 16,199:
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t i ! ^ i l The decommissioning and termination surveys were conducted in four phases. - Phase 1 included l the termination survey of the internal components of the main turbine, the Turbine Building and j associated systems, site grounds, and exterior site structures. Phases 2 and 3, included the l j Reactor Building Suppression Pool, Phase 2 systems and the Radwaste Building. Phase 4 ] addressed the remaining portions of the Reactor Building and associated systems. l i ! l ~
. i i The NRC Headquarters' Division of Waste Management r=W that the Environmental i
4 - Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and ! 3 Education (ORISE) conduct confirmatory radiological surveys of the SNPS decommissioning i
- project. Since Febmary 1993, ESSAP has completed the confumatory survey of the turbine i 4
i internal components, the Turbine Building, site pour 4s, exterior site structures, Radwaste : 4 ; 2 Building, Suppression Pool, and Phase 2 systems. The results of these surveys are the subject ; j of separate fmal, or draft reports (Vitkus 1993a; Vitkus 1994a). The final confirmatory survey l l for Phase 4, the Reactor Building, has now been completed and is the subject of this report. ! i l 1 i
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, SITE DESCRIFFION l i ! The SNPS is located in the Town of Brookhaven, New York on the north shore of long Island, ! l l approximately 80 km (50 miles) east of La Guardia Airport and the confluence of the East River j and long Island Sound (Figure 1). The SNPS is located on a 32.4 hectare (80 acre) portion of f ; a a larger 202 hectare (ha) parcel of land owned by the IlLCO. The site is bounded on the north ! by long Island Sound, on the east by the Wading River Marshland, on the west by other IJLCO > ! property, and on the south by Route 25A. A cyclone fence encloses the 8 ha site secured area. Within this boundary are the buildings and grounds classified as the Restricted Area, also known l l l es the power block, where radiological controls were necessary (Figure 2). Each of the buildings ) l addressed during the confirmatory process are located here and are shown on Figure 2 as the ;
! t Turbine Building, the Radwaste Building, and the Reactor Building. Constmetion of the Reactor l 2 i j Building is predominately stmetural steel and concrete. Total floor space of the building is approximately 7,800 m' (84,000 ft2 ) divided among 8 levels at elevations 8'-0",40'-0", 63'-0", !
4 i 1 78'-0",112'-0",128'-9",150'-0", and 175'-0". The Reactor Building housed the nuclear steam ! supply system which included the Reactor Pressure Vessel and its associated auxiliary and safety - 1 :'
' Shoreham Reactor Buildmg and Phase 4. February 16, im 2 t
i systems. Major structural components included the prunary containment system, spent fuel - storage pool, dryer / separator pool, polar crane and building sumps. The major auxiliary and ; safety systems included the reactor core isolation cooling, high pressure coolant injection, core spray, stand by liquid control, reactor water cleanup, fuel pool cooling and clean up, and primary containment atmospheric control systems. Termination surveys have been performed in' accordance with Draft NUREG/CR-5849 ; (Berger 1992). LIPA classified plant systems and building surfaces into two categories that are ; based on the potential for residual contamination. The two area categories, referred to as ; affected or unaffected, are defined as follows: "affected areas are those areas of the SNPS that are potentially contaminated or have known contamination, or a system which circulated, stored or processed radioactive materials such that they could become contaminated, or experience, neutron activation, or where records indicated spills or other occurrences may have resulted in contamination; unaffected areas are those portions of the SNPS that are not expected to contain residual radioactivity" (LIPA 1994a). Area classification was determined by radiological use history, environmental monitoring activities, and the results of the previous characterization survey. Affected and unaffected areas are further subdivided into survey units. Survey units ! are categorized as structures (floors, walls, ceilings, and exterior surfaces of piping and equipment), plant systems (equipment and piping internals), and ouMoor areas (grounds and building exteriors). In addition, affected survey units also had sub-classifications as suspect or non-suspect, and may also have been classified as alpha affected if involved with fuel handling i or storage. Phase 4 of the decommissioning addressed a total of 117 survey units, of which 100 were structures, and 17 were systems. OBJECTIVES The objectives of the confirmatory survey were to provide independent document reviews and radiological data for uw by the NRC in evaluating the adequacy and accuracy of the licensee's procedures and termination survey results. Shoreham Reanor Buildmg and Phase 4 - February 16,1995 3
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t . l DOCUMENT REVIEW { j i e. ESSAP reviewed UPA's release records for those survey units selected for confirmatory survey ) ! (UPA ' 1994b). Documents were reviewed for adequacy, accuracy, completeness, and 4
- consistency. Data were reviewed for appropriateness of calculations and interpretations relative
- .to the guidelines. l I PROCEDURES i i,
i A survey team from ESSAP visited the SNPS during the period October 31 through No':-mber I 4,1994 and performed independent visual inspections, measurements, and sampling of survey ) j units associated with the Reactor Building. Table 1 lists, and Figures 3 through 6 show, the ! structural survey units selected for confirmatory surveys. Survey unit designators were alpha-f numeric with the first characters designating the type of unit, structure (building or area specific, RB = Reactor Building, PC = Pnmary Containment, RW = Radwaste Building [for this report, j refers specifically to the Bioshield wall blocks that are presently stored in the Radwaste Building]), or system (designated as SU), followed by a three-digit numeric reference. Subunits were given an additional two-digit designation preceded by X. Eleven survey units were selected [ randomly and twenty were selected by the NRC site representative. Field survey activities were
- conducted in accordance with a site-specific survey plan submitted to and approved by the NRC 1
l as well as applicable sections of the ESSAP Survey Procedures and Quality Assurance Manuals j (Vitkus 1994b). The following procedures apply to survey units selected for independent confirmatory surveys. Additional information regarding selection of confirmatory survey units j and the implementation of this plan may be found in the general site confirmatory survey plan 1 4 (Vitkus 1993). 1 i j 1 a )' 4 l 4 Shorcham Reactor Buildmg and Phase 4. February 16.1995 4 1 ]
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SURVEY PROCEDURES l t Reference System UPA established the grid system that ESSAP used for referencing measurement and sampling locations. The grid size or reference interval established by LIPA for a given survey unit was - dependent upon the classification of the survey unit (affected vs. unaffected) and surface (floor, lower wall, upper wall, ceiling, or equipment). Typically, the established grid consisted of' I m X 1 m grid blocks on the floors and lower walls (up to 2 m). ESSAP referenced measurement and sampling locations on ungridded surfaces (upper walls and equipment) to the floor or lower wall grid. Systems were referenced by drawings or by prominent cou yarznts. Surface Scans Surface scans for alpha, beta, and gamma activity were performed over 100% of floor and lower , wall surfaces and up to 50% of equipment surfaces within each structural survey unit selected for confirmation. Additional scans were performed over portions of upper wall, ceiling, and system surfaces, as well as locations such as drains, where material may have settled or accumulated. Accessible portions of each confirmatory system survey unit were also scanned. Locations of elevated direct radiation detected by scans were marked for further investigation. Scans were performed using gas proportional, Geiger-Mueller (GM), zine sulfide (ZnS), and/or l sodium iodide (Nal) detectors coupled to ratemeters or ratemeter-scalers with audible indicators. Surface Activity Merasuree.ee3 For each structural survey unit, ESSAP performed a minimum of 30 direct measurements for total beta surface activity. ESSAP also performed additional direct measurements at locations of elevated direct radiation detected by surface scans. At measurement locations, where the avIrage NRC surface contamination guideline was exceeded, the size of the contaminated area and the average activity in the contiguous 1 m2 area was also detennined. Measurements were performed using gas proportional and/or GM detectors coupled to ratemeter-scalers. A smear shortham Reactor Building and Pinse 4 - February 16. Im 5
- sample for determining removable gross alpha and gross beta activity levels were collected from -
each direct measurement location. Figures 7 through 29 show measurement and sampling locations. l Exposure Rate Measurements , 2 Exposure rate measurements were performed within each survey unit at each accessible floor l direct measurement location, excluding units RB071 x01 and RB109 and system interiors which ! 3 wire inaccessible for this type of measurement @gures 7 through 19,23 through 25, and 30). l All exposure rates were measured at 1 m above surfaces using a pressurized ionization chamber (PIC). Background exposure rates were previously determined during the confirmatory survey of the Turbine Building (Vitkus 1994a). f i
- Systems i LIPA provided access points into each system or system component listed in Table 1. Beta and gamma surface scans were performed within the accessible portions of each system or
. component, followed by beta direct measurements and smear samples. The total number of I direct measurements performed and smears collected was dependent upon component size and l accessibility and ranged from 5 to 45 measurements 'per system. Scans and direct measurements l were performed using gas proportional, GM, and/or NaI detectors coupled to ratemeters or j i ratemeter-scalers. Figures 31 through 39 show measurement and sampling locations. )
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l l l Miscellaneous Material Samoles 1 Per the request of the NRC site representative, ESSAP collected a water sample from the porous 1 concrete sump and a sediment sample from the truck bay trench (Figures 34 and 35).
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Shoreham Reactor Buildmg and Phase 4 - February 16. IM 6
I Sample Analysis and Data Interpretation ( Samples and data were returned to ESSAP's Oak Ridge, Tennessee facility for analysis and ; interpretation. Smear samples were analyzed for gross alpha and gross beta activity using a low background proportional counter. Direct measurement and removable activity results were j 2 i j converted to units of disintegrations per minute per 100 square centimeters (dpm/100 cm). I Surface activity levels which exceeded the badyvend distribution, referred to as the critical i l l level (I,), were then eviis;imi for those radionuclides that may not be quantified using field ! l e instrumentation, specifically Fe-55, H-3, and/or to a lesser extent Ni-63. The respective j correction factors used were dre@ by LIPA and were based on the fraction of total activity ( contributed by each radionuclide in each particular survey area (LIPA 1994a). Water and - , I sediment samples were analyzed by solid state gamma spectrometry. Spectra were reviewed for l Co-60 and any other identifiable photopeaks. Results were reported in units of picoeuries per ; liter (pCi/1) for water and picoeuries per gram (pCi/g) for the sediment. Exposure rates were reported in units of microroentgens per hour (pR/h). The 95% confidence level ( .), in accordance with NUREG/CR-5849, was calculated for surface activity and exposure rates for each survey unit selected for confirmation. A direct comparison of the ESSAP and LIPA survey unit results was also performed. Additional information concerning survey and analytical procedures may be found in Appendices A and B. DATA EVALUATIONS AND COMPARISONS The results of each survey unit sampled were statistically tested. The goal of the test was to determine; rich a given confidence level, whether the LIPA survey data was not biased low compared to ESSAP. The null hypothesis _was that in a survey unit, surface activities as calculated by LIPA were greater than or equal to those determined by ESSAP, i.e., H : pLIPA 2: ESSAP. This hypothesis was tested at the 95% confidence level (0.05 level of significance). If the 3ypothesis was rejected at that confidence level, the alternative hypothesis was accepted i.e., Hg LIPA < ESSAP. The test statistic, t, was calculated using the-followiug equation: l Sicreham Reactor Builduig and Phase 4 - February 16, 1995 7
- , _ . , ~ _
- e. X-K I
. (n,-1) sj + (nt-1)s! 'n ,+ nt ' I
% Dz*Dz ~2 s Det i i
where: M is the LIPA surface activity mean for a survey unit l
& is the ESSAP surface activity mean for the same survey unit
- ni. is the number of IlPA direct measurement data points
) ng is the number of ESSAP direct measurement data points , St ., Ss are the standard deviations. The calculated t was then compared to the critical value of Student's t-distribution (one-tailed) : fsr the appropriate degrees of freedom at the 95% confidence level (0.05 level of significance). If the H,: LIPA 2: pESSAP was rejected, then ESSAP evaluated adslitional options and , I alternatives and conferred with the NRC as to the recommended approach. i , j FINDINGS AND RESULTS ( i DOCUMENT REVIEW i l < ESSAP's review of the termination survey final report and release recortis for those survey units ; selected for confirmatory survey itxticated that the termination plan had been appropriately followed with no significant deviations. Data were appropriately converted, tested, and presented. ! SURFACE ACTIVITY GUIDELINES l
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The surface activity results that follow were compared with the following guidelines: ) ~ i I Shoreham Reactor Buildmg and Phase 4 - February 16. 1995 8
- l i
i. l l The beta guidelines are: l Total Activity 5,000 B-y dpm/100 cm2, averaged over 1 m2 2 15,000 B-y dpmIl00 cm2 , maximum in a 100 cm ,,, .. Removable Activity 1,000 B-y dpm/100 cm2 . The alpha guidelines are: Total Activity 5,000 a dpm/100 cm 2, averaged over 1 m2 2 2 15,000 a dpm/100 cm , maximum in a 100 cm ,,,, Removable Activity 1,000 a dpm/100 cm2 In addition, the NRC has approved site-specific surface contamination guidelines for H-3 and Fe-55, particularly in activated concrete and steel (Pittiglio 1994). These guidelines are: Total Activity 200,000 dpm/100 cm2, averaged over 1 m 2 2 2 600,000 dpm/100 cm , maximum in a 100 cm ,,,, Removable Activity 2 1,000 dpm/100 cm For those survey units involving mixtures of radionuclides LIPA substituted the gross activity guideline (GAG) value, which is based on the " sum of fractions rule" found in Appendix A of Shoreharn Reactor Buildmg and Phase 4 - February 16, im 9 I
- . . ~ _ . . .
, i l- , i Draft NUREG/CR-5849, rather than the above generic NRC guidelines (LIPA 1994a; Berger 1992). These confirmatory survey units, together with the GAG values are listed below: , SU001. Reactor Accamhly lower Bowl and RB072x01. Drver/Sanarator Pool-l 11,100 dpm/100 cm2 , averaged over 1 m2 , I 33,000 dpm/100 cm2 , maximum RWO73 x02. Bioshield Concrete ; 2 2 95,900 dpm/100 cm , averaged over 1 m ; 2 287,700 dpm/100 cm , maximum I e RWO73 x03. Bioshield Steel 3 76,900 dpm/100 cm2 , averaged over 1 m 2 ; 230,700 dpm/100 cm2 ,' maximum ;
- l RB068. Soent Fuel Pool l
9,400 dpm/100 cm2 , averaged over 1 m 2 , 2 28,300 dpm/100 cm , maximum i f SURFACE SCANS l l , Surface scans identified a number of locations with above background residual beta surface , l activity in three survey units, RB068 (Spent Fuel Storage Pool), RB072 x01 (Dryer /Separa:or Pool), and SU001 (Reactor lower Bowl). Additional investigation of each location determined l that two of the locations, each measuring less than 100 cm 2 in total area, were either at the average GAG or exceeded the maximum GAG values. Both of these were located in survey unit RB068. The surface activity levels were 9,400 dpm/100 cm2 (average GAG is 2 2 9,400 dpm/100 cm ) and 32,000 dpm/100 cm (maximum GAG is 28,300 dpm/100 cm2 ). LIPA personnel remediated each location and ESSAP then performed post-remedial action surface scans and direct measurements. These surface scans and measurements indicated that areas were adequately remediated. The surface activity levels at the other locations were less than the Shorcham Reactor Buildmg and Phase 4 - February 16, 1995 10 1
h
-applicAle guidelines but IlPA elected to remediate these areas as well. There were no other areas of elevated direct gamma or beta activity detected in the remaining confirmatory survey I units.
I l SURFACE ACTIVITY LEVELS 1 1 The data reported below and in Table 2 is the difference between the gross field sample counts and area background. The difference was then corrected for detector efficiency and geometry, _ l
- j. sample count time, and contributions fr'o m Fe-55, H-3, and Ni-63 as appropriate (when the net count rate exceeded the background distribution [LJ). Actual values are reported, including s -- negative surface activity levels, which occurred when the field count rate was less than the i
i background. Of 7% measurements,180 exceeded the 1,. t i i '
- The results of total and removable surface activity levels for each confumatory survey unit are ,
1
- summarized in Table 2. Total beta activity levels after additional remediation for the stmetural j survey units, excluding RWO73X02 and -X03 which are discussed below, ranged from -760 to
- 5,600 dpm/100 cm 2. The highest fmal direct measurement was located in the Spent Fuel Storage 2
. Pool (RB068) which had an average GAG of 9,400 dpm/100 cm . Prior to the additional e remediation, residual activity levels ranged up to 8,600 dpm/100 cm (average GAG of 2 2
11,100 dpm/100 cm2 ) and 32,000 dpm/100 cm (maximum GAG of 28,300 cpm /100 cm 2
)n l
{ survey units RB072X01 and RB068, respectively. Removable activity levels ranged from -1 to 8 dpm/100 cm2 for gross alpha and from -7 to 23 dpm/100 cm for gross beta. The mean 2 { j residual activity in stmetural survey units, as presented in Table 4, ranged from -340 to 450 dpm/100 cm 2and from -0'.7 to 3.5 dpm/100 cm2for total and removable gross beta activity, i respectively. The'mean activity levels, prior to additional remediation, were 670 dpm/100 cm2 ! in RB072X01 and 1,500 dpm/100 cm2 in RB068. I 1 f 2 Total beta activity in surveyed systems ranged from -680 to 2,500 dpm/100 cm . The removable { activity levels ranged from -1 to 5 dpm/100 cm2 for gross alpha and from -5 to 35 dpm/100 cm2 l for gross beta. The mean beta activity levels for systems ranged from -70 to 390 dpm/100 cm 2
- for total activity and from -0.3 to 2.8 dpm/100 cm2for removable gross beta activity (Table 5).
i ! se n.- sams .w n=a . ran-y 26. tws 11 a
. - .. ~
1 1 1 Total beta activity levels for the Bioshield Wall Blocks (RWO73X02 and -X03) ranged from -410 to 15,000 dpm/100 cm2. The average GAGS for these survey units were 95,900 dpm/100 cm 2 2 l for RWO73X02 and 76,900 dpm/100 cm for RW073X03. Removable activity ranged from -1 ( 2 2 to 1 dpm/100 cm for alpha and from -5 to 1/ dpm/100 cm for beta. The mean beta activity l levels were 6,400 and 55 dpm/100 cm2 for RWO73X02 and -X03, respectively. The mean gross beta removable activity was 0.4 dpm/100 cm2 for both units. , EXPOSURE RATES i : i i i The interior background exposure rates, as determined during the Turbine Building survey, j ranged from 4 to 5 R/h and averaged 5 R/h at 1 m (Vitkus 1994a). Individual gross exposure ; rate's within the Reactor Building and adjacent to the Bioshield Blocks ranged from 2 to 6 R/h at 1 m (Table 3). The net mean exposure rates for the stmetural survey units, ranged from I l -3 to 0.9 pR/h (Table 4). i l RADIONUCLIDE CONCENTRATIONS IN MISCELLANEOUS SAMPLES Cobalt-60 concentrations in the Porous Concrete Sump water sample and the Reactor Building l Truck Bay trench residue samples were <4.2 pCi/l and <0.2 pCi/g, respectively. No other
- significant photopeaks were identified.
i i COMPARISON OF RESULTS WITH GUIDELINES i i i The confirmatory survey results were compared with both the data provided by llPA and the
- i. NRC guidelines for release to unrestricted use. The NRC's Regulatory Guide 1.86 provides the i guidelines for acceptable surface contamination levels used to determine whether a licensed facility may be released to unrestricted use. These guidelines are summanzed in Appendix C.
4 l j The applicable guidelines are those for beta-gamma emitters and the alpha contamination guidelines are those for uranium and associated decay products. The beta-gamma guidelines are: l 3 - i I I i l Shoreham Reacor Buildmg and Phase 4. February 16, 1995 12 i _. --. .- - - - . . _ . .1
.- - ~ . . - . . . . . . . . . - .
?
- i
.l Total Activity 5,000 B-y dpm/100 cm 2, averaged over 1 m 2 ,
2 2 15,000 8-y dpm/100 cm , maximum in a 100 cm ,,,, , I Removable Activity l 2 1,000 8-y dpm/100 cm , The alpha guidelines are: j 2 i Total Activity- ' i ' 2 2 5,000 a dpm/100 cm , averaged'over 1 m 15,000 a dpm/100 cm , maxunum m a 100 cm2 area 2 i t i 3 i !
~
Removable Activity l j 1,000 a dpm/100 cm2 In addition, the NRC has approved site-specific surface contamination guidelines for H-3 and j Fe-55, particularly in activated concrete and steel (Pittiglio 1994). These guidelines are: b f Total Activity l 200,000 dpm/100 cm2, averaged over 1 m2 3 2 2 ] 600,000 dpm/100 cm , maximum in a 100 cm ,,,, 5 Removable Activity I 1,000 dpm/100 cm2 p. For those survey units involving mixtures of radionuclides LIPA substituted the gross activity ) guideline (GAG) value, which is based on the " sum of fractions rule" found in Appendix A of
- Draft NUREG/CR-5849, rather than the above generic NRC guidelines (LIPA 1994a; Berger 1992). These confirmatory survey units, together with the GAG values that the data were
[ . compared with and tested against, are listed below: 1-t Shoreharn Reactor Building and Phase 4 February 16,1995 13 1 n
1 1 SU001. Reactor Accamhly I.ower Bowl and RB072x01. Diver /hnarator Pool . .\
.2
! 11,100 dpm/100 cm , averaged over 1 m 2 33,000 dpm/100 cm2 , maximum l !' i t 1 l RWO73 x02. Bioshield Concrete j 2 2 95,900 dpm/100 cm , averaged over 1 m l t
- 287,700 dpm/100 cm2-, maximum i
RWO73 x03. Bioshield Steel !
- - 76,900 dpm/100 cm2, averaged over 1 m2 t'
j 230,700 dpm/100 cm2,' maximum ! .- t RB068. Spent Fuel Pool l h 9,400 dpm/100 cm2, averaged over 1 m2 ) t I j 28,300 dpm/100 cm2 , maximum l: l l Removable activity guidelines for each of the above remained at 1,000 dpm/100 cm2 , ! The exposure rate guideline currently being used by the NRC is 5 pR/hr above background, i measured at I m above the surface (U.S. NRC 1984). t i I As previously discussed, the detection sensitivities of the field instruments are such that the insidual Fe-55 activity cannot be detected. In addition,11PA determined that residual H-3
- l. n.d/or Ni-63 activity also must be accounted for within several survey units (LIPA 19946).
I Therefore for those surveys that LIPA identified as such, total and removable surface activity j measurements were corrected for Fe-55, H-3, and/or Ni-63. The mean of the corrected surface ! activity level for each survey unit was then calculated and the survey unit data was tested at the 95% confidence level (upper confidence level [UCL]), relative to the guidelines, in accordance with Draft NUREG/CR-5849 (Berger 1992). These results are provided in Table 4. The l ESSAP and LIPA data sets were then compared and evaluated relative to the established
- conditions.
l l sn-.nno nemor sundins and n=e 4. February is, i,25 14 t
, , . . , _ , m,s.,,, . . , ,,,--rre--w~ w--T-"*
t A comparison of the ESSAP mean surface activity levels to the UPA mean activity levels j showed that the ESSAP mean was statistically less than or equal to the respective mean j determined by UPA for 24 of the 30 confirmatory survey units, thereby satisfying the data . cvaluation and comparison conditions. However, the ESSAP mean surface activity value was f statistically greater than the UPA mean for survey units PC005, PC007, PC012, RB072X01, i RWO73X02 and SUO60; therefore, additional evaluation was necessary. f l i For the three Primary Containment survey units (PC005, PC007, and PC012) where the ESSAP i surface activity mean was higher than the UPA mean, UPA qualified the release record data as containing excessive negative measurements. According to UPA, this was due to observed, f background levels being lower than the generic site backgrounds that were used for surface . activity measurement conversions.' As a result, the mean survey unit total activity levels were biased low. The maximum surface activity level obtained by ESSAP for these survey units was 910 dpm/100 cm2 in PC007, and the maximum survey unit mean and UCL, found in PC005 ; 2 were 310 dpm/100 cm2 and 350 dpm/100 cm , respectively. In addition, the ESSAP total surface activity mean was less than or only slightly higher than the detection limits of the instrumentation, in all three units. Based on these factors, it is ESSAP's opinion that for the i above survey units, the observed difference in means would not affect the conclusions UPA , reached for remaining survey units. i i In survey unit RB072X01, the ESSAP surface activity mean of 450 dpm/100 caf exceeds the
)
2 UPA mean of 140 dpm/100 cm . It should be noted that ESSAP performed 5 of the 30 direct l l measurements at locations of elevated direct radiation identified by surface scans. The surface , 2 i activity at these locations ranged from 1,800 to 4,000 dpm/100 cm . Inclusion of these I 1 ! measurements in the determination of the average surface activity level results in a mean that - i is biased high and may not be representative of the average survey unit residual activity level l which UPA reported, i The ESSAP mean for survey unit RWO73X02 was 6,400 dpm/100 cm2 versus the UPA mean of 2,200 dpm/100 cm 2. To the extent practical, ESSAP selected direct measurement locations on the concrete portion of the shield blocks on the basis of the maximum direct radiation levels Shoreham Reactor Building and Phase 4 - February 16. Im 15
)I h
- noted while performing surface scans. UPA selected measurement locations systematically. I i
Therefore, similar to RB072X01, the ESSAP mean for this survey unit is high relative to the : UPA determined average, and the difference in means in not considered significant when ! I
- compared to the applicable guidelines (average GAG of 95,900 dpm/100 cm2 ) for this survey h
unit. ! 1 i l 2 For survey unit SUO60, the ESSAP total surface activity mean was 360 dpm/100 cm with a i maximum level of 2,500 dpm/100 cm2 . The corresponding LIPA values were 174 dpm/100 cm2 and 3,247 dpm/100 cm2 . UPA had not qualified the data for this unit as containing excessive i negative values. Therefore, the UPA surface activity levels, mean activity level, and UCL were evaluated to determine the potential impact on the UPA reported status of this and the other f , Phase 4 survey units, relative to the guidelines. The difference between the ESSAP and UPA j 2 , means was 213 dpm/100 cm . If this difference in activity levels were applied to each Phase ! i 4 survey unit, overall surface activity levels would not be significantly altered and the j i conclusions reached, that the radiological status of this and all other Phase 4 survey unit satisfy l ! the guidelines, would remain valid. I i Survey unit RB068, the spent storage pool, was an alpha affected area. Although the total
- activity levels provided in the Tables of this report indicate beta activity, the instrumentation ESSAP used in this survey unit detects both alpha and beta radiation. ESSAP evaluated the confirmatory data developed for this unit, relative to the potential alpha radiation contribution l at the location with the highest surface activity level. If all of the net activity present at this j location was to be attributed to alpha contamination, the maximum alpha surface contamination level in the survey unit would be approximately 2,600 dpm/100 cm 2, which is below the average alpha activity guideline of 5,000 dpm/100 cm 2-4 1
Comparison of the ESSAP survey data directly with the guidelines showed that overall, total surface activity levels within each survey unit satisfied the guidelines at the 95 % confidence level ! (p, or UCL). The maximum UCL was 2,900 dpm/100 cm2 n survey unit RWO73X02. This comparison also included an evaluation and data testing of the preremediation surface activity levels in survey units RB068 and RB072X01. The evaluation determined that prior to the Sbrcham Reactor Building and Phase 4 - February 16, 995 16 4 I J _ , . . _ . . . ~ . _ . m, - . . .
- c. _ _, , _ _ , . ,m.. m, _ _ , , , .- ,-i
4 i additional remediation the average GAGS (9,400 dpm/100 cm2 for RBM8 and. 11,100 dpm/100 cm2 for RB072X01) were satisfied at the 95% confidence level and no i additional measurements were required to demonstrate compliance. The preremediation UCLs ! were 3,100 dpm/100 cm2 for RB068 and 1,200 dpm/100 cm2 for RB072x01. There were no 3 4 final direct measurements that exceeded the applicable NRC generic or site-specific GAG average guideline values. All removable activity was below the guideline at the 95% confidence h level. 1 i Exposure rates were compared with those obtained by LIPA, and tested at the 95% confidence l level, relative to the 5 pR/h above background guideline currently being used by the NRC. The 1 Reactor Building and Bioshield Block exposure rates were comparable to background exposure rate levels and confirmed the findings presented by UPA. i 1
- There was no detectable Co-60 in the sediment sample and therefore the site-specific guideline of 8 pCi/g is satisfied. There is not a site-specific guideline for Co-60 in water for comparison; however, the water sample collected did not contain a detectable level of Co-60.
l l
SUMMARY
i i The Environmental Survey and Site Assessment Program performed confirmatory survey
- I g activities for the Reactor Building at Shoreham Nuclear Power Station in Brookhaven, New York. Confirmatory activities included document reviews, and during the period of October 31 l.
through November 3,1994, ESSAP performed independent surface scans, surface activity l l measurements, exposure rate measurements, and miscellaneous sampling. j i l
- l i i 4
ESSAP's confirmatory survey results support the data presented in the UPA termination survey I e report. These results indicate that total and removable surface activity levels and exposure rates l l were below the NRC guidelines for release to unrestricted use. Statistical tests of data sets l l further support the conclusion that each survey unit satisfies the guidelines at the 95 % confidence I level. l i j Shoreham Reactor Buildmg and Plase 4 - February 16. IM 17 I i
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273-035 (2) e w.,;. g I i ELEVATION 63'- 0" l I I (" an 0 +
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I 273-037 (2) l l-l- RB057 I 1
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. FIGURE 6: Reactor : Building, Floor Plans of 150'-9" and 175'-9" Elevations -
Structural Units Surveyed mm m .a no. 4. r*n-y is, im 23 j i
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! . 7F i g SINGLE-POINT l LOWER WALLS AND FLOOR l g SINGLE-POINT d l UPPER WALLS AND CElUNG O FEET 6 l 4 EXPOSURE RATE m3 l OMETERS 2 l l FIGURE 7: Reactor Building, Primary Containment 63' NW G/A (PC004) - Measurement and Sampling Locations m a cw naag .no mn 4. rebmuy is, ses 24
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l FIGURE 14: Reactor Building, 40' SW G/A (RB014) - Measurement and Sampling Locations ! sm a cw auums a mic 4. Febm u.1995 31 1 l l i
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$ EXPOSURE f%TE o 4 METERS FIGURE 18: Reactor Building, RMCU Regen /Non-Regen HTX's Room (RB061) -
Measurement and Sampling Locations l Shorchem Reactor Building and Phaic 4. February 16.1995 35
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I O 4 1, METERS l l FIGURE 19: Reactor Building, Spent Fuel Storage Pool, Floor, North and West Wolls l (RB068) - Measurement and Sampling Locations Shoreham Reactor BuMmg and l'hase 4. February 16.1995 36 i
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$borcham Reactor Duildag and Phase 4. February 16,1995 42
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MGURE 30:MeasurementRodwoste Building, Locat ions 52' SE G/A Shield Blocks (RWO73X02 and XO3) - f Shoreham Reackw BuMmg and Phase 4 - February 16, IM5 47 l J j
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;. m m cun notiai o en e 4. February 16.1995 49 4
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Measurement and Sompting Locations
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Measurement and Sampling Locations she ne== se.: na rho 4 - redn-y a ms 52 I, . _ _ . . . , . . . _
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SU058-11-1 SUO58-6-1 .] SUPPRESSION POOL SUPPRESSON POOL SUPPRESSION POOL - MEASUREMENT / SAMPLING LOCATIONS NOT TO SCALE I g SINGLE-PotNT FIGURE 36: Reactor Building, Reactor Primary Containment Components (SUO58);- Measurenment and Sampling Locations shonhams Reactor Buddag and Phase 4. Feenery 16,1995 53 l-l .
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TABLE 1 REACTOR BUILDING SURVEY UNITS SELECTED FOR CONFIRMATORY SURVEY SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Survey Affected (A)/ Structure / Unit / Component " 'I ""' Unaffected (U) System PC004 Primary Containment - 63' NW A stmeture PC005 Sub-Pile Room A stmeture PC007 Primary Containment - 78' NE A structure PC008 Primary Containment - 78' SE A stmeture PC012 Primary Containment - 109' NE A structure RB004 Reactor Building - 8' SE A structure RB006 Reactor Building - 8' SW A structure RB014 Reactor Building - 40' SW A structure RB037 Reactor Building - 78' SW A stmeture RB038 West Accumulator Aisle A stmeture RB057 Fuel Pool Clean-up Pumps Room A structure RB061 RWCU Regen /Non-Regen HTX's Room A structure RB068 Spent Fuel Storage Pool A stmeture RB071 x01 Reactor Cavity - 150' A stmeture RB072x01 Dryer / Separator Storage Pool a stmeture RB103 Reactor Building - 175' A stmeture RB106 Reactor Building - 175' A stmeture RB109 Polar Crane A structure RWO73 x02 Bioshield Wall Blocks (Concrete) A structure RWO73 x03 Bioshield Wall Blocks (Steel) A structure SU001 Reactor Assembly 1.ower Bowl A system SU002 Nuclear Boiler Main Steam Relief Valve A system 1B21-RV-095G Shoreham Reactor Buildmg and Phase 4 - February 16, IM 57
TABLE 1 (Continued) REACTOR BUILDING SURVEY UNITS SELECTED FOR CONFIRMATORY SURVEY SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Survey Affected (A)/ Structure / Unit /Ceponent EN ""'~ Unaffected (U) System SU004 CRD Hydraulic Control A system SU014x05 #953 Misc. Embedded Drain Piping A system 1G11-TK-190 SU014x12 RB Porous Concrete Sump #P-224A A system SU023 Misc. Building Storm Drains - A system SUO58 Reactor Prunary Containment A system SUO60x09 Fuel Pool Cooling Room A system SUO60x13 RWCU Valve Chambers Room A system SUO60x23 Reactor Building ~ Air from Drywell A system IT46-ADV-039A SUO60x24 Reactor Building Vent Dump ' A system i l I j l Shortham Reactor Building and Phase 4. February 16. Iws 58
SUMMARY
OF SURFACE ACTIVITY LEVELS REAC1DR CUILDING { SHOREHAM NUCLEAR POWER STATION 3 BROOKHAVEN, NEW YORK l' Number of Total Activity Range Removable Activity I Location" Measurement (dpm/100 cm ) . 2 Range (dpm/100 cm2) 1 Locations Betabd Alphad Beta' f Reactor Building PC004 Primary Containment 30 -230 to 410 -1 to 1 -5 to 14 y PC005 Primary Containment, Sub-Pile Room 30 57 to 580 -1 to 3 -4 to 18 g PC007 Primary Containment 30 -340 to 910 -1 to 5 -4 to 10 PC008 E. ry Containment 30 -420 to 270 -1 to 3 -5 to 6 PC012 Primary Containment 30 -100 to 530 -1 to 6 -4 to 9 g RB004 Reactor Building 30 -760 to -9 -1 to 3 -7 to 3 RB006 Reactor Building 30 -640 to -150 . -1 to 3 -4 to 8 RB014 Reactor Building -30 -220 to 370 -1 to 3 -5 to 10 RB037 ' Reactor Building 30 -470 to -61 -1 to 5 -3 to 12 RB038 West Accumulator Aisle 30 -670 to 580 -1 to 6 -5 to 10 RB057 Fuel Pool Clean-up Pumps 30 -450 to -22 -1 to 8 -5 to 9 RB061 RWCU HTX's Room 30 ~ 570 to 330 .-1 30 1 -5 to 5 RB068 Spent Fuel Storage Pool 34 (-300 to 32,00078 -1 to 5 ~ ' 3 to 21
-300 to 5600 RB071 x01 Reactor Cavity 30 -470 to 680 -1 to 3 -7 to 23 RB072x01 Dryer / Separator Storage Pool 30. (-390 to 8600f" -1 to 6 -3 to 21
-390 to 4000 RB103 Reactor Building 30 -500 to 530 -1 to 3 -5 to 12 RB106 Reactor Building 30 -190 to 550 -1 to 3 -4 to 6 RB109 Polar Crane and Crane Rail Area 30 -570 to 770 -1 to 3 -3 to 14
SUMMARY
OF SURFACE ACTIVITY LEVElcS REACTOR BUILDING I SIIOREIIAM NUCLEAR POWER STATION , BROOKHAVEN, NEW YORK l l h A Number of Total Activity Range Removsbie Activity I Location" Measurement (dpm/100 cm2) Range (dpm/100 cm') Locations PL. .6
- Alpha d Beta' A
SU001 Reactor Assembly Lower Bowl 30 -160 to 2200 -1 to 3 -5 to 10 f SUOO2 Nuclear Boiler System 5 -38 to 340 -1 to 1 -3 to I h SUOO4 Hydraulie Control System 18 -460 to 300 -1 to 5 -5 to 3 3 SUO14 x 12 Sump P-224A 6 -110 to 730 -1 to 1 -1 to 5 SU0l4 x05 Embedded Drain Piping 10 -380 to 340 -1 to 1 -4 to 5 SUO23 Misc. Building Storm Drains 26 -680 to 2400 -1 to 1 -5 to 13 g; SUO58 Primary Containment Valves 17 -190 to 1000 -1 to 5 -4 to 35 SUO60 Reactor Building Ventilation 45 38 to 2500 -1 to 5 0 to 35 SUO61 Stand-By Ventilation 19 -76 to 590 -1 to 5 -3 to 12 Radwaste Building RWO73 x02 Bioshield Wall Blocks (Concrete) 25 -130 to 15,000 -1 to 1 -5 to 17 RWO73x03 Bioshield Wall Blocks (Steel) 25 -410 to 66001 -1 to 1 -5 to 17
' Refer to Figures 7 through 39.
b MDAs = 200 to 1200 dpm/100 cm2 .
- Activity levels corrected for Fe-55, H-3, and Ni-63 as appropriate.
d MDA = 12 dpm/100 cm2,
'MDA = 16 dpm/100 cm 2, f
Represents surface activity levels prior to additional remediation. 2 8 GAG values are 9,400 dpm/100 cm average and 28,300 dpm/100 cm2 maximum. h GAG values are 11,100 dpm/100 cm2 average and 33,300 dpm/100 cm2 maximum. i GAG values are 95,900 dpm/100 cm2 average and 287,700 dpm/100 cm2 maximum. i GAG values are 76,900 dpm/100 cm2 average and 230,700 dpm/100 cm2 maximum.
TABLE 3
' EXPOSURE RATES REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK
- h. Number of Measurement Net Exposure Rate Range IAcations at I m GdUh)*
Reactor Building PC004 10 -3 to -2 PC005 8 -1 PC007 7 -3 PC008 8 -3 PC012 2 -2 RB004 10 -2 to -1 RB006 4 -2 to -1 RB014 10 0 to 1 RB037 9 -1 RB038 6 -1 RB057 6 -1 to 0 RB061 10 -1 to 0 RB068 9 -1 to 1 RB071 x01 NA6 NA RB072x01 9 0 to 1 RB103 17 0 to 1 RB106 9 0 to 1 RB109 NA NA-Radwaste Building RWO73 x02 5 1 RWO73 x03 5 0
- Refer to Figures 7 through 18, 23, 24, 25, and 30.
b S te background exposure rate is 5 R/h.
- NA=Not Applicable, measurements not performed.
sn-.6.m ac== suiioine .no th 4 - r bruary 26. im 61
TABLE 4 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit Su- ari y . Total Beta Activity (dpm/100 cm')*
# of Direct Measurements 30- 30 30 30 30 Mean(X) 36 310 100 -120 23 LIPA X -38 -26 -78 70 -144.4 #=
86 -350 '189 -73 63 Generic Guidelines or GAG Satisfied - Yes Yes Yes Yes Yes Removable Beta Activity (dom /100 cm') b
# of Smears 30' 30 30 '30 30 Mean(X) 0.3 0.1 0.7 -0.4 -0.2 ; IlPA X- 7.0 9.9 2.7- 4.8 7.4 F=
1.4 1.5 1.2 0.5 0.6 1,000 dpm/100 cnf Guideline Satisful Yes Yes Yes Yes Yes Frpnanre Rates at 1 m (gR/h)
# of Exposure Rate Measurements 10 8 7 8 2 Net Mean (X) -2.6 -1.2 -3.0 -2.8 -2.4 IlPA X -1.0 -0.3 -0.5 -0.9 0.2 F=
-2.5 -1.2 -2.9 -2.8 -1.8 5 R/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied Shoreham Reactor Buildmg and Phase 4 - February 16. Im 62
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES
; REACIDR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit Summary RB004 RB006 RB014 RB037 RB038 2
Total Beta Activity (dpm/100 c n)"
# of Direct Measurements 30 30 30 30 30 Mean (X) -290 -340 150 -210 -190 LIPA X -48 -220 56 47 86 F=
-250 -300 190 -180 -110 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes 2
Removable Beta Activity (dpm/100 cm) ,b
# of Smears 30 30 30 30 30 Mean(X) -0.7 1.1 0.7 1.5 0.1 LIPA X 3.1 6.6 5.1 5.5 5.1 Fa 0.1 1.9 1.8 2.7 1.0 1,000 dpm/100 cm Guideline Satisfied 2
Yes Yes Yes Yes Yes Emm_ ire Rates at 1 m (gR/h)
# of Exposure Rate Measurements 10 4 10 9 6 Net Mean (X) -1.4 -1.7 -0.7 -0.9 -1.1 LIPA X -0.3 0.1 -0.2 -0.1 -0.6
#=
-1.3 -1.1 -0.7 -0.7 -0.9 5 R/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied Shacham Remm Buuding and Phase 4 - February 16. Im 63
l TABLE 4 (Continued) CGNFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit
""""U RB057 RB061 RB068 RB071x01 RB072x01 Total Beta Activity (dpm/100 cm')" # of Direct Measurements 30 30 34 30 30 Mean(I) -240 -170 (1500)* 43 (670)'
M 42 IlPA X 75 55 290 36 140 F=
-210 -110 (3100)' 60 (1200)*
250 760 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes Removable Beta Activity (dpm/100 crd)*'
- of Smears 30 30 34 30 30 Mean(I) 0.2 -0.6 2.7 1.5 3.5 LIPA X 3.6 9.2 17 4.6 27 F=
1.5 0.2 4.3 3.4 5.3 1,000 dpm/100 cmr Guideline Satisfied Yes Yes Yes Yes Yes Fmna_ ire Rates at 1 m (;4R/h)
- of Exposure Rate Measurements 9 10 9 NA 0.4 Net Mean (X) -0.8 -0.7 0.0 NA 0.4 IlPA X 0.0 -0.1 0.0 -0.3 0.1 Fa
-0.6 -0.6 0.2 NA 0.6 5 pR/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied Shoreham Reactor Building and Phase 4. February 16.1995 64
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit
"""U RB103 RB106 RB109 RWO73x02 RWO73x03 2
Total Beta Activity (dpm/100 cm)*
# of Direct Measurements 30 30 30 25 25 Mean(X) 130 79 100 6400 55 LIPA X 64 55 85 2200 2700 Fa 210 130 210 2900 520 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes Removeble Beta Activity (dpm/100 cm)O 2 # of Smears 30 30 30 25 25 Mean(X) 0.4 1.2 0.8 0.4 0.4 LIPA X 7.2 4.9 6.8 15 5.1 Fa 1.7 2.0 1.9 1.8 1.8 1,000 dpm/100 cm Guideline Satisfied 2
Yes Yes Yes Yes Yes Exposure Rates at I m (pR/h)
- of Exposure Rate Measumments 17 9 NA 5 5 Net Mean (X) 0.5 0.5 NA 0.9 0.9 LIPA X -0.3 -0.2 -0.8 0.3 0.4 Fa 0.5 0.7 NA 1.2 1.2 5 pR/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied Activity levels corrected for Fe-55, H-3, and Ni-63 as appropriate.
All alpha removable activity was less than 12 dpm/100 cm 2~ Represents survey unit radiological status prior to additional remediation. Shorcham Reactor Buildmg and Ptase 4 - February 16. tw5 65
l TABIE 5 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS REAC'IDR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit N SU001 SU002 S0004 SU014x12 SU014x05 Total Beta Activity (dpm/100 cuf)*
# of Dimet Measumments 30 5 18 6 10 Mean(X) 170 110 -70 130 ' 65 LIPA I 470 260 210 960 -130 M=
250 260 4 400 110 Genenc Guidelines or GAG Satisfied Yes Yes Yes Yes Yes Removable Beta Activity (dpm/100 cmh'*
- of Smears 30 5 9 5 8 Mean(I) 0.6 0.6 -0.2 0.2 0.2 IlPA X 14 5.2 15 13 14 Fa 1.7 2.0 1.6
_2.7 2.4 Conditions and 1,000 dpm/100 cnf Yes Yes Yes Yer Yes Guideline Satisfied Shoreharn Reactor Building and Phase 4. February 16. 1995 66
TABLE 5 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS REACIOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit Summary SUO23 SUO58 SUO60 SUO61 Total Beta Activity (dpm/100 cm')*
# of Direct Measurements 26 17 45 19 Mean(X) 50 32 390 -280 LIPA 240 210 170 590
- g. 240 150 530 360 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Removable Beta Activity (dpm/100 cm')*
# of Smears 15 17 45 19 Mean(X) -0.3 2.2 2.8 1.1 LIPA I NA 7.4 17 18 Fa 17 6.1 4.5 2.6 1,000 dpm/100 cm' Yes Yes Yes Yes Guideline Satisfied Activity levels corrected for Fe-55, H-3, and Ni-63 as appropriate.
'All alpha removable activity was less than 12 dpm/100 cm2 , l Stwcham Reactor Building and Phase 4 - February 16, IM ' 67
REFERENCES Berger, J. D. Manual for Conducting Radiological Surveys in Support of License Termination, Draft, NUREG/CR-5849. Prepared by Oak Ridge Associated Universities, Oak Ridge, Tennessee for the Nuclear Regulatory Commission. June 1992. Iong Island Lighting Company, Brookhaven, New York. Shoreham Nuclear Power Station Site Characterization Program Final Report. 1990. Iong Island Power Authority, Brookhaven, New York. Shoreham Decommissioning Project, Termination Survey Plan, Revision 3. July 1994a. Iong Island Power Authority, Brookhaven, New York. Shoreham Decommissioning Project Termination Survey Final Report, Volumes 1 and 2. October 1994b. Pittiglio, C. L. ? etter to A. J. Bortz, long Island Power Authority. June 7,1994. U.S. Nuclear Commission. Guidance and Discussion of Requirements for an Application to Terminate a Non-Power Reactor Facility Operating License, Revision 1. September 1984. Vitkus, T. J. Ietter to D. Fauver, U.S. Nuclear Regulatory Commission. November 4,1993. Vitkus, T. J. Confirmatory Survey of the Turbine Building, Site Grounds, and Site Exteriors, Shoreham Nuclear Power Station, Brookhaven, New York. Oak Ridge Institute for Science and Education, Oak Ridge, Tennessee. September 1994a. Vitkus, T. J. Ietter to D. Fauver, U.S. Nuclear Regulatory Commission. October 25,1994b. t Shoreham Reactor Buildmg and Phase 4 - February 16, IM 68
+
1 APPENDIX A ' MAJOR INSTRUMENTATION Shoretuun Reactor Buildmg and Phase 4 - February 16.19A
APPENDIX A MAJOR INSTRUMENTATION The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the authors or their employers. DIRECT RADIATION MEASUREMENT Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM) Eberline " Rascal" Ratemeter-Scaler Model PRS-1 (Eberline, Santa Fe, NM) Imdlum Ratemeter-Scaler Model 2221 (Ludlum Measurements, Inc., Sweetwater, TX) Detectors Eberline GM Detector Model HP-260 Effective Area,15.5 cm2 (Eberline, Santa Fe, NM) Eberline ZnS Scintillation Detector Model AC-3-7 Effective Area,59 cm2 (Eberline, Santa Fe, NM) Imdlum Gas Proportional Detector Model 43-37 Effective Area,550 cm2 (Imdlum Measurements, Inc., Sweetwater, TX) Shoreham Reanor Buildmg aiut Phase 4 - February 16,1995 A-1
i Ludlum Gas Proportional Detector Model 43-68 Effective Area,100 cm2 (Ludlum Measurements, Inc., Sweetwater, TX) Reuter-Stokes Pressurized Ion Chamber Model RSS-111 (Reuter-Stokes, Cleveland, OH) Victoreen Nal Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal (Victoreen, Cleveland, OH) LABORATORY ANALYTICAL INSTRUMENTATION High Purity Extended Range Intrinsic Detectors Model No: ERVDS30-25195 (Tennelec, Oak Ridge, TN) Used in conjunction with: Izad Shield Model G-11 (Nuclear lead, Oak Ridge, TN) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT) High-Purity Germanium Detector Model GMX-23195-S,23% Eff. (EG&G ORTEC, Oak Ridge, TN) Used in conjunction with: lead Shield Model G-16 (Gamma Products, Palos Hills, IL) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT) I.ow Background Gas Proportional Counter Model LB-5100-W (Oxford, Oak Ridge, TN) Shoreham Reactor Buildmg aral Phase 4 - February 16, IM5 A-2
APPENDIX B SURVEY AND ANALYTICAL PROCEDURES Shoreham Reactor Buildmg anxl Phase 4 February 16. 1995
APPENDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES Surface Scans Surface scans were performed by passing the probes slowly over the surface; the distance berwaa the probe and the surface was maintainul at a minimum - nominally about I centimeter (cm). A large surface area, gas proponional floor monitor was used to scan the floors of the 2 surveyed areas. Other surfaces were scanned using small area (15.5 cm ,59 cm2 or 100 square 2 centimeters [cm ]) hand-held detectors. Identification of elevated levels was based on increases in the audible signal from the recording and/or indicating insuument. Combinations of detectors and instruments used for the scans were: Alpha - gas proponional detector with ratemeter-scaler ZnS scintillation detector with mtemeter-scalcr Beta - gas proportional detector with ratemeter-scaler GM pancake detector with ratemeter-scaler Gamma - NaI scintillation detector with ratemeter Surface Activity M - --- := Measurements of total beta activity levels were performed using Geiger-Mueller (GM) and gas proponional detectors with portable ratemeter-scalers. Count rates in counts per minute (cpm), which were integrated over 1 minute in a static position, were convened to activity levels in disintegrations per minute /100 square centimeters Shoreham Reactor Buildmg and Phase 4 February 16, im B-1
2 (dpm/100 cm ) by dividing the net rate by the 4 x efficiency and correcting for the active area of the detector. The beta activity background count rates for the GM and gas proportional detectors ranged from 15 to 37 cpm and from 100 to 207 cpm, respectively. Beta efficiency factors were 0.17 for the GM detectors and ranged from 0.21 to 0.23 for the gas proportional detectors. The effective window areas for the GM and the gas proportional detectors were 15.5 2 cm and 100 cm 2, respectively. Surface activity measurements in all confumatory survey units, except for RB068, RWO73 x02, RWO73 x03, and SU001 that exceeded the normal background distribution were corrected for the iron-55 contribution by multiplying the dpm/100 cm2 field activity level by a factor of 1.2. For the remaining survey units, correction factors for Fe-55, H-3, and/or Ni-63 were as follows: RB068, measured dpm/100 cm2 times 2.3 RWO73 x02, measured dpm/100 cm2 times 35.9 RWO73 x03, measured dpm/100 cm2 times 24.2 SU001, measured dpm/100 cm2 times 2.6 LIPA based each of these correction factors on the relative concentrations of contaminants in the construction material of the various survey units. The instmment response level at which the detector output could be considered above background was defined as the critical level (IJ. This level was defined for each detector / instrument combination as follows: 1.96 " "PI' '#""' '"
+ ackground count rate
, 8 Sample count time Background count time (Detector Efficiency) (Detector Geometry)
Removable Activity Measurements Removable activity levels were determined using numbered filter paper disks,47 millimeters (mm) in diameter. Moderate pressure was applied to the smear and approximately 100 cm 2 of Shorcham Reactor Building and Phase 4 - February 16. lW5 B-2
the surface was wiped. Smears were placed in labeled envelopes with the location and other pertinent information recorded. Exposure Rate Measurements Measurements of gamma exposure rates were performed using a' pressurized ionization chamber (PIC) set I meter from the surface. Water Sampling
- Approximately 3.8 liters of water was collected, the' sample.was transferred to a plastic corwainer, sealed, and labeled in accordance with ESSAP survey procedures.
Sediment Sampling Approximately I kilogram of sediment was collected from the trench, placed in a plastic container, sealed and labeled in accordance with ESSAP survey procedures. ANALYTICAL PROCEDURES Removable Activity Smears were counted on a low background gas proportional system for gross alpha, and gross beta activity. Gm= =a St+:1. = -iry Sediment and water samples were dried, mixed, crushed, and/or homogenized as necessary, and a portion sealed in 0.5-liter Marinelli beaker or other appropriate container. The quantity placed in the beaker was chosen to reproduce the calibrated counting geometry. Net material weights - were determined and the samples counted using intrinsic germanium detectors coupled to a pulse - height analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the Shoreham Reactor Buildirg and Phase 4 - February 16. Im B-3
analyzer system. The energy peak used for determining the activity of the radionuclide of concern was: Co-60 1.173 MeV Spectra were also reviewed for other identifiable photopeaks. UNCERTAINTIES AND DETECTION LIMITS The uncertainties associated with the analytical data presented in the tables of this report represent the 95% confidence level for that data. These uncertainties wem calculated based on both the gross sample count levels and the associated background count levels. Additional uncertainties, associated with sampling and measurement procedures, have not been propagated into the data presented in this report. Detection limits, referred to as minimum detectable activity (MDA), were based on 2.71 plus 4.66 times the standard deviation of the background count [2.71 + (4.66VBKG)]. ' Although data is reported as actual values, including negative values, in the document text and table, the MDAs for total and removable activity levels are provided in the footnotes of applicable tables. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides in samples, the detection limits differ from sample to sample and instrument to instrument. CALIBRATION AND QUALITY ASSURANCE Calibration of all field and laboratory instmmentation was based on standards, traceable to the National Institute of Standards and Technology (NIST), when such standard were available. In cases where they were not available, standards of an industry recogmzed organization were used. Calibration of pressurized ionization chambers was performed by the manufacturer. Analytical and field survey activities were conducted in accordance with procedures from the following documents of the Environmental Survey and Site Assessment Program: Shoreham Reactor Buildmg and Phase 4 - February 16. 1995 B-4
1
- Survey Procedures Manual, Revision 8 (December 1993)
-* 12boratory Procedures Manual, Revision 8 (August 1993)
+ . Quality Assurance Manual, Revision 6 (July 1993)
The procedures contained in these manuals were developed to meet the requirements of DOE Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to assess processes during their performance. Quality control procedures include:
- Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.
Participation in EPA and DOE /EML Quality Assurance Programs. Training and certification of all individuals performing procedures. Periodic internal and external audits. t Shorehara Reactor Buildag and Phase 4 - February 16,1995 B-5
APPENDIX C REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS Shoreharn Reactor Buildmg and Phase 4 - February 16, 1995
U.S. ATOMIC ENERGY COMMISSION June 1974 REGULATORY GUIDE DIRECTORATE OF REGULATORY STANDARDS REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION important to the safety of reactor operation is no longer required. Once this possession-only license is issued, Section 50.51, " Duration oflicense, renewal," of 10 reactor operation is not permitted. Other activities CFR Part 50, " Licensing of Production and Utilization from the reactor and placing it in storage (either onsite Facilities," requires that each license to operate a or offsite) may be continued. production and utilization facility be issued for a specified duration. Upon expiration of the specified A licensee having a possession-only license must period, the license may be either renewed or termmated retain, with the Part 50 license, authorization for by the Commission. Section 50.82, " Applications for special nuclear material (10 CFR Part, 70, "Special termmation oflicenses," specifies the requirements that Nuclear Material"), byproduct material (10 CFR Part must be satisfied to terminate an operating license, 30, " Rules of General Applicability to Licensing of including the requirement that the dismantlement of the Byproduct Material"), and source material (10 CFR facility and disposal of the component parts not be Part 40, " Licensing of Source Material"), until the inimical to the common defense and security or to the fuel, radioactive components, and sources are removed health and safety of the public. This guide describei, from the facility. Appropriate admmistrative controls methods and procedures considered acceptable by the and facility requirements are imposed by the Part 50 Rigulatory staff for the termmation of operating license and the technical specifications to assure that licenses for nuclear reactors. The advisory Committee proper surveillance is performed and that the reactor on Reactor Safeguards has been consulted concerning facility is maintamed in a safe condition and not this guide and has concurred in the regulatory position. operated. B. DISCUSSION A possession-only license permits various options and procedures for decommissioning, such as When a licensee decides to terminate his nuclear mothballing, entombment, or dismantling. The reactor operating license, be may, as a first step in the requirements imposed depend on the option selected. process, request that his operating license be anaded to restrict him to possess but not operate the facility. Section 50.82 provides that the licensee may
'Ibe advantage to the licensee of converting to such a dismantle and dispose of the component parts of a possession-only license is reduced surveillance nuclear reactor in accordance with existing regulations.
requirements in that periodic surveillance of equipment For research reactors and critical facilities, this has USAEC REGULATORY GUIDES g,,,,,,,, ,,,, ,, neguietory c des are usued to desenbe and make avoiiable to the pubhc **"'ed to m vs. Am enway canen.sm, wesen, o.c. 20sn methods acceptable to the AEC reguietary staff of implement ng specific parts Attenkn: hw W %wy stands 4 Cornnwnts and suggesh tw as the commasson's reguistoons. to arsgenets technsoues used by the staff in knptwenents m these glades are encouraged and should be sent to the evaluating speciflC prov ' " "
.pp.ce,,s egee,wyproblem.s a .reorr,ot postuiated sub.t t.esscendents, o
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- 2. Rosench and test nonctors 7. Transportation s and o efle new enf expor ence Note: Section electronically reproduced from photocopy. C-l
nsually meant the disassembly of a reactor and its b. A description cf measures that will be taken to l shipment organization f:r further use. The site from prevent criticality or reactivity changes and to g which a reactor has been removed must be minimize releases of radioactivity from the facility. y decontaminated, as necessary, and inspected by the Cornmission to determine whether unrestricted access c. Any proposed changes to the technical can be approved. In the case of nuclear power specifications that reflect the possession-only facility I reactors, dismmtling has usually been accomplished by status and the necessary disassembly / retirement shippiog fuel offsite, making the reactor inoperable, activities to be performed. j and disposing of some of the radioactive components. 7 d. A safety analysis of both the activities to be Radioactive components may be either shipped accomplished and the proposed changes to the off-site for burial at an authorized burial ground or technical specifications. l secured on the sne. Those radioactive materials remaining on the site must be isolated from the public e. An inventory of activated materials and their by physical barriers or other means to prevent public location in the facility, access to hazardous levels of radiation. Surveillance is necessary to assure the long term integrity of the 2. ALTERNATIVES FOR REACTOR barriers. The amount of surveillance required depends RETIREMENT upon (1) the potential hazard to the health and safety of the public from radioactive material remammg on the Four alternatives for retirement of nuclear reactor I site and (2) the integrity of the physical barrien. facilities are considered acceptable by the Before areas may be released for unrestricted use, they Regulatory staff. These are: g must have been deconenmmated or the radioactivity l must have decayed to less than prescribed limits a. Mothballing. Mothballing of a nuclear reactor (Table 1). facility consists of putting the facility in a state of protective storage, in general, the facility may be The hazard associated with the returned facility is left intact except that all fuel assemblies and the evaluated by considering the amount and type of radioactive fluids and waste should be removed remaining contamination, the degree of confinement of from the site. Adequate radiation monitoring, the remaining radioactive materials, the physical environmental surveillance, and appropriate security security provided by the confinement, the susceptibility procedures should be established under a to release of radiation as a result of natural phenomena, possession-only license to ensure that the health and and the duration of required surveillance. safety of the public is not endangered. C. REGULATORY POSITION b. In-Place Entombment. In-place entombment consists of sealing all the remammg highly [ 1. APPLICA*IiON FOR A LICENSE TO POSSESS radioactive or contaminated components (e.g., the BUT NOT OPERATE (POSSESSION-ONLY pressure vessel and reactor internals) within a LICENSE) structure integral with the biological shield after having all fuel assemblies, radioactive fluids and A request to a nend an operating license to a wastes, and certain selected components shipped possession-only license should be made to the Director offsite. The structure should provide integrity over of Licensing, U.S. Aamic Energy Commission, the period of time in which significant quantities Washington, D.C. 20545. The request should include (greater than Table I levels) of radioactivity remain the following information: [ with the material in the entombment. appropriate and continuing surveillance program An
- a. A description of the current stras of the facility. should be established under a possession-only license.
r' Note: Section electromcally reproduced from photocopy. C-2
l i c. Rennovel of Radianctive. D+=--- - ts and = b. De physical barners to unauthor~cxd entrance ( Diamantling. All fuel assemblies, radioactive fluids into the facility, e.g., fences, buildings, wuaed doors, . ) ] and waste, and other materials having activities and access openings, should be * ;9=' at least l [ above accepted unrestricted activity levels (Table 1) quarterly to assure that these bamers have not I { - should be removed from the site. The facility deteriorated and that locks and lockmg apparatus are !
- j. owner may then have unrestncted use of the site intact.
' with no requirement for a license. If the facility .
]
j owner so desires, the remander of the reactor c. A facility radiarian survey should be performed l facility may be dismantled and all vestiges removed at least quarterly to verify that no radioactive matenal i . and disposed of. is escaping or - being transponed through the l 4 cans =6n=an barners in the facility. Sampling should l
- d. Conversion to a New Noelear System or a be done along the most probable path by which l l Fossil Fuel Systeen. His ahernative, which applies radioactive matenal such as that stored in the inner l only to nuclear power plants, utilizes the existing n=tamment segions could be transponed to the outer l l
- - turbine system with a new steam supply system. regions of the facility and ultimaneiy to the environs. l l The original nuclear steam supply system should be j i separated from the electric generstag r,ystem and d. An envtrnamental radiarian survey shouki be *
$ disposed of in accordance with one of the previous performed at least seminnmintly to verify that no l 4 I three retirement alternatives. significant amounts of radiation have been released to the environment from the facility. Samples such as
- 3. SURVEILLANCE AND SECUPTY FOR THE soil, vegetation, and water should be taken at locations J l RETIREMENT ALTERNATIVES WHOSE for which statistical data has been estabhshed d'2 ring FIN AL STATUS REQUIRES A reactor operations, j POSSESSION-ONLY LICENSE
! e. A site representative should be designated to be ! A facility which' has been licensed 'under a responsible for controlling authorised access into and 4 possession-only license may contain a significant movement within the facility, i amount of radioactivity in the form of activated and contaminated hardware and structural matenals. f. Admimarrative procedures should be established ' Surveillance and cornmensurate security should be for the notifirmenan and reporting of aboonnel
- provided to assure that the public health and safety are occurrences such as (1) the entrance of an unanthartaed i not endangered. person or persons into the facility and (2)'a significant
- a. Physical security to prevent inadvenent exposure change in the radiation or cantammarian levels in the
{ ! of personnel should be provided by multiple locked facility or the offsite environment. l barriers. The presence of these barriers should make i it extremely difficult for an unauthorized person to gain 3. The following repons should be made: access to areas where radiation or contammarionlevels exceed those specified in Regulatory Position C.4. To (1) An annual report to the Director of j i prevent inadvenent exposure, radiation areas above Licensing, U.S. Atomic Energy Comminalon, l 5 mR/hr, such as near the activated primary system of Washington, D.C. 20545, describing the results of the [ a power plant, should be appropriately marked and environmental and facility radiation surveys, the status i should not be accessible except by cutting of welded of the facility, and an evaluation of the performance of [ closures or the disassembly and removal of substantial security and surveillance measures, j structures and/or shielding material. Means such as a I remote-readout intrusion alarm system should ~ be (2) An abnormal occurrence report to the jl provided to indicate to designated pw,& el when a Regulatory Operations Regional Office by telephone j physical barrier is penetrated. Security personnel that within 24 hours of discovery of an abnormal provide access control to the facility may be used occurrence. The abnormal occurrence will also be l instead of the physical barners and the intrusion alarm reponed in the annual report described in the preceding i systems, item. { meses: section electronically reproduced from photocopy. C-3
,, _. - . , .- . . - . - . _ . _ . - - , , , . , - . . . _ . , - .-,,c. -
~ , _ .
- h. Records or 1rgs relative to the following items d. Upon request, the Commission may authorize a l should be kept and retamed until the license is licensee to relinquish possession or control cf premises, !
ternunated, after which they must be stored with other equipment, or scrap having surfaces contaminatM in plant records: excess of the limits specified. This may include, but is not limited to, special circumstances such as the (1) Environmental surveys, transfer of premises to another licensed organintion
'(2) Facility radiation surveys, that wm continue to wd with radioactive materials.
Requests for such authorizauon should provide: )
- (3) Inspections of the physical barriers, and 1 (4) Abnormal occunences. (1) Detailed, specific information describing the I j premises, equipment, scrap, and radioactive
- 4. DECONTAMINATION FOR RELEASE FOR conmninanta and the nature, extent, and degree of 1 I
UNRESTRICTED USE residual surface contammation. ! If it is desired to termmate a license and to (2) A detailed health and safety analysis indicatir.g l eliminate any funher surveillance requirements, the that the residual amounts of matenals on surface areas, 3 facility should be sufficiently decontammarM to prevent together with other considerations such as the ! risk to the public health and safety. After the prospective use of the premises, equipment, or scrap, , decontamination is satisfactorily accomplished and the are unlikely to result in an unreasonable risk to the j site inspected by the Commission, the Commission may health and safety of the public. l
- authorize the license to be termmated and the facility j abandoned or released for unrestncted use. De c. Prior to release of the premises for unrestncted
- licensee should perform the decontammation using the use, the licensee should make a comprehensive. 1 j foll
- wing guidelines: radiation survey establishing that contamination is within the limits specified in Table 1. A survey report
- a. The licensee should make a reasonable effort to should be filed with the Director of Licensing, U.S.
eliminate residual contamination. Atomic Energy Comminaion, Washington, D.C. 20545, j with a copy to the Director of the Regulatory
- b. No covering should be applied to radioactive Operations regional Office having jurisdiction. De surfaces of equipment of structures by paint, plating, or report should be filed at least 30 days prior to the other covering material until it is known that planned date of abandonment. The survey report i contammation levels (determmed by a survey and should
l documented) are below the limits specified in Table 1. In addition, a reasonable effon should be made (and (1) Identify the premises; j documented) to funher mmimne contamination prior to any such covering. (2) Show that reasonable effort has been made to ; i reduce residual contamination to as low as practicable J l c. The radioactivity of the interior surfaces of levels; pipes, drain lines, or ductwork should be determined l by making measurements at all traps and odier (3) Describe the scope of the survey and the general appropriate access points, provided contamination at procedures followed; and ' these locations is likely to be representative of I contamination on the interior of the pipes, drain lines, (4) State the finding of the survey in units specified or ductwork. Surfaces of premises, equipment, or in Table 1. scrap which are likely to be contammated but are of such size, construction, or location as to make the After review of the report, the Commission may surface inaccessible for purposes of measurement inspect the facilities to confirm the survey prior to should be assumed to be contammated in excess of the granting approval for abandonment. permissible radiation limits. Note: Section electronically reproduced from photocopy. C-4
c i 5, REACTOR RETIREMENT PROCEDURES i As indicated in Regulatory Position C.2, several alternatives . are acceptable for reactor facility ; retirement. If minor disassembly or "mothballing' is l planned, this could be done by the existag operatag l and maintenance procedures under the license in effect. Any planned actions involving an unreviewed safety question or a change in the narhnical specifications : i should be reviewed and approved in accordance with , the requirements of 10 CFR 5 50.59. I If major structural changes to ' radmactive components of the facility are planned, such as removal cf the pressure vessel or major'campanants of the primary system, a dismantlement plan including the information required by I 50.82 should be submitted to , L the Commission. A dismantlement plan should be l submitted for all the alternatives of Regulatory Position ! C.2 except mothballing. However, minor diamanamMy activities may still be performed in the absence of such a plan, providd they are permitted by existing operating - and - maintenance procedures. 'A dismantlement plan should include the followag:
- a. A description of the ultimate status of the facility -
- b. A description of the dismantling activities and the precautions to be taken.
! c. A safety analysis of the dismantling activities including any effluents which may be released. l d. A safety analysis of the facility in its ultimate l status. l Upon satisfactory review and approval of the dismantling plan, a dismantling order is issued by the Commission in accordance with 5 50.82. When dismantling is completed and the Commission has been notified by letter, the appropriate Regulatory Operations Regional Office inspects the facility and verifies completion in accordance with the dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the: Commission may terminate the license, if possession-only license under which the dismantling activities have been conducted or, as an alternative, may make application to the State (if an Agreement State) for a byproduct materials license. blose: Section electronically reproduced from photocopy. C-5
TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS Nuclide* Average *' Maximum *d Removabic*d U-nat, U 235, U-238, and associated decay products 5,000 dpm a/100 cm2 15,000 dpm a/100 cm 2 1,000 dpm a/100 cm2 Transuranics, Ra-226, Ra-228, , D-230,'Th-228, Pa-231, Ac-227,1-125,1-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th-nat, R-232, St-90, Ra-223, Ra-224, U-232,1-126,1-131,1-133 1,000 dpm/100 cm2 3,000 dpm/100 cm2 200 dpm/100 cm2 Beta-gamme emitters (nuclides with decay modes other than alpha emission or l spontaneous fission) except Sr-90 and l others noted above. 5,000 dpm h/100 cm2 15,000 dpm h/100 cm2 1,000 dpm h/100 cm2 l
"Where surface contammation by both alpha- and beta-gamma-emitting nuclides exists, the limits established for alpha- and beta- gamma-emitting nuclides should apply iPatly.
l 'As used in this table, dpm (disintegrations per mmute) means the rate of emission by radioactive matenal as deteramed by
- correcting the counts per minute observed by an appropnate detector for background, efficiency, and geometnc factors l associated with the instrument ation.
' Measurements of average corrammant should not be averaged over more than I square meter. For objects ofless surface area, the average should 5.,e derived for each such object.
*Ibe maximum conrammarion level applies to an area of not more than 100 cm2 .
'ne amount of removable radioactive material per 100 cm 2of swface area should be determmed by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive matenal on the wipe l
with an appropriate instrument of known efficiency. When removable contammarion on objects of less surface area is
&;sruirsi, the peninent levels should be reduced proportionally and the entire surface should be wiped.
Note: Section electronically reproduced from photocopy. C-6}}