ML20081A668
| ML20081A668 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 02/17/1995 |
| From: | Laudeman M, Payne A, Vitkus T OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20081A654 | List: |
| References | |
| CON-FIN-A-9076 ORISE-95-B-80, NUDOCS 9503150243 | |
| Download: ML20081A668 (74) | |
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I I I I I I I I The Oak Ridge Institute for Science and Education (ORISE) was established by the U.S. Depanment of Energy to undenake 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. Depanment 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. This repon 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 owned rights. Reference herein to any specific commercial product, process, or service by trade name, mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement or recommendation, or favor by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. I-I
ORISE 95/B-80 CONFIRMATORY SURVEY l OF THE RADWASTE BUILDING, SUPPRESSION POOL, PHASE 2, AND PHASE 3 SYSTEMS SHOREHAM NUCLEAR POWER STATION .
. BROOKHAVEN, NEW YORK 7
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 ! l U.S. Nuclear Regulatory Commission Headquarters Office i Sponsored by the Division of Waste Management ' FINAL REPORT l FEBRUARY 1995 l 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. Oak Ridge Institute for Science and Education performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy. Shortham Brookhavam, NY . Febnery 16,1995
._ ~ ._. . , .. - _
l 1 i l CONFIRMATORY SURVEY I OF THE RADWASTE BUILDING, SUPPRESSION POOL, AND PHASE 2 SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK l ! Prepared by: Date: '
/S~[W '
T. J. Vitkus,$rpject Leader Environmental Survey and Site Assessment Program Reviewed by: OU kM
- Date: xlP7) W M. J. Laudeman, Radiochemistry Laboratory Supervisor l Environmental Survey and Site Assessment Program i
Reviewed by: h twe. Date: Q //(o / 97' A. T. Payne, Quality Assuradce Officer l Environmental Survey and Site Assessment Program l l < l l Reviewed by: _ s Date: 4//7d5 W. L Beck, Acting Program Director ' Environmental Survey and Site Assessment Program i i l SWBroukhaven, NY - Fet ruary 16,1995
L l l l ACKNOWLEDGEMENTS The author would like to acknowledge the significant contributions of the following staff members: FIFJD STAFF ; E. H. Bjelland E. H. Montalvo . J. R. Morton J. L. Payne i I LABORATORY STAFF R. D. Condra J. S. Cox M. J. I2udeman CLERICAL STAFF l D. A. Adams R. D. Ellis K. E. Waters ILLUSTRATOR T. D. Herrera l l I I Shoreham Broc4 haven, NY Febnaary 16,1995 l l _
l l < l' I i- , l TABLE OF CONTENTS PAGE List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii List of Tables ..............................................iv . Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v . Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l Site Description ............................................. 2 , f Obj ectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 \ Document Review and LIPA Procedure Surveillance . . . . . . . . . . . . . . . . . . . . . . . 4 , Procedures ................................................ 4 i Findings and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 S u m mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 !1 l References ...............................................49 l
- l i
l Appendices: l 1 \ l Appendix A: Major Instrumentation ; i l Appendix B: Survey and Analytical Procedures l Appendix C: Regulatory Guide 1.86 Termination of Operating Licenses for Nuclear Reactors i t t 1 i l l !' ShoretmDrookhaven NY . February 16,1995 1
LIST OF HGURES i PAGE FIGURE 1: location of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 17 ! FIGURE 2: Plot Plan of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . . I8 FIGURE 3: Radwaste Building, Elevations 15'6"/19'6" Floor. Plan-Surveyed Areas ...................... . . . . . . . . 19 t FIGURE 4: Radwaste Building, Elevation 37'6" Floor Plan-Surveyed Areas ...................................20 FIGURE 5: Radwaste Building, Elevation 50'6"/52'6" Floor Plan-Surveyed Areas ...................................21 FIGURE 6: Reactor Building, Elevation 8'0" Floor Plan, Suppression i Pool-Surveyed Areas ...............................22 1 FIGURE 7: Radwaste Building, Radwaste 15' Nonh Hallway (RWO13)- I Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . 23 FIGURE 8: Radwaste Building, Cation / Anion Regen. and Resin Storage i Tanks Area (RWO17)-Measurement and Sampling locations . . . . . . . 24 i FIGURE 9: Radwaste Building, Liner Fill Stations /BW Storage Rooms (Cubicle A) (RWO23)-Measurement and Sampling Locations . . . . . . . 25 l l FIGURE 10: Radwaste Building, Waste Evap./ Regen.Evap. Distil. Room l (RWO40)-Measurement and Sampling locations . . . . . . . . . . . . . . . 26 FIGURE 11: Radwaste Building, Radwaste 37' North Hallway (RWO42)- l l Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 27 l FIGURE 12: Radwaste Building, "B" RW O/G HEPA After Filter Area l (RWO69A)-Measurement and Sampling Locations . . . . . . . . . . . . . . 28 FIGURE 13: Radwaste Building, "C&D" Dryer Skid Area (RWM9B)- Measurement and Sampling Imcations . . . . . . . . . . . . . . . . . . . . . 29 FIGURE 14: Radwaste Building, Off Gas Desiccant Dryers Area (RWO72)- Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . 30 FIGURE 15: Reactor Building, Suppression Pcol, NW Quadrant (SP004) Lower Walls and Floor-Measurement and Sampling locations . . . . . . 31 Shorsam-Brodhavca, NY - February 16,1995 il l l l , .-
i l LIST OF FIGURES (Continued) l PAGE l FIGURE 16: Reactor Building, Suppression Pool, NW Quadrant (SP004) Upper i Walls-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . 32 ; FIGURE 17: Reactor Building, Suppression Pool, Area Inside Vessel Pedestal (SP005)-Measurement and Sampling Iocations . . . . . . . . . . . . . . . . 33 FIGURE 18: Reactor Building, HPCI Valves E41-01V-3049 and 3050 (SUO12)-Measurement and Sampling locations . . . . . . . . . . . . . . . 34 FIGURE 19: Radwaste Building, Radwaste Influent Drain System, Drain Sump Tank-054 (SU14 x09)-Measurement and Sampling locations ..............................35 FIGURE 20: Radwaste Building, Equipment Drain System, , Drain Sump Tank-071 (SU014 x 10)-Measurement and Sampling Locations ..............................36 l FIGURE 21: Radwaste Building, Radwaste Equipment / Components, Flat
- Bed Floor Drain Filter 1G11-FL-012 (SUO14x12)-Measurement and Sampling Locations ..............................37 FIGURE 22: Radwaste Building, Radwaste Equipment / Components, l l Waste Collector Tank 1G11-TK-10A (SUO14 x12)- l Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 38 FIGURE 23: Reactor Building, Reactor Water Clean-up System Components (SUO15)-Measurement and Sampling locations ...............39
, FIGURE 24: Radwaste Building, Condensate Demineralizers, Tank 1N52-DE-002E (SUO43)-Measurement and Sampling Locations . . . . . . . . . . . . . . . 40 l l l 1 i I i f Shortharn-Brodhaven, NY - February 16,1995 jii 1 i
t i LIST OF TABLFR i PAGE TABG 1: Summary of Surface Activity Levels . . . . . . . . . . . . . . . . . . . . . . . 41 t TABG 2: Interior Exposure Rates . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . . 43 TABG 3: ' Embedded Piping Program Summary . . . . . . . . . . . . . . . . . . . . . . 44 TABG 4: Confirmatory Radiological Status Summary-Structures . . . . . . . . . . . 46 1 l TABG 5: Confirmatory Radiological Status Summary-Systems . . . . . . . . . . . . 48 1 l-t I I I l l i I I l l I l t I ( l l NBroaumven, NY . Febnnry 16, 1995 iy
t ABBREVIATIONS AND ACRONYMS l ASME American Society of Mechanical Engineers , cm2 square centimeter cpm ' counts per' minute
- DOE Department of Energy dpm/IO0 cm2 disintegrations per minute per 100 square centimeters -
- EML Environmental Measurements Laboratory . . ) EPA Environmental Protection Agency l . ESSAP Environmental Survey and Site Assessment Program ft2 square feet ha hectare
- GM Geiger-Mueller .
km kilometer 1, critical level ; LILCO Long Island Lighting Company i LIPA long Island Power Authority 1 m meter i
- m2 square meter y MDA minimum detectable activity 1 Nal sodium iodide l
! NIST National Institute of Standards and Technology l 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 i i SU# system survey unit designation l pR/h microroentgens per hour l 1 3
)
1
- l l
l i 1 1 I l ! l i ' abarcham-Brookhaven. NY - February 16,1995 y i j
i l CONFIRMATORY SURVEY OF THE RADWASTE BUILDING, SUPPRESSION POOL, i PHASE 2, AND PHASE 3 SYSTEMS : SHOREHAM NUCLEAR POWER STATION l BROOKHAVEN, NEW YORK INTRODUCTION AND SITE HISTORY The Long Island Lighting Company (LILCO) constnicted 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 198.1. Low power. testing, in accordance with U.S. Nuclear Regulatory Commission (NRC) license No. NPF-82 -l (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
\
t January 1989, at which time power generating operations were terminated. The total operating l , i I 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 j 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 that were identified during characterization studies included iron-55 (Fe-55), cobalt-60 (Co-60), nickel-63, and j smaller quantities of tritium, carbon-14, nickel-59, manganese-54, zinc-65, and europium-152.8 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 l 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 i ShBrookhaven, NY . February 16,1995 i
i l-f:llowed by termination surveys of the remainder of the structures and systems located within . { the Turbine 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. Phase 4 will address the Reactor Building. i i 1 l It is the policy of the NRC to perform confirmatory surveys of facilities that have undergone > l decommissioning and have requested NRC license termination. The NRC Headquarters' ) Division of Waste Management has requested that the Environmental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) f ) conduct confirmatory radiological surveys and related activities for the SNPS decommissioning t 4 3 project as the various decommissioning milestones are completed. The results of the
- confirmatory survey of Phase 1, the turbine intemal components and the Turbine Building, Site ;
l Grounds, and Building Exteriors, are the subject of separate reports.2.s This report describes I the results of the confirmatory process that has been completed for the Radwaste Building , (Phase 3), Suppression Pool (Phase 2), and Phase 2 systems. l SITE DESCRIPTION u I SNPS is located in the Town of Brookhaven, New York on the north shore of Long Island, ; approximately 80 km (50 mile) east of La Guardia Airport and the confluence of the East River l and Long Island Sound (Figure 1). Reactor and supporting operations were conducted within f a 32.4 ha (80 acre) portion of a larger 202 ha LILCO owned parcel of land that is bounded on the north by Iong 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 i Area, also known as the power block, where radiological controls had been necessary-(Figure 2). Each of the buildings that have been or will be addressed during the confirmatory j surveys are located here and are shown on Figure 2 as the Turbine Building, the Reactor j Building, and the Radwaste Bailding. Radwaste Building construction is predominately of 4 2 concrete and structural steel with a total floor space of approximately 4,700 m (51,000 ft2 )
) which is divided between three levels at elevations 15'6"/19'6", 37'6", and 50'6"/52'6" Shareham-Breakhaven.NY February 16,1995 2 i i
i (Figures 3 'hrough 5). The systems and equipment housed within the Radwaste Building include the condensate demineralizers, the liquid radwaste system, the solid radwaste storage area, the crane and truck bay, the makeup water treatment plant, chemical support system, and a portion of the off-gas radwaste system. The Suppression Poolis 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 surveys have been performed in accordance with Draft NUREG/CR-5849.d 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 l l
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, a total of 77 survey units were 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 Shoreham-Brookhaven. NY . February 16,1995 3 I . - , _ . . --
l l 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.8' 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
- applicable procedures and records for both the calibration of instruments used in, and the survey j data generated for, three 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 i
- During the period August 22 through 25,1994, an ESSAP team visited the SNPS and performed independent visual inspections, measurements, and sampling of the Radwaste Building, Suppression Pool, Phase 2, and Phase 3 Systems. Surveys were performed in accordance with j a survey plan submitted to and approved by the NRC.7 Nine structural survey units and either
! the complete or components of four system survey units were selected for confirmatory surveys. j Survey units were selected either randomly by ESSAP or based on recommendations of the NRC l ) site representative. Survey unit designators are alpha-numeric with the first figures designating l the type of unit, structural (building specific, RW=Radwaste, SP = Suppression Pool), or system, 4 followed by a three-digit numeric reference. Subunits are given an additional two-digit i 4 l j designation preceded by X. The survey units selected and the respective classification for each l were: l l 4 I Shoreham-Brookhaven, NY - February 16.1995 4 1
- 1 i
i Affected (A)/ Structure / System / j Survey Unit Survey Unit Name Unaffected (U) Building Grounds ' RWO13 Radwaste 15' North Hallway A structure RW017 Cation / Anion Regen and Resin A structure Storage Tanks Area RWO23 Liner Fill Stations /BW Storage 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 SU012 High Pressure Coolant Injection A system ; Valves E41-01V-3049 and 50 1 SU014XO9 Radwaste Influent Drain System, A system : Drain Sump Tank-054 SU014X10 Radwaste Building Equipment A system ' l Drain System, Drain Sump Tank-071 ! l SU014X12 Radwaste Equipment / Components, A system Flat Bed Floor Drain Filter 1G11-
- FL-012 and Waste Collector Tank i
l 1G11-TK-10A i SU015 Reactor Water Cleanup System A system j SUO43 Condensate Demineralizers, Tank A system ; IN52-DE-002E Figures 3 through 6 indicate the structural survey units surveyed. Confirmatory surveys for SUO12, SUO14X09, SU014X10, SU014X12, and SUO43 involved individual component (s) rather t than the entire survey unit. l Shorehann-Breakbeven, NY February 16,1995 5 l i I
. - ~ . . . . - , . , , . . .,=me
Field survey activities were conducted in accordance with the applicable sections of the ESSAP Survey Procedures and Quality Assurance Manuals. The following procedures apply to survey units selected for iPt confirmatory surveys. SURVEY PROCEDURES Refemnee Systen 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 Mt upon the survey unit classification (affected vs. unaffected) and surface (floor, lower wall, upper wall, ceiling, or quipment). Typically, floor and lower wall grid blocks were 1 m x 1 m. Upper scrfaces and equipment were referenced to either these grids or other prominent building features. Systems were referenced by either a distance from a specific point, by drawings, or by prominent components. Surface Scans ( Surface scans for alpha, beta, and gamma activity were performed over 100% of ficor 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. Imations of elevated direct radiation detected ( by scans were marked for further investigation. Scans were performed using gas proportional, GM, and Nal detectors coupled to ratemeters or ratemeter-scalers with audible indicators. Surface Activit) Measurements For each structural survey unit, ESSAP performsd a minimum of 30 direct measurements for total beta surface activity. ESSAP also pOfwad M41donal direct measurements at locations of elevated direct radiation detected by surface scans. Alpha surface activity measurements were not required because the selected survey units were not classified as alpha affected, and no alpha
- e . xv. ra-,y a. im 6 L
contamination was identified by surface scans. MeasuremcNi wre performed using gas: proportional or GM detectors coupled to 1. muster-scalers. A smear sample for determining zernovable activity levels was collected from each direct measurement location. Figures 7 through 17 show measurement and sampling locations. Expecure Rate Measurunents F-== e 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 ditect measurements and smear samples. The total number of direct measurements performed and smears collected was dependes.t upon component size and ame==ibility 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 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 embedded piping survey design and performance, review of the data and records for the Shosham-Broekhoven. NY Fatmanry 16.1995 7
termination surveys of selected embedded piping subunits, and inelq=ndent measurements and sampling of selected sections of piping. Specific procedures for each activity are described below.
. The cr'"vst'.on records for the detector / instrument combinations, referred to as " pipe crawlers, used dunng 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 UPA 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 i@nt , measurements and calculations to confirm the level of, and distribution of, Co-60 activity on two of the custom sources that UPA used for the calibration check on the pipe crawler detector assemblies. This'information was used to evaluate the appropriateness of the UPA stated' detector efficiencies. Once these parameters were established, ESSAP confirmed the UPA computer-generated surface activity measurement data conversions to disintegrations per minute 2 , per 100 square centimeters (dpm/100 cm ), critical level calculations, and the action level calculations used by UPA 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 were then compared to the UPA-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. UPA was also requested to wm n. ran-y u. ms 8
i resurvey portions of the selected embedded piping subunits and the results compared with the l original termination survey data. .! I SAMPLE ANALYSIS AND DATA INTERPRETATION ; 1 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 - l 2 cf dpm/100 cm . Because Fe-55 cannot be adequately detected with field instrumentation, a -l correction factor of 1.2 was applied to those surface activity measurements that exceeded l background distribution levels, referred to as the critical level (IJ. 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.s.: The 95% confidence level (p.), in accordance I 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 - l was also performed. Exposure rates were reported .in microroentgens per hour ( R/h). ! Additional information concerning major instrumentation and survey and analytical procedures , is provided in Appendices A and B. DATA EVALUATIONS AND COMPARISONS The results of each survey unit sampled were statistically tested. The goal of the test was to determine, with a given confidence level, whether the LIPA survey data was not biased low compared to ESSAP. The null hypothesis was that in a survey unit, surface activities as calculated by LIPA were greater than or equal to those determined by ESSAP, i.e., H : LIPA 2 ESSAP. 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., H 4 : LIPA < ESSAP. The test statistic, t, was calculated using the following equation: m -a em... n . F.6, ,y i6, i,95 9
- , . _ _, ._,-.m.__ _ - . , , - . . . _ . .
i X-X z i (n,-1)s* + (nt -1)s* rn , + nt '
) "s * *i ~2 r "s"z ,
where: { is the LIPA surface activity mean for a survey unit { is the ESSAP surface activity mean for the same survey unit nt is the number of LIPA direct measurement data points na is the number of ESSAP direct measurement data points St, S, are the standard deviations. The calcnhted t was then compared to the critical value of Student's t-distribution (one-tailed) for the appropriate degrees of freedom at the 95% confidence level (0.05 level of significance). ) If the H,: LIPA 2: ESSAP was rejected, then ESSAP evaluated additional options and alternatives and conferred with the NRC as to the recommended approach. HNDINGS 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 correWee.' 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 were appropriately converted, tested, and presented. SURFACE SCANS Alpha, beta, and gamma surface scans identified one small area of elevated direct beta radiation on a support bracket in SP004. The surface activity level at this location was 2 3,800 dpm/100 cm . All other surface scans were comparable to background levels. mm=, w - r.i ry 16, i,95 10
l SURFACE ACTIVITY LEVEIJ!i , The results of total and removable surface activity levels are summarized in Table 1. The data reported below and in Table 1 is the difference between the gross field sample counts and area ; background. The difference is then corrected for detector efficiency and geometry, sample count . ! time, and contributions from Fe-55 (when the net count rate exceeded the background l distribution [I,]). Actual values are reported including negative surface activity levels, which , occurred when the field count rate was less than the background. Of 414 measurements, , 160 exceeded the I,. i Total beta activity levels for the structural survey units ranged from -310 to 3,800 dpm/100 cm2 . j The highest direct measurement was on the SP004 support bracket discussed previously. l Removable activity levels ranged from -1 to 6 dpm/100 cm2 for alpha and -7 to 81 dpm/100 cm2 for beta. The mean residual activity in structural survey units ranged from 14 to i 510 dpm/100 cm 2 and -1.3 to 3.4 dpm/100 cm 2 for total and removable beta activity, ! l respectively. i 2 l Total beta activity levels in the surveyed systems ranged from -760 to 1,600 dpm/100 cm . The 2 2 removable activity levels were -1 to 8 dpm/100 cm for alpha and -6 to 14 dpm/100 cm for l 2 beta. The mean beta activity levels for systems ranged from -310 to 98 dpm/100 cm for total activity and -1.3 to 1.7 dpm/100 cm2 for removable activity. EXPOSURE RATES As determined during 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 R/h at 1 m. The average gross exposure rates for all survey units ranged from 4 to 5 R/h at I m. The net survey unit exposure rates ranged ! from -1 to 1 Rh at I m. Table 2 provides a summary of the exposure rates. l l wm. xv . r.6, ,y i6. i,95 11 l ! ! l l
, -.. .. . -. . J
EMBEDDED PIPING l Review of the embedded piping instrumentation calibration and calibration check; operational l source cross-checks; survey procedure review and surveillance; and review of data conversions, l 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. Specifically, ESSAP performed cross-checks of calibration source activity as well as LIPA reported detector efficiencies using representative pipe calibration geometries. The action level, in counts per minute, that LIPA developed and used for determining when small areas of residual activity in excess of the guidelines were present in a section of pipe was also confirmed. A confirmation of the embedded piping survey design was also conducted. For this confirmation, ESSAP evaluated the LIPA requirements for total number of measurements in a given pipe section and compared the required number to the actual number of measurements that LIPA performed. Because of the unique design of the pipe-crawler detector assemblies, ESSAP also evaluated the appropriateness of surface activity data conversions taken from gross field counts through the final reported surface activity levels. The final evaluations of the LIPA embedded piping program included field observation of embedded piping survey procedures, a comparison of original data and data generated when LIPA l resurveyed three sections of pipe, and independent measurements by ESSAP for comparison with LIPA data. Table 3 provides a summary of the information developed during these evaluations. Surface scans performed in each piping section did not identify any locations of elevated direct radiation. The ESSAP total activity results for each subunit are as follows: SU014X02 (#366), l the range was -320 to 120 dpm/100 cm2 and the mean was -20 dpm/100 cm2 ; SU014X09 (#850), the range was -360 to 40 dpm/100 cm2 and the mean was -140 dpm/100 cm2 ; and SU016X01 (#20), the range was 400 to 690 dpm/100 cm2 and the mean was 540 dpm/100 cm2, l COMPARISON OF RESULTS WITH 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 s w a-o. .m.ra,-yis,m3 12
I guidelines for t.cceptable surface contamination levels used to determine whether a Ecensed facility may be released to unrestricted use. These guidelines are summarized in Appendix C. t- The applicable guidelines are those for beta-gamma emitters of which Co-60 and Fe-55 am the i primary contaminants at SNPS. The residual surface activity guidelines are: 4 Total Activity 2
- 5,000 dpm #-y/100 cm2, averaged over 1 m 15,000 dpm S-7/100 cm2 , maximum in 100 cm 2 ,
Removable Activity 3 1,000 dpm #-7/100 cm2 ! As previously discussed, the detection sensitivities of thc field instruments are such that the , msidual Fe-55 activity cannot be detected. Therefore, tote and removable surface activity l measurements were corrected for Fe-55 when appropriate. The mean surface activity level for each survey unit was calculated and the survey unit data tested at the 95% confidence level ( , l or upper confidence level [UCL]), relative to the guidelines, in accordance with Draft ] i NUREG/CR-5849. These results are provided in Tables 4 and 5. No direct measurements 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 i 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 4 therefore been satisfied for these survey units. However, the ESSAP mean was greater than the
. LIPA mean for survey units RWOl3, RW017, RWO23, RWO42, RWO72, SP004, SUO43, SUO14X09, and SU014X10; therefore, additional evaluation was necessary.
a mamm. xv . re-ry is, i,95 13
l ! 7 l- Of the nine units where the ESSAP total surface activity mean was higher than the UPA mean, . ; UPA qualified the release record data for survey units RW013, RW017, RWO23, SP004, and : ; {' SUO43 as containing excessive negative measurements. Accordmg to UPA, this was due to; l 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 l 2 units was 3,800 dpm/100 cm in SP004, and the maximum survey unit mean and UCL, found" l in RWO23, was 300 dpm/100 cm2 and 400 dpm/100 cm2 , respectively. Although the condition j cf the ESSAP mean being less than the UPA mean was not satisfied for these survey units, l both termination and confirmatory survey surface activity levels are below the 5,000 ; dpm/100 cm2average guidelme. 1 l The ESSAP surfa::e activity means for survey units RWO42,- RWO72, SU014X09, and l SUO14X10 were statistically higher than those for UPA, and LIPA had not 'quali5ed the data I I for these units as biased low. Therefore, the UPA surface activity levels, mean actWhy kvels, l' I j 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 cm2 (SUO67) and maximum UCL of 812 dpm/100 cm2 (RWO61) were within sc@ble criteria. The maxinium observed difference 2 cf 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. For embedded piping, ESSAP's surface scans and direct measurement results did not indicate the presence of residual surface activity within the embedded piping, with most surface activity i l levels comparable to background and all levels below the minimum detectable activity (MDA) of the instrumentation. The comparison of the LIPA and ESSAP embedded piping . direct measurement data indicates two discrepancies. First, the LIPA mean activity levels (Table 3) for two of the three sections of pipes investigated do not agree for the original survey and the
. resurvey of SU014X02 (#366) and SU014X09 (#850). The probable cause of the difference in l
I wn m=.m aan-y 26, im 14 1
( . - - .. . . . .-- ._ - , - . . _ _ _ _ _ _ I l the means 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 SU014X02 (#366) was 22 counts per minute (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 SUOl6X01 (#20) were comparable. l l The second discrepancy was identified when the comparison of the means indicated ~that the ESSAP mean was statistically greater than the LIPA mean for each of the three data sets. 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 measurement location consists cf a 93 square centimeter (cm') area. LIPA's reported total activity is the additive' activity contribution from six 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. Based on these factors, LIPA's conclusions as to the overall radiological status of the SNPS embedded piping _ appear to be appropriate. Exposure rates were compared with those obtained by LIPA and tested at the 95% confidence l level, relative to the 5 R/h above background guideline currently being used by the NRC (Table 4).' The Radwaste Building exposure rates were comparable to background exposure rate . levels and confirmed the findings presented by LIPA. ; l l
SUMMARY
ESSAP performed confirmatory activities for the Radwaste Building, Suppression Pool, and 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 ! 25, 1994, independent surface scans, surface activity measurements, exposure rate measurements, and operational surveillance were performed. i war.*h n . r*n-y 16.1995 15 l
The survey results confirm the results of the LIPA termination surveys. These findings 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. NBreatlaven, NY . I;chruary 16,1995 16
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# EXPOSURE RATE 0
WETERS ! FIGURE 8: Redwoste Building, Cation / Anion Regen, and Resin Storoge Tanks d Area (RWO17) - Measurement and Sampling Locations j wamven, ny - February 16,1995 24 2
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6 i METERS FIGURE 13: Rodwoste Building, "C&D" Dryer Skid Area (RWO698) - l Measurement and Sampling Locations i NBrookhavm.NY February 16,1995 29
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258-017 (2) I w l N / OPENNG ONLY f-- ]- 7-7 UNACCESSBLE TO DO UP & DOWN STREAM 6- 6 _/ \__ f , w PROBE C(NERS ENTIRE VALVE ATTACHMENT
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) FIGURE 18: Reactor Building, HPCI Valves E41-01V-3049 and 3050 (SUO12) - ) Measurement and Sampling Locations shardiams-Brookhaven, NY - February 16,1995 34 7 ~
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$A O METERS 1 FIGURE 19: Rodwoste Building, Rodwoste influent Drain System, Droin Sump Tank 054 (SU14XO9) - Measurement and Sampling Locations ww. ny r*=ry i6, ms 35 l
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h FIGURE 21: Rodwoste Building, Rodwoste Equipment Drain Filter 1G11-FL-012 (SUO14X12) / Components,
- Measurement Flat Bed and Sampling Floor Locations shoreham-Brodhsvem, NY . February 16,1995 37
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METERS FIGURE 23: Reactor Building, Reactor Water Clean-up System Components (SUO15) - Measurement and Sampling Locations Shoreham-Brookhaven NY . February 16,1995 39
l i 258-022 (3) l l l SOUTHWEST HALF NORTHEAST HALF i i ENT INFLU P!PE Q PIPE (( ~ l ~
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TABLE 1
SUMMARY
OF SURFACE ACTIVITY LEVEIE RADWASTE BUILDING, SUPPRESSION POOL, AND PHASE 2 SYSTEMS e SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Removable Number of Total Activity Range Activity Range I4 cation
- Measurement -(dym/100 cm') (dpm/100 cm2)
Locations Beta *** Alpha, Beta
- RW913[15' North Hallwayl 2 '
~
4 3 Floor 11 10 to 540 -1 to 1 -5 to 1-Iower 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 L ' Floor 10 -200 to 360 -1 to 3 -3 to 9 Iower 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 RWG23719' 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 RW640,15' Evaporation Distil Room Floor 13 - 220 to 440 -1 to 3 -5 to 9 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 RW642,~37' North Hallway. Floor 14 450 to 920 -1 to 1 -4 to 4 lower Walls 6 230 to 460 -1 to 5 -4 to 3 Upper Walls 3 190 to 280 -1 to 5 0 to 3 Equipment 7 -150 to 520 -1 to 1 -5 to 3 Shoreham-Broaboven. NY . February 16.1995 41
TABLE 1 (Continued) SUMh1ARY OF SURFACE ACTIVITY LEVELS RADWASTE BUILDING, SUPPRESSION POOL, AND PHASE 2 SYSTEMS SHOREHAM 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 b,e RWO69, HEPA After Filter Area Floor 11 110 to 490 -1 to 4 -3 to 6 Iower Walls 9 -62 to 130 -1 to 3 -4 to 10 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 Floor 11 410 to 920 -1to3 -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. Juad. - 31 -310 to 3800 -1 to 6 -5 to 81 SP005, Suppression Pool In.-Pedestal 35 -210 to 350 -1 to 3 -7 to 8 SUO12, Hi Pressure Coolant Injection 4 -440 to -200 -1 to 1 -5 to 8 SUO14x09, Tank 54, Influent Drain 31 -320 to 870 -1 to 1 -3 to 6 SUO14 x10, Radwaste Equipment Drains 11 -760 to 1600 -1 to 1 -4 to 6 SUO14x12, kadwaste Equipment / Comp. 50 -720 to 440 -1 to 5 -6 to 14 SUO15, Reactor Water Cleanup System- 30 -520 to 1100 -1 to 5 ,,4 to 5 SUO43, Condensate Demineralizer 12 -270 to 300 -1 to 8 -5 to 3 ' Refer to Figures 7 through 24. 6 Beta activity levels corrected for Fe-55 contribution as appropriate. 'MDAs = 250 to 1100 dpm/100 cm2, d MDA = 12 dpm/100 cm2, 'MDA = 16 dpm/100 cm 2, Ahorehan-Brookheven. NY - February 16, 1995 42
TABLE 2 INTERIOR EXPOSURE RATES RADWASTE BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK I4 cation
- I4 cations at 1_n (pR/h?
RW013 11 -1 to 0 , RW017 6 -1 to O RWO23 8 -1 to 0 RWO40 6 -1 to 0 RWO42 13 0 to 1 RWO69 6 0 to 1 RWO72 11- 0
' Refer to Figures 7-14.
b Site background exposure rate was 5 R/h. NBrecibeven, NY . February 16,1995 43
TABLE 3 EMBEDDED PIPING PROGRAM
SUMMARY
SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK p .v Pipe Emciency 4' Pipe LIPA 6 detector Pipe Crawler Assembly 0.134* ESSAP HP-260 0.1466 ~ 8" Pipe LIPA 6 War Pipe Crawler Assembly 0.152' ESSAP HP-260 0.1636 Description LIPA Expected /Actuaf - ESSAP Confirmation 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 spotd) 2408 cpm /830 cpm' 2410 cpm
> Survey n sp ; ,
Required Number of Measurements Confinned Number of
""'I "
(Per SP No. 66x020.11) Measurements SU016x01 (#20) 14 15 SUO14x02 (#366) 28 30 SUO14x09 (#850) 26 28 E Data Convenibs LIPA ESSAP Confinnation Location Surface Activity Critical Imvel Surface Activity Critical I2 vel (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm2) (d&!100 cm 2) 20-1-3 -290 270 -290 270 20-5-2 -298 270 -298 270 366-2-1 -382 175 -382 l'/5 366-5-2 -271 181 -271 181 366-9-3 -338 178 -338 178 handhaves. NY - February 16,1995 44
l TABLE 3 (Continued) EMBEDDED PIPING FROGRAM
SUMMARY
SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK
-Data Conversion(Continued):
LIPA ESSAP Confirmation I4 cation Surface Activity CriticalI* vel Surface Activity Critical level (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm?) 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 Rssults Survey Unit Data Origination SU0l6x01 (#20) SUO14x02 (#366) SUO14x09 (850) LIPA Termination Total Activity Range -468 to 169 -401 to -111 -379 to 16 Surv y (dpm/100 cm2)
-245 -240 -196 Mean Activit{)
(dpm/100 cm LIPA Resurvey Total Activity Range -395 to -121 -469 to -271 -581 to -315 (dpm/100 cm )
-245 -337 -488 Mean (dpm/100Activity) cm ESSAP Survey Total Activity Range 400 to 690 -320 to 120 -360 to 40 (dpm/100 cm2 )
540 -20 -140 Mean Activit{) (dpm/100 cm "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 Assembly Calibration Check Data Sheet (SPF66 x020.11) for detector assembly IZ12-100PC4-0001.
d Action level developed to identify small areas of contamination measuring less than 100 cm2and matching a pipe-crawlers individual " pancake" detector active area.
'830 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 gen: rated report.
hMm, NY . h* nary 16, 1995 45
TABLE 4 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES, RADWASTE BUILDING AND SUPPRESSION POOL ~ SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiologleal Survey Unit'
""" RW013 RW017 RW023 RW640 RWO42 TM M M~(M00'd':
~
# of Direct Measurements 30 30 30 30 30 Mean(I) 82 ~14 300 240 430' LIPA I -170 -130 13 110 ~280, F= 140 60 400 290 510 5,000/15,0000 dpm/100 cm' Yes Yes Yes Yes Yes Guidelines Satisfied
~
Removable Beta Activity (dpm/100 enf)*
# of Smears 30 -30 30 30 30 Mean(I) -0.5 0.1 -0.4 0.2 -1.3 LIPA I 7.5 2.6 3.9 3.2 0.3 Fa 0.4 0.9 0.6 1.2 -0.5 2
1,000 dpm/100 cm Yes' Yes Yes Yes Yes Guideline Satisfied Exposure Rates at.1 m (gR/h)"
# of Exposure Rate Measurements 11 6 8 6 13 Net Mean (I) -0.7 -0.9 0.4 -0.5 0.2 Net LIPA I -0.2 0.5 0.4 0.0 0.2 F- -0.6 -0.7 -0.1 , -0.2 0.3 5 R/h Above Background Yes Yes Yes Yes Yes Guideline Satisfied wav., NY . Fenmy 16, IM 46
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES, RADWASTE BUILDING AND SUPPRESSION POOL SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit'
"""7 RWO69 RWO72 SP004 SP005 TotalBeta ActMty (doma/100 amp)? *
# of Direct Measurements 30 30 31 35 Mean(I) 93 510 110 100 LIPA I 360 280 -220 76 Fa 140 580 350 150 5,000/15,0000 dpm/100 cm2 Yes Yes Yes Yes Guidelines Satisfied Ranovable Beta ActMty (dpm/100 ce?)O
# 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 F= 1.2 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 ' -
Net Mean (I) 0.3 0.0 - - Net LIPA X 0.1 0.1 - - F= 0.4 0.2 - - 5 R/h Above Background Yes Yes - - Guideline Satisfied
' Refer to Figures 7 through 17.
b All al# reMle activiy a less than 12 dpm/100 cm'.
- = Measurements not performed.
m.,*.--m. xy rer=ry a. ms 47
TABLE 5 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Surrey Unit' Summary SUO12 SUO14x09 SUO14x12 SUO14x10 SUO15 SUO43 Total Bet 5 Activity (dpm/100 cnf) -
# of Direct Measumments 4 31 11 50 30 12 Mean(X) -310 -48 26 80 -32 66 LIPA I 150 -170 -170 11 87 4.6 F=
-190 42 170 , 180 59 160 5,000/15,0000 dpm/100 enf Yes Yes Yes Yes Yes Yes GuiMines Satisfied Removable Beta" Activity (dpm/100 cal)b .:
# of Smears 4 11 7 50 25 12 Mean(I) 0 0.9 -1.3 1.7 0.2 -0.3 LIPA I 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 dpm/100 cm 2, Shoreham-Brookhsven NY . February 16, 1995 48
REFERENCFE
- 1. I.ong Island Lighting Company, "Shoreham Nuclear Power Station Site Characterization Program Final Report," May 1990.
- 2. T. J. Vitkus, ORISE, " Confirmatory Survey of the Turbine Internal 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 l Radiological Surveys in Support of License Termination," NUREG/CR-5849, June 1992.
- 5. Long Island Power Authority, "Shoreham Decommissioning Project, Termination Survey Plan, Revision 1," April,1993.
i
- 6. Long Island Power Authority, "Shoreham Decommissioning Project Termination Survey Final Report, Volumes 1 through 5," September,1993.
- 7. Letter 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. Letter from M. R. Landis, ORISE to D. Fauver, U.S. Nuclear Regulatory Commission, "Shoreham Decommissioning Project, Termination Survey Plan, Revision O, Shoreham l 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.
I l
%reham-Brookhaven.NY February 16,1995 49 l
i I l i l l l l l I t APPENDIX A l l MAJOR INSTRUMENTATION ! l l l l l l l [ l l I \ l Shortham-Broothaven. NY - February 16,1995 l
APPENDIX A MAJOR INSTRUMENTATION The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the authors or their employers. DIRECT RADIATION MEASUREMENT Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM) Eberline " Rascal" Ratemeter-Scaler Model PRS-1 (Eberline, Santa Fe, NM) Ludlum Ratemeter-Scaler Model 2221 (Ludlum Measurements, Inc., Sweetwater, TX) Detectors Eberline GM Detector Model HP-260 Effective Area,15.5 cm2 (Eberline, Santa Fe, NM) Ludlum Gas Proportional Detector Model 43-37 Effective Area,550 cm2 (Ludlum Measurements, Inc., Sweetwater, TX) Ludlum Gas Proportional Detector Model 43-68 l Effective Area,100 cm 2 (Ludlum Measurements, Inc., Sweetwater, TX) hBem, NY . February 16,1995 A-1 D
Reuter-Stokes Pressurized Ion Chamber Model RSS-111 (Reuter-Stokes, Cleveland, OH) Victoreen NaI Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal
- (Victoreen, Cleveland, OH)
LABORATORY ANALYTICALINSTRUMENTATION Iow Background Gas Proportional Counter
- Model LB-5100-W (Oxford, Oak Ridge, TN) sheruham-Breakhaven. NY - February 16, 1995 A-2
I APPENDIX B SURVEY AND ANALYTICAL PROCEDURES 3 Shoreham-BrmAhsven, NY - February 16,1995
\ APPENDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES Surface Scans Surface scans were performed by passing the probes slowly over the surface; the distance 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 cm 2
,59 cm2 or 100 cm 2) 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 Beta - gas proportional detector with ratemeter-scaler pancake GM detector with ratemeter-scaler Gamma - Nal scintillation detector with ratemeter Surface Activity Measumments Measurements of total beta activity levels were performed using GM and gas proportional detectors with portable ratemeter-scalus. Count rates (cpm), which were integrated over 1 minute in a static position, were converted to 2 activity levels (dpm/100 cm ) by dividing the net rate by the 4 x efficiency and correcting for the active area of the detector. The beta activity background count rates for the GM and gas w a v ny-r.6, ,yi6.:995 B-1
pivpunional detectors ranged from 22 to 34 cpm and from 129 to 174 cpm, respectively. Beta efficiency factors 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 100 cm 2, respectively. Surface activity measurements which exceeded the normal background distribution were 2 corrected for the Fe-55 contribution by multiplying the dpm/100 cm field activity level by a factor of 1.2. The instrument response level at which the MMM output could be considered above background was defined as the critical level (Q. This level was defined for each Wm/ instrument combination as follows: gy Sample count rate Backround count rate h ample count time Backround count time L* , (Detector Effciency) (Detector Geometry) Removable Activity Measurements Removable activity levels were determined using numbered filter paper disks, 47 mm in diameter. 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. Exposure Rate Measurements Measurements of gamma exposure rates were performed using a pressurized ionization chamber (PIC). ANALYTICAL PROCEDURES Removable Activity Smears were counted on a low background gas proportional system for gross alpha, and gross beta activity. wa-u . m . m-y is, im B-2
UNCERTAINTIES AND DETECTION LIMITS The uncertainties associated with the analytical data presented in the tables of this report represent the 95% confidence level for that data. These uncertainties 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 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)]. Although data is reported as actual values, including negative values, in the document text and tables, the MDAs for total and removable activity levels are provided in the footnotes of applicable tables. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclide in samples, the detection limits differ from the sample to sample and 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 recognized organization were used. Calibration of pressurized ionization chambers was performed by the manufacturer. Analytical and field survey activities were conducted in accordance with procedures from the following documents of the Environmental Survey and Site Assessment Program: Survey Procedures Manual, Revision 8 (December 1993) Laboratory Procedures Manual, Revision 8 (August 1993) Quality Assurance Manual, Revision 6.1 (November 1993) Shorehem-Breakhaven NY - February 16, 1995 B-3
l 1 The procedures contained in these manuals were developed to meet the requirements of DOE 4 i Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to assess ' j processes during thir performance. . I j Quality control procedures include: -)
- Daily instrument background and check-source measurements to confirm that {
J equipment operation is within acceptable statistical fluctuations.
- Participation in EPA and DOE /EML Quality Assurance Programs. j
- Training and certification of all individuals performing procedures.
{ I
- Periodic internal and external audits.
1 1 + t i i l l 3 4 1 NB@ven. NY - February 16, 1995 B-4
= j + i 4 i 4 i 1
?
s
- APPENDIX C
- . REGULATORY GUIDE 1.86, TERMINATION OF OPERATING 4
LICENSES FOR NUCLEAR REACTORS i i E i i 1 l i e 1 i l
- I e
I 1 l 2 ) 3 4 f 1
- 1 i
e . . I l i 1 4 i d Shoreham-Brodhaven, NY February 16,1995 i i i
U.S. ATOMIC ENERGY COMMISSION June 1974 REGULATORY GUIDE DIRECTORATE OF REGULATORY STANDARDS REGULATORY GUIDE 1,86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION important to the safety of reactor operation is no longer required. Once this possession-only license is issued, Section 50.51, " Duration 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 licen.see having a possession-only license must period, the license may be either renewed or termmated retain, with the Part 50 license, authorizatior for by the Commission. Section 50.82, " Applications for special nuclear material (10 CFR Part, 70, "St ucial termmation oflicenses,* specifies the requirements that Nuclear Material"), byproduct material (10 CFR Part must be satisfied to termmate 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 recurity or to the fuel, radioactive components, and sources are removed health and safety of the public. This guide describes from the facility. Appropriate admmistrative controls methods and procedures considered acceptable by the and facility requirements are imposed by the Part 50 Regulatory staff for the termination of operating license and the technical specifications to assure that licenses for nuclear reactors. The advisory Committee proper surveillance is performed and that the reactor on Reactor Safeguards has been consulted concerning facility is 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, he may, as a first step in the requirements imposed depend on the option selected. process, request that his operating 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 reseech reactors and critical facilities, this has USAEC REGULATORY GUIDES Cap.= or put*shed guida mey t,e obtamed by request indicetmo etw ewon Reguistory ouedes we issued to descree and mee evmistne to sta e desired to the U.s. Atomw Energy C. _ _ _ , Wastengton D.C. 2o645. methods acceptstne to ste AEC regulaiory staff of impiamentmo spectie parts Anente Dirma of Reguimwy standards. Canunus and suggenens for of the commesion's reguistions, to origmete sectnques used by the staff in enprovenwnts in sneu gWen are encanaged and armuid tie sent to em evaluetmo specifs problems or postuisted acc6 dents, or to provide guldance to " " * ' " ' '
*"'*"0' '
apphcants. Reguistory Guedes are not substitutes for regulations and 2 C Maff-those a out in t gwdes w a ttwy at or e gWdes are in Wie tonDwing hn broad evleens. fmdmos reawaste to the msuance or conimuence of a pwmit or imense by saw
*""*"" 1. Power Reactors 6. Products
- 2. Research and test Reactors 7. Transportaten Putsshed guides wili be revised peredcativ, as approprote, to accommodate 3. Fuses and Materiais Fachtes S. occupational Health comments and to rettect new mfornmeteon or experience. enta and mg Reew L
Note: Section electronicaHy reproduced from photocopy. C-l
usually meant the disassembly of a reactor and its b. A description cf measures that will be taken to shipment organization for further use. The site from prevent criticality or reactivity changes and to which a reactor has been removed must be minimize releases of radioactivity from the facility. decontaminated, as nacesanry, and inspected by the Commission to determme whether unrestricted access c. Any proposed changes to the technical can be approved. In the case of nuclear power specifications that reflect the possession-only facility reactors, dismantling has usually been accomplished by status and the necessary diammaembly/ retirement shipping fuel offsite, making the reactor moperable, activities to be performed. and disposing of some of the radioactive components.
- d. A safety analysis of both the activities to be accomplished and the proposed changes to the Radioactive components may f.>e either shipped technical specificat ons off-site for burial at an authorized burial ground or secured on the site. Those radioactive materials c. An inventory of activated matenals and their remaming on the site must be isolated from the public locanon in the facility.
by physical barriers or other means to prevent public access to hazardous levels of radiation. Surveillance is 2. ALTERNATIVES FOR REACTOR nace=<ary to assure the long term integrity of the RETIREMENT bamers. The amount of surveillance required depends upon (1) the potential hazard to the health and safety of Four alternatives for retirement of nuclear reactor the public from radioactive maternal remaming on the facilities are considered acceptable by the site and (2) the integrity of the physical barriers. Regulatory staff. 'Ibese are: Before areas may be released for unrestricted use, they must have been decontaminated or the radioactivity a. Mothballing. Mothballing of a nuclear reactor must have decayed to less than prescribed limits facility consists of putting the facility in a state of (Table 1), protective storage, in general, the facility may be left intact except that all fuel assemblies and the The hazard associated with the returned facility is radioactive fluids and waste should be removed evaluated by considering the amount md type of from the site. Adequate rachation monitoring, remaining contamination, the degree of confinement of environmental surveillance, and appropriate security the remaining radioactive materials, the physical procedures should be established under a security provided by the confinement, the susceptibility possession-only license to ensure that the health and to release of radiation as a result of natural phenomena, safety of the public is not endangered. and the duration of required surveillance,
- b. In-Place Estaenbment. In-place entombment C. REGULATORY POSITION consists of sealing all the remainmg highly radioactive or contammanent components (e.g., the
- 1. APPLICATION FOR A LICENSE TO POSSESS pressure vessel and reactor internals) within a BUT NOT OPERATE (POSSESSION-ONLY structure integral with the biological shield after LICENSE) having all fuel assemblies, radioactive fluids and wastes, and certain selected components shipped A request to amend an operating license to a offsite. The structure should provide integrity over possession-only license should be made to the Director the period of time in which significant quantities of Licensing, U.S. Atomic Energy Commission. (greater than Table I levels) of radioactivity remam Washington, D.C. 20545. The request should include with the material in the entombment. An the following information: appropriate and continuing surveillance program should be established nuder a possession-only
- a. A description of the current status of the facility. license.
Nota: Section electronically reproduced from photocopy. C-2
l :
- c. Rem:vr.1 of Radioactive. Components and b. The physical barriers to unauthorized entrance i Dismantling. All fuel assemblies, radioactive fluids into the facility, e.g., fences, buildings, welded doors, ;
- and waste, and other materials having activities and access openings, should be inspected at least
! above accepted umestricted activity levels (Table 1) quarterly to assure that these barriers have not j 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. j l with no requirement for a license. If the facility i l owner so desires, the remamder of the reactor c. A facility radiation survey should be performed l l facility may be dismantled and all vestiges removed at least quanerly to verify that no radioactive material j and disposed of, is escaping or being transported through the {
i containment barriers in the facility. Sampling should l
- d. Conversion to a New Nuclear System or a be done along the most probable path by which l l
l Fossil Fuel System, nis alternative, which applies radioactive material such as that stored in the inner l only to nuclear power plants, utilizes the existing contamment regions could be transponed to the outer i turbine system with a new steam supply system. regions of the facility and ultimately to the environs. { Re original nuclear steam supply system should be i separated from the electric generating system and d. An environmental radiation survey should be ! disposed of in accordance with one of the previous performed at least semiannually to verify that no j l three retirement alternatives. significant amounts of radiation have been released to l l the environment from the facility. Samples such as ! i 3. SURVEILLANCE AND SECURITY FOR THE soil, vegetation, and water should be taken at locations ; l RETIREMENT ALTERNATIVES WHOSE for which statistical data has been established during l FIN A L STATUS REQUIRES A reactor operations. j POSSESSION-ONLY LICENSE I c. 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. j amount of radioactivity in the form of activated and j. contaminated hardware and structural materials. f. Administrative procedures should be established l Surveillance and commensurate security should be for the notification and reponing .of abnormal j 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 signiticant
- a. Physical security to prevent inadvenent exposure change in the radiation or contammation levels in the of personnel should be provided by multiple locked facility or the offsite environment.
barriers. The presence of these barriers should make it extremely difficult for an unauthorized person to gain g. The following reports should be made: j access to areas where radiation or contammation levels
^
exceed those specified in Regulatory Position C.4. To (1) An annual report to the Director of
- 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 thould 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.
, structures and/or shielding material. Means such as a i remote-readout intrusion alarm system should be (2) An abnormal occurrence repon to the i provided to indicate to designated personnel Ten a Regulatory Operations Regional Office by telephone physical barrier is penetrated. Security person.. I that within 24 hours of discovery of an abnormal j provide access control to the facility may be used occurrence. The abnormal occurrence will also be j instead of the physical barriers and the intrusion alarm reponed in the annual repon described in the preceding systems. item. 1 ] Not:: Section electronically reproduced from photocopy. C-3
,m , , _ , ,
. - . . ..,m. . - - _ , . ,--,.-,,-...,,..,-..w . w~.r y
- h. Records or 1:gs relative to the f:llowing items d. Qa request, the Commission may tuthorize i 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. His may include, but is not limited to, special circumstances such as the (1) Environmental surveys, transfer of premises to another licensed organization that will continue to work with radioactive materials.
(2) Facility radiation surveys, Requests for such authorization should provide: (3) Inspections of the physical barriers, and (1) Detailed, specific informatiac describing the Premises, equipment, scrap, and radioactive (4) Abnormal occunences. contammants and the nature, extent, and degree of
- 4. DECONTAMINATION FOR RELEASE FOR residual surface contammation.
UNRESTRICTED USE (2) A detailed health and safety analysis indicating If it is desired to termmate a license and to that the residual amounts of materials on surface areas, eliminate any further surveillance requirements, the together with other considerations such as the facility should be sufficiently decontammated to prevent prospective use of the premises, equipment, or scrap, risk to the public health and safety. After the are unlikely to result in an unreasonable risk to the decontammarion is satisfactorily accomplished and the health and safety of the public. , site inspected by the Commission, the Commission may authorize the license to be terminated and the facility e. Prior to release of the premises for unrestricted abandoned or released for unrestricted use. The use, the licensee should make a comprehensive licensee should perform the decontammation using the radiation survey establishing that contammation is following guidelines: within the limits specified in Table 1. A survey repon should be filed with the Director of Licensing, U.S.
- e. The licensee should make a reasonable effon to Atomic Energy Commission, Washington, D.C. 20545, eliminate residual contamination, with a copy to the Director of the Regulatory Operations regional Office having jurisdiction. The
- b. No covering should be applied to radioactive repon should be filed at least 30 days prior to the surfaces of equipment of structures by pamt, plating, or planned date of abandonment. The survey repon other covering material until it is known that shoult contamination levels (determined by a survey and documented) are below the limits specified in Table 1. (1) Identify the premises; In addition, a reasonable effon should be made (and documented) to further minimize contamination priot to (2) Show that reasonable effort has been mad.:
- any such covering. reduce residual contamination to as low as practics.w levels, -
)
- c. The radioa::tivity of the interior surfaces of pipes, drain lines, or ductwork should be determined (3) Describe the scope of the survey and the general l by making measurements at all traps and other procedures followed; and appropriate access points, provided contammation at j these locations is likely to be representative of (4) State the finding of the survey in units specified l contamination on the interior of the pipes, drain lins, in Table '.
l or ductwork. Surfaces of premises, equipment, or I scrap which are likely to be contaminated but are of After review of the report, the Cornmission may l l such size, construction, or location as to make the inspect the facilities to confirm the c cy prior to j surface inaccessible for purposes of measurement granting approval for abandonment. should be assumed to be contaminated in excess of the I permissible radiation limits. I 1 \ l Note: Seuion electronicaHy reproduced from photocopy. C-4 I I
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- 5. REACTOR RETIREMENT PROCEDUREC i
As indirmad in Regulatory Position C.2, several I a:ternatives are acceptable for reactor facility j i retirement. If minor diannaembly or "mothballing"_is ; i l pinnnad, this could be done by the caisting operaung i and matanenmare procedures under the license in effect. 'l Any planned actions involving an 6 ea. I safety 1 question or a change in the nachnical sperifications l should be reviewed and approved in accordance with ts of 10 CFR I 50.59. I; the %.J. I If major structural changes to radioactive , components of the facility are planned, such as removal ! cf the pressure vessel or major components of the f primary system, a diamantlement plan including the ; information required by i 50.82 should be submitted to l
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the Com====iae. . A danmantiement plan should bc submined for all the alternatives of Regulatory Position { C.2 except mothballing. However, minor dinamaembly activities may stil! be performed in the absence of such . a plan, provided they are permined by existing ; operating and maintenance procedures. A ; l diamantlement plan should include the followag: i 1 \ 1 l - a. A description of the ultimme status of the facility l
'b. A description of the diamantling activities and
[ the precautions to be taken. l
- c. A' safety analysis of the dismantling activities including any effluents which may be released.
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- d. A safety analysis of the facility in its ultimate status. I Upon satisfactory review and approval of the dismantling plan, a dismantling order is issued by the Commission in accordance with i 50.82. When dismantling is completed and the Commission has been
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notified by , letter, the appropriate Regulatory i Operations' Reaional Office mspects the facility and i verifies completion in accordance with the 1 dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the Commission may termmate the license. If possession-only license under tyhich the dismantling activities have been conducted ; cr. as an alternative, may make application to tne State I
- (if an Agreement State) for a byproduct materials l license.
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' Note: Section electronicelly reproduced from photocopy. C '
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TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS' Nuclide* Average *d Maximum *d Removable ** l U-nat, U-235, U-238, and [ 2 associated decay pmducts 5,000 dpm a/100 cm 2 15,000 dpm a/100 cm 1,000 dpm a/100 cm2 ! Transuranics, Ra-226, Ra-228, ; Th-230, Th-228, Pa-231, ! Ac-227,1-125, I-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th-nat, Th-232, Sr-90, Ra-223, t Ra-224, U-232,1-126, I-131, I-133 1,000 dpm/100 cin2 3,000 dpm/100 cm2 200 dpm/100 cin* Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 L ' * , others noted above. 5,000 dpm Sy/100 cm2 15,000 dpm Sy/100 cm2 1,000 dpm Sy/100 cm2 i j 'Where surface contammarion by both alpha- and beta-gammadtting nuclides exists, the limits established for alpha- and l beta-l gamma-emitting nuclides should apply independently.
*As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive matenal as determmed by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated
! cith the instrumentation. ! ' Measurements of average contammant should not be averaged over more than 1 square later. For objects ofless surface i area, the average should be derived for each such ooject.
"The maximum contarmnation level applies to an area of not more than 100 cm2 . 'The amount of removable radioactive material per 100 cm2 of surface area should be determmed by wiping that area with dry filter l or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an l appropriate instrument of known efficiency. When removable contammation on objects ofless surface area is determmed, I the peninent levels should be reduced proportionally and the entire surface should be wiped.
l NotJ: Section electronically reproduced from photocopy, C-6
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