ML20062D396
| ML20062D396 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/24/1977 |
| From: | NEWPORT NEWS INDUSTRIAL CORP. (SUBS. OF NEWPORT NEWS |
| To: | |
| Shared Package | |
| ML17053A245 | List: |
| References | |
| NNI-77-7NP, NUDOCS 7811220191 | |
| Download: ML20062D396 (109) | |
Text
{{#Wiki_filter:',m , n g (q Newport News Industrial Co@cienon b- hb REPORT NO. EI/NNI-77-7-NP JUNE 24, 1977 TOPICAL ltKni 7 IhR-lE PADWASTE EUIE EDl.UIm SYSTE4 l PREPARED FOR: e UNITED STATES NUCLEAR REGULATORY COMMISSION
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. WASHINGT0ft, D.C.
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( NOTICE This report was prepared by Energy Incorporated in cooperation with Newport News Industrial Corporation. Neither Newport News Industrial Corporation, nor Energy Incorporated, nor any of their employees, nor (- any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, nor assumes any legal liability or responsibility
. for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, nor represents that its use would not infringe privately owned right's.
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- ABSTRACT This document describes the RWR-1 volume reduction system ~ for application to radioactive waste produced at nuclear electrical generating stations. .
This process utilizes a fluidized-bed to calcine liquids and incinerate solids within a connon chamber but with separate modes of operation. Volume reduction is projected to vary from 5 to 1 for filter sludges to 80 to 1 for compacted combustible solid waste. g- Typical waste stream data from Boiling Water Reactors and Pressurized Water Reactors are examined. Decontamination factors (DFs) across
, similar systems are reviewed and DFs for the RWR-l System are developed.
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Anticipated releases to the atmosphere are computed for normal and abnormal operations and shown to be within the prescribed limits. The RWR-1 System components, instrumentation and controls, materials of construction, and operating characteristics are described. It is shown that the components and controls have been chosen on the basis of operating reliability and safety. ( e C = 6 . t; i __- ~,__ __ __ , , _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ -
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TABLE OF CONTENTS PAGE ABSTRACT 4 l.0 INTRODUCTION j 1.1 Purpose j 1.2 Scope 2
.; 1.3 Applicability 2 1.4 History and Background 2
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2.0 DESIGN BASES 4 2.1 General 4 2.2 Light Water Reactor Wastes which are Volume 5 Reducible 2.2.1 Types of Wasta S 2.2.2 Maximum and Expected Volumes and Activities 5 of Vasta 2.2.3 Specific Activities 9 2.3 RWR-1 System Sizing 10 2.3.1 Introduction 10 2.3.2 Wasta Generation Rates 11 2.3.3 Selected RWR-l D Capacity 11
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2.4 RWR-l D System Performance and Operating Conditions 12 2.5 Seismic Classification of the RWR-1 Systen 13 2.6 Quality Group Classification 14 2.6.1 Scope 14
, , 2.6.2 Description of Quality Control System 15 V 2.7 Compliance with Federal Regulations 19 2.7.1 10CFR20 - Standards for Protection Against 19 Radiation 2.7.2 10CFR50 - Licensing of Production and 21 Utilization Facilities 2.7.3 10CFR71-Packaging of Radioactive Material 23 For Transport and Transportation of Radioactive Material Under Certain Conditions 11
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2.7.4 NRC Regulatory Guides 23 3.0 SYSTEM DESCRIPTION 46 3.1 General 46 3.2 RWR-1 Systm Procest 9escription 47 3.2.1 Incineration and Calcination 47' 3.2.2 Off-gas cleanup 49 3.3 RWR-1 3 Components Description 52 3.3.1 Feed System 52
; 3.3.2 Incinerator /Calciner 54 3.3.3 Off-gas Cleanup Syst m 54 .. 3.3.4 Component Design 56 3.4 Off-gas Cleanup System 56 3.5 RWR-1 Syste Instruentation and Controls 57 3.5.1 Reliability Effects 57 .
3.5.2 Instrumentation Types 58 3.5.3 Control System Description 58 3.6 Material Selection 61 3.7 System / Plant Interfaces - 63 3.8 RWR-l D Syst m Operation 65 3.8.1 Introduction 65 3.8.2 Cm6ustible Waste Incineration Mode 65 3.8.3 Resin / Sludge Incineration Mode 65 ( 3.8.4 Liquid Waste Calcination Mode 66 3.8.5 Off-gas Systa 66 3.9 RWR-1 3 System Reliability 66 3 3.10 RWR-l D System Internal Decontamination 67 3.11 RWR-1 3 System Maintenance 67 4.0 ENVIRONMENTAL IMPACT ANALYSIS 79 4.1 General Background , 80 4.2 Decontamination Factors 80 4.2.1 Decontamination Factors for Particles 81 4.2.2 Decontamination Factors for Iodine 83 111
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4.2.3 Release Rates from Maximum Expected Operations 85 4.2.4 Dose Rates and Concentrations Resulting 85 from Maximum Expected Operations 4.3 Exposures from Anticipated Transients and Postulated 86 Accidents 4.3.1 Anticipated Transients 86 4.3.2 Postulated Accidents 86 4.3.3 Exposure at the Site Boundary from Anticipated 89 Transients and Postulated Accidents
- 5. 0 RWR-1 SYSTEM BENEFITS 100 I
.. , 5.1 On-Site Benefits 100 l l 5.2 Off-Site Benefits 100
6.0 REFERENCES
101 APPENDIX A 105 APPENDIX A REFERENCES 109 or i
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.- LIST OF FIGl!RES FIGUR!( TITLE PAGE 2-1 lion-Gaseous Wastes fra a BWR with Deep-Bed 40 Condensate Cleanup 2-2 Non-Gaseous Wastes fra a BWR with Powdered Resin 41 Condensate Cleanup 2-3 Non-Gaseous Wastes fra a PWR with Deep-Bed 42 Condensate Cleanup 2-4 Non-Gaseces Wastes from a PWR with Pcwdered Resin 43
, Condensate Cleanup 2-5 RWR-l* General Equipment Arrangement 44 3-1 Olagram of RWR-l D Process 76 3-2 R1iR-l
- General Piping and Instrumentation Diagram 77 O
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LIST OF TABLES TA8LE - TITLE P PAGE 5 2-1 ESTIMATED ACTIVITY, VOLUME AND SPECIFIC ACTIVITY
' 26 i 0F RADWASTE ,
' 2-2 ESTIMATED NUCLIDE CONTENT OF BWR SOLID WASTE 27 2-3 ESTIMATED NUCLIDE CONTENT OF PWR SOLIO WASTE 28 2-4 ESTIMATES OF SOLID RAUWASTE GENERATION RATES 29 J
2-5 SOLID RA0 WASTE SHIPMENT RATES FOR BWRS 30
' 2-6 SOLID RADWASTE SHIPMENT RATES FOR PWRS 32 e '2-7
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REACTOR DATA USED FOR OBTAINING THE AVERAGES FOR 34 CASE B 2-8 REACTOR DATA USED IN 08TAINING THE AVERAGES FOR 3S CASE C
' 2-9 NUCLIDE COMPOSITION OF TOTAL SOLID RADWASTE FOR 36 CASE E 4
2-10 NUCLIDE CGlPOSITION OF WET RADWASTE SHIPPED FROM 37 11 REACTORS 2-11 ESTIMATES OF SPECIFIC ACTIVITIES OF WET WASTES 38 AND RESINS 2-12 RWR-1 E CAPACITY ' ' 3 39 3-1 RWR-1 SYSTEM COMPONENTS 69 3-2 EQUIPMENT CODES 71
~ 3-3 CGlPONENT DESIGN DATA 3 72 3-4 RWR-1 MMOR COMPONENT MATERIALS LIST 74 3-5 INTERFACES OF RWR-1# SYSTEM AND PLANT 75 4-1 SOLID PARTICLE DECONTAMINATION FACTORS FOR WCF 91 4-2 DECONTAMINATION FACTORS FOR THE RWR-1* SYSTEM 92
.' 4-3 EMISSION RATES, BOUNDARY CONCENTRATIONS. AND DOSE 93 RATES FRGl NORMAL OPERATION
'- 4-4 ANTICIPATED TRANSIENTS AND THEIR CONSEQUENCES 95 4-5 RESINS AND TOTAL WET SOLIDS SHIPMENT RATES 98
'4-6 DOSES AT THE SITE BOUNDARY FROM THE POSTULATED 99 MAXIiG CREDIBLE ACCIDENT E
1.0 INTRODUCTION
This document is the Genaric Licensing Topical Report for the RWR-l E System. The RWR-l* Systen provides a new method of radweste management and disposal which results in improvements in the cost and effectiveness ' of radweste disposal for light water power reactors. The RWR-l D System has been designed and developed to be capable of dealing with the nuclear power industry's problems of increasing waste volumes, greater activity levels, and more stringent regulations regarding containerization, transportation, and disposal.
, _ The RWR-I D System is being developed jointly by Energy Incorporated and
- Newport News Industrial Corporation. The RWR-1 D System incorporates a
. proprietary design that combines a single-chamber fluidized-bed incinerator and calciner that can substantially reduce the volume of both liquid and solid radwestas such as concentrated chemical wastes, filter sludges, spent ion exchange resin beads, rags and other similar materials.
Operation of the RWR-1 D System reduces all liquid and combustible solid radweste to anhydrous granular solids. This residue is compatible with existing solidification agents as well as many agents currently under study. The RWR-1 D System capabilities result in several benefits for the 1 . nuclear power industry which include cost savings for containers, transportation, and disposal as well as an extension of available space at disposal sites. From huiiihn' engineering considerations, it results in increased safety in handling, transportation and disposal, and in reduced
'. radiation exposure to operating and maintenance personnel.
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,' ~ 1.1 Purpose .
The purpose of this Generic Lice ::ing Topical Report is to provide specific design bases, systsa descriptions, and an Environmental Impact Analysis of the RWR-1 D System to allow generic licensing of the systam by the Nuclear Regulatory Conssission. The RWR-1 System is capable of "RWR-1 1s a trademark of Newport News Industrial Corporation and Energy Incorpor" a. I
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meeting the current pubite and governmental standards on radwaste disposal methods and is capable of dealing with more stringent regulations regarding containerization, transportation, and disposal. ; 1.2 Scope ' This G6neric Licensing Topical Report deals with those overall issues of licensing a solid waste processing system which are not specific to the location of the system. Applications for specific plants shall be
, submitted dealing with plant-specific issues. The report is designed in such a mannte that plant-specific applications will only have to ascertain compliance to the generic design basis and confim the generic safety analyses.
1.3 Acolicability This report, being generic in nature, is intended to apply to all light water reactors. For this reason an extensive effort has been expended to develop source term design criteria for the RWR-1D System in such manner that all present and future designs of the light water reactor will generate radioactive waste at rates which will be less then those used for the present analysis of the RWR-1 System. Moreover, the D RWR-1 System is applicable to other foms of low-level and intermediate level wastes not necessarily generated by the LWR. However, for the purposes of the licensing effort, only the applicability to the LWR will
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be discussed in this generic report.
". 1.4 History sad Background , The basic processes of liquid calcination and combustible waste incineration which are used in the RWR-1 System have been used i_n industrial p1';nts for decades. Fluidized bed calcination of radioactive wastes w s developed during the period 1952-1959 at the Idaho National Engineering Laboratory. Use of calcination for ifquid radwaste reduction was first demonstrated in an engineering scale facility, the Waste Calcining l
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Facility (WCF), at the Idaho Chemical Processing Plant in 1963.(2) The successful operation of the WCF has demonstrated that liquid wastes can be routinely calcined into a granular free-flowing powder which can l subsequently be handled in a simplified manner. Since 1963, the WCF has ' handled over 2.5 million gaiS' of radim live aqueous wastes which were calcined to approximately C.,2 cubic feet of solids.(3) A batch operated fluidized bed calciner was designed and built as part of the Midwest Fuel Recovery Plant (MFRP) at Morris, Illinois, for General
'l ElectricCampany.I4) The plant did not go into operation, but this was not due to radioactive waste concerns. Another batch calcination process.
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the Potcal process, was demonstrated on a fully radioactive basis in the , WSEP program from 1966-1970. This process was developed at ORNL for the specific purpose of solidification of high level liquid wastes, although it did not use a fluidized bed process.54) Incineration of combustible radioactive wastes has been in use as a disposal technique since 1948 when a pilot plant incinerator and off-<jas cleanup system were built at Mound Laboratory.(5,6,7) Early systems were adaptations of standard refuse incinerators and did show that considerable volume reduction in waste handling was possible. Data taken in the early 1960's at the General Electric Atomic Power Equipment Department in San Jose, California, showed that s99% of the activity of the incinerated wastes remained in the ash.(5) Similar data are reported from an incinerator at Pratt and Whitney Aircraft where approximately S 99.1-99.98% of the activity remained in the ash.(6) Various methods have been utilized in off-gas treatment systems including .' . very simple dry scrubber systems and very complex arrangements having both wet and dry scrubbers, water sprayed filters, and HEPA filters. The RWR-1 D System has been designed to utilize the most proven components and a general arrangement that experience has shown to be effective and relatively maintenance free. 3
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- 2. 0 DESIGN BASES The design bases of,the RWR-1 System can be considered to be of three distinct types. The first type relates to system location and interfaces with the comunercial nuclear power installation. The second trpe relates to system operation, which includes not only types of wastes processed, but also expected volumes of wastes, flows of different categories of wastes, and the associated configuration and through-put. The third type is related to the legal regulations placed on the function, operation,
,, and hardware configuration, s 2.1 General s -
The RWR-1 Systen is designed to meet the needs of utilitias operating nuclear power reactors. Radwaste handling is becoming an ir: creasing burden at these installations, and a system which reliably and safely - reduces the volume of waste to be handled, shipped, and buried offers significant advantages over current practices. The RWR-1 D System is designed to reduce the volume of radwaste by incineration of the combustible waste and calcination of the liquid non-combustible waste. Spent fuel rods and hardware items such as broken control rods and contaminated tools are, of course, not' amenable to treatment by the RWR-1 System. Progress in fluidized-bed technology now allows both incineration and l calcination to take place in the same vessel. This is one of ti:e design bases of the RWR-1 System and is described in Section 3.2. The other j generic design bases relate to G types and amounts of waste which the system must handle and the requirements for safe and reliable operation.
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Insuring that personnel operating the RWR-l D System receive minimal
. occupational doses is also a primary consideration in design of this system. Finally, the system is designed to satisfy,all the applicable requirements and regulations of the U.S. NRC and other governmental agencies.
2.2 1.ight Water Reactor Wastes Which Are Volume Reducible Data are available from operating light water reactor systems to characterize the types, volumes, and activities of radweste which is volume reducible. 2.2.1 Types of Waste The solid waste generated at a nuclear power plant has been classified as" wet"wasteor" dry" waste.(8) The " wet" wastes result from treatment s processes which remove radioactive contaminants in cooling system and fuel storage pool water, from decontamination, and from other sources of ( contaminated water. This type of waste generally contains at least some water and consists mainly of spent domineralizer resins, evaporator bottoms and filter sludges. The " dry" waste consists of ventilation air filters, contaminated clothing, cleaning swipes, paper, miscellaneous hardware and laboratory wastes. This type of waste normally has a much lowr specific activity than the wet waste and it is generally combustible, if hardware items are removed. Hardware items, such as contaminated equipment, tools, and other metal itens are generally excluded from this discussion.
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The methods of processing these wastes with the RWR-1 System are discussed in Section 3.2 of this report. 2.2.2 Maximum and Expected Volumes and Activities of Waste The characteristics of radioactive waste from light water power reactors l, can be described within reasonable generic ranges, although they vary between plants. Volumes and radioactivity content of radwasta have been
. estimated by two recent reports on radioactive wastes: WASH-1258(8) and ERDA-76-43.OI Table 2-1 summarizes the volumes and maximian specific activities estimated by WASH-1258 for a typical 3500 Nt BWR and for a typical 3500 MWt PWR. The BWR volumes apply to case 3 in WASH-1258 4
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and 6 in WASH-1258 and are the largest volumes estimated. Tables 2-2 and 2-3 contain estimates of solid waste activity, Ifsted by nuclide. WASH-1258 assumed 180 days decay prior to shipment, which allowed many of the short-lived radionuclides to decay away. With the D RWR-1 System, storage of radweste for_a period on the order of a year before processing is not likely, so the 180 day decay period is not appropriate. For the purposes cf this report, it is assumed that the
. radweste is processed by the RWR-l* System every four weeks, so that the average decay period before processing would be 14 days.
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The final column in both Tables 2-2 and 2-3 contains the activity for each nuclide adjusted for 14 Jays decay. This adjustment is easily made since radioactive decay is descrited by an exponential relation: A180
- A14 exp[-1(180 - 14)]
where - A180 = generation rate with 180 days decay (Curies / year) A34 = generationratewith14daysdecay(Curies / year)
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I" A = = decay constant (1/ days)
,' HL = half life (days)
, . As an example, consider Sr-89' in Table 2-2: A180 = 49 C1/yr. HL = 52.7 days, and Aj4 = expi,(-in(2 /52.7 (180 - 14)J = 435 Cf/yr.
ERDA-76-43 I9) distinguishes between reactors with deep-bed condensate clean up and those with powdered resin condensate cleanup. The ERDA-76-43 estimates are given in Figures 2-1 through 2-4. The WASH-1258 estimates are for a 3500 megawatt thermal (MWt) reactor while those of ERDA-76-43 6
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are for a 1000 megawatt electric (m e) reactor. Since a 1000 W e plant is approximately equal to 3200 Nt, the values from ERDA-76-43 must be multiplied by a factor of 35/32 before a valid comparison can be made. WASH-1258 provides an analysis by nuclide but ERDA-76-43 does not. Therefore, only the data from WASH-1258 are used in this section. Radweste generation rates throughout this report are nomalized to 3500 MWt. For example, if an 1825 Mt plant shipped 1036 Curies in a certain year, the rate is given as 1986 C1/yr = 1036 (3500/1825).
.,, Five different cases for radweste generation rates are considered herein.
They are listed in Table 2-4. Case A is based upon WASH-1258, adjusted for 14 days decay. Cases B, C and D are based upon a survey of the
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semi-annual operating reports of nuclear power reactors. The data through the end of 1972 has been analyzed in ORNL-4924.0 0) This early data was not particularly helpful in the preparation of this report since the 1500 MWt to 3300 MWt reactors, which comprise almost all the reactors currently operating, had accumulated very little operating time by the end of 1972. Therefore, cases B, C and D are based upon radwaste shipment data primarily from 1975 and the first half of 1976. Tim ' material available on microfiche from the U.S. NRC Public Document Room was examined for all operating coussercial reactors in the U.S. from the beginning of 1975 onward. Data for the second half of 1976 was generally not available, and some data for 1975 was also unavailable. For Monticello,
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Nine Mile Point, Oyster Creek, Pilgrim and Point Beach, which all had relatively high shipment rates, data for a year or two prior to 1975 was obtained to insure that the shipment rates for 1975 were not atypical. i This radweste shipment data is suser.arized in Tables 2-5 and 2-6. These tables contain data from only those reactors which had received their operating licenses before 1975. Examination of semi-annual reports shows that the radwaste shipment rate increases with time during the first few years of the reactor's operation.- Thus, the data from the first year or two of operation was not considered to be significant. For Case B, all the reactors in Tables 2-5 and 2-6 for which there was data are considered, with three exceptions (see Table 2-7). The exceptions l
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are Big Rock Point, Humboldt, and Lacrosse. These three reactors are relatively old, are less than 250 st, and were considered to be a special case. They are listed in Tables 2-5 and 2-6 for completeness, but are not considered further. Because most reactors ship much less solid radwaste during their first year or two of operation, Case C excludes those reactors which received their operating licenses in 1973 or later. The reactors considered in constructing the averages for Case C are listed in Table 2-8. The Surry and Turkey Point reactors are borderline cases, which might well have been included. However, the data concerning volume shipped from Surry appears te be inconsistent for May 1975, so it was decided to exclude
- the data from all four reactors. Had the data Surry 1 and 2 been included, the activity rate would have been slightly higher (1830 Cf/yr), but if the data from Surry 1 and 2 and the data from Turkey Point 3 and 4 had been included, the rate would have been considerably lower (1600 Ci/yr).
A comparison of cases B and C in Table 2-4 indicates that excluding data i from the newer reactors results in a significant increase in the average amount of radwaste shipped. Case D is the maximian amount for each category, as shown by Tables 2-5 and 2-6. The reactor site for each category is given in Table 2-4 A comparison of Cases B, C and D with Case A indicates that the volume' estimates of WASH-1258 are rather low. This may be due to the fact that the data upon which WASH-1258 is based was preliminary in some respects, !. or it may be due to radwaste practices which differ from those assumed l* in WASH-1258. These reasons, plus the variability in cladding integrity, ! may explain why WASH-1258 appears to have underestimated the activity in
.' BWR radweste and overestimated the activity in PWR radwaste. In comparison to both the WASH-1258 and the ERDA-76-43 estimates, the PWRs have shipped considerably less radwaste, both in volume and in activity, than the BWRs. It should be noted that Case A assumes a unifonn 14-day delay between time of production and the time of assessment. On the other hand, cases B, C and D are not generation rates but shipment rates from the site, and the delay between production and shipment is not known.
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Case E has been selected as a worst case for use in computing site boundary doses and concentrations resulting from normal operations. The highest annual rate of activity generation, 9500 Ci/yr, has been taken
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from Cases A, B, C and D and applied to both BWRs and PWRs. The volume rate is the maximum for each type of reactor. -l The miclide composition of che 9500 Cf/yr is given in Table 2-9. It represents a compromise between the nuclide composition of WASH-1258 (Tables 2-2 and 2-3) and the reported compositions of actual shipments (Table 2-10). Unfortunately, the BWR (Monticello) and the PWR (Point Beach) which shipped the most activity did not report solid radwaste composition by nuclide in their semi-az.inual reports. Table 2-10 indicates that the nuclide composition of solid radwaste varies considerably from reactor site to reactor site. The volumes shown for Case E are worst case (maximum) volumes. For a given total activity, the minimum voltme will obviously correspond to the maximum specific activity. Thus, Case E is not the worst case from ! the standpoint of specific activity. The specific activity is not so important for the solid radwaste as a whole as it is for each individual type of weste, so maximum specific activities are considered for each type of radwaste in Section*4.3. 2.2.3 Specific Acc1vities Specific activities .are given in Tables 2-5 and 2-6. Some sites listed wet and dry wastes separately, and some identified only resin shipments. l' while others did not identify the specific type of waste at all. Therefore, the survey of the specific activies of radwaste shipped from operating
.' reactors was necessarily incomplete. The highest specific activities are found in resins and other types of wet waste. The information for resins and wet waste is stamarized in Table 2-11. Shipments with specific
- activities greater than 1000 Ci/m3 do not constitute a great deal of l
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l voltme. The Pilgrim shipment with a specific activity of at 1434 Cf/m3 consisted of only 0.38 m3 (13.6 ft 3) of sludge containing 553 C1, and the Point Beach shipment with a specific activity of 1168 C1/m3 consisted ! of only 2.1 m3 (73 ft3 ) of resin containing 2415 C1. Therefore, it is ! not reasonable to consider that the feed to the RWR-1 System is comprised of material at a very high specific activity for more than a very short time, if at all, since different batches of the same type of waste may be mixed in the holding tank before inciner! tion. Further, if station radweste persennel know that a batch of waste has a high specific activity,
;. they may elect not to process it in the RWR-1M System in order to keep the surface radiation levels from the radwaste containers below some desirable level. Even if a batch of high specific activity waste were reduced in volume with the RWR-1 E System, it would be very unlikely that the residue fram this batch alone would fill up tne product container.
Since radweste with specific activity greater than a few hundred Curies per cubic meter comprises only a small portion of the solid radwasta, the material filling up the remainder of the product container would probably be fairly low in specific activity and the total activity in the container might well fall within acceptable limits. If a product container receives so much activity that it cannot be shipped as low-level waste after the usual storage period, it must either be stored for a longer period or shipped in a shielded cask. 2.3 RWR-1 D system Sizine 2.3.1 Introduction D The RWR-1 Systen has been designed so that it can process the wastes ~ from a typical reactor by operating less than 75% of the time on an around the clock basis. Extra capacity is included' to permit time for systen maintenance and to' ensure sufficient capacity for periods of abnormally high waste generation. This section discusses the basis for the selected throughput rates which in turn determine the size of the RWR-1 components. 10
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2.3.2 Waste Generation Rates The principle documents used in this report for determining the activities of the expected waste feed to the RWR-l E System were WASH-1258(8) and ERDA-76-43.UI The total waste volumes reported in WASH-1258 were 3 225-325 m /yr (for 3500 W t reactors, see Table 2-4). ERDA-76-43 reports somewhat higher volumes: 371-684 3m /yr (for 1000 We reactors, < see Figures 2-1 through 2-4). These " generic" estimates of volumes may be compared with the actual volumes shipped from reactor sites as listed in Tables 2-5 and 2-6. The rates at which volumes have been shipped 3 vary from 77 m /yr to 3538 3m /yr (nonnalized to 3500 Wt). It would not be prudent to base the size of the RWR-1
- System on a volume generation
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3 rate as high as 3000 m /yr. In the first place, few of the operating reactors are as large as 3500 E t. Secondly, plants which generated much smaller volumes of radwaste would be forced to consider systems which would stand idle too much of the time to be economically feasible. The ERDA-76-43 radwesta volume generation rates have been selected as the basis for detennining the size of the RWR-l D System. The volume through-put rates which result from designing the RWR-1 D System on this basis.are such that the majority of nuclear power plants can process their waste by operating less than three fourths of the hours in the year. 2.3.3 Selected RWR-1 E Capacity The tabulation of waste generation rates by type of waste in ERDA-76-43 l,' is adequate to detennine time of operation of the RWR-1 System required in each mode. The wastes from a BWR using deep-bed condensate cleanup
, represent the hiahest volume generation rate of the four cases presented
( in ERDA-76-43: I9 684 m 3
/yr. If the RWR-1 System is sized so that it 3
can process these 684 m /yr in 184 (24-hour) days,of operation (50% capacity), waste processing for the other three cases in ERDA-76-43 requires from 93 to 137 (24-hour) days per year as shown in Table 2-12. 11 4
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Two hundred and sixty-four days of operation per year is considered to be maximum for the RWR-1 System. The capacity of the RWR-l
- System 3
operating 264 days per year is 1230 m /yr for a PWR with deep-bed condensate cleanup. The maxistan operating capacities for the RWR-1D System with wastes from all four cases considered in ERDA-76-43 are given in Table 2-12. 2.4 RWR-l D System Perfonnance and Operating Conditions
, The RWR-1 System is designed to incinerate and calcine in the same single-chamber vessel with the same inert bed. It has been sized to
' incinerate 91 kilograms (200 pounds) per hour of combustible dry waste or 68 kilograms (150 pounds) per hour of resins and sludges, or to calcine 132 liters (35 gallons) per hour of liquid wasta. Processing reduces these wastes to a dry, granular solid product. An effluent containing a very small amount of radioactivity is released in gaseous form to the atmosphere. The resulting concentrations at the site boundary are shown in Section 4.2 to be much less than the limiting concentrations.
During incineration the fluidized bed operates at about 1000*C and during calcination it operates at about 400*C. The hot off-gas passes through the dry cyclone at temperatures slightly below the bed temperature. In the quench tank the off-gas is cooled to about 70*C. The off-gas remains between 40*C and 70*C throughout the remainder of the off-gas cleanup system. j The pressures in the incinerator /calciner and the off-gas cleanup system are maintained below ambient pressure in all operating modes. The
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pressures range from 29 kPa (4.2 psig) below ambient at the inlet to the off-gas blower (the exit of the off-gas cleanup system) to near at:nospheric in the fluidized bed. . The RWR-1 Systen volume reduction factors are estimated to be: 12
- - ..-- ..--. . T _ __ _r . -.: = 2 . = = = -
_ _ = __
Compacted dry combustible solids 80 Spent resin 18 Concentrated liquids 8 Filter sludge 5 Decontamination factors used for computing emissions are 4 x ld for 4 particles and 1 x 10 for fodine. Estimates of the decontamination factors expected to be observed in operation are considerably higher than these very conservative values used for computing emissions. i 2.5 Seismic Classification of the RWR-l* System The RWR-1 System will be designed, fabricated, erected, and tested in such a manner that it will meet the seismic design standards of the U.S. Nuclear Regulatory Commission. The RWR-1 D System is capable of satisfying all the applicable regulations and guidelines for radioactive waste manager : systems located in light-water-cooled nuclear power plants. A primary dssign basis for the RWR-l D System is to minimize the exposure of operating personnel and the general public to radiation in the event of a natural disaster such as an earthquake. The U.S. NRC document ivhich detemines the seismic design classification is Effluent Treatment Systems Branch Technical Position ETSB No.11-1
-, (Rev.1).I") ETSB No.11-1 states in Part B. Sections I and III, that
" Equipment and components used to collect, process or store liquid (or solid) radioactive wasta need not be designed to the seismic criteria
,- given in Section V." Section II of Part B requires that "Those portions of the gaseous radweste treatment system which by design are intended to
' store or islay the release of gaseous radioactive waste, including portions of structures housing these systems should be designed to the seismic design criteria given in Section V." While the RWR-l D System will handle radioactive waste in solid, ifquid, and gaseous fom, storage vessels will contain only liquids and solids. There is no gaseous holdup or storage capability in the system. Therdare, the equipment and components of the RWR-1 System will not be designed to seismic criteria.
13
- :---.=m---=---- --
~~ ~ ~ ~ ~ ~ ~ ~ ~-
, 1 ETS8 No.11-1, Part 8, Sections I and III do require that "The foundations and adjacent walls of structures that house the liquid or solid radwaste
_ system should be designed to the seismic criteria described in Section V to a height sufficient to contain the liquid inventory in the building." , Section V, Paragraph b defines the seismic design requirements for i buildings housing radweste systems. However, Section V, Paragraph c l allows an optional construction method:
"In liiu of the requirements and procedures defined above, optional shield structures cor. tructed around and supporting the radweste systems may be erected to protect the radweste systems from effects
. of housing structural failure. If this option is adopted, the
' ~
procedures described in Section V.h only need to be applied to the shield structures while treating the rest of the housing structures as non-seismic Category I." In conclusion, the RWR-1 D System will not be designed to seismic criteria; however, the foundations and adjacent walls that house the RWR-1D system must be designed and constructed to the seismic criteria given in Section V.b of ETSB 11-1 (Rev.1) to a height sufficient to contain the liquid inventory of the storage tanks within the building, or optional
~
shield structures surroun' ding and supporting the RWR-1D System must be erected to protect the system from housing structural failure as described in Section V.c of ETSB 11-1 (Rev.1). , 2.6 Quality Group Classification In accordance with NRC Branch Technical Position ETSB No.11-1 (Rev.1)(II)
~ the following Quality Assurance Program will be implemented for the RWR-1 N System. '
2.6.1 Scooe Design, fabrication, inspection, and testing of the RWR-1 System will be accomplished in accordance with the codes and standards listed in 14
l
. l
. Table 1 of ETSB 11-1. To assure that all requirements promulagated for the system are met, a quality assurance system has been established in accordance with SE Section VIII, Division 1.02) This system is maintained by each department assumin.g responsibility for quality -
subject to control and audit by quality assurance personnel. 2.6.2 Description of Quality Control System Operation of the quality control system for the RWR-1 E System is the
. responsibility of the Quality Assurance Department.
2.6.2.1 Authority and Responsibility The Quality Assurance Manager has overall responsibility for coordination and isolamentation of the QA system. He has the authority to identify and resolve quality control problems. He is independent of any other activity and reports directly to the Vice-President, Engineering. 2.6.2.2 Organization An organization chart which shows relationships between management and l the Quality Assurance Department and all areas concerned with the . quality 1 l system is a. part of the detailed quality assurance procedure. Departments ( _ implementing the quality system are Purchasing, Engineering, Design, Manufacturing, Field Installation, and Inspection and Testing. 2.6.2.3 Drawings, Design Calculations and Specification Control g All design calculations, 'rocedures, p and drawings are subject to review I - by a person or persons other than the preparer. The reviewer verifies incorporation of connents by his signature. All documents must be approved and signed by manageme.t. I 15 l l 1 wm .--m- --e----- g-W#gnag - -vy'" ~
s
. The records control system precludes possession of an incorrect revision of internal documents or customer specifications by any individual. All recipients of documents are recorded and must verify, by their signature, receipt of each subsequent revision.
2.6.2.4 Material control Purchase orders are all subject to review by the Quality Assurance Capartment to ascertain incorporation of quality requirements. Material c certification and test reports are also reviewed by QA prior to issue of material. Material or items which require traceability are assigned a 7 unique number applied at receipt. All incoming material is inspected by k - a qualified inspector and inspection results are reported to Quality Assurance. . Material is stored and handled in a manner to prevent damage or deterioration. Special storage and handling instructions are furnished by Engineering when required. Cleaning and preservation is in accordance with requirements of design specifications. 2.6.2.5 Examination and Inspection Program m. Instructions on drawings and in procedures specify inspections, tests, and examinations to be perfomed and refer to detailed procedures which
,- must be followed in performance of inspections and examinations. All inspections, tests and examinations are documented by reports which are
~
reviewed by Quality Assurance. t . l
. 2.6.2.6 Correction of Nonconfomities ,
Nonconfomities are reported by inspectors using a Nonconfomity Report fom. These reports are submitted to the Engineering Department for resolution. If it is determined that evaluation by another department 16
is required, that department will be consulted by Engineering. Nonconformity Reports are reviewed by Quality Assurance to ascertain adequacy of the resolution and are signed by Inspection and Quality Assurance when all corrective action has been completed. Nonconfoming material is tagged to indicate that it is in a hold status and is physically isolated from satisfactory, in process material. 2.6.2.7 Calibration of Measurement and Test Equipment Controls Controls governing calibration examinations of measuring and test
~
. equipment are providad in a written procedure. All tools and equipment I
requiring calibration are periodically recalled, calibration is recorded, and itans are labeled with the next calibration due date. Calibration . is to standards traceable to the National Bureau of Standards. 2.6.2.8 Records Retention Quality documentation is maintained by the records contml center for a
, length of time determined by customer specifications. ; Documentation retained is that indicated in the next paragraph and as follows:
(1) Design Specifications (2) Design Calculations
. (3) Certified Materiai Test Reports (4) Radiographic Film .
(5) Heat Treatment Records 17
(6) Test Reports (7) Drawings (8) Data Reports - 2.6.2.9 Sample Foms Sample foms are included in the detailed quality assurance procedure with requirements for their use. These forms include but are not limited to the following: (1) Personnel and Procedure Qualification Forms (2) Inspection Reports (3) NDE Reports (4) Purchase Order Form (5) Job Order Form (6) Tags (7) Material Issue Records (8) Nonconformity Report (9) Documented Distribution Record 2.6.2.10 Shioment All shipments are inspected by Quality Assurance to ensure the adequacy of packaging and secureness. m - 13
2.7 Compliance With Federal Reculations i The RWR-1 System will be in compliance with federal regulations concerning protection of personnel against radiation and other technical - and legal licensing requirements for production and utilization facilities. These regulations are promulgated by the Nuclear Regulatory Comission and other agencies of the federal government. 2.7.1 10CFR20 - Standards for Protection Against Radiation The RWR-l E System design results in very low radiaticn levels. When RR -l
- is operated in accordance with prescribed procedures, at no time will the limits of 10CFR20 Appendix B be exceeded. The dose rates to personnel operating the RWR-l* System cannot be computed on a generic
. basis, but the individual cells formed by the concrete shield walls and
- the operation of the System at less than atmospheric pressure facilitate keeping the operational dose rate below the levels required by 10CFR20.
The emissions from speration of the RWR-l* System result in concentrations
~
and dose rates at the site boundary which are well below the limiting values for concentrations and dose rates as set by 10CFR20 for unrestricted areas. The e ximum concentrations and dose rates are presented in
. Section 4.2 of this report.
2.7.1.1 10CFR20.101 - Exposure of Individuals to Radiation in Restricted Areas To allow compliance with this regulation, the components of the RWR-1 System shall be placed behind concrete walls to attenuate the radiation which my be mitted from these components. The modular design of the RW-l D System allows the components of the system to be placed in cells. The shield walls of the cells minimize the radiation exposure of operating personnel both during opera.lon at the remote console and during maintenance. Figure 2-5 shows a typical arrangement of the components and the shielding walls. 19
- . - v- : - . = - -
I 2.7.1.2 10CFR20.103 - Exposure of Individuals _to Concentrations of Radioactive Material in Restricted Areas Concentration limits for radioisotopes in restricted areas are listed in 10CFR20, Appendix 8. Table I. Compliance with these limits will be achieved by minimizing leakage rates fror.1 the RWR-l TM System to surrounding areas during operation. Storage hoppers, feed systems, product transfer mechanisms, etc., will be isolated by equipment air seals. In addition, the RWR-1 System is designed to operate at a negative pressure with y respect to its surroundings; thereby, further reducing the possibility of leakage of radioactive material to restricted areas. Area radiation
. monitors are intended to continuously monitor the room air.
L 2.7.1.3 10CFR20.105 - Permissible levels of Radiation in Unrestricted Areas This regulation specifies the following dose rate limits in unrestricted areas: (1) 2 millirem per hour, (2) 100 millirem in any seven consecutive days, and s (3) 0.5 rem per calendar year. The RWR-1 " System can be located and operated so that the dose rate in the closest unrestricted area will be less than these limits.
~
2.7.1.4 10CFR20.106 - Radioactivity in Effluents to dnrestricted Areas Radioactivity in effluents released to unrestricted areas from the operation and maintenance of the RNR-I IM System will be limited to concentrations "as low as is reasonably achievable." The concentrations at the site boundary will be below these specified in 10CFR20, Appendix 3, 20 -
-_ .._. 7 ,__._.
s - Table II. Section 4.2 of this report provides an analysis of the concentrations at a site boundary from the normal operation of the RWR-1* System. Concentrations due to anticipated transients and postulated accidents are considered in Section 4.3. These analyses confim this design basis. 2.7.2 10CFR50 - Licensing of Production and Utili:stion Facilities The RWR-1 System will be installed at a facility meeting the requirements of 10CFR50 and will itself meet all applicable requirements.
, 2.7.2.1 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants The System principle design criteria are listed in Section 2.0 of this document. Certain plant specific considerations will be covered on an individual basis by the plant owner.
2.7.2.1.1 Criterion 60 - Control of Releases of R::dioactive Materials to the Environment Tne concentrated liquid waste storage tank, the resin storage tank, the sludge storage tank, and the dry combustible waste hopper all are usually part of the reactor plant radwaste system. They shall be designed to accommodate anticipated operational occurrences, such as unscheduled desineralizer changeouts. The high treatment efficiency of the RWR-1 off-gas cleanup system minimizes the release of gaseous effluents to the ~ atmosphere. In case a portion of the off-gas cleanup system should fail, the RWR-1 System has the capability of recirculating the off-gas through the cleanup system instead of releasing it to the at:nosphere. The only significant solid eff".c9t is the product which is removed from the dry cyclone to the plarit radwaste system for imobilization or storage. Purged scrub liquid returns to the plant liquid racwaste system. Condensate frcm the condenser which is not needed to maintain the voltme of the scrub liquid is returned to the plant water system. 21
a No liquid releases fion the RWR-lE System directly to the environment are envisaged. The RWR-1 System includes an internal holdup tank for scrub liquid. 2.7.2.1.2 Criterion 63 - Monitoring Fuel and Waste Storage Appropriate instrumentation will be provided to detect conditions that may result in excessive radiation levels within the RWR-l TM System. The TM RWR-l System will be equipped with controls designed to sense and s activate an alarm upon the occurrence of a wide variety of off-normal operational conditions. A part of the controls will be an annunciator ,
,- - panel which will provide identification of the causes of any alarm. ,
(' i Corrective action will be taken either automatically or manually, depending i on the potential seriousness of the occurrence. 2.7.2.1.3 Criterion 64 - Monitorino Radioactivity Releases Off-gas from the RWR-l TM System is routed to the plant stack. Therefore, the plant monitoring system will be used to monitor these releases. In addition, a separate PWR-l TM System radioactivity monitor will be located in the off-gas exhaust line to the plant stack. 2.7.2.2 10CFP,50 Appendix B - Quality Assurance Criteria for Nuclear
, Power F ar.:s and Fuel Reprocessing Plants The RWR-1 System is cesigned to permit compliance with these criteria. . The quality assurance criteria for the RWR-1 System have been established in accordance with U.S. NRC ETSB 11-1 (Rev.1).UI) This quality t
assurance p v; ram is presented in Section 2.6,of this repo.-t. f i 22 l Y - 9%a yde
- e a -
. 2.7.2.3 10CFR50 Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation _to Meet the Criteria "As low As Is Reasonably Achievable" for Eadioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents ,
The RWR-l* System design results in very low releases to the atmosphere. The dose rates, as shown in Section 4.2, are well below, and should be maintainable below, the limits set in Appendix I. The radioactive effluents produced by the RWR-I D System during normal operations will
.. be so small that their addition to other effluents from a nuclear power plant should have no significant effect upon the ability of the plant to satisfy the requirements of Appendix I. Specific compliance will be addressed on a plant by plant basis.
2.7.3 10CFR71 - packaging of Radioactive Mate. rial For Transport and Transportatff n of Radioactive Material Under Certain Conditions It is expected that the solid residue (s), or product (s), of the RWR-lD System will be packaged and transported to a licensed disposal site. The responsibilities for this belong to the user (owner / licensee). The specific compliance shall be discussed on a plant by plant basis in individual applications. 2.7.4 NRC Regulatory Guides l 2.7.4.1 Regulatory Guide 1.21 - Measuring. Evaluating, and Reporting Radioactivit/ in Solid Wastes and Releases of Radioactive Materials in Liouid and Gasecus Effluents from Light-Water-
, Cooled Nuclear Power Plants l The specific compliance to this guide shall t:e di: cussed on a plant by plant basis in individua' applications. Provisions shall be made to monitor and to limit tne radiation from each package of solid waste in order to pernit the operator to control radiation exposure to personnel and to meet the regulatory requirements of 10CFR71 and of 49CFR.
l l 23 l
2.7.4.2 Regulatory Guide 1.110 - Cost-Benefit Analysis for Radwaste Systems for Lieht-Water-Cooled Nuclear Power Reactors l The RWR-l
- System is an addition to a nuclear power station radioactive -l waste handling system, the purpc5e of which is to reduce handling and disposition costs. An ecological benefit is the fact that the land required for isolation of the wastes is reduced by at least a factor of ten. The radiological dose effects resulting from the addition of RWR-l* to a. radweste facility are complex. The gaseous emissions to the atmosphere are insignificant, therefore no cost is realized from that source. A cost benefit results from the reduced number of shipments.
-( . . U.S. NRC Regulatory Guide 1.110 ser.s forth guidelines for preparation of cost / benefit analyses for radweste systems. This guide, however, does not land itself well to radweste systsss of this type which do not directly change the dose to the pcplation in the immediate vicinity of the station. As a result, no do'e s credits result from reduced radweste shipments and reduced repository land connitments. Direct costs include the fixed costs associated with purchasing the RWR-l* System and the construction of a building and support equipment. Any increase of mintenance costs as compared with the previous system must also be included.
" Benefit" is more difficult to quantify than cost. The most obvious item is reduced cost of operation. This results from the reduced annual values of radioactive waste that must be handled and shjpped and the reduction in space required for storage. A less tangible benefit of the RWR-l* System is reduced radiation exposure of operating personnel.
24
--_r.....-,..,.i._, _ . . . _ _- . - - - - - - . . . _ - . - _ _ _ ~ _ - _ . - - . - _
l 2.7.4.3 Reculatory Guide 8.8 - Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable The RWR-l E e System equipment design is consistent with the guidelines set forth. Means of interior surface decontamination will be provided via a separate spray system in larger components which handle radioactive materials. The overall design includes considerations of equipment location, shielding, personnel access, etc., which should pennit the
. operation and maintenance of the RWR-1 System with a minimum of occupational radiation exposure. The design criteria incorporated in the RWR-1 D
System will permit the operator to effectively implement the
' ~
philosoph of NRC Regulatory Guide 8.8. e e e 9 e 0 25
l l '. . TABLE 2-1 ESTIMATED ACTIVITY, VOLUE, AND SPECIFIC ACTIVITY OF RA0 WASTE ACCUMULATING IN ONE YEAR
- Type of Waste Characteristic Units BWR PWR 3
Volime m 166 125 3 ft 6000 4500 55-gal drums 800 600 ht 3 Specific C1/m <14 <75 3 Activity C1/ft <0.4 <2 3 Volume m 125 94 3 ft 4500 3300 55-gal drums 600 450 Ory 3 Activity Ci/m ".03 .03 3 Activity Cf/ft yj y; 3 Volume m 291 219 3 ft 10500 7800
, 55-gal drums 1400 ,
1050 Total Activity Curies 2900 9500
*These values are for a 3500 MWt reactor and are based on WASH-1258.(3)
The BWR volume estimates are for case 3; taken from Table 2-47, p. 2-157. The PWR volume estimates are for cases 4, 5, and 6; taken from Table 2-49, p. 2-159. The specific activity limits for wet wastes (as a whole) are from the same tables, and apply to all cases for the BWR and to all cases except Case I for the PWR. The total activity is from Tables 2-2 and 2-3 and has been adjusted for 14 days decay. 26
TA8LE 2-2 f ESTIMATED NUCLIDE CONTENT OF 8WR SOLID WASTE
- Generation Rate Generation Rate
, Cf/yr Cf/yr Nuclide Half Life (180 days Decay) (14 days Decay)
Sr-89 52.7 d 49 435(2) Sr-90 27.7 y 90 91 Zr-95 65 d 1.1 6.5
~ . .(
Ru-103 368 d 2.1 2.9 Te-127m 109 d 1.6 4.6 Cs-134 2.0 y 130 152 Cs-137 30.0 y 120 121 Ce-141 32.5 d 0.25 8.6 Ce-144 284 d 5.1 7.6 Cr-51 27.8 d 0.60 37.6 lei-54 303 d 13 19 , Fe-55 2.6 y 800 903 Fe-59 45.6 d 6.6 82.3 Co-58 71.4 d 130 651
~
Co-60 5.3 y 200 212 (no days decay) I-131 8.065 d 442 133 I-133 20.8 h 275 0.004
', 2867
*The values for 180 days decay are from WASH-1258,(8) Table 2 46, p. 2-156.
The values for 14 days decay have been computed 'from the 180 days decay . values explained in the text. The fodine values have been computed by the method explained in Appendix A. 27
TABLE 2-3
;4 ESTIMATED NUCLIOE CONTENT OF PWR SOLID WASTE
- Generation Rate Generation Rate -
Ci/yr Cf/yr Muclide Half Life (180 days decay) (14 days decay) i ^ Sr89 52.7 d 2.0 17.8(2)
~~~ ~'
~~Sr90~ 27.7 d 2.7 172 Zr-95 65 d 0.64 3.8
*s Ru-103 39.5 d 0.10 1.8 Ru-106 368 d 1.7 2.3
(, Te-127m 109 d 11 31.6
-.Te-129a 34.1 d 1.3 38.0 Cs-134 2.0 y 2800 3278 Cs-137 30.0 y 2700 2729
._.Ce-144 284 d 4.2 6.3 Cr-151 27.8 d 0.35 22.0 Mn-54 303 d 22 32.2 Fe-55 2.6 y 200 226 Fe-59 45.6 d , 1.8 22.4 Co-58 71.4 d' 120. 601 Co-60 5.3 y 280 297
(~- (no days decay) I-131 8.065 d 6803 2044 l I-133 20.8 h 1037 .0142 9525
*The values for 180 days decay are from WASH-1258,(8) Table 2-48, p. 2-158.
- The values for 14 days decay have been computed from the 180 days decay values as explained in the text. The iodine values have been computed by the method explained in :Nedix A.
- m. . --
28 l l
N . .! f . l TABLE 2-4 ! l ESTIMATES OF SOLID RA0 WASTE GENERATION RATES NORMALIZED TO 3500! W t i Generation Rates Generation Rates Case Description from a 3500 W t SWR from a 3500 W t PWR t 3 Cl/yr m /yr C1/yr m3 /yr 1 A Values from WASH-1258 corrected 2875 375 9500 225 for 14-day rather than 180-day decay. See Tables 2-1, 2-2, and 2-3, 1 1 8 Average of available data on radwaste 1750 1300 1150 750 shipped for all reactors except three early small reactors (Big Rock Point. Humboldt8ay,andLacrosse). See Tables 2-5, 2-6, and 2-7. \ U C Average of available data on radwaste 2875 1850 1825 750 i. shipped for all reactors which are ' greater than 500 W t and which had their operating license prior to 1973. The Surry and Turkey Point sites have been excluded because each second unit received its operating Ilcense in 1973. See Tables 2-5, 2-6, and 2-8. D Maximum site. For each type of Monticello Dresden Point Beach Yankee Rowe reactor and each category (activity 8575 3550 4950 1525 or volume) the site which ships the maximum has been chosen. See Tables 2-5 and 2-6. . E Selected worst case. See Table 2-7. 9500 3600 9500 1600
,. ?
I - 4 TA8tt 2-5 SOLIO RADWASTE SHIPl H T RATES FOR BWRs Specific Radweste Actigity Shipment 8 ate
- Cl/m- Time Perid N555 Date of 3 Resin Ilsed for Site Unit Supplier MJ.1. Op. L ic. Ct/yr m /yr Total Wet h Only feta Notes Arnold GE 1593 2-74 173 571 .30 1975 Big Rock Point GE 240 8-62 14,800 1975 No volumes given.
Browns Ferry 1 GE 3293 6-73 885 # density 15 For 2 GE 3293 6-74
- 71) 679 1*06 3.5 '10 55 1975
" hip,resir." only.
3 GE 3293 7-76 Unit 3 not considered for normalization. Brunswick 2 GE 2436 12-74 292 1489 .20 .30 .04 fjfneI fh6 Cooper 2831 1.60 .01 GE 1-74 550 407 1.35 dj f9f6 Dresden I GE 700 9-59 Jan.1975 Extraordinary shipment of 2 GE 2527 12-69 658 3538 .19 .34 .05 to 6300 Cl in August 1975 has 3 GE 2527 1-71 June 1976 been excluded. Fitzpatrick GE 2436 10-74 Zu 835 .29 .34 .03 fj If9f6 liatch GE , 2436 8-74 208 329 .63 .86 .02 fby f6 ilumleidt Bay CE 220 8-62 686 2014 .34 .08 .45 Millstone 1 GE 2011 10-70 3669 2645 1.39 3.51 .02 Jjan J 96 Mor.ticello GE 1670 9-70 8578 700 12.25 fj"Ef6 II3I Nine Mlle Point GE 1850 8-69 4848 906 5.35 7.62 .04 1974 - 1976 [2] Oyster Ereek GE 1930 4-69 2853 1869 1.53 104 f," 6 [3]
. -v , .*
'..j e
i TAett 2-5 (Contd.) Speelfle Radweste Actigity shipment Rate
- Ci/o Time Period NSSS Bate of 3 Resin used for Site unit Supplier Mit op. L ic. Cl/yr m /yr Total j We h Only Data Notes Psach Botton 1 GE 3293 8 73 2 GE 3293 7-74 115 310 .37 1975 Pilgria GE 1998 6-72 5940 900 6.6 13.1 1.96 I ne 6 and Dr densit based onJuly1955-June 5976 ^
! Quad Cities 1 GE 2511 10-71 an. 1975 2 GE 2511 3-72 1840 906 2.0
-June 1976 Vermont Yankee GE 1593 3-72 77 880 .08 .23 .02 June - k . Only me Semi-annual
, 1975 Report found.
IIl Nnticello: If only July 1974 - June 1976 data are used, normaltred rates are 10355 Cl/yr and 729 m 3/ year. [2]MineNilePoint: If 1970 - 1976 data are used, normalized rates are only 2474 Cl/yr and 773 m 3 / year, but operating licence was received in August 1969. I3}0ysterCreek: If only January 1975 - June 1976 data are used, normalized rates are 3647 Cl/yr. and 2213 m /yr. I*I Pilgrie: If only July 1975 - June 1976 data are used porealized rates are 13604 Cl/yr and 1633 m3 /yr.
'Non.allred to 3500 MWt.
l l l
s r TAatt 2-6 50 LID RA0 WASTE SHIPIENT RATES FOR PWRs specific Radweste Activity shipment Rate
- Cl/m3 Time Period NSSS Date of 3 Re'iTn used for Site unjt Supplier MWt Op. L ic. Cifyr, m /yr Total Wet Dry Only Data Notes Arkansas B&W 2568 5-74
~
No information found. Calvert Cliffs 1 Cf ' 2560 7-74 Does not appear to have 2 CE 2560 8-76 shipped anything yet. Conn Yankee W 1825 6-67 1986 1333 1.49 12.3 .01 1975. 1976 Cook W 3250 10-74 No infonnation found. Fcrt Calhoun CE 1420 5-73 132 1046 .12 .106 .16 Jan 1975-m June 1976 Ginna W 1520 9-69 267 963 .28 Indian Point i B&W 615 3-62 2 W 2758 10-71 2080 645 3.22 4.69 .04 1975 3 W 3025 12-75 Unit 3 not considered for normalization. Kawaunee W - 1650 12-73 Very few shipments so far. La Crosse A-C 165 7-67 No data. m ine Yankee CE 2440 9-72 1892 309 6.12 9.63 .13 ,"ne 6 Millstone 2560 8-15 J8 ** 2 CE 3 550 .005 Oconee I B&W 2568 2-73 2 B&W 2568 10-68 865 771 1.12 120 Jan.-June 1975 3 B&W 2568 7-74 Jan.-June 1976 Palisades CE 2000 3-71 259 1276 .20 4 1975
~
TABLE 2-6 (Contd.) Specific Radweste Activity Shipment Rate
- C1/m3 ilme Period N555 Date of Resin used for Unit 3
$lte Supplier M Op. L ic. C1/yr m /yr Total jWe h Only Data flotes Point Beach I W 1518 10-70 2 4927 235 21.0 343 Jan. 1974 W 1518 11-71 -July 1976 Excluding July-Dec. 1975 Prairle Island I W 1650 8-73 Jan.1975 2 W 1650 10-74 78 248 .32
-June 1976 Rancho Seco B&W 2772 8-74 No data found.
RIbinson 2 W 2200 7-70 2.8 J' I 1446 521 4.8 .25 M h6 Sea Onofre W 1347 3-67 117 338 .35 1975
""Y f f,h 1895 1975 Suspicious volume data.
Ihree Nile Island B&W 2535 4-74 356 632 .56 1975 Turkey Point 3 W '
' 2200 7-72 4 W 2200 108 794 *14 Jan. 1975 4-73 -June 1976 Yantte.Rowe W 600 7-60 19 1534 .01 .02 .002 1975 Zion 1 W 2760 4-73 2 W 2760 25 1075 *02 Jan. 1975 11-73 -June 1976 *Normaltzed to 3500 MWt.
?
o s'
- TABLE 2-7 I
REACTOR DATA USED IN 08TAINING INE AVERAGE FOR CASE 8 (SEE TA8LE 2-4) ! BWR PWR Arnold Conr.ecticut Yankee Browns Ferry 1 & 2 Fort Calhoun Brunswick 2 Ginna Cooper Indian Point 1 & 2 Dresden 1, 2 & 3 Maine Yankee Fitzpatrick Oconee 1, 2 & 3
. Hatch Palisades
~
Millstone 1 Point Beach 1 & 2-Monticello Prairie Island 1 & 2 Nine Mile Point
- Robinson 2 Oyster Creek San Onofre Peach Botta 1 & 2 *Surry 1 & 2 Pilgrim Three Mile Island Quad Cities 1 & 2 Turkey Point 3 & 4 Verinont Yankee Yankee - Rowe Zion 1 & 2 20 Units .
*24 Units
*0ue to apparently inconsistent data for the volumes shipped from Surry ,
the Surry data has been used only for computing the average activity shipped.
~
e h 9 34
l TABLE 2-8 REACTOR DATA USED IN O8TAINING THE AVERAGES FOR CASE C (SEE TABLE 2-4) 8WR PWR
'l Dresden 1, 2 & 3 Connecticut Yankee l Millstone 1 Ginna Monticello Indian Point 1 & 2 Nine Mile point Maine Yankee Oyster Creek Palisades Pilgrim Point Beach 1 & 2 ouad Cities 1 & 2 Robinson 2 Verimont Yankee San Onofre 11 Units Yankee - Rowe 11 Units O
O
, g l
e O 35
. e 8 '
e e 0 TA8LE 2-9 NUCLIDE COWOSITION OF TOTAL SOLID RA0 WASTE FOR CASE E
- ~ ~
(SEETA8LE2-4): A SELECTED WORST CASE FOR A
~
3500 Et REACTOR 8WR PWR C@I, 1 Ci/yr 1 CR-51 2000 21.1 100 1.1 m-54 800 8.4 400 4.2 FE-55 100 1.1 100 1.1
.FE-59 300 3.2 100 1.1
~
CO-58 1800 18.9 2700 28.4 CD-60 1300 13.7 1800 18.9 ZN-65 400 4.2 100 1.1 SR-89 80 0.8 20 0.2 SR-90 20 0.2 80 0.8 I-131 300 3.2 2000 21.1 CS-134 700 7.4 700 7.4 C5-137 1400 14.7 1300 13.7 CE-141 300 3.2 100 1.I TOTAL 9500 9500 ss.. n 6 e h 36 _ - _ - _ _ - - _ . . . ._ _ _ _ _ . ..___________._.-.__1
i l
! , i i
i , i l i . i l t . i TABLE 2-10 NUCLIDE COMP 051 TION (5) 0F WET RADWASTE SHIPPED FROM ELEVEN NUCLEAR REACTOR SITE 5*
- BWRs PWRs lstone Nine Mlle Vermont Conn Indian Maine Prairie Nuclide Brunswick Cooper Hatch Ml)fl Point Pilgrim Point j
Vankee Yankee Yankee Island Robinson CR-51 37 39 40 17 MM-54 5 10 31 3 16 2 9 3 IE-59 3 C0-58 35 20 10 22 20 64 11 CO-60 4 20 Both Co=31 w 49 22 52 29 80 4 7 22 M ZN-65 2 1 27 24 TC-99N 5 C5-134 6 25 6 5 29 28 18 C5-137 9 47
, , 15 14 5 43 45 24 Cf-141 8
*Ihe data is for the same periods shown in Tables 2-5 and 2 6.
total radweste as well (excluding spent fuel, control rods, strud gal components and stellar hardware itmas).Stace the bulk of the activity is in G
TA8LE 2-11 ESTIMATES OF SPECIFIC ACTIVITIES OF WET WASTES AND RESINS BWR 3
# 3
. _ . _ . . .. - .______ Description of Source Ci/m Ci/m _ Maximum for all wet wastes from 30 100 WASH-1258 Maximum for all wet wastes for 7.3 17 ERDA-76-43 (including non-combustible slurries) Maximum class of resin from 229 647 ERDA-76-43 Average wet waste from Tables 2-5 2.5 5.2 end 2-6 Average of resin shipments (site Oyster Creek Point Beach with maximum) 104 343 Maximum found for any ' Pilgrim Point Beach single shipment * (Clean-upSludge) (Resin) 1434 1168
*0nly those reactors which listed activity and volume for each shipment could be examined.
S 0 38
l i
.s i '
TA84E 2-12 - RWR CAPACITY
- 1 Days per year Waste Generation Required to TM RWR-1 Capacity Condensate Cleanup from ERDA-76-43 Process ERGA- at 264 day / year Reactor Type Method (m3/yr) 76-43 Wastes (m3/ yr)
BWR Deep Bed 684 t 184 980 BWR Powdered Resin 445 120 980 PWR Deep Ped 586 137
, 1230 PWR Powdered Resin 371 93 1050
*The volumes generated for the different ty Mrs of waste are those presented in ER04-76-43.I9) It is
! assumed that the RWR-IIM System operatas tiree shifts (24 hours) per day, and that the distribution l of tiie total volume among the various types of waste does not vary with the total volume. U$ f
PRIMMtY idASTE PER 11Me-year RA0!O. T WCIDEPACTED
- m. ,0. "r *c'U" rm -
4 POWDetED 2.8 (10 COMeusT. av l FILTER-DEMINERALIZER ; an:As , 1 l
=
l CORE
... . . . . - ... _ . u e FILTER-GENINERALIZER 2 SEAD RESINS 4.3 1000 COfeUST. av
. r CONDENSU h
BEAD RESIN 5 9.9 5.0 C0feUST. av DEMI ZER REGENERANT
. CHEMICALS I
Y
- - O!RTY PREC0AT DEEP-8ED e PURIFIED LIOUID LIWI W FILTER " UNE d DEM!N!UtALIZEu* SEAD RESINS 2.8 5.0 C:1MOUST. av W TES I
2 SLURRY 4 55 W COMRUST. O!550t.VED SOLIDS. MOSTLYNag 280 40 L!'lu:0 av
. r PREC0AT SLUDGE.
O!ATOMCEDUS EARTil OR CELLULO5!C FIBERS OR Pt3fDERED RES!NS 42 30 COMRUST. av L 'I O r T _. DEEP-8ED
- PURIFIED LIOU!O FILTER DEMINERALIZER
- BEAD RES!NS 5.7 10 COMBUST. av WASTES i
PREC0AT SLUDGE 57 10 COMnuST. my POWDERED RESIN
. MISCELLANE0uS DRY HEPA FILTERS. C0uen?!. iv W TES .
CHARCOAL. PLASTIC. arc PAPER. WOOD. METAL. M*J.Com%T.
. RU88ER 280 <5.0 sfitgo ,,
FIGURE: 2-l
- Mon-Geseous Vestes Fras a %R with Oeep-Sed Condensate C7eanup
* (This figure is dressi frzer EADA46-43 (9). Vol.1. Fig. 2.7. p. 2.22. The total weste for this case amounts to 684 m3 containing 1115 Ct.)
40 t __.__'_.___..__ El_ .E' [ _ _. __ ._ i 1. 2 _ '~~___^_ _ . .
1 I PRIMRY WASTE PER Gue-year UNCopFACTED RA0!0-FUEL VOLUME ACTIVITY 3 F0W4 W C1 DPE p .
- POWDERED FILTER.0EMINERALIZER 2 RESIN 2.0 <10 COMBUST!8LE sy CORE -
n R LTER e ! M IZ D ; POWDERED ' RESIN 4.8 1100 COMBUSTIR E sy
- CONDEN5ER h
FILTER. _ POWEDERED DEMINERALIZER ~ RE5!N 94 70 COMBUSTfR E ly DIRTY LIQUID CARTRIDGES 2.3 1.0 ComeusT;RE h WASTES 9 PREC0AT FILTER
- EVAPORATOR 4 DEEP-8ED
, LIQUID PUR!FIED 00t! net 4LIZER l
BEAD RESINS .28 1.0 CW SUSTIRE fy
$ LURRY 455
; DISSOLVED 50LIOS.
MOSTLY Na 2 50, .14 (10 NON-CupeU5 TIRE
; PREC0AT SLUDGE LITID sy ORkLU OR POWDERED RESIN b! 14 <5.0 C0peusT!8LE av T
D--- -e -e PURIFIED WASTES gg
-
- LIQUID l
; SEAD RE5!NS 5.7 <IO COMeUST!R E sy a RESIN SLU0r:E 42 <10
. COMBUSTIBLE av M!$CELLANEOUS - HEPA FILTERS. COM8uST!8tE !v m W TES - CHARCOAL A850R8ERS ar.d CLOTH!M. PLASTIC. '80N-COPCUSTISLE
. PAPER. W000. 'tETAL. 230 c5 a SOLIO av RUSOER 4
FIGURE: 2-2* h Ga*=aue Watt ** h a # 8 with M a**d 4**fa #= t-"ata
- LlaADuiL.
(This figure was drawn from ERDA-76-43 (9). Vol.1. Fig. 2.8. p. 2.s1. containing 1223 Cf.) The total weste for this case amounts to 445 e 41 i
--.m,- -_-- , -,.. -.P_ e#3._ _..__,.---w,.,_.w,.-r--.-,,.gme., _,,.e_
, , , -_ . , - - - , _ .-,,, ,gw,._ ~._ , _ . - - - . - _ - _ , , _
PRIMRY IdASTE Pst Gide-year FUEL _ UNCOWACTED RADIO-POIR. VOLLPE ACTIVITY FONI P>8 C1 TYPE ~ ~ DEEP-8ED ODt!N G ALIZH 800 RE5!NE 1.4 5.0 CCMUSTIK E sv i
- CARTR!DGE - FILTER FILTER ~ CARTRIDGES 14 5.0 COMBUST!KE sv
,, CORE 6 g CHEMICAL & VOLIME CONTROL SYSTDt; -e 80 0 RESINS 8.5 5500 C04U5T!8LE Sy CLEAN LIQUID idASTES
. FII.TER
=
% CARTRIDGES .28 500 CamusTIRE 8Y 3g ; EVAPORATOR SLURRY 5.7 5.0 NON T!KE lENHA M NmDT 7 : 80 0 RE5!NS 8.5 < 20 L TIRE 87 h I
- FILTER CARTRIDGES 2.8 10 COMBUST!8LE 87 CON 0EMSER h
DEEP-8ED 00t!N GALIZER 2 8G 0 RES!NS 9.9 <l.0 COMBUSTIBLE SY 6 REGENERANT CHEMICALS [ CARTRIDGES 2.8 1.0 COMBUSTIBLE sv OIRTY CARTRIDGE L!r1UID r FILTER r CARTRIDGES .!S 2.0 COMBUST!BLE av g h SLURRY d os O!550LVED 50LIOS, NON-C0f*SUSTISLF
" EVAPORATOR 4 5TLY H 80 310 <10 LIOU!D sy 3 3 - h
- PURIFIED LIQUID
. OEEP-BED DDt!NDALIZER ; BGO RESINS 5.7 < 1. tl COM8uSTICLE sy
. HEPA FILTERS, C M us!!htE 8v ORY WA5TES CHARCOAL, CLOTHIM and PLASTIC. PAPER. WGD NON combust!3LE METAL RUB 8ER 230 <5.0 50 LID 8v FIElRE: 2-3* %n/asseus Wastes 'm a M with Deeo-Bed Condensate CaenulL
* (This figure was drawn from ERDA-75-43 Vol.1. Fig. 2.10, p. 2.25. The total weste for this case amounts to 5861r3 containing 6065 Cf.)
42 i
PRIMARY niASTE PER Gue-year pg IRCoppACTED RADIO-POE VOLyME ACTIVITY FOWt 3*8 Cf TYPE DEEP-4E0 DDI!!IERALIZER 1 SEA 0 RESIIIS 'l 1.4 5.0 CIMOUSTIBLE sv
% CARTRIDGE ; FILTER FILTER CARTRIDGES .14 5.0 COMBUSTIBLE 8v CORE
! n p CHEMICAL & VOLtmE CollvlWL SYSTEM:
- I N W I"3 3.5 5500 CaptuSTIBLE sy CLEAll LIQUID nlASTES
- FILTER CARTR!DGES .28 500 COMBUSTIBLE sy i ~
p e EVAPORATOR SLURRY 5.7 5.0 NON-COMBUSTIBLE LIQUID 8v STEAM Mm L BEAD RESINS 8.5 (20 GUIERATOR COMeuSTIBLE ev TREAT 7EIT , 2 FILTER CARTRIDGES 2.8 <10 a COMeuSTIBLE av CON 00tSER h FIL M - m !N R IZR 5 POWDERED RESINS 93 5.0 COMBUSTIBLE sv
- CARTRIDGES 2.3 <.10
' COMeuSTI8LE !?
DIRTY ' L UU ;
; CARTRIDGES .28 2. 0 $0MOU5T!8LE 8v h
SLURRYs101 EVAPORATOR - DISSOLVED SOLIDS.
' MOSTLY H 30 17 2.0 g 3 3 NON-COMRUST!8LE LIOUID av OEEP 8ED PURIFIED LinUID DEMINERALIZER r SEAD REs!MS 1.4 <1 0.
COMBUSTIBLE 87 g,y HEPA FILTERS.
- CONBUSTIBLE sy W TES CHARCOAL, CLOTHING.
and PLASTIC. PAPER. WOOO. NON-COMBilST!BLE METAL,RU88ER 230
~
<5.0 SOLIO av FIGURE: 2-4*
Mon-Gassous Wattne % e e OWR wft$ su r +; #etie cane etate C 1.ane en ( This figure was dreist from ERDA 43. Vol.1. Fig. 2.9. p. 2.24. 371 m3 containing 6060 Cf.) The total weste for tais case amounts es 43 i p-,..7.- ,-.,--- - - ,,y..- - ~ . - ,,m.- y -,,y -
. . , - - , , - - . , . . - . . . _ [ .- *$ ,--.wm-i-w--- LL
1 l Figure 2-5 Sheet 1 General Equipment Arrangement This figure contains proprietary information 44
j Figure 2-5 Sheet 2 General iquipment Arrangement ,
.t A
This Figure contains Proprietary Infonnation 45 l l 1
*t'G9* * ^ - - -n+-~-,__. - - _ . - - - - - -
- l 3.0 SYSTEM DESCRIPTION 3.1 General The RWR-l* System is based on advanced fluidized bed technology using an inert bed media to incinerate and calcine within a single-chamber process vessel. The bed media has been demonstrated to be functional at temperatures required for incineration and calcination. The RWR-l D reduces the need to monitor particle size which is usually necessary in fluidized bed calcination processes. Supplemental energy required for wet solids incineration and calcination is provided by fuel oil burners.
Process off-gas leaving the incinerator /calciner vessel is cleaned in a mechanical dry cyclone and a wet scrubbing system. The wet scrubbing system is comprised of a spray quench tank, a high energy venturi scrubber followed by a wet cyclone, a condenser and a mist eliminator (demister). Gaseous fission products are adsorbed by the scrub liquid and a solid adsorbent bed. Particulate material is removed by the dry cyclone, wet scrub systen and HEPA filters. Cleaned off-gas is vented to the atmosphere while the product, a dry granular residue from the dry cyclone, is removed for storage and shipment as reduced volume radioactive waste. Scrub liquid is recycled through the liquid waste system. The RWR-l D System permits the processing of liquid and combustible wastes such as evaporator concentrates, shoe covers, wood, paper, plastics, mops, etc., as.well as spent ion exchange resins and filter sludges. Lubricating and transformer oils and other low level plant wastes can also be processed. The operating temperature and residence time in the bed assures complete destruction of all organic wastes fed to the unit. Three different operating. modes are utilized to process a range of waste types: liquid waste calcination, combustible waste ~ incineration, and ion exchange resin and filter sludge incineration. The incineration modes are separated due to differences in the control system action. Mode selection instrumentation is provided so that only one operating B 46
mode can occur at any one time. Each mode of operation has a separate feed system. When operating in a given mode, the other feed systems are rendered inoperable by interlocking controls. 3.2 RWR-l D System Process Description . This section of the Licensing Topical Report describes the processes which take place in the incinerator /calciner during the volume reduction of radioactive wastes and the processes which take place in the off-gas cleanup system during the removal of residue and during the cleaning of the off-gas stream before release to the atmosphere. The flow paths are shown in Figure 3-1. 3.2.1 Incineration and Calcination The processes upon which the RWR-1 System is based are incineration and calcination. Advances in fluidized-bed technology allow the incineration and calcination to take place in the same process vessel. The bed is composed of an inert heat-transfer medium which is identical for both the incineration and calcination modes of operation. The high heat capacity or thermal inertia of the bed gives the RWR-l D System high temperature stability. , In the incineration mode, the basic process is one of combustion. Almost all of the feed is non-radioactive organic material which is intimately mixed with trace quantities of radioactive molecules. Radioactive carbon and radioactive hydrogen are almost entirely absent in the feed, so if the organic molecules are broken down into non-radioactive carbon dioxide and water and allowed to pass off as harmless gases, then the
. bulk of the waste will have been removed leaving the radioactive material behind. The bulk of the activity in many types of waste are in nuclides of the elements such as strontium, managanese, iron', and cobalt which I
will fann oxides in the hot oxidizing atmosphere of the incinerator. These oxides are solids at the incineration temperature, so an efficient volume reduction process depends upon complete combustion and effective 47
separation of gases and solids in the effluent. The separation takes place in the off-gas system and is discussed in the next subsection. The fluidized-bed combustion process is a very efficient one. The constant agitation of the bed particles with the small pieces or droplets of the feed material results in a rapid rise of the waste material temperature. The air which maintains the bed in its fluid state provides an ample supply of oxygen, and complete combustion is enhanced by supplying over-fire air above the top of the bed. The rubbing action of the bed wears off the brittle oxidized surface which forms on the larger waste particles and virtually ensures that the center of particles do not remain unoxidized. This bed rubbing action also wears off any liquid coatings which may form on the bed particles. These coatings fom
- because wastes such as plastic melt before oxidizing. Some wastes such as sludges and slurries do not have sufficient caloric content to maintain the bed at the desired temperature. In this case additional heat is provided by the injection of fuel oil. The themal inertia of the bed material itself means that the system is relatively insensitive to variations in the exact caloric content of the feed.
i In the calcination mode, the basic process is one of evaporation or l , drying: heat is' used to drive off water as a vapor, leaving behind an ! incombustible residue. Spraying the liquid waste into the process
,, vessel creates droplets which are heated rapidly by their contact with the hot bed particles. As the temperature rises the water evaporates leaving dried ' waste material on the individual bed particles. This dried waste material is ground off the bed particles by the agitation of the bed and is elutriated from the process vessel to the dry cyclone.
ihe calcination process is endothemic, and heat is supplied by the combustion of fuel oil in the fluidized bed. The supply of heat by direct combustion has two advantages over the supply of heat by electric heater or heat exchangers: it eliminates the hot, metallic, heat transfer surfaces which are prone to fouling, and it easily accomodates any organic material which may be contained in the liquid waste. 48
.-_y -.. ., , _ , _ _ - . __ .- ..-,,-.y, . , , ' - , - - - -* - - . , - - . , . . ~ ^
, , . , . , i".*'"* - F* T t EPP F M - . - , - g g.- --
The use of an inert bed means that the bed material does not have to be changed when switching from incineration to calcination. The size of the bed particles is not greatly affected by contacts in the bed due to the hardness of the material fra which the bed particles are made and . the soft friable nature of the dry calcine. This means that the size distribution of the bed medium does not have to be monitored continuously. The R1R-1 E Syste is also largely independent of the chemical composition of the waste because of the inertness of the bed. This is important in l an application such as a radweste volume reduction system for a utility (- because the chemical composition of the waste is rarely known with any great accuracy.
- 3.2.2 Off-gas Cleanup The processes which take place in the off-gas cleanup system depend upon which part of the systs is being considered. All the processes are designed to reove material from the effluent of the incinerator / calciner.
All the off-gas cleanup system components except the adsorber are designed to reove particulate matter. The scrub liquid and the adsorber are designed to renove halogens because of the presence of radioactive iodine nuclides. The first itas in this system is the dry cyclone, which reoves particles
,. from the hot gas exhausted from the incinerator /calciner. These particles fall by gravity into the product container. The swirling motion of the air in a cyclone causes a centrifugal force to act upon the particles so that they migrate to the wall. The air velocity is lower close to the wall due to the boundary effect, so the particles slit.e downward along the wall to the particulate exit. The gas stream exits upward fram the
' top center of the cyclone.
Following the dry cyclone is the quench tank. The processes here are cooling and wetting. The off-gas passes upward through the tank while scrub liquid is sprayed in from nozzles near the top of the tank. The turbulent gas motion in the tank causes intimate contact between the gas 49
and liquid streams. One effect of this contact is the cooling of the gas; another is the wetting of most of the remaining particles in the off-gas stream. The larger droplets fall to the bottom of the tank and return to the scrub liquid tank, taking the wetted particles with them. Smaller droplets are swept out of the quench tank with the off-gas stream. Following the quench tank is a venturi scrubber, which wets those few particles which were not wetted in the quench tank. The process is one of condensation. The off-gas leaving the incinerator /calciner contains a great deal of water vapor. Passage through the quench tank added more water vapor and cooled the gas so that the off-gas is saturated or
- nearly saturated den it reaches the venturi scrubber. More scrub liquid is sprayed into the gas as it passes through the throat of venturi to ensure saturation. As the off-gas passes through the venturi throat, the pressure drops, which causes an increase in the amount of moisture which the gas can hold in vapor form, so evaporation occurs. As the gas enters the divergent section of the venturi, the velocity decreases and the pressure increases, resulting in condensation of water vapor. This condensation causes the existing droplets to get larger, but it also causes new droplets to form on the unwetted particles which serve as l condensation nuclei.
The purpose of the quench tank and the venturi scrubber is to wet as many of the particles as possible because it is easier to remove a liquid droplet than it is to remove a much smaller solid particle. The following pieces of equipment are designed to remove the droplets and the particles they contain. It is irrelevant whether the particle was composed of soluble notarial' and has gone into solution in the droplet or the particle was composed of insoluble material and is retained as a
, wetted solid within the droplet - in either case the particle moves with the droplet. In the process of creating these droplets, an inanense liquid surface area was created in the off-gas stream. The chemical additives to the scrub liquid have been designed to enhance gas liquid reactions which will draw gaseous halogens, especially iodine, from the 50
~
gas into the liquid. The large surface area due to the process of droplet formation creates the ideal conditions for these reactions to proceed rapidly. Once the iodines are captured they are retained in the solution long enough to decay to stable xenon. The first item which removes liquid droplets from the off-gas steam after the venturi scrubber is the wet cyclone. It operates on the same principles as the dry cyclone. The second droplet removal mechanism is a condenser, which is a shell and tube heat exchanger that cools the off-gas. The cooling here is substantial, and the droplets grow in size to the point where a significant amount of the water in the off-gas is removed by gravitational or momentum effects. In the gravitational l' ,, removal mechanism, the drops are so large thay they fall to the bottom of the vessel as rain, in the momenttui removal mechanism, the droplets are not large enough to fall out of the off-gas stream, but they are large enough so that when the gas changes direction suddenly, the droplets inginge upon the wall or other solid material which has caused the direction change. Following the condenser is a demister, which removes drops solely, by the momentian renoval effect. The gas is passed through a filter of woven j fibers and in doing so the. air molecules make many rapid turns. The liquid drops are too large to turn as sharply and collide with the filter fibers. The drop 1ets then run down the fibers to the wall of the desister and from there to the drain which returns the liquid to the scrub liquid system. ( Following the desister is a heater which heats the gas stream. The purpose of this is to evaporate any remaining droplets and so protect
.. the HEPA filter from plugging due to moisture overloading. The HEPA filter removes particles which have escaped unwetted or which were formed when the few droplets which were 'not removed in the demister were evaporated. The filter is composed of a medium with very small pores, and the particles are removed by impingement. Following the first HEPA is an fodine adsorber which removes halogens by adsorption. The halogen 51
, _ , . ,..,,.,e -w ~- ' " ~ ~ ~ ~ ^ ' ' ' ' ~ " "
v'- ' ' ' ' ' " ~ ' ' " ' ~ ~ ' ' ~ ^ ^ ' '
atoms are held on the surface cf the material by a chemical bond with the adsorbing agent and decay to stable xenon. A second HEPA filter. l following the adsorber, is purely a redundant safety item which will ; capture particles if a failure occurs in the first filter. l 3.3 RWt-1 3 Components Descriptions
~
The RWL-1 System can be divided into three major subsidiary systems as shown on Figure 3-1. These are the (1) feed (2) incineration / calcination and(3)off-gascleanupsystems. Table 3-1 lists the components which make up the complete RWL-1 System. Noted are those items which are
, always furnished and those which are optional. The standard system components are shown in Figure 3-2.
The feed system delivers radweste to the incinerator /calciner where the volume reduction takes place at elevated temperatures. Off-gas from the incinerator /calciner is scrubbed and filtered to renove pollutants before leaving the systen via the stack. Radioactive residues leave tM system as dry granular solids which are packaged for further processing or storage. 3.3.1 Feed System The feed system consists of three distinct subsystens, which supply the incinerator /calciner with three types of wasta: low-level combustible waste, spent resins and sludges, and liquid waste. 3.3.1.1 Combustible Feed Subsystem The combustible feed sub-system consists of a shredder, a storage hopper, a mechanical feed device and an 1. stion valve. The initial step in handling combustible waste is to reduce the size of the individual pieces of waste so that it can be transported into the incinerator efficiently. This is accomolished with a shredder located directly above the storage hopper so that the shredding and hopper loading operations 52 f y--r ----n- vemr-wv- ry'----n-n-ynew-g--n- tm-T-rm+, , ,- _ y-,-v,-, - - - - - , , -
U occur simultaneously with minism effort and exposure. The storage hopper is large enough to contain a month's supply of shredded wasta. It will be sealed after filling. The possibility of combustion products filtering back through the feed system is remote because the waste is injected into a region of the incinerator where the pressure is lower . than ambient pressure. A mechanical feed device will automatically meter waste from the hopper into the incinerator through an isolation valve, which is used to provide a positive air tight seal when the low level combustible feed system is not in use.
, '3.3.1.2 Resin Sludge Feed Subsystem The resins and sludges are fed to a mixing / dewatering tank, then through a metering device into the incinerator. The mixing / dewatering tank provides a continuous supply of waste material to the metering device.
Mechanical agitation of the wasta helps prevent bridging, compaction, or adhesicn to the tank walls. The water that results from dewatering can be returned to the slurry pumping, system as makeup water or pumped to the liquid wasta storage tank. A positive displacement , mechanical feed device meters the waste as it is injected into the incinerator. Compaction in this ' device causes a second dewatering cycle which further reduces the liquid content of the wasta. This system can be continuously operated since dewatering, metering, and injection are simultaneous operations. The rate of feed is automatically regulated.
- 3. 3.1. 3 Liouid Feed Subsystem -
The liquid waste feed subsysten consists of a paping/ metering / injection system. Contaminated liquid wastes will be pumped directly from plant storage facilities to the incinerator to allow a continuous calcining operation. The paping/ metering / injection systen consists of a pump, strainer, meter and atomizing nozzles. The metering device automatically regulates the liquid waste flow. The self-cleaning strainer minimizes l 53 l l i._-- __ _ __ ...-._ r r -- _ _ - . -. _ . . _ _ n-- o - - .c----------
the chances that injection lines and the nozzles will become plugged. The atomizing nozzles, which can be rootely cleaned, atomize the liquid waste as it is. injected into the incinerator. In addition, this feed system will be maintained at a high taperature to retard the p/ecipitation of dissolved solids from the waste stream. - 3.3.2 Incinerator /Calciner The major components of this subsidiary systs are the fluidized bed l - vessel, the fluidizing air blowers and air inlet nozzles, and the fuel t ~ oil pumps and nozzles. The fluidized bed vessel contains the inert bed medita and is the place in which the incineration and calcination processes ( , take place. The three feed subsystems inject the feed into this vessel for processing. Air from the fluidizing blowers entars the bottom of the vessel to keep the bed at its proper height and to supply oxygen for combustion. Additional air is injected above the bed to enhance complete combustion. The fluidized bed vessel is tall enough to provide adequate residence time above the bed for combustion to take place before the entrained particulate matter is exhausted into the dry cyclone. Hea: for endothermic reactions and for pre-heating the' bed is provided by the - combustion of liquid hydrocarbon fuel in an auxiliary burner in the bed.
. 3.3.3 Off-gas cleanup System The initial component of the off-gas cleanup system is the dry cyclone as shown in Figure 3-1. The dry cyclone renoves the bulk of the solid material in the off-gas and transfers it by gravity to the product container. The off-gas entrance is at the side near the top so that a circular swirling motion is imparted to the gas. The off-gas exits . through the top center of the cKlone and product exits out the bot *am.
The quench tank is a large vessel with provisions for gas entry at the bottom and exit at the top. Scrub liquid is sprayed'in thrcugh nozzles at the top of the tank and exits by gravity through a drain in the bottom. 54
The venturi scrutar is essentially a piece of pipe with a nozzle which causes faster flow and lower pressure in the throat. It contains ports for spraying scrub solution into the off-gas at the throat. The wet. cyclone is similar to the dry cyclone, but renoves liquid droplets instead of solid particles. The liquid runs down the side of the cyclone to a drain in the bottom. The condenser is a shell and tube heat exchanger. Cool water is run through the heat exchanger in order to reduce the off-gas temperaturs, The desister is a tank containing a fiber air straining device in which the gas passes through many layers of a material which has small pores. The HEPA filters are constructed of a material with very small openings where particles are renoved by impingenent. Each filter has a roughing and a finish section. The (' ,
'adsorber is a solid adsorption bed containing a medium onto which an adsorbing chemical has been deposited.
The wet scrubbing system fonus a major part of the off-gas system. It is composed of the scrub liquid system and the quench tank, venturi scrubber, wet cyclone, condenser and denister. The scrub liquid systas is composed of a scrub liquid holding tank, pumps, a heat exchanger, and the pipes and nozzles thoegh which the scrub liquid is injected into the quencr: tank and the venturi scrubber. Drain lines return the scrub liquid from the wet cyclone, condenser, and denister. Since much of the heat f' rom the combustion process in the incinerator /calciner is given up to the scrub liquid, scrub liquid is circulated through a heat exchanger to keep the temperature of the scrub solution from getting too high. Condensate from the denister can be returned to the scrub liquid tank to keep the voltase of the scrub liquid up if necessary, or it may be sent
. to the plant's condensate holdup tank. A side stream of scrub solution is recycled back to the liquid wasta holding tank or the sr.ation evaporators to be processed during the liquid waste calcination mode. This recycle prevents an infinite buildup of solids containing fission and waste
, products in the scrub liquid since solid:: are continually entering the scrub liquid as the wet scrubbing system renoves particulate material from the off-gas.
L 55 l l
The high-temperature process vessels, the incinerator /calciner vessel, the dry cyclone, and connecting pipes will be insulated to limit building air conditioning requirements. The incinerator and much of the off-gas cleanup system will be made of corrosion-resistent material as discussed in Section 3.6. - 3.3.4 Component Desian The RidR-1 3 System is classed as a radioactive waste system and as such !, will be designed within the guidelines of the USNRC Effluent Treatment System Branch Technical Position ETS8 No.11-1. (Rev.1) as discussed in Section 2.5 of this report. The recommendations from Table 1 of ETS8 l () ,
'No.11-1 are presented in Table 3-2.
The major system components as shown in Figure 3-1 are listed in Table 3-3. The operating conditions and volume are listed for each item. 3.4 Off-gas Cleanup System A large part of the RidR-1 System is devoted to cleaning the off-gas stream afte- f t leaves the incinerator /calciner. The components of this system have been described in Section 3.3, and the processes which take place in these components are discussed in Section 3.2. The components have been selected and designed to clean the off-gas in a highly efficient manner so that the emissions to the atmosphere are "as low as is reasonably achievable". Decontamination factors (DFs) are detemined in Section 4.2. Two values of each DF are given, a value which is expected when the RidR-l* System is in actual operation at a utility plant, and a lower, more conservative, value used in computing releases to the atmosphere. 4 The overall 0Fs used in calculating emissions are 4x10 for particulate 4 material ant' 1x10 for iodine. , 56
3.5 RM-1* Systa Instrumentation and Controls 3.5.1 Reliability Effects 3 The RW-1 Systa reliability is enhanced by an extensive instnamentation ' and control syste. This instrumentation and control system provides the operator information for process control and also for monitoring the performance charactaristics of the components. Monitoring the performance of the components allows the operator to anticipate many problens before
., a syste shutdown becomes necessary. For example, the scrub liquid strainer pressure differential is recorded in the control room. The
. . , , operator is able to observe the effects of plugging and then remotely
( , change to the redundant strainer before a condition such as loss of scrub liquid flow develops. Another feature which is included in the instrumentation and control syste is a two-level alarm and protective action. When a process parameter drifts outside of the normal operating band, the first indication is an alarm which notifies the operator of the probles. If corrective action is not made, a second alars, set slightly further outside the control band is actuated. This second alarm is accompanied by automatic protective action. By using this two-level action, many process problems can be corrected by the operator before protective action is initiated and possible process shutdown results. , u l Another feature of the RW-1 instraentation and control systen which l ,. leads to improved reliability is the roote control capability of the i systen. Process flows are automatically controlled so that operator's attention can be directed to the anWeipation of process excursions
' rather than maintenance of steady flow conditions. , All redundant equipment can be remotely valved in or out and remotely started. This feature ,
minimizes the operator actions away from the control room. 57 l l 1 l 1 _ __.__...... _ _ _. ~ __ . _ _ _ _. _ .._ . . , _ _ . -_ ___ _ . _ _ _ _ _ _ _ _ _ . . . -. _ ._
n , s. e , e l 3.5.2 Instrumentation Types The sensors used to monitor process taperatures and pressures are thermocouples and pressure transmitters which have well established reliability. The transmitters, controllers, and taperature recorders ' are all solid-state electrical devices of high reliability. Two types of automatic control valves are used - motor driven and pneumatic. Pneumatic valves are widely used for flow centrolling operations. The electrical signal from the automatic controllers is transmitted to the valve where an I/P transducer converts the electrical signal to the required pneumatic pressure range. This combination of electrical
, controllers and pnematic flow control valves takes advantage of the I. -
strengths of each component. Motor driven valves are used for automatic isolation valves where modulation is not required. The use of these electrically driven valves minimizes the problem of transmission of pneumatic air signals. 3.5.3 Control Systm Description The M -1 3 Systan's instrumentation and controls maintain process parameters within the limits which assure safe and efficient systen operation. The safe operation of the M-I D Syste is provided by control sequencing and interlocks which prevent improper operation and effect automatic syste shutsown if syste parameters are not maintained within prescribed limits. Alams are provided to alert the operator o'f abnomal conditions. The efficient operation of the RWR-1 Systen is ,. accomplished by control devices which maintain the system temperatures, flow rates, and pressures within prescribed limits. Figure 3-2 is a piping and instruentatEn diagram (P&ID) for the system. The following sections describe the protection features of the control syste. 1 58 e e ==
3.5.3.1 Incinerator /Calciner Instrumentation and Control 3.5.3.1.1 Vessel Instrumentation and control i The incinerator /calciner vessel is instrumented for the measurment of
.I temperature and pressure. Thermocouples or equivalent are located in the fluidized bed to measure and allow control of the bed tamperature and to activate the interlock system, if necessary. Should the bed temperature drop below the operating range, an alans is activated.
, Should the bed temperature continue to drop and approach a condition in which operation is detrimental to the process, a second alarm is sounded and auteutic corrective action is initiated. The same two level (alarm /
t . protective) feature applies to high temperature operation. Thennocouples in the vapor space above the fluidized bed are also equipped with two-level high taperature alarm / protection. A temperature logic system is incorporated to prevent spurious initiation of protective action. Pressure taps are located in the incinerator /calciner vessel which connect to both absolute and differential indicators. These taps are used to measure absolute pressure in the vessel, the fluidized bed density, and the fluidized bed pressure differential. The fluidized bed height can be derived from these measurenents. 3.5.3.1.2 Control During Dry Cambustible Waste Incineration l During the incineration of combustible wastes the fluidizing air and l ,. overfire air flows are set at predetermined levels to maintain optimum fluidization and combustion conditions required for normal expected combustible waste composition. Waste is fed at a controlled rate for these optimum conditions. Startup of the RWR-1 D System requires a definite sequence of events. The off-gas subsystem is placed in operation first. Only after the bed j has been preheated to a pre-set incineration temperature can the combustible j waste feeder be started. l l 59 i
- I l
i Failure of the fluidizing blower scrub liquid pump or off-gas blower i will result in an automatic shutdown of the system. 3.5.3.1.3 Control During Resin and Sludge Incineration Mode controls 1 Resin and sludge incineration is controlled in the same manner as combustible waste incineration. The operational difference is that bed temperature will be controlled by the fuel burner firing rate.
- 3. 5. 3.1. 4 Control nuring Liquid Waste Calcination The energy for liquid waste calcination is supplied by hydrocarbon fuel.
( . Neste flow is automatically controlled at a predetermined rate and fuel flow is automatically regulated to maintain the appropriate bed temperature. The high and low temperature protective features of the controls terminate the flow of liquid waste and fuel. System startup procedures require that the off-gas system be in operation and the bed preheated to the pre-set calcination temperature before liquid waste feed can be . initiated. A shutdown sequence initiates purging of the liquid waste feed lines with water before the fuel flow is stopped. This water purge insures against line plugging during or after the shutdown secuence. 3.5.3.2 Off-Gas System Instrumentation 3.5.3.2.1 Quench Tank 4 Scrub liquid flow to the quench tank nozzles is automatically controlled at a rate which is optism for efficient gas scrubbing and cooling. The tamparatures of the scrub solution and the gas leaving the quench tank are recorded. A liquid level indicator and alarm at the bottom of the tank aid in control of ifquid level. A high temperature sensor in the gas exit line is provided to indicate quench system malfunctions and initiate protective action. 60
3.5.3.2.2 Venturi Scrubber Scrub liquid flow to the venturi scrubber is automatically controlled to ensure optimum particle reoval. 3.5.3.2.3 Condenser-Demister The tamperature of ga: leaving the condenser is controlled by regulating the cooling water flow rate through the condenser. 3.5.3.2.4 Heater
's
- s. -
The heter power input will be regulated to maintain a constant differential ' tamperature. A temp 4 ature limit . witch insures that the heater does not heat the exit gas above the tanperature which could damage the HEPA filter. 3.5.3.2.5 Filters The roughing and HEPA filter system are equipped with pressure differential indicators to indicate the degree of filter loading. - 3.5.3.2.6 Off-Gas Blower Flow rate, tasperature, pressure, humidity, and radiation monitors detemine the conditfDn of the gas stream released to the stack. A
, sample station is provided for analytical determination of the composition j , of amitted gases. An alam and automatic shutdown devices will be l activated if the radiation detection equipment, located at the blower exit, detects excessive radiation levels. ,
3.5 Material Selectic; i, The most corrosive conditions occur in the incinerator /calciner, the dry cyclone, and the gas cuct 'eaving the cyclone. Temperatures in the i 61 _y 3,- , -, .. .._s.,=.,m. . - _ ryy . . ,,_.. _., . . - - _ . - . - , - -
range of 1000'C together with acidic residues result in highly corrosive , conditions requiring that these components be fabricated with corrosion resistant alloys such as Inconnel 601. Other metals are also being considered for these components. Samples of these metals will be evaluated in corrosion tests. Final selection will be made following these tests. The quench tank, associated nozzles, and venturi scrubber will be exposed to corrosive chlorides; however, the presence of the scrub solution will result in the forsation of a protective liquid film on these items.
., They will be fabricated with 304 stainless steel.
The commercial components of the scrub solution and off-gas systems will O - be fabricated from 304 or 316 stainless steel, whichever is the manufacturer's standard. The solid waste feed system components, fluidizing air system, off-gas blower, fuel systems, motors, control panels, and other minor components not subject to corrosive conditions will be fabricated with carbon steel. Table 3-4 summarizes the component materials and alternate candidata materials which have been' tentatively selected.
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It is recognized from past experience in the field of incineration that material problems have occured in systems similar to RWR-1*. A review of existing radieste incineration equipment, LA-6252,(5) indicates the
, magnitude of the corrosion probism. It is known that suitable alloys exist for the construction of the components exposed to these severe corrosive and thermal conditions. However, tradeoffs between availability,
' ease of fabrication, servicability, process chemistry control and cost nest be considered. .
i The high operating temperature of the incinerator and hot gas duct may require external cooling or the selection of high temperature refractory linings. Also, the possible formation of hcl from the combustion of i 1 62
plastic may pose a threat to prolonged life of austenitic stainless steels. It is believed that scrub liquid chemistry can ce adjusted to correct this proble but further untarial studies are pi? aed. The results of these tdies will be included as an addendim6 to this report when available. i 3.7 System / Plant Interfaces 3 The RW -1 Systee boundary is defined by a series of interfaces with the plant radwasta syste and other non-radioactive service piping. These interfaces are listed in Table 3-5 and illustrated in Figure 3-2. The radioactive weste streams entering the RWt-1 E from the station are: l - (1) Concentrated aqueous solutions (bottoms) from the liquid radteste evaporator - the interfaces are the inlet isolation valves on the pumps between the plant liquid radwaste storage tank and the incinerator /calciner. (2) Spent resins from the deineralizers - the interface is an isolation valve just upstream of the dewatering tank. (3) Combustible radwaste from handling, cleaning or maintenance - the interface is a door over the dry waste shredder. (4) Filter sludges from precoat filters - the interface is an d isolation valve just upstream of the dewatering tank.
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Radioactive waste streams leaving the RWt-1 3 are: (1) Off-gas scrub solution tank purge - the interface is an isolation valve just downstream of the purge flow control valve. (2) Floor drain and equipment drain wastes - the interface is the entrance to the drains, which are connected to the station liquid radwaste system. 63
.. 1 (3) Ventilation exhaust from the building housing the RWR-1 3 l System - the interfaces are grates which lead to the station HVAC exhaust syste.
(4) Water from dewatering the spent resins and filter sludges - - the interface is an isolation valve between the dewatering tank and the plant liquid radwasta syste. (5) Off-gas to the stack - the interface is an isolation valve
, downstream of the off-gas blower.
(6) Condensate from the desister - the interface is an isolation (,
, valve between the desister and the plant condensate water systan.
(7) Solid product - the interface is an isolation valve between the dry cyclone and the product container, The non-radioactive connections with the station are: (1) Electric power - the interfaces are the output connections of the motor control center (s). (2) Cooling or service water - the interfaces are isolation valves located between the plant supply and return and the RWR-l* Systam. (3) Instrument and service air - the interfaces are isolation valves between the plant supply systems and the RWR-1 System. (4) Fuel oil - the interfaces are isolation valves upstream of the fuel peps. (S) Plant or ambient air - the interfaces will be the inlet to the i fluidizing blowers, l 64
I (6) HVAC makeup air into the building housing the RWR-1* System - the interfaces will be the grates from the building HVAC inlet system. (7) Choical addition - the interface will be an isolation valve - upstream of the scrub solution tank. 3.8 RWR-1 Systm Operation 3.8.1 Introduction The RWR-l E System is a highly instrumented remotely operated system l'.,
'~
, ' utilizing components with demonstrated reliablity. This section describes automatic control system actions and required operator actions during normal operation. The RWR-l
- System has three operational modes for the three waste feed systems: combustible waste, resins and sludges, and liquid wastes. Startup in all three modes requires preheating the bed to the operating temperature using the fuel burner. Control set points and protective system actions are specific to the mode of operation.
3.8.2 Combustible Waste Incineration Mode Complete combustion of combustible wastes require adequate air (oxygen) ' and high taaperatures. These requirements are met in the RWR-l
- System by automatically controlling the waste feed, fluidizing air and overfire air rates at pre-set flows. The dry product (ash) elutriated out of the fluidized bed and captured by the dry cyclone accumulates in the product container.
3.8.3 Resin / Sludge Incineration Mode t . Adequate air for incineration of ion exchange resins and filter sludges is achieved by automatic control of flows for the resin / sludge feed, the fluidizing air, and the atomizing air. The temperature of the fluidized bed is automatically controlled by a fuel burner to assure complete combustion. 65
3.8.4 Liquid Waste Calcination Mode l In this mode the fluidized bed temperature is controlled by the fuel burner. The liquid waste and fluidizing air flows are automatically controlled at prescribed values. Since there is no waste combustion in -' the overfire area in this mode, no overfire air is required. 3.8.5 Off-cas System
. The scrub liquid temperature and scrub liquid flows to the quench tank and venturi scrubber are automatically maintained at optimum values.
The scrub liquid composition is controlled by the scrub liquid recycle
'( , rate to the station liquid wasta system. This flow is automatically controlled at a pre-set value which is selected to give the optimum scrub liquid composition.
The amount of water vapor condensed from the off-gas is controlled by the condenser exit tanperature. The operator has the option of either conto 111ng this temperature at the value which gives a constant scrub liquid systen inventory or running the condenser exit at its lowest achievable temperature. By running at the lowest achievable temperature there is excess condensate which is available for return to the station for cleanup and reuse. 3.9 RWR-1 Systen Reliability The investment in a nuclear reactor systen is very large; therefore, ensuring a high capacity factor with high reliability in support equipment, such at a radweste systen, is imperative. In the RWR-1 System, high reliability comes from minimizing the amount of equipment with moving parts and using redundant equipment where moving parts cannot be avoided. The fluidized bed provides high processing efficiency without the mechanical complexity of a moving grate incinerator or a scraped wall evaporator. The off-gas system is composed of low maintenance equipment such as cyclones and quench tanks instead of high maintenance items such as bag filters or packed colunns. 66 l
*w _m'er-p-3., . --_ -mN hpe --e--- w. e y r - -,,---www - - .-.
~
Components such as blowers and pumps require periodic maintenance as well as occasional replacement of parts. The actual vendor equipment selected for inclusion in the RWR-1 # ystem S will be items with proven high reliability. Further assurance of high system reliability comes , from the use of redundant pumps and blowers. In sunnary, equipment selection for the RWR-1# System is based upon high reliability. In areas where equipment selection alone cannot ensure good reliability, redundant equipment is installed. 3.10 RWR-1 3 System Internal Decontamination
'The RWR-1 3
System is designed to minimize possible crud traps. However, because of the nature of the materials handled, the potential exists for a buildup of radioactive contamination in the form of scale and solid material. To facilitate maintenance in the RWR-1* System, nozzles for remote decontamination by solution spray are built into the major process vessels. During a decontamination cycle, the excess decontamination solution will be routed to the plant liquid wapte system prior to calcination in the RWR-1# System during the next operating campaign. 3.11 RWR-1 System Maintenance The RWR-1 System components are designed for high reliability and low maintenance. However, equipment with moving parts such as pumps and [ control valves will require periodic maintenance. The ease of maintenance of this equipment is greatly influenced by the location of the components in relation to one another. The RWR-1 System is based upon the philosophy of placing the basic function components as modules in different cells. For example, the main process cell holds the incinerator /calciner and dry cyclone. The 67
wet scrubbing cell contains the quench tank, venturi scrubber, condenser and desister. The scrub liquid pumps and most of the valves are located in cubicals outside the major ')rocessing cells. Thus, they are accessible for maintenance without decontaminating the complete system. The final off-gas treatment equipment (HEPA filters and iodine adsorbers) are ' located in individual cells to permit ease of maintenance and replacement. Figure 2-5 shows the plan location of the major components. ; i Major maintenance would occur only after a decontamination cycle using the internal decontamination headers located in all major process vessels. Depending upon the residual radiation fields, either direct maintenance or a combination of direct plus rente maintenance will be used. All
. weldments are of the butt type and flanges are minimized to reduce possible locations where crud buildup might occur. Socket welds are not used. Provisions have been made for retracting the screw feeders used in the feed systems.
e h l Oe a 68
TABLE 3-1 RWR-l
- SYSTEM COMPONENTS Normally No. Furnished Component Required With RWR-1
- 1. Solid Waste Shredder 1 *
- 2. Solid Waste Feed Bin & Screw 1
- 3. Solid Waste Block Valve 1
- 4. Liquid Waste Storage Tank 1 Optional
- 5. Liquid Waste Feed Pump 2
- 6. Liquid Waste Strainer 2
- 7. Liquid Wasta Control Valve 2
- 8. Liquid Waste Feed Nozzle 2
- 9. Resin / Sludge Mixing Tank 1
- 10. ~ Resin / Sludge Metering Pump & Valve 1
- 11. Dewater Liquid Return Pump 1
- 12. Fuel Oil Burner, All Valves, Pumps &
Ducting 2
- 13. Incinerator /Calciner Vessel 1
- 14. Fluidizing Air Blower 2
- 15. Fluidizing & Overfire Air Control Valves 5
- 16. Dry Cyclone 2
- 17. Dry Cyclone Product Shutoff Valves 2
- 18. Dry Product Storage Hopper 1 Optional
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- 19. Dry Cyclone Product Duct Between Yalves 1
- 20. Fluidizing Air Ducting - From Valves to Incinerator /Calciner Vessel 1
- 21. Fluidizing Air Ducting - From Blower to Valves 1
- 22. Hot Gas Duct 1
- 23. Quench Tank 1
- 24. Quench Tank SP :) .40 xzles 6 69
TABLE 3-1 (Contd) ' Normally No. Furnished Component Required With RWR-1 3 i
- 25. Venturi Scrubber 1 *
- 26. Wet Cyclone 1
- 27. Off-Gas condensor 1
- 28. Demister 1
- 29. Scrub Solution Tank 1
- 30. Off-Gas Duct Connecting Quench Tank.
Venturi, Wet Cyclone, Condenser, Demister, Pre-heater, HEPA Filters, Adsorber and Off-Gas Blower -
- 31. Scrub Solution Piping .
- 32. Scrub Solution Pump 2
- 33. Scrub Solution Valves s51
- 34. Off-Gas Valves s34
- 35. Air Preheater 1
- 36. Raw Water Valves $12
- 37. HEPA & Roughing Filten & Housing 4
- 38. Iodine Absorber 2 ~
- 39. Off-Gas Blower 2
- 40. Off-Gas to Stack, Recirculation &^
Makeup Air Ducts - ( 41. Fuel Oil System Including Filter, Storage Tank and Piping 2 Optional '
.. 42. Control Panel, Controls & Instrumentation.
Including Transmitters & Detectors 1
- 43. Motor Control Center 1 Optional
- 44. Fan and Pump Drive Motors - Electric -
- 45. . Solenoid and Motor Yalve Operators -
- 46. Scrub Solution Cooler 1
- 47. Sample Valves -
Optional Blank indicates this itam is furnished as a part of RWR-1 . l l 70 l l
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s , . . ; i 2 ' - . i .- ] TABLE 3-2 - EQUIPENT CODES , i Equipment Codes l Walder , Design and Qualifications ' Inspection Fabrication Materials [2] and Procedure and Testing Pressure Vessels ASE Code ASME Code ASE Code ASE Code Section VIII. Div. I Section II Section IX Section VIII. Div. 1 Atmospheric or ASE Code I33 ASME Code E43 ASE Code ASE Code E33 0-15 psig tanks Section III. Section II Section IX Section 111 Class 3. or API 620 & Class 3 or API 620; 650' AWWA D-100 650 AWWA D-100 llent Exchanger ASME Code ASE Code ASME Code ASE Code Section VIII. Div. I Section III Section IX Section VIII. Div. 1 3 and TEMA Piping and Valves ANSI 31.1 ASTM or ASME Code ANSI 8 31.1
- ASE Code Section IX Section II Pumps Manufacturer's Ell ASME Code ASME Code ASME E3l Standards Section !! or Section IX Section III Manufacturer's (asrequired) Class 3; or
. Standard Hydraulic Institute Notes:
Ell Manufacturer's standard for the intended service. Hydrotesting should be 1.5 times the design pressure. [2] Material Manufacturer's certified test reports should be obtained whenever possible. El ASME Code Stamp and material traceability not recuired. [4] Fiberglass reinforced plastic t.anks may be used in accordance with Part M. Section 10. ASME Boller and Pressure Vessel Code, for applications at ambient temperature. S
. j s .
\
- s Table 3-3 Component Design Data
-s J'
l l [ This Table contains proarietary infomation 72
l Table 3-3 Component Design Data (continued) v This Table contains proprietary infcmation l 73
Table 3-4 RWR-1 E Major Component Materials List
- F o
l' This Table contains proprietary information l 74
TABLE 3-5 INTERFACES OF RWR-l
- SYSTEM WITH PLANT SYSTEM INTERFACE Radioactive Wastes In:
- 1. Concentrated Liquid Inlet isolation valves between plant concentrated waste storage tank and liquid waste pumps.
- 2. Spent Resin Inlet isolation valve between plant resin slurry tanks & dewatering tank.
3 Combustible Solids Door into waste shredder from plant.
- 4. Filter Sludge Inlet isolation valve between plant resin slurry tank and dewatering tank.
Radioactive Wastes Out:
- 1. Off-gas Scrub Liquid Outlet isolation valve between flow centrol valve & concentrated waste storage tank in plant. -
- 2. Floor & Equipment Drain Entrance to drains connecting with plant liquid waste system.
- 3. Ventilation Air Grates leading to plant HVAC system.
- 4. Resin Dewatering Liquid Outlet isolation valve between dewater-ing tank and plant liquid waste system.
- 5. Off-gas Outlet isolation valve between off-gas blower and plant stack.
- 6. Denister Condensate Outlet isolation valve between denister and plant condensate system.
- 7. Solid Product Outlet isolation valve between dry cyclone and plant furnished product container.
Non-Radioactive Connections:
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j 1. Electric Power Outlet connections from motor control center to loads.
- 2. Service Water Inlet and outlet isolation valves between plant and RWR-1TM,
- 3. Instrument & Service Air Inlet isolation valves between plant supply and RWR-1TM.
- 4. Fuel Oil Isolation valves between the plant tanks and the fuel pumps.
- 5. Ambient Air Inlet to fluidizing blowers.
- 6. HVAC Makeup Air Inlet grates from plant HVAC System.
- 7. Chemical Addition Inlet isolation valve between plant and scrub solution tank.
75 l l l
Figure 3-2 Sheet 1 Piping and Instrumentation Diagram l This Figure contains proprietary information l i 77 \
Figure 3-2 Sheet 2 Piping and Instrumentation Design This Figure contains proprietary infonnation 9 78
4.0 ENVIR0lWENTAL IMPACT ANALYSIS This section describes the impact of the operation of the M-1* System on the general envircreent. An adverse impact may result from the release of gaseous, liquid or solid radioactive material in large, uncontrolled amounts. The M -1 System has been designed on the basis of safety and containment so that no credible accident results in the release of more than a few hundred Curies of airborne activity, this amount of activity being based on several worst case assumptions. The
, nonnal operating atmospheric releases are so low that even with very conservative assumptions the resulting doses are orders of magnitude less than typical background doses. Liquid releases to the envirorment
' are not envisaged frca the M-1 3 Systen, eit.her under norwal operating conditions or as the result of accidents. Liquid spilled from the M -1 3 System due to accidents will be contained by the building housing the system.
The solid material from the dry cyclone is placed in a product container and removed in a controlled manner for innobilization or storage. The final destination of the product is burial at a regulated site. Two typos of releases to .the atmosphere are considered, nomal and
. abnormal. To determine the nomal operating releases, the off-gas cleanup system decontamination factors are detemined and applied to the maximum activity feed rates developed earlier (see case E of Table 2-4).
Abnomal releases may result from either anticipated transients or postulated accidents. Anticipated transients are considered to be non-catastrophic failures, sore of etch may be expected to occur at some time during the life of the M-1 D System. Postulated accidents are catastrophic events whose occurrence is considered credible though unlikely. 79 __ _ _ . _ . _ _ _ . . . . ~ .
4.1 General Background Most systems designed to incinerate or calcine radioactive wasus have employed some system to reduce the releases to the atmosphere, c1though some small, early installitions had no off-gas cleaning whstsoever. A review here of the dozens of previous incinerators is not needed because of the recent sununaries contained in LA-6252(5) and ERDb76-43,(6) The older reviews, WASH-ll68(7) and IAEA Safety Series No. 28N)mayalso be consulted. Four incinerator or calciner systems have off-gas cleanup
. systems which closely resemble the RWR-1 System. They are the second incinerator built at Harwell (U.K.)(5,14,15,16) , the Waste Calcining Facility (WCF)(2,17-20) at INEL, the New Waste Calcining Facility at' INEL,(21,22) and the Aeropep System.(23,24) Data of sufficient detail from Hansell was not located and NWCF is not built yet, so data from WCF aad Aeropep was used. When available, data from WCF is preferred because it comes from long-duration production runs, whereas the Aeropep data is from test operations of relatively short duration.
4.2 Decontamination Factors The off-gas clean-up system is d: scribed in Sections 2 and 3.3 and Figure 3.1. The quench tank, venturi scrubber, wet cyclone, condenser
. and desister make up the wet scrubbing system. While there are some differences in the efficiency with which the clean-up system removes various elements, the bulk of the activity will be contained in elements which will fann stable oxides in the incinerator /calciner, f.e. Cr, Ma, Fe, Co, Sr Zr, and Ce. Cesium may be in the oxide or hydroxide form depending upon dether hydrolysis takes place. The overall effect of l
the off-gas cleanup system will be equivalent regardless of whether casitse leaves the calciner/ incinerator as an oxide or as a hydroxide.
- These oxides are all in solid phase at the highest temperatures expected l fn the RWR-1 process, and therefore will remain as " solid particles l throughout the off-gas system. Iodine is present in velatile forms, and i its removal from the off-gas is considered separately.
80
l 4.2.1 Decontamination Factors for Particles l The Ria-1 3 off-gas clean-up syste is similar to that used on the Wasta Calcining Facility (WCF) at Idaho National Engineering Laboratory (fonnerly National Reactor Testing Station) (2,17-20) The two systems are similar in that the incinerator /calciner is followed by a dry cyclone, a quench tank, a vr.. turi scrubber and a wet cyclone. After this first wet cyclone, the WCF Syste has a second cyclone whereas the RlR-1D System has a condenser and a duister. Following the second cyclone, the WCF System has a heater, an adsoruer, a third cyclone, a second heater, and a filter. Following the duister, the RWR-1 off-gas clean-up system has a heater, a filter, an adsorber and a second filter. The difference
.- in the arrangment of the heater (s), filter (s) and adsorber are not significant. The major difference is that the wet cyclone following the venturi scrubber was followed by a second wet cyclone in the WCF System while the wet cyclone following the venturi scrubber is followed by a condenser and a demister in the RlR-l* Systen.
2 The Decontamination Factors (DFs) reported by WCF are sumarized in Table 4-1 for the five processing campaigns. A DF for the dry cyclone ' is specifically identified only in the report for the fifth campaign.(20) The DFs for the RWR-1 System are given in Table 4-2. Two values are listedforparticlesandfoi* iodine; an ex;,ected value and a lower, more conservative value which is used in computing emissions and site boundary concentrations. The efficiency of the dry cyclone was established using data from sodim sulfate calcination feasibility studies.(25) Analysis of solids collected from the cyclone bottom outlet indicate that approximately 82% of the , mass of the total solids entering the dry cyclone are collected, with the reminder leaving as fines with the gas stream. Therefore, ass ming that the non-gaseous radionuclides exit the dry cyclone in direct proportion to the total solids mass, the DF shown in Table 4-2 can be expected for particulate matter. Similar DFs have been measured in other dry cyclones.(26) For use in computing emissions, however, a more conservative value has been selected. 81
.s..
The size distribution of solids entering and leaving the dry cyclone in the feasibility studies was used to detennine the particle size distribution entering the quench tank. The potential of the quench tank for removing solid insoluble particles from the gas stream is dependent upon the particles striking liquid droplets. Knowing the particle size distribution ' and the expected droplet size distribution in the quench tank, the fraction of the particles which strike a droplet may be computed by standard procedures.(26) Some of the droplets, however, are carried out of the quench tank with the gas stream. Assuming that the mass of the particles collected by the scrub solution leaves the quench tank in proportion with the liquids leaving the quench tank, the OF shown in Table 4-2 for the nuclides in solid form has been computed. Once the
. solid particle strikes the droplet, its fate is that of the entire droplet and is independent of the solubility of the particle in the liquid comprising the droplet.
A particle wetting efficiency has been calculated for the venturi scrubber, using the particle size distribution leaving the quench tank and standard methods.(26) The unwetted particles reaching the wet cyclone continue with the gas stream and the wetted particles, soluble or insoluble, leave thp wet cyclone in proportion to the liquids leaving the cyclone. l So few particles will remain unwetted at the wet cyclone that the relative l liquid flow rates from the cyclone become the dominant factors in det: training the DF across the venturi and wet cyclone. Almost all of the liquid remining in the off-gas downstream of the wet cyclone is removed by the condenser and the demister.(20 Since the wetted particles will leave
,, the exhaust stream in the same proportion as the liquid, the DF across , the condenser and demister is quite high. '; the DF across the entire wet scrubbing systen is listed in Table 4-2.
It is the product of the DF across the quench tank, the DF across the venturi scrubber and wet cyclone, and the DF across the condenser and demister. A much lower value has been used in computing the emissions, so that there may be no question about the fact that the emissions are conservatively cverestimated. 82
For the particle sizes measured in preliminary tests,(25) measured effiencies greater than 99.95% have been reported for HEPA filters.(28,29,30) The DF expected is shown in Table 4-2 and a much lower value has been entered for use in calculating emissions. The second HEPA filter in the system is purely a back-up filter and no credit for its existance is taken. 4 Table 4-2 lists the overall DF of 4 x 10 which will be used in computing particulate material emissions from RM-l D This overall DF is many
, orders of magnitude below the expected overall DF. In addit. ion, it is 4
lower than the lowest OF measured at WCF. Therefore, 4 x 10 is thought to be a conservative figure. 4.2.2 Decontamination Factors for Iodine The radionuclides I-131 and I-133 are not mentioned by WASH-1258 because of its 180 day decay basis and their short half-lives. For the same reason, they are not often found in analyses of radwaste shipments. I-131 and I-133 have been included in Tables 2-2 and 2-3 however, because
'they are expected to be present when the.radwaste is fed into the RW-1D System. I-131 has also been included in Table 2-9 for the same, reason.
At tt.e temperature of the. incinerator, iodine will be present primarily
. as I 2 r as HI,'both volatile species. Therefore, the iodine will react differently than the particles, and a separate consideration of 0Fs for fodine is required.
Table 4-2 lists two values for each iodine decontamination factor, a value dich should be seen in actual operation, and a lower, conservative value used for computing emissions from the RW -1 E System. On the basis of tests made with the Aeropep systen, a DF of 2 is expected across the dry cyclone, but for the purpose of computing enissions it is assumed that no fodine at all is in particulate for$i and therefore there is no iodine renoval by the dry cyclone. 83
In the feed to the R W-13 Syste, fodine may be present as elemental iodine (12), hydrogen fodida (HI), hypoiodus acid (HOI), or it may be in particulate or organic fom. After passage through the highly oxidizing atmosphere of the incinerator /calciner no organic forms reain. It was shown in tests with the Aaropep system (24) that the rooval efficiency of a wet scrubbing system was lower for HOI than it was for either almental or particulate iodine. Therefore it is conservative to assume that all the iodine is present in the fotis of hypoiodus acid. The Aeropep tests indicated that a DF of 4 was obtained across one wet contacting stage, a venturi scrubber. The RW-l* Systa: contains two wet contacting stages in series, a quench tank followed by a venturi scrubber. Therefore a DF of 16 is a very conservative estimate of the
. OF across the wet scrubbing system in the RWR-1 System, and this value is used for computing emissions.
More efficient fodine reoval than this is expected in actual operation since a proprietary fodine r eoval additive is being developed for the RW-1 D System. Tests have been conducted (31) in which a DF greater than 30 was repeatedly obtained in an experimental apparatus which contained one wet contacting stage. The feed gas stream was rich in CO 2 and poor in 02 , as is the case with the off-gas from the incinerator /calciner, and contained iodine as at about 100 ppe. Since the RWR-l E Systae contains two wet contacting stages in series, the DF might be expected 3 to be on the order of 10 . However, a more conservative value has been entered in Table 4-2 for the expected perfomance DF of the wet scrubber. _ The radiciodine which goes into solution in the scrub liquid will decay
, with a half-life of about 8 days. Therefore most of the radiciodine will decay to stable xenon isotopes while in the scrub liquid system.
' The xenon atoms will be vented through the aff-gas exhaust. Almost all the fodine which succeeds in passing through the we,t scrubbing system will be caught in the silver-loaded adsorber: Table 4-2 lists the DF expectad across the adsorber. A DF of this magnitude is well within the capabilities of currene technology (32,33) . A value lower than the expected 0F has been used in computing iodine emissions. The overall 0F 84
e
- for iodine to be used in cesiputing missions is 1 x 104. It is expected that the actual DF of the operating systen will significantly exceed this value.
4.2.3 Release Rates from Maximum Expected Operations ' In Table 4 3, the rate of feed from Table 2-9 is divided by the DFs to obtain the rate of release from the RnR-1 System. Since the rate of feed is conservatively high and the OFs are conservatively low, the {, calculated release rates are conservatively high. Actual operating releases are expected to be considerably lower than those listed in Table 4-3 due to these conservatisms. Even with these assumptions that overestimate the activity released, the total activity released is computed to be less than 0.5 Ci per year. This may be compared to the thousands of Curies of noble gases estimated to be emitted each year by a 3500 MWt plant.(8) 4.2.4 Dose Rates and Concentrations Resulting from Maximum Expected Operations The site boundary has been assumed to be 1,000 meters from the point of release. The dilution factor (x/Q) used,1.2x10-5 3,cf,3, is the 4 to 30 day value for a ground-level release from Regulatory Guide 1.3.(34) Using this value for an aanual average dilution factor is very conservative because the average x/Q varies inversely with the length of time over which the average is taken. Site boundary concentrations calculated
; from the release rates, with this x/Q, are listed in Table 4-3. For most nuclides the concentrations are four or five orders of magnitude below the limiting concentirations specified by 10CFR20. The nuclide with the highest concentration has a concentration more than three '
orders of magnitude less than the limiting concentration. l The last three columns of Table 4-3 list the whole body, thyroid and lung dose rates for the concentrations at the site boundary. It has been assumed for the purposes of this report that the recipient is ? -
)
~ ~ _ _ _ . . . . . . . - - . - - - - _ .- - - - _ _ - _ _ - . _ - - - - _ - _ - , - - _
,_ v located at the 1000 meter site boundary and is breathing at a rate of 20 3
m / day. The dose factors used have been taken from Regulatory Guide 1.109.(35) The dose rates are very low even with respect to background radiation. The highest dose rate is that to the thyroid with the feed assumed from a 3500 MWt BWR. This dose rate is less than one millirem per year. Due to the many conservatisms in the computation of the dose rates, the expected dose rates will, in all probability, be much lower j than those listed in Table 4-3. 4.3 Exposures from Anticipated Transients and postulated Accidents
, In, this section the dose received frem radiation due to abnormal occurances
- in the RW -1
- System is considered for a hypothetical person standing at the site boundary. Anticipated transients are those events which can .
reasonably be expected to occur during the operating life of the RWR-1* Systae. This class of events includes such occurances as leaks in pipes and tanks, pump failure, and plugs in nozzles. Accidents,, 1 the c+.her hand, are considered to be more catastrophic events which a e not expected to occur, but which must be considered possible. An exampi! of an accident is the complete, sudden rupture of a tank. As might be expected, the consequences of the postulated accidents are generally more serious than the consequences of the anticipated transients. Even so, with very conservative assumptions, the maximum dose from the maximum credible accident is shown to be only slightly greater than the maximum annual t occupational dose limit. l 1 4.3.1 Anticipated Transients Transients resulting from the non-catastrophic failure of some component of RW-1 may be expected during the 1ifetime of the' system. Because
^
the complete instrumentation systen is coupled to alarms and interlocks, most of these transients will cause the RWR-1 system to shut down quickly with little or no radiation consequences. Table 4-4 lists all the transients which appeared to be reasonably credible. 86
The first two cases in Table 4-4 concern leaks in liquid lines or tanks. There, leaks will cause contaminated liquid to accumulate on the floor l of the radweste building. The leaks in themselves will not affect the l performance of the M-1 3 Syste, and the operator will shut down the system. Because the radioactivity will be in the liquid or in particles contained in the liquid, no significant gaseous releases from the building i containing the M-1* System are expected. Since the foundation of the I building is of seismic design and is capable of containing the entire liquid inventory within the building, no liquid release is expected.
. There are no adverse radiological consequences associated with leaks in the incinerator /calciner or with leaks in the off-gas systa (Item 3 in
, T.able 4-4) because the pressure in the entire system is slightly below atmospheric pressure.
Items 4, 5 and 6 in Table 4-4 concern plugs in the scrub liquid lines. If one of these lines should plug, either a low reading from the flow sensor in the line or a high reading from the liquid level indicator in the quench tank or in the wet cyclone will cause the system to shut down. If the iodine adsorber or one of the HEPA filters plugs (due to excessive moisture loading or particle loading), the increased pressure drop across the adsorber or filter d11 be sensed. If the adsorber or filter should blow out, the final radiation monitor will register an increase in radiation levels and shut down the incinerator /calciner while recirculatitig the exhaust gas back through the off-gar clean-up systan.
. The last ten items listed in Table 4-4 have little potential, if any, for distributing radioactive material to the general environment. These transients either tend to keep radioactive material within the RWR-1
- Systee or they tend to shut the systen down. Out-of-range sensor readings will either shut down the system automatically or initiate both auditory and visual warnings to the operator.
87 _=. _ _ = . _ . . , - -. . - . . _ _ - . . . . . _ - , . - . - - .
aq , 4.3.2 Postulated Accidents A number of potential accidents have been considered, such as rupture of the resin-sludge dewatering tank, f9cture of the dry waste hopper, rupture of the calciner/ incinerator, leaks in the off-gas system, rupture - of the scrub liquid tank and failure of any component in the off-gas systaa. None of these accidents was as potentially serious as the violent rupture of the product container. This is due to the fact that no other portion of the system can accumulate as much activity in dry forie. The only other component of RW-1* which can accumulate much activity is the scrub liquid tank. If this tank were to rupture, the activity would be released into the radwaste building in wet form. In
, this forre it would be much less likely to pass into the air and escape from the building than it would be in the dry, granular form of the material in the product container. The tanks which accumulate resin, sludge or liquid wasta prior to processing in the RW-l* System might contain a significant amount of activity,,but they are outside the E
RW-1 System and their contents are wet. Therefore, the postulated maximum credible accident is the violent rupture of the product container. For the purpose of this analysis it is assumed that the product container is a 55-gallon drum. The 55-gallon (0.208 m )3drum has been chosen
. because the specific activities in the product are high enough to make larger containers infeasible. Further, the volume reductions occurring in the RW -1 3 process eliminate the necessity for containers larger than 55-gallon drums to contain the product.
[ By the following analysis it has oeen deterinined that 1670 Ci is the l upper limit to the activity that will be contained in one drum. Table 4-5 shows all those reactor sites which list resins or resins and sludges separately as well as those which differentiate between wet and dry wastes and which shipped more than 200 Ci/yr of wet waste. A review of this material indicates that it would be extremely unlikely that the RWR-1 3 Systan would process more than 2000 Ci or less than 5 m3 in one month. Therefore, the maximum specific activity, before p m essing. l 38
}
3 would be 400 Cf/m . While individual shipments of resin or sludge have had higher specific activities than 400 Ci/m3 (see Table 2-11), they did not amount to a great deal of volume, as discussed in Section 2.2.3. Other shipments during the month were greater in volume and much lower in specific activity so that the average specific activity of resins and ' 3 sludges for the month was much less than 400 C1/m . Further, it is ass' aed in this analysis that the volume reduction is the maxima envisaged for wastes for other than the dry combustible waste, 20:1. A voltme reduction of 20:1 is greater than the 18:1 voltme reduction expected for spent resin, and has been used here to conservatively increase the specific activity in the drum. Use of a 20:1 volume reduction factor gives a specific activity in the product container of 8000 Cf/m3 , or _ i670CiperSS-gallondrum. This implies that more than one sixth of the maximus amount of activity for the entirs year accumulates in one drum (1670/9500 = 0.166), which is unlikely to occur. It would seen conservative to asstne that ten percent of the granular ash in the product container escapes fram the building containing the RWR-1E I Systen and remains airborne long enough to reach the site boundary. l 4.3.3 Exposure at the Site Boundary from Anticipated Transients and Postulated Accidents The potential for the release of radioactive material to the general environment from the postulated maximum credible accident is greater than that from any of the anticipated transients. This is due to the fact that the worst anticipated transient results in the release of the
^ radioactive waste in wet form. In this form it is less mobile and less likely to migrate to the general envirorment than it is in dry fonn.
The possibility of the release of dry radioactive material by an anticipated
;. transient is essentially zero because the entire off-gas syston is operated at a pressure slightly below ambient pressu,re.
The maximum abnonnal release from the RWR-1* System, then, is that from the postulated maximum credible accident, wnich is the violent rupture of the product container at a time when the container, assumed to be a 1 89 { - t
s *, .
- SS-gallon drum, is filled with product material which has a specific activity greater than that likely to occur under any reasonable circumstances.
Do es computed assuming that ten percent of the maximum credible epntents of the drum (167 C1) escape from the building in two hours are given in - Table 4-6. The dilution factor, x/Q used is the 0-2 hour value from Regulatory Guide 1.3 ) for 1000 m and a ground level release. The dose factors were taken from Regulatory Guide 1.109,(35) and the breathing rate 3 was 20 m / day. The nuclide distribution has been assumed to be that
. shown in Table 2-9.
The maximum accidental doses are seen to be quite small, all of them are
,, under 6 ras. This may be compared to the annual occupational permitted dose of 5 rem. Even if 167 Curies were to escape, it is unlikely that the thyroid doses would be as high as shown in Table 4-6. It was assumed in computing these doses that I-131 is present in the product container in the same proportion that it is present in the feed. Actually, due to the volatility of iodine, a much smaller fraction of iodine than of the other nuclides will be removed in the dry cyclone. As shown earlier in this section, the I-131 is captured in the scrub solution and adsorber where it decays to stable Xe-131. Therefore the radiciodine will comprise a smaller fraction of the. activity in the product container than it did
. In the feed. In any event, it is very conservative to assume that one product container will accumulate 1670 C1, so the likelihood of accident doses as high as those of Table 4-6 is very low, givea the rupture of the product container.
0
- e 9
e 90
, , - , - -y99--,-, ..,-gpi.e wy .7-- T-w - - -- m,%~- w+~~- - - - - - - - -
,. , . s ,,
~ -s V .) ,,.-
' .i TABLE 4-1 .
SOLID PARTICLE (0XIDE) DECONTAMINATION FACTORS FOR THE OFF-GAS CLEAN-UP SYSTEM OF THE WASTE CALCINING FACILITY Dry Wet Scrubbing Basis Reference Cyclone System Adsorber Filter Overall Average of Sr-90 and Cs-137 2 NR 645 6 10 220 1.4x10 Measurements During the First Processing Campaign Sr-90 Measurements During 19 2750 8 NR 10 3000 1.6x10 the Second Processing Campaign Sr-90 Heasurementt During 20 NR 1000 10 1500 1.5x10
- the Third Processing l Campaign Sr-90 Heasurements During 21 NR 2760 8 630 1.4x10
- the Fourth Processing Campaign
. i Cs-137 Measurements During 22 3 25 3 1300 3.0x10 5 the Fifth Processing Campaign NR = Not Reported. In the first procassing campaign, a DF of 6.25 is reported for the calciner and dry cyclone together. For the second, third and fourth campaigns, no DF is reported for the dry cyclone, alone or in combination.
.. v, ..
I .. . l 1 l 6 h ( t . Table 4-2 Decontamination Factors For the Off-Gas _ Cleanup System of the RWR-1 System w
~
t
~
This Table contains proprietary infomation ae 92
(> b .- i , TABLE 4-3 LHISSION RATES. BOUNDARY CONCENTRATIONS. AND DOSE RATES RESULTING FROM THE ' OPERATIONSYSTEM OF THE
- RWR-1#
FOR A 3500 W t BWR Rate of Decontamination Rate of Concentration Boundary Whole Body Thyroid Nuclide Lung Feed Factor Release Limit Concentration Oose Rate Oose Rate Oose Rate C1/yr mC1/yr pC1/m 3 pCf/m 3 mres/yr ereelyr mram/yr CR-51 2000 4
- 4x10 50 80000 .019 HN-54 800 4 4x10 20 1000 .007 4
.010 FE-55 100 ~4x10 2.5 30000 .001 .0001 .001 FE-59 300 4 4x10 7.5 2000 .003 .003 C0-58 4 1800 4x10 45 2000 .017 .015 00-60 4 1300 4x10 32.5 300 .012 .0002 .067 g ZN-65 400 4 4x10 10 2000 .004 .0002 .003 SR-89 80 4x10 4 2.0 300 .000R .001 SR-90 20 4x10 4 0.5 30 .0002 .0011 .002 1-131 300 4 1x10 30 100 .014 .0002 .124 CS-134 700 4 4x10 17.5 400 .007 .0044 .001 CS-137 1400 ' 4x104 35 500 .013 .0052 4 .001 CE-141 300 4x10 7.5 5000 .003 .001 Total 9500 260.0 .0114 .124 .103
\ ,
l TABLE 4-3 (Contd.) FOR A 3500 W t PWR Rate of Decontamination Rste of Concentration Boundary Whole Body Thyroid Lung Nuclide Feed Factor Release Limit Concentration Oose Rate Dose Rate Oose Rate C1/yr mci /yr 3 pC1/m pCl/m ares /yr ares /yr ares /yr CR-51 100 4x10 4 2.5 80000 .001 HN-54 400 4 4x10 10 1000 .004 .005 FE-55 4 100 1x10 2.5 30000 .001 .0001 .001 FE-59 4 100 4x10 2.5 2000 .001 .001
- CO-58 2700 4x104 67.5 2000 .026 .022 CO-60 4 1800 4x10 45 300 .017 .0002 .093 7N-65 4 100 4x10 2.5 2000 .001 .001 i
I SR-89 20 4x10 4 0.5 300 .0002 SR-90 4 80 4x10 2 30 .0008 .0042 .007 I-131 4 2000 1x10 200 100 .076 4014 .828 4 CS-134 700 4x10 17.5 400 .007 .0044 .001 C5-137 4 1300 4x10 32.5 500 .012 .0048 .001 4 CE-141 100 4x10 2.5 5000 .001 i Total 9500 387.5 .0153 .828 .131 l
*The rate of feed is taken from Table 2-9. The Decontamination Factors are from Table 4-2. The Concentration Limit l
i 1s that for unrestricted areas and is taken from 10CFR20, Appendix 8. Table !!; when the soluble and insoluble ' limits vary, the lower limit has been used. The boundary concentration has been computed using the 4-30 day x/Q value for 1000n from Regulatory Guide 1.3, Figure 3 (ground level release). The dose rates have been computed assuming that the recipient stands at the boundary breathing at a rate of 20 m3 /day. The dose factors have been taken from Regulatory Guide 1.109. Table C-1.
1 \
- 4 . .f i> -
t TAttE (-{ , ANilCIPATED TRAllSIENTS F0ft THE RWR-1" SYSTEM AND THEIR CONSEQUEleCES TVPE OF PA0 CESS Ile0ICATICII PA0 CESS IS j CASE COMPONENT (H FAlluRE Il0EDIATE CONSEQUEllCE OF FAltuRE SHUT 00681 BY RADIATI0llAL CGIISEQUEIICE 1 1 Re-in-Sli.dge Tank or Line Leak Contaminated Water en Room Air Monitor Alarm Operator Minor, confined to Radmaste ! Dewatering-Tank or Line Floor Building j Liquid Weste Tank or Line 2 Scrub Liquid Tank Leak Contaminated Water on Room Air Monitor Alarm Operater Minor, confined to Radmaste i Floor tullding 3 Incinerator /Calciner Leak sullding Air Flows fato System Pressure and Operator Mone Of f-Gas Piping Off-Gas System Component Olfferential Pressure Indicators 1 4 Scrub L1 quid Feed Lines Plug Loss of Cooling (n High Temperature in the Quench Tank Exhaust Minor ] Quench Tank, or - quench Tank Exhaust. Temperature Interlock Decrease of Decontam- Venturi Enhaust and or the Venturt Scrub i laation factors in Wet Cyclone Exhaust. Liquid Flow Interlock i sa Off-gas Cleanup System Low Flow Alarms for or the Quench Tank Quench lank and Scrub Liquid Flow Venturl Scruh Liquid. Interlock or Enhaust i> High Radiation Alam Radiation Interlock In the Enhaust 5 Scrub Liquid Return Lines Plug Scrub Liquid Accumu- High Level Alare in Operator alone lates in Quench Tank Quench Tank or Wet or Wet Cyclone Cyclone or low Level Alarm in Scrub Liquid Tank . i 6 Coolant Line to or from Plug Scrub Liquid High Scrub Liquid Operator or Quench Sciub Liquid Cooler None Temperature Rises Temperature and High Tank Exhaust Temperature Quench Tank and Venturl Interlock Exhaust Temperature Indicators i 7 itEPA or lohae Adsorber Plug increased Pressure High Pressure Differ- Operator 1 tone Drop ential Indication or Alare. Low Pressure Indication at Off-Gas Blouer Inlet. l i l
i o
~
. .c TABLE 4-4 (Contd.)
T@E OF P90CE55 IllDTCATION Pil0CE55 15 CASE f0HPONENT(51 FAlltsE IletEDIATE CONSEQUEhCE OF FAILURE SHUT 006018V RADIATIONAL CONiTOUEIICE 8 IIEPA or Iodine Adsorber Blowout Increased Activity High Radiation Alarm or Operator None Downstream High Pressure Differ-ential Indication and Alare prior to Blowout 9 Off-Gas lleater Electri- misture Plugs HEPA Low Heater Temperature Operator Minor cal Filter Differential and High HEPA Pressure Differen-tial Indication or Alam 10 Dry Cyclone Plug Particles in Dry High Pressure in Dry Operator Hone Cyclone Product Exit Pipe 11 Coolant Line to Condenser Plug
- Motsture Plugs HEPA High Temperature in Operator None Filter the condenser Exhaust.
High Tamperature Alam e
- in the Heater Exhaust or High HEPA Differen-tial Pressure Indica-tion or Alam 12 Fluidizing Blower Insaf- 8ed Settles Fluidizing Air Low Fluidizing Air None ficient Flow Alam Interlock or Fluidizing Flow Blower Interlock i
13 Of f-Gas Blowu Insuf- Bed Settles High Off-Gas System Fluidizing Air Inter- None ficient Pressure and Low lock or Off-Gas Blower Flow Fluidizing Air Flow Interlock 11 Scrub Liquid Pump Insuf- loss of Cooling in Low Flow Alams for
~ Quench Tank or Venturi None ficient Quench Tank Decrease Quench Tank and Venturl Scrub Liquid Low Flow Flow of Decontamination High Temperature Interlocks or Quench Factors in the Quench Tank Tank Exhaust Temperature Enhaust. High Radiation Interlock. Exhaust Alarm in the Stack Radiation Interlock, or Line Scrub Liquid Pump Interlock
i
~
1 TABtE4-4(Contd.) IVPE CF PIIOCf5T11iDTCATI0el PB0Ciss Is ] CASE COMPONENI(SJ FAllDRE IletEDIATE CONSEQUENCE OF FAILURE $ HUT 00MN BV RADIATIONAL CONSEQUENCE Ib Fuel Pump Insuf- Bed Temperature Drops Low Fluidized Bed Fluidized Bed Tempera- None i ficient Temperature ture Interlocks Flow 16 Restn-Sludge Feed Pump Insuf- Iso m ste Feed Low W ste Feed Alarms Operator None Liquid Feed Pump ficient Flow 17 Dry Waste Screw feeder Stops No Wste feed Waste Feed Hopper Low Operator or fluidized None Level Alarm Fluidized Bed Low Temperature Bed Low Temperature Interlock Alarm la Plant Compressed Air low Loss of Air to Atomize I.ow Atossizing Air Flow Operator None Pressure fuel and Operate Valves or Low Pressure Alarms, a Multiple Alarms as
' Control valves Move ,
in Fall-Safe Direction i I t
s -: a TABLE 4-5 RESINS AND TOTAL WET SOLIOS SHIPMENT RATES * . Shipment Rate Site Type of Waste Ci/yr m 3/yr Cf/mo m 7,, 3
, Browns Ferry 1 & 2 Low-level Resin 280 185 23 15 High-level Resin 387 7 32 0.6 Total Resin 667 192 53 16 t, Brunswick,2 Total Wet Solids 267 879 22 73
' Conn Yankee Total Wet Solids 1970 160 164 13 Cooper Total Wet Solids 549 344 . 46 29 Dresden 1, 2 & 3 Total Wet Solids 570 1668 47 139 Fitzpatrick Total Wet Solids 241 718 20 60 i Indian Point Total Wet Solids 2071 442 173 37 Maine Yankee Total Wet Solids ' 1878 195 156 16 Millstone 1 Total Wet Solids 3635 1034 303 86
.Nine Mile Point Resin 1493 38 124 3 Evaporator Bottoms 891 467 74 39 Filter Sludge 2454 123 204 52 Total Wet Solids 4838 628 403 52 Oyster Creek Resin
~
1857 18 155 1.5 Pilgrim Total Wet Solids 12244 970 1020 81 Point Beach Resin - 4804 14 400 1.2 Robinson - 1389 291 116 24 ERDA-76-43 Total Wet Solids -8WR 1331 . 182 111 15 PWR 6055 141 505 12
*These radwaste shipment rates are for reactor sites which distinguished resin or total wet solids (resins, sludge, and evaporator bottoms) and which shipped over 200 Ci/ year. The rates includedare for nomalized to }gQ>O MWt. Values from ERDA-76-43 (powdered resin system) have been comparison.\
98
f .
\
i TABLE 4-6 - 4 00SES AT THE SITE BOUNDARY RESULTING FROM THE POSTULATED MAXIMUM CREDIBLE ACCIDENT
- 4
, BWR PWR Nuclides** Nuclides** 00se Making 00se Making (rem) Major (rem) Major ORGAN Contribution Contribution Bone 0.69 SR-90 CS-137 2.04 SR-90 Liver 2.03 FE-59 1.07 FE-59 Whole Body 0.31 CS-137,CS-134 0.39 CS-137,CS-134 Thyroid 0.87 I-131 5.74 I-131 18 Lung 2.86 C0-60 3.62 C0-60,C0-58 C0-58,141-54
*The doses are based upon the asstaption that 167 Curtes (10% of the activity in the product The x/Q used is the 0-2 hour value for container) 1000 m from escapes from Regulatory the radwaste Guide 1.3, Figurebuilding 3 ground(.level release). The dose factors have been taken from Regulatory Guide 1.109 Table C-1. The breathing rate Was 20 m3/ day.
**The nuclide distribution in the escaping material is assumed to be that shown in Table 2-9.
5.0 Rim-1 3 SYSTEM BENEFITS 5.1 On-Site Benefits The envirormental and economic benefits of any voltme reduction system
- are very much a function of site specific considerations. However,
.__.certain qualitative assessments can be made with some validity on a generic basis.
I.' The RWR-1 volume reduction systen reduces the number of containers that mst be handled by the plant operating personnel and, therefore.
.some reduction in occupational radiation exposure can be expected. This N -
volume reduction will also resuit in a reduction in the amount of on-site radweste storage space required. . The volume reduction achieved with this system suggests that, in general, substantial reductions in the cost of radwaste disposal can be expected. In some cases, these cost savings can amount to over 50% of the total present annual radweste disposal costs. - 5.2 Off-Site Benefits
~
The reduced volumes of radwaste indicate a reduction in .the number of
.. shipments of radwaste to licensed disposal sites. This should therefore decrease the probability of radioactive releases to the environment due to transportation accidents.
s Volume reduction at the origin results in reducing the rate at which
~
available space will be exhausted at existing disposal sites. Volume reduction by an overall factor of approximately ten suggests that presently available disposal sites could continue to accept wastes for a longer period of time; perhaps doubling present predictions. Thus it would appear that the RWR-l* volume reduction systen has the potential for substantial, beneficial impacts upon the environrcental and economic aspects of radwaste disposal. 100
6.0 REFERENCES
- 1. Richard C. Corey, ed. Principles and Practices of Incineration.
Wiley-Interscience, New York,1969. p. 239.
- 2. R. E. Commander, et al. Operation of the Waste Calcining Facility with Highly Radioactive Aqueous Waste: Report of the First Processing Campaign. 100-14662, Phillips Petroleum Co., Idaho Falls. ID, June 1966.
- 3. W. A. Freeby. Interim Report: Fluidized-Bed Calcination of
- . Simulated High-Level Commercial Wastes. ICP-1075, Allied Chemical
- Corp., Idaho Falls, ID, June 1975.
- 4. U.S. Energy Research and Development Administration. Alternatives for Managing Wastes from Reactors and Post-Fission Operations in List Fuel Cycle. ERDA-76-43. USERDA, Washington, DC, May 1976.
~
V. 2, p. 6.13-6.34.
- 5. B. L. Perkins. Incineration Facilities for Treatment of Radioactive Wastes: A Review. LA-6252, Los Alamos Scientific Laboratcry, Los Alamos, let, February 1976.
- 6. U.S. Energy Research and Development Administration, op. cit.,
V. 2, Ch. 9.
,, 7. U.S. Atomic Energy Commission. Incineration of Radioactive Solid Wastes. WASH-1168, USAEC, Washington, DC, August 1970.
' 8. U.S. Atomic Energy Condssion, Of rectorate of. Regulatory Standards.
Final Environnental Statenent Concerning Proposed Rule Making t Action: Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Lew As Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Power j Reactor Effluents. WASH-1258, USAEC, Washington, DC, July 1973. i V. 1, Ch. 2. 101
- 9. U.S. Energy Research and Developme.it Administration, op. cit., V.
- 1. Sec. 2.2. I
- 10. A. H. Kibbey and H. W. Godbee. Critical Review of Solid Radioactive Waste Practices at Nuclear Power Plants. ORit.-4924, Oak Ridge l
National Laboratory, Oak Ridge, TN, March 1974. i i
- 11. U.S. Nuclear Regulatory Commission Effluent Treatment Systes Branch. Branch Technical Position 11-1: Design Guidance for
- Radioactive Waste Managment Systes Installed in Light-Water Cooled Nuclear Reactor Power Plants, Revision 1, Novaber 24, 1975.
.'.s.
- 12. American Society of Mechanical Engineers. Boiler and Pressure Vessel Code. Section VIII. Civision 1: Rules for Construction of Pressure Vessels. ASME, New York,1974.
- 13. Internation Atomic Energy Agency. Managenent of Radioactive Wastes at Nuclear Power Plants. (IAEA Safety Series 28). IAEA, Vienna, 1968.
- 14. R. H. Burns, G. W. Clare, and J. H. Clarke. Radioactive Waste Control at the United Kingdom Atomic Research Establishment. Harwell. AERE-R-5900. Atomic Energy Research Establishment, Harwell, England,1968.
- 15. International Atomic Energy Agency. 3 e Volume Reduction of Low-Activity Solid Wastes. (IAEA Technical Reports Series 106).
IAEA, Vienna,1970.
- 16. R. H. Burns and J. H. Clarke. "Harwell Experiences in Waste Management",
[ - in Management of Low- and Intennediate-level Radioactive Wastes, Proceedings of the Spiposium held at Aix en-Provence, Septaber 7-11, 1970. IAEA, Vienna,1970. p. 419.
- 17. G. E. Lohse and M. P. Hales. Second Processing Camoaign in the Waste Calcining Facility. IN-1344, Idaho Nuclear Corp., Idaho Falls, ID, March 1970.
102 -- . . - . - . ~ . - - -. . . - . - . - - - . - - - - - . - . - . - - - - - - - - . -
- a. ..
- 18. C. L. Sendixsen, G. E. Lohse, and M. P. Hales. The Third Processing Campaign in the Wasta Calcining Facility. IN-1474 Idaho Nuclear Corp., Idaho Falls, ID, May 1971. p. 23,
- 19. J. A. Wielang, G. E. Lohse, and M. P. Hales. The Fourth Processing '
Campaign in the Waste Calcining Facility. ICP-1004, Allied Chemical Corp., Idaho Falls, ID, March 1972, p. 25.
- 20. J. A. Wielang and W. A. Freeby. The Fifth Processing Campaign _
i
,- in the Waste Calcining Facility. ICP-1021, Allied Chemical Corp., Idaho Falls, ID, June 1973, p.19.
_ ~21. U.S. Atomic Energy Commission. Draft Environmental Statement: New Waste Calcining Facility, National Reacter Testing Station. Idaho. WASH-1531, USAEC, Washington, DC, January 1974. 22- R. R. Smith and J. A. Carter. Final Design Criteria for New Waste Calcining Facility (NWCF), (Allied Chemical Corp. Internal Report ACI-176, January 1976.)
- 23. Topical Report: Fluid Bed Dryer and Aeropeo Solidification System.
AECC-1, Aerojet Energy Conversion Co., Sacramento, CA, February 21, 1975.
- 24. Amendment 2 to Topical Report: Fluid Bed Dryer and Aeropep Solidification System. AECC-1, Amend. 2 Aerojet Energy Conversion
( Co., Sacramento, CA, September 30, 1975.
- 25. L. G. Gale et al. Final Report - Liquid Waste Calcination
~
l Feasibility Tests - Task I-C. Energy Inc., Idaho Falls, ID, February 1976. ,
- 26. J. M. Marchello and J. J. Kelly. Gas Cleaning for Air Quality Control. Marcell Dekker, New York,1975, p. 258-283, 314-319.
103 i i
- - - - - - + - , - -
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_ - - - - - , , - ---,,-,c- ------..--,-,.,.---.w--w -
t .
- 27. C. A. Burchsted, J. E. Xahn, and A. B. Fuller. Nuclear Air Cleaning Handbook: Design, Construction, and Testing of High-Efficiency Air Cleaning Systems for Nuclear Application. ERDA-76-21, Oak Ridge National Laboratory Oak Ridge, TN, March 1976, p. 64-72. .
- 28. Flanders Air Filtration Systems. Flanders Filters, Inc., Washington, NC, 1975.
, 29. C. A. Burchsted, op. cit., p. 42-43.
- 30. J. D. DeField and H. J. Ettinger. " Efficiency Testing of the Air Cleaning System for a High Temperature Reactor", in Treatment of Airborne Radioactive Wastes, Proceedings of the Symposium held in New York, August 26-30, 1968. IAEA, Vienna, 1968. p. 265.
- 31. R. Vance. Liouid Scrub Solution Study For Radiodine Retention (Proprietary). EI-77-4, Energy Inc., Idaho Falls, ID, June 1977.
- 32. U.S. Energy Research and Development Adminstration, op. cit., V.2, -
Sec. 13.2.
- 33. J. G. Wilhelm, J. Furrer, and E. Schultes. " Head-End Iodine Removal from a Reprocessing Plant with a Solid Sorbent", Proceedings of the 14th ERDA Air Cleaning Conference, Sun Valley, Idaho, August 2-4,1976, (CONF-760822, V.1), p. 447-477.
- 34. U.S. Nuclear Regulatory Comission. Regulatory Guide 1.3: Assumptions
- Used for Evaluating the Poteatial Radiological Consequences of a Loss l
. of Coolant Accident for Boiling Water Reactors. Revision 2. June 1974.
- 35. U.S. Nuclear Regulatory Commission. Regulatory Guide 1.109: Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I.
March 1976. l l 104
. .'. 1 APPENDIX A This appendix presents the derivations of the iodine activities found in Tables 2-2 and 2-3. The bases for these calculations are flow rates, concentrations, and decontamination factors taken from tilREG-0016(A1) and '
NbREG-0017.(A2) These values are for nominal 3400 MWt reactors, whereas the other values in Tables .2-2 and 2-3 are for nominal 3500 left reactors. j In light of the many assumptions made, and other factors which are not known with great accuracy, this difference is not felt to be significant. i The concentration of radiciodine in the domineralizer resin may be
, computed once the following values are known: radiofodine concentration
_.. in the reactor coolant, flow rate of coolant through the domineralizer, _._ _ _ iodine collection efficiency of the domineralizar resin, and the iodine decay rate. Separate determinations are given for BWRs and PWRs. Boiling Water Reactors Reactor water concentration at the nozzle where it leaves the reactor vessel: 2 x 10-2 uti 1-133/p 2 x 10~3 uCi I-131/ p (A1, p. 2-3) Cleanup domineralizer flow rate: 1.3 x 105 lb/hr (A1, p. 2-5)
- Reactor coolent demineralizar decontamination factor:
10 (A1, p. 2-36) The rate, G, at which radiofodine is removed by the resins may now be computed. For iodine-131: 105
l s. . l l l 5 G = (.9)(1.3 x 10 lb/hr)(5 x 10~3 uCf/gn) x ' (453.6gn/lb)(24hr/ day)(C1/106uCi) = 6.4 C1/ day and for iodine-133: G = (.9)(1.3 x 10 5)(2 x 10-2 uC1/gn) x 6 (453.6 gm/lb)(24 hr/ day)(C1/10 uti) = 25.5 C1/ day Assume that the bulk of the radiciodine appearing in the feed to the RWt-l* Systen has been collected from the primary coolant by the domineralizer resin.
. ___Let:
A = activity on the domineulizer resins _ _ __ _ _ and 1 = decay constant Then , h=G-AA describes the change of activity on the resins. At equilibrium, h = 0, l , therefore. G = 1A. f. (' The resin beds are typically changed cut several times a year, so the
, period between bed changeouts is measured in months, while the half-lives of I-131 and I-133 are on the order of days. Therefore, most of the iodine nuclides removed by the resins will have decayed away before the resins are removed. The activity upon removal will be the equilibrium activity, l A = G/1 106 l
~
I for I-131: A = (6.37 Ci/dayl(8.065 days) = 74.1 Ci I ni,2) and for I-133: - A = (25.5 Ci/ day)(20.8 hours) = 31.9 Ci In(2)(24 hours / day) Assuming that the resin bed is changed out once every two months, the e activity removed each year on the resins is six times the activity just computed. The rate is, theref. ore: . I . 445 C1/yr for I-131 191 C1/yr for I-133 Note that these are the rates for the resins as renoved from the demineralizer beds and no allowance has been orde for decay before processing in the R)R-1* System. Pressurized Water Reactors The logic for the fodine found on the resins in a PWR is the same as that for the BWR, above. The following data was applied for the PWR case, however.
, Concentration in primary coolant with U-tube steam generators:
2.7 x 10-I sci I-131/gm (A2, p. 2-3)
?
3.8 x 10"I uCf I-133/gm 107
. i, .
Concentration in primary coolant with once-through steam generators 2.7 x 10'I uCi I-131/gm (A2, p. 2-5) 3.8 x 10'I aCl I-133/gm Notice that the radiciodine concentrations are the same regardless of whether the reactor uses a U-tube or a once-through steam generator. C 7 Reactor coolant letdown flow: f , 3.7 x 104 lb/hr (A2, p. 2-7) Resin Oecontamination Factor: 10 (A2, p. 2-41) Using the same calculational procedures and assumptions as for the BWR, the generation rates from a PWR are: l 6803 Ci/yr for I-131 1037 C1/yr for I-133 I These are the generation rates for the resins as removed from the denineralizers. The half-lives of these two nuclides are so low that 5 even a few days' storage before processing will reduce the activity considerably. The greater generation rates for PWRs with respect to l e . BWRs are primarily due to the higher concentrations in the PWR primary
- coolant as opposed to the BWR coolant. This is largely a function of assumptions made about leak rates through the fuei cladding.
1 108
R
.. .+ ,
. t APPENDIX A REFERENCES A1. U.S. Nuclear Regulatory Commission. Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors (BWR-GALE Code), NUREG-0016, April 1976.
A2. U.S. Nuclear Regulatory Comission. Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized g Water Reactors, (PWR-GALE Code), NUREG-0017, April 1976. E w r spq T, . O , 5, C . , I-O 109
~ '. -
Natural Resources Defense Council,Inc.
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917 25TH STREET, N.W. WASHINGTON, D.C. 20005 so 737-5000 - , Westem OgEce
~Non,Terk Ofice 8345 YALz - October 25, 1978 ,,, , , , ,,,, 4,,, ,ra,7,7 FALO ALYO, CALIF. g(3o6 ,
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a . - s. U.S. Nuclear Regulatory.. Commission - s.y. - - Washington, D.C. 20555"J>. --
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.. y:. _:: . ., -
Attn: Secretary of the Commission . :..L.::.r.} .. ..
-] ~
Dear Sir,
Please provide me with any technical information
,~, you have describing the health ahd safety consequences of operating the low level waste incinerator planned for installation at Nine Mile Point Unit No. 1 in New York.
Have you, or do you intend to prepare an environmental assessment of this technology? If not, why not? If so, please send me a copy when it becomes available. Given the public concern over the installation of this incinerator, an environmental impact statement, or at least an environmental assessment would appear appropriate. t - Sincerely, i b b b.
~
Thomas B. Cochran O b TBC/ps - y.. .
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