ML20081B594
| ML20081B594 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 02/17/1995 |
| From: | Audeman A, Payne A, Vitkus T OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC |
| Shared Package | |
| ML20081B582 | List: |
| References | |
| CON-FIN-A-9076 NUDOCS 9503160266 | |
| Download: ML20081B594 (95) | |
Text
{{#Wiki_filter:c i., . s. CONFIRMATORY SURVEY.
-OF THE REACTOR BUILDING 1 AND' PHASE 4 SYSTEMS SHOREHAM NUCLEAR POWER STATION'
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[ DOCKET No. 50-322] T.J. VITKUS Prepared for the [ Division of Waste Management Headquarters Office U.S. Nuclear Regulatory Commision d _f
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i b The Oak Ridge Institute for Science nnd Education (ORISE) was established by the U.S. Department of Energy to undertake national and international programs in science and engineering education, training and management systems, energy and environment systems, and medical sciences. ORISE and its programs are operated by Oak Ridge Associated Universities (ORAU) through a management and operating contract with the U.S. Department of , Energy. Established in 1946, ORAU is a consortium of 88 colleges and universities. NOTICES The opinions expressed herein do not necessarily reflect the opinions of the sponsoring institutions of Oak Ridge Associated Universities. 1
'Ihis report was prepared as an account of work sponsored by the United States Government. Neither the United States Government nor the U.S. Department of Energy, nor any of their employees, makes any warranty, expressed 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 l 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 l i
e t, , %. CONFIRMATORY SURVEY OF THE ' REACTOR BUILDING AND PHASE 4 SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK i Prepared by T. J. Vitkus Environmental Survey and Site Assessment Program Energy / Environment Systems Division Oak Ridge Institute for Science and Education Oak Ridge, Tennessee 37831-0117 l Prepared for the i U.S. Nuclear Regulatory Commission IIeadquarters Office Sponsored by the Division of Waste Management i FINAL REPORT FEBRUARY 1995 This report is based on work performed under an Interagency Agreement (NRC Fin. No. A-9076) between the U.S. Nuclear Regulatory Commission and the U.S. Depanment of Energy. Oak Ridge Institute for . Science and Education performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy. Shoreham Reactor Bailding and' Phase 4. February 16, 1995
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CONFIRMATORY SURVEY OF THE REACTOR BUILDING AND PHASE 4 SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Prepared by: ' Date:c2!/'/!77 T. J. Vitkus, $fject Izader / / Environmental Survey and Site Assessment Program Reviewed by: [A wk. Low k w Date: 1} 17/ W M. J. Laudeman, Radiochemistry Laboratory Supervisor Environmental Survey and Site Assessment Program l l Reviewed by: mn Y. k h A. T. Payne, Quality Assurancs Officer Date: d f//o' f 7 l l Environmental Survey and Site Assessment Program ; 1 ' i 1 i Reviewed by: Date: e2//7/95 W. L. Beck, Program Director Environmental Survey and Site Assessment Program I Shoreham Reactor Buildmg and' Phase 4 - February 16. 1995
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t,?. 3 i l ACKNOWLEDGEhENTS J l The author would like to acknowledge the significant contributions of the following staff members: , FIELD STAFF ; T. L. Bright l T. D. Herrera ! A. L. Mashburn ) E. H. Montalvo l J. L. Payne l l LABORATORY STAFF R. D. Condra J. S. Cox M. J. Laudeman CLERICAL STAFF D. A. Cox R. D. Ellis K. E. Waters ILLUSTRATOR T. D. Herrera i Shoreham Reactor Building ar[ Phase 4 - February 16, 1995
tl, . . t +- 1 L TABLE OF CONTENTS PAGE List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv List of Tables ..............................................vii Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . viii Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . 1 Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Obj ect i ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Document Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Procedures ................................................ 4 Findings and Results .......................................... 8 Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 S ummary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 References ................................................68 Appendices: Appendix A: Major Instmmentation Appendix B: Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86 Termination of Operating Licenses . for Nuclear Reactors t Shoreham Reactor Building and Phase 4. February 16,1995 iii . F
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LIST OF FIGURES 1 PAGE l 1 FIGURE 1: Location of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 18 l i FIGURE 2: Plot Plan of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 19 i t FIGURE 3: Reactor Building Floor Plans of 8'-0" and 40'-0" Elevations-- ! Structural Units Surveyed .............................20 j FIGURE 4: Reactor Building Floor Plans of 63'-0" and 78'-0" Elevations- ! Structural Units Surveyed .............................21 , FIGURE 5: Reactor Building Floor Plans of 112'-0" and 126'-0" Elevations- { Structural Units Surveyed .............................22 i FIGURE 6: Reactor Building Floor Plans of 150'-9" and 175'-9" Elevations-Structural Units Surveyed ........ ....................23 l i FIGURE 7: Reactor Building, Primary Containment 63' NW G/A (PC004)-- ! Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 24 l 1 FIGURE 8: Reactor Building, Primary Containment Sub-Pile Room (PC005)- l Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 25 l l FIGURE 9: Reactor Building, Primary Containment 78' NE G/A (PC007)- l Measurement and Sampling Imcations . . . . . . . . . . . . . . . . . . . . . . 26 j FIGURE 10: Reactor Building, Primary Containment 78' SE G/A (PC008)- l Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 27 l FIGURE 11: Reactor Building, Primary Containment 103' SE G/A (PC012)- f Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 28 ; FIGURE 12: Reactor Building, 8' SE G/A (RB004)-Measurement and i Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 . i FIGURE 13: Reactor Building, 8' SW G/A (RB006)-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 , l FIGURE 14: Reactor Building,40' SW G/A (RB014)-Measurement and Sampling locations .
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FIGURE 15: Reactor Building,78' SE G/A (RB037)-Measurement and ! Sampling Locations . . . . . . . . . . . . . . . . . . . .............32 l shoreham neaaor suuing and Phase 4. February 16,1995 iv
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LIST OF FIGURES (Continued) PAGE FIGURE 16: Reactor Building, West Accumulator Aisle (RB038)-Measurement and Sampling Locations ..............................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)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . 35 ! FIGURE 19: Reactor Building, Spent Fuel Storage Pool, Floor and North 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 l FIGURE 21: Reactor Building, Reactor Cavity Floor (RB071 x01)-Measurement and Sampling Locations ..............................38 ! FIGURE 22: Reactor Building, Reactor Cavity Walls (RB071 x01),-Measurement , and Sampling Locations. .............. .............. 39 ) FIGURE 23: Reactor Building, Dryer /S:parator Storage Pool (RB072x01),- ; Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 40 j
.i FIGURE 24: Reactor Building,175' South G/A (RB103)-Measurement and l Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 FIGURE 25: Reactor Building,175' NW G/A (RB106)-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 FIGURE 26: Reactor Building, Polar Crane (RB109)-Measurement and Sampling Iecations . . . . . . . . . . . . . . . . . . ..............43 FIGURE 27: Reactor Building, Polar Crane Rail Area (RB109)-Measurement ;
44' and Sampling tocations .............................. FIGURE 28: Radwaste Building,52' SE G/A Shield Blocks (RWO73 x02 and x03)~ . Measurement and Sampling IAcations . . . . . . . . . . . . . . . . . . . . . . 45 FIGURE 29: Radwaste Building, 52' SE G/A Shield Blocks (RWO73 x02 and x03)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 46 Shortham Reactor Building and' Phase 4 - February 16. 1995 V
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j LIST OF FIGURES (Continued) PAGE FIGURE 30: Radwaste Building, 52' SE G/A Shield Blocks (RWO73 x02 and x03)- ! Measurement Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 l FIGURE 31: Reactor Building, Reactor Asserdly Lower Bowl (SU001)- i 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 (SUO14)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 51 J FIGURE 35: Outdoor Building Storm Drain Manways and Trenches Components (SUO23)-Measurement and Sampling Locations ................ 52
' FIGURE 36: Reactor Building, Reactor Primary Containment Components (SUO58)-
Measurement and Sampling Locations . . . . . . . . .............. 53 ; FIGURE 5.': Reactor Building, Reactor Primary Containment Components (SUO58)- , Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 54 i FIGURE 38: Reactor Building, Ventilation Components (SUO60)-Measurement I and Sampling Locations ..............................55 l FIGURE 39: Reactor Building, Ventilation Components (SUO61)-Measurement and Sampling Locations .. ...........................56 l l 1 Shoreham Reactor Building and' Phase 4 - February 16, 1995 Vi
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' LIST OF TABLES L PAGE i- TABLE 1: Reactor Building Confirmatory Survey Units.. . . . . . . . . . . . . . . . . . 57 i TABLE 2: Summary of Surface Activity levels . . . . . . . . . . . . . . . . . . . . . . . 59 l
TABLE 3: Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61
. TABLE 4: Confirnntory Radiological Status Summary-Structures . . . . . . . . . . . 62 l-TABLE 5: Confirmatory Radiological Status Summary-Systems . . . . . . . . . . . . 66 1
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- ABBREhTIONS AND ACRONYMS .
- ASME : American Society of Mechanical Engineers ,
cm2 square centimeter ' 3:, Co-60 cobalt cpm counts per minute DOE Department of Energy. , 2 dpm/100 cm disintegrations per minute per 100 square centimeters ! E M L- Environmental Measurements Laboratory EPA ~ Environmental Protection Agency ESSAP Environmental Survey and Site Assessment Program ! Fe - iron-55 ft2 square feet H-3 tritium ha hectare GAG gross activity guideline GM Geiger-Mueller , km kilometer ! L, critical level LILCO Long Island Lighting Company ; LIPA - Long Island Power Authority m meter ; m2 square meter MDA minimum detectable activity , NaI sodium iodide l Ni-63 nickel-63 ! NIST National Institute of Standards and Technology i NRC . Nuclear Regulatory Commission l ORISE Oak Ridge Institute for Science and Education PC#- primary containment structural survey unit designation pCill picocuries per liter pCi/g picocuries per gram QA Quality Assurance QC Quality Control RB# Reactor Building structural survey unit designation RW# Radwaste Building structural survey unit designation SNPS Shoreham Nuclear Power Station SU# system survey unit designation , UCL upper confidence level- { pR/h microroentgens per hour ZnS zinc sulfide 1 Shorcham Reactor Building anfPhase 4 - February 16. 1995 viii q --, w , e ,- , , - - . - ., ,--
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CONFIRA1ATORY SURVEY OFTHE REACTOR BUILDING AND PHASE 4 SYSTEhiS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, MW YORK INTRODUCTION AND SITE HISTORY The long Island Lighting Company (LILCO) constructed a boiling water reactor known as the Shoreham Nuclear Power Station (SNPS). The plant was designed to provide a gross electrical output of 849 Megawatts and uchieved initial criticality in Febmary 1985. The U.S. Nuclear Regulatory Commission (NRC) License No. NPF-82 (NRC Docket File No. 50-322) issued for the facility allowed reactor ope,ations at power levels not to exceed 5% of full power. Low power testing in accordance with the license then commenced in July,1985 and continued ; intermitt:ntly 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 i exposure. The irradiated fuel, which was a standard low enrichment (2 to 3% uranium-235) ; uranium fuel, was removed from the reactor vessel and placed into the spent fuel pool in August 1989. Subsequently, the fuel was shipped off site during 1993 and 1994. i Various reactor components, piping systems, and other equipment became radiologically 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 (LILCO 1990). The Long Island Power Authority (LIPA) was established to decomraission the facility and release the site for unrestricted use. LIPA developed a decommissioning plan, approved for implementation by the NRC in June 1992, which included decontamination or removal of contaminated ponions of the reactor and other plant systems and equipment. A major consideration of the decommissioning plan was to maintain the integrity, when possible, of plant stmetures end systems. Shoreham Restor Building am! Phase 4. Fetutary 16. 1995
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The decommissioning and termination surveys were conducted in four phases. Phase 1 included the termination survey of the internal components of the main turbine, the Turbine Building and associated systems, site grounds, and exterior site structures. Phases 2 and 3, included the Reactor Building Suppression Pool, Phase 2 systems and the Radwaste Building. Phase 4 addressed the remaining portions of the Reactor Building and associated systems. The NRC IIeadquarters' Division of Waste Management requested that the Environmental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) conduct confirmatory radiological surveys of the SNPS decommissioning project. Since February 1993, ESSAP has completed the confirmatory survey of the turbine internal components, the Turbine Building, site grounds, exterior site structmes, Radwaste Building, Suppression Pool, and Phase 2 systems. The results of these surveys are the subject of separate final, or draft reports (Vitkus 1993a; Vitkus 1994a). The final confirmatory survey for Phase 4, the Reactor Building, has now been completed and is the subject of this report. SITE DESCRIPTION The SNPS is located in the Town of Brookhaven, New York on the north shore of Long Island, approximately 80 km (50 miles) east of La Guardia Airport and the confluence of the East River and Long Island Sound (Figure 1). The SNPS is located on a 32.4 hectare (80 acre) portion of a larger 202 hectare (ha) parcel of land owned by the LILCO. The site is bounded on the north by Long Island Sound, on the east by the Wading River Marshland, on the west by other LILCO 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 as the power block, where radiological controls were necessary (Figure 2). Each of the buildings addressed during the confirmatory process are located here and are shown on Figure 2 as the Turbine Building, the Radwaste Building, and the Reactor Building. Construction of the Reactor Building is predominately structural steel and concrete. Total floor space of the building is approximately 7,800 m2 (84,000 ft2 ) divided among 8 levels at elevations 8'-0",40*-0",63'-0", 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 Shortham Rcactor Building and he 4 - February 16, 1995 2
t, . . *p . systems. Major structural components included the primary 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 l (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 potentiait' contaminated or have known contamination, or a system which circulated, stored or processed radioactive materials such that they could become contaminated, or experience, I 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 l residual radioactivity" (LIPA 1994a). Area classification was determined by radiological use history, environmental monitoring activities, and the results of the previous characterization 1 survey. Affected and unaffected areas are further subdivided into survey units. Survey units l are categorized as stmetures (floors, walls, ceilings, and exterior surfaces of piping and equipment), plant systems (equipment and piping internals), and outdoor areas (grounds and l building exteriors). In addition, affected survey units also had sub-classifications as suspect or l non-suspect, and may also have been classified as alpha affected if involved with fuel handling or storage. Phase 4 of the decommissioning addressed a total of 117 survey units, of which 100 were structures, and 17 were systems. l OBJECTIVES The objectives of the confirmatory survey were to provide independent document reviews and I radiological data for use by the NRC in evaluating the adequacy and accuracy of the licensee's procedures and termination survey results. Shoreham Reactor Buildmg and Phase 4 - February 16. 1995 3
t, . . d. DOCUMENT REVIEW ESSAP reviewed LIPA's release records for those survey units selected for confirmatory survey (LIPA 1994b). Documents were reviewed for adequacy, accuracy, completeness, and consistency. Data were reviewed for appropriateness of calculations and interpretations relative to the guidelines. PROCEDURES A survey team from ESSAP visited the SNPS during the period October 31 through November 4,1994 and performed independent visual inspections, measurements, and sampling of survey 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- ) numeric with the first characters designating the type of unit, structure (building or area specific, l RB = Reactor Building, PC = Primary 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 l l 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 as well as applicable sections of the ESSAP Survey Procedures and Quality Assurance Manuals ! (Vitkus 1994b). The following procedures apply to survey units selected for independent confirmatory surveys. Additional information regarding selection of confirmatory survey units and the implementation of this plan may be found in the general site confirmatory survey plan (Vitkus 1993). Shorcham Reactor Buildmg an[ Phase 4 - February 16.1993 4
1,A %. SURVEY PROCEDURES I Reference System LIPA 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 j 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 ! t 1 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 components. ! 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. l I Scans were performed using gas proportional, Geiger-Mueller (GM), zine sulfide (ZnS), and/or sodium iodide (Nal) detectors coupled to ratemeters or ratemeter-scalers with audible indicators. i S_urface Activity Measurements i 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 l average NRC surface contamination guideline was exceeded, the size of the contaminated area ! and the average activity in the contiguous I m 2area was also determined. Measurements were : performed using gas proportional and/or GM detectors coupled to ratemeter-scalers. A smear ; l Shnrrham Reacu Dung amlThue 4 - February 16. 1995 5 ! I
(. . : s, . . 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. Exposure Rate Measurements Exposure rate measurements were performed within each survey unit at each accessible floor direct measurement location, excluding units RB071 x01 and RB109 and system interiors which were inaccessible for this type of measurement (Figures 7 through 19,23 through 25, and 30). 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). Systems L1PA 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 direct measurements performed and smears collected was dependent upon component size and accessibility and ranged from 5 to 45 measurements per system. Scans and direct measurements were performed using gas proportional, GM, and/or Nal detectors coupled to ratemeters or ratemeter-scalers. Figures 31 through 39 show measmement and sampling locations. Miscellaneous Material Samples Per the request of the NRC site representative, ESSAP collected a water sample from the porous concrete sump and a sediment sample from the truck bay trench (Figures 34 and 35). Shoreham Reactor Dullding and Ptuise 4 - February 16.1W5 6
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Samole Analysis and Data Internretation ; Samples and data were returned to ESSAP's Oak Ridge, Tennessee facility for analysis and j interpretation. Smear samples were analyzed for gross alpha and gross beta activity using a' low l background proportional counter. Direct measurement and removable activity results were f 2 converted to units of disintegrations per minute per 100 square centimeters (dpm/100 cm), i Surface activity levels which exceeded the background distribution, referred to as the critical t level (L), were then corrected for those radionuclides that may not be quantified using field instrumentation, specifically Fe-55, H-3, and/or to a lesser extent Ni-63. The respective l correction factors used were developed by LIPA and were based on the fraction of total activity ; contributed by each radionuclide in each particular survey area (LIPA 1994a). Water and j sediment samples were analyzed by solid state gamma spectrometry. Spectra were reviewed for : Co-60 and any other identifiable photopeaks. Results were reported in units of picocuries per f liter (pCi/l) for water and picoeuries per gram (pCi/g) for the sediment. Exposure rates were ! reported in units of microroentgens per hour ( R/h). The 95% confidence level ( ), in l f accordance with NUREG/CR-5849, was calculated for surface activity and exposure rates for t 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 i determine, with a given confidence level, whether the LIPA survey data was not biased low f 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., Hg LIPA h pESSAP. This hypothesis was tested at the 95% confidence level (0.05 level of ; significance). If the hypothesis was rejected at that confidence level, the alternative hypothesis was accepted i.e., H3 : LIPA < pESSAP. The test statistic, t, was calculated using the following equation: Shoreham Reactor Ituildmg and l'hase 4. l'ebruary 16,1995 7
/
., = 1 i T, - K l; t= (ng-1) S, + (nt-1) S$ f ng+n' t l
% ng+n-2 z nn gz ,
r l where: j M is the LIPA surface activity mean for a survey unit j K is the ESSAP surface activity mean for the same survey unit ] no is the number of LIPA direct measurement data points I na 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) l for the appropriate degrees of freedom at the 95% confidence level (0.05 level of significance). f If the Ho: pLIPA 2: ESSAP was rejected, then ESSAP evaluated additional options and f alternatives and conferred with the NRC as to the recommended approach. I FINDINGS AND RESULTS l l DOCUMENT REVIEW ESSAP's review of the termination survey final report and release records for those survey units selected for confirmatory survey indicated that the termination plan had been appropriately I followed with no significant deviations. Data were appropriately converted, tested, and 1 l presented. l i i SURFACE ACTIVITY GUIDELINES j The surface activity results that follow were compared with the following guidelines: , I l Shorchain Reacaw Building and Phase 4 - February M.1995 8 l l
-~ _ - -.. . _ . . - - -. ..- - . . - . - - ..~ . - . , . . . . . -
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- The' beta guidelines are:
i I i Total Activity i 5,000 8 9 pm/100 d cm ,2averaged over 1 m2 ] , 2 2 15,000 Sq~dpm/100 cm , maximum in a 100 cm area i i
.I Removable Activity j 1,000 Bq dpm/100 cm2 ;
t 4 The alpha guidelines are: I Total Activity ! 5,000 a dpm/100 cm2, averaged over 1 m2 -l 15,000 a dpm/100 cm2 , maximum in a 100 cm2 area l Removable Activity j 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:
- l
-i Total Activity .
q 200,000 dpm/100 cm2, averaged over 1 m2 j 600,000 dpm/100 cm2 , maximum in a 100 cm2 area l j Removable Activity 1,000 dpm/100 cm2 ; i 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 Shorcham Reactor Buildatg and Phase 4. February 16.1995 9
w .; ,. 5s p.- I p j @; Draft NUREG/CR-5849, rather than the above generic NRC guidelines (LIPA 1994a; Berger j
- 1992). These confirmatory survey units, together with the GAG values are listed below: l SU001. Reactor esembiv lower Bowl and RB072x01. Drver/Senarator Pool !
11,100 dpm/100 cm2, averaged over 1 m2 33,000 dpm/100 cm2 , maximum RWO73x02. Bioshield Concrete j 2 2 95,900 dpm/100 cm , averaged over 1 m 287,700 dpm/100 cm2 , maximum RWO73 x03. Bioshield Steel' [ 2 2 76,900 dpm/100 cm , averaged over 1 m l i 230,700 dpm/100 cm2 , maximum i l RB068. Soent Fuel Pool 9,400 dpm/100 cm2, averaged over 1 m2 28,300 dpm/100 cm2 , maximum i l
- SURFACE SCANS l i
Surface scans identified a number of locations with above background residual beta surface. l activity in three survey units, RB068 (Spent Fuel Storage Pool), RB072x01 (Dryer / Separator Pool), and SU001 (Reactor lower Bowl). Additional investigation of each location determined ! that two of the locatie's, each measuring less than 100 cm2 in total area, were either at the . average GAG or - wtd the maximum GAG values. Both of these were located in survey unit RB068. Tine surface activity levels were 9,400 dpm/100 cm 2 (average GAG is 2 9,400 dpm/100 cm2 ) and 32,000 dpm/100 cm 2 (maximum GAG is 28,300 dpm/100 cm ). LIPA i i personnel remediated each location and ESSAP then perfonned post-remedial action surface j scans and direct measurements. These surface scans and measurements indicated that areas were I adequately remediated. The surface activity levels at the other locations were less than the Sharcham Reactor Building and Phase 4 - February 16.1995 10
*p.
applicable guidelines but LIPA 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 units. SURFACE ACTIVITY LEVELS 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, sample count time, and contributions from Fe-55, H-3, and Ni-63 as appropriate (when the net count rate exceeded the background distribution [1,]). Actual values are reported, including negative surface activity levels, which occurred when the field count rate was less than the background. Of 796 measurements,180 exceeded the I,. The results of total and removable surface activity levels for each confirmatory survey unit are summarized in Table 2. Total beta activity levels after additional remediation for the structural survey units, excluding RWO73X02 and -X03 which are discussed below, ranged from -760 to 5,600 dpm/100 cm .2 The highest final 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 2 remediation, residual activity levels ranged up to 8,600 dpm/100 cm (average GAG of 11,100 dpm/100 cm2 ) and 32,000 dpm/100 cm2 (maximum GAG of 28,300 cpm /100 cm2 ) in 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 cm2 for gross beta. The mean residual activity in structural survey units, as presented in Table 4, ranged from -340 to 450 dpm/100 cm2and from -0.7 to 3.5 dpm/100 cm 2for total and removable gross beta activity, respectively. The mean activity levels, prior to additional remediation, were 670 dpm/100 cm 2 in RB072X01 and 1,500 dpm/100 cm2 n RB068. Total beta activity in surveyed systems ranged from -680 to 2,500 dpm/100 cm2 . The removable activity levels ranged from -1 to 5 dpm/100 cm2for gross alpha and from -5 to 35 dpm/100 cm2 1 for gross beta. The mean beta activity levels for systems ranged from -70 to 390 dpm/100 cm2 l 2 for total activity and from -0.3 to 2.8 dpm/100 cm for removable gross beta activity (Table 5). Shoreham Reactor Buildmg and Nic 4 - February 16, 1995 11
1,, . *p* Total beta activity levels for the Bioshield Wall Blocks (RWO73X02 and -X03) ranged from -410 to 15,000 dpm/100 cm 2. The average GAGS for these survey units were 95,900 dpm/100 cm2 for RWO73X02 and 76,900 dpm/100 cm2 for RWO73X03. Removable activity ranged from -1 to 1 dpm/100 cm2 for alpha and from -5 to 17 dpm/100 cm2for beta. The mean beta activity 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 The interior background exposure rates, as determined during the Turbine Building survey, i ranged from 4 to 5 pR/h and averaged 5 pR/h at 1 m (Vitkus 1994a). Individual gross exposure f rate's within the Reactor Building and adjacent to the Bioshield Blocks ranged from 2 to 6 pR/h at 1 m (Table 3). The net mean exposure rates for the structural survey units, ranged from
-3 to 0.9 pR/h (Table 4).
RADIONUCLIDE CONCENTRATIONS IN MISCELLANEOUS SAMPLES Cobalt-60 concentrations in the Porous Concrete Sump water sample and the Reactor Building Truck Bay trench residue samples were <4.2 pCi/l and <0.2 pCi/g, respectively. No other significant photopeaks were identified. COMPARISON OF RESULTS WITH GUIDELINES l The confirmatory survey results were compared with both the data provided by LIPA and the NRC guidelines for release to unrestricted use. The NRC's Regulatory Guide 1.86 provides the guidelines for acceptable surface contaminatien levels used to determine whether a licensed facility may be released to unrestricted use. These guidelines are summarized in Appendix C. 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: Shoreham Reactor Building and Phase 4 - February 16,1995 12
' g,, , . *: . . i Total Activity [ 5,000 B'y dpm/100 cm 2, averaged over 1 m2 15,000 B-y dpm/100 cm2 , maximum in a 100 cm2 area ; h Removable Activity ! 2 1,000 B-y dpm/100 cm The alpha guidelines are: , Total Activity 5,000 a dpm/100 cm 2, averaged over 1 m2 15,000 a dpm/100 cm2 , maximum in a 100 cm2 area i Removable Activity ! 1,000 a dpm/100 cm2 q
-s 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 i 200,000 dpm/100 cm2, averaged over 1 m2 600,000 dpm/100 cm2 , maximum in a 100 cm2 area f Removable Activity f 2 1,000 dpm/100 cm , For those survey units involving mixtures of radionuclides LIPA substituted the gross activity f 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 (IJPA 1994a; Berger i 1992). These confirmatory survey units, together with the GAG values that the data were i compared with and tested against, are listed below: i Shortham Reactor Buildmg and thase 4 - February 16.1995 13 ;
.i,r
SU001. R.eactor Assembly Lower Bowl and RB072x01. Drver/Senarator Pool 11,100 dpm/100 cm2 , averaged over 1 m2 33,000 dpm/100 cm2 , maximum Rwo73 x02. Bioshield concrete 95,900 dpm/100 cm2, averaged over 1 m 2 287,700 dpm/100 cm2 , maximum RWO73 x03. Bioshield Steel - 76,900 dpm/100 cm2, averaged over 1 m 2 230,700 dpm/100 cm2 , maximum RB068. Spent Fuel Pool 9,400 dpm/100 cm2 , averaged over 1 m2 28,300 dpm/100 cm2 , maximum 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 R/hr above background, measured at 1 m above the surface (U.S. NRC 1984), As previously discussed, the detection sensitivities of the field instmments are such that the . residual Fe-55 activity cannot be detected. In addition, LIPA determined that residual II-3 and/or Ni-63 activity also must be accounted for within several survey units (LIPA 1994a). Therefore for those surveys thst LIPA identified as such, total and removable surface activity 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 ' ESSAP and LIPA data sets were then compared and evaluated relative to the established conditions. Shoreham Reactor Buihhng and' Phase 4 - February 16.1995 14
y, *v - A comparison of the ESSAP mean surface activity levels to the LIPA mean activity levels showed that the ESSAP mean was statistically less than or equal to the respective mean determined by LIPA for 24 of the 30 confirmatory survey units, thereby satisfying the data evaluation and comparison conditions. However, the ESSAP mean surface activity value was statistically greater than the LIPA mean for survey units PC005, PC007, PC012, RB072X01, RWO73X02 and SUO60; therefore, additional evaluation was necessary. For the three Primary Containment survey units (PC005, PC007, and PC012) where the ESSAP surface activity mean was higher than the LIPA mean, LIPA qualified the release record data as containing excessive negative measurements. According to LIPA, this was due to observed, 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 were 310 dpm/100 cm2 and 350 dpm/100 cm2 , 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 above survey units, the observed difference in means would not affect the conclusions LIPA reached for remaining survey units. In survey unit RB072X01, the ESSAP surface activity mean of 450 dpm/100 cm 2exceeds the LIPA mean of 140 dpm/100 cm2 . It should be noted that ESSAP performed 5 of the 30 direct measurements at locations of elevated direct radiation identified by surface scans. The surface 2 activity at these locations ranged from 1,800 to 4,000 dpm/100 cm . Inclusion of these l measurements in the determination of the average surface activity level results in a mean that is biased high and may not be representative of the average survey unit residual activity level l which LIPA reported. The ESSAP mean for survey unit RWO73X02 was 6,400 dpm/100 cm2 versus the LIPA 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 DuitJmg adPhase 4. February 16, 1995 15
noted while performing surface scans. LIPA selected measurement locations systematically. Theefore, similar to RB072X01, the ESSAP mean for this survey unit is high relative to the LIPA de: ermined average, and the difference in means in not considered significant when compared to the applicable guidelines (average GAG of 95,900 dpm/100 cm2 ) for this survey unit. For survey unit SUO60, the ESSAP total surface activity mean was 360 dpm/100 cnf with a maximum level of 2,500 dpm/100 cm2. The corresponding LIPA values were 174 dpm/100 cm 2 and 3,247 dpm/100 cm2 . LIPA had not qualified the data for this unit as containing excessive negative values. Therefore, the LIPA surface activity levels, mean activity level, and UCL were evaluated to determine the potential impact on the LIPA reported status of this and the other Phase 4 survey units, relative to the guidelines. The difference between the ESSAP and LIPA means was 213 dpm/100 cm2 . If this difference in activity levels were applied to each Phase 4 survey unit, overall surface activity levels would not be significantly altered and the conclusions reached, that the radiological status of this and all other Phase 4 survey unit satisfy the guidelines, would remain valid. 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 at the location with the highest surface activity level. If all of the net activity present at this 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-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 in 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 Shorcham Reactor Buildmg andhase 4 - February 16,1995 16
e 1,. . ;. . additional remediation the average GAGS (9,400 dpm/100 cm2 for RB068 and 2 I 11,100 dpm/100 cm for RB072X01) were satisfied at the 95% confidence level and no additional measurements were required to demonstrate compliance. The preremediation UCLs were 3,100 dpm/100 cm2 for RB068 and 1,200 dpm/100 cm 2 for RB072x01. There were no 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 level. 4 Exposure rates were compared with those obtained by LIPA, and tested at the 95% confidence level, relative to the 5 R/h above background guideline currently being used by the NRC. The Reactor Building and Bioshield Block exposure rates were comparable to background exposure rate levels and confirmed the findings presented by LIPA. 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.
SUMMARY
The Environmental Survey and Site Assessment Program performed confirmatory survey 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 through November 3,1994, ESSAP performed independent surface scans, surface activity measurements, exposure rate measurements, and miscellaneous sampling. ESSAP's confirmatory survey results support the data presented in the LIPA termination survey report. These results indicate that total and removable surface activity levels and exposure rates were below the NRC guidelines for release to unrestricted use. Statistical tests of data sets further support the conclusion that each survey unit satisfies the guidelines at the 95 % confidence level. Shoreham Reactor Buildug and Phase 4 - February 16.1995 17
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Measurement and Sampling Locations Shortham Reactor Building and Phase 4 - February 16,1995 45
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i FIGURE 31: Reactor Building, Reactor Assembly Lower Bowl (SU001) - Measurement and Sampling Locations ; t Shoreham Reactor Inuilding and Phase 4. February 16,1995 48
6, c 275a,031 (2) o o o 9 o o G ' o o 9 9 o o o O MEASUREMENT / SAMPLING LOCATIONS NOT TO SCALE
$ SINGLE-POINT FIGURE 32: Reactor Biulding, Nuclear Boiler Main Steam Relief Valve (SUOO2) -
Measurement and Sampling Locations Shoreham Reactor Building and Phase 4 Ichruary 16.1995 49
- g. 2739029 (2) o o n v
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$ SINGLE-POINT FIGURE 33: Reactor Building, CRD Hydraulic Control Components (SUOO4) - l Measurement and Sampling Locations j Shortham Rearkw Building and Phase 4 - February 16,1995 50
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Measurement and Sampling Locations i Shorcian Reactor Building and l'hase 4 - February 16,1995 51 l
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E SEDIMENT l NOT TO SCALE l FIGURE 35: Outdoor Building Storm Drains, Manways, and Trenches, (SUO23) - Measurement and Sornpling Locations Shortham Reactor Building and' Phase 4 Febniary 16,1995 52 l
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9 SUO58-14-1 3 SUO58-15-1 Y U SUO58-11-1 SU058-16-1 SuoS8-17-1 e w 2 SUO58-12-1 SUO58-11-1 SUO58-6-1 SUPPRESSION POOL SUPPRESSION POOL SUPPRESSION POOL MEASUREMENT / SAMPLING LOCATIONS NOT TO SCALE g SINGLE-POINT FIGURE 36: Reactor Building, Reactor Primary Containment Components (SUO58) - Measurenment and Sampling Locations shonham Reactor Buildmg and Ibse 4. Febmary 16, 1995 53
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lOI ! s ! leo l SUO58-8-1 SUO58-9-1 5U058-10-1 l I MEASUREMENT /SAMPUNG LOCATIONS NOT TO SCALE ! 9 SINGLE-PolNT l I l FIGURE 37: Reactor Building, Reactor Primary Containment Components (SUO58) - Measurement and Sampling Locations Shoreham Reactor Building an[ Phase 4. February 16,1995 54 i
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$ SINGLE-POINT FIGURE 38: Reactor Building, Ventilation Components (SUO60) -
Measurement and Sampling Locations
$behmn Rmmr BuiWmg E Phm 4. Febnary 16, 1995 55
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9 SUO61XO7 o e e o e N / \ 4 M / 11 12 f3 ELEVATION 125'-0* MEASUREMENT / SAMPLING LOCATIONS NOT TO SCALE e SINGLE-POINT FIGURE 39: Reactor Building, Ventilation Components (SUO61) - Measurement and Sampling Locations Shorrbam Reactor Duilding a ui Phase 4 February 16,1993 56
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TABLE I i REACTOR BUILDING SURVEY UNITS - SELECTED FOR CONFIRMATORY SURVEY SIIOREHAM NUCLEAR POWER STATION i BROOKIIAVEN, NEW YORK Survey Affected (A)/ Structure / ! Survey Unit Name Unit / Component Unaffected /U) System , PC004 Primary Containment - 63' NW A stmeture ! PCb05 Sub-Pile Room A structure i PC007 Primary Containment - 78' NE A structure ; PC008 Primary Containment - 78' SE A structure . PC012 Primary Containment - 109' NE A structure RB004 Reactor Bui'djng - 8' SE A structure RB006 Reactor Eaihting - 8' SW A structure RB014 Reactor Building - 40' SW A structure IG037 Reactor Building - 78' SW A structure RB038 Westpumulator Aisle A structure RB057 FueliW1 L' lean-up Pumps Room A structure ! RB061 RWCU Regen /Non-Regen IITX's Room A structure RB068 Spent Fuel Storage Pool A structure RB071 x01 Reactor Cavity - 150' A structure i RB072 x01 , Dryer / Separator Storage Pool A structure RBiO3_ Reactor Building - 175' A stmeture ! RB106 Reactor Building - 175' A structure , RB109 Polar Crane A structure : i R,WO73 x02 Bioshield Wall Blocks (Concrete) A structure t RWO73 x03 Bioshield Wall Blocks (Steel) A stmeture SU001 Reactor Assembly Lower Bowl A system f SU002 Nuclear Boiler Main Steam Relief Valve A system i IB21-RV-095G Shoreham Reac;ar !!uildmg and i'hase 4 - Fd,rimry 16. 1995 57
- s. . /,.
TABLE 1 (Continued) REACTOR BUILDING SURVEY UNITS SELECTED FOR CONFIRMATORY SURVEY ; SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK t Survey Affected (A)/ Structure / Survey Unit Name Unit /Compenent Unaffected (U) System SU004 CRD Hydraulic Control A system SUO14 x05 #953 Misc. Embedded Drain Piping A system 1G11-TK-190 SUO14 x12 RB Porous Concrete Sump #P-224A A system SUO23 Misc Building Storm Drains A system SUO58 Reactor Primary 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 Shoreham Reactor Buildmg and thase 4. February 16, IW5 58
TAELE 2 .';
SUMMARY
OF SURFACE ACTIVITY LEVELS , ,.. REACTOR BUILDING 2 ir SIIOREHAM NUCLEAR POWER STATION f BROOKHAVEN, NEW YORK I a E Number of Total Activity Range Removable Activity Location" Measurement (dpm/100 cm2) Range (dpm/100 cm2) I 1 Locations Betab Alphad Beta' E } Reactor Building f PC004 Primary Containment 30 -230 to 410 -1 to 1 -5 to 14 f PC005 Primary Containment, Sub-Pile Room Primary Containment 30 30 57 to 580
-340 to 910
-1 to 3
-1 to 5
-4 to 18
-4 to 10 j PC007 PC008 Primary 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 IITX's Room 30 -570 to 330 -1 go 1 -5 to 5 RB068 Spent Fuel Storage Pool 34 (-300 to 32,000)f 8 -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 8600)" -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
TABLE 2 (Continued) '.
SUMMARY
OF SURFACE ACTIVITY LEVELS .. . REACTOR BUILDING .:. ( SIIOREIIAM NUCLEAR IEVER STATION li BROOKIIAVEN, NEW YORK R E $ Number of Total Activity Range Removable Activity Range (dpm/100 cm 2) { Location" Measurement (dpm/100 cm2) i Locations Betab
- Alpbad &ta'
[
? SU001 Reactor Assembly lower Bowl 30 -160 to 2200 -1 to 3 -5 to 10 f SU002 Nuclear Boiler System 5 -38 to 340 -1 to 1 -3 to 1 f;: SUOO4 Hydraulic 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 SUO14 x05 Embedded Drain Piping 10 -380 to 340 -1 to 1 -4 to 5 SUO23 Misc. Building Storm Drains 25 -680 to 2400 -1 to 1 -5 to 13
$ 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 RWO73 x03 Bioshield Wall Blocks (Steel) 25 -410 to 6600i -1 to 1 -5 to 17
' Refer to Figures 7 through 39.
b MDAs = 200 to 1200 dpm/100 cm2, j
- 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, I
Represents surface activity levels prior to additional remediation. 8 GAG values are 9,400 dpm/100 cm2 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.
' GAG values are 95,900 dpm/100 cm2 average and 287,700 dpm/100 cm2 maximum.
iGAG values are 76,900 dpm/100 cm2 average and 230,700 dpm/100 cm2 maximum.
.,,, ' .* 4 TABLE 3 EXPOSURE RATES REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION . BROOKHAVEN, NEW YORK Number of Measurement Net Exposure Rate Range Location. Locat,ons i at 1 m (gR/h) 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 NA ' NA ; RP072x01 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 Site DaCkground exposure rate is 5 R/h.
'NA=Not Applicable, measurements not performed.
Shorcham Reacun Building amfPhase 4 February 16.1995 61
s, . . *
*\ ;
TABLE 4 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES REACTOR BUILDING . SIIOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit PC004 PC005 PC007 PC008 PC012 Total Beta Activity (dpm/100 cul)"
# of Direct Measurements 30 30 30 30 30 Mean(X) 36 310 100 -120 23 ;
IlPA X -38 -26 -78 70 -144.4 Fa 86 350 189 -73 63 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes : Removable Beta Activity (dpm/100 cnf)"'b
# of Smears 30 30 30 30 30 ,
Mean(X) 0.3 0.1 0.7 -0.4 -0.2 , L1PA X 7.0 9.9 2.7 4.8 7.4 Fa 1.4 1.5 1.2 0.5 0.6 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes Exposure 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 ;
IJPA X -1.0 -0.3 -0.5 -0.9 -0.2 ; Fa -2.5 -1.2 -2.9 -2.8 -1.8 , 5 pR/h Above Background Guideline Yes Yes Yes Yes Yes , Satisfied i i i 1 Shoicham Reactor Buildwg andhase 4 - February 16.1995 62
. 1
- s, . - -- TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES REACTOR BUILDING < SIIOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit RB004 RB006 RB014 RB037 RB038 Total Beta Activity (dpm/100 cm 2)" ;
# of Direct Measurements 30 30 30 30 30 ;
Mean (I) -290 -340 150 -210 -190 LIPA X -48 -220 56 47 86 Fa -250 -300 190 -180 -110 I Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes Removable Beta Activity (dpm/100 cal)a,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 cnf Guideline Satisfied Yes Yes Yes Yes Yes ; Exposure Rates at I m (gPJ10
# of Exposure Rate Measurements 10 4 10 9 6 Net Mean (X) -1.4 -1.7 -0.7 -0.9 -1.1 f LIPA I -0.3 0.1 -0.2 -0.1 -0.6 Fa -1.3 -1.1 -0.7 -0.7 -0.9 ;
5 pR/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied h Shoreham Reactor BuildinC and' Phase 4 February 16, 1995 63
,, a _
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES REACTOR BUILDING SIIOREIIAM NUCLEAR POWER STATION l BROOKHAVEN, NEW YORK Radiological Survey Unit , RB057 RB061 RB068 RB071x01 RB072x01 . Total Beta Activity (dpm/100 cm 2)"
# of Direct Measurements 30 30 34 30 30 .
Mean(X) -240 -170 (1500)* 43 (670)* 85 450 , LIPA X 75 55 290 36 140 Fa -210 -110 60 (3100)' (1200)* 250 760 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes t Removable Beta Activity (dpm/100 enf)a,b
# of Smears 30 30 34 30 30 Mean(X) 0.2 -0.6 2.7 1.5 3.5 LIPA X 3.6 9.2 17 4.6 27 i Fa 1.5 0.2 4.3 3.4 5.3 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes Exposure Rates at 1 m (gR/h)
# of Exposure Rate Measurements 9 10 9 NA 0.4 Nct Mean (I) -0.8 -0.7 0.0 NA 0.4 i LIPA I 0.0 -0.1 0.0 -0.3 0.1 ,
F= -0.6 -0.6 0.2 NA 0.6 5 pR/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied i Stuweham Restor Building and the 4 - February 16. 1995 64
s, . * ~ TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES i REACTOR BUILDING SIIOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit
""**U RB103 RB106 RB109 RWO73x02 RWO73x03 i Total Beta Activity (dpm/100 cnf)* '
# 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 Removable Beta Activity (dpm/100 cni)**b
# 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 cnf Guideline Satisfied Yes Yes Yes Yes Yes Exposure Rata at i m (gR/h) i
# of Exposure Rate Measurements 17 9 NA 5 5 Nct Mean (X) 0.5 0.5 NA 0.9 0.9 i 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-55,11-3, and Ni-63 as appropriate. -
b All alpha removable activity was less than 12 dpm/100 cm2 , ;
' Represents survey unit radiological status prior to additional remediation. ,
i i l Shoreham Reactor fluilding and Phase 4 - February 16. lWs 65
,g ,
. 4 TABLE 5 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS REACTOR BUILDING SIIOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit
"** U SU001 SU002 SU004 SU014x12 SU014x05 Total Beta Activity (dpm/100 cid)"
# of Direct Measurements 30 5 18 6 10 Mean(X) 170 110 -70 130 -65 LIPA X 470 260 210 960 -130 Fa 250 260 4 400 110 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes Removable Beta Activity (dpm/100 cal)"6
# of Smears 30 5 9 5 8 Mean(X) 0.6 0.6 -0.2 0.2 0.2 LIPA X 14 5.2 15 13 14 F= 1.7 2.0 1.6 2.7 2.4 Conditions and 1,000 dpm/100 cid Yes Yes Yes Yes Yes Guideline Satisfied Shorcham Reactor DuMmg am! Phase 4 - February 16, 1995 66
k, , TABLE 5 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit Summary SUO23 SUO58 SUO60 SUO61 Total Beta Activity (dpm/100 cuf)"
# of Direct Measurements 26 17 45 19 Mean(X) 50 32 390 280 LIPA 240 210 170 590
, 240 150 530 360 ,
Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Removable Beta Activity (dpm/100 cnf)*d'
# of Smears 15 17 45 19 Mean(X) -0.3 2.2 2.8 1.1 LIPA X NA 7.4 17 18 l
A 17 6.1 4.5 2.6 l 1,000 dpm/100 enf Yes Yes Yes Yes Guideline Satisfied 0 Activity levels corrected for Fe-55, H-3, and Ni-63 as appropriate. b All alpha removable activity was less than 12 dpm/100 cm2, ) l Shorcham Reactor Buildmg and Phase 4 - February 16. 1995 67
. 4 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.
Long 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. Long Island Power Authority, Brookhaven, New York. Shoreham Decommissioning Project Termination Survey Final Report, Volumes 1 and 2. October 1994b. Pittiglio, C. L. Letter 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. Letter 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. Letter to D. Fauver, U.S. Nuclear Regulatory Commission. October 25,1994b. l l Shoretuun Reactor Buikhng and Phase 4. February 16.1995 68
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- APPENDIX A MAJOR INSTRUMF.NTATION P
l Shoreham Reactor Buildmg and Phase 4 - February 16. 1995 l 1 i
I,. Ie I APPENDIX A i
! MAJOR INSTRUMENTATION i 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 l Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM) Eberline " Rascal" Ratemeter-Scaler ! Model PRS-1 (Eberline, Santa Fe, NM) ! Ludlum Ratemeter-Scaler - Model 2221 ! (Ludlum Measurements, Inc., Sweetwater, TX)
- Detectors t 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) Ludlum Gas Proportional Detector M@l 43-37 ~ E.ketive Area,550 cm 2 , (Ludlum Measurements, Inc., , Sweetwater, TX) ! i Shorcham Reactor Buildmg and Phase 4 - February 16. 1995 A-1
i >Q, 14
- Ludlum Gas Proportional Detector Model 43-68 Effective Area,100 cm 2 (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: Lead 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 i (Gamma Products, Palos Hills, IL) and Multichannel Analyzer 3100 Vax Workstation , (Canberra, Meriden, CT) Low Background Gas Proportional Counter Model LB-5100-W l (Oxford, Oak Ridge, TN) ! Shorcham Reactor Building and Phase 4 February 16, 1995 A-2 l u 1 l i
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APPENDIX B SURVEY AND ANALYTICAL PROCEDURES i l Shoreham Reactor Building and Phase 4. February 16.1995
x,l'.. l I APPENDIX B SURVEY AND ANALYTICAL PROCEDURES i SURVEY PROCEDURES Surface Scans Surface scans were performed by passing the probes slowly over the surface; the distance , between the probe and the surface was maintained at a minimum - nominally about 1 centimeter i 1
. (cm). A large surface area, gas proportional 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 i 2 centimeters [cm ]) hand-held detectors. Identification of elevated levels was based on increases ! in the audible signal from the recording and/or indicating instrument. Combinations of detectors and instruments used for the scans were: Alpha - gas proportional detector with ratemeter-scaler : ZnS scintillation detector with ratemeter-scaler i Beta - gas proportional detector with ratemeter-scaler > GM pancake detector with ratemeter-scaler I Gamma - NaI scintillation detector with ratemeter Surface Activity Measurements Measurements of total beta activity levels were performed using Geiger-Mueller (GM) and gas ; proportional detectors with portable ratemeter-scalers. Count rates in counts per minute (cpm), which were integrated over 1 minute in a static position, were converted to activity levels in disintegrations per minute /100 square centimeters ; i Shoreham Reactor Building and Phase 4 - February 16. 1995 B-1
h,, . o (dpm/100 cm2 ) 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 cm2 and 100 cm 2, respectively. Surface activity measurements in all confirmatory survey units, except for RB068, RWO73 x02, RWO73 x03, and SU001 that exceeded the normal background distribution were corrected for 2 the iron-55 contribution by multiplying the dpm/100 cm f eld 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 instrument response level at which the detector output could be considered above background was defined as the critical level (I,). This level was defined for each detector / instrument combination as follows: 1.96 ampl e count rate , Background count rate
% 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 Buildmg and Phase 4 Febriary 16. 1995 B-2
Vx,, '. e f the surface,was wiped. Smears were placed in labeled envelopes with the location and other pertinent information recorded. ; I Exposure Rate Measurements
.l t
Measurements of gamma exposure rates were performed using a' pressurized ionization chamber l (PIC) set 1 meter from the surface, j i Water Sampline Approximately 3.8 liters of water was collected, the sample was transferred to a plastic ; container, sealed, and labeled in accordance with ESSAP survey procedures. Sediment Sampline ; Approxim.aely kilogram of sediment was collected from the trench, placed in a plastic container, sealed and labeled in accordance with ESSAP survey procedures. ANALYTICAL PROCEDURES : i Removable Activity Smears were counted on a low background gas proportional system for gross alpha, and gross beta activity. , Gamma Spectrometry 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 germani_m detectors coupled to a pulse height analyzer system. Background and Compton stripping, peak search, peak identification, f and concentration calculations were performed using the computer capabilities inherent in the ! i Shortham Reactor Buildmg azul Phase 4. February 16. 1995 B-3 J
- y i
analyzer system. The energy peak used for determining the activity of the radionuclide of
^
concern was: , Co-60 1.173 MeV 1 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- i represent the 95% confidence level for that data. These uncertainties were calculated based on i both the gross sample count levels and the associated background count levels. Additional ! i uncertainties, associated with sampling and measurement procedures, have not been propagated into the data presented in this report. ! i 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 i other radionuclides in samples, the detection limits differ from sample to sample and instrument f to instrument. [ i CALIBRATION AND QUALITY ASSURANCE { Calibration of all field and laboratory instrumentation was based on standards, traceable to the { National Institute of Standards and Technology (NIST), when such standard were available. In l cases where they were not available, standards of an industry recognized organization were used. I Calibration of pressurized ionization chambers was performed by the manufacturer. t i Analytical and field survey activities were conducted in accordance with procedures from the j following documents of the Environmental Survey and Site Assessment Program: i l Shoreham Reactor Building and Phase 4 - February 16. IM O-4 i i
- 4. ,_.
- Survey Procedures Manual, Revision 8 (December 1993) 1 1
- Laboratory Procedures Manual, Revision 8 (August 1993) l l
- Quality Assurance Manual, Revision 6 (July 1993) l 1
The procedures contained in these manuals were developed to meet the requirements of DOE l Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to assess processes during their performance. i I Quality control procedures include:
- Daily instrument background and check-source measurements to confirm that f equipment operation is within acceptable statistical fluctuations.
- Participation in EPA and DOE /EML Quality Assurance Programs. f
- Training and certification of all individuals performing procedures.
- Periodic internal and external audits. !
I I i f I t I f Shoreham Reactor Buildmg and Phuc 4 - February 16,1995 B_j
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APPENDIX C ( REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS l l l 1 1 l l l \ . 1 Shoreham Reactor Building and Plane 4 - February 16, 1995
e Q < = U.S. ATOMIC ENERGY COMMISSION Juna1974 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 of license, renewal," of 10 reactor operation is not permitted. Other activities CFR Part 50, " Licensing of Production and Utilization from the reactor and placing it in storage (either onsite Facilities," requires that each license to operate a or offsite) may be continued. production and utilization facility be issued for a specified duration. Upon expiration of the specified A licensee having a possession-only license must period, the license may be either renewed or terminated retain, with the Part 50 license, authorization for by the Commission. Section 50.82, " Applications for special nuclear material (10 CFR Part, 70, "Special termination of licenses," 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 comp; ants, and sources are removed health and safety of the public. This guide describes from the facility. Appropriate administrative controls methods and procedures considered acceptable by the a' ? facility requirements are imposed by the Part 50 Regulatory staff for the termination of operating beense 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 consuhed concerning facility is maintained in a safe condition and not this guide and has concurred in the regulatory position. operated. B. DISCUSSION A possession-only license permits various options and procedures for decommissioning, such as When a licensee decides to terminate his nuclear mothballing, entombment, or dismantling. The reactor operating license, lie may, as a first step in the requirements imposed depend on the option selected. process, request that his opera:mg license be amended to restrict him to possess but not operate the facility. Section 50.82 provides that the licensee may The advantage to the licensee of converting to such a dismantle and dispose of the component parts of a possession-only license is reduced surveillance nuclear reactor in accordance with existing regulations. requirements in that periodic surveillance of equipment For research reactors and critical facilities, this has USAEC REGULATORY GUIDES g,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,t,,,,,,,,,,,,,, Regulatory Guidea are issued to desutbe and rnake evallable to the public desired to the U.S. Atomic Energy Commission. Washington, D.C. 20545, methods ecceptable to the AEC reDuistory staff of implementmg specific parts Attention: Director of Regulatory Standards. Comments and suggestions for I of the Commession's 'soulations, to originate techniates used by the staff en Wwnenu ln Wee om n oma# W sw h sent m W evaluating specsfee probsems or postuinted ecc6 dents, os to prov6de guidance to Secretary of the Comm saron. U.S Atom 6c Energy comm6ssion. Washmpton, apphcents. Regulatory Gu6 des are not subelltutes for regulations and D.C. 20646. Attention: Ch6ef, Public Proceedmgs Staff. compl*ance with them is not ,equired. Methods and solutions different from those set out in the guides will tre acceptabte if they provide e basis for the The puedes are assued in the following ten broad divisions. l j find requate to the sesuance or contmuence of a permet or hcense by the
- 1. Power Reactors 6. Products
- 2. Research and test Reactors 7. Transportation i Published guides will be revised poeiodicahy. as appropriate, to accommodate Nels and Mamals Fames A occupatmat he comments and in renect new informanon or empenence. n and S n0 n t Rev6ew f , p p, ,
I Note: Section electronically reproduced from photocopy. C-l
=
- o*o usually meant the disassembly of a reactor and its b. A description of measures that will be taken to l shipment organization for further use. The site from prevent criticality or reactivity changes and to l' which a reactor has been removed must be minimize releases of radioactivity from the facility.
decontaminated, as necessary, and inspected by the Commission to determine whether unrestricted access c. Any proposed changes to the technical l can be approved. In the case of nuclear power specifications that reflect the possession-only facility reactors, dismantling has usually been accomplished by status and the necessary disassembly / retirement shipping fuel offsite, making the reactor inoperable, activities to be performed. and disposing of some of the radioactive components.
- 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.
secured on the site. 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 remaining on the Four alternatives for retirement of nuclear reactor site and (2) the integrity of the physical barriers, facilities are considered acceptable by the Before areas may be released for unrestricted use, they Regulatory staff. These are: must have been decontaminated or the radioactivity , must have decayed to less than prescribed limits n. 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 Entomhment. In-place entombment consists of sealing all the remaining highly
- 1. APPLICATION 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 amend 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. Atomic Energy Commission, the period of time in which significant quantities Washington. D.C. 20545. The request should include (greater than Table 1 levels) of radioactivity remain the following information: with the material in the entombment. An appropriate and continuing surveillance program
- a. A description of the current status of the facility, should be established under a possession-only license.
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- c. Removal of Rndloretite. - Components and b. The physical barriers to unauthorized entrance Dismantling. All fuel assemblies, radioactive fluids into the facility, e.g., fences, buildings, welded doors, l and waste, and other materials having activities and access openings, should be inspected at least i above accepted unrestricted activity levels (Table 1) quarterly to assure that these barriers have not should be removed from the site. The facility deteriorated and that locks and locking apparatus are owner may then have unrestricted use of the site intact.
with no requirement for a license, if the facility owner so desires, the remainder of the reactor c. A facility radiation survey should be performed - I facility may be dismantled and all vestiges removed at least quarterly to verify that no radioactive material and disposed of. is escaping or being transported through the containment barriers in the facility. Sampling should ;
- d. Conversion to a New Nuclear System or a '
be done along the most probable path by which Fossil Fuel System. This alternative, which applies radioactive material such as that stored in the inner l only to nuclear power plants, utilizes the existing containment regions could be transportc.d to the outer ; turbine system with a new steam supply system. regions of the facility and ultimately to the environs. ! The original nuclear steam supply system should be separated from the electric generating system and d. An environmental radiation survey should be l disposed of in accordance with one of the previous performed at least semiannually to verify that no ; three retirement alternatives. significant amounts of radiation have been released to the enviromnent from the facility. Samples such as
- 3. SURVEILLANCE AND SECURITY FOR THE soil, vegetation, and water should be taken at locations !
RETIREMENT ALTERNATIVES WIIOSE for which statistical data has been established during [ FIN A L STATUS REQUIRES A reactor operations. f POSSESSION-ONLY LICENSE
- e. A site representative should be designated to be A facility which has been licensed under a responsible for controlling authorized access into and possession-only license may contain a significant movement within the facility.
amount of radioactivity in the form of activated and ; contaminated hardware and structural materials. f. Administrative procedures should be established l Surveillance and commensurate security should be for the notification and reporting of abnormal i provided to assure that the public health and safety are occurrences such as (1) the entrance of an unauthorized not endangered. person or persons into the facility and (2) a significant ,
- a. Physical security to prevent inadvertent exposure change in the radiation or contamination levels in the !
of personnel should be provided by multiple locked facility or the offsite environment. ; barriers. The presence of these barriers should make I it extremely difficult for an unauthorized person to gain g. The following reports should be made: [ access to areas where radiation or contaminationlevels exceed those specified in Regulatory Position C.4. To (1) An annual report to the Director of l prevent inadvertent exposure, radiation areas above Licensing, U.S. Atomic Energy Commission, 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 l 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. l structures and/or shielding material. Means such as a , remote-readout intrusion alarm system should be (2) An abnormal occurrence report to the provided to indicate to designated personnel when a Regulatory Operations Regional Office by telephone 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 instead of the physical barriers and the intrusion alarm reported in the annual report described in the preceding systems. item. 1 Note: Section electronically reproduced from photocopy. C-3
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. 4 !
- h. Records or logs relative to the following items d. Upon request, the Commission may authorize a should be kept and retained until the license is licensee to relinquish possession or control of premises, terminated, after which they must be stored with other equipment, or scrap having surfaces contaminated in plant records: excess of the limits specified. This may include, but is not limited to, special circumstances such as the (1) Environmental surveys, transfer of premises to another licensed organization (2) Facility radiation surveys, that will continue to work with radioactive materials.
Requests for such authorization should provide: (3) Inspections of the physical barriers, and (4) Abnormal occurrences. (1) Detailed, specific information describing the premises, equipment, scrap, and radioactive
- 4. DECONTAMINATION FOR RELEASE FOR contaminants and the nature, extent, and degree of UNRESTRICTED USE residual surface contamination, if it is desired to terminate a license and to (2) A detailed health and safety analysis indicating ,
eliminate any further surveillance requirements, the that the residual amounts of materials on surface areas, facility should be sufficiently decontaminated to prevent together with other considerations such as the risk to the public health and safety. After the prospective use of the premises, equipment, or scrap, decontamination is satisfactorily accomplished and the are unlikely to result in an unreasonable risk to the site inspected by the Commission, the Commission may health and safety of the public. authorize the license to be terminated and the facility abandoned or released for unrestricted use. The e. Prior to release of the premises for unrestricted licensee should perform the decontamination using the use, the licensee should make a comprehensive following guidelines: radiation survey establishing that contamination is within the limits specified in Table 1. A survey report
- a. The licensee should make a reasonable effort to should be filed with the Director of Licensing, U.S.
eliminate residual contamination. Atomic Energy Commission, Washington, D.C. 20545, with a copy to the Director of the Regulatory
- b. No covering should be applied to radioactive Operations regional Office having jurisdiction. The surfaces of equipment of structures by paint, plating, or report should be filed at least 30 days prior to the other covering material until it is known that planned date of abandonment. The survey report contamination levels (determined by a survey and should:
documented) are below the limits specified in Table 1. In addition, a reasonable effort should be made (and (1) Identify the premises; documented) to further minimize contamination prior to any such covering. (2) Show that reasonable effon has been made to reduce residual contamination to as low as practicable
- c. The radioactivity of he interior surfaces of levels; pipes, drain lines, or ductwork should be determined by making measurements at all traps and other (3) Describe the scope of the survey and the general appropriate access points, provided contamination at procedures followed; and these locations is likely to be representative of contamination on the interior of the pipes, drain lines, (4) State the finding of the survey in units specified or ductv ork. Surfaces of premises, equipment, or in Table 1.
scrap which are likely to be contaminated but are of such size, construction, or location as to make the After review of the report, the Commissiort may surface inaccessible for purposes of measurement inspect the facilities to confirm the survey prior to should be assumed to be contaminated in excess of the granting approval for abandonment. permissible r.tdiation limits. Note: Section electronicah reproduced from photocopy. C-4
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- 5. REACTOR RETIREMENT PROCEDURES As indicated in Regulatory Position C.2, several alternatives are acceptable for reactor facility retirement. If minor disassembly or "mothballing" is planned, this could be done by the existing operating and maintenance procedures under the license in effect.
Any planned actions involving an unreviewed safety question or a change in the technical specifications should be reviewed and approved in accordance with the requirements of 10 CFR $ 50.59. If major structural changes to radioactive components of the facility are planned, such as removal of the pressure vessel or major components of the primary system, a dismantlement plan including the information required by i 50.82 should be submitted to the Commission. A dismantlement plan should be submitted for all the alternatives of Regulatory Position C.2 except mothballing. Ilowever, minor disassembly activities may still be performed in the absence of such a plan, provided they are permitted by existing operating and maintenance procedures. A dismantlement plan should include the following:
- a. A description of the ultimate status of the facility
- b. A description of the dismantling activities and the precautions to be taken.
- c. A safety analysis of the dismantling activities including any effluents which may be released.
- d. A safety analysis of the facility in its ultimate status.
Upon satisfactory review and approval of the dismantling plan, a dismantling order is issued by the Commission in accordance with 9 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. Note: Section electronically reproduced from photocopy. C-5
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TABLE 1 L ACCEPTABLE SURFACE CONTAMINATION LEVELS - l i [, - J E
' Nuclide' Average6 ' Maximum6d Removabic6 '
U-nat, U-235, U-238, and , 2 2 i associated decay products 5,000 dpm a/100 cm 15,000 dpm a/100 cm 1,000 dpm a/100 cm2 t Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa 231, l Ac 227, I-125, I-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th-nat, Th-232, Sr-90, Ra-223, , Ra-224, U-232, I-126, I-131,1-133 1,000 dpm/100 cm2 3,000 dpm/100 cm2 200 dpm/100 cm2 Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 and l others noted above. 5,000 dpm By/100 cm2 15,000 dpm #7/100 cm2 1,000 dpm Sy/100 cm2
'Where surfxe contamination by both alpha- and beta-gamma-emitting nuclides exists, the limits established for alpha- and beta- ganuna-emitting nuclides should apply independently.
6 As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation.
' Measurements of average contaminant should not be averaged over more than 1 square meter. For objects of less surface area, the average should be derived for each such object.
dThe maximum contamination level applies to an area of not more than 100 cm2 .
The amount of removable radioactive material per 100 cm2of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
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