ML20235J611
| ML20235J611 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/31/1987 |
| From: | Buchanan D, Eichfeld S GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20235J576 | List: |
| References | |
| 3527-006, 3527-006-R05, 3527-6, 3527-6-R5, NUDOCS 8710020030 | |
| Download: ML20235J611 (180) | |
Text
{{#Wiki_filter:' TER asn_oos ngy s ISSUE DATE September 1987 l 0 178- , D uen C NITS DIVISION TECHNICAL EVALUATION REPORT < l FOR Submerged Demineralized System 1 t COG ENG dONbA - PATE Bl27le7 RTR O ^b - M bATF. 691-?> 7 l 7 COG ENG MGR e DATE 8[
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5 E. E 3 E e71062003o 870928 E DR ADOCK 050 o 5 DOCUMENT PAGE 1 OF
No. ! e Nuclear 3527-006 Title Ter.hnical Evaluation Report for Pa ge of Submerged Demineralized System 2 Rev. *
SUMMARY
OF CHANGE Approval Date 0 Initial issue per GPU Nuc, lear letter 4400-82-L-0066. 4/82 1 Reissue per GPU Nuclear letter 4410-83-L-0122. 6/83 2 Reissue per GPU Nuclear letter 4410-84-L-0109. 7/84 Incorporates changes required by S-ECMs 1151 (Revision 0 through 3),1163 (Revision 0 through 3),1110 Revision 0, 1140 Revision 0,1159 Revisior, 0, and 1141 Revision 0. 3 Annual Update. 8/85 Incorporates changet made by S-ECM 1110 Revisions 0 and 1, ECAs 072, 042, 047, 041, and 102. 4 Annual Update. S/86 Incorporates changes made by S-ECM 1058 Revision 2, ECAs 041, 072, 087, and 312, 5 Annual Update. 9/g7 Incorporates updates to the valve and component lists; updatas hictorical data; deletes the Early Defueling DWC React 3r " ssel Filtration System which is no longer applit al e . O. 7 8 ~ ei E Y 6
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i l l 1 1 1 1 J TECHNICAL EVALUATION REPORT l l SUBMERGED DEMINERALIZATION SYSTEM l 1 1 i 1 i l 1 l l l r l l'
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TEP 3S27-006 CONTENTS- ., Chapter 1 Summary of Treatraent' Plan I 1.1 Project S: ope 1.2 Identification of Radionuclides and Radioactivity Levels q 1.3 Alternatives Considered 1.4 Description of the Decontamination Process 1.4.1 General 1.4.2 SDS Operating Description j Chapter 2 Summary of Health and Environmental Effects l 2.1 Occupational Exposure During Routine Operation 2.1.1 Exposure Planning 2.2 Exposures to the Public During Routine Operation of the SDS and 1 k EPICOR-II 2.3 Evaluation of Unexpected Occurrences l 2.4 Industrial Health and Safety 2.4.1 Public Safety ; 1 2.4.2 Occupational Safety 2.5 Non-Radiological Environmental Effects 2.6 Ultimate Waste Disposition i i e
TER 3527-006-Chapter 3 Process Description 3.1 Introduction l 3.2 Ion-Exchange Concepts
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3.3 Ien-Exchange Materials 3.4 Resin Selection Criteria l 3.5 Predicted Performance of Ion-Exchangers , l 3.6 Monitoring of Ion-Exchangers j 4 Chapter 4 Design Basis ! I l l 4.1 Introduction i 4.2 Components of the SDS Waste Processing System l 4.3 Submerged Demineralization System Criteria 4.3.! Design Basis l l 4.3.2 Process 4.3.3 Performance i 4.3.4 Capacity 4.3.5 Performance and Design Requirements 4.3.6 Piping System 4.3.7 Vessels and Tanks 4.3.8 Shielding Design 4.3.9 Leakage 4.3.10 Building and Auxiliary Services Interfaces 4.3.11 Controls and Instrumentation 4.4 System Operatioiial Concepts ,
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TER 3527-006 Chapter 5 System Description and Arrangement '{ 5.1 Demineralized System 5.1.1 Influent Water Filtration 5.1.2 Ion Exchanger Units 5.1.3 Leakage Detection and Processing 5.1.4 EPICOR-II i 5.1.5 Monitoring Tank System ] 5.1.6 Off-Gas and Liquid Separation System i 5.2 Sampling and Process Radiation Monitoring System j 5.2.1 Sampling System 5.2.2 Process Radiation Monitoring System 5.3 Ion-Exchariger and Filter Vessel Transfer in the Spent Fuel Pool 5.4 Arrangement of the Water Treatment System in the Fuel Storage Pool 5.5 Liner Recombiner and Vacuum Outgassing System Chapter 6 Radiation Protection 6.1 Ensuring Occupational Radiation Exposures are ALARA 6.1.1 Policy Considerations 6.1.2 0esign Considerations 6.1.3 Operational Considerations 6.2 Radiation Protection Design Features 6.2.1 Facility Design Features 6.2.2 Shielding ; I 6.2.3 Ventilation 6.2.4 Area Radiation Monitoring Instrumentation 111 L____-__________-_____________-______-____.-
TER 3527-006 l 6.3 Dose Assessment 6.3.1 On-site Occupational Exposures 6.3.2 Off-site Radiological Exposures Chapter 7 Accident Analyses l l 7.1 Inadvertent Pumping of Containment Water into the Spent Fuel Pool 7.2 Pipe Rupture on Filter Inlet Line (above water level) ! 7.3 Inadvertent Lifting of Prefilter Above Pool Surface 7.4 Inadvertent Lifting of Ion Exchanger Above Pool Surface 7.5 Inadvertent Drop of SDS Shipping Cask i Chapter 8 Conduct of Operatiens i 8.1 System Development 8.2 System Preoperational Testing 8.3 System Operations . l 8.4 System Decommissioning l References 1 l l l l l iv l l
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TER 3527-006 Appendix No.~1 - RC Processing Plan with the RCS in a Partially Orained Condition Appendix No. 2 - Internals Indexing. Fixture Processing System (Deleted) Appendix No. 3 - Fuel Transfer Canal Oraining System 1 Appendix No. 4 - Fuel Transfer Canal Shallow End Drainage System l Appendix No. 5 - Early Defueling DHC ReactorVessel Filtration System (Deleted) l t i i 4 a i-i l t i l 1 i I V k l l i i u )
TER 3527-006 { I Chapter 1 l 1 Summary of Treatment Plan 4
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1.1 Project Scope . l To date the SDS system has processed almost 4.5 million gallons of l contaminated water, including; 650,000 gallons of Reactor Building sump 1 water, 500,000 gallons from RB decon and 1,200,000 gallons of RCS l water. The continued decontamination of THI-2 includes the repeated processing of the IIF/RCS using the Letdown / Makeup Method or the Reactor ! Vessel Filtration System (DHCS). The activity level of this water is i given in Table 1.1. In addition, Reactor Building Decon water or water from other sources may be processed through SDS as necessary. This report describes the Submerged Demineralized System (SDS) and the work associated with the development of the system for the expeditious clean-up and disposition of the contaminated water mentioned above. Specific design features of the system include:
- 1. Placement of the operating system in the spent fuel pool to take advantage of shielding provided by the water in the pool.
- 2. Radioactive gas collection and treatment prior to release.
- 3. Liquid leak-off collection and treatment.
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- 4. Underwater placement of ion-exchange vessels into a' shipping cask-without removal from the spent fuel ~ pool.
-5. Use of existing EPICOR-II~ equipment for polishing of SDS effluent,.
as. required. ; I 1.2 Identification of Radionuclides and Radioactivity Levels i
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Water samples were taken from the reactor coolant system and the-containment sump, and were analyzed to identify specific radio'nuclides and concentrations. Typical results are listed in Table 1.1'. The Reactor Coolant System (RCS) and containment's' ump specific radionuclides and concentrations are based upon actual sample data ~taken. The RCS i activity decreases due to radioactive decay and leakage from the RCS. i However, RCS' activity may increase during processing shutdown due to leaching. 1.3 Alternatives Considered i i During the early phases of developing a system for the control, clean-up, and disposition of the contaminated water located in the containment building of THI-2, several methods or alternatives were evaluated. These alternatives were grouped into two categories: I (1) -those with no volume reduction, and (2) those with volume reduction. , [ 0398B/LC l 1 : l~ _J
TER 3527-006 Presented beloc, are the alternatives considered with a discussion and conclusion about each. Alternative I: Leave Contaminated Water in Containment Indefinitely (No Volume Reduction) Discussion i A. Containment Sump Water
- 1. The sump water'contains radionuclides concentrations as depicted in Ta01e 1.1. The existence of this may cause some l l
increase in radiological exposure problems during the recovery program, i.e., increased exposure to recovery program personnel, increased contamination levels, and increased possibility of airborne radioactivity.
- 2. The presence of the contaminated sump water would prevent ;
decontamination of the lower levels of the containment building. l l B. Reactor Coolant System Water i The presence of the contaminated water in the reactor coolant ! I system would inhibit disassembly of the reactor and impede defueling operations. 1 ! l l 039BB/LC. j u __--._--_______---_--_-J
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Conclusion:
Alternative I is not deemed feasible for the following reasons:
- 1. The potential for increased personnel exposure exists. Therefore, compliance with the principles of ALARA is not possible.
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- 2. Facility decontamination and defueling operations are seriously inhibited or perhaps prevented.
- 3. Continued storage of the contaminated water in the containment sump for increased periods of time increases the probability that leakage from the building may occur. Leakage of contaminated water from the reactor building surp may threaten the public health and safety.
- 4. Continued storage of the water in the containment building for an l extended period of time is undesirable. The primary isotopes of concern (Cs-137 and Sr-90) exhibit decay half-lives of l approximately 30 year. Storage in the containment sump for l
approximately 300 years would be required for 10 half-life decay. Maintenance of containment integrity for this interval of time cannct be assured. Alternative II: Transfer Water to On-site Storage Facility (No Volume Reduction) 0398B/LC
TER 3527-006 Discussion:
- 1. To safely contain the contaminated water, the construction of an on-site liquid radwaste storage facility would be required.
- 2. Additional radiation areas on the plant site would be created if a liquid radwaste storage facility were built.
- 3. Estimates indicate the construction of a liquid radwaste storage l
facility would require two to three years, at a minimum. l
- 4. A liquid radioactive waste transfer system for the transfer of the contaminated water from the various locations to the waste storage complex would be required.
- 5. Handling and pumping operations tray involve leakage and the spread of contamination.
- 6. Disposal of the water prior to natural decay is required because of I the long radicartive decay half-lives. This alternative is not representative of an acceptable long-term solution. I l
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Conclusion:
Based on the above discussion, Alternative II is not a feasible method. Alternative III: Solidification and Disposal (No Volume Reduction) 0398B/LC L------------------------_ _ - - - - - - - - - _ - - - - _ _ _ - - - - - . - -
TER 3527-006 i Discussion: I
- 1. The construction of an on-site solidification facility would be !
required. l l
- 2. Based on 1,000,000 gallons of contaminated water originally to be processed, a 30-gallon availability of water volume in a 55-gallon drum, 70% availability, 24-hour / day operation, and a 45 minute {
l cycle time, the processing time may exceed four years.
- 3. Based on 1,000,000 gallons of contaminated water originally to be I processed and a 30-gallon availability of water volume in a 55-gallon drum. The number of drums of solidified waste that would be generated would exceed 33,000. Handling, transportation and disposal of this extremely large quantity of solidified waste would be prohibitively expensive and violate basic principles of minimizing radioactive waste volumes.
- 4. The handling evolution required to solidify the contaminated water may involve substantial radiation exposure to personnel.
l l S. The potential for leakage and contamination problems may be substantial in operating a solidification facility for processing this contaminated water in this manner.
Conclusion:
Based on the above considerations, Alternative III is not considered to be feasible. 03988/LC l ] l l l
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'TER 3527-006' hiternativeIV: Submerged Demineralized System (SDS) in the'"B" Spent-'
Fuel Pool'and EPICOR-II System (Volume Reduction) Discussion:
- 1. The system'would be' capable'of. concentrating fission' products on a medium to effectively remove those products from the water.
- 2. Processing contaminated' water would result in concentrated waste requiring additional shieloing.
- 3. The system' incorporates remote operability: features.
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- 4. Design, construction and operation would allow for: relatively short -
t lead times. I
- 5. The system would require minimal maintenance. l
- 6. The 505 is amenable to location within the Spent Fuel Pool which would utilize the shielding capability of the pool water.
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- 7. Containe:s of highly loaded ion exchange media arising from operation of the SDS would not be acceptable at shallow land 1
disposal sites. The SDS design and selection of' ion exchange media j allows volumes of such highly loaded media to be minimized.to permit interim storage and probable ultimate disposal ~ in a geological repository. It is b?lieved that the EPICOR-II liners, .. generated as a result of polishing the SDS effluent, will be- a suitable for shallow land disposal because of their low curie i content. l 0398B/LC l 1 \' > r oa-...; e .n .
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- 8. The EPICOR-II system, used in conjunction eith SDS, cill provide the capability to remove trace quantities of radionuclides from the SDS effluent.
== Conclusion:== Based on the above considerations, Alternative IV is an acceptable method for decontamination. Alternative V: Evaporation (Volume Reduction) Discussion: j l
- 1. Evaporation would req" ire the design and construction of a new facility.
- 2. Due to the nature of the contaminated water to be processed the design of the facility would be complex to allow for maintenance of tne processing system and personnel radiological protection. The construction of the facility may require at least four years.
- 3. Evaporation provides the ability to process a wiCe range of ;
i chemical contaminants. I 1 l
== Conclusion:== Evaporation is an acceptable alternative for processing the contaminated waste waters. Based on the long construction time'of the facility and inherent potential for higher occupational exposure due to increased maintenance requirements, this <ernative is less desirable than Alternative IV, Submerged Demineral r System (505) coupled with the EPICOR 11 system. 0398B/LC
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TER 3527-006: 11'.4 Description of'the Decontamination Process
- 1. 4.1 - General
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Analysis;of the alternatives previously presented has're'sulted-
'in the. determination tnat,.of the two alternative categories considered.. volume reduction is. appropriate for the' disposition of contaminated > water. This conclusion,was reached' based on the. considerations'that ' volume' reduction:
' 1. . fixes the contaminants
- 2. concentrate's'the. activity. .
- 3. minimizes storage and disposal space Of the volume reduction category, the Submerged Demineralized system (505) in conjunction'with EPICOR II for final polishing, or Alternative IV, was chosen as the'most appropriate process for the following reasons:
- 1. Basic design simplicity.
- 2. High performance for decontaminating liquids, i.e.,
decontamination factors up to 107 , or higher.
- 3. Amenable to placement under water to take advantage of shielding properties of the water 1
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-4. Ability td implement cater processing in a timely' fashion
'for_ support of the overa11' objective of fuel removal. ,
- 5. Ability to use existing' proven plant structures,.
equipment:and technology.for containment of-the pro:essed: water.and final process polishing.(EPICOR-II) :
- l The 505 with EPICOR II is an ion-exchange process expected-to provide decontamination factors of up to 107 forfcesium and '
105 for strontium.(see Table 3.1). thus removing the
. majority of the; activity from the water prior to placement.in' the Processed Water Storage Tanks,'or usageEfor continued-1 decontamination or makeup to the RCS.
1.4.2 SOS Operating Description Figure 1.1 shows a block diagram of.the process flow of the Submerged Demineralized System'(SDS) with the EPICOR II System. Radioactive water enters the SDS via the RCS manifold. This source of water can pass through two cartridge or sand type filters for removal. of particulate-matter. Sample connections are provided on the influent and effluent of the filters, and influent to the ion-exchange' system to determine radionuclides content;and concentrations of the water to be processed.
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'The.first part of the SDS ion-exchange system consists of up.
to six underwater vessels (24 1/2 in. x 54'1/2 in.)'. Each vessel contains approximately 8 cubic fee.t of homogeneously m'1xed IE-96 and LINDE-A zeolit'e ion exchange media? Zeolite: media volumes and mixtures may be changed to reflect different processing scenarios (The resin mix is specified by. { Radiochemical Engineering on the form. included in OP 4215-OPS-3527.16). Inlet, outlet, and vent connections are s
'made with remotely operated coupilngs. The vessels ~are l
arranged in.two parallel trains with three columns in each 1 trdin. Flow may be directed through one train of!three- I vessels or through,both trains in parallel. Loading of the vessels will be controlled by feed batch size, residence time, 1 influent and effluent. sample analysis, and continuous monitoring. , i The second part of the SOS-ion exchange system consists of two. i parallel sand filter Vessels located underwater and immediately downstream of the zeolite beds. These sand filters will contain a mixture of sand and-are intended to remove system effluent particulate, primarily zeolite fines. The columns are intended to be operated singly. i Present SDS operations are envisioned to provide for l radionuclides loading of the zeolite media to a maximum of 134 137 60,000 Ci of Cs and Cs at the time of shipping. 03988/LC l L- _ . _ _ _ _ _ _ _ _ _ _
o TER'3527-006~ Thisfloading level)is based on' restrictions imposed. based on-the shielding provided by the Chem-Nuclear 1-13C II shipping cask. From the point of. view of minimizing waste volume
. generation it is desirable ixrload the zeolites to these higher levels.
When the desired bed loading.is achieved ~onLthe'first bed'of the' train, the feed flow to the' train will be. stopped, the bed' will be flushed with' clean water, and'the first bed will:be disconnected and moved to the storage rack in thel spent feel pool:using the pool area crane. .The.second and third' beds will be disconnected, moved to the first and second positions, respectively. A new lon exchanger vessel is then installed in the third. position. Following. installation of the new ton-exchanger, the treatment of the contaminated-water will recommence. This operational concept, which is the. currently. intended mode of operation, has eliminated the potential for. valving errors.and also minimizes the possibility of an unexpected radionuclides'" breakthrough" which could decontaminate the water already processed. This mode of operation may change if the processing scenario changes. Additionally some processing operations will require fewer than three (3) ion exchange units per train to achieve desired decontamination factors, in these cases jumpers will be installed to bypass the unused positions. l
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TER.3527-006: When the.SDS is' processing con %aminated sump cater, the effluent from the " cation" sand filters can~be sent to. EPICOR-II.for polishing. When processing reactor coolant the~ effluent may be routed to. installed tankage for injection back into the Reactor Coolant System as a source of makeup or..to EPICOR for polishing. The spent ion-exchangers and filters of SDS will be retained under water in the spent fuel pool until removed. ~To transport-spent ion-exchangers, they will~be bulk 1 dewatered,' vacuum dewatered, and catalyst recombiner added, and loaded into shleided casks while under water and removed - from the spent fuel l pool. Following decontamination of-the l cask surface, the cask can then'be loaded.onto a trailer ~for- i j transportation. l i
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TER 3527-006 TABLE 1.1 Typical Results of Analysis from the Reactor Coolant System Water and the Containment Sump Water Radionuclides Concentraticas (pC1/ml) Reactor Coolant RB Sump Isotope System Decon Sampling Date (4/87) (2/87) H3 0.038 0.12 Sr-90 1.1 0.11 Sb-125 0.039 0.015 Cs-134 0.003 0.082 Cs-137 0.16 3.5 l pH 7.61 7.48 Boron 5340 ppm 3297 ppm f l l Na 1560 ppm 548 ppm 1 l I i 0398B/LC
I TER 3527-006' i m Chapter 2 Summary of Health and' Environmental ~ Effects-2.1 ' Occupational Radiation Exposure During Routine Operation The SDS has.been designed to maintain radiation exposures _to operat'ing personnel as low as reasonably achievable. To implement the.ALARA-concept, the following features have been incorporated into the SDS design. o Shielding has been designed to limit whole. body dose rates in operating areas to less than 1 mrem /hr. The filters and ion-exchangers are located approximately 16 feet underwater.for j shielding. Components and piping carrying high activity water not-contained underwater in the fuel pool have been provided with' shielding to maintain external dose rates to acceptable levels. o Controls and instrumentation are located in low radiation areas.
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o Components containing high activity water have been designed for . 1 venting exhaust gases to the SDS Off Gas System. The Off-Gas System will minimize the potential for excessive airborne-radioactivity releases in the work areas and.to the environment.
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Additional design and operational ALARA features are given in Section 6. t 0398B/LC l
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'The~ occupational exposure for the EPICOR-11 system eas assessed in NUREG-0591. .The occupational radiation exposure for'the EPICOR-II system will be lower for the processing.of the effluent from the SDS than previously processed by EPICOR-II since the influent activity to the EPICOR-II from the SDS has been substantial.ly reduced by processing the' radioactively contaminated. water through'the 505.
2.1.1 Exposure Planning Several activities will..be implemented prior.to and shortly after, the SDS start up to assure occupational exposures.are minimized. These activities include:
-o Review of operating, maintenance and surveillance procedures to assure precautions and prerequisites are adequate.
1 o Review of the installed system to identify potential' problems during operation and the implementation of corrective actions.. o Operational evaluations during preoperational testing and system training will be performed to update exposure estimates. o Determination of radthtion dose rates during-normal operations and maintenance evolutions will be performed. 0398B/LC
TER 3527-006 As these reviews are completed, operating and surveillance frequencies can be established; total occupational exposures can be updated for the various activities during SDS operation. Th h exercise' will permit review of those activities estimated to yield the highest man-rem expenditure. Pre-examination to assure that every reasonable effort is l 1 expended to minimize personnel exposure may include the following j considerations: o Reduction of the frequency of operation o Temporary or additional shielding o Tool modification o Procedure modification o Personnel training to reduce work time o Component modifications 2.2 Exposures to the Public During Routine Operatio_n of the SDS and EPICOR-II Refer te Chapter 6 for information on exposures to the public from routine operatica of the SDS and EPICOR-II processing. 2.3 Evaluation of Unexpected Occurrences i The radiological assesstant of unexpected occurrences includes the analysis of five hypothetical accidents that are pott 01ated to occur during operation of the system. i l 1 i 03988/LC ! l \ L
1 TER 3527-006 ! The first accident is an inadvertent pumping of RCS water into the fuel storage pool untti a total of 225 gallons of radioactive water is released to the pool. No exposures occur to the-public since the contaminated water is contained in the pool. The maximum exposure rate at a distance of six feet above the pool surface is estimated to be 4.2 i mR/ hour. Since the release of water occers underwater, no significant internal exposures are expected for workers. The primary impact of the accidsnt is the contamination of water in the Spent Fuel Pool (223,000 gallons). (Refer to Section 7.1) I ihe second hypothetical accident assumes a pipe is ruptured end RCS water is sprayed into the building and fuel storage pool. It is j possible that workers could be contaminated, however,-prompt implementation of emergency procedures would minimize radiation exposures. The radioactive materials would be contained within the building except small amounts of radionuclides that would become airborne and subsequently be released through the inonitored station discharge. This dirborne radionuclides release would not result in significant exposures to the public. (Refer to Section 7.2) The third hypothetical accident evaluated considers the inadvertent falsing of a loaded prefilter above the pool surface. The de,te rate at a distance of 15 feet from the source is estimated to be 21 Rem / hour and could result in a dose of approximately 1.8 rem to workers who remain in the area for a five minute period. (Refer to Section 7.3) 0398D/LC
LTER3527-006- l 1 Theifourth' hypothetical accident evaluated considers'the inadvertent raising of a loaded zeolite _ ton exchanger;above the pool surface. The dose rate at'a distance of'70 feet from the source is estimated to be t i approximately 340 Rem /hr. .(Refer to Section 7.4) l The final hypothetical accident considers;the inadvertent drop of the I SDS shipping cask containing a. loaded zeolite ion exchanger. The SDS' shipping cask is assumed to be dropped from'the maximum height of'the fuel. Handling Building crane to the EL 30S' floor. The dose rate resulting from a complete.-rupture of the SDS shipping cask at'a distance' i of 20 feet is approximately 340 Rem /hr and assumes rupture of both the cask and the. vessel. The'small amounts of radionuclides assumed to betoms airborne would not result in significant exposures to the public. Also there would not be a'significant effect from direct i radiation exposure to the public. (Refer to Section 7.5) l The evaluation of unexpected occurrences for the EFICOR-II system was l analyzed in NUREG-0591. The potential releases from processing SDS effluent water will be significantly lower because of the lower concentration of water being processed through EPICOR-II from-the 505. i (See Table 3.1) l 1
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TER 3527-006. I 2.4 Industrial Health and Safety i
.l 2:4.1 Public Safet,y.
Operation of the Submerged Demineralized System poses no risk i from an industrial safety standpoint to the general pt'blic for the following reasons: 1 i 1
- 1. Lifting and handling activities described take place within the THI complex.
l 2. Hazardous chemical species, flammable or explosive substances, heavy industrial processes, and concentrated manufacturing activities are not involved in the installation or operation of the SDS. 1
- 3. No toxic substances are used in the SDS. !
2.4.2 Occupational Safety 4 l During the operation of the SDS, operating personnel will
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adhere to station requirements for occupational safety. Structural equipment and operating equipment used shall meet ! l Occupational Safety and Health Administration requirements as applicable. Personnti protective equipment that would be required for the operation of the SDS will be utilized in j accordance with standard station procedures. 0398B/LC
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'2.5 Non-Radiological Environmental Effects j I
I Adverse. environmental effects from the construction and operation of the SDS are not anticipated. The system will be installed and operated in l ( an existing, on-site facility and thus will not require any change in j i land-use. Adaittonally, the system is designed in such a manner as to ! I allow zero discharge of liquid effluents to receiving waters. The final l disposition of tbe processed water will be determined at a later date. l l Solid wastes (spent ion-exchangers, etc.) generated by the SDS will be l i stored and held uritil final disposal is accomplished. )
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I i l 2.6 Ultimate Haste Disposition ! l l Radioactive material generated as a result of the accicent at TMI is j I currently restricted to disposai at the commercial disposal site operated by U.S. Ecology at Hanford, Washington. SDS vessels meeting the criteria for disposal at this site will be disposed of by shallow land ourial at this location. SDS vessels not meeting the Hanford Site I criterte will be classified as abnormal waste and disposed of by the l Department of Energy in accordance with the Memorandum of Understanding dated July 15,1 Al, between the Nuclear Regulatory Commission and the Department of Energy Naling with the disposition of solid nuclear waste from the cleanup of TMI Unit 2. l l l 1 i 0398B/LC
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TER 3527-006 l Chapter'3 I l Process Description 1
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3.1 Introduction l l A combined filtration-ion. exchange process has been selected as the' method for. treating radioactive water contained in the reactor coolant , system and containment building. The filter ion-exchange method has
.been used successfully to reduce l quantities of radionuclides in the :j process effluent.to levels that.are in compliance with'10 CFR 20 and.10 l .i
'CFR 50. 4 Furthermore, experiments conducted at ORNL, documented in ORNL report TM-7443, provide evidence that SDS processing, followed by EPICOR-Il-polishing, should provide an effective method for wa+er decontamination.
The initial processing of the waste water is filtration for the removal of solids to optimize the subsequent ion-exchange process. Filtration is believed to be necessary to protect the zeolite beds from particulate in the sump and RC3' water. After filtration, radioactive ion removal from the waste water involves the use of ion-exchange materials. The two or three ion-exchange'
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l columns (per train) contain homogeneously mixed inorganic zeolite l material which effectively removes essentially all of the cesium and 03SBB/LC '^
TER 3527-006 much of the strontium. Other trace levels of: radionuclides are alsol partially removed by.the zeolite media. 'The radioactivity content in. the effluent stream of each bed is used to_ determine when the bed is: expended and replaced. Final demineralization of the contaminated; sump water and selected batches of RCS water is intended to be by the EPICOR-II system. Essentially, all remaining radionuclides excluding.t'ritium are expected to be removed from the water during this process step. 3.2 Ion-Exchance Concepts Ion-exchangers are solid inorganic ~and organic materials containing. exchangeable cations or anions. When solutions containing tonic species are in contact with the resin, a stoichiometrically equivalent. amount of-ions are exchanged. As an eyample, an ion-exchanger in the sodium (Na+) form will " soften" water by an ion-exchange process. Hard water containing CaC12 is " softened" by this exchange mechanism which reraoves the Ca" ions from solution and replaces them with Na+ ions. In a similar manner, Sr" and Cs* tons are exchanged with the - Na+ ions from the solid zeolite material. Characteristic properties of lon exchangers involve micro-structural features contained in a framework held together by chemical bonds andfor lattice energy. Either a positive or negative electric surplus charge is carried within this framework which must be compensated for by ions of opposite sign. Because the exchange of ions is a diffusion process-0398B/LC
l TER 3527-006 within the structura' framework, it does not conform to normal chemical reaction kinetics. The preference of ion-exchangers for a particular specie is due to electrost& tic interactions between the charged framework and the exchanging ions which vary in size and charge number. The decontamination factor (DF) is the ratio of the concentration in the 1 influent stream to that in the effluent stream and is used for determining the efficiency of a purification process for radionuclides removal. The fellowing equation is a qualitative expression for the removal of a singis ionic specie from solution. DF - _, 1 1 - Kn0Ew CV f where: Q - Total exchange capacity (meq/ml wot resin) n - fraction of Q used E, = Equivalent weight of the nuclide under consideration C f a Nuclide concentration (weight / volume) V - Feed throughput (number of ion-exchange bed volumes) K - Unit conversion constant Important variables which are considei6d as part of the evaluation of ion-exchangers for decontainination are ion exchange media type, selectivity and capacity, concentration of the species to be rembved, total composition of the feed stream, and the preserce of contaminants. Operating parameters such as resin bed size, flow rate, flow distribution, pH, and temperatures cre spitified for the ion-exchange beds in order to maxim 1ze removal of the contaminating 16ns. 0398B/LC
q TER 3527-006 ;
. Specifications.which have'been defined for this purification ( process includei (1) The flow rate to provide an acceptable residence time ~for ton diffusion j and exchange to occur.
(2) The cross-sectional area of the lon-exchange media.to provide an- j i acceptable linear velocity through the bed. i l
;(3)- The bed depth to result in an acceptable press'ure, drop. i 1
(4) A uniform flow distribution and a uniform media distribution to reduce the potential for~ channeling. (5) The ion-exchange media bead size te minimize atrition and large pressure I I drops. (6) The curie loading to satisfy personnel exposure, radiation damage, transportation, and storage regulations. l
.j (7) The cation form and the amount of lon-exchange media impurities to maximize removal of specific nuclides. ;
3.3 Ion-Exchanae Materials The ion-exchanger media selected for use in this processing system are - an inorganic zeolite material that is commercially available and known as Ion Siv IE-96 (Na* form of IE-95), and LINDE-A, to be used for SDS and cation and anion resins to be used in EPICOR II. . ; I I 03988/LC j 1
\
a TERI3527 006
-Zeolites are aluminosilicate with frametork structures enclosing large.
~
and uniform cavities .Because of their narrow, rigid, and. uniform p' ore-size, they can'also act as " molecular sieves" to.' sorb small molecules,
~
but to exclude molecules that are larger'than the opening in the crystal framework. Other. media'are also being evaluated. Should'our plans change with. regard to ion exchange media to be employed, th'e-NRC will be/ notified, ! 1 l Organic ton exchange restns are typically geis and are_. classified as 1 cross-linked polyelectrolytes. Their framework, or matrix, consists of an irregular, macromolecular,.three-dimensional network of hydrocarbon.: l chains. In cation exchangers, the matrix carries ionic groups such as 50". C00~, (P0 )", and in anion exchangers groups such as 3 2 NH+, 4 Na+, H+ are carried. The framework of the organic resins,. in contrast to that of the zeolites, is a flexible random' network which-is elastic, can be expanded, and is made insoluble'by introductier, of 1 cross-links which interconnect.the various hydrocarbon chains. The extent of crosslinking establishes the mesh width of the matrix and, l th0s, the degree of swelling and the ion mobilities within the resin'. This, in. turn . determines the ton exchange rates and electric conductivity of the resin. l u Since the mechanism of the ion exchange process involves the stoichiometric exchange of ions between the exchanger and the-solution- ! while electrical neutrality is maintained, the rate determining step is 1 controlled by the interdiffusion of ions within the framework of the. . i i 0398B/LC'- ! 4 l, J
TER 3527-006 l1
'lon-exchanger. Since the: rate of lon exchange is' determined by diffusion ~pr'ocessec, rate laws are derived by. applying well-known- I 1
diffusion equations t.) ion-exchange systems. However, complications 1 j arise from diffusi.;n-induced electric forces, from' selectively specific interactions, and changes in swelling such that rate laws are applicable for only a few limited cases, Experimental efforts have been conducted' at the Savannah River Laboratory to ir.vestigate the kinetics of. cesium ' and strontium ion-exchange with the, zeolite exchanger. Cesium was absorbed so rapidly that only rough estimates of the diffusion parameter could be obtained. The resulting equation, used to calculats. column: pe,formance, did not involve kinetic parameters but was suitable-to ) described the equilibrium column behavior. y l i-3.4 Resin Sele,ction Criteria. ., j
)
-Technical information obtained from previous use of various ion-exchange materials and the results Of recent experimental work with simulated and actual water samples from Three Mile Island were used to' support the selection of tpecific ton exchange materials for this processing system. The performance of an ton exchange system is, controlled by the physical and chemical properties of the exchange material.as well as by the operating cenditions specified in Section 3.2. The important criteria which were used in the ton exchanger selection process included:
(1) Exchange capacity (2) Swelling equilibrium (3) Degree of crosslinking (4) Resin particle size 27 - 0398B/LC j
r
' ,,, TER 3527-006 j l
(5) ' Ionic selectivity j (6); Ion-exchange.' kinetics (7) Chemical,.radiolytic and physical stability (8) ' Previous demonstrated performance (EPICOR-II)- : l Experimental studies with reactor coolant' water have been conducted to supportandverifythe.selectionofthesefon-exchangers;;referjto'ORNL-J TW7448. Further, onsite studies'have been performed'to support and verify selection of the ion-exchange media. The. decontamination factors i for the major contaminants were' measured using a number of chadidate ion-l l exchangers including the organic resins, HCR-5 and. SBR-OF, and'the zeolite ION SIV IE-96 and LINDE-A. The results indicated the most favorable type of ion exchange media to be'used in the cle'nup a process ; weretheavailablecation-antonresinsincombinationwiN~thezeolite l exchanger. Furthermore, as a result of processing in excess.of 4,400,000 gallons of l radioactively contaminated water from the Auxiliary Building, Reactor Building and RCS, we are confident that the SDS, with EPICOR-II used as a polishing system for treatment of SDS effluent, will continue to provide an effective means to decontaminate the contaminated waters. EPICOR-II resin loadings may be altered to improve. polishing effectiveness, if required. 28 - 0398B/LC l L _ _-- _ _ -_-___-_- __ _-- _._ _ _ - _ - - - .
T1R 3527-006 3.5 Predicted Performance of Ion-Exchanger _s The concentrations of radionuclides in samples of water from the Reactor Coolant System have been measured. Those radionuclides still detectable in June,1984 include Sr-90, Cs-134 Cs-137, and Sb-125. The expected performance of the SDS ion-exchangers, and the EPICOR-II lon exchangers is shown in Table 3.2. The concentrations of strontium and cesium are expected to be significantly reduced by processing through the SDS and EPICOR-II system. Table 3.1 is included to provide historical data on Reactor Building Sump water' processing. Antimony is expected to pass through the SDS ion exchangers and will end l Up as the predominant gamma emitter in the solutior, entering the EPICOR-II system. The Concentration of Sb-125 in the containment ! building sump sample is approximately 0.015 microcuries per milliliter. l l 3.6 Monitoring of Ion Exchar.gers Methods which may be; used to monitor the effectiveness of the ion j exchangers include liquid tampling and in-line radiation detectors. 1.iqu!d samples of feed and effluent streams can also be used to I establish the approximate curie 3oadings in the loaded beds. ! l \ ( 0398B/LC
1 y i j] 1! j Il!l ,l] iI n A A
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5 7 L. _ u R - E' - . X l O 8E06E876 4 C f f I 83.1 E 4.EEE - - - 6 5 E P 02<I 322I 0 E a h 42434 t - - - - - r 853EEEEE u 01 1. .00 o 8.E E F 02S41 1 1 4 _ ) _ s s _ m m _ a r _ e u - _ rd l 342434 _ te d - - - - - - sb oc r 65EEEEEE _ 6 se st e i h T
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- t
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a 3sonr o r S u o t S 021 41 1 1 4 Eoh h i t i e 0 _ Litt: t B a o l .l p 3 _ B R r e Aas w m - Tre o - t Z a. ~ t ml n s. nuf l e el a c l co s c n sl nv o ea _ o u i 4201 4 o r c t - - + + - l r - cd u o s 85 EEEEE pe yb e n t t '.
- 8. E 501 750 mv t
i s n e 1 F 02241 1 1 4 ao s t i i0n H u o v0o i2c l f it a t f ue cr n E t u ae o cd lf aa u t t d e s a r 1 2
+ +
a 'y . np ooc c B e EE ds A ( t l 8 8 2 93
- 32 eo sr i 0. ' at F 0b5bb1 1b bc _
2p e. 8s 9 1 a 1 ym m 1 2 ra
+ + ag EE u ry d 8 2 93 e 8 - 0 32 bb e e F 0b5bb1 1 b Fe _
l fb oa i sf ai _ t l n ma
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e i q _ d i ubsse tt _ l c orRSCCC CS65474 I
" o N u 0002334 N 36 91 1 1 1 1 b .
l' ;iliI,ll ' L
TER 3527-006
,. .. TABLE 3.2-Actual activity concentrations .in.SDS' a process streams after.200 bed volumes.through each zeolite-bed-(Based on continuous flow through two_ zeolite columns)'
RCS Processing ;q - 3 Effluent' concentrations. a UCi/ml. Zeolite columns Nuclide Feed filter Fi r s t '- Second Sand Filter'- 60Co <2.0E 2.2E-3 1.2E < 1.6E-4' <2E-4 ' 90S r .3.4 3.1 .0.084 2.8E-3 3.0E-3 106Ru 2.3E-2 <2E-2 .<5.2E-3 <1.5E <1.7E-3' 125Sb 0.16 0.15. 0.15- 0.14 0.15. 134Cs 0.025 0.023 1.2E <1.1EL4 <1.2E-4 137Cs 0.56 0.51 3.0E <1.7E-4 (l'.6E d 144Ce <1.2E-2~ (1.2E-2 <4.5E-3 ' <1.8E-3. ' <2.0E-3 a In pCi/mi as of June 1984 based on actual samples.. b Not quantifiable by gamna spectroscopy due to overall' sample activities. I
')
1 ; 4 1 l l 0398B/LC l l l l
r- - TER 3527-006 Chapter 4 Submerged Demineralized System Design Basis 4.1 Introduction The Submerged Demineral'ization System (SDS)'is an underwater ion-exchange system'which has'been specifically desig'ned to process-higher-level waste waters *, with inherent system features for reduction of occupational and environmental exposures. The SDS is submerged in the spent fuel pool (1) to provide. shielding'during operation,'(2) to-permit access to the system during demineralized changeout; (3) to: minimize the hazard from potential accidents, and (4) to utilize-an existing Seismic Category I facility. In conjunction with the SDS, the EPICOR-II system may be used to provide final polishing of the SDS effluent water for removal of trace quantitles of radionuclides. Design features for SDS include:
- 1. A prefilter and final filter in series', followed by two parallel trains of 2 or 3 zeolite ion-exchangers in series. These j ion-exchangers are followed by two " cation" sand filters in i parallel followed by the EPICOR-II equipment. This combination of filters and ion-exchangers achieves the desired process flow rates 1
and decontamination factors (DF's).
- Higher-level waste waters are those contaminated waters having gross activity; concentrations in excess of 100 pCi/ml.
0398B/LC
^
TER 3527-006-r
.2. . Series operation, logic that allocs-for sequencing the :
.c o demineralization units to prevent activity breakthrough in thef final zeolite bed. and maximize activity loading on spent beds to accompitsh the best possible activity concentration.
~
The' design objectives are as follows:
- a. A totally integrated system that is as independent.as possible'from existing waste systems at the Three Mile Island plant. The SDS is j a temporary system for the recovery of TMI-2. '
l 1 1'
~
! b,- A system that has the capability to, reduce the' fission product' concentration in the contaminated water and has optional. , capabilities for removing chemical contaminants to permit future disposition of the concentrated waste form; s
- c. A system that could be' operated with' a minimum of exposure to personnel and a negligible risk to the public. '
- d. A system that could accomplish the objective listed above in a- _
timely and cost effective manner. j
- e. A system that incorporates known and demonstrated processing equipment, materials and techniques. (EPICOR-II).
)
i l4 0398B/LC . 1 i _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1._ .U
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L4.2 Compogentpof Mf SDS Waste Pdocessic$ Systemi , r . . .. /.
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' The SDS f(compWed of, the Tollowing(cEnponent.s,1all of which 'will,b'e,
' ' ,f * * , q IQw s. ~,
located in the Ua'N12 E fuelIpool, .or. iiithehar91 din 1ty of.the B fuel
\- 5 .f
' pool.,L(See Figure 5.6, General Layout Plan.')
?l 5f >
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- 1. Fe"ed filterin'g system; :+
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p " 2 yy[f,'paralOg1 lon' exchange trains, eacM comprised ~of tw 8' cubic-feet (nominal)'of 10-cubic-fcAt
,y > _ ves,sels~ loaded wit c unogeneously mired (E-96 and LINDil-A zeolite exchange mpdia;-
./ a >< p '
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4., A monitoring and 5fsp'. nJ' system .for' control of demineralized unit (, > .r
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< + r
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?
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#yi radiation shielding for piping, valves, sampling, and monitoring ]
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/- ll e systems; %
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t 9t ;Y " Two monitoring tanks for collecting treated wated I u(,, ( ,n
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1 i f 1 1 7. :.,e - An of[f-. gas Jyttem for rtreating and filtering gasep and vent air-i j
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0398B/LC ). q+ ;, d $ 4
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TER 3527-006i
- 8. -A Liner Recombiner.and Vacuum Outgassing System (LRVOS') designed to
.j eliminate the potential of a combustible' hydrogen and oxygen l mixture existing in the SDS liners - 'j . l
- 9. Associated pipi.ng. valving, and structural _ supports required.for
-placement of system components;.
I
- 10. Auxiliary systems inc1' ding u underwater ion-exchange' column storage, a dewatering system, and analytical equipment; l'
l 11. . Vent system.to: allow for venting of stored vessels..
'l 1 The EPICOR-II system is e vnstream of the SDS process flow stream for removal of trace fission products that are not removed in the ion exchange media of the SDS.
4.3 StLb merged Demineralized System Desian Criteria 1 4.3.1 Design Basi,s Regulatory guidance followed during the design of the l l- Submerged Demineralization System was extracted from the following documents: i-o U.S. Nuclear Regulatory Guide 1.140 dated March,1978 o U.S. Nuclear Regulatory Guide 1.143 dated July,1978.
~0393B/LC l
.1
-s TER 3527.-006
~
U.S. Nuclear l Regulatory Guide 8.8, dated June,1978-o lj o .U.S. Nuclear Regulatory' Guide 8.10, dated May, 1977' ., 4..
- o- U.S. Nuclear, Regulatory Guide 1.21 Revision.1, June 1974; o 'Coce of Federal' Regulations,;10 CFR 20; Standard.for-Protection Against Radiation -
o Code of Federal Regulations; 10.CFR-50,. Licensing of. ') ' Production and Utilization Facilities. 4.3.2 Process , The design shall provide for operations and maintenance'in. l such a manner as to maintain' exposures'to plant personnel.to levels which are "as low as is reasonably achievable", in' ; accordance with Regulatory Guide 8.8. 4.3.3 Performance i j The isotopic inventory for the water to be. processed'is . Summarized in; Table 1.1. The SDS followed by the EPICOR-II' j systems is designed and operated such as to reduce the average : Isotopic specific activity of the treated waste streams. The expected performance of these systems is given in Table 3.2. , a
- 36 .
0398B/LC u r.
i TER 3527-006L .;
-{'
4.3.4- Capac i ty.-
!l 1
Flow' Rate - 5'to 30 GPM.(up to:15.GPM per train). 'The' system wil_1 have.the ability to operate. continuously. .(subject to' . periodic maintenance. shutdown). , 1 1 4.3.5' Performance and Desian Requirements
.The following system requirements have been incorporated'into the design of'the SDS.
o Leak Protection and Containment o Shielding (Beta, Gamma) o Ventilation o Functional Design and Maintainability.
- o Criticality Concerns o Decontamination - Decommissioning 4.3.6_ Piping System (piping, valves and pumps)
- 1. The mechanical and structural design criteria and fabrication of piping systems and piping components are specified in ANSI B31.1,'1977 Edition with Addendum through Winter 1978 or, ANSI b31.1,'1980 for components added after 1980, and Table 1 of Regulatory Guide'1.143, 2, Piping system design shall be based on a maximum of 150 psi at 100*F.
0398B/LC \ G _
TER 3527-006 1
- 3. Piping runs are generally designed to permit water flushing.
- 4. Instrument connections to piping systems are located to provide clearance for attachment, operation and maintenance.
4.3.7 Vessels and Tanks l 1. The mechanical and structural design criteria and fabrication of vessels and tanks will be in accordance with the requirements of the ASME Boiler and Pressure vessel Code, Section VIII, Division 1, 1977, Addendum i through Hinter 78.
- 2. The vessels shall be of two types:
I
- a. Primary ion-exchangers shall contain approximately l eight (8) cubic feet of zeolite ion exchange media for the purpose of removing cesium and strontium from the waste water. Should our processing scenario be changed it may be necessary to alter the volume of the 2eolite media. Should changes occur, the NRC will be informed.
l 0398B/LC 1
,< 1 TER 3527-006J b .' fnfluentfande" cation"sandfilteruqitsareplanned to contain: cartridge; type fil.ter: assemblies.or sand
' capable of removing particles greater!than- .
.J approximately 10 microns. SDS effluent filter.
capability has been provided to incorporate the capability to filter'out ion-exchange media' fines. from the process stream should fines carryover occur. 'f
- 3. The SDS ion-exchangers and filters shall be capable of; functioning submerged under approximately 16 feet of. water j within the s' pent fuel, pool.
~4 . The ion-exchangers shall be designed for 15 GPM-nominal.
process ~ rate, filters shall be designed for 50 GPM nominal;- volume velocity through the loaded ion-exchangers~shall'be' limited to prevent channeling or breakthrough.
- 5. Pressure loss through the ion-exchangers'should not exceed 15-psi when operating at 5 GPM with clean resins.
.l
- 6. The lon-exchangers shall be equipped with a lifting arrangement compatible with the spent fuel pool crane to permit movement of the vessels in the pool.
- 7. The 10-cubic-foot vessels will be equipped with all required-nozzles, including inlet, outlet, vent connections, and fill q and sluicing connections.
l l
'0398B/LC=
i
-j i
l.. L____- _ _ _ _ - - _ _ _ _ _ _ - _ - _ - - - - - - - - _ - _ _ _ - - - - - - - - - ---- o
'TER 3527-006!
- 8. Each ion-exchanger shall.be equipped with all Internals.
required for media" distribution, dewatering, an'd venting. .
- 9. Design Condition-
- a. The 10-cubic-foot vessels will be compatible with the-
_ piping design conditions of;150.psig at 100*F. 'Tha vessel design conditions!for. continuous operation will-be, at least, equivalent to the piping design conditions. 1
- b. The following additional design conditions have been l
imposed: o Overall Height 54~1/2 inches o Overall Diameter 24 1/2 inches-o ~ Materials Stainless Steel o Height will have negative buoyancy (loaded with ton-exchange media)
- 10. Testing The vessels shall be hydrostatically tested at 1.5 times the design pressure per ASME Section VIII.
1 i a j 0398B/LCL i
.I-
i
.TER 3527-006 !
4.3.8 Shielding Design The shielding shall be designed to reduce levels resulting from the SDS to less than imR/hr, general area. The shielding for the EPICOR-II equipment is adequate for the processing of the SDS effluent because the SDS effluent water activity will be lower than the activity level of the water for which EPICOR-II shielding was originally designed. 4.3.9 Leakage To minimize the operational impact of activity that can potentially leak from bad process connections to fuel Pool B, SDS vesself are contained in secondary containment enclosures. Pool water is continuously drawn through these enclosures and passed through separate ton exchangers (Leakage containment). This design prevents the pool water from ; eventually attaining high level concentrations of radionuclides. Monitoring of potential leakage is accomplished through the established SDS Sampling System. 4.3.10 Building and Auxiliary Service Interfaces I The SDS has been designed to meet the following building interface requirements. l 0398B/LC 1 __ _ _ _ _ _ _ _
TER 3527-006 y 1 :. All" components of. the SDS located. in the Fuel Handling l Building do not exceed the normal load' capacities of'the- . cranes'in this area. ;The' Fuel Handling Building, I auxiliary and main cranes' have capacities:'of 15 tons and - 110 tons, respectively. l
- 2. .The SDS will' operate:in the ambient conditions of-the fuel Handling' Building as supplied by the' building heating, ventilating and' air conditioning system,'and .,.
lighting system.
- 3. Auxiliary services supplied to the SDS'are from the
, Demineralized Water, Electrical' Distribution, Instrument
-Air and Service Air Systems.
- 4. During installation of the system, no equipment was permanently attached to the fuel pool liner and no penetrations were made in the fuel pool liner.
- 5. Structural support for.the system will be designed to
-take the dynamic and static loads associated with the normal operation of the system.
0398B/t.C i
- TER 3527-006.
.4.3.11. ~ Controls'and instrumentation.
1
'4.3.11 1 General System Description l '. The control and'in'strumentation. systems shall be designed to control and monitor the various normal'
. process functions throughout the system and'will
~
permit a safe, orderly. shutdown of.the' system.
- 2. The controls and instrumentation systems will enable.
the operators.to perform the. designated functions efficiently and safely.
- 3. Where portions of.the process must be operated .
remotely, sufficient instrumentation shall be included to assure safe operation'and permit analysis of a process upset or remote detection of equipment malfunction.
- 4. Control and instrumentation systems shall be-categorized as: (1)' controls and instrumentation.
systeras essential for the maintenance of. process fluid confinement, and (2) process controls instrumentation systems essential for the determination of process operating parameters. 03988/LC' _ _ - - - - _ - . - _ _ _ _ _ _ _ _ - - . - _ - - - _ - - - - - - - _ - - - - - - _ - . - . - - - - - _ -- . -- -- - L- - - -_- _-_ J
l TER 3527-006L
- 5. ' Radiation monitoring'and surveillance instrumentation' essential for. the protection of-
~
operating personnel, the publie and the environment is provided. 4.3.11 2 Performance and Design Reapirements'- l
- 1. Remote controls and' instrumentation shall h' ave provisions for remote connection'of electrical leads.
I
- 2. Alarms and/or indicators are provided for adequate.
surveillance of process operation. 3, Process-connected instrumentation shall be
' constructed of material compatible with that u' sed for the construction of the process equipment,
- 4. Electrical wiring shall-be designed in such a manner as to minimize noise and spurious signals.
- 5. Instrumentation identification and_ numbering should follow the standards and practices of the Instrument Society of America (ISA). ]
- 6. . Radiation monitor's shall be provided for the detection of gamma radiation. In-line radiation ,i monitors were installed to monitor beta radiation, however to date have not been used or maintained, !
nor are they planned to be. l 0398BILC' l l
TER 3527-006
- 7. Specific instruments shail be designited to function i in a fail-safe mode and will 31ert to a failure condition.
4.4 System Operational Concepts l ) The following is a summary operation description. This operating sequence depicts the processing scenarlo,as currently planned and could be changed based on operating experience. t l The SOS process logic as currently planned, is based on the follcuing steps: i l
- 1. Ion-exchanger units will be preloaded with'new ion exchange media prior to placement in the system. The ion exchanger units will utilize a homogeneous mixture Of zeolite reedia. !
- 2. Water will be introduced to fill and vent the ion-exchange units.
- 3. These preloaded SDS ion-exchange units will be lowered into the Unit 2 spent fuel pool and placed in the containment enclosures.
- 4. Inlet and outlat header connections will be made to the ion-exchange units.
- 5. The ion-exchange system isolation valves will be opened and treatment of the contaminated waste stream will begin at low flow rates until system integrity and acceptable out water quality are verified. i l 0398B/LC l
1 l l
TER 3527-006-6 .1 The flow rate to the ion-exchange. units will be increased on'a: gradual basis until the desired operational; flow rate is achieved.
- 7. When the first ibn-exchange' bed.becomes dspleted,'theu ' nit will be f
. flushed with processed water to ensure that radioactive waste water in tM system piping is purged prior' to disconnecting the quick disconnects'on the demineralized unit.
1
- 8. The ion-exchange unit will be decoupled remotely via the use of- ;
quick disconnects and will be stored in'the. spent fuel pool. i However, loading directly into a cask prior to shipment is possible.
- 9. After the first lon-exchange unit has been removed, the second ion exchange unit will be placed into the position of the first unit.,
and the third ion exchange unit will be moved.to the second position. A new lon-exchange unit will be installed in the-third position. In some instances fewer than three (3) lon-exchange units will be required to achieve the desired decontamination factors. In these cases, jumpers will' be installed to bypass the unused positions. l
)
l p l ! 0398B/LC l l __z_.1______ _ _j
TER 3527-006 Chapter S System Description and Arrangement 5.1 Deminera112er System I 5.1.1 Influent Water Filtration I A flow diagram of the waste water influent system is shown in Fig. 5.1 Contaminated water is pumped into the SDS from the containment surap, the RCS, the fuel transfer canal, or liquidwaste (WDL) tanks. The containment sunn will employ the presently instalted SWi-P-1 pump (jet pump). Two filters have been installed to filter out solids in the untreated contaminated water before the water is processed by the ion-exchangers. These f!1ters will he either cartridge or sand type. The cartridge filter' elements are protected by , l 3/16 inch perforated metal plate serving as a roughing screen. The prefilter has 125 micron filter cartridges to remove debris and suspended solids from the contaminated l li water. The design of the final filter is similar to the prefilter except that the filter cartridge is designed for removal of suspended solids of greater than 10 microns in size from the contaminated water, The two sand filters are loaded in layers. The first layer is 200 pounds of 0.85 mm sand and the second layer is 700 pounds of 0.45 mm sand. Borosilicate 0398B/LC l
3 J TER 3527-006 ] glass with a norms) Boron content of:22% is'added uniformlyL - through the sand to prevent potential criticality. The flow; , j capacity through each filter is 50'gpm.- Reverse' flow through filters is prevented by a check valve'in the supply line to each filter. Each filter is housed in a~ containment enclosure to enable:
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leakage' detection and confinement of poterittal leakage. Th e .-- ' filters are submerged in the spent fuel pool for shielding! ! considerations.
' Influent waste water may be sampled from a shielded sample. box-located above the water level to determine the activity.of ;
-l:
contaminated water prior to.and following filtration. ' l Inlet, outlet, and vent connections'on the filters are made j with quick disconnect valved couplings which are remotely operated from the top.of the pool. Inlet-outlet pressure l gauges are provided to monitor and control solids loading. Load limits for the filters are based on filter differential pressure, filter influent and effluent sampling, and/or the j surface dose limit for the filter vessel. A flush line is l attached to the filter inlet to provide'a source.of water for flushing the filters prior to removal. 0398B/LC j
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, , TER 3527-006 5.1. 2 . Ion Exchanger Units l
A' flow diagram of' the ion exchange' manifold and primary ion-exchange columns 1s shown in Fig. 5.2. ;This system j consists of six underwater columns (24:l/2 in. 1 54 1/2 in.),- i
~
each containing eight cubic feet of homogeneously mixed Ionf 'l Siv IE-96 and LINDE-A. zeolite n.edia'and'twoLunderwater. columns containing sand filter media. The six zeolite beds are - divided into two trains each containing three beds (A, B, C,) with.: piping and valves provided to~ operate either train individually or both trains in parallel. ! 1 The effluent fro'n the first parallel train of 'three zeolite beds flows through either of the " cation" sand filters. . Jumpers are provided to permit fewer =than four (4) vessel per. train operation. An in-line radiation monitor measures the activity level of the water exiting the cation exchanger. The i' I valve manifold for controlling the operation of the primary ion exchange columns is located above the pool, inside a shielded enclosure that contains a built-in sump to collect i l leakage that might occur. Any such leakage is routed back to the RCS manifold. A line connects to the inlet of each primary exchanger to provide water for flushing the exchangers a when they are loaded. Radionuclides loading of ion exchange vessels is determined by analyzing:the influent and effluent I from each exchanger. Process water flow is measured by instruments placed in the line to each.lon exchange train. a a 0398B/LC j 1
I TER 3527-006 When process $ng containment sump water, effluent'from the.'SDS is directed to the. EPICOR-II polishing unit, if fdesire6.. When. the'SDS is;to be utilized to process reactor. coolant,.the effluent can be valved'into .the RCS clean-up manifold then . back intolthe Reactor Coolaret System via installed tankage, < bypassing EPICOR-II. q 5.1. 3 - Leakage Detection and Process!ng Each submerged vessel is located'inside a secondary-containment box that contains spent fuel pool water. During operation'the secondGry containment lid is closed. This lid is tiotted to pertait a calculated quantity of pool' water to flow past the vessels and connectors. Pool water from the containment boxes is continuously monitored to detect leakage and is circulated by a pump through one of the two leakage. containment ton-exchangers (See Figure E.2).. Any leakage which occurs during routine connection and disconnectioft of the quick-disconnects will be captured by the containment boxes, diluted by pool water, and treated by ica-exchange i before being returned to the pool, i 5.1.4 EPICOR-II I EPICOR-II (Figure 5.3) can provide final treatment of water ) after the water 1s processed through the SDS. When processing containment sump water, the processing plan is to polish.with-EPICOR-II. When processing RCS water, EPICOR II may be used 0398B/LC. I
y _ TER 3527-006 :
- 1 a
as necessary.to_ remove Antimony 125 before being returned to' ' RCS (prior chemi. cal adjustment will be requirsd). EPICOR-II! consists of filters.' ion-exchangers'and receiver tanks. 'ine' 'l purpose lof EPICOR-II;is to femove trace fission products they 1 may be present in the water. .The EPICOR-II safety' assessment' s is provided in NUREG-0591. 5.1.5 Monitorina Tank System -l
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Effluent from the SDS ion-exchanger can. flow into one of two j monitoring tanks (Figure 5.4) or in the~ case of RCS processing, directly to one of three RCBT's. The purpose of - the monitoring tank system is to collect treated water. Eac h -- 1 monitor tank is' equipped with a sparger and. tank level-indicators that will automatically shut the inlet to the tank should a high level condition exist. Water in the monitoring l tanks can be transferred back for reprocessing by SDS or used-as flush water in the SDS, or directed to existing tankage. i 1 5.1.6 Off-Gas and Liould Separation System
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An off-gas and liquid separation system collects gaseous and I liquid wastes resulting from the operation of the Nater treatment system. The off-gas system is illustrated in Figure 5.5. Gaseous effluent lines from the ion exchange vessels, 4 sampling glove boxes and shielded valving manifolds are i connected to the off-gas system. Gaseous effluent.is passed through a mist eliminator in the off-gas separator tank before-being treated by an electric off-gas heater to reduce the . 0398B/LC ] u i
TER 3527-006 off-gas relative humidity to 70%s A roughing filter and two HEPA filters are provided for further treatment. Air is moved through the system by a centrifugal blower rated at 1000 cfm. The discharge of this blower will be monitored ard routed to the cristing fuel Handling Building HVAC fystem. Hoisture collected by the off-gas system and waste returned from the continuous radiation monitoring system is directed into a separator tank. At the top of the tank a mist eliminator separates moitture from effluer.t gas prior to the gas entering the Off-gas treatment system. The tank is located in the surge pit and is covered with a concrete and lead shield. The level in the tank will bt- indicated and controlled manually to return collected water to the RCS manifold for reprocessing. Offgassing of the RCBT's during processing of the RCS to the RCBT's is handled by established station procedures involving the Haste Gas Decay Tanks. Discharge from these tanks is filtered through HEPA filters before being released through the station vent. 5.2 Samplina and Process Radiation Monitorina System The sampling glove boxes are shielded enclosures which allow water samples to be taken for analysis of radionuclides and other contaminants. The piping entering the glove boxes contains cylinders that permit draining a predetermined amount of sample into a collection bottle. Cylinders are purged by positioning valves to permit the water to flow through them and return to a waste drain header and into the 1 i 0398B/LC
1~ ' TER 3527-006 off-gat' separator. tank. :A water line connects to the inlet'of the sample cylinders to allow the line to be. flushed after a sample has been , taken. .. 5.2.1 Samplina System l Sampling of the SDS process.to monitor performance is
~
accomplished from three shielded' sampling glov'e boxes. .One glove bo'x is for sampling the filtration system, the second.is L 'for sampling the feed and effluent for.the first zeolite'. bed if there is significant breakthrough'of the first zeolite . bed ! 3 and tne third for' sampling the effluents of the remaining
.i
-zeolites beds.
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-1 The entire sampilng secuence is performed in shielded glove.boxesIto :q 1 minimize the possibility of inadvertent leakage and spread of contamination during routine operation. 1 i !
5.2.2 Process Radiation Monitoring System l The SDS is eqcipped with a process radiation monitoring system which provicrt indication of the radioactivity concentration. l , in the process flow stream at the effluent point from each ion 1 i l exchanger vessel. The purpose of this monitoring. system is to i provide indication and alarm of radionuclides breakthrough of the jor, exchange media. 0398B/LC
TER 3527-006_.
'5.3 lon-Exchanger and Filter Vessel Transfer in the Fuel' Storage Pool Prior to system. operation, lon exchanger.and filter vessels-are placed' inside the containment boxes and connected with quick-discorSect couplings. When it is. determined that'.a. vessel 'is loaded with -
radioactive contaminants to predetermined limits as.specified in the Process Coittrol Program, the system will; be flushed with low-ectivity. processed water. This procedure flushes away.~ waterborne radioactivity,. thus minimizing.ihe potential for-loss of contaminants.into the pool water while secoupling vessels. Vessel decoupling is accomplished remotely. Vessels are transferred using the existing fuel handling
, j 1 . \
crane utilizing a yoke attached to~a,long shaft. The' purpose of this yoke-arm assembly !s to prevent inadvertent' lifting of.the,lon exchange bed or filter vessel'to a 1.eight. greater than eight feet below'the , surface of the water in the pool, This device is a safety tool"that < l will mechanically prevent lifting a loaded vessel out.of.the. water j shielding and. preclude the possibility of accidental exposure of' l operating personnel. The ion-exchange vessels are arranged to. provide series processing l through each of the beds; the influent waste water.is ticated by the bed , in position "A", then by the bed in position "B", then by the bed in 'l position "C" and finally either of the " cation" sand filters "A" or , "B". The first vessel in each train (position A) will load with l radioactive contaminants first. The loaded vessel will then.be stored. l untti transfer to a shielded cask. At no time during the operation of the system wil? a loaded vessel be taken out of the pool before it has ] 1 been placed in a shielded cask. The loaded cask will be transferred from the pool with the overhead crane. . 1
- .03988/LC
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y 1 TER 3527-006L l 54- !' Arrangement of the Water Treatment System in the Fuel Storace Pool i Figure.5.6 illustrates the arrangement of'the SDS in the fuelistorage' pool (viewed from above). The. filters, and zeolite' ion exchanger-- t
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- vessels, are. located underwater in. containment = enclosures in the "B"'
da spent fuel. pool. These enclosures ~and the exchangers-are supported alongonesideof.thepoolon~a.structurAls'teelrackthatisattached-to the' pool curb. The racks act as a support for the. system:and also
.provides an operating platform from which'the remote connections can be J made. .The off-gas system is mounted on the curb near the surge tank area. -0 A dewatering station is located in-the "B" SFP cask pit b'elow the water level and is used for displacing the' water from expended columns and 1
j : filters and dewatering.them prior to placement.in the cask.~ An underwater storage rack,. designed to handle 60 expended vessels'is. f located in the pool. This storage capacity' allows processing to continue without interruption due to handling operations or vessel disposal or shipping. Stored IX vessels-'will be vented via-a common header connecting to the liquid separation module.to continually vent 3
.q gas byproducts that may be generated in the' vessels during storage. l 5.5 Liner Recombiner and Vacuum Outcassing System (LRVOS) .
J t The Liner Recombiners and Vacuum Outgassing System (LRV05) is designed ; to eliminate the potential of a combustible Hydrogen and Oxygen mixture j existing in the SDS Liners. This will facilitate the ultimate shipment and burial of the SDS Liners. '
'0398B/LC-L L2__ _ _ _ _ - - _ . _ - _ _ - _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . : ._ _ _ _ _ _ _ _ _ . . - - _ _ _ _ . _ _ - _ _ _ _ _ _ _ _ -
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TER 3527-006' The LRVOS.will perform.the following operations'while maintaining.the-.
-normal operating depth of wa'ter between the operators-and.the SDS liner. '
l 1
-1. Reduce' water in.the'.SDS liner using vacuum outgas'si_ng'to ensure enhanced operation of the recombiner catalyst. i l
- 2. Allow sampling of the liner gas at atmospheric pressures, f
- 3. Provide capability to inert the SDS Liner with Argon or N2 to-approximately_10 psig prior to tool' removal. This will prevent any water intrusion during tool decoupling.
- 4. Provide a means to remotely insert' the recombiner catalyst into the SDS liner vent port. The catalyst is retained inside the__ liner by' the. internal vent port screen.
- 5. Provide sufficient recombiner catalyst to recombine the hydrogen and oxygen produced by radiolosis of the water remaining in the liner.
- 6. Provide vacuum to defueling canisters at the DS to allow canister gas sampling.
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TER 3527-006 Chapter 6 Radiation Protection 6.1 Ensurino Occupational Radiation Exposures are ALARA 6.1.1 Policy Considerations The objectives with respect to SDS operations are to ensure that operations conducted in support of the on-going demineralization program are conducted in a radiologically safe manner, and further, that operations associated with radiation exposure will be approached from the standpoint of maintaining radiation exposure to levels that are as low as reasonably achievable. I During the operational period of the system, the effective control of radiation exposure will be based on the following considerations' 1
- 1. Sound engineering design of the facilities and equipment.
- 2. The use of proper radiation protection practices, j including work task planning for the proper use of the appropriate equipment by qualified personnel.
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The SDS,was.specifica11yldestened to maintain ekposureito; 7 W' operating (pPJ,onnel.rcto v.tlow as; reasonably achtevable. i *l;: ,a .. l- . To -
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y, . implement ~ th.'s corfdp't'the components carrying highiledeli [. 3 , '
.activitywaterwillbepw;videdwithadditionalshieldingJor r
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are submerg? pin. the spentjfuel pool. Shielding bassbeen a av ~
,- designsd to 11mitLwhole bod) body exposure ratesiin operating i+ -
9 areas ?# approximately 11mR/hr. ..In' addition, components ; o., , carryinjhigh}evelprocessfluidshave'beenedesignedfor-f~ r : > [~. exhat.ij/ to the SDS off-gas system. 9This method of off-gas g treatmentwthminimizethepotentialifor'airbornereleasesin r , the work areas;, ' Thespecificdesignfeaturesutilizedinmeetinh.th'is
- requirement are discussed in detail'in-Section._6.2.1.
s 6.1.3 Operationalfpnsiderations a The-system design reflects the following operational ALARA jconsiderationu
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- 1. Exposure of personnel. servicing a specific component'on
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the SDS will be redt!ced by providing: shielding between the individua1' components that constitute substantial I radiation sources =to the. receptor. -j
- 2. The exposure of-personnel who operate valves on the SDS.
wil1~ be reduced through the use.of reach rods :through' ,
~
lead and' steel shield boxes. l
- 3. Controls for the SDS will be located in. low. radiation l zone s '.
l 4. Airborne radioactive material concentrations.will be' minimized by routing the off-gas effluent from the SDS,to the TMI ventilation system for.further treatment.
- 5. The sampling stations for the feedstream and filters that L contain high levels of radioactive materials will be exhausted through the SDS ventilation system.
- 6. All sampling is performed in shielded glove boxes to minimize the possibility of inadvertent leakage and spread of contamination during' routine operation.
~03988/LC
I TER-3527-006 6.2 Radiation Protection Design Features j 6.2.1 Facility Design Features The system is designed to take maximum advantage of station features already in place and Operational in terms of protection of the public. In addition, design features- j provided by the system are intended for the reduction of l ! releases of radioactive material to the environment. The following features provide for protection of individuals from ! l radiological hazards during normal operations from external exposure and unanticipated operational occurrences, such as spills.
- 1. The SDS primary demineralization units are housed under approximately 16 feet of shielding water in the TMI-2 spent fuel pool.
1
- 2. The entire process and all equipment is housed in the i Auxiliary and Fuel Handling Buildings which are Seismic Category I structures with air handling and ventilation .
systems designed to mitigate the consequences of radiological accidents. k
- 3. The system is designed in such a manner as to allow zero discharge of 11guld effluents. The effluent processed water will be stored on the TMI site until final disposition has been determined.
0398B/LC 1 i d
TER 3527-006.
- 4. The off-gas system effluent will.be filtered and'
. monitored before input to existing ventilation exhaust-systems.
- 5. Filters, primary lon-exchange. beds, " cation" sand filters, and their. associated couplings'are operated.in containment devices'. Teach containment' devic'e is connected to a pump manifold and.a continuous flow of.
approximately.10 GPM is maintained through each-containment. The combined flow from the. containment-enclosures is then processed through a-separate ion exchange column and then discharged back to the spent fuel pool. U
- 6. Loaded vessels will be placed in a shielded cask underwater.
- 7. To the extent possible all-welded stainless steel l construction le specified to minimize the potential for leakage.
- 8. Lead or equivalent shielding is'provided for pipes, valves, and vessels (except those located under water) where necessary for personnel protection.
- 9. Design of a sequenced multi-bed process - three (3) be'ds in series to preclude breakthrough and contamination.of the' outlet stream.
03988/LC u___________________ _.-______:_--_-_.-----__.
TER 3527-006 i
- 10. The entire process stream is designed with appropriate 1
pressure indicators, j
'l
- 11. Inlet, outlet andIvent connection.are made'with remote
. operated-valved q'ulck release. couplings; 6.2.2 Shielding The minimum shielding thickness required for radiological protection.has been designed to reduce levels'in occupied areas to less than l'mR/hr.~ Operating panels and-instrumentation racks are located away from potential sources of radiation or. adequate shielding is provided to meet radiological exposure design limits.
All movements of the vessels out of the fuel pool will be performed utilizing a shielded transfer cask. l 6.2.3 Ventilation The ventilation and off-gas system provided to service the SDS is designed to minimii; airborne radiological releases to the environment. Among these design features are:
- 1. Manual level controlled off-gas separator tank with mist eliminator'to receive vent connections from the ion l
exchange and filter vessels, sample glove boxes, piping manifolds, and the dewatering station. 0398B/LC
l TER 3527-006 l i
- 2. Roughing filter with differential pressure indication. i
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- 3. Two HEPA filters with differential pressure indication.
l 1 l
- 4. A centrifugal off-gas blower with flow indication.
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- 5. Sample ports for monitoring the system and DOP test ports for HEPA testing.
- 6. The effluent of the SDS off-gas system is routed to the l existing TMI-2 ventilation system Exhaust, which is filtered again through the Fuel Handling Building exhaust HEPA filters prior to discharge from the plant.
1 6.2.4 Area Radiation Monitoring Instrumentation j General area radiation monitors have been provided which will be utilized to alert personnel of increasing radiation levels during normal operations or maintenance activities. ' { j J 6.3 Dose Assessment i 6.3.1 On-site Occupational Exposures i 0398B/LC
, TER 3527-006-t Normal Operation' During the operation ofJ the' Submerged. Demineralization' System, :
. there are operations'that involve! occupational exposures, but precautions have been taken'in the design stage to' minimize personnel exposures. Major' operational lactivitiesLinvolving g
- I such exposures are'as follows: 'l
. A. Sampling operations- 1 l B. System start-up valveLalignment C. Spent-vessel-changeoutl
'0 D. Cask; removal,~ decontamination'and' survey operations E. System maintenance. J
.]
F. Vessel dewatering
-j Decommissioning 1
1 The SDS detailed decommissioning plan is being developed in conjunction with the operating procedures for the: system. 1 However, the modular design of the system is conductive to 1 y disassembly while minimizing exposure to personnel. l l l l 0398B/LC l l l ou_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - . _ . _ .
- TER 3527-006'
'6.3.2 Off-site Radioloalcal-Exposures:
S_o,urce Terms for Licuid Effluents
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Liquid effluent from the system will be returned to station. tankage.for further disposition, therefore,-~no liquid source term is. required for;this report. '
- l. Radiological source terms for. potentia 1 environmental releases are dependent on:the processing schedule. proposed for SDS ,
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l and/or EPICOR-2. Up to this timetEPICOR-2 has not been used l for RCS processing,;but recent. elevations in'the Sb-125 concentration in the RCS may necessitate the use of EPICOR-2' l to remove this contaminant. The assumption made here for j potential source term generation purposes is that both SDS and
'EPICOR-2 will be dedicated to processing RCS. Miscellaneous small batches of' liquid waste may be processed by EPICOR-2,-
but would be infrequent since liners dedicated for RCS more 1 than likely could not be used for other waste' streams. I i 1 \ ! Experience with previous operations within the RCS show that ' l ininor disturbances within the reactor vessel'.give rise to .! increased concentrations-of a select number of isotopes which become candidates for potential releases to systems involved i L i 0398B/LC-
TER 3527-006 in RCS decontamination and therefore, potentially to the- 1 I environment. A history of concentrations of the major # radiologically significant isotopes with time is shown in figure 6-1. Not reflected in this figure are the increases in l Ce-144 and alpha concentrations that accompany disturbances within the RCS. Sample analysis results, tabulated below, I show typical concentrations resulting from RCS disturbances I i l I I I i i 0398B/LC 1 I l I f
. i ilR 3527-006 i
Radiochemistry Analysis Results for RCS Sample of 4/9/84 1
)
(Sample #84-04966) I l l Concentration Isotope (uCi/ml) Uncertainty Ag-110m- <l.5E-2 Ce-144 1.1E+0 4.0E-2 i Co-60 1 7E-1 1.0E-2 Cs-134 2.3E-1 1.0E-2 Cs-137 4.9E+0- 4.2E-2 Ru-106 3.2E-1 5.8E-2 {
)
Sb-125 5.5E-1 3.1E-2 ! gross a 1.2E-3 6.1E-4 gross 6 1.9E+1 2.6E-1 H-3 3.5E-2 2.2% Sr-90 9.9E+0 35% l 1 The increased concentration of Ce-144 and associated alpha ! l l i activity is expected for RCS disturbances and is due to a l l colloidal suspension of finely divided fuel fines resulting from the accident. Concentration el2vations of alpha bearing activity, and Ce-144, are projected to be much more significant than reflected in the table above. Short term concentration spikes may increase a factor of 103 or more depending on operations in the R.V., However, for purpose of 1 l 0398B/LC
LTERL3527.-006 '1 potential source term generation, these time averaged
~
concentrations'are assumed to be as tabulated above except.for-tritium which remains fair 3y stable at 0.04 pC1/m1, neglecting radioactive decay. H l Source Terms for Gaseous Effluenti When the'SDS Technical Evaluation Report was originally written a methodology was conceived for the definition of 3 gaseous effluent source terms resulting'from SDS/EPICOR-2 processes. This methodology used defhndable, but highly - 1 l conservative assumptions for defining. gaseous effluent source ' terms. Since the beginning of SDS operation in August 1981, a significant amount of operating experience has yielded effluent data that allows more reasonable gaseous effluent source terms. The effluent data applicable to the EPICOR-2 l i and-SDS operations is reviewed in the following section for purposes of. arriving at gaseous source terms appropriate to i the proposed. future operations of these two systems. -] , 1 l l )
. i A review of the 6/83 version of the SDS TER shows that, l
according to Table 6.2, the following quantities of the applicable isotopes would have been released to the ; l environment over the previous 27.5 months of SDS operation i l through the off-gas system had the release values been correct. l-0398B/LC ; 1 i
< ' TER'2527-006' l 1
.fsotope Quantity'-(uCi) I l
i H-3 5.20 x.108 Ci' j Sr-90^ 11.5pC1
.j I-1291 '4,125 pC1' ;
Cs-134 31.6 pCi ] Cs-137 280!pC1 ! 1 i
, , l Review of.these values agains't airborne effluent release 'l
-reports, shows the projected: releases from the SDS'off-gas j i
system'to be highly conservative, Because the. data applicable' l to the SDS Off-Gas system has been' reduced so that the amount attributable to this system can be separated--from other sources, the following sour'ces attributable to the future 1 SDS/Epicor-2 operations are based on previous operations of-
] 1 these systems. Processed water concentrations, the ultimate
~
'i l
source of airborne. effluent concentrations, for previous l operations will differ from water concentrations to be processed in the future. This' initial water concertrationi difference has been factored into the projected release' values considered for this evaluation. 1 1 l J I I l 4 0398B/LC-j l 1
-1 i TER 3527-006 SDS Off-Gas System Releases for' the Period 09/IS/81' to 12/31/83 ~
SDS Off-Get Particulate.& Tritium Releases Particulate and tritium data as measured by tho.0ff-Gas' PING-1A &'H-2 bubblers was assembled for.the period 9/14/81 to 12/P 183. The total amount of Tritium l ! released through the.off-gas system for this period wf.s 7.iBE-1 Curies. l The total particulate attributed.to sampling through the PING-1A at the
'l off-gas. system was 3.15E-7 curies of Cs-137 and 2.52E-8 curies of Cs-134. j Cs-134 appeared > LLD on one instance between 12-14-81 and.12-21-81.
l The SDS Off-gas system feeds to the exhaust. ventilation of the Fuel Handling Building at 1000 cfm. The point of insertion into the Fuel Handling Building exhaust is before the HEPA filters, therefore, no increase in particulate is j seen at the station vent. In addition, the Fuel Handling Building exhaust is diluted by a factor of 3 by the time it' reaches the station vent. Table 6.1 lists the dates of positive particulate samples identified as Cs-137. l As a condition to startup of SDS, a tritium sampler in'the off-gas system was-required. A sampling unit which consists of two Fisker-Milligan bubblers in series was installed downstream of the pump of.the PING-1A'in the SDS off-gas j system. The total cumulative curies released through the off-gas system was L integrated for the time period 09/14/81 to 12/31/83 and is 7.18E-1 curies of tritium, Table 6.2 lists the H-3 ciJries by month and compares amounts released ; from the station vent, the SDS amount as a fraction of the Station Vent 1 ) Release and the curies of H-3 released through FPICOR-2. 1 70 - 03980/LC
m , , TER 3527-006' T. Table'6.3showsenvihonmental'releasecalculationsfortheproposedRCS processing through SOS and EPICOR-2. The values of' column 3 of the table'6.3 are' about a factor of 100 lower than.would have been estimated by the method-of the original SER but are considered to st11'1 be conservative. ~The values
~
in column 3 are the assumed values for.the release ~ rate to the environment. The values in column 4 are the concentrations at a downwind distance of 0.5 miles froin the station vent', assuming atmosphere dispersion is calculated by: the most restrictive data published in~NUREG-0683,-(Table H-3). The highest value of X/0 from this' table.in 3.996 E-6:sec/M3, - Using this factor and the'- dose conversion factor for tritium from Reg. Guide'1.109, an. inhalation' dose was calculated for the most restrictive recipient, an adolesc9nt. This' dose-was calculated to be 1.5'x 10-5 mrem /yr. As shown by the value of summation of the Cr/MPCx'at the bottom of column 6, the' total maximum yearly average concentration for.all'the isotopes is 16.5' million times more restrictive than allowable under the guidelines of 10 Cf3 20 using the more restrictive of the " soluble"/" insoluble" form of each isotope. 0398B/LC
l l TER 3527-006 Figure 6-1 ras nc
- ru <
o o m A e . o m A e a o l 3TI A m I I I I I I A i 1. I- l 6 S, s ji li 8 o. s rs d==~ . l i Y :e ' Dg I l3li e - -- e s y i*.p s , eg 3e 5,i: _ C3.e eu e m 3 d:= - - m-r g t - 1ll
, I,
. i.
kgI $O il! = m a I _ _~. (
.e.
.g 2* * ' I
( i, ga ill) <- s
- <w = : ,r g:il An !! " !
h'i,8!! 5W 8d, w ill.ii I"C" g a _ ( % IaX m - s - m I 3 N
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- 12. =
TER 3527-006 Table 6.1 Positive Particulate Samples Identified as Cs-137 Dates Curies of Cs-137 Curies of Cs-134 9-28-81 to 10-5-81 3.17E-9 - 12-7-81 to 12-14-8i 1,64E-8 - ! 12-14-81 to 12-21-81 3.49E-7 2.52E-8 12-21-81 to 12-28-81 2.88E 9 - 1-18-82 to 1-25-82 4.53E-9 - 6-14-82 to 6-27-82 4.46E-9 - 9-20-82 to 9-27-82 6.16E-9 - 9-25-83 to 10-2-83 1.73E-8 - 11-20-83 to 11-27-83 1.09E-8 - Total 3.15E-7 Curies of Cs-137; 2.52E-8 Curies of Cs-134 0398B/LC i i l l l I
TER 3527-006
' Table 6.2 Station Tritium Release Values i
SDS Ping 1A SDS Ping 1A Station-Vent H-3, fraction EPICOR-II Dates H-3. Ci H-3. Ci of Station Vent H-3. C1 9-14-81 to 9-30-81 2.99E-2 5.24E-1 0.0367 2.91E-1 Oct. 81 5.71E-2 3.25E0 0.0176 1.03E-2 Nov. 81 1.17E-1 1.30EI 0.0090 1.20E-2 Dec. 81 6.64E-2 1.14E0 0.0582 3.10E-2 . Jan. 82 5.70E-2 5.77E0 0.0099 3.06E-2 Feb. 82 Har. 82 2.12E-2 3.54E-2 1.68E-1 3.97El 0.1262-0.0009 5.77E-3 7.71E-1
)l Apr. 82 2.72E-2 1.80E0 0.0151 2.30E-3 '
May 82 1.02E-2 6.31E0 0.0016 1.26E-3. Jun. 82 9.80E-3 3.06E0 0.0032 6.39E-3 ; Jul. 82 8.50E-3 1.42E0 0.0060 6 '. 58 E- 3 1 Aug. 82 2.17E-2 1.40E1 0.0016 1.11E-2 ] Sep. 82 8.80E-3 1.48E1 0.0006 1.30E-2 Oct. 82 1.38E-2 1.17El 0.0012- 1.33E-1 Nov. 82 2.84E-2 1.88E0 0.0151 6.50E-2 i Dec. 82 2.05E-2 1.02E1 0.0020 2.02E-2 l l
] !
Jan. 83 1.44E-2 3.83E0 0.0038 3.00E-2 Feb. 83 1.08E-2 8.04E0 0.0013 1.01E-2 Mar. 83 1.05E-2 3.58E0 0.0029 6.20E-3 i Apr. 83 3.00E-2 3.03E0 0.0099 1.02E-3 May 83 7.80E-3 1.61E0 0.0048 3.71E-3 ' Jun. 83 2.13E-2 1.33E1 0.0016 4.82E-3 Jul. 83 9.50E-3 2.13E0 0.0045 3.56E-3 l Aug. 83 7.00E-3 3.15E0 0.0022 1.04E-2 i l Sep. 83 1.33E-3 2.60E0 0.0005 9.10E-3 i Oct. 83 2.34E-2 2.15E0 0.0109 4.24E-3 l Nov. 83 3.48E-2 2.41E0 0.0144 < LLD l l Dec. 83 1.38E-2 2.83E0 0.0049 < LLO I 1 Total 7.175E-1 177.4 .---- 1.44 l Ci/ month 2.61E-2 6.45 .---- 5.22E-2 1 I I i l l 0398B/LC 1
---- - - - z
TER 3527-006 i Environmental Release Calculations for the Propo' sed RCS Processing Through SDS and EPICOR-2 The amount of RCS to be processed over a years time is projected to be 1.3 x 106 gallons. Concentrations of the various radionuclides in this volume are assumed to be as tabulated below. Table 6.3 RCS Processing Release Parameters Conc. Conc. at'O.5 units 10 CFR 20 Cx Isotope (pCi/ml) Ci/sec. (C1/m5) Table II Col. 1 RPCx i Ag-110m <1.5E-2 <4.7E-18 <1.9E-23 3E-10 (6.3E-14 Ce-144 1.1E+0 3.4E-16 1.4E-21 2E-10 7.0E-12 Co-60 1.7E-1 5.3E-17 2.?E-22 3E-10 7.0E-13 Cs-134 2.3E-1 7.0E-17 2.8E-22 4E-10 7.0E-13 l Cs-137 4.9E+0 1.5E-15 6.0E-21 SE-10 1.2E-11 ! Ru-106 3.2E-l 9.9E-17 3.9E-22 2E-12 2.0E-12 Sb-125 5.0E-1 1 6E-16 6.4E-22 9E-10 7.1E-13 Sr-90 9.9E+0 3.2E-15 1.3E-20 3E-11 4.3E-10 H-3 3.5E-2 2.9E-9 1.2E-14 2E-7 6.0E-8 U-235* 3.8E-7 1.2E-22 4.8E-28 4E-12 1.2E-16 i U-238* 2.4E-6 7.4E-22 3.0E.27 3E-12 1.0E-15 Pu-238' 4.7E-7 1.5E-22 6.0E-28 7E-14 8.6E-15 l Pu-239* 8.4E-4 2.6E 19 1.0E-24 6E-14 1.7E-11 1 Pu-240* 2.1E-4 6.5E-20 2.6E-25 6E-14 4.3E-12 Pu-241* 1.4E-2 4.3E-18 1.7E-23 3E-12 5.7E-12 ! Am-241* 1.4E-4 4.3E-20 1.7E-25 2E-13 8.5E-13 Np-237* 1.1E-7 3.4E-23 1.4E-28 1E-13 1.4E-15 Np-339* 1.7E-8 5.3E-24 2.lE-29 2E-8 1.1E-21 f (Gross a) (1.2E-3) [3.7E-19) (1.5E-24) (2E-14) .----- Cx - 6.05E-8 TOTAL MPCx Values calculated according to the Ce-144/ fuel ratio value is calculated by the ORIGEN Computer code as programmed for the THI-2 Operational history and a decay time of 5.5 years. i
)
-l
, i l 0398B/LC l i
.__.____.____.____________m
TER3527-Od'6 Chapter 7 Accident' Analysis
,Because of.the' inherent safety features of the Submerged Demineralized System and maximum utilization of existing. site' facilities, potential accidents which involve the release of radionuclides to the environment are minimized.
Hypothetical accidents during system operations are-proposed and evaluated in l 'the following assessment. The following accident analysis lhas been performed 1 based on the assumption that zeolite beds are radiologically loaded;to 60,000 C1. Should higher radiological loadings be determined to be appropriate, the accident analysis will be reassessed using-the higher radiological. loadings. l 7.1 Inadvertent pumping of RCS water into the spent fuel pool. { l l Assumptions: 4 The effluent line from the final filter develops a leak and is not ; detected immediately. Contaminated water is released into the pool at a rate of 15 gpm for a period of 15 minutes, (225 gallons ~~or.~15 curies). l l It is assumed that the total activity is made up of 0.2Ci of Cs-134 and 4.2 Ci of Cs-137, 0.94 Ci of Ce-144, 8.4 Ci of Sr-90, and 0.5 Ci'of Sb-125 (based upon the measured concentrations as reported in Chapter 6). Analysis of the accident also assumes uniform mixing in 233,000 gallons of-pool water and results in pool water contamination 0398B/LC l l l
, , s 1
'TER.3527-0064 .,
~ ~
levels'ofLO.017'pC1/mloftotilactivity'orof0.0075.pCi/ml'of
. gamma emitters. This value'is only about 37."of the ialue. calculated for-the same accident' assuming RB-" sump" water was inadvertently pumped;into the fuel pool water.
Occupational Exposure Effects: The' dose rate is calculated to'an individual on the' walkway at.a point. i three feet above the surface of the water using the ISOSHLD-II. computer. -l code. 'The depth of-water in the pool is 38 feet. The calculated' maximum exposure rate at three feet above the surface is 4.2 mR/hr' . After such an accidental leak the pooliwould contain.-1 millicurie of. .{ alpha activity. Such a leak would require that more-stringent contamination' control procedures would ha@ to-be-installed to prevent' > alpha activity from leaving the pool. Cleanup of the pool would' require passing the water through 2 specially prepared'4x4 liners; one similar to the SDS liners and one similar to the EPICOR. Off-site Effects: l
.I A review of previous SOS operation shows that this accident does not l-release measurable activity to the environment.
l l ( No significant increases in the site boundary direct gamma exposure c l 1evel is expected as a result of this hypothetical accident due to the-i spent fuel pool configuration and inherent shielding properties of the pool side walls and the distance to the site boundary. 03988/LC
n- y , d b TER 3527-006!, ,
Conclusions:
s This hypothetical accident is evaluated'under-conservative assumptions. 1
,y Although the analysis of this-hypothetical accident provides results I
that indicate radiation field'of 4.2 mR/hr.at a level three feet above' j the pool surface, area radiation monitor alarms wo'uld indicate-its- i presence. Personnel would'be evacuated to ensure that occupational exposures are limited. ~l a
.l Off-site radiological consequences potentially resulting1from this ]
I hypothetical accident are insignificant l q
-1
]
7.2 Pipe rupture'on filter inlet line (above water leve1F 1 Assumptions: A pipe rupture occurs in the inlet line to the filters above water level at the southeast corner of the pool. -The. leak proceeds for fifteen minutes before the pump is stopped. Contaminated water sprays from around the lead brick shielding. A total of 28 gallons'of water is spread onto a surface area of 100 ft.2 and'340 gallons of-contaminated- ! water are drained into the pool., It is further assumed that'the_ contaminated water contains 0.065 Ci/ gallon.of activity in the same ! concentration ratios that were assumed for the previous hypothetical accident. 03988/LC _ _ _ = _ _ - _ . _ - - _ _ _ _ - ____ -
- n v
TER 3527-006- l
^
Occupational Exoosure Effect's: > a g l
-As'a. result of this. hypothetical ~ accident, five significant effects are: I postulated:
l
- 1. The maximum gamma: exposure rate at'.the surface ofsthe contaminated :)
l u
. floor. area is calculated to.be 100~ mrem /hr.
- 2. The maximum beta exposure rate at a' point three feet above:the
]
surface of the contaminated floor area is estimated to.be I 560 mrad /hr. 1
- 3. The exposure rate from the surface of the. contaminated spent. fuel pool waters, at a point three feet above.the. surface, would be
. J approximately 6,3 mrem /hr gamma, and ~32' mrad /hr beta.
1
- 4. The pool water would contain about-1.5 millicuries of alpha-activity, and- '
-l l l 5. the floor surface would be contaminated with about 0.2 millicuries I 1 l l of alpha activity. !
u l l I Offsite Effec'ts: ;
-To calculate off-site concentrations it is' conservatively. assumed-.that l
- l. 0.1% of the activity sprayed from the pipe becomes airborne within:the l l
l Fuel Handling Building. This airborne activity is evacuated from the l 03988/LC i __a_-_:-____---_
TER 35i7-006 Fuel Handling Building by the FHB H&V system which is filtered through HEPA filters before the airborne effluent reacncs the environment. The offsite concentration is maximized by assuming'the activity is evacuated from the FHB in a 15 minute time period and, consequently, the hypothetical release to the environment occurs over a 15 minute period. Release parameters for this accident are as tabulated below. Credit has been taken for only 1 of the 2 HEPA filter banks of the FHB exhaust filter system. l
Conclusions:
Analysis of this hypothetical accident, show that even unoer the conservative assumptions of the accident, the effluent concentrations, for a period of 15 minutes, are calculated to reach a level such that j the summation of the individual. C1/MPCg values is 79% of the allowable. Credit for the neglected HEPA filter'and a less conservative l l X/0 would reduce this fraction to an even lower value. , i i i f 03988/LC l l
- TER 3527-006-Release Parameters-for a RCS Pipe. Spray-Leak Accident EA Concentration (C1/K3 )
Release rate -Station Vent- (at 610m.with' Isotope to FHB (ci/s) Release Rate-(ci/s) X/0-1'.3x10-3 s/M3 )* Cx/MPCx ,
'I Ag-110m- <2.4E-8 <2.4E-11 <3.1E-14~ (1.0E-4 .
Ce-144 1.8E-6 1.8E-9 2.3E-12' .1.2E-2 ]q Co-60 2.7E-7 2.7E-10 3.5E-13 1.2E-3 L Cs-134 3.7E-7 3.7E-10 4.8E-13 1.2E-3
.Cs-137 7.8E-6 7.8E-9 .1.0E-11 2.0E-2 Ru-106 5.1E-7 5.1E-10 6'. 6 E 3.3E-3 i Sb-125 8.0E-7 8.0E-10 1.0E 1.1E-3 l Sr-90 1.6E-5 1.~ 6 E-8 2.1E-11. 7 ~. 0E 1 H-3 . 5.6E-8 5.6E-11 7.3E-11 ~3.7E 4:
U-235 6.1E-13 6.1E-16 7.9E-19 - 2.0E-7 i- 'U-238 -3.8E-12 3.8E-15 :4.9E 1.6E 4 Pu-238 7.5E-13 7.5E-16 '9.8E-19 1.4E-5 Pu-239 1.3E-9 1.3E-12 1.7E-15' 2.8E-2i Pu-240 3.4E 3.4E-13 -4.4E-16 ^7.3E-3 , 1 Pu-241- 2.2E-8 2.2E-11 2;9E-1A- :9.7E-3< -1 Am-241 2.2E-10 2.2E-13 -2.9E-16 1.5E-3' NP-237 1.8E-13 1.8E-16 :2.3E 2;3E NP-239 2.7E-14 2.7E-17 3.5E 1.8E-12
'C1 - 0.786. -!
TOTAL ETCj l 3
- The X/Q value chosen for.this analysis _(l.3x10-3 gfg ) was used because ofL the short dura.lon of the release.. This precluded the use of the annual average-X/Q. .
4 As shown at the bottom of column 5, the summation of the 'Cx is only 797. of the MPC x l specified 1.0 for this scenario. J i 0398B/LC
)
W ' . I 3,
'R'. t , < TER 3527-006.
Even though high surface contamination' levels exist at'the floor area. and the spent filel pool waters.are contaminated such t' hat-the' total' body-could be exposed to relatively high~ radiation leve)s, area radiation. monitors would indicate the presence cf high.radiatibn. Personnel would) h be evacuated from the area to ensure that occupational' exposures.are i limited. I 7.3 Inadvertent liftino of prefilter above pool' surface I
-l Assumptions: .I
'j It is assumed that due to a failure in the crane, control; system, the j 1'
over head crane moves toward the loading bay after pulling one expended filter to the maximum height of eight feet below the pool surface. As the crane moves toward the bay, the handling tool hits the end of the i pool and the filter is dragged from the water exposing operating i l personnel. Analysis of the accident is performed by using a point source. I approximation and calculating the dose rate at a distance of 15 feet from the filter. The calculated dose rate is 21 Rem /hr and is based on i an assumed filter loading of 1000 curies. l 1 Occupational Exposure Effects: As the filter assembly nears the surface of the spent fuel pool water-l area, radiation monitor alarms will be sounded announcing the presence ! of high radiation fields. Personnel would be evacuated from the area to ensure that occupational exposures are limited. 0398B/LC
1
.TER 3527-006~ j 1
Off-site Effetts:: Airborne contamination as a- result of th)s hypothetical accident would : not occur since-the particulate activity.is fixed on the filter elements which are contained within the filter. housing'. The increase in'the radiation: level at the site boundary would not be L significant due'to the shielding characteristics'of the fuel building - ( .. walls and~the distance to the site boundary. l.
Conclusions:
l The public health and safety is not compromised as a consequence of this I ! hypothetical accident. 7.4 Inadvertent lif ting of zeolite ion exchanger above r3ool~ surface-Assumptions: It is assumed that due to multiple failures, a zeolite vessel is lifted . I from the pool resulting in the exposure of plant operating personnel. 1 Analysis of the accident.is performed by modeling the zeolite ion' q exchanger bed in cylindrical geometry and. calculating the dose rate at a . distance of 20 feet from the sutface cf the zeolite ion exchanger. The i calculated dose rate is approximately 340 Rem /hr based.on an estimated i zeolite ion exchange bed loading of approximately 2730 Curies of , Cesium-134 and approximately 51.900 Curies of Cesium 137.- ; 0398B/LC-
e- -,, TER 3527-006 y .
. Occupational Exposure Effects:
1 As the' zeolite vessel nears the surface'of-the. spent'.fuei pool water, area radiation monitorialarms will. automatically sound; announcing the presence of highi radiation fields. Personnel would be evacuated from the area- to reduce' occupational: doses. Airborne contamination wouldinot . occur since the' activity is' fixed on the'zeolites. e Offsite Effects: Air'orne-contamination b as a result sf this-hypothetig l' accident'would z Y not occur since the activity is contained on the.zeolites-which are contained in the. ion exchanger vessel.7 Th e increase'in the: radiation. level at the site boundary would not be.significant due to the' shielding provided by the fuel Handling Building walls and the distance-to thel site boundary.
Conclusions:
The public health and safety is.not endangered as a-result of-this. hypothetical accident. Occupational exposures are minimized by evacuation of the area. 0398B/LC o
~
i
'TER 3527-006-7.5 Inadvertent Drop of SDS Shippino Cask ,
+
I l Assumptions:
, 1 l
It is assumed that:due to a failure in SDS shipping cask handlingt l .
- , -equipment an-SDS cask containing a zeolite ton exchange'r is dropped from l
- the Fuel Handling Building-(FHB) crane to the floor at EL 305'.. The SDS shipping cask is assumed to drop'from the maximum cratie lift height' .
Upon. Impact with the floor at EL 305', the SDS shipping cask is' assumed b to experience rupture as well as' rupture of'the zeolite vessel, thus exposing the dewatered zeolite resins to the FHB atmosphere. The radiation source is approximately 2730 Curier of Cs-134 and j approximately 51,900 Curies of Cs-137 on the. zeolite ion exchange-l l media. The contribution from other~ isotopes on the. zeolite media and residual containment building sump water (Table 1.1) in the ion exchange media is negligible; it is' assumed that a factor of 10~4 of the isotopes are instantaneously released to the FHB atmosphere. This o assumption is conservative because the' isotopes are absorbed onto the zeolite media. The fuel Handling. Building HEPA filte'rs are assumed to have an efficiency of 99%. Occupational Effects: Assuming that the SDS shipping cask ruptures ~~ completely exposing the-zeolite ton exchanger containing the activity mentioned abov'e, the calculated dose rate is approximately 340 Rem /hr at a distance of 20 feet. Upon the rupture of the cask, radiation monitors will sound 039CB/LC - ___ _ - _ _ = _ _ _ _ _ _ _ _ _ _ _ _ .
m-- - x g p w
-~ ~
J > g hE
, gf* 3-f[x sf TER35s7-006
;MD
- / ,
L
, Jannouncing the 'phsencef of high radiation lf telds. -- PersonnelEwould tie. .
- , .,,. yT .
j l y. 'evacuatec'lfrom:the area;to redude' radiation exposures. Airbornca 1 \
. t . > . . . .
.\
o , contamination will not' occur if.the zeoliterion exchange vessels remains- j l T ~ l N i L intacti- WIN. the assumption that'. the> vessels rupture andiradiciactive
- g. -.
31 material'bec5mes" airborne,theairborneactivitylwillbe'reducedto ' ' 4 L.
, acceptable levels'by the fuel Handling Building HVAC System prior to-
- l. ?
- l. atmospheric. release.
] ,
Operational Effectr.: ! j ;
- 1. . Impact on.systeins, structures and congonents has been considered - q r ,.
q j which could possibly' result in adversely affecting t:ig zability.to - Y ' operate'these Reactor Plants safely transfer load or. unload fuct 4 , ufely, .or maintain these Plants in a safe cold shutdown condition. J w e
- 2. - Analysis has been conducted which demonstrates-that a postulateo.
SDS Cask drop along'the proposed travel path would not adversely p affect either TMI1 Unit 1 or Unit 2. It r Off-Site Effects: s The increase in radiation level at the site boundary would not be n significant due to the shielding provided by the FHB walls and the distance.to the site boundary, if the SDS cask ruptures exposing the l zeolite ion exchanger. With the issumption that radioactive material l
, escapes, the whole body dose due'to the released activity at the site m i s
boundary will be less than 1 mrem for both beta'and gamma radiation. a 1 h I
- 86'- 0398B/LC.
^
q h i 4Jh
TER 3527-006
Conclusions:
The public health and safety are not compromised as a consequence of this hypothetical accident. 1 l i 0398B/LC i i
, TER 3527-006 ,
References Campbell, 0.0 . E.D. Collins, L.J. King; J..B,'Knauer,?" Evaluation ~of the Submerged Demineralized System (505) Flowsheet for Decontamination of; High-Activity-Level Water at the Three Mile Island Unit 2 Nuclear Power Station," ORNL/TM-744:, July 1980. Clark, W.E., "The Use'of Ion-Exchange to Treat Radioactive Liquids in: Light-Water-Reactor Nuclear Reactor Power Plants," NUREG/CR-0143, ORNL/NUREG/TM-204 (August 1978). Ga, J. H., E. W. Murbach, and'A. K.'H1111ams, 1979,." Experience and Plans for Effluent Control at LHR Fuel Reprocessing Plants", in Proc. Conf. on Contro111na Airborne Effluents from Fuel Cycle Plants,'AICHE Topical Meeting. Lin, K. H., "Use of Ion-Exchange for the' Treatment of Liquids in' Nuclear Power Plants," ORNL-4792 (December 1973). Lin, K. H., " Performance of Ion-Exchange Systems in Nuclear Power Plants," AICHE Symposium, Ser. H (152), pp 224-35 (1975). Willingham, H. E., 1972, "The Vitro Engineering ISOSHLD User's Manual," VITRO-R-153. U.S.DepartmentofHealth, Education,andWelfare,1970,RadiolobicalHealth' ; Handbook, U.S. Government Printing Office, Washington, D.C. l 4
- 88 '
0398B/tC l
~
TER 3527-006L ;y Chapter-8 .j T
]
- )
. Conduct-of Operations. , 3 j l
l o ,
'l 1The SDS program for operations 3 1s divided into a phased approach. These <
- phases'are:
8.1 System Development
-]
l 1 System development activities have been. performed to assure that' ) components are developed.specifically.to meet the-conditions imposed at.- TMI and perform in the intended manner. The ion-exchange process is a well un'derstood process. .Even though ion-exchange media have been in use for approximately 50 years or more,. ! a development program was conducted at the Oak Ridge National' Laboratory, the-results of which are: documented in ORNL TM-7448, to .j ensure that the media selected for use at TMI provided optimized performance characteristics of various media-using samples'of the waters-to be processed at'TMI. In.some cases, SDS' effluent'.will be polished by i EPICOR-II.- Additional development. effort has been expended to verify that media loading and dewatering can be accomplished in the intended manner and. that the remote tools, necessary for the coupling and de-coupling of the l vessels, operates in the intended manner. l 0398B/LC
TER 3527-006 8.2 System Preoperational Testing Prior to use in the SDS each vessel will be hydrostatically tested in conformance with the requirements of applicable portions of the ASME Boller and Pressure Vessel Code. Upon completion of construction, the entire system will be pneumatically tested to. assure leak-free operations. The system will be tested to an internal pressure of no less than 1.5 times the design pressure. Individual component operability will be assured during the i preoperational testing. Motor / pump rotation and, control schemes will be verified. The leakage collection sub-system, as well as the gas collection sub-system, will be tested to verify operability. filters for the treatment of the collected gaseous waste will be tested prior to initial operation. System preoperational testing will be accomplished in accordance with approved procedures. 505 system testing will be-approved by the GPUN Start-up and Test Manager. ! 8.3 System Operations System operations will be conducted in accordance with written and approved procedures. These procedures will be applicable to normal system operations, emergency situations, and required maintenance i evolutions. l 03988/LC
, , TER.3527-006c Prior?to.SDS' operation,'. formal ^ classroom instruction will be provided to.
~
, systems operations personnel to ensure that adequate knowledge is. gained, to enable safe and efficient. operation. 'During system operations
-on-going operator evaluati_ons will-be: conducted.to ensure continuing-safe and efficient system operation.
l 1. 8.4 System Decommissioning l L The decommissioning plan for SDS is being developed. An outline of the l planned approach to decommissioning is shown below. The basis for the decommissioning plan is that the Submerged
' Demineralization System is a temporary system; its~ installation and removal will cause no permanent plant changes.
- 1) Equipment and interconnecting piping will be decontaminated': the-levels to which decontamination is accomplished will depend on'the intended disposition of individual items, 1.e., disposal or reuse.
l
- 2) The system will be disassembled, component by component.
l
- 3) Major system components can be stored for later use or disposed of l at a licensed burial facility. l
.i
- 4) Small components, such as valves, piping, instruments, etc. can be l
disposed of as radioactive waste. ; 1 'l
.0398B/LC
TER 3527-006 Appendix No. I to Submerged Demineralized System Technical Evaluation Report REACTOR COOLANT PROCESSING PLAN WITH THE REACTOR COOLANT SYSTEM IN A PARTIALLY DRAINED CONDITION I i l I i f
TER 3527-006 CONTENTS Chapter 1 Summary of Treatment Plan 1.1 Project Scope i 1.2 Current RCS Radionuclido Inventory and Chemistry 1.3 RCS Processing Description Chapter 2 RCS Processing Plan Design Criteria 2.1 Introduction 1 l 2.2 Design Basis ! 2.2.1 Submerged Demineralized System 2.2.2 Interfacing Systems 2.3 RCS Process Plan Goal Chapter 3 System Description and Operations 1 3.1 Introduction 3.1.1 Submerged Demineralized System 3.1.2 Interfacing Systems 3.2 RCS Water Processing Preparation 3.2.1 RCS Preparation 3.2.2 SPC Operation 3.2.3 Reactor Coolant Liquid Haste Chain 3.3 RCS Water Letdown and Injection 3.4 RCS Processing by SDS ! 3.4.1 RCS Hater Filtration 3.4.2 RCS Water Demineralization 3.4.3 Leakage Detection and Processing 1
)
l i I
)
t
- TER 3527-006:
CONTENTS (continued) 4
, . Chapter 3' System Description'and Operations .(continued):
3.4 RCS Processing by SDS (continued). 3.4.4. Off. Gas and_ Liquid. Separation System
'3.4.5' . Sampling and Process Radiation Monitoring System 3.4.5.1 . Sampling System 3.4.5.2 Process Radiation Monitoring System 3.4.5.3 Transuranic Element'Honitoring-
'3,4.6 Ion-Exchanger:and Filter. Vessel Transfer 3.5 Zeolite' Mixtures 3.6 Waste Produced l .
l 4.1 RCS Processing Safety Assessment 1 l 1 1 1 j I
-4 l
-. J i
i
'TERi3527-006 Chapter 1- ']j i
SUMMARY
OF' TREATMENT PLAN , 4 1.1 Project Scope-
.The decontamination of the TMI-2 Reactor Coolant System (RCS) requires the processing of the radioactive contaminated water to _
reduce the activity-'therein. The-present activity level of_this water is given.in Table 1.1. To date, in excess of 1,000,000 gallor.s of water have been processed from the RCS. The feed and bleed operation via the Submerged Demineralized System (SDS) has reduced the radionuclides concentration of the RCS water;: . l specifically the Cs-137 concentration has been reduced from 14.0-pCi/cc to the present value of approximately 0.016 pCi/ct. l l
)
This report describes the processing of the RCS by the SDS while j i maintaining the RCS in the partially' drained, open condition. The j 1 design features of this processing method will utilizei l l
- 1. proven processing capabilities of the SDS, and
- 2. Existing plant systems in support of the SDS.
1 0400B/LC
, _.TER 3527-006 1.2' Current RCS Radionuclides Inventory and Chemittry Watersampleshavebeen'takencontinuouslyffromtheRCS-to. identify.
specific radionuclides.and concentrations, and plant chemistry. Typical'results;are listed in Table.1.1. This~ data is based on
' actual samples taken. RCS activity is decreasing.due to radioactive' decay and . leakage:from the.RCS which is being.made~up
- by injection.of clean water into the RCS,'and due to batches'which have been removed for SDS' processing. Figure-121.shows-how activity for the major;nuclides has. decreased with'.. respect to time. Currently Sb-125 concentrations:have risen to. radiologically-significant levels due to changing RCS chemistry parameters'. -
The: Sb-125.will be removed by batching water from SOS through.EPICOR using organic resins. i 0400B/LC
1 l Figure 1:2.- TER 3927-006
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TER 3527-006-1.3~ RCS Processina Description 1 On a batch basis, radioactive RCS water is letdown ~to a-Reactor- . Coolant Bleed Tank.(RCBT) while clean water is injected into the RCS from another RCBT. -RCS. water is'then pumped from the' receiving 1 .. RCBT through the prefilter and-final filter. 'RCS water then goes-through the RCS manifold and the SDS ion exchangers. The. effluent from the ion exchangers is routed through'the cation ~ sand filter to another RCBT'for chemical sadjustment,,1f necessary, and injection back into the RCS as makeup. .The above process'is repeated until the RCS water is decontaminated. EPICOR II may be used for : q processing selected batches of RCS water unless needed for chloride 1 I control. .j H l The processing of the RCS will use the existing filter and ion exchangers of the SDS. Existing sampling connections.will.be used-on the influent and effluent of the filters and ion.exchangers to determine radionuclides and chemical composition of the'RCS before and after processing. As described in the SDS TER, the prefilters, final. filters, and cation sand filters are for the removal of particulate matter. The j prefilter and final filter are followed by a series'of ion exchange l vessels containing about 8 cubic feet of zeolite ion exchange media. Location, operation, and handl1ng of these vessels remains-unchanged from the mode of operation used for processing of the Reactor Building sump water and the RCS. water as described in the SDS TER. 0400B/LC-
.j TER 3527-006 TABLE 1.1 RCS RADIONUCLIDES'AND CHEMISTRY DATA (May 1986)
ISOTOPE RADIONUCLIDES CONCENTRATION pCi/cc H-3 0.085 Sr-90 2.4 .i Cs-134 0.01 Cs-137 0.35 t pH 7.58 j Boron 5460 ppm Ha 1500 ppm- l l 4 1 i
.l 1 d
l l i l l 1 0400B/LC
l
-TER 3527-006! 1 Chapter =25 ,
4; -RCS PROCESSING PLAN'bESIGN CRITERIA'
.:j 2.1' ' Introduction --
^
)
1 This.RCSProcessingPlan.isdesignedtousethsSuomerged Demineralized.'. System (SDS) and portions.of existing plant. liquid'
-radwaste disposal systems'to' decontaminate the:RCS water..-This- )
1 will reduce plant personnel and off site radiation exposures. LThe' j < design objectives of.this processing plan are.tol utilize: ]
- 1. A system that.is as. independent'as possible from existing
. radioactive waste systems at THI-2. The SDS portion'of..this "
plan is a temporary system for the recovery of TMI-2. Only
.small sections of existing THI-2 plant. systems willibe used. i l
- 2. A system that has proven performance'in processing radioactive.
waste. The SDS portion of this. processing plan has
~)
successfully decontaminated the . Reactor Building sump arid the i l RCS water. l 1 1 2.2 Desion Basis 1 l 2.2.1 Submerged e Demineralized System l W The. Submerged Demineralized System was designed in , accordance with the following regulatory documents: o J0400B/LC' .
TER 3S27-006 1.. Code of Federal Regulations,10CFR20 Standard for Protection ag& inst Radiation.
- 2. Code of Federal' Regulations,.10CFR50, Licensing of Production and Ut!11zation Facilities.
- 3. U.S. Regulatory Guide 1.21, dated June 1974. I
- 4. U.S. Regulatory Guide 1.140, dated March 1978. ll 1
- 5. U.S. Regulatory Guide 1.143, dated July 1978. .
- 6. U.S. Regulatory Guide 8.8, dated June 1978. '
- 7. U.S. Regulatory Guide 8.10, dated May 1977.
i The design basis for the SDS is presented in greater detail in Chapter 4 of this TER. l h Interfacing Systems l ' 2 . '2 . 2 The interfacing systetas with the SDS in the'RCS 1 l Processing system are: l
- 1. Radwaste Disposal (Reactor Coolant Liquid) System
- 2. Reactor Coolant Makeup and Purification System
- 3. Auxiliary and Fuel Handling Buildings Heating Ventilation and Air Conditioning Systems
- 4. Nitrogen Supply System l
1 S. Decay Heat Removal System i 6 Haste Gas System
- 7. Standby Pressure Control System
- 8. Spent Fuel Coo' ling System
- 9. Instrument Air System 0400B/LC
-TER.'3527-006 :
oThedesigncriteriafor,these;systemsTiexcept'SPC)are-I presented-in Chapter;3 of the THI-2'FSAR. Conformance to these criteria-is presented'in the Wespective sectior5
~
for-these systems in'the.TMI-2 FSAR. Standby Pressure Contr'ol' System data may.be found in.the THI Recovery System Descriptions and TER'.s. 1 2.3 RCS Processino Plan Goal L l .. The goal of the RCS Processing Plan is to reduce the; total
^
l radionucitde concentration of Cs in'the RCS to"less than 1.- , i pCi/cc. The'RCS Chemistry will be maintained'as follows as'a ;j
-1 minimum:
i i Chlorides < 5 ppm pH > 7.5 but < 8.4 Bore,n > 4950 ppm ,
.k i
1 The processing of water through'the SDS'is not expected to.have any l undesirable effect on the chemical characteristics of the RCS-l ! i water. Maintaining proper chemistry of the makeup water will l ensure that there will be no adverse effects on the RCS with 1 respect to corrosion. The boron concentration of the makeup will l also ensure that sufficient boron is present to maintain'the core in a non-critical safe condition. Sampling of the RCS water will be. continued in accordance with approved operating procedure. l l l l 0400B/LC- j i
TER 3527-006 H Chapter 3 l SYSTEM DESCRIPTION AND OPERATIONS j 3.1 Introduction l d i This RCS Processing Plan is designed specifically for the j controlled decontamination of the radioactive water in the RCS and the treatment of the radioactive gases and solid radioactive waste l which are produced. This plan will use the SOE as the means of j decontamination of the RCS with support from other existing plant systems, t i 1 3.1.1 Submerged Demineralized System f The SDS consists of a liquid waste processing system, an off gas i system, a monitoring and sampling system, and solid waste handling system. The liquid waste processing system decontaminated the RCS water by a process of filtration and demineralization. The off gas system collects, filters, and adsorbs radioactive gases produced i during processing, sampling, dewataringc and spent SDS liner- l l venting. The sampling system provfdes measurement of process performance. The solid waste handling system is provided for moving, dewatering, storage, and loading of filters and demineraliter vessels into the shipping cask. The SDS will be unchanged from that described in the SDS TER. l 9- 0400B/LC
TER 3527-006 : L3.1.2LInterfacina Systems'
~
' Interfacing with the SDS are existing plant' systems, as given in Section 2.2. The Reactor Coolant Liquid. Waste Chain:provides-a staging location for the SOS.for-collecting'dnd , injection of RCS water from and to the-RCS. The Fuel. Handling Building and:.
Aux!11ary Building HVAC systems provide tempered ventilating air; and controlled air movement-to~ prevent spread of. airborne'.. contamination with the plant and to:the outside environment. Thea
~
Nitrogen Supply system provides N 2
,for blanketing:the Reactor.
Coolant Bleed Tanks. -The Makeup and. Purification and Spent-Fuel
~ Cooling Systems provide piping for the transfer.of. the waste-water. The Waste Gas System processes the gases from the vents from the RCBT's. The Instrument Air System provides air pressure for air-operated valves in the. Interfacing Systems. The Standby.
Pressure. Control System, installed as.a. temporary THI-2: recovery system, will be u'ed as a backup system to ensure a source of. I additional makeup to the RCS. 3.2 RCS Water Processing Preparation 3.2.1 RCS Preparation i The RCS will be maintained in a partially drained condition vented to atmosphere. Its water level may vary from Elevation 347' to 323'6" depending on the needs for access to the reactor vessel. l 0400B/LC 1
i TER 3527-006
~
The minimum water level-is expected to;be 323'6" (l' abov'efthe.
. reactor vessel flange). -
At this level and at all levels above this, the Waste Transfer j
. pumps will be used'to inject RCS. makeup water into the RCS for the {
RCS cleanup process. The maximum discharge. pressure of these pumps + :
.is 74 psig at a flow rate of'40.gpm. Flow to the,RCS'will be. 1 controlled by valve HDL-V-36A or'36B depending on which waste- ,
1 transfer pump is used for feed and if necessary..MU-V-9. 'MU-V10' I will also be open to permit makeup flow.to;the RCS; 'The: flow rate to the RCS will be maintained at less than'5.gpm to match the letdown flow rate. Minor adjustments in flow rate will be made to-maintain the'RCS water level within the.' limits required. i l The decay heat analysis as reported in Appendix B THI-2 Decay Heat Removal Analysis, April 1982, submitted as a part.of-the Safety Evaluar.lon for Insertion of a Camera into the Reactor Vessel Through a Leadscrew Opening Rev. 2 July 1982, 1s applicable.for the RCS processing described herein. The' average incore coolant temperature will be liniited to'less than 170'F. This criterion was adopted as a conservative value for the recovery' program to maintain a positive margin to bolling. 1 i i 1 0400B/LC'
-J
'TER-3527-006' 3.2.2 SPC Operation i The~. Standby Pressure Control: System. (SPC) will serve as' a' backup-system to ensure that the RCS level is maintained during RCS.
processing. 3.2.3 Reactor Coolant Liauid Waste Chain:, i Prior to starting-_RCS water processing, an.RCBi will.be filled with more than 50,000 gallons of borated, suitable, processed water-. The radionuclides and chemistry' data for this water will.be similar to that'used for RCS makeup during the previous RCS' processing period. Chemicals will be added to this water if. required to - ensure that this water complies'with the plant chemistry specified . y in Section 2.3.
-3.3 RCS Water Letdown and Injection RCS letdown will be performed by a bleed and feed process of-simultaneously removing the radioactive RCS water.and injecting borated processed water at the same flow rate to maintain RCS water volume constant. The bleed and feed process will be' controlled from the Control Room in coordination with the Radwaste Control i l
Panel. The RCS water is letdown through the normal letdown line on the loop cold leg before Reactor Coolant Pump RC-P-IA. The letdown- ) rate is S gallons per minute if_ the waste transfer pumps art used l a l 04008/LC. ) 4 u
l' TER 3527-006 or 10 op:n if 'a nealy installed sandpiper pump (fig. 3.4), thich is normally disconnected, is used. The RCS water is letdown through the letdown coolers to a RCBT. The plugged block orifice and isolated Makeup Cemineralizers and Filters are bypassed. As the RCS water is letdown, simultaneously the borated processed water ; located in another RCBT is' injected into the RCS. After the RCBT l hcs been filled to more than 50,000 gallons, the letdown and injection of water from and to the RCS will be secured. The RCBT will be recirculated prior to processing. After recirculating, decontamination of the RCS radioactive water by the SDS will j i commence. 3.4 RCS Processing By SDS 3.4.1 RCS Hater Filtration j l Two filters have been installed to filter out solids in the j l untreated contaminated water before the water is processed by the q w lon exchangers. Both filters are sand type. The two sand filters l are loaded in layers. The first layer is 0.85 mm sand and the I second layer is 0.45 mm sand. Mixed uniformly with the sand is approximately 6 pound'; borosilicate glass which is at least 22 weight percent boron. The loading of these filters may be changed if applicable. The purpose of the borosilicate is to prevent the possibility of criticality should any fuel fines be transported in the let down. The flow capacity through each filter is 50 gpm. Reverse flow through filters is prevented by a check valve in the supply line to each filter. 0400B/LC
TER 3527-006 Each filter is housed in a containment enclosure to enable leakage detection and confinement of potential leakage. The filters are submerged in the spent fuel for shielding considerations. Contaminated water can be pumped through the filters and the RCS manifold to the ion exchangers. Influent waste water may be sampled from a shielded sample box located above the water level to determine the activity of contaminated water prior to and following filtration. l Inlet, outlet, and vent connections on the filters are made with quick disconnect valved couplings which are remotely operated from the top of the pool. Inlet / outlet pressure gauges are provided to monitor and control solids loading. Load limits for the filters i are based on filter differential pressure, filter influent and effluent sampling, and/or the surface dose limit for the filter vessel. A flush line is attached to the filter inlet to provide a source cf water for flushing the filters prior to removal. I 3.4.2 RCS Water Demineralization This system consists of eight underwater columns (24 1/2" x 54 1/2"), each capable of containing eight cubic feet inorganic zeolite sorbent. Homogeneously mixed Ion Siv IE-96 and LINDE-A 0400B/LC
1 TER 3527-006- -j
. )
zeolite are the medias of choice to efficiently imn:obilize the-Cesium and Strontium in the RCS. Six zeolite beds are divided into , i two trains each containing two or three beds (A, B, C) with piping j 4 and valves provided to operate either train individually or both ) trains in parallel. 1 The effluent from the zeolite trains flows through the remaining
" cation" sand vessel. Jumpers are provided to permit 2, 3, or 4 vessels per train operation. An in-line radiation monitor measures j the activity level of the water exiting the last ion exchanger vessel. The valve manifold for controlling the operation of the primary ion exchange columns is located above the pool, inside a shielded enclosure that contains a built-in sump to collect leakage that might occur. Any such leakage is routed to the off gas bottoms separator tank and pump. A line connects to the inlet of 1
l each ion exchanger to provide water for flushing the ion exchangers . I 9 when they are loaded. Radionuclides loading of ion exchange vessels ! is determined by analyzing the influent and effluent from each exchanger. Process water flow is measured by instruments placed in the line to each lon-exchange train. The effluent from the " cation" sand I vessel is routed back tc a RCBT, as shown in Figure 3.3. The l remaining SDS equipment and EPICOR II are not used for RCS water I processing. 0400B/LC
, ~ ;
t; . 1
.TER 3527i006
' Periodic; sampling'of;the.procesststream tillioccurfduring:the j 1
processing.of a batch of water. At the completio'niof processing.a
~
batch, the_ contents of'the receiv'ing.RCBT{ dill'be sampled to : i determine acceptability-for injections;ofothis. water into'the RCS. If the'watertis within specification,_ittis-injected into the'RCS. < j y
.l The types of samples to.be taken;at RCBT afterlietdown and. prior to. i reinjection are shown in-Table 3.l'. j i
3.4.3 Leakage Detection and-Processing l l Each submerged vessel is. located inside a : secondary containment box 'l that contains spent fuel pool water. During oper'ation'the secondary containment lid is closed.- This lid is slotted to permit a calculated quantity of pool water to flow past the vessels and l connectors. Pool water from the containment boxes'is con'inuously: t ;; monitored to detect leakage and is circulated by a pump through onei 1 I of the two leakage containment ion exchangers. Any leakage which ! occurs during routine connection and disconnection of the quick-disconnects will be captured by.the containment boxes,
' diluted by pool water, and treated by lon exchange.before being. j returned to the pool.
l 1 0400B/LC 1 _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ ____________..__.______i'___________________._______._.____.____ _ _ _ _ _ _ . _ _ . . - ._ . _ _
+ {TER3527-006j 3.4.4 Off Gas and Liauld Separation System'i An off gas;andL11guid separation. system coll'ects gaseous and liquid E
wastes resulting from the. operation of the water-treatment ~ system. 3.4.5 Sampling and Process Radiation Mcnitoring System ' l ! The sampling glove boxes are shielded enclosures wh'ich allow water-l samples to be taken for analysis of radionuclides and other .j
.i contaminants. .The piping entering the. glove' boxes permits the -{
withdrawal of'a volume: limited amount of sample into a collectiorr 1 bottle. Cylinders are purged by positioning valves to: permit the water to flow through them and return to a waste' drain header and j
.into the off gas separator. tank. A water line connects to the I sample line to allow the line to be flushed after a. sample has been. j taken, ,
.1 The entire sampling sequence is performed in shielded glove' boxes 'j to minimize the possibility of inadvertent leakage and spread of-contamination during routine operation.
1 1 3.4.5.1 Sampling System Sampling of the SOS process to monitor performance is-accomplished from three shielded sampling glove boxes. One glove box is for sampling the filtration system, the second is l 0400B/LC. q q i
. y
, 3 n. ; ,
-TER-3527-006' 'i
,f ;
. for sampling' the feed and effluent for the.first zeolite bed. .
and th'e. third from'samplingLthe effluents of the' remaining 1 zeolite beds and the " cation'! sand filter. ; 1
'l i
3.4.5.2 ' Process Radiation Monitorino System ; i l The SDS is' equipped with a' process. radiation monitoring system '! which provides indication of the radioactivity concentration ;) 1 in the process flow stream at the effluent point from'the last
' ion exchanger vessel. The purpose of this monitoring system is to provide indication and alarm of radionuclides breakthrough.
1 3.4.5.3 Transuranic Element Monitoring a I Filter and process' train samples are being analyzed for. j l 1 isotopes of Uranium and Plutonium. -)
. J 3.4.6 Ion Exchanger and Filter Vessel Transfer in the Fuel Storace Pool
']
Prior to system operation,. ion exchanger and filter vessels are- 3 placed inside the containment boxes and connected with . l quick-disconnect couplings. When it is determined that a vessel is I loaded with radioactive contaminants to predetermined limits as specified in the Process Control Program,'the system will.be u
- IS - 0400B/LC j l
a l
t TER 3527-006 flushed with low activity' processed water. 'This procedure flushes away waterborne radioactivity' y , thus minimizing the potential.for loss of contaminants into the pool water while decoupling' vessels'.- Vessel decoupling is accomplished remotely. . Vessels.are-transferred using-the-existing fuel handling crane utilizing a yoke attached to a'long shaft. The purpose of this yoke-arm. assembly.is . '
.to prevent: inadvertent lifting of the' ion exchange bed or filter vessel to a. height greater than eight feet below the surface of the water in the. pool. Thisdehiceisasafetytoolthatwill mechanically prevent lifting'a loaded vessel out of the water shielding and preclude the possibility of accidental exposure.of operating personnel.
The ion exchange vessels are Erranged to provide series processing through each of the beds; the influent waste water is treated by. the bed in position "A", then by the bed in position "B", then by. the bed in position "C", and finally by the bed in the " cation" sand filter "A" or "B" position. 3.5 Zeolite Mixtures The SDS ton exchangers will contain a uniform mixture of IONSIV-96 l and LINDE-A ion exchanger media. These two zeolites were' selected for their proven capabilities while processing Reactor Building Sump water to remove radionuclides. IONSIV-96 primarily removes 0400B/LC' l
.~
y , o i- .
< ~; ,'g-;..
, . . . . 'S
'TER 3527 006.
k . . f". . .
;jg "the isotopes.of-Cesium'andtLINDE-A' removes the isotopes.of Strontium.. .
.j 3 y ,
TheratiocfI)oadingthetwotypesiof'ionexchan'germedia'willbe 0 , . . .4 , . _ .. r# determined by experimental 1' data'to~ determine the optimum. loading.
- g. ,
.) ,
e;W 1, s ,
,g ,; l-Periodic samp1tntf of the . process stream will'Ne used to verify the o performance of the ion exchange media. If.necessary revisions'will' o 'y .
- be-made to the' loading ratio $ conditions warrant to achieve the' 2
. f ... . L proper decontamination factort. Verification of the performance ofr the iofe exchange' media will'.be~made in'accordance with.the Process
,,g
' ' ? - Control Plan.
3.6- -Haste Produced-Based on operating experience processing the Reactor' Building sump
" water, the useful life of a zeolite resin bed is in excess of' 100,000 gallons of waste water processed. At this point the DF of the zeolite bed for Strontium goes to 1..
l r,
'4 f h
0400B/LC
,j., ' .
YI '" .
. , .L
TER 3577-006 Table 3.1 RCBT WATER SAMPLII;G RCBT LETDOWN SAMPLE RCBT INJECTION SAMPLE Gamma Scan Gamma Scan Gross Beta - Gamma Gross Beta - Gamma Sr-90 Sr-90 pH pH at 77'F Conductivity Conductivity Boron Boron Na Na C1 C1 , Sulfates Sulfates H-3 H-3 0xygen Fluorides I 0400B/LC i
m astn=ees g . t%;_I ?1, , t yj;@ i j e anpg-( , l u
,3 C' i #a i l'
l g _ g%;
@ GL. @ .@ @ D @ @ -c g -- -
)
-', , p g- l 4,3 D.
r 1 r."q :=~ _;' . N, h.! Il : ' ni'b; l : fk iWI ' 3i @l l tj l DL-g'. d. v' i I![ -
,4 g eg g-f g. .
- 3 [4gW!,Qj$)lf ? !$U if- l ij ipI/ "N
~ ~ '
l ,-} 4 5d;se ti
'lNd!i? _ [ id f l ni$
'IS ,b h h k b
< n.hksh o .r w ~w, w .w ,
IL hhh$hffl. rl . h' es i= @gWhhk,gA na, u sk +V')Q*Jj)Aj . c* v'l hgVki
, K .'. J % d QQL b S' hhLL -
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TER 3527-006 f4.1, .RCS Processino Safety Assessment: y Processing of'the RCS while in a partially diained conditi,on does' not present a' unique safety concerr. The actual processing ~of ( Reactor Coolant is' adequately.' addressed:in the SDS, Technical Evaluation Report and the. maintenance of the Reactor Coolant System in a' partially drained condition is adequately addressed in the' Quick took Safety Evaluation. The only evolution not previously addressed is the simultaneous feed and bleed of the Reactor Coolant System'in a partially drained configuration; .During this evolution, RCS water level will be monitored and maintained.by operating procedures.- Such procedures will maintain the water q 1evel to within-'six (6) inches of the predetermined level set-point. At the present RCS level, to permit-incore inspections, this level is 210" 1 6". This level'is the same as that
. established for the Quick Look program and will be monitored i.n a similar fashion. Thus this evolution will not increase the probability of occurrence or consequences of an accident previously evaluated or create the possibility of a'different type accident, nor will the margin of safety as defined in the basis for any i
Technical Specification be reduced. l i l l 0400B/LC
TER 3527-006 i k Appendix No. 2 to Submerged Demineralized System Technical Evaluation Report Title Internals Indering Fixture Processing System June 1983 (Deleted) l I l 1 l
' TER 3527-006 j l I i i Appendix No. 3 to
- Submerged Demineralized. System Technical Evaluation Report TITLE FUEL TRANSFER CANAL' DRAINING SYSTEM (FCC)
JUNE 1985 i 1 i l l I I 1 1 I
)
TER 3527-006 1 1 CONTENTS Chapter 1 Summary of Treatment Plan ) 1.1 Project Scope- . j 1.2 Current Fuel Transfer Canal Activity and Chemistry { 1 1.3 FCC Processing Description _ Chapter 2 FCC Processing Plan Design Criteria 4 1 2.1 Introduction -{ 2.2 Design Basis 2.2.1 SDS 2.2.2 Interfacing Systems 2.3 FCC Processing Plan Goal Chapter 3 System Description and Operations 3.1 Introduction 3.1.1 SDS 3.1.2 Interfacing Systems 3.2 FCC Transfer Operations 3.3 FCC Instrumentation 3.4 FCC Processing by SDS 3.4.1 FCC Water Filtration 3.4.2 FCC Water Demineralization 3.4.3 Leakage Detection 1 i 4
TER'3527-006' , 5 CONTENTS (continued)- Chapter 3 System Description and Operations (continued) . 3.4 FCC Processing by:SDS (continued) 3.4.4 .Off Gas and Liquid Separation System'
- 3.'4. 5 - Sampling and Process Radiation Monitoring' System
,3.4.6 Ion-Exchanger and Filter Vessel Transfer in the Fuel Storage Pool
' 3. 5' ZeoliteHixtures 1 i
Chapter 4 Radiation Protection , ) i 4.1 . Ensuring.0 occupational Radiation Exposures are'ALARA
.4.1.1 Overall Policy. '
4.1.2 SDS Design and Operation. 4.1.3? Existing Plant Considerations, 4.2 Dose' Assessment 4.2.1 On-Site Assessment
-4.2.2 Off-Site Radiological Exposures.
Chapter 5 Conduct of Operations ; 5.1 System Performance 5.2 System Testing 5.3 System Operations Chapter 6 Additional Accident Scenarios 6.1 Possible Accident Scenarios 6.2 Design Features to Mitigate Effects of Casual _ty Events a l
i TER 3527-006 1 i Chapter 1 i
SUMMARY
OF TREATMENT PLAN j 1.1 Project Scope The capability to maintain water clarity and radionuclides concentrations in the Fuel Transfer Canal (Deep End) duringdefuelingoperationsmustbel l available. The design features of this processing method are: I i
- 1. Use of the proven processing capabilities of the SDS.
- 2. Use of existing plant systems in support of SDS. i
- 3. Use of FCC-P-1 (canal drain pump).
i
- 4. Use of DWC system piping. l l
1.2 Current Fuel Transfer Canal Activity & Chemistry f 1 i Water samples are taken weekly to monitor radionuclides activity and chemical parameters of the Fuel Transfer Canal. Current results are listed in Table 1.1. Activity decreases due to decay, however activity in water may increase due to leaching from plenum or activity on canisters being transferred through the Fuel Transfer Canal. 0402B/LC
TER 3527-006 1.3 FCC-Processing Description figure 1.1 shows a block diagram of the FCC processing flow path The Fuel Transfer Canal may be processed on a continuous basis through the SDS pre & final filters, one or both trains of ion exchangers, and the cation sand filter with the effluent routed back to the FTC or the 'A' t Spent Fuel Pool. In addition the FTC may be processed through the'SDS to any of the RCBT's. The FCC processing will use the existing SDS. filters j and ion-exchangers. Existing sampling capabilities will be used to monitor the process as in past processing. Further information on the SDS system may be found in the main sections of the TER. l 9 l I 0402B/LC 4
c TER~-3527-006 Table 1.1 FTC Radionuclides and Chemistry Data-(05/14/86) Co60 3 x 10-4 pCi/ml Sr90 3.3 x 10-2 pC1/ml Rul06 < 6.1 x 10-5 pCi/ml Sb125 3.3 x 10-3 pCi/ml Cs134 3.2 x 10-4 pCi/ml Cs137 1.3 x 10-2 pC1/m1' Cel44 < 4.8 x 10-5 pC1/ml Boron 4990 ppm Turbidity 1.25 NTU I i 0402B/LC l 1 l 1 l _
TER 3527-006' ,l Il
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1 TER 3527-006 j l Chapter 2 FCC PROCESSING PLAN DESIGN CRITERIA 2.1 Introduction ( l J The FCC Processing Plan is designed to use a high capacity submersible pump (FCC-P-1), the Submerged Demineralized System, and portions of the Defueling Water Cleanup System to maintain water clarity and activity j s levels in the Fuel Transfer Canal. The design objectives are: l l I
- 1. A system to maintain FTC clarity and radioactivity lemels.
- 2. A system that is as independent as possible from existing plant systems. The only portions of this system that are not temporary recovery systems are plant services connections (water, air, electric) and if water is to be processed to a RCBT, the inlet i
header to the RCBT's (WDL System). 1 l l 3. A system that has proven performance in radioactive waste l processing. The SDS has successfully decontaminated to date almost l 4.5 million gallons of contaminated liquids. l 2.2 Design Basis 2.2.1}_DS D The design basis of the SDS is presented in Chapter 4 of the SDS TER. l 1 04028/LC 1 u __
1 o s , TER'3527-006-
~
- 2.2'.2 Interfacing Systems'~ '
s TheLinterfacing. systems with the'SDS in-the FCC Processitig;iystem" are:
- 1. Reactor Coolant Liquid Waste System
- 2. Reactor, Auxiliary and' Fuel Handling Buildings Heating,.
Ventilation'and. Air Conditioning Systems.
- 3. Haste Gas System
- 4. Plant Air,' Electric, Nitrogen and Demin. Water Systems
- 5. RB Jet Pump-
- 6. DHCS-Reactor Vessel filtration System
- 7. Fuel Transfer Canal Shallow End Drainage System.
(- l l The Design Criteria.for Systems 1-4 above are presented in the { . THI-2 FSAR. System; 5 thru 7 are covered in the SDS System Description.
]
l 2.3 FCC Processing Plan Goal The goal of the FCC Processing Plan is 1) to maintain Gross By. activity ) less than 1 x 10-4 pC1/mi to minimize the'suurce term and 2) to , maintain turbidity less than INTU to maintain underwater visibility. The i processing of this water through SDS has no effect on the chemical characteristics of the water. l l l 04028/LC
l TER 3527-005 Chapter 3 1 SYSTEM DESCRIPTION AND OPERATIONS. ; 3.1 Introduction The FCC is designed to maintain liquid levels and water clarity in the fuel Transfer Canal. 3.1.1 Submerged Demineralized System The SDS consists of a liquid waste processing system, an off gas .3 ! system, a monitoring and sampling system, and solid waste handling l j system. The liquid waste processing system decontaminated the FTC 1 water by a process of filtration and demineralization. The off gas , d system collects, filters and absorbs radioactive gases produced during processing, sampling, dewatering and spent SDS liner i l venting. The sampling system provides measurements of process performance. The solid waste handling system is provided for moving, dewatering, vacuum drying, inerrization, storage, and loading of filters and demineralized vessels into the shipping cask. l 0402B/LC
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* 'TER 3527-006.
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3.1.2. Interfacing Systems. .,
, .j Canal water is transferred from.the FTC;usingia' commercially
.available high capacity submersible pump. This pump (Canal-Drain '
~ Pump FCC-P-1) takes suction in~the 4" drain, located in the, deep end of the canal. A 1 1/2 inch 10 rubber hose with quick-disconnect:
two way shut-off type' fitting connects the discharge'.of the pump to. .j \ , i 1 the fuel transfer canal' drain .mantfold. The manifold serves 'as a tie-in' point for' 3 systems; the' Reactor - Bldg. Basement Pump system, the. fuel, transfer canal, drain; system, and.the FTC Shallow End Drainage System.- Double' isolation of the-FCC processing system from the'other two is provided by ball. valves FCC-V003 and FCC-V002 in addition.to disconnected / capped < connections located on each of the other= branches of the manifold. From the manifold,.the system uses an existing flow path' through . Reactor Building penetration R-626, Fuel Handling Building penetration 1551 to tie-in and interface with the SDS system. I Power for the pump.is supplied from distribution panel PDP-6A, - breaker #12. 1 Flow from the FTC may be manually throttled via CN-V-FL-1 or ) CN-V-FL-3 in SDS if desired. , i I l l
- 8- 04028/LC
q c
;TER 3527-006' The fuel Handling Buildingc Auxiliary BuildingiLand Reactor..
. Building HVAC systems provide' tempered ventilating. air and' controlled air movement to prevent spread offairborne contamination-
, with the plant-and'to the outside environment. {The Nitrogen Supply 1 system provides N 2
.for blanketing the Reactor Coolant Bleed 3
Tanks, should the system effluent be routed to them. The Waste Gas
. 1 System processes the~ gases from the vents from the RCBT's~.
l-1- The principal' components of the.SDS are located in Spent fuel Pool- I "B", as.shown in Appendix' No. 2 Figure 3.1. 'The' piping and' components of the systems interfacing with the SDS are' located in~ y the Fuel Handling and' Auxiliary Buildings; Tanks, pumpsc.v'alves,- 1 l piping, and instruments are-located in controlled access areas. Components and piping containing'significant radiation sources are-located in shielded cubicles, such as the Reactor' Coolant Bleed - Tanks'and the Waste Transfer pumps-WDL-P-5A and'WDL-P-5B (see. I Appendix No. 2 Figure.3.2). i
)
3.2 FCC Transfer Operations ., l 3.2.1 Normal Operations 1 The FTC will be filled with RCS grade' water. .The function.of the .)
)
FCC system is to possible a controlled means of draining or I processing this water from the canal.
~
1 I 0402B/LC. I
1 TER 3527-006 - To start the-FCC processing stens, the valves must.be aligned and the SDS must be configured per the approved operating procedure and both the connections from SHS-P-1 and DWC-P-1 must be I disconnected. The pump FCC-P-1 is started and valves are operated per the procedure to ensure effluents is routed where desired. 3.3 FCC Instrumentation Pump FCC-P-1 is controlle6 via hand indicating switch FCC-HIS-1, which is located on SDS control panel CN-PNL-1 at El. 347'-6" of the fuel handling > building. The switch starts and stops the pump and shows, via a light, that power is being delivered to the pump. The starter FCC-STR-1, for the pump is mounted adjacent to panel CN-PNL-1. Pressure gauge FCC-PI-3 is provided on the canal drain manifold to sense 1 the line pressure downstream of the manifold isolation valves. l l Fuel Transfer Canal Water Level FCC-LI-102 is provided by a bubbler (FCC-VICV-104) through proportional controller FCC-LT-102, with Local Level indication FCC-LIS-103 which also actuates high/ low level alarms (FCC-LAHL-103). 0402B/LC
?
TER 3527-006._ l 3.4 FCCProcessindbySOS' , u 3.4.1-FCC Water Filtration 'l Two filters'are' installed to filter out solids in the untreated contaminated water before the water is processed by the ion-exchanger. The filters'are-loaded in layers using'various' sand 1 sizings to' optimize. filter. performance. Mixed uniformly with the i .. sand is approximately 6 pounds of.boros111cate glass;whiIchlis at-least 22 weight percent boron,.to prevent.the remote' possibility of; ,
- criticality should any fuel fines be transported to the filters.
1 The filters'and theirl containment enclosures, sampling, etc._are unchanged from that-in previous sections'of this TER. - l i 3.4.2 FCC Water Demineralization' t This system consists of two trains of ion exchangers consisting of 2 or 3. ton exchangers each. Each ion exchanger contains eight cubic feet of ion organic zeol.ite sorbent. Piping (and valves exist allowing operation of either train individually or both in parallel. The effluents from the two trains of ion.exchangers.is routed through one of two sand filters installed in the " cation" positions. These sand filters were installed in. place of the original cartridge type post filter, and,is.used to trap zeolite-r fines and improve effluent clarity. These-ion-exchangers, their containment enclosures, sampling, etc.'are discussed in more detail in previous sections of this TER. 04028/LC
m
-TER 3527-006.
3.4.3 Leakage Detection and Processing Each submerged vessel is. located inside a secondary containment. box
.that contains spent fuel pool water. During operation the.
secondary' cont'inment a lid is closed. This111d is slotted to permit a' calculated quantity of; pool water to. flow past the vessels and-connectors. Pool water from the containment boxes;is continuously-monitore'd to detect. leakage and is circulated by a' pump >through one of the two-leakage containment,lon-exchangers. Any leakage which-occurs during. routine connection and' disconnection'of.thei quick-disconnects will'be: captured by the containment boxes, diluted.by pool water,.and' treated by ion-exchange before being-
]i returned to the pool- .
)
1 3.4.4 Off-Gas and Liquid' Separation System 1 An off-gas and liquid separation system collects gaseous and.1iquid j i wastes resulting from the operation of the water treatment system. l
,)
3.4.5 Sampling and Process Radiation Monitorina System The sampling-glove boxes are shielded enclosures which allow water samples to be taken for analysis of-radionuclides and other contaminants. The piping entering the glove boxes. permits the withdrawl of a volume limited amount of sample into a collection bottle. Cylinders are purged by positioning valves to permit the water to flow through them and return to a waste drain header and ; into the off-gas separator tank. A water line connects to the- -1 i sample line to allow the line to be flushed after a sampic has been ' taken.
.0402B/LC- ;
'l
,L TER'3527-006 The entire sampling sequence is performed in shielded glove. boxes to minimize the possibility'of inadvertent leak' age and spr'ead ofL contamination'during routine operation. .:
.. g 14.5.1.' Samplina System;
,f Sampling of the.'SDS process to monitor performance is<
accomplished from.three shielded sampling glove boxes. .Onel ) glove box is for sampling th'e filtration system,.the second is? l for. sampling the feed an'd' effluent for the first zeolite bed-and the' third from sampling ~the; effluents of the remaining. l zeolite. 3.4.5.2 Process Radiation Monitoring System The SDS is equipped with a process radiation-monitoring system i which provides indication of the radioactivity concentration. in the process flow stream at the effluent point from the last ion exchanger vessel. The purpose of-this monitoring system is to provide indication and alarm of radionuclides breakthrough. l
- 13 - 0402B/LC
TER 3527-006-3.4.6 Ion-Exchanger and Filter Vessel Transfer in the Fuel Storage Pool
' Prior to system operation, ion exchanger and fil'ter vessels'are placed inside the. containment boxes and connected with quick-
' disconnect couplings. 'When.it is determined.that a. vessel.is loadea with radioactive contaminants to predetermined. limits as'
-specifledgin the Process Control' Program, the system will be flushed'with low-activity processed water. This procedure flushes 'i away waterborne. radioactivity, thus minimizing the potential-'for l~ loss of contaminants into the pool water while decoupling vessels.
Vessel decoupling is. accomplished remotely. Vessels-are transferred using the existing fuel' handling crane utilizing a yoke
-attached to a long. shaft. The purpose of this yoke-arm assembly is-to' prevent inadvertent lifting of the. ion exchange bed or filter vessel to a height greater than eight feet below the surface of the-water in the pool. This device is a safety tool that will mechanically prevent lifting a loaded vessel out of the water shielding and preclude the possibility of accidental' exposure of operating personnel.
The lon-exchanger vessels are arranged to provide series processing through each of the beds; the influent waste water-is treated by the bed in position "A", then by the bed in position "B", goes through a bed or jumper in position "C", and finally by the sand' filter in position "D". 04028/LC M
.TER 3527-006-.
L: i '"- The SOS-ion exchangers will contain a unifors misture of IONSIV-96 and. LINDE-A ion'exch' anger media. These_two.zeolites were. selected for their. proven capabilities while processing Reactor lBulldtng. Sump water tol
. remove radionuclides. IONSIV-96'primarily removes the' isotopes of CesiumL and LINDE-A removes the-isotopes.of. Strontium.
The ratio of, loading the.two types of ion exchanger medi.a wil.l'be determined by experimental' data to determine.the.optimu'm' loading. Periodic sampling.of the' process stream'will be used 'to verify. the ; performance'.of the ion exchange media. If necessary, revisions will:be
-made to the loading ratios if conditions-warrant to achieve the proper decontamination factors.
d u l i
- 0402B/LC ]
q
.1
q r TER 3527-006'- I CHAPTER'4-: s Radiation Protection. 4.1 Ensurina' Occupational Radiation < Exposures are ALARA. i 1 4.1.1 Overall Policy' -
-The objectives with respect.to FCC processing-' operations are~to-ensure that operations conducted in support of the on-going- .
demineralization program are' conducted in a" radiologically safe! manner, and further,.that operations: associated with radiation a exposure will . be ' approached from the .. standpoint.-.of maintainirig radiation exposure to' levels that.'are as. low as reasonably achievable. I
~
During the operational period of the' system.the' effective. control of radiation exposure will be based on the following considerations:
- 1. Sound engineering design of the facilities and equipment.
- 2. The use of proper radiation protection practices, including
-{
work task planning..for the proper use of the! appropriate i equipment by qualified personnel. .) l
- 3. Strict adherence to the radiological-controls procedures as developed for THI-2. l i
i 0402B/LC
., - w. m 73 ,
TER3527-006)' h 7 L 4.1.2 SDS Desiary and-Operation? ' M J M :.
}e A The SDS design'and operational considerations arelgiven in Chapter 1
y 6 of theLSDS TER. These' design and operational considerations and J features-Dremain unchanged from.this-evaluation. , q Theradiationdoseexp6sures:toplant'personnelduringfCC' processingwillbelowerduet6.the'factthattheradionuclide' L concentrationintheFTCwaterissignificant1yilowerthanlthose l experienced during' processing of< Reactor' Building sump.. water. The :j design basis for shielding the-SDStequipment.is to' reduce' radiation. 1 levtls to less than 1: mrem /hr using the radionuclides concentration: 'I
.j , .
of 200 pCi/cc of predominately. Cesium. The radionue.lide concentration'of. Cesium-in the'FTC water is currently much.less than 0.1 pCi/cc. -
.c 4.1.3 Existino 11 ant Considerations The radiation protection features for the existing plant system- j which interface with the SDS are described in Chapter 12 of the TMD 2 FSAR. The existing radiation shielding within the Auxiliary Building for the following systems is adequate to reduce the '
radiation levels to below the design basis of 2 mrem /hr in are~as requiring access: 1
- 1. -Reactor Coolant Liquid Waste Chain '
1
- 2. Haste Gas System-
.l 04028/LC-i
.TER 352710061 i
'4.2 Ocse Assessment ~ j 4.2.1 On Site Assessment Operation of the SDS'in.the FCC processing mode is expected to require intermitted processing of the FTC as required to maintain water clarity and Gross By activity <1 x 10'4 pCi/ml as required. Based on current experience with the SDS this amount of processing is expected to result in a negligible exposure for SDS 1
operat'ing area activities. 4.2.2 Off-site Radiological Exposures Source Terms for Liouid Effluents
'l I
Liquid effluent.from the system will be returned to st'at'on i tankage for further disposition. The efore, no liquid source term is identified for this evaluation, i Source Terms for Gaseous Effluents i e The plant vent system is a potential pathway for carrying airborne ; radioactive material and release. Radionuclides in the gaseous effluent arise from entrainment during transfer of contaminated ;
'I water to various tar.kt, filters, ion exchange units, and also from
]
water sampling. For further information, see.section 6.3.2 of the ) SDS.TER. .l 1 04028/LC !
)
I
;d-TER35F?,f006:
Chapter 5: CONDUCT OF OPERATIONS
.e t 4
') .
5.1, f SystemPe4ormance-4 By procevying.thi Reactor,Bu11' ding sump water and RCS successfully u 7 assurant has heen granted that components developed specifically~to meet' the corvir'.'ons imposed E[TMI' will perform in the intended manner.1
.s t ~se f
v s'/ .f / -
'Theion-excF4[ge' process:ivawellunderstoodprocess.'ThelSDShas demonstratedthathighdecontamination-factorscan.beachievedbytheuse" of zeolite ion exchange media.
. ?>E During FCC processing, the SDS system flow rates may be higher than during all previous processing. An eight hour test was performed to assure thtt these increased flowrates will not adversely affect zeolite performa'ryc'e . Also; calculations have'been performed by ORNL-to demonstrate that system performance will not'be jeopardized.c Although.
radionuclides breakthrough may occur sooner in.the batch,;it will' progress more slowly. This breakthrough will be allowed to occur'to extend' reolite life (minimize wastes) since the effluent is routed back to the fuel Transfer Canal. Zeolite media loading and dewatering can be accomplished in the intended manner and remote tools, necessary for the coupling and decoupling of the-vessels, operate in the intended manner.
- 19'- 0402B/LC l' . i a
____.._ _ __ _ _ ___. _ - l
TER 3527-006 , i 5.2 System Testing Prior to use in the SDS each vessel will be hydrostatically tested in ' conformance with the requirements of applicable portions of the ASME ] Boiler and Pressure Vessel Code. Upon completion of construction, the entire system was pneumatically tested to assure leak-free operations. 1 The system will be retested prior to IIF processing at the. design pressure. 1 Individual component operability will be assured during the preoperational testing. Motor / pump rotation and, control schemes wi'll be verified. The leakage collection sub-system, as well'as the gas collection sub-system, will be tested to verify operability. Filters for j the treatment of the collected gaseous wsste will be tested prior to initial operation. System preoperational testing will be accomplished in accordance with approved procedures. 5.3 System Operations System operations will be conducted in accordance with written and approved procedures. These procedures will be applicable to normal system operations, emergency situations, and required maintenance evointions. 1 04028/LC 1 1
YER 3527-006 Prior to FCC operation, formal classroom instruction will be provided to systems operations personnel to ensure that adequate knowledge is gained to enable safe and efficient operation. During system operations on-going operator evaluations will be conducted to 6:nsure continuing safe- ] 1 and efficient system operation. i 1 i. i l l l i t l l l 1 0402B/LC
y
. TER- 3527-00'6 ? l j
Chapter 6 , -i
. ADDITIONAL ACCIDENT SCENARIOS 6.1 Possible Accident Scenarios
~
6.1.1 A breech of.the system pressure boundary while delivering water 9 i
.from the fuel: transfer canal could result,in additional contamination of' reactor building surfaces..
6.1.2 Introduction lof reactoE'duilding. sump water 1ntoithe' fuel transfer
~
canal would contaminate the canal and could re'sult in a potential
'l criticality problem. U,
- l. 1 6.2 Desian Features-to Mitiaate Effects of Casualty Events l,
6.2.1 A hose or pipe. break ~will result .initoss of line pressure. , 1 i Pressure and flow indication are provided at,various. locationson-the pump discharge flowpath. The piping and hoses are hydrostatically tested to 1.5 times their maximum operating i 1 pressure per ANSI B31.1. To ensure pressure boundary integrity, .] i a hoses are to be inspected prior to operation of the FCC' canal drain network.
.)
4 l
. 2 2. . - '0402B/LC' I
TER 3527-006i 6.2.2 The fuelftransfer; canal drain l system'and the:Iff processingLor fuel transfer canal shallow end. drainage system connections of;the canal drain manifold contain double'isoistion, which' includes-'a check-valve in each line. This'is to prevent reactorLbuilding sump and flush water froni being delivered into the canal. In addition, the.
~
coupling. connections'on the canal drain and fuel transfericanal-shallow end drainage' branch lines of the manifold are 1 1/2-inches' _ , and' incorporate:a.two-way shut-off feature. . All.other. manifold coupling connections, including theLreactor. building basement jet-pump system connection, are 1-inch diameter., This prevents.. connecting a 1 1/2-inch pump' discharge hose,to'.the.1-inch RB. basement jet pump system connection which does not include-a' check valve. QC is to verify that each hose'is connected to t'e h proper -r manifold branch connection prior to system turnover, e l u i i
- 23.- 04028/LC I' y )
E. . a
TER 3527-006 REFERENCES
- 1. SDS System Description Appendix 18.
- 2. THI-2 Radiochemistry Summary Sheet, Sample No. 86-07106 dated f 1
May 14, 1986. ) i
- 3. Bechtel Dwg. No. 2-M75-DHC04, Schematic Diagram: Interim fuel !
Transfer Canal Processing System. I
.. 1 I
i i
)
l i i l 1 1 I l l l 1 0402B/LC l
1 TER 3527-006 I Appendix No. 4 l to -l I Submerged Demineralized System j
- l Technical Evaluation Report 1
l TITLE { FUEL TRANSFER CANAL SHALLOW END ORAINAGE SYSTEM j JUNE 1985 j l l l l l i l l l 1 l j
TER 3S27-006 CONTENTS- l Chapter 1 Summary Plan 1.1 Project Scope 1.2 FTC (Shallow End) Activity and Chemistry i 1.3 Shallow End Drainage Description - Chapter'2 Design Criteria 2.1 Introduction 2.2 Design Basis j 2.2.1 SDS i 2.2.2 Interfacing Systems ,. 2.3 System Goal l . Chapter 3 System Description and Operations [ 3.1 Introduction 3.1.1 SDS 3.1.2 Interfacing Systems 3.2 Shallow End Drainage Operations 3.2.1 Normal Operations 3.3 Shallow End Drainage Instrumentation 3.4 Shallow End filtration by SDS i i 3.4.1 Filtration l 3.4.2 Leakage Detection and Processing i 3.4.3 Off Gas and Liquid Separation System . 3.4.4 Sampling System 3.4.S Filter Vessel Transfer in the Fuel Stcrage Pool j k 4 l l I i
'TER 3527-006' CONTENTS (continued)- ,
Ch' apter 4 Radiation' Protection 4.1. Ensuring' Occupational' Radiation Exposures are ALARA 4.1.1- 'Overall Policy; . 4.1.2 SDS Design and. Operations
.4.1.3 Existing Plant Considerations 4.2 Dose. Assessment 4.2.1. On. Site Assessment a 4.2.2 Off Site Assessment Chapter.'5 Conduct'of Operations:
5 .1 ' System Performance 5.2 System Testing
-5.3 System Operations-Chapter 6 Accident Scenarios 6.1 Casualty Events 6.2 Design Features to Mitigate Effects of Casualty Events b
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TER 3527-006 Chapter'1
SUMMARY
PLAN 1.1 -Project Scope The capability to transfer water from the shallow end of'the' Fuel Transfer Canal to the deep end or a RCBT is necessary to deal _with FTC dam leakage or overflow, orLinleakage from some other source. This report is presented as an addendum to the previously submitted SDSIERtoprovidedetailsofthetransferof:waterfromtheFTC shallow end. 1.2 FTC (shallow end)' Activity Chemistry There are a number of sources which may contribute water to the shallow end (IIF Leakage, FTC dam leakage, decon, etc.) and therefore it is impossible to. state the actual activities or 1' chemistry of the water to be transferred. However water from all of these sources has been transferred / processed through SDS in the l past, and any possible sources have'been covered in detail in previous sections of this TER. I
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I TER 3527-006 j 1.3 Shallow End Drainage Description l 1 Figure 1.1 shows a block diagram of the shallow end drainage flow a paths. The shallow end of the FTC may be transferred to the deep end of the FTC, or the RCBT's, with or without filtration through l the SDS pre- & final-filters. l \ 1 l l l l l 0403B/LC
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1 TER 3527-005 f. Figure 1.1 -l
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. TER 3527-006L
. Chapter 2 i ;
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. DESIGN' CRITERIA-
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..l 2 .' l' Introduction-lq
'The FTC Shallow End Drainage System is' designed to.use a submersible pump.
1 (OHC-P-1) previously used'as'the-IIF-Processing' Pump,'and. portions of: 1 the SDS Feed & Filtration' subsystem, the Reactor Coolant Liquid Hastes Disposal' System and the fuel Transfer Canal ~ Drain System. The_Shal. low End. Drainage System design objectives are: f 1)- capability to drain shallow end'by transfer toldeep end or existing tankage.
- 2) as independent from existing plant system as: possible.
- 3) use SDS.or portions thereof.
2.2 D_esion~ Basis. 2.2.1 SDS ' See Chapter 4 of the SDS TER. 0403B/LC-q ________._______z________z._____n-______.___;______ __ _ _ . _
TER 3527-006 2.2.2 Int'erfacing Systems The' interfacing systems with the SOS in'.the Shallow End Drainage l a System are: 'l
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l 1)- Reactor Coolant. Liquid Haste System- , i
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Reactor, Auxiliary and fuel Handling Building. Heating, j Ventilation'and Air. Conditioning System. 3). ' Waste Gas System': Plant Services-(Air, Electric, Nitrogen and Demin. Water)-
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4) S) RB Jet Pump
- 6) DWCS-Reactor' Vessel.FiltrationSystem
- 7) FCC - Fuel Transfer Canal Drainage System.
The design criteria for systems 1-4 are presented in.-the'TMI-2 FSAR. TheremainingsystemsarecoveredbyLtheSDSSystem Description. 2.3 System Goal The goal is to provide a system. capable of t' transferring water from the shallow end of the RTC back to the deepend or the_RCBT's.. l
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TER 3527-006 -i
, Chapter 3 SYSTEM DESCRIPTION AND OPERATIONS' 3.1- Introduction -
The FTC Shallow End Drainage System is' designed to allow pumping of water i from the shallow end back to the.-deepend or to.the RCBT's'for future -l processing as necessary. 3.1.1 SDS The portions of the SDS used, consist of a liquid filtering system, an off gas system and a sampling system. The liquid filtering: i H
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system if used removes solids from the transfer stream. The-off gas system collects, filters and absorbs radioactive gases produced during sampling, dewatering and vessel venting. .The sampling; system provides measurements of filtration performance. 3.1.2 Interfacing Systems .i Canal water is transferred from the shallow end uslog a l commercially available submersible pump. This pump,.DHC-P-1 ! (formerly the IIF processing pump) is installed on Elev 308'-O and takes suction in an existing 4" drain located in the New fuel Pit. A 1 1/2" 10 rubber hose with quick-disconnect two way shut-off fitting connects'the p0mp discharge to the FCC manifold. 0403B/tc t I
TER 3527-006 The'manifoldserverasatie-inpoint'for4 systems;theRBlet Pump, the Fuel-Transfer. Canal. Drain. System (FCC), the DHCS-Reactor Vessel Filtration- System (EarlylDefueling). and the FTC Shallow End Drainage System. Double isolation of allLother' systems'from the
-Shallow End-Drainage system.is provided by manifold isolation'.
valves and the disconnecting of the SHS, FCC and DHC-hoses from the. manifold. From the manifold, the discharge. hose is either routed to the FTC deep end or through.the existing flow path through.RB' penetration R-626'and FHB' penetration 1551 to the RCS manifold at l SDS. Power for the pump'is. supplied from circuit II.of i distribution panel PDP-6-A. From the RCS' manifold,. flow may be- _, filtered through the SDS Pre and Final filters, or may bypass the filters. The receiving tank in either case is one of the'RCBT's'. The fuel Handling, Auxiliary and Reactor Building's HVAC systems provide tempered ventilating air.'and controlled air movement to
. prevent the spread of airborne contamination within the plant or-to the environment. The Nitrogen system provides'N2 for. blanketing
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the RCBT's when transferring to them. The Waste. Gas system stores and processes the gases from the RCBT vents. 3.2 Shallow End' Drainage Operations 3.2.1 Normal Operations The fuel transfer canal shallow end drainage system is~a temporary- '; modification in'the reactor building designed to pump water from the shallow end of the canal and deliver the water to the deepend of.the canal or the reactor coolant bleed tanks (RCBT)'s. 0403B/LC- -
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=.TER 3527-006:
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During defuelingLoperations.Jthe"shallod end.of the FTC'may require-
' drainage as a result of: leakage,-spills, or deliberate flooding of- ;
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the canal -This system provides the means to accomplish this j drainage. ' j The fuel transfer canal shallow end ' drainage operation is started { l and stopped by opening or-closing val _ve FCC-V003 and using on/off
" hand switch DHC-HIS-1. This,.in turn, automatically. starts or stops pump DHC-P-1.
3.3 Shallow End Drainage' Instrumentation
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Pump DHC-P-1 is controlled via hand l indicating switch DHC-HIS-1, which is located on SDS control panel CN-PNL-1-at El._347'-6" of the fuel handling building. The switch' starts and stops'the pump and contains indicating lights'for pump status. The starter, DHC-STR-1, for'the pump is mounted 1 adjacent to panel CN-PNL-1. l A local emergency stop switch, DHC-HS-1, is located in'the Reactor Building near the pump on El. 347'-6". This local switch overrides the indicating switch, and the pump can be started again only after the' local I switch has been reset. 3 Pressure gauge FCC-PI-3 is provided on the canal drain manifold to sense the line pressure downstream of the manifold isolation valves. l' , 0403B/LC 1'
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TER 3527-006
' Air-operated valve FCC-V003 is interlocked with Ne pump such that:the ]
valve must be opened before the pump will Start. .; A:high' level alarm is provided at control panel CN-PNL-1 to inform:the operator to begin draining the. pit. A' low level alarm is.'also provided at CN-PNL-1 to inform the operator to stop the' pump. The' low level alarm d will.not alarm when the' pump.1s'off. i 4 3.4 Shallow End Filtration by SDS. 3.4,1-Filtration i Two sand filters are installed to remove solids from the canal water prior to storage in tanks for future processing. The filters q l contain layers of variously sized sand uniformly mixed with , borosilicate glass which'is added to preclude criticality ! concerns. The filters and related SDS subsystems are unchanged i from that discussed in previous sections of'this TER. , l 3.4.2 Leakage Detection and Processina .! q The filters are located inside submerged containment boxes which are monitored and recirculated through the SDS Leakage Containment l System which is unchanged from that discussed in previous sections ] 1 of this TER.
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3.4.3 Off-Gas and Liould Seperation System An off-gas and liquid separation system collects gaseous and liquid wastes resulting from the operation of the filtration system and sampling. 0403B/LC 1
LTER'3527-006 3.4.4 Sampling System - Sampling'of the flitration influent and' effluent'to monitor filter- '( performance is accomplished using'.the shielded High; Rad filter:- k j Sample Glove Box. This system is discussed in de'ta'il elsewhere in this TER. j 3.4.5 Filter' Vessel Transfer in the fuel Storace Pad Prior to system operation filter vessels are placed inside1the. containment boxes and connected to the system'using quick-disconnect couplings. . When it.is determined that a filter is i L loaded with solids (based on op), the. filter is' flushed with 1 low-activity processed water,' transferred to a storage location'in the pool or the dewatering station, and replaced with a new filter.
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-TER 3527 06
. 1 Chapter 4 d j
RADIATION' PROTECTION' (
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's 4.1 Ensurino Occupational Radiation Exposures are ALARA ..
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-1 l 4.1.1 Overall' Policy-l
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l The objectives with respect to Shal. low End Drainage Operations are l a to' ensure operations are conducted in a~ radiologically' safe-manner and radiation exposure will.be maintained asilow as-reasonably achievable. The effective control of' radiation exposure will be based on the following considerations: q 1, Sound engineering design of facilities and equipment,
- 2. Use of proper radiation protection practices and qualifiable-l personnel. j l
- 3. Strict adherence to THI-2 radiological controls procedures.
.1 4.1.2 SDS Deslan and Operation The SDS design and operational considerations are given in Chapter 6 of the SDS TER. These design and operational .
considerations and features remain unchanged from this evaluation. , The radiation dose exposures to. plant personnel during ! Shallow End Drainage operations will be lower due' to the fact that 1 1 0403B/LC !
TER 3527-006 activities of canal eater should be significantly lower than that experienced during processing of RB sump or initial RCS processing. The SDS shielding l design basis is levels less than 1 mr/hr using 200 pC1/cc Cesium. I 4.1.3 Existing Plant Considerations l The radiation protection features for the existing plant and systems which interface with the SOS are described in Chapter 12 of the THI-2 FSAR. , 4.2 Dose Assessment 4.2.1 On Site Assessment Operation of the SDS Filtration system in the FTC Shallow End Drainage mode may be required intermittently to drain the shallow end of the canal. Based on past SDS operating experience, the exposure for SDS operating area activities due to this operation is expected to be negligible. 4.2.2 Offsite Assessment Source Terms for Liould Effluents All liquid effluent from the system will be retained in station tankage. Source Terms for Gaseous Effluents The plant vent system is a potential pathways for gaseous or airborne release, see section 6.3.2 of this TER. 0403B/LC l
'TER:3527-006 i [Cha'ter5' p
CONDUCT OF OPERATIONAL
- 5.1 System Performance Past processing experience i.e., filtering RB sump and Tank Farm water,
. assures that the' filtration system will perform in the intended. manner.
l~ l -During Shallow En'd Drainage Operations,.the flow rates'through the filters.may appr'oach 50 gp'm. Filters will be taken out of service on high differential. pressure. Filter changeout and dewatering.can be= accomplished in the intended manner using remote long-handled tools. 4 5.2 System Testina Prior to use, each SDS vessel will.be hydrostatically' tested in conformance with the. requirements of applicable portions of the ASME Boiler and Pressure Vessel Code. Upon completion of construction, the' entire system was pneumatically tested to assure leak-free. operations. Individual component and subsystem operability was preoperationally. tested satisfactorily in accordance with approved procedures, q i 0403B/LC l
TER'35k7-006L 5;3': System Operations'
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System operations will be. conducted inLacco'rdance with written and.-
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approved. procedures. These pro'cedures wili.be applicable to normal. system operations, emergency operations., Land required maintenance-evolutions. 'During system operations'on-going operator training and evaluation will.be conducted to ensure continuing safe and efficient system operations. l 1 l
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?TER-3527-006
. Chapter 6 ACCIDENT SCENARIOS t;
6.1- Casualty Events'
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6.1.1 A breech of the; system pressure boundary.while removing water from-
.the shallow end of the fuel transfer canal could result in additional contamination of reactor building surfaces.
'6.1.2 Introduction of this water into'the fuel transfer canal could contaminate the canal.
6.2 Design features to Mitigate Effects of Casualty Events 6.2.1 A hose'or pipe break will result .in loss of line . pressure-, Pressure and flow indication are-provided at various locations on the pump discharge flowpath. The piping.and hoses are l hydrostatically tested to 1.5' times their maximum operating .; l 1 ' l pressure per ANSI B31.1. To ensure pressure boundary integrity, hoses are to be inspected prior to operation;of the canal shallow end drainage network.
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6.2.2 The. fuel transfer canal and the shalloc end drainage system branch - connections of the fuel canal drain manifold'contain double j isolation, which includes a check' valve in each line. -This is to'
' prevent reactor building sump and flush water from being delivered into the' canal. :In addition, the coupling connections on the canali drain and shallow end drain lines of.the manifold are-111/2-inches-and incorporate a two-way shut-off feature. All other manifold-coupling. connections, including the reactor building SWS. system.
connection, are'l-inch diameter. This prevents connecting a 1 1/2-inch pump discharge hose to the 1-inch SHS system' connection which does'not include a check. valve.- QC is'to verify that each hose is connected to the proper manifold branch. connection prior to system. turnover. 0403B/LC _._m . . , , . . . . , . ,- ,, , , , -. . . . . . . . . .. ..A
3 TER 3527-006 J i, l 1 1 Appendix No. 5 i to I i l Submerged Demineralized System Technical Evaluation Report l i TITLE ! 1 f EARLY DEFUELING DWC REACTOR VESSEL FILTRATION SYSTEM l JUNE 1985 j l , (Deleted) j c 1 I l i 1 l l l s l I l l _ _ _ _ _ . _ _ _ _}}