ML20077M846
| ML20077M846 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 10/31/1994 |
| From: | Vitkus T OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20077M839 | List: |
| References | |
| CON-FIN-A-9076 NUDOCS 9501130191 | |
| Download: ML20077M846 (70) | |
Text
{{#Wiki_filter:. __ _ . .. .. DRAFT REP RT 1 CDNFlRMATORY SURVEY
; F}THE:RADWASTE BUILDING, SUPPRESSION POOL, AND PHASE 2 SYSTEMS SHOREHAM NUCLEAR POWER STATION Bh00KHAVE'N, NEW YORK
[ DOCKET No. 50-322] T.J. VITKUS Prepared for the DMson of Waste Management Headquarters Office w U.S. Nuclear Regulatory Commision
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. t CONFIRMATORY SURVEY OF TIIE RADWASTE BUILDING, SUPPRESSION POOL, t AND PIIASE 2 SYSTEMS '
SIIOREllAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK 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 Prepared for the U.S. Nuclear Regulatory Commission Headquarters Office , Sponsored by the Division of Waste Management DRAFT REPORT , OCTOBER 1994 i i This draft report has not been given full review and patent clearance, and the dissemination of I its information is only for official use. No release to the public shall be made without the approval of the Office of Information Services, Oak Ridge Institute for Science and Education. This report is based on work performed under an Interagency Agreement (NRC Fin. No. A-9076) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy. 1 Oak Ridge Institute for Science and Education performs complementary work under contract ' number DE-AC05-760R00033 with the U.S. Department of Energy. 1 Shcreham inrathaven. NY . Ock+er 20,1994 b:\essap\ reports \shoreham\brookhaven .001
ACKNOWLEDGEMENTS The author would like to' acknowledge the significant contributions of the following staff members: FIELD STAFF ' E. H. Bjelland E. H. Montalvo J. R. Morton J. L. Payne LABORATORY STAFF R. D. Condra J. S. Cox M. J. Laudeman CLERICAL STAFF D. A. Cox R. D. Ellis K. E. Waters i f ILLUSTRATOR i T. D. Herrera i Shortham-Bnsthaven. NY Ockter 20,1994 h:kssapurports\shoreham%rookhaven.001
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TABLE OF CONTENTS ! 1 i PAGE ! Li st o f Fig u res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i i i List of Tables ..............................................iv v Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v ; Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ! Site Description ............................................. 2 Obj ec ti ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Document Review and LIPA Procedure Surveillance . . . . . . . . , . . . . . . . . . . . . . . 4 ; Procedures ......................,. .. ,,,,,,,,,,,,,,,,,,,,,4 Findings and Results i
................... 9 j Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 S u m mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 [
P References ...............................................47 r Appendices: i Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures i Appendix C: Regulatory Guide 1.86 Termination of Operating Licenses ! for Nuclear Reactors I I 1 j l l hBem& haven, NY . Octatier20,1994 1 b:\essap\ reports \shoreharaibrmahaven.001 j i l I k
LIST OF FIGURF3 , PAGE i l FIGURE 1: Location of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 15 S FIGURE 2: Plot Plan of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 16 FIGURE 3: Radwaste Building, Elevations 15'6"/19'6" Floor - Plan-Surveyed Areas ...............................17 FIGURE 4: Radwaste Building, Eevation 37'6" Floor Plan- - Surveyed Areas ...................................18 FIGURE 5: Radwaste Building, Elevation 50'6"/52'6" Floor Plan-Surveyed Areas ...................................19 FIGURE 6: Reactor Building, Elevation 8'0" Floor Plan, Suppression Pool-Surveyed Areas ...............................20 FIGURE 7: Radwaste Building, Radwaste 15' North Hallway (RW013)- [ Measurement and Sampling IAcations . . . . . . . . . . . . . . . . . . . . . . 21 FIGURE 8: Radwaste Building, Cation / Anion Regen. and Resin Storage ! Tanks Area (RW017)-Measurement and Sampling Locations . . . . . . . 22 FIGURE 9: Radwaste Building, Liner Fill Stations /BW Storage Rooms (Cubicle A) (RWO23)-Measurement and Sampling Locations . . . . . . . 23 FIGURE 10: Radwaste Building, Waste Evap./ Regen.Evap. Distil. Room (RWO40)-Measurement and Sampling IAcations . . . . . . . . . . . . . . . 24 FIGURE 11: Radwaste Building, Radwaste 37' North Hallway (RWO42)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 25 FIGURE 12: Radwaste Building, "B" RW O/G HEPA After Filter Area (RWO69A)-Measurement and Sampling Locations . . . . . . . . . . . . . . 26 : FIGURE 13: Radwaste Building, "C&D" Dryer Skid Area (RWO69B)- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 27 FIGURE 14: Radwaste Building, Off Gas Desiccant Dryers Area (RWO72)- Measurement and Sampling IAcations . . . . . . . . . . . . . . . . . . . . . . 28 , 1 FIGURE 15: Reactor Building, Suppression Pool, NW Quadrant (SP004) IAwer Walls and Floor-Measurement and Sampling IAcations . . . . . . 29 Hwrt6am Brathaven NY . Orteer 20.1994 ii h *essap\rtports%orcharn\brotAhaven.001
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s... . e'~ : LIST OF FIGURES (Continued) '! i PAGE l 1 FIGURE 16: Reactor Building, Suppression Pool, NW Quadrant (SP004) Upper ; Walls -Measurement and Sampling Locations . . . . . . . . . . . . . . . . . 30 FIGURE 17: Reactor Building, Suppression Pool, Area Inside Vessel Pedestal (SP005)-Measurement and Sampling locations . . . . . . . . . . . . . . . . 31 i FIGURE 18: Reactor Building, HPCI Valves E41-OlV-3049 and 3050 (SUO12)-Measurement and Sampling Imcations ...............32 ! FIGURE 19: Radwaste Bailding, Radwaste Influent Drain System, ; Drain Sump Tank-054 (SU14 x09)-Measurement ! and Sampling Locations ..............................33 ! FIGURE 20: Radwaste Building, Equipment Drain System, ! Drain Sump Tank-071 (SU014 x 10)-Measurement i and Sampling Locations ..............................34 I FIGURE 21: Radwaste Building, Radwaste Equipment / Components, Flat Bed Floor Drain Filter 1G11-FL-012 (SUO14 x12)-Measurement i and Sampling Imcations ..............................35 . 1 FIGURE 22: Radwaste Building, Radwaste Equipment / Components, Waste Collector Tank 1011-TK-10A (SUO14x12)- , Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 36 FIGURE 23: Reactor Building, Reactor Water Clean-up System Components ! (SUO15)-Measurement and Sampling locations ...............37 FIGURE 24: Radwaste Building, Condensate Demineralizers, Tank 1N52-DE-002E ; (SUO43)-Measurement and Sampling Locations ...............38 ! i l wn.*um. w - oen 20 im iii a.,wa.e=+-a. mmi :
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i i LIST OF TABLES PAGE TABLE 1: Summary of Surface Activity levels . . . . . . . . . . . . . . . . . . . . . . . 39 - l TABLE 2: Interior Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , . 41 ! TABLE 3: Embedded Piping Program Summary . . . . . . . . . . . . . . . . . . . . . . 42 , TABLE 4: Confirmatory Radiological Status Summary-Structures . . . . . . . . . . . 44 , TABLE 5: Confirmatory Radiological Status Summary-System. . . . . . . . . . . . . 46 l A r i c h a W t 1 i l
%=bam BmAbam, NY . Ockdet 20, IW jv h:kssap\npons\shcham\bmihaven21
- i a b ABBREVIATIONS AND ACRONYMS ac acre
- ASME American Society of Mechanical Engineers r em2 square centimeter '
cpm ' counts per minute DOE. Department of Energy , dpm/100 cm 2 disintegrations per minute per 100 square centimeters EML Environmental Measurements Laboratory EPA Environmental Protection Agency ESSAP Environmental Survey and Site Assessment Program ft2 square feet ha hectare GM Geiger-Mueller km kilometer L, critical level LlLCO Long Island Lighting Company > LIPA Long Island Power Authority m meter m2 square meter MDA minimum detectable activity mi mile NaI sodium iodide , NIST National Institute of Standards and Technology NRC Nuclear Regulatory Commission ; ORISE Oak Ridge Institute for Science and Education QA Quality Assurance QC Quality Control RW# Radwaste Building structural survey unit designation SNPS Shoreham Nuclear Power Station SP# Suppression Pool structural survey unit designation - SU# system survey unit designation { I i Shmban-Brutahaven, NY Ockeer 20,1994 V h;kasap\ reports \shortham\broc&havca.001
a CONFIRMATORY SURVEY OF TIIE RADWASTE BUILDING, SUPPRESSION POOL, AND PIIASE 2 SYSTEMS SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK INTRODUCTION AND SITE IIISTORY The Long Island Lighting Company (LILCO) constructed a boiling water reactor, known as the Shoreham Nuclear Power Station (SNPS), which was designed to provide a gross electrical output of 849 megawatts. Reactor criticality was achieved in February 1985. Low power testing, in accordance with U.S. Nuclear Regulatory Commission (NRC) License No. NPF-82 (NRC Docket File No. 50-322), which permitted reactor operations at levels not to exceed 5% ' of full power, commenced in July 1985. Reactor operations continued intermittently until January 1989, at which time power generating operations were terminated. The total operating history was equivalent to 2.03 effective full power days of fuel exposure. Irradiated fuel, which was a standard low enrichment (2 to 3% uranium-235) uranium fuel, was subsequently removed from the reactor vessel and placed into the spent fuel pool in August 1989. Various reactor components, piping systems, and other equipment became radiologically contaminated as a result of reactor operation. The primary contaminants which were identified during characterization studies include iron-55, cobalt-60, nickel-63, and smaller quantities of tritium, carbon-14, nickel-59, manganese-54, zinc-65, and europium-152.' The long Island Power Authority (LIPA) was established to decommission the facility and release the site for unrestricted use. LIPA's decommissioning plan was approved for implementation by the NRC in June 1992 and includes decontamination (grinding, high pressure washing, etc.) or removal of contaminated portions of the reactor and other plant systems and equipment. A major consideration of the decommissioning plan is to maintain the integrity when possible, of plant structures and systems. Activities involved with the decommissioning and termination surveys will be conducted in 4 phases. The initial phase involved the termination survey of the internal components of the main turbine; which has since been followed by Shmhern-Bra & haven, NY . October 20,1994 b:\easap\reportsinhartharnerookhaven.001
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termination surveys of the remainder of the structures and systems located within the Turbine l Building as well as the site grounds and building exteriors. Phase 2 included the termination survey of the reactor suppression pool and several systems. Phase 3 involved the termination surveys for the Radwaste Building. ! 1 It is the policy of the NRC to perform confirmatory surveys of facilities that have undergone decommissioning and have requested NRC license termination. The NRC Headquarters' Division of Waste Management has requested that the Environmental Survey and Site t Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) conduct confirmatory radiological surveys and related activities for the SNPS decommissioning ; project as the various decommissioning milestones are completed. The results of the . confirmatory survey of Phase 1, the turbine internal components and the Turbine Building, Site I Grounds, and Building Exteriors, are the subject of separate reports.2.3 This report describes i the results of the confirmatory process which has been completed for the Radwaste Building (Phase 3), Suppression Pool (Phase 2), and Phase 2 systems, i I SITE DESCRIPTION j SNPS is located in the Town of Brookhaven, New York on the north shore of Long Island, approximately 80 k.m (50 mi) east of La Guardia Airport and the confluence of the East River f; and long Island Sound (Figure 1). Reactor and supporting operations were conducted within a 32.4 ha (80 ac) portion of a larger 202 ha LILCO owned parcel ofland that is bounded on the ; I north by IAng 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 I krown as the power block, where radiological controls were necessary (Figure 2). Each of the I buildings that have been or will be addressed during the confirmatory surveys are located here , and are shown on Figure 2 as the Turbine Building, the Reactor Building, and the Radwaste i Building. Radwaste Building construction is predominately of concrete and structural steel with i a total floor space of approximately 4,700 m2 (51,000 ft 2) which is divided between three levels at elevations 15'6"/19'6",37' 6", and 50'6"/52'6" (Figures 3 through 5). The systems and : i mo h m. ru . m+n 20 im 2 a. pw.ho-6 a-*h.=mi i
equipment housed within the Radwaste Building include the condensate demineralizers, the liquid radwaste system, the solid radwaste storage area, the crane z.nd truck bay, the makeup water treatment plant, chemical support system, and a portion of the off-gas radwaste system. The Suppression Pool is located on the 8' level of the Reactor Building and is constructed of steel plate (Figure 6). Surfaces and components within the buildings remain essentially intact following decommissioning activities. Termination sarveys have been performed in accordance with Draft NUREG/CR-5849.4 LIPA has classified plant systems, building surfaces, and outside areas into two categories for survey, which are based on the potential for residual contamination. The two categories, referred to as affected or unaffected, are defined as follows: "affected areas are those areas which are potentially contaminated or have known contamination, or a system which circulated, stored or processed radioactive materials such that they could become contaminated, or experience neutron activation, or where records indicated spills or other occurrences may have resulted in contamination; unaffected areas are those portions of the SNPS that are not expected to contain residual radioactivity."5 Area classification was determined by radiological use history, environmental monitoring activities, and the results of the previous characterization survey. Affected and unaffected areas are further subdivided into survey units. Survey units are categorized as structures (floors, walls, ceilings, and exterior surfaces of piping and equipment), plant systems (equipment and piping internals), and exterior areas (grounds and building exteriors). In addition, affected survey units also have sub-classifications as suspect or non-suspect, and may also be classified as alpha affected if involved with fuel handling or storage. For the Radwaste Building, Suppression Pool, and Phase 2 Systems, there were a total of 77 survey units addressed, of which,60 were structures and 17 were systems. Sixty-five of these survey units were classified by the licensee as affected. OBJECTIVES The objectives of the confirmatory activities were to provide independent document reviews, review and perform field observations of the LIPA procedures for embedded piping surveys, and Entuun Ilmthoven, NY - Ockda 20. m 3 a. re*=uem*6m.m
develop radiological data for use by the NRC in evaluating the adequacy and accuracy of the licensee's procedures and termination survey results. DOCUMENT REVIEW AND LIPA PROCEDURE SURVEILLANCE ESSAP reviewed LIPA's termination survey procedures and the termination survey release records for those survey units selected for confirmatory survey.5,6 Documents were reviewed for adequacy, accuracy, completeness, and consistency. Data were reviewed for appropriateness ' of calculations and interpretations relative to the guidelines. In addition, ESSAP reviewed the l applicable procedures and records for both the calibration ofinstruments used in, and the survey data generated for,3 representative sections of embedded piping. Together with observational surveillance of the resurvey of one section of selected embedded piping, the documentation was evaluated for appropriateness and consistency in field application. PROCEDURES During the period August 22 through 25,1994, an ESSAP team visited the SNPS and performed i independent visual inspections, measurements, and sampling of the Radwaste Building, Suppression Pool, and Phase 2 Systems. Surveys were performed in accordance with a survey plan submitted to and approved by the NRC.7 Nine structural survey units and either the complete, or components of,4 system survey units were selected for confirmatory surveys. Survey units were selected either randomly by ESSAP or based on recommendations of the NRC site representative. Survey unit designators are alpha-numeric with the first figures designating the type of unit, structural (building specific, RW=Radwaste, SP = Suppression Pool), or system, i followed by a three digit numeric reference. Subunits are given an additional two digit l designation preceded by X. The survey units selected and the respective classification for each were: l 1 Whhm. NY - Oc*r 20, IW 4 WaupWohhanWahmmt
4* .' . Survey Unit Survey Unit Name Affected (A) Structure / System / Unaffected (U) Building Grounds RW013 Radwaste 15' North Hallway A structure RW017 Cation / Anion Regen and Resin A structure Storage Tanks Area RWO23 Liner Fill Stations /BW Storege A structure Rooms (Cubicle A) RWO40 Waste Evap./ Regen. Evap. Distil. A structure Room RWO42 Radwaste 37' North Hallway A structure RWO69 "B" RW O/G HEPA After Filter A structure Area RWO72 Off Gas Desiccant Dryers Area A structure SP004 Suppression Pool - NW Quadrant A structure SP005 Suppression Pool - Area Inside A structure Vessel Pedestal SUO12 High Pressure Coolant Injection A system : Valves E41-OlV-3049 and 50 SUO14 x09 Radwaste Influent Drain System, A system Drain Sump Tank-054 SUO14 x10 Radwaste Building Equipment A system i Drain System, Drain Sump Tank-071 l SU014 x12 Radwaste Equipment / Components, A system Flat Bed Floor Drain Filter IG11- i FL-012 and Waste Collector Tank l 1Gil-TK-10A l SUO15 Reactor Water Cleanup System A system SUO43 Condensate Demineralizers, Tank A system IN52-DE-002E Figures 3 through 6 indicate the structural survey units surveyed. Confirmatory surveys for SU012, SU014x09, SUO14x10, SU014x12, and SUO43 involved individual component (s) rather than the entire survey unit. Shehmn Buthavm, NY - October 20, im 5 6:w-rver w wn-S-Ahm.m ;
I 1 Field survey activities were conducted in accordance with the applicable sections of the ESSAP Survey Procedures and Quality Assurance Manuals. Appendices A and B provide additional information regarding instrumentation and procedures. The following procedures apply to survey units selected for independent confirmatory surveys. SURVEY PROCEDURES Reference System LIPA established the grid system which ESSAP used for referencing measurement and sampling locations. The grid size or reference interval established by LIPA for a given survey unit was dependent upon the survey unit classification (affected vs. unaffected) and surface (floor, lower wall, upper wall, ceiling, or equipment). Typically, floor and lower wall grid blocks were 1 m x 1 m. Upper surfaces and equipment were either referenced to these grids or other prominent building features. Systems were referenced by either a distance from a specific point, by drawings, or 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. Additional scans were performed over portions of upper wall, ceiling, and/or system surfaces where material may have settled or accumulated. Locations of elevated direct radiation detected by scans were marked for further investigation. Scans were performed using gas proportional, GM, and NaI detectors coupled to ratemeters or ratemeter-scalers with audible indicators. Surface Attivity Measurements For each structural survey unit, ESSAP performed a minimum of thirty direct measurements for total beta surface activity. ESSAP also performed additional direct measurements at locations of elevated direct radiation detected by surface scans. Alpha surface activity measurements were l ShwehBmumn, NY - Ockeer 20,1994 6 WmpgaWimchamWmumn001
not required as the selected survey units were not classified as alpha affected and there was no alpha contamination identified by surface scans. Measurements were performed using gas proportional or GM detectors coupled to ratemeter-scalers. A smear sample for determining removable activity levels was collected from each direct measurement location. Figures 8 through 17 show measurement and sampling locations. Exposure Rate Measurements Exposure rate measurements in structural survey units were performed at each accessible floor grid block where ESSAP had performed direct surface activity measurements. All exposure rates were measured at 1 m above surfaces using a pressurized ionization chamber (PIC). Figures 7 through 14 show measurement locations. Background exposure rates were previously determined during the confirmatory survey of the Turbine Building.' Systems LIPA provided access points into each system or system component listed on page 5 of this report. Beta and gamma surface scans were performed within the accessible portions of each system or component, followed by 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 4 to 30 measurements per system. Scans and direct measurements were performed using gas proportional, GM, and/or NaI detectors coupled to ratemeters or ratemeter-scalers. Figures 18 through 24 show measurement and sampling locations. Embedded Piping i Confirmation of the LIPA embedded piping termination survey program was accomplished through on-site review and observation of the LIPA procedures for instrument calibration and i embedded piping survey design and performance, review of the data and records for the ; 1 l l I Shehese Brmahaven. NY - Oekdict 20, !W 7 h&mpWehmhamWMhaven 001 l l
9 - termination surveys of selected embedded piping subunits, and independent measurements and sampling of selected sections of piping. Specific procedures for each activity are described below. The calibration records for the detector / instrument combinations, referred to as " pipe crawlers," used during termination survey data acquisitions for three selected embedded piping subunits were compared with, and evaluated against, the required calibration and operational check-out procedures and standards. These LIPA instrumentation procedures included the following: Control of Health Physics Instrumentation (SP Number 61X080.01), Chi-Square Test and Control Chart Review (SP Number 61X081.01), Detector Calibration (SP Number 66X020.11), and Instrument Calibration (SP Number 66X022.02). In addition, the recalibration of one detector / instrument combination was observed. ESSAP then performed independent measurements and calculations to confirm the level of, and distribution of, Co-60 activity on two of the custom sources which LIPA used for the calibration check on the pipe crawler detector assemblies. This information was used to evaluate the appropriateness of the LIPA stated detector efficiencies. Once these parameters were established, ESSAP confirmed the LIPA 2 computer generated surface activity measurement data conversions to dpm/100 cm , critical level calculations, and the action level calculations used by LIPA for identifying " hot spots" while surveying embedded piping. Three sections of embedded piping were selected for confirmatory evaluation and independent surveys. Embedded piping subunit selection was based on recommendations of the NRC site representative. ESSAP performed surface scans and direct measurements for beta activity over a length of each selected pipe run, nominally 4 m in either direction of the access point. Access was gained through sumps, drains, traps, or openings where pipe sections had been removed. A total of 17 direct measurements were made. The surface activity data collected was then compared to the LIPA generated data from the corresponding section of each pipe run and the NRC surface activity guidelines. In addition, for each of these subunits, the Termination Survey Design (LIPA Procedure SP Number 67X001.10) was compared to the field records and evaluated for appropriateness and consistent implementation. LIPA was also requested to wnom=. m - ocu+n 20. m. 8 w..wwawwww-m
resurvey portions of the selected embedded piping subunits and the results compared with the original termination survey data. SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data were returned to ESSAP's Oak Ridge, Tennessee laboratory for analysis and interpretation. Smears were analyzed for gross alpha and gross beta activity using a low background proportional counter. Smear and direct measurement data were converted to units of disintegrations per minute per 100 square centimeters (dpm/100 cm2). Because Fe-55 can not be adequately detected with field instrumentation, a correction factor of 1.2 was therefore applied to those surface activity measurements that exceeded background distribution levels, referred to as the critical level (Lc). LIPA developed, and the NRC approved, the use of this correction factor based on the observed Co-60 to Fe-55 activity ratio identified in characterization samples.8 8 Exposure rates were reported in microroentgens per hour ( R/h). Additional information concerning major instrumentation, sampling equipment, and survey and analytical procedures is provided in Appendices A and B. The 95 % confidence level ( .), in accordance with NUREG/CR-5849, was calculated for surface activity and exposure rates for each survey unit selected for confirmation. A direct comparison of the ESSAP and LIPA survey unit results was also performed. FINDINGS AND RESULTS DOCUMENT REVIEW ESSAP's review of the termination survey plan indicated that the document provided an adequate description of survey methodologies and general approaches. Comments were provided to the NRC in a January 12,1993 correspondence.' ESSAP's review of the termination survey final report, and release records for those survey units selected for confirmatory survey, indicated that the survey plan had been appropriately followed with no significant deviations. Data was appropriately converted, tested, and presented. n-a.a ou ny-a w ,20, m 9 u.. pwun-w.a...e __--.--___..---_--_-_____-__-__-________--_____-_=
i .+ . l SURFACE SCANS Alpha, beta, and gamma surface scans identified one small area of elevated direct beta radiation on a support bracket in SP004. All other surface scans were comparable to background levels. SURFACE ACTIVITY LEVELS The results of total and removable surface activity levels are summarized in Table 1. Total beta activity levels for the structural survey units ranged from -310 to 3,800 dpm/100 cm2 . The highest direct measurement was on the SP004 support bracket. Removable activity levels ranged from -1 to 6 dpm/100 cm2for alpha and -7 to 81 dpm/100 cm2 for beta. The mean residual activity in structural survey units ranged from 14 to 510 dpm/100 cm2 and -1.3 to 2 3.4 dpm/100 cm for total and removable beta activity, respectively. " Total beta activity levels in the surveyed systems ranged from -760 to 1,600 dpm/100 cm2 . The removable activity levels were -1 to 8 dpm/100 cm2 for alpha and -6 to 14 dpm/100 cm2 for beta. The mean beta activity levels for systems ranged from -310 to 98 dpm/100 cm2 for total activity and -1.3 to 1.7 dpm/100 cm2 for removable activity. EXPOSURE RATES As determined curing the Turbine Building survey, the interior background exposure rates ranged from 4 to 5 R/h and averaged 5 R/h at 1 m. Individual gross exposure rates within the Radwaste Building ranged from 4 to 6 pR/h at 1 m. The average gross exposure rates for all survey units ranged from 4 to 5 pR/h at 1 m. Table 2 provides a summary of the exposure rates. EMBEDDED PIPING Review of the embedded piping instrumentation calibration and calibration check; operational source cross-checks; survey procedure review and surveillance; and review of data conversions, Wrtham4rtdhaven, NY - October 20, im 10 hwwwmwnxis.=mi
critical levels, and action levels showed that all facets of the embedded piping survey program were performed in accordance with procedures and that the radiological status of embedded piping had been accurately presented. Table 3 provides a summary of the information relative to the following: LIPA calibration check source activity levels and efficiencies, action levels versus to instrumentation alarm settings that identify locations of elevated direct radiation requiring additional investigation, number of direct measurements performed for each embedded piping subunit to meet termination survey plan requirements, data conversion confirmation, and comparison of activity levels for sections of three embedded piping subunits where LIPA and ESSAP performed direct measurements. Surface scans performed in each piping section did not identify any locations of elevated direct radiation. The ESSAP 'o:.tl activity results for each subunit are as follows: SU014X02 (#365), 2 the range was -320 a n J dpm/100 cm and the mean was -20 dpm/100 cm 2; SU014X09 (#850), the range was -360 a 40 dpm/100 cm 2and the mean was -140 dpm/100 cm 2; and SU016X01 (#20), the range was 400 to 690 dpm/100 cm2 and the mean was 540 dpm/100 cm2 . COMPARISON OF RESULTS WITII GUIDELINES 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 contamination 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 of which Co-60 and Fe-55 are the primary contaminants at SNPS. The residual surface activity guidelines are: Total Activity 5,000 dpm S-7/100 cm2, averaged over 1 m2 15,000 dpm #-y/100 cm2 , maximum in 100 cm 2 Removable Activity 1,000 dpm #-y/100 cm 2 s%new.. nr . oas. 20, im 11 a.w pwu%w.wiwm.mi
1 As previously discussed, the detection sensitivities of the field instruments are such that the residual Fe-55 activity can not be detected. Therefore, total and removable surface activity measurements were corrected for Fe-55 when appropriate. The mean surface activity level for each suney unit was calculated and the survey unit data tested at the 95% confidence level [ upper confidence level (UCL)], relative to the guidelines, in accordance with Draft NUREG/CR-5849. These results are provided in Tables 4 and 5. There were no direct measurements which exceeded the average or maximum total activity guideline. Overall, surface activity levels within each survey unit also satisfied the guidelines at the 95 % confidence level. All removable activity was below guidelines at the 95 % confidence level. The maximum removable activity identified was 81 dpm/100 cm2, 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 6 of the 15 confirmatory survey units. The conditions established have therefore been satisfied for these survey units. However, the ESSAP mean was greater than the LIPA mean for survey units RW013, RWO17, RW023, RWO42, RWO72, SP004, SUO43, SU014X09, and SUO14X10; therefore, additional evaluation was necessary. Of the 9 units where the ESSAP total surface activity mean was higher than the LIPA mean, LIPA qualified the release record data for survey units RW013, RW017, RWO23, SP004, and SUO43 as containing excessive negative measurements. The LIPA stated cause was due to observed, lower backgrounds levels than the generic site backgrounds which 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 2 units, was 3,800 dpm/100 cm in SP004, and the maximum survey unit mean and UCL, found 2 in RWO23, was 300 dpm/100 cm and 400 dpm/100 cm2 , respectively. Although the condition of the ESSAP mean being less than the LIPA mean was not satisfied for these survey units, both termination and confirmatory survey surface activity levels are below the guidelines. Shortham Benhavte. NY - October 20, 1994 12 s:se ap\ reports \shoreham\brnothavm001
The ESSAP surface activity means for survey units RWO42, RWO72, SU014x09, and SU014x10 were statistically higher than those for LIPA and, LIPA had not qualified the data for these units as biased low. Therefore, the LIPA surface activity levels, mean activity levels, and UCLs for all remaining survey units were evaluated to determine the potentialimpact on the LIPA reported status of the Phase 2 and Phase 3 survey units, relative to the guidelines. The maximum LIPA survey unit mean of 2,317 dpm/100 cm 2 (SU%7) and maximum UCL of 2 812 dpm/100 cm (RWO61) were within acceptable criteria. The maximum observed difference 2 of ESSAP and LIPA means was 330 dpm/100 cm . If this difference in activity levels were applied to the above, overall surface activity levels would not be significantly altered and the conclusions reached, that the radiological status of the Phase 2 and 3 survey units satisfies the guidelines, would remain valid. The comparison of the LIPA and ESSAP embedded piping direct measurement data indicates two discrepancies. First, the LIPA mean activity lev.:Is for 2 of the 3 sections of pipes investigated do not agree for the original survey and the resurvey of SUO14X02 (#366) and SU014X09 (#850). Secondly, a comparison of the means indicates that the ESSAP mean is statistically greater than the LIPA mean for each of the three data sets. The probable cause of the difference in the raeans of the LIPA surveys may be attributed to variations in background levels. The difference in the average activity levels between the initial survey and resurvey of SUO14X02 (#366) was 22 cpm (3.7 cpm per detector) and for SU014X09 (#850) the difference was 9 cpm (1.5 cpm per detector). The initial survey results and resurvey results for SU016X01 (#20) were comparable. The observed difference in the surface activity level means between LIPA and ESSAP may be the result of the differences in detector geometry. Each LIPA direct 2 measurement location consists of a 93 cm area. LIPA's reported total activity is the additive activity contribution from 6 locations distributed around the interior circumference of each 0.3 m section of embedded piping where a measurement was performed, whereas ESSAP performed 2 one direct measurement, representing a 15.5 cm area, every 0.3 to 0.6 m. Overall, surface scans and direct measurement results do not indicate the presence of residual surface activity within the embedded piping, with most surface activity levels comparable to background and all levels below the minimum detectable activity (MDA) of the instrumentation.
%sdam Bruthavea, NY . Octater 20, im 13 6*-r w .h-+=w-th.-mi
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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 h (Table 4).' The Radwaste Building exposure rates were comparable to background exposure rate - levels and confirmed the findings presented by LIPA. l
SUMMARY
l ESSAP performed confirmatory activities for the Radwaste Building, Suppression Pool, and f Phase 2 systems at the Shoreham Nuclear Power Station in Brookhaven, New York. : Confirmatory activities included document reviews, and during the period August 22 through f 25, 1994, independent surface scans, surface activity measurements, exposure rate j measurements, and operational surveillance were performed. ! The survey results confirm the results of the LIPA termination surveys. These findings indicate f that total and removable surface activity levels and exposure rates were below the NRC . f s guidelines for release to unrestricted use. Statistical tests of data sets further support the i conclusion that each survey unit satisfies the guidelines at the 95% confidence level. I l l l i I mamaham. NY - OcWe 20, IW 14 WmpVeymuhhsham.001
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FIGURE 18: Reactor Building, HPCI Volves E41-01V-3049 and 3050 (SUO12) - Measurenent and Sanpling Locations l Shoreham DrnAhaven, NY - Ock&r 20, Im 32 w.-,,*%h Ser..ooi i
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METERS FIGURE 23: Reactor Building, Reactor Vater Clean up Systen Components (SUO15) - Measurement and Sanpling Locottons hwrham-Droukhavra. NY - Ockkr 20, im 37 h*me w pore..ix,th wnth.m m
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MEASUREMENT / SAMPLING LOCATIONS NOT TO SCALE e DIRECT MEASUREMENT i FIGURE 24: Radwaste Building, Condensate Denineralizers, Tank 1N52-DE -002E j (SUO43) - Measurenent and Sanpling Locations sama.m-huma, NY - Ru,er 20, im 38 h*.=,w.www.u mun
TABLE 1
SUMMARY
OF SURFACE ACTIVITY LEVELS RADWASTE BUILDING, SUPPRESSION POOL, AND PIIASE 2 SYSTEMS SIIOREIIAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Removable Number of Total Activity Range Activity Range < Location
- Measurement (dpm/100 cm') (dpm/100 cm 2)
Locations Beta Alpha' Betad RW013,15' North Itallway Floor 11 10 to 540 -1 to 1 -5 to 1 Lower Walls 9 -260 to 260 -1 to 1 -4 to 4 Upper Walls 5 -210 to 100 -1 to 3 -5 to 3 Equipment 5 -260 to 450 -1 to 1 -3 to 8 RW017,15' CAT /AN/REGN Tank Floor 10 -200 to 360 -1 to 3 -3 to 9 Lower Walls 10 -100 to 190 -1 to 1 -4 to 3 Upper Walls 7 -100 to 38 -1 to 1 -5 to 0 Equipment 3 -200 to -14 -1 to 3 -4 to 3 RWO23,-19' Cubicle A, Liner Stations Floor 9 370 to 1100 -1 to 1 -4 to 3 Walls 11 -250 to 490 -1 to 3 -5 to 5 Upper ' Walls 3 -170 to 230 1 0 to 8 Equipment 7 170 to 200 -1 to 1 -5 to 3 , RWO40,115' Evaporation Distil Room ; Floor 13 220 to 440 -1 to 3 -5 to 9 i Walls 7 87 to 330 -1 to 1 -4 to 3 { Upper Walls 4 -26 to 180 -1 to 3 -1 to 5 Equipment 6 -43 to 660 -1 to 3 -5 to 1
-RWO42,~ 37' North Hallway ^
Floor 14 450 to 920 -1 to 1 -4 to 4 1 Lower Walls 6 230 to 460 -1 to 5 i
-4 to 3 i Upper Walls 3 190 to 280 -1 to 5 0 to 3 Equipment 7 -150 to 520 -1 to 1 -5 to 3 Skwth Bm& haven, NY - October 20, im 39 w=rwww-u ve..wi
TABLE 1 (Continuet
SUMMARY
OF SURFACE ACTI' Y LEVELS RADWASTE BUILDING, SUPPRESSION POO \ND PHASE 2 SYSTEMS SIIOREIIAM NUCLEAR POWL ;TATION BROOKIIAVEN, NEW Y0 '( Removable Number of Total Activity Range Activity Range Location" Measurement (dpm/100 cm 2) (dpm/100 cm 2) Locations Betah Alpha' Betad RWO69,~ IIEPA After Filter Area: Floor 11 110 to 490 -1 to 4 -3 to 6 Lower Walls 9 -62 to 130 -1 to 3 -4 to 10 i Upper Walls 5 -48 to 95 -1 to 1 -4 to 3 Equipment 5 -210 to 140 -1 to 1 -1 to 3 RWO72, Gas Desiccant Dryer Area i Floor 11 410 to 920 -1 to 3 -4 to 8 Lower Walls 9 250 to 550 -1 to 3 -5 to 5 Upper Walls 3 270 to 380 -1 t 3 -5 to 1 Equipment 7 81 to 580 -1 to 3 -5 to 3 SP004. Suppression Pool NW.-Quad.' - 31 -310 to 3800 -1 to 6 -5 to 81 SP005, Suppassion Pool In. Pedestal 35 -210 to 350 -1 to 3 -7 to 8 SUO12?III Pressum Coolant hdection ~ 4 -440 to -200 -1 to 1 -5 to 8 ! SUO14 x 09/ Tank'54, Influent Dralmii 31 -360 to 870 -1 to 1 -3 to 6 SUO14'x 10, Radwaste Equipment Drains' 11 -320 to 530 -1 to 1 -4 to 6 SUO14 x 12f Radwaste Equipment / Comp. - 50 -760 to 1600 -1 to 5 -6 to 14 SUO15, Reactor Water Cleanup Systemt 30 -520 to 1100 -1to5 -4 to 5 SUO43, Condensate Demineralizer 12 -270 to 360 -1 to 8 -5 to 3
' Refer to Figures 7 through 24.
b MDAs = 250 to 1100 dpm/100 cm 2,
*MDA = 12 dpm/100 cm 2, )
d i MDA = 16 dpm/100 cm2. I Shoreham BmAbsm, NY Odda 20, im I 40 n. rwim=w w w=u.m m : I 1
1 TABLE 2 INTERIOR EXPOSURE RATES RADWASTE BUILDING SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK g g,,, Number of Measurement Exposure Rates ' Locations at 1 m (gR/h) RW013 11 4 to 5 RW017 6 4 to 5 RWO23 8 4 to 5 RWO40 6 4 to 5 RWO42 13 5 to 6 RWM9 6 5 to 6 RWO72 11 5
' Refer to Figures 7-14.
l i wwwv., m . om 20, im 41 w -pwp-uw=a.ar=*h.va m
l l TABLE 3 EMBEDDED PIPING PROGRAM
SUMMARY
SIIOREIIAM NUCLEAR POWER STATION , BROOKIIAVEN, NEW YORK Calibratilon Pipe Efficiency 4" Pipe LIPA 6 detector Pipe Crawler Assembly 0.134' ESSAP HP-260 0.1466 8" Pipe LIPA 6 detector Pipe Crawler Assembly 0.152' ESSAP HP-260 0.163b Description LIPA Expected / Actual
- ESSAP Confinnation 4" Pipe Crawler Calibration Operational Check 44/48 (dpm/cm2) 44 (dpm/cm2)
Source Activity (dpm/100 cm2) Action Level (for a 15,000 dpm/15 cm2 hot spot) 2408 cpm /830 cpmd 2410 cpm
. Survey Design Required Number of Measurements Confirmed Number of 8""*Y "I (Per SP No. 66x020.11) Measurements SU0l6x01 (#20) 14 15 SUO14 x02 (#366) 28 30 SUO14 x09 (#850) 26 28 Data ConversionY LIPA ESSAP Confirmation Location Surface Activity Critical Level Surface Activity Critical Level (dpm/100 cm 2) (dpm/100 cm 2) (dpm/100 cm2) (dpm/100 cm 2) 20-1-3 -290 270 -290 270 20-5-2 -298 270 -298 270 366-2-1 -382 175 -382 175 366-5-2 -271 181 -271 181 366-9-3 -338 178 -338 178
%amusm. NY - me 20, tW 42 Wasap%hhamhdam 001
.'5 ,_ g TABLE 3 (Continued)
EMilEDDED PIPING PROGRAM
SUMMARY
SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK l Data Conversion' (Continued)- ' LIPA ESSAP Confirmation Location Surface Activity Critical level Surface Activity Critical 12 vel ) (dpm/100 cm 2) (dpm/100 cm ) 2 (dpm/100 cm 2) (dpm/100 cm2) 850-3-2 -508 257 -508 257 850-5-1 -556 254 -556 254 850-6-3 -411 263 -411 263 850-8-1 -556 254 -556 254 - Survey Results Survey Unit Data Origination SU0l6x01 (#20) SU014 x02 (#366) SUO14x09 (850) LIPA Termination Total Activity Range -468 to 169 -401 to -111 -379 to 16 Survey (dpm/100 cm2)
-245 -240 -196 Mean (dpm/100 Activity) cm LIPA Resurvey Total Activity Range -395 to -121 -469 to -271 -581 to -315 j (dpm/100 cm2) ;
I
-245 -337 -488 Mean (dpm/100 Activity) cm ESSAP Survey Total Activity Range 400 to 690 -320 to 120 -360 to 40 (dpm/100 cm2) 540 -20 -140 Mean (dpm/100 Activity) cm i
*Per Table 4.3 Shoreham Decommissioning Project Termination Survey Plan, December 1993. .
b Determined using an LIPA NIST traceable flexible mylar Co-60 source placed inside of a 4" and 8" piping spool piece.
' Determined from 8/19/94 LIPA Multiple G-M Detector Assesly Calibration Check Data Sheet l (SPF66x020.11) for detector assembly IZ12-100PC4-0001. !
d830 cpm is the LIPA alarm setting for the 6 detector assembly /ratemeter-scaler combination, where additional investigation of possible hot spots is required.
' Data points analyzed were representative of information collected during the resurvey of embedded piping subunits performed by LIPA. Data was tracked for accuracy from the field records to the final computer generated report.
%weham Braahaven, NY - OcM*r20,1994 43 hMasap\reportsi nhorehamibrathaven .001 l
TABLE 4 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES, RADWASTE BUILDING AND SUPPRESSION POOL SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK Radiological Survey Unit *
""** U RW013 RW017 RWO23 RWO40 RWO42 Total Beta' Activity (dpm/100 cm)2~-
# of Direct Measurements 30 30 30 30 30 Mean(X) 82 14 300 240 430 LIPA X ,
-170 -130 13 110 280 Fa 140 60 400 290 510 Conditions and 5,000/15,0000 No/Yes No/Yes No/Yes Yes/Yes No/Yes dpm/100 cm2 Guidelines Satisfied Rembvable Beta Activity (dpm/1002 cm)b '
# of Smears 30 30 30 30 30 Mean(X) -0.5 -0.1 -0.4 0.2 -1,3 LIPA X 7.5 2.6 3.9 3.2 0.3 Fa 0.4 0.9 0.6 1.2 -0.5 1,000 dpm/100 cm2 Yes Yes Yes Yes Yes Guideline Satisfied Exposure Rates at 1 m'(pR/h) ' #
# of Exposure Rate Measurements 11 6 8 6 13 Mean (X) 4.2 4.1 4.6 4.5 5.1 LIPA X 5.8 6.5 6.4 6.0 6.2 Fa 4.4 4.3 4.9 4.7 5.2 Conditions and 5 R/h Above Yes Yes Yes Yes Yes Background Guideline Satisfied Shorrham Brmahavea, NY - Ockhnr 20, 1994 44 h:ksiap\ reports \ahortham%molhaven.001
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES, RADWASTE BUILDING AND SUPPRESSION POOL SIIOREHAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK Radiological Survey Unit' SN RWO69 RWO72 SP004 SP005 < Total Beta Activity (dpm/100 cn/) ' l
# of Direct Measurements 30 30 31 35 Mean(I) 93 510 110 100 t LIPA I 360 280 -220 76 ;
Fa 140 580 350 150 _. Conditions and 5,000/15,0000 Yes/Yes No/Yes No/Yes Yes/Yes dpm/100 cm2Guidelines Satisfied Removable BAta 'Actid[(dpm/100 cuf)b i ,
# of Smears 30 30 31 35
, Mean(I) 0.3 0.1 3.4 -0.2 LIPA I 3.4 2.9 9.0 5.0 Fa f.2 1.2 8.0 0.6 2
1,000 dpm/100 cm Yes Yes Yes Yes Guideline Satisfied , Exposure Rates at I 'm (gR/h) -
# of Exposure Rate Measurements 6 11 -c _
i Mean (I) 5.3 5.0 - - LIPA I , 6.1 6.0 - - Fa 5.4 5.2 - - 5 pR/h Above Background Yes Yes - - Guideline Satisfied
" Refer to Figures 7 through 17.
- b All alpha removable activity was less than 12 dpm/100 cm2 ,
- == Measurements not performed.
w r oa.., ny - me 20, im 45 u e-ea s m w- a = wi
TABLE 5 CONFIRMATORY RADIOIDGICAL STATUS
SUMMARY
-SYSTEMS SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK Radiological Survey Unit
- Summary SUO12 SUO14 x 09 SUO14 x 09 SUO14 x 10 SUO15 SUO43 Total Beta Activity (dpm/100 cnh
# of Direct Measurements 4 I 31 11 50 30 12 Mean(X) -310 -42 49 98 -21 90 LIPA X 150 -170 -170 11 87 4.6 Fa
-190 50 210 200 74 200 Conditions and Yes/Yes No/Yes No/Yes Yes/Yes Yes/Yes No/Yes 5,000/15,000(, .pm/100 cm2 Guidelines Satisfied Removable Beta Activity'(dpm/100 cm')h'
# of Smears 4 11 7 50 25 12 Mean(X) 0 0.9 -1.3 1.7 0.2 -0.3 IJPA X 7.2 13.9 6.6 13.1 5.8 16.7 Fa 6.9 2.8 1.3 2.6 1.0 1.0 2
1,000 dpm/100 cm Yes Yes Yes Yes Yes Yes Guideline Satisfied
' Refer to Figures 18 through 24.
b ! All alpha removable activity was less than 12 dptrd100 cm2, l i Sharrham Brmtbevea. NY - Ockda 20, IW 46 hhwW6mehadnxthavmM1
d * . REFERFACES
- 1. Iong Island Lighting Company, "Shoreham Nuclev Power Station Site Characterization Program Final Report," May 1990. ,
- 2. T. J. Vitkus, ORISE, " Confirmatory Survey of the Tmbine Intemal Components, Shoreham Nuclear Power Station, Brookhaven, New York," July 1993.
- 3. T. J. Vitkus, ORISE, " Confirmatory Survey of the Turbine Building, Site Grounds, and Site Exteriors, Shoreham Nuclear Power Station, Brookhaven, New York," September 1994.
- 4. J. D. Berger, Oak Ridge Associated Universities, Draft " Manual for Conducting Radiological Surveys in Support of License Termination," NUREG/CR-5849, June 1992.
5. Iong Island Power Authority, "Shoreham Decommissioning Project, Termination Survey Plan, Revision 1," April,1993. 6. Iong Island Power Authority, "Shoreham Decommissioning Project Termination Survey Final Report, Volumes 1 through 5," September,1993, i
- 7. Ixtta from T. J. Vitkus, ORISE to D. Fauver, U.S. Nuclear Regulatory Commission, " Final Confirmatory Survey Plan for the Shoreham Nuclear Power Station, Brookhaven, New York -
Docket File No. 50-322," November 4,1993. 8. Letter from D. N. Fauver, U.S. Nuclear Regulatory Commission, to T. Vitkus, ORISE, July 1,1993.
- 9. Irtter fn>m M. R. Landis, ORISE to D. Fauver, U.S. Nuclear Regulatory Commission, "Shoreham Decommissioning Project, Termination Survey Plan, Revision O, Shoreham Nuclear !
Power Station, October 1992," January 12, 1993.
- 10. U.S. Nuclear Regulatory Commission, " Guidance and Discussion of Requirements for an Application to Terminate a Non-Power Reactor Facility Operating License," Revision 1, '
September 1984 l w n.*w= en .cano ro, im 47 mw rw=>mwn *m.= m
e e APPFEDIX A MAJOR INSTRUMFNTATION Mhwa, NY - Ockter 20, im h:ksaap\nportsiabarchamitiemmg
APPENDIX A MAJOR INSTRUMENTATION The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the authors or their employers. DIRECT RADIATION MEASUREMENT Instnunents 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) Ik4ectors Eberline GM Detector Model HP-260 Effective Area,15.5 cm2 (Eberline, Santa Fe, NM) Ludlum Gas Proportional Detector Model 43-37 Effective Area,550 cm2 (Ludlum Measurements, Inc., Sweetwater, TX) Ludlum Gas Proportional Detector Model 43-68 Effective Area,100 cm2 (Ludlum Measurements, Inc., Sweetwater, TX) un*= we. . ny - ona= 20. nw A-1 h*. p % m w e = wi
r. j ,. , . . i Reuter-Stokes Pressuriztxi Ion Chamber ; Model RSS-111 (Reuter-Stokes, Clev: land, OH) Victoreen NaI Scintillaticn Detector Model 489-55 3.2 cm x 3.8 cm Crystal ! (Victorcen, Cleveland, OH) ' LABORATORY ANALYTICAL INSTRUMENTATION - Low Background Gas Proportional Counter , Model LB-5100-W (Oxford, Oak Ridge, TN) l t i I P t 1
\
4 t J s Shardume-Brockhoven, NY - October 20,1994 A-2 h*. c s, econ.s.horehamsbrooth vc=.001 : I l
s l I i APPENDIX B SURVEY AND ANALYTICAL PROCEDURES r l i hm Bmhvea, m . ock*er 20,1W4 b:Vasap\reposts\shoreham\bnxthaven.001 ;
.-.e., ,iJ.-
APPENDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES Surface Scatu Surface scans were performed oy passing the probes slowly over the surface; the distance between - the probe and the surface was maintained at a minimum - nominally about I cm. A large surface area, gas proportional floor monitor was used to scan the floors of the surveyed areas. Other surfaces were scanned using small area (15.5 cm2 , 59 cm2 or 100 cm ) 2hand-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 ' f Beta - gas proportional detector with ratemeter-scaler pancake GM detector with ratemeter-scaler Gamma - NaI scintillation detector with ratemeter 9 Sprface Activity Measurrments Measurements of total beta activity levels viere performed using GM and gas proportional detectors with portable ratemeter-scalers. Count rates (cpm), which were integrated over 1 minute in a static position, were converted to t 2 activity levels (dpm/100 cm ) by dividing the net rate by the 4 r efficiency and correcting for the j active area cf the detector. 'Ihc beta activity background count rates for the GM and gas proportional l detectors ranged from 22 to 34 cpm and from 129 to 174 cpm, respectively. Beta efficiency factors m wah.m. NY 4 O*en 20, W B-1 h hasap % h ham u m u n
l l ranged from 0.16 to 0.17 for the GM detectors and from 0.21 to 0.23 for the gas proportional detectors. The effective window for the GM and the gas proportional detectors were 15.5 cm2 and j 100 cm2, respectively. Surface activity me.surements which exceeded the normal background distribution were corrected ' for the Fe-55 contribution by multiplying the dpm/100 caf field activity level by a factor of 1.2. The instmment response level at which the detector output could be considered above background was ; defined at the critical level (IJ. This level was defined for each detector /instmment combination as follows:
;y Sample count rate , Background count rate h Sumpl e count time Backgrcund count time L* ,
(Detector Fhiciency) (Detector Geometry) . Removable Activitv Measurements Removable activity levels were determined using numbered filter paper disks, 47 mm in diameter. . 2 Moderate pressure was applied to the smear and approximately 100 cm of the surface was wiped. Smears were placed in labeled envelopes with the location and other pertinent information recorded. ! Exposurt Rate Measumnents ' Measurements of gamma exposure rates were performed using a pressurized ionization chamber I (PIC). ANALYTICAL PROCEDURFE i Removable Activhv l Smears were counted on a low background gas proportional system for gross alpha, and gross beta l . activity.
% w m = ,nr . m w . 20 i m B-2 a vem%mmmi
. - . . . . . . - . . . ~ - _ - . . _.
UNCERTAINTIES AND DE7ECTION LIMITS
'Ihe uncertainties associated with the analytical data presented in the tables of this report represent the 95 % confidence level for that data. These uncertainties were calculated based on both the gross +
sample count levels and the associated background count levels. Additional uncertainties, associated with sampling and measurement procedures, have not been propagated into the data presented in this I report. Detection limits, referred to as minimum detectable activity (MDA), were based on 2.71 plus 4.66 times the standard deviation of the background count [2.71 + (4.66/BKG)]. When the activity was i determined to be less than the MDA of the measurement procedure, the result was reponed as less than MDA. Because of variations in background levels, measurement efficiencies, and contribudons from other radionuclide in samples, the detection limits differ from the sample to sample and I instrument to instrument. CALIBRATION AND QUALITY ASSURANCE Calibration of all field and laboratory instrumentation was based on standards, traceable to NIST, when such standard were available. In cases where they were not available, standards of an industry n: cognized organization were used. Calibration of pressurized ionization chambers was performed by the manufacturer, t Analytical and field survey activities were coinducted in accordance with procedures from the following documents of the Environmental Survey and Site Assessment Program: Survey Procedures Manual, Revision 8 (December 1993) Laboratory Procedures Manual, Revision 8 (August 1993) , I Quality Assurance Manual, Revision 6.1 (November 1993) l l w mm. wy . oc*, n. im B-3 a-rw,m.mww= m j l
-- ~-
l
i
^
The procedures contained in these nanuals were developed to meet the requirements of DOE Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to nueu processes during I
.t dieir performance.
Quality control procedures include: e Daily instrument background and check-source measurements to confirm that ! equipment operation is within acceptable statistical fluctuations. ; Participation in EPA and DOE /EML Quality Assurance Programs. Training and certification of all individuals performing procedures. f Periodic internal and extemal audits. i s h h t
?
i i f 4 WMham NY Mu 20. Im B-4 6% ,w.h.h m..ool l
. . . . . . - - . - . . - - . . , .- - . - - -- .-, , - . - , -n .
APPENDIX C REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS-hBnmahaven, NY % 20,1994 h:kasaP%\ahorcham\brookhavenMI l
D.S. ATOMIC ENERGY COMMISS!ON Jun31974
' 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 liceme is issued, Section 50.51,
- Duration of license, sv:newal," of 10 reactor operation is not permitted. Other activities from CFR Part 50,
- licensing of Production and Utilization the reactor and placing it in storage (either onsite or Facilities," nquires that each license to operate a offsite) may be continued.
producuon and utilization facility be issued for a specified duration. Upon expiration of the specified period, the A licemee having a possession-only liceme must retain, license may be either renewcxi or termmated by the with the Part 50 license, authorization for gecial nuclear Commission. Section 50.82, " Applications for ternunation matenal (10 CFR Part, 70, "Special Nuclear Material"), of licmses," specifies the requirements that must be bypmduct material (10 CFR Part 30, " Rules of General satisfied to terminste an operatmg liceme, including the Applicability to Licensing of Byproduct Material"), and requirement that the dismantlement of the facility and source material (10 CFR Part 40, "Licen-ing of Source disposal of the component parts not be inimical to the Material"), until the fuel, radioactive components, and common defmse and security or to the health and safety sources are removed from the facility. Appropriate of the public. This guide describes methods and administrative controls and facility requirements are procedures considered acceptable by the Regulatory staff imposed by the Part 50 liotcr and the technical for the termination c f operating licenses for nuclear specifications to assure that proper surveillance is reactors. 'Ihe advisory Committee on Reactor Safeguards performed and that the reretor facility is maintained in a has been consultext concerning this guide and has safe condition and not operated. concurred in the regulatory position. A possessionenly license permits various options and B. DISCUSSION procedures for decommissioning, such as mothballing, eatombment, or dismantling. The requirements imposed When a licensee decides to terminate his nuclear depend on the option selected. reactor operatmg license, he may, as a first step in the process, requent that his operating license be amended to Section 50.82 provides that the licensee may dismantle restrict him to powcss but not opeaste the facility. The and dispose of the componmt parts of a nuclear reactor in advantage to the licmsee of converting to such a accordance with existing regulations. For research possession-only license is reduced surveillance reactors and critical facilities, this has usually meant the requirements in that peakxlic surveillance of equipment disassembly of a reactor and its shipment organization for USAEC REGULATORY GUIDES c ,,,, ,, ,,,,,,, , ,,, ,,, , ,,,,,, ,,,,,,, ,,,,,,,,,,, mes. om is.u .. c.a en. mau w.w. h. pian. "# ~ M. M. We, Cm WW,.n, D.C. MM. m.~. uc , .,. ., n mr. ^=~ o -'- ~ ~v =-a-a c- - --~ '-
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- f fwther use. 'Ibe site from which a reactor has been c. ' Any proposed changes to the technical specific .tions I 3
- removed must be ha**% as necessary, and that reflect the possession-only facility status and the mapected by the Conmussion to d.t .J.w whether necessary A 'Ny/ retirement activities to be unrestncted === can be approved. In the case of performed nuclear poww reactors, diamentling has usually bem '
accomplished by duppmg fuel offsite, melang the reactor d. A safety analysis of both the ' activities to be ' inoperable, and dupoemg of some of the radioactive accomplished and the proposed changes to the technical components. specifications I i i
- e. A inventory of activated matenals and thdr 5 Radioactive componets may be either shipped off-site location in the facility. !
for burial at an authonaed burial ground or secured on the site. 'Ihose radioactive matenals remanung on the site 2. ALTERNATIVES FOR REACTORRETIREMENT must be isolated from the public by physical barners or other means to prevent pubhc access to hazardous levels Four alternatives for retiremmt of nuclear reactor of rad.ation. Surveillance is neccesary to assure the long facilities are considered acceptable by the Regulatory : term integrity of the barriers. The amount of surveillance staff. 'Ibese are: required depends upon (1) the potential hazard to the health and safety of the public from radioactive matenal a. Mothhalling. Mothballing of a nuclear reactor remaining on the site and (2) the integrity of the physical facility consists of putting the facility in a state of ! harriers. Before areas may be seleased for unrestricted protective storage. In general, the facility may be left une, they must have been decontaminated or the intact except that all fuel assembbes and the radioactive radioactivity must have decayed to less than prescribed fluids and waste should be nrnoved from the site. limits (Table 1). Adequate radiation monitoring, mvuonmental i surveillance, and appropnate security procedures 1 h hazard associated with the returned facility is should be estabhshed under a possessionely license i evaluated by conndenng the amount and type ofremamag to meure that the health and safety of the public is act contamination, the degree of confinement of the remantig endangered. radaoactive matennis, the laysical security provided by the confinemmt, the susceptibility to relcane of radiation as a b. In-Place Entomhment. In-place entombment result of natural phenomma, and the duration of requued consists of sealing all the remmuung highly raaioactive ! surveillance. or contanunated vis (e.g., the pressure vessel and reactor intemals) within a structure integral with f C. REGULATORY POSITION the biological shield after having all fuel assemblies, radioactive fluids and wastes, and certam selected
- 1. APPUCATION FOR A UCENSE TO POSSESS w- =ts shipped offsite. b structure should BUT NOT OPERATE (POSSESSION-ONLY provide meegrity over the penod of time in which UCENSE) significant quantities (greater than Table 1 levels) of radioactivity remam with the matenal in the A request to amend an operatmg licase to a entombment. An .py.ogirde and continuing possessionely beense should be made to the Duector of survedlance program should be estabhshed under a Ucensing, U.S. Atomic Energy Commission, Washmgton, possennon cely bcase.
D.C. 20545. b request should include the followmg information* c. Removal of Baaaactive. Components and Dismantling. - All fuel assemblies, radioactive fluids
- a. A description of the current status of the facility. and waste, and other matenals having activities above accepted unresencted activity levels (Table 1) should be l
- b. A description of measures that will be taken to removed from the site. The facility owner may then prevent criticality or reactivity changes an i to nummiae have unresticted use of the site with no requuement releases of radioactivity from the facility, for a license. If the facility (mnet *o desues, the l l
Note: Section electronicaHy reproduced froen pnotocopy. C-2
remamder of the reactor facility may be dismantled and barriers in the facility. Sampling should be done along the 3
- all vestiges reuxwed and disposed rif. most probable path by which radioactive material such as that stored in the inner contammmt regions could be
- d. Conversion to a New Nuclear Systan or a Fossil transported to the outer regions of the facility and l Fud Systan. This alternative, which applies only to ultinately to the envuons.
nuclear power plants, utilizes the existing turbme system with a new steam supply system. The original d. An environmental radiation survey should be nuclear steam supply system should be separated from performed at least semiannually to verify that no the electric generatmg system and disposed of in significant amounts of radiation have been released to the accordance with one of the previous three retirement envuonment from thc. facility. Samples such as soil, attematives. vegetation, and water should be taken at locations for which statistical data has been established during reactor
- 3. SURVEILLANCE AND SECURITY FOR TIIE operations.
RLTIREMENT ALTERNATIVES WIIOSE FINAL STATUS REQUIRES A POSSESSION-ONLY e. A site representative should be designated to be I LICENSE responsible for controlling authonzed access into and movement within the facility. A facility which has been licensed under a ! possession-only liceme may contain a significant amount f. Administrative procedures should be established for of radioactivity in the form of activated and cont = mind ~1 the notification and reporting of abnormal occunences hardware and structural materials. Surveillance and such as (1) the entrance of an unauthorized person or commensurate security should be provided to assure that persons into the facility and (2) a significant change in the the public health and safety are not endangered. radiation or contamination levels in the facility or the
- a. Physical security to prevent inadvertent exposure of offsite environment.
personnel should be providal by multiple locked barriers. W presence of these barriers should make it extremely g. The following reports shoulo b made: j difficult for an unauthorizal person to gain access to areas I where radiation or contamiaation levels exceed those (1) An annual report to the Director of Licensing, l specified in Regulatory Position C.4. To prevent U.S. Atomic Energy Commission, Washington, D.C. I inadvertent exposure, radiation areas above 5 mR/hr, such 20545, describing the results of the mvironmental and as near the activated primary system of a power plant, facility radiation surveys, the status of the facility, and an should be appropriately marked and should not be evaluation of the performance of security and surveillance I accessible except by cutting of welded closures ot the measures. disassembly and removal of substantial structures and/or shielding material. Means such as a remoto-readout (2) An abnormd necurrence report to the Regulatory l intmsion alarm system should be provided to indicate to Operations Regional Office by telephone within 24 hours designated personnel when a puysical barrier is penetrated. of discov ry of an abnormal occurrence. 'Ihe abnormal 1 Security personnel that provide access control to the occurrence will also be reported in the annual report facility may be u=xl instead of the physical barriers and described in the precedmg item. the intrusion alarm systems.
- h. Records or logs relative to the following items
- b. The physical barriers to unauthorized entrance into should be kept and retained until the license is termmated, the facility, e.g., fences, buildings, welded doors, and after which they must be stored with other plant records access opemngs, should be inspected at least quarterly to suave that these barrierr have not deteriorated and that (1) Envuonmental surveys, locks and lockirig apparstus are intact.
(2) Facility radiation surveys,
- c. A facility radiation survey should be performed at (3) 1nspections of the physical barners, and least quarterly to verify that no radica:tive matenal is escaping or being transpxtal through :he contamment (4) Abnormal oxurrences Note: Station a4ctrorneally reproduced from photocopy. C-3
m
- 4. DECONTAMINATION FOR RELEASE FOR (2) A detailed heahh arxl safety analysis indicating tha
- i.
- UNRESTRICTED USE the residual amounts of matenals on surface areas, together with other considerations such as the prospective If it is desired to terminate a license and to eliminate use of the premuu equipment, or scrap, are unlikely to any further surveillance requirerrents, the facility should rault in an unreasonable risk to the health and safety of be sufficimtly demtaminated to prevent risk to the public the public.
health and safety. After the decontamination is satisfactorily accomplished and the site inspected by the e. Prior to release of the premises for unrestrictal use, Commission, the Commission may authorize the license to the licensee should make a comprehensive radiation survey be terminated and the ficility abandoned or released for establishing that contammation is within the limits specified unrestricted use. h licensee should perform the in Table 1. A survey report should be filed with the daymtamination using the following gu.1 "nes: Director of Licensing, U.S. Atomic Energy Commission, Washington, D.C. 20545, with a copy to the Director of
- a. The licmsee should make a reasonable effort to the Regulatory Operations regional Office having eliminate residual contamination. jurisdiction. The report should be fded at least 30 days prior to the planned date of abandonment. h survey
- b. No covering should be applied to radioactive report should:
surfaces of equipment of structures by paint, plating, or other covering material until it is known that contamination (1) Idmtify the premises; levels (determined by a survey and documented) are below the limits specified in Table 1. In addition, a reasonable (2) Show that reasonable effort has been made to effort should be made (and docummtal) to further reduce residual contamination to as low as practicable minimize cmtanunation prior to any such covering, levels;
- c. The radioactivity of the interior surfaces of pipes, (3) Describe the scope of the survey and the general drain lines, or ductwork should be determined by making pmcedures followed; and measurements at all traps and other appropriate access points, provided contamination at these kxations is likely (4) State the finding of the survey in units specified in to be representative of 2 contamination on the interior of Table 1.
the pipm, drain lines, or ductwork. Surfaces of premises, equipnxnt, or scrap which are likely to be contaminst~1 After review of the seport, the Commission may but are of such size, construction, or location as to make inspect the facilities to confirm the survey prior to granting the surface inaccessible for purposes of measurement approvcJ for abandonment. should be assumed to be contaminated in excess of the pennissible radiation limits.
- 5. REACTOR RETIREMENT PROCEDURES
- d. !
Upon request, the Commission may authorize a As indicated in Regulatory Position C.2, several licmsoe to relinquish possession or control of premises, altematives are acceptable for reactor facility retirement. ajuipment, or scrap having surfaces contamin*yI in if minor disassembly or "mothballing" is planned, this excess of the limits specified. This may include, but is not could *oe done by the existing operating and maintmance l liraited to, special circumstances such as the transfer of procedures under the license in effect. Any planned premises to another licased organization that will continue actions involving an unreviewed safety question or a to work with radioacFve materials. Requests for such change in the technical specifications should be reviewed authorization should provide: and approved in accordance with the requirerrats of 10 CFR i 50.59. (1) Detailed, specific information describing the premises, equipmmt, scrap, and radioactive contaminants If major structural changes to radioactive components and the nature, extmt, and degne of residual surface f of the facility are planned, such as removal of the pressure contamination, vessel or major components of the primary system, a dismantlemmt plan including the information required by Note: Section electron;ce#y reoroduced from photocopy. C-4
58
- I 50.82 should be submitted to the Comnumion. A
- - ,o, _
alternatives of Regulatory Position C.2 except nothballing. IIowever, minor diennaembly activities may still be !
. performed in the abamce of such a plan, provided they are permitted by existing operating and maintenance procedures. A dismantlemmt plan simuld include the following:
- a. A descriptkm of the ultimate status of the facility F
- b. A description of the dismantling activities and the precautions to be taken.
- c. A safety ardysis of the dismantling activities including any efflumts 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 imued by the Commision in accordance with ! 50.82. What dismantling is completed and the Comrmasion 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 Commision may termmate the license. If pomension<mly bcenne under which the dismantling activities have becs o=Imai or, as an attemative, may { make application to the State (if an Agreemmt State) for a bypmduct matenals license. i i Note: Section electronically reproduced from photocopy. C-5 i 5
,, . 3 ,
TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS Nuclide* Average
- Maximum
- Removable
- U-nat, U-235, U-238, and associated dway products 5,000 dpm a/100 cnf 15,000 dp.n a/100 cnf 1,000 dpm a/100 caf Transuranics, Ra-226, Ra-228, Th-230, n-228, Pa-231, Ac-227,1-125, I-129 100 dpm/100 cnf 300 dpm/100 cnf 20 dpm/100 cnf Th nat, Th-232, Sr-90, Ra-223, Ra-224, U-232, I-126, I-131. I-133 1,000 dpm/100 cnf 3,000 dpm/100 caf 200 dpm/100 cnf Beta-gamma emitters (nuclides with decay nudes odier than alpha emission or spontaneous fission) except Sr-90 and others noted shove. 5,000 dpm Sy/100 cnf /
15,000 dpm 07100 cuf / 1,000 dpm 07100 enf
*Where surface contammation by both alpha- and beta-gamma <mitting nuclides exists, the limits established for alpha- and b gamnu-emitting nuclida should apply indgendently.
'As used in this table, dpm (disintegrations per minute) means the rate of emisr nn by radioacCve material as deter correcting the counts per minute obe:ved by an appropriate detector for backgmund, efficiency, and geometric factors associated with the instrunustation.
" Measurements of average contaminant should not be averaged over more than I square meter. For objects ofless surface are the average dould be derived for each such object.
"Ihe maximum contamination level applies to an area of not more than 100 enf.
2
'Ihe anumnt of renovable radioactive mater:al per 100 cm of surface area should be determined by wiping that an:a wit or soft aborbmt paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropnate instrummt of known efficiency. When removable contamination on objects ofless surface area is determined, the pertmmt levels should be reduced pmportionally and the entire surface dxmld be wiped.
l l l 1 1 Note: Section electronically reproduced frorn photocopy. C-6 4 1 _.}}