ML20138Q267
| ML20138Q267 | |
| Person / Time | |
|---|---|
| Issue date: | 11/07/1985 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20138Q140 | List: |
| References | |
| FOIA-85-481 PROC-851107-04, NUDOCS 8511180049 | |
| Download: ML20138Q267 (562) | |
Text
{{#Wiki_filter:-- I ll INSPECTION AND ENFORCEMENT TRAINING CENTER PRESSURIZED WATER REACTOR , RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM STUDENT's MANUAL Og118 0 9 851107
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r~ - l i TABLE OF CONTENTS Section Title Page
0.0 INTRODUCTION
. ................ 0.1-1 1.0 LIQUID RADI0 ACTIVE WASTE PROCESSING SYSTEM. . 1.1-1 2.0 GASEOUS WASTE PROCESSING SYSTEMS. . . . . . . 2.1-1 3.0 SOLID WASTE PROCESSING SYSTEM . . . . . . . . 3.1-1 4.0 VENTILATION SYSTEMS . . . . . . . . . . . . . 4.1 -1 5.0 RADIATION MONITORING SYSTEM . . . . . . . . . 5-1 6.0 CHEMICAL AND VOLUME CONTROL SYSTEM. . . . .. 6-1 7.0 BORON THERMAL REGENERATION SYSTEM . . . ... 7-1 8.0 BORON RECYCLE SYSTEM. . . . . . . . . . . . . 8-1 9.0 RADIATION SOURCES . . . . . . . . . . . . . . 9-1 10.0 CHEMISTRY . ................. 10.1-1 l 1
l
.r PRESSURIZED WATER REACTOR l
RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT. SYSTEM i
- CHAPTER 0.0 i
INTRODUCTION A a 7 a f 1 a!
~
TABLE OF CONTENTS Section Title Page
0.0 INTRODUCTION
. . . . . ............ 0.1-1 0.1 GENERAL . . . . . . . . . . . . . ...... 0.1-1 0.2 PROCESSING SYSTEMS. . . . . . . . . . . . . . 0.2-1 0.2.1 Liquid Radioactive Waste. . . . . . . . . . . 0.2-1 0.2.2 Solid Radioactive Waste . . . . . . . . . . . 0.2-2 0.2.3 Gaseous Radioactive Waste . . . . . . . . . . 0.2-3 0.3 SUMt1ARY . . . . . . . . . . . . . . . . . . . 0.3-1 i
i. INTRODUCTION l' GENER'AL A major aspect of nuclear power plant operation is manage- ): ment of the radioactive waste generated as a by-product of nuclear power. ' Of all the problems associated with the nuclear power industry, probably none is so chronic and i its solution so controversial as that of management of the radioactive wastes generated. Management of these
- wastes is complicated not only because of their diverse
- physical and chemical characteristics, but also because i
- of the level and duration of containment required for some of the radioactive constituents.
The development of facilities and equipment to handle and
- process radioactive waste has provided the nuclear industry with capability to treat, process, store or dispose of -
t , ! the radioactive wastes within applicable regulatory require-ments. It is not the intent of this course to present a l " Standard System," for it is clearly recognized that there ' ~ are many equipment combinations which meet the perfomance r objective. 4 1 A number of designs, concepts and operating systems-will l be presented, in addition, applicable Regu,latory Guides f were considered in the development of this text. , 0.1-1
. . _. _ _ _ __. ,_ ._ .,~._ _ _. _ .. _ . . _ . _ - -
I i
)
f The quantities of radioactive waste generated by operation and maintenance activities are dependent upon several factors: type of equipment, equipment arrangement, and operating philosophy. The design, construction, and operation of the Radioactive Waste Processing System shall minimize the release of radioactive material to the environs and in-plant areas which could result in an undue risk to the health and safety of the public or station personnel. The Radioactive Waste Processing System shall be designed and constructed to collect, process, and/or package wastes in a timely fashion to avoid adverse effects to unit capacity or availability while maintaining radiation ex-posures to operating and maintenance personnel "as low as is reasonably achievable" (ALARA). Since plant availability may be affected by the system sizing and redundancy, excess processing and storage capacity as well as component redundancy are largely an economic consi-deration. Successful ~ operation stems from providing adequate storage, sufficient processing capacity, and flexibility in routing feed and process streams. Equipment redundancy and consi-deration of the requirement for sampling, operation, and maintenance while minimizing operator exposure must be an
/
integral part of the design philosophy. 0.1-2
The concentration of radionuclides in the reactor coolant is a function of the reactor power level, the fuel burn-up, type of fuel cladding, the integrity of the fuel cladding, impurities and chemical additives in the reactor-coolant, the reactor coolant volume, and the rate of reactor coolant purification. Appendix I of 10 CFR 50 requires consideration of popula-tion doses from discharged radionuclides at a much lower level than previously permitted. Design objectives are in the range of 3 to 15 MREM / year per reactor, hence any pertinent environmental radiation measurements will have to be extremely sensitive. It also requires the licensee to compute population doses on the basis of effluent measurements and calculational models of radionuclide released to environment. 4 a i ( 0.1-3 l l
I 1 0.2 PROCESSING SYSTEMS 0.2.1 Liquid Radioactive Waste The Liquid Radioactive Waste Processing System collects and processes potentially radioactive wastes for recycle or for release to the environment. Provisions are made to sample and analyze fluids before they are discharged. Based on the laboratory analysis, these wastes are either released under controlled conditions via the Cooling Water System or retained for further use. A permanent record of liquid releases is provided by analysis of known volumes of waste. The bulk of the radioactive liquids discharged from the Reactor Coolant System is processed by the Boron Recycle System. This limits input to the Liquid Radioactive Waste Processing System and results in processing relatively small quantities of generally low-activity level wastes. 9 The Liquid Radioactive Waste Processing System is arranged to recycle as much reactor grade water entering the system as possible. This is implemented by the segregation of waste sources to prevent the intermixing of , liquid wastes. These sources are the miscellaneous waste, chemical waste, detergent waste, and secondary waste. The miscellaneous waste system handles floor drains, aerated equipment drains, and sampling system. The Chemical i 0.2-1 i t
i Waste System handles drains from the Radio Chem Lab, chemical cleaning wastes, and nondetergent decontamination wastes. The Detergent Waste System handles soap and deter-gent wastes. The Secondary Waste System handles steam generator blowdown and turbine-building drains. The processing methods used are filtration, evaporation, ion exchange, and reverse osmosis, j l t For the purpose of this text, the Liquid Radioactive Waste Processing System begins at the interfaces with the Reactor Coolant pressure boundary and the interface valve (s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the Waste Solidification System and . and at the point of recycle back to storage for reuse. 0.2.2 Solid Radioactive Waste Most solid radioactive wastes result from the water treatment or purification of liquid streams in the plant. Solid wastes are broadly classified as either wet or dry. Wet wastes mostly comprise spent demineralizer resins, evapor-ator concentrates, filter sludge, resin cleaner sludge, and reverse-osmosis concentrates.
)
0.2-2
/-
Dry wastes include paper, rags, plastics, clothing, and ventilation air filters. These are processed with deconta-mination steps and packaging appropriate to the particular component. Wet solids are treated in these five basic steps: 1) col-lection, 2) pretreatment, 3) solidification-agent handling,
-4) mixing and packaging, 5) container handling. Solids pretreatment is basically a volume-reduction process, serving to minimize the quantity to be solidified and shipped off-site.
The Solid Radioactive Waste Processing System begins at the interface with the Liquid Radioactive Waste Processing System boundary, at the inlets to the spent resin, filter j sludge, evaporator concentrate, and phase separator tanks. All radioactive or contaminated materials, including liguid filter elements, spent resin beads, filter sludge, evaporator and reverse osmosis concentrates, and dry radioactive , wastes shall be processed in appropriate portions of the { Solid Radioactive Waste System. The system terminates , at the point of loading the filled drums and other containers I on a vehicle for shipping off-site to a licensed burial site. 0.2.3 Gaseous Radioactive Waste The Gaseous Radioactive Waste Processing System is designed ' to remove fission produced gases from the reactor coolant 0.2-3 i _ - , . . , . - . _ , _ . _ . . , - _ _ , , _ _ . _ . , - _ . . . - - - _ - . _ . - - - - _ _ . , , . _ , . , - ~ , . . . . - - . . - _ , - - - - _ . . ,
in the volume control tank. Process methods to be used are either the Pressurized Tank Storage System or charcoal adsorption, both methods store or delay waste gas for decay prior to release to the environment. The Gaseous Radioactive Waste Processing System begins at the point of discharge from plant components, systems and , equipment designed for the removal of radioactive gas from the Reactor Coolant System. The system terminates at the point of introduction into the plant ventilation exhauststream(s). 8 0.2-4
0.3
SUMMARY
In this chapter of the PWR Radwaste Manual we have briefly presented an introduction to the (ALARA) "As Low As is Reasonably Achievable"-program and to the limits of Appendix I of 10 CFR 50. Also presented is a brief de-scription of the Liquid, Solid, and Gaseous Radioactive Waste Processing Systems. They will be more fully described in the following chapters. Other pertinent topics such as radiation monitoring, ventilation, and containment will also be described. 4 0.3-1 [
TABLE OF CONTENTS Section Title Pace 1.0 LIQUID RADI0 ACTIVE WASTE PROCESSING SYSTEM. . 1.1-1
1.1 INTRODUCTION
. . . .............. 1.1-1 1.2 PROCESSING METHODS. . . . . . . . . . . . . . 1.2-1 1.2.1 Fi l t ra ti on . . . . . . . . . . . . . . . . . . 1.2-1 1.2.2 Evaporation . . . . . . . . . . . . . . . . . 1.2-2 1.2.3 Ion Exchange. . . . . . . . . . . . . . . . . 1.2-6 1.2.4 Reverse Osmosis . . . . . . . . . . . . . . . 1.2-7 1.3 SYSTEM OPERATION. . . . . . . . . . . . . . . 1.3-1 1.3.1 Reactor Grade Water (Recycle) . . . . . . . . 1.3-1 1.3.2 Non-Reactor Grade Water (Waste) . . . . . . . 1.3-3 1.4
SUMMARY
. . . .............. . 1.4-1 a
i
LIST OF FIGURES Figure Title Page 1.2-1 Liquid Radioactive Waste Processing System Flow Diagram. . . . . . . . . . . . 1.2-9 1.2-2 Cartridge Filter . . . . . . . . . . . . . . 1.2-11 1.2-3 Precoat Filter . . . . . . . . . . . . . . . 1.2-13 1.3-1 Deaerated Reactor Grade Water System . . . . 1.3-7 1.3-2 Miscellaneous Waste System . . . . . . . . . 1.3-9 1.3-3 Chemical Waste System. . . . . . . . . . . . 1.3-11 1.3-4 Detergent Waste System . . . . . . . . . . . 1.3-13 1.3-5 Secondary Waste System . . . . . . . . . . . 1.3-15 1.4-1 Liquid Radioactive Waste Processing System . . . . . . . . . . . . . . . . . . 1.4-3 l iii
[,- . . . . .
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PRESSURIZED llATER REACTOR RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM CHAPTER 1.0 LIQUID RADI0 ACTIVE WASTE PROCESSING SYSTEM
1.0 LIQUID RADI0 ACTIVE WASTE PROCESSING SYSTEM
1.1 INTRODUCTION
The Liquid Radioactive Waste Processing System collects and processes potentially radioactive waste for recycle or for release to the environment. This system begins at the interfaces with the reactor coolant pressure boun-dary and the interface valve (s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the Waste Solidification System, and at the point of recycle back to storage for reuse. Provisions are made to sample and analyze fluids and based on laboratory analysis, these wastes are either recycled for reuse, released under con-trolled conditions via the Cooling Water System or retained for further processing. The Liquid Radioactive Waste Pro-cessing System is arranged to recycle and reuse as much reactor grade water entering the systems from the Reactor Coolant System as possible by processing this water through the Boron Recovery System. This limits input to the balance of the Liquid Radioactive Waste Processing System and results in processing of relatively small quantities ~ of generally low activity level wastes for release to the environment. 1.1-1
The Liquid Radioactive Waste Processing System shall be designed to meet the requirements of 10 CFR 50 Section 50.34a and the limits specified in 10 CFR 20; control and monitoring of releases shall meet the requirements of 10 CFR 50, Appendix A. The Liquid Radioactive Waste Processing System shall be designed, construci.ed and operated so that no single active failure or administrative failure can result in a release to the environment or exposure to the general public ex-ceeding 10 CFR 7) guide lines. The Liquid Radic tctive Waste Processing System shall be designed, constructed and operated such that radiation exposure to operating and maintenance personnel may be maintained "As low as practicable" in accordance with 10 CFR 20. " Standards for protection against radiation" 10 CFR 20 states that licensees should make every reasonable effort to maintain, exposures to radiation as far below the limits specified in that part as reasonably achievable. Proper planning, design. construction, and operation of a Liquid Radwaste Processing System is necessary to insure the licensee can meet the criterion that exposures of station personnel to radiation during routine operation will be "As low As is Reasonably Achievable" (ALARA). " Reasonably Achievable" is judged by considering the state of technology and the economics of improvements in relation to all of
/
1.1-2
the benefits from these improvements. The goals of the ALARA effort are 1) to maintain the annual dose to indi-vidual station personnel as low as reasonably achievable, and 2) to keep the annual integrated (collective) dose
, to station personnel (i.e., the sum of annual doses [ex- -
press in man-rems] to all station personnel) as low as reasonably achievable. Equipment and components within the Liquid Radioactive Waste Processing System shall be located and arranged to reduce radiation exposure to plant personnel during operation and maintenance. All tanks
- and process equipment shall be shielded so. that normally
{ occupied areas are in a radiation field less than 2.5 MREM /hr. 3 Process equipment shall be located and arranged to provide space for removal and replacement of components and the
- equipment itself. Process subsystems shall be arranged 1
to reduce the length of piping runs. Process equipment such as filters, ion exchangers and evaporators should be i located in individual cells. Equipment design which facilitates rapid disconnect for removal and replacement or ease of in-place maintenance shall be used. In order to help the licensee meet the criterion "As low as practicable" for radioactive material in light-water-cooled nuclear power reactor effluents Appendix I of 10 CFR 50 was established. This is a numerical guide for design objectives and limiting conditions for operation. 1.1-3
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It requires the licensee to compute population doses on the basis of effluent measurements and calculational models. The design objective is in the range of 3 to 15 MREM /yr per reactor. I I 4 l . 1.1-4
1.2 PROCESSING METHODS (Figure 1.2-1) The requisite decontamination factors for liquid waste processing are obtained by the selection and combination of a number of unit operations. At the present time, the principal unit operations for treating radioactive liquid waste are evaporation, ion exchange, filtration, and reverse osmosis. 1.2.1 Filtration Filtration is defined as the separation of undissolved particulate, suspended solids from a fluid mixture by passage of most of the fluid through a septum or membrane-that retains the solids 'on or within itself.
- 1) Cartridge Type Filter (Figure 1.2-2)
Cartridge filters are designed with replaceable ele-ments which are discarded when contaminated. These elements are usually constructed of pressed paper, matted fibers, or porcelain materials. In general, these types of elements are best suited for the re-moval of gross contamination from low pressure, low temperature systems.
- 2) Precoat Type Filter (Figure 1.2-3)
Precoat filters use elements designed as retainers to prevent a precoat material from being flushed down-stream. The actual filtering of system contamination is accomplished by the precoat material. When the 1.2-1
filter becomes contaminated, the precoat material is replaced. This type of filter can provide high levels of filtration efficiency, but is limited by temperature and pressure extremes. In addition, the large quan-tities of precoat material which must be disposed of can present significant problems in terms of handling and disposal. 1.2.2 Evaporation Evaporation is used to remove the solvent from its impuri-ties, both dissolved and suspended, and is widely used for reducing waste volumes and radioactive nuclides in liquid effluents. . Evaporation can be used on solutions having vastly different compositions and concentrations; however, it is most effectively used on waste solutions having high concentrations of impurities. Evaporation is the process by which a solution or slurry is concentrated via boiling away the solvent. It is a unit operation that has wide application in the nuclear industry for reducing waste volumes and for reducing the amount of radioactive nuclides in liquid effluents. Evaporation is usually used for radioactive wastes that require the high' degree of separation between volatile and nonvolatile components that it o'ffers or for wastes that are not amenable to treatment by low-temperature
. ]
1.2-2
l operations such as filtration or ion exchange. An evap-orator consists basically of a device to transfer heat ! 1 for boiling the solution or slurry and a device to separate the vapor phase from the liquid phase. The principal ele-ments involved in evaporator design are heat transfer, vapor-liquid separation, volume reduction and energy uti-lization. In the design of evaporators for concentrating radioactive liquids, vapor-separation is the most important factor because decontamination of the liquid is the most important objective and heating costs and volume reduction are relatively less important. A small amount of entrain-ment can contaminate the condensed vapor and reduce the decontamination to unsatisfactorily low levels. The device in which heat transfer takes place is termed a heating element, many evaporators are equipped with ex-ternal heaters, heaters may be vertical or horizontal, long- or short-tube. Circulation of liquid past the heating surface may be induced by the density variations b' ought on by boiling (a natural-circulation evaporator) or by mechanical means such as a pump (a forced-circulation evaporator). The device in which vapor-liquid disengagir.g takes place is called a flash chamber, a vapor head or sometimes a body. Evaporators used to concentrate radioactive wastes com-monly have a large flash chamber so that the flow rate of the vapor is slow, a condition which can induce efficient vapor-liquid separation. 1.2-3
There are several modes in which evaporators can be operated: batch, semibatch, and continuous. In batch operation, the evaporator initially contains the entire quantity of liquid to be processed. In semibatch operation, the feed in con-i tinually added to maintain a constant level in the evapora-tor until the entire charge reaches a final density for drumming in the Radioactive Solid Waste System. In contin-uous operation, feed and product (bottoms and distillate) flows are kept constant; operation is at steady state. Decontamination factor for radwaste evaporators are ex-pressed as the feed / distillate concentration ratios. i In evaporating radioactive solutions effective entrainment separation is required to avoid contaminating the condensate. Equally important in evaporating radioactive liquids is operation under conditions to suppress the volatilization of radioactive materials and some organics that can ha~ve high vapor pressures. Evaporators can separate water from solids very effectively, and a system decontamination 4 5 factor of 10 - 10 is generally expected for a single-effect evaporator separating water from a nonvolatile solute. Decontamination factors are decreased by four basic factors: entrainment, splashover, foam, and _ volatilization of solute. j
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1.2-4
Entrainment is liquid suspended in the vapor as fine drop-lets that are carried along with the rising vapor stream. The extent of entrainment losses from an evaporator depends, for the most part, on the vapor velocity and the size dis-tribution of the droplets. The larger d*cplets are a major source of entrainment losses unless they settle back into the liquid or are removed by entrainment separators (de-entrainmentdevices). Evaporators generally have devices incorporated or attached to remove the entrained larger droplets that do not settle back. Splashover consists of the carryover of large parcels of bottoms into the condenser. It occurs at very high boilup rates when boiling becomes violent and erratic. Foam implies a mass of stable bubbles formed in or on the surface of the bottoms. Among the causes of foam in an evaporator are traceslof organics, finely divided solids, and dissolved gases. It leads to an increase in entrainment by raising the effective liquid level. Volatilization of solute is a concern in the operation o~f evaporators with liquid radioactive waste since the design objectives for release of radioactive materials to the environment are low. Several elements found in radwaste, such as iodine, are volatile to varying degrees. Methods
. 1.2-5
l l for decreasing the volatility include adjusting the PH and i RED 0X potential of the solution or slurry to produce new conditions under which they are nonvolatile. 1.2.3 Ion Exchange Ion exchange resins are insoluble high molecular weight polyelectrolytes which can reversibly exchange their mobile ions of equal charge from the surrounding' solution. Those ' exchangers in which the anionic portions are able to react, or are mobile, are anion exchangers; those in which the cationic portion is mobile, cation exchangers. Resins used in radioactive liquid waste applications are either in the hydrogen or in the hydroxyl form. Deionization with mixed bed ion exchange resins is capable of producing a water of exceptional purity. This technique involves the passage of an aqueous solution through an intimate mixture of a cation resin in the hydrogen form and an anion resin in the hydroxide form. By using strong acid and strong base,known as Type I, resins, it is possible to consis-tently produce an effluent having a conductivity of less 0 On occasion, the effectiveness of than 1 umho at 25 C. an ion exchange material may be impaired due to the accu-mulation of insoluble materials, such as oil, colloids or particulates, on the surface and in the interior of the ion exchange particles. It is best to avoid or minimize
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1.2-6
fouling by use of suitable pretreatment filtration. Ion exchange is most cost effective when used on low conduc-
'tivity selutions; above approximately 100 umho/cm (at 25 C),
reverse osmosis or evaporation is more effectively utilized. 1.2.4 Reverse Osmosis Reverse osmosis is a membrane process that, to oversimplify, acts as a molecular filter to remove almost all dissolved minerals and dissolved organics with molecular weights over 100 and all biological and colloidal matter from water. With this process, pure water is separated from a salt solution by a semipermeable membrane which passes water readily, but retards the flow of ionic solids. Osmosis is the natural process whereby pure water flows through the membrane from a dilute solution into a more concen-trated one, thereby diluting the latter. However, if pressure is applied to the salt side at a level that is greater than this osmotic pressure of the salt solution, 4 pure water will ' flow across the membrane from the salt solution. At.the present time, reverse osmosis has found extensive use in radioactive service for the treatment of laundry and detergent waste. i 1.2-7 4
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1.3 S_Y1 TEM OPERATION The Liquid Radioactive Waste Processing System consists mainly of a reactor grade water or recycle section and a non-reactor grade water or waste section. 1.3.1 Reactor Grade Water (Recycle) Figures 1.3-1 and 1.3-2) The recycle section is provided to process reactor grade water which enters the Liquid Radioactive Waste Processing System:
- 1) Deaerated tritiated Reactor Coolant System water i
; collected directly in the reactor coohnt drain tank may be routed directly to the boron recycle holdup tanks for processing in,the boron recovery system or routed to the miscellaneous waste tank.
- 2) Tritiated and aerated water is collected in the 4
miscellaneous waste tank. The major contributors of water which can be recycled are: , 1) Sampling Station Radioactive Waste l 2) Aerated Systems and Equipment drains i
- 3) Reactor Coolant Auxiliary System drains
- 4) Emergency Core Cooling $futem drains
- 5) Containment Cooling System drains
- 6) Process and Component Cooling System drains.
l Considerable surge and processing capacity is incorporated in the recycle portion of the Liquid Radioactive Waste i 1.3-1
Processing System to accomodate abnormal operations;e.g., leaks which may develop in the reactor coolant and auxiliary systems. The system to process miscellaneous waste shall include evaporation or equivalent processes. Evaporation may be preceded by filtration to minimize the deposition of par-ticulate matter on evaporator components, such as sensory elements and heat exchanger surfaces. Ion exchange may be used to " polish" the evaporator distillate. Evaporation is selected as a primary processing method due to the normally high decontamination factors which can be achieved for ionic and particulate matter. The basic composition of the liquid collected in the -
- miscellaneous waste tank is boric acid and. water with some radioactivity. Liquid collected in this tank is evaporated 4
to remove radio-isotopes, boron, and air from the water so that it may be reused in the Reactor Coolant System. Evaporator bottoms are normally drummed unless found accep-table for boric acid recycle. The condensate leaving the miscellaneous waste evaporator may pass through the miscel-laneous waste condensate polishing demineralizer and then enter monitor tank #1. When a sufficient quantity of water has collected in the monitor tank #1, it is normally trans-ferred to-the primary water storage tank for reuse. Samples
]
1.3-2
. . . = - ._ ~ . _ _ . .- . . - . .
are taken at sufficiently frequent intervals to assure 4 proper operation of the system to minimize the need for reprocessing. If a sample indicates that further proces-sing is required, the condensate may be returned to the miscellaneous waste tank for additional evaporation.
- 1.3.2 Non-Reactor Grade Water (Waste) l 4
The waste section is designed-to collect and process non-reactor grade liquid wastes. These include chemical wastes, detergent wastes and secondary system wastes.
- 1) Chemical Waste (Figure 1.3-3)
The input sources are: radiochemistry laboratory 4 drains, chemical cleaning waste, decontamination waste, God other liquid radioactive wastes which contain high concentrations of chemicals. 4 The chemical waste may be treated separately, but it may be processed either in the Solid Waste System
- or in the Miscellaneous Waste System. If processed
, in the Miscellaneous Waste System, it should be on i 2 a-batch basis. Chemical wastes may be_directly pro-cessed in the' Solid Waste System if they have a very high solids or radioactivity content. If a separate chemical waste system-is provided, it shall be designed l for filtration,.followed by evaporation. I 1.3-3
Ion excnange should be used to polish the effluent. Provisions shall be included to route treated waste to a monitor tank for analysis prior to release or recycle.
- 2) Detergent Waste (Fiaure 1.3-4)
Input sources include: laundry, personnel deconta-mination, and other liquid radioactive wastes con-taining detergents and soaps. Detergent waste collection systems should be segregated from other waste collection systems to reduce operational problems with processing equipment, and shall provide
- filtration as a minimum. Alternatives should be in-c.luded to route treated detergent waste to a monitor tank for analysis prior to release to the environment or processed directly in the Solid Waste System.
- 3) Secondary System Waste (Figure 1.3-5)
The input sources are: steam generator blowdown, tur-bine building drains, ion exchange spent regenerant and filter waste. If separately provided, the Steam Generator Blowdown Processing System shall include filtration and ion exchange or evaporation and routed to a radiation monitor prior to release to the environ-me4. The Miscellaneous Waste System may be utilized l 1.3-4
to treat turbine building drains and steam generator blowdown. The Chemical Waste System may be utilized to treat ion exchanger spent regenerants and filter wastes. e 9 1.3-5
l DEAERATED REACTOR GRADE WATER SYSTEM RCP PRESSURIZER RCS SIS REFUEL CVCS LETDOWN :t 2 SE AL RELIEF LOOP ACCUM CANAL HE AT EXCHANGER LEAK 0FF TANK DRAINS DRAINS DRAINS u u y u y u J L JL P VALVE VESSEL
" FLANGE Y
LEAK 0FF LEAKRFF REACTOR COOLANT ORAIN TANK PUMP INSIDE CONTAINMENT OUTSIDE CONTAINMENT l i
, TO TO l
g MISCELLANEOUS : : BORON RECOVERY WASTE TANK SYSTEM
}
P Y. l
MISCELLANEOUS WASTE SYSTEM OUT000R CONTROLLEO ARE A WASTES ECCS AER ATEO SYS. FUEL FLOOR REACTOR SAMPLE STATION A 0 EQUIP. HANDUNG ORAINS ORAINS SYSTEM COOLANT RA010 ACTIVE RHR AU X. SYSTEMS WASTE v v v v v u u y CONTAINMENT COMPONENT WASTE STE AM GEN. COOLING AND AUX. COOLING DISPOSAL BLOWOOWN SYSTEM SYSTEM RECYCLE AU X.10 N WATER SYSTEM FROM MONITOR EXCHANGER ANO FILTER NOTE:[ TANK WASTE MAY GO TO SECONDARY u i, e u ir WASTE TANK ' ir MISCE LL AN EO US WASTE TANK FROM CHEMICAL WASTE TANK PUMP
=
MISCELLANEOUS 10 N FILTER - WASTE - TO EVAPORATOR EXCHANGER RECYCLE y L MONITOR TANK
=1 TO :
RELEASE Figure 1.3-2 1.3-9
CHEMICAL WASTE SYSTEM (NONDETERGENT) DECONTAMINATION WASTE RA010 CHEM. CHEMICAL CLEANING DEMIN REGENERANT L AB ORAINS WASTE SOLUTIONS NOTE: MAY GO T0 p SECOND ARY WASTE TANK CHEMICAL WASTE TANK w
~
PUMP C TO CHEMICAL MISCELLANEOUS : : l FILTER WASTE ION
"^
WASTE SYSTEM EVAPORATOR U u TO SOLIO WASTE SYSTEM [ MONITOR TANK y, N O. 2
% TO RECYCLE 2 TO MISC. TO WASTE TANK " RELEASE I
W ?
DETERGENT WASTE SYSTEM PERSONNEL . EQ UIPMENT LAUNDRY DECONTAMINATION DECONTAMINATION 2 o DETERGENT WASTE TANK PUMP I w" FILTER TO SOLID REVERSE WASTE < : OSMOSIS SYSTEM 1. 1 i r y TANK
, TO ,
y RELEASE ' l
SECONDARY WASTE SYSTEM STE AM GEN TURBINE BLOG. DEMIN REGENERANT BLOWOOWN ORAINS SOLUTIONS NOTE: NMAY GO TO [
\
MISC. WASTE \ NOTE:
" " " MAY GO TO TANK CHEM WASTE o TANK SECONDARY WASTE TANK PUMP FILTER o p 10N EXCHANGER EVAPORATOR o
l MONITOR l TANK NO.2 TO RELEASE Figure 1.3-5 1.3-15
1.4 5UMMARY (Figure 1.4-1) In 1971, AEC issued its proposed Appendix I to 10 CFR 50. It required plants to reduce the amount of radioactivity 4 released to the environment to a level "As Low as Practi-cable," and also establisted limitations. The regulation states that the design objective for radioactive releases 4 with liquid effluents is to limit the annual exposure for an individual in an unrestricted area to not more than three millirem to the whole body, and not more than ten millirem to any organ. To further reduce the activity 4 of water ready for release meant additional processing and purification, after which the water usually can be reused within the plant.
~lhe other change in environmental regulation was one re-lated to condenser cooling water. Several plants nave had to convert from once-through cooling to a closed cooling-water system - usually involving either a cooling tower or a lake. Allowing only for periodic blowdown, this effectively reduced the dilution flow by 90 - 95%.
Since discharge limits had applied to the concentration of radioactivity, this resulted in reduction of the total quanitity of radioactive effluent that may be discharged e within existing regulations. i Generally, the water entering the radwaste system comes from equipment drains, floor drains, water-treatment waste, laboratory and sample drains, and liundry drains. 1.4-1 l
1 i Water management, a key element of the program, aims at j reducing radioactive-water releases to "As Low As is Reasonably Achievable." Minimizing liquid discharges during normal plant operation requires that the plant main-i tain a water balance, and maximizes internal water reuse. 4 Adequate storage and processing equipment are essential
- to water recycle.
W G 1.4-2
R R R K OGE O E N FNI TS F T ESA E A TA L CESW L F E
. C YC C Y Y G C OS C R A O ERI E A R T RPM R P O I
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_ . - _ _. . _ . - . . .-. ._. . . _. i. . _. _ .[- - . I I PRESSURIZED WATER REACTOR RADIOACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM 9 CHAPTER 2.0
< GASE0US WASTE PROCESSING SYSTEMS e
*e v r r
- ~ . -. - - _ _ . _ ~ . . - ..
A i 4 TABLE OF CONTENTS i Section Title Page i 2.0 GASE0US WASTE PROCESSING SYSTEMS. . . . . . . 2.1-1
2.1 INTRODUCTION
. . . . . . . . . . . . . . . . . 2.1-1 1 2.2 STORAGE AND RELEASE SYSTEM. . . . . . . . . . 2.2-1 i 2.2.1 Introduction. . . . . . . . . . . . . . . . . 2.2-1
'2.2.2 System Design and Operation . . . . . . . . . 2.2-4 2.2.3 Components. . . . . . . . . . . . . . . . . . 2.2-7 l 2.2.4 Summary . . . . . . . . . . . . . . . . . . . 2.2-8
- 2. 3 ' . VOLUME REDUCTION SYSTEtiS. . . . . . . . . . . 2.3-1 2.3.1 Introduction. . . . . . . . . . . . . . . . . 2.3-1 i 2.3.2 System for Use with Continuous VCT Purge. . . 2.3-2 2.3.2.1 Introduction. . . . . . . . . . . . . . . . 2.3-2 2.3.2.2 System Design and Operation . . . . . . . . 2.3-3 2.3.2.3 Estimated Releases . . . . . . . . . . . . . 2.3-5 i
'2.3.2.4 Components. . . . . . . . . . . . . . . . . 2.3-8
'2.3.3 Systems for. use with CVCS Gas Strippers . . . 2.3-9 2.3.3.1 I n trod u cti on . . . . . . . . . . . . . . . . 2.3-9 2.3.3.2 System Design and Operation . . . . . . . . 2.3-10 2.3.3.3 Components. . . . . . . . . . . . . . . . . 2.3-15 j 2.3.4 Summary . . . . . . . . . . . . . . . . . . . 2.3-19 2.4 CHARC0AL ADS 0RBER SYSTEMS . . . . . . . . . . 2.4-1 2.4.1 I ntroducti on . . . . . . . . . . . . . . . . . 2.4-1 2.4.2 System Design and Operation . . . . . . . . . 2.4-1 2.4.2.1 Ambient Charcoal. . . . . . . . . . . . . . 2.4-1 2.4.2.2 Cryogenic Charcoal. . . . . . . . . . . . . 2.4-3 2.4.3 S umma ry . . . . . . . . . . . . . . . . . . . 2.4-4 l 2.5 MISCELLANE0US GASE0US WASTE SYSTEM FUNCTIONS. 2.5-1 2.5.1 I n t rod uc ti on . . . . . . . . . . . . . . . . . 2.5-1 2.5.2 Plant Nitrogen' Systems. . . . . . . . . . . . 2.5-2 2.5.3 Plant Hydrogen System . . . . . . . . . . . . 2.5-2
-2.5.4 Air . Ejector Off-Gas System. . . . . . . . . . 2.5-3
'2.5.5 Hydraulic Baler Package Vent System . . . . . 2.5-3 2.5.6 Gas Analyzer. . . . . . . . . . . . . . . . . 2.5-4 l
5 i f i 'i l i . I _ _ . _ ._._2,~ _. - _ _ , - , _ . . _ . _ _.- .. m . ;,... _ . , _ . _ . _ , _ , . _ _ . . . , ,
LIST OF TABLES Tables Title Page i 2.1 -1 PWR Gaseous Radioactive Waste Sources Expected and Design Basis Quantities of I-131 and Noble Gases Prior to Treatment 2.1-7 2.1 -2 PWR Gaseous Radioactive Waste Processing System Design Basis Inputs. . . . . . . . . 2.1-11 2.2-1 Estimated Annual Gaseous Release by Isotope . 2.2-9 2.3-1 Gaseous Activity Released to Vent Stack from Leaks in the Gaseous Waste Processing System. . . . . . . . . . . . . . . . . . . 2.3-21 2.3-2 Accumulated Radioactivity Per Unit in the Gaseous Waste Processing System After Forty Years Operation . . . . . . . . . . . 2.3-23 2.3-3 Expected Accumulated Radioactivity Per Unit in the Gaseous Waste Processing System
, After Forty Years Operation . . . . . . . . 2.3-25 2.3-4 Reduction in Reactor Coolant System Gaseous Fission Products Per Unit Resulting from Normal Operation of the Gaseous Waste P roces s i ng Sys tem . . . . . . . . . . . . . 2.3-27 2.3-5 Sources, Volumes and Flow Rates to the Gas S u rge Hea de r. . . . . . . . . . . . . . . . " 3-29 2.3-6 Specific Activities of Sources to the GSH During Normal Opera tion . . . . . . . . . . 2.3-31 LIST OF FIGURES Figures Title Page 2.1-1 PWR Storage and Release - Waste Gas Decay Tank. . . . . . . . . . . . . . . . . . . . 2.1-15 2.1-2 Typical PWR Gasec - Waste Processing System (Storage and Release) . . . . . . . . . . . 2.1-17 l
iii
N LIST OF FIGURES (CONTINUED) Figures Ti tle Page 2.3-1 PWR Storage and Release - Nitrogen Recycle. . 2.3-33 2.3-2 Typical PWR Gaseous Waste Processing System
- (Storage and Release with Nitrogen Recycle) 2.3-35
! 2.3-3 Typical PWR Gaseous Waste Processing System (Storage and Release with N Recycle and CVCS Gas Stripping) . . . 2 . ....... 2.3-37 1 i 2.3-4 Waste Gas Processing System Fission Gas Accumulation Based on Continuous Core Oper-ation at 3565 MWt with 1% Fuel Defects, Stripping Efficiency 100% . . . . . . .' . . 2.3-39 i 2.3-5 Estimated Waste Gas Processing System Fission l Gas Accumulation Based on Full Power Oper-ation of 3565 MWt . . . . . . . . . . . . . 2.3-41 2.4-1 PWR Ambient Charcoal. . . . . . . . . . . . . 2.4-7
- 2.4-2 Typical PWR Gaseous Waste Processing System (Ambient Charcoal ) . . . . . . . . . . . . . 2.4-9
- 2.4-3 PWR Cryogenic Charcoal . . . . . . . . . . . . 2.4-11 i
2.5-1 Miscellaneous Gaseous Systems . . . . . . . . 2.5-7 2.5-2 ' Gas Analyzer System . . . . . . . . . . . . . 2.5-9 ) i, 4 i f i i l
-iv .
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1 2.0 GASEOUS WASTE PROCESSING SYSTEMS
-2.l' INTRODUCTION--
.The Gaseous Radioactive Waste Processing Systems for a
~~
5 pressurized water reactor begin at the point of discharge ~ from plant components, systems and equipraent designed to i remove radioactive gases from.the reactor coolant system and reiated auxiliary systems. The Gaseous Waste
- Processing Systems terminate at the point of introduction into the plant ventilation exhaust system.
3 In a PWR, waste gases originate from several sources. These sources may be intermittent or continuous depending upon plant design. Tables 2.1-1 and 2.1-2 show the design . bases flows and activities for the various sources. A brief description of each of these sources is given below: Volume Control Tank (VCT) - The hydrogen concentration in , the reactor coolant system is controlled by the absorption of hydrogen as the RCS letdown is sprayed into the hydrogen i ~ gas space of the VCT. Since the reactor coolant contains
~
radioactive gases, the volure control tank gas space will
. accumulate them in various concentrations. Gas from the f
volume control tank may be either continuously purged or
- periodically. vented to the waste gas processing system.
Reactor Coolant Drain Tank (RCDT) - The Reactor Coolant - !- Drain Tank is used-to collect liquid' effluents from various - 1 2.1 -1 i' i
,, --- , , , , , . - - , , , ,,,* , . , , -.v.-.- ~ . . , , , . , ~ - , - - . , , . - -. ,, . . - - ,
reactor coolant system controlled leakoffs. Radioactive gases accumulate in the RCDT gas space and enter the Gaseous Radioactive Waste Processing System through the tank vent. Activity from this source is highly variable and depends upon process and system design. Gas Stripper - The degassifier or gas stripper removes dissolved gases from liquids (reactor coolant, boron recycle system, etc.) and discharges them to the waste gas processing system. Flow rates and activity from this source depend upon system design. Cover Gases - Cover gases (usually nitrogen) used to exclude air from various tank vapor spaces to limit oxygen concentration in the water or to prevent the formation of explosive mixtures with hydrogen, may contain radioactive gases. These gases enter the Gaseous Radioactive Waste Processing System as tank levels are increased. Pressurizer Relief Tank (PRT) - Radioactive gases from any pressurizer relief valve lifting or leakage enter the PRT. Periodically, the PRT is vented to the gaseous waste system. The gas space in the PRT is normally filled with nitrogen so most of the gas vented will be nitrogen with some fission gases. In some plant designs the PRT and RCDT functions are combined. Y 2.1-2 I
Miscellaneous Sources - Gases from other sources such as reactor vessel head and pressurizer venting or fuel inspection (sipping) operations may discharge to the Gaseous Radioactive Waste Processing Systems. Air Ejector Exhaust - The main condenser air ejector exhaust consists of air and water vapor which may contain radioactive material in the event there is steam generator tube leakage. Gases from this source are not normally treated in the
- Gaseous Radioactive Waste Processing System.
Steam Generator Blowdown Tank Vent - During plant operation, flashed steam from the steam generator blowdown flash tank is normally released directly to the environment. This release may contain radioactive noble gases, radiciodines, and radioactive particulates. Equipment Vent Header System - Aerated vents from equipment containing radioactive liquids may be piped to a separate header system for release. These discharges may contain radioactive noble gases, radioiodines, and particulates. As can be seen from the preceding list, the gases processed by the waste gas system consist mainly of hydrogen and nitrogen with small amounts (by volume) of fission gases from the various process streams. No aerated gases are processed to preclude the formation of explosive oxygen-hydrogen mixtures. The radioactive gases of primary interest l l l 2.1-3 I
are Xe 133 and Kr 85 as these are the only ones present in significant amounts which have relatively long half lives. These waste gases typically are combined in a Gaseous Radioactive Waste Processing System for common treatment. There are a number of process designs which meet the performance objectives for gaseous radwacte systems. The Gaseous Radioactive Waste Processing Sys = = selected should be designed to process the waste gas streams in combinations appropriate to the sources of the specific plant design. Sections 2.2 through 2.4 describe four examples of typical Gaseous Radioactive Waste Processing System designs. The systems preser.ted consist of three process designs using
. tank storage systems and one process design using charcoal adsorption. The illustrated process designs differ in the method of gas stripping of primary coolant, the method of startup cperations and the method of handling gases from other sources.
One of the differences shown in the illustrated process designs is the use of a recombiner to react the hydrogen ^ in the waste gas with separately supplied oxygen. Where used, this feature reduces the quantity of waste gas thereby permitting either a reduction in the size of the 2.1-4 i
l I tank storage or delay system or an increase in the hold-up or delay time compared to systems having the same storage volume but no recombiners. Another of the differences between illustrated process designs is the method of the reactor coolant gas stripping. In one case gases are removed from the volume control tank by a continuous hydrogen purge. In the other a gas strip-per located in the reactor coolant purification flow path removes gases continuously. 1 Each of the described systems will be designed to meet the following criteria:
- 1. Radioactive materials in gaseous effluents from the
; plant collectively meet the design objectives given in 10 CFR 50, Appendix I, and limits specified in 10 CFR 20. Control and monitoring of release of radio-active materials to the environment will meet the requirements of 10 CFR 50, Appendix A, Design Criteria 60 and 64.
- 2. Accidental release of radioactive material from a single component would not result in an off-site dose which would exceed the guidelines of 10 CFR 100.
- 3. Radiation exposures to plant operation and maintenance personnel will be maintained "as low as reasonably l
achievable." l l 2.1-5 l
TABLE 2.1-1 PWR GASEOUS RADI0 ACTIVE WASTE SOURCES EXPECTED AND DESIGN BASIS QUANTITIES OF I-131 AND N0BLE GASES DRIOR TO TREATMENT SOURCES ACTIVITY (CURIES /YR) N0BLE GASES 10 DINE-131 Expected Design Basis Exoected Design Basis Gaseous Waste Prccessing System (10) 4 Tank Storage System (6) 3.8 x 10 Negligible Negligible Without CVCS Stripping Adsorption System or ( ) 2.5 x 10 5 Negligible Negligible Volume Reduction System With Partial Stripping At Volume Control Tank Adsorption System or (8) 3.2 x 10 5 Negligible Negligible Volume Reduction System With CVCS Stripping Air Ejector Exhaust For Plants with GRWPS(2,3,6,9) 260 .027 2.6 Tank Storage System Without CVCS Stripping For , Plants With GRWPS(2,3,7,9) 56 .027 2.6 Adsorption System or Volume Reduction System With Partial Stripping At Volume Control Tank For Plants With GRWPS(2,3,8,9) 29 .027 2.6 Adsorption System or Volume Reduction System With CVCS Stripping Blowdown Flash Tank Vent (4,5) U-Tube Steam Generators Negligible .18 17 With AVT Without Condensate Polishing U-Tube Steam Generators Negligible .085 8.1 With AVT With Condensate Polishing 2.1-7
- ~ . . ~ .. .
i' i TABLE 2.1-1 (continued) NOTES: 1. All sources are for a single 3400 mwt plant. Adjustment of sources for other power levels are based upon ANSI N237.
- 2. Primary to secondary leak rate 100-pounds (450 kilograms) per day j expected,1200 pound (550 kilograns) per day design basis.
- 3. Volatile iodine fraction in primary coolant .05 totally vaporized in steam generators and becomes soluble in condenser with partition t factor of 0.15 for volatile species and zero for non-volatile
{ species. Iodine in condensate converted to non-volatile species. i
- 4. -Steam generator blowdown rate,1% of main steam flow for AVT, and q not-applicable for once through steam generators. Partition factor =
l .05 if steam is not condensed prior to venting. -
- 5. Negligible source for systers including vent condensers venting to 1 main condenser, or blowdown coolers without flash tanks.
i 6. Expected activity concentrations based upon primary coolant activity
- concentrations given in ANSI N237 and complete stripping of boron
- control letdown flow 9 500 lb/hr (230 kilograms /hr).
- 7. Expected activity concentrations based upon primary coolant activity j
concentrations given in ANSI N237 (modified by formulae given in ANSI N237 for CVCS stripping). Calculations based upon 40% strip-ping of letdown stream 9 37000 lb/hr (17000 kilograms /hr).
- 8. Expected activity concentrations based upon primary coolant activities
- concentrations given in ANSI N237 (modified by formulae given in ANSI j N237 for CVCS stripping). Calculations based upon 100% stripping of
- letdown stream 9 37000 lb/hr (1700 kilograms /hr).
- 9. Design basis iodine activity concentrations are equal to eight times 1 the expected activity concentrations.
- 10. Design basis noble gas quantities shall be based upon 1% failed fuel l and the spectrum specified by the supplier of the nuclear steam supply system.
l t . 2.1-9 o
., . --,-m ,, .---_ , _ - - , . . ~ - . , - , , ,,,, ,,.---,-..--.,,-m. - _ - - - , , . _ , . . , . . - - . - . _ . , - . . , , - _
TABLE 2.1-2 PWR GASE0US RADI0 ACTIVE WASTE PROCESSING SYSTEM DESIGN BASIS INPUTS PROCESS SOURCE COMPOSITION QUANTITY _ Normal Range Annual Quantity Standard Standard Standard Standard Cubic Liters Cubic Liters Feet / Min Per Min Feet Volume Control Tank With Continuous Purge for H() 2
.7-1.4( ) .33 .66 300,000(3) 140,000 Stripping Intermittent Purge H,N r 0-4 0-2 3,000 1,400 2 2 Mixtures Degassifier/ Stripper Boron Recycle Stripper H .25-1.34 .123 3,600 1,700 2
(Intermittent) CVCS Letdown Stripper H 2
.3 .8(4) .14 .38 150,000(5) 70,000 (Continuous)
Reactor Coolant Drain Tank Bith Controlled Liquid H,N p' or 0-5 0-2.5 200(6) 95 2 Level (Intermittent) Mixtures Mith Varying Liquid Level H,N or 0-5 0-2.5 12,000 I7) 5,700 2 2 Mixtures Pressurizer Relief Tank (9) N 0-40(8) 0-20 2 Fuel Inspection I9) N 4 2 2,000 950 2 2.1-11
. 1 I
TABLE 2.1-2 (continued) NOTES: 1. Changes to 100% Nitrogen during shutdown degassing.
- 2. Based upon a maximum flowrate of 40SCFM (20 standard liters /second)
(maximum capacity of compressor) to the recombiner, and a maximum hydrogen concentration of 4% in the recombiner effluent.
- 3. Based upon one start-cp and shut-down operation; uses 100% stripping and adsorber efficiency, a pressure of 2 atmospheres absolute in Volume Control Tank, hydrogen concentration 25-35 cc/kg at start-up and 5 cc/kg at shutdown.
- 4. Based upon stripping of coolant in boron recycle system with hydrogen concentration of 25-35 cc/kg and flowrate of 75-286 gpm (5-18 liters /
second).
- 5. Based upon stripping approximately 2% of the boron recycle flow.
- 6. Based upon 300 gallons / day of reactor coolant leaking to the Reactor Coolant Drain Tank with one-half of the hydrogen leaving solution and entering the Gaseous Radioactive Waste Processing System.
- 7. Based upon a liquid flowrate of 300 gallons / day (1140 liters / day) for 300 days / year.
- 8. Based upon the continuous operation of a compressor with a 40 SCFM (20 standard liters /second) capacity.
- 9. Certain plant desigas to not release waste gas from this source.
- 10. See Table 2.1-1 for Activity Tenns.
2.1-13
PWR STORAGE AND RELEASE - WASTE GAS DECAY TANK RCS VENTS SURGE TANK FILTER VCT VENT , G AS STRIPPER
~
A-BRS HOLDUP TANK VENT COMPRESSOR m TO PLANT VENT j( L 1 P 1 P 1 P HEPA l ' ' ' d b FILTER u 4 f f f f 2. S STORAGE T ANKS s N i I , r , r i r J L J b d L f i
CVCS G AS ST RIPPE R ? VOLUME CONTROL TANK p TO Pt ANT VENT HOLOUPTANKS. p_ MANUAL _ j( CONTROL BORIC ACIO EVAPORATOR O- r--- H1 lCCWl l PRESSunt2ER REllEF T ANK 3 p l RE ACTOR COOLANT ORAIN TANK = COMPRESS RS ~ l RAD WASTE EVAPORATOR L- lCCWl MONITOR 4 GAS ANALYZER RETURN 3 p
- VCT GAS SAMPLE RETURN 3 PL ANi hyHEADER 3 s r o
) J J J J , ,
7 I
) J ) JA )J )J )J i A A % % A N A H[H[,[] H[H:, _ ] H[H:,[] H[ , ] qH[, ]H [H[, ]
D r% 6 s$ : r% 5 s4 :s4 r% 4 P J P [ 3 2 I COVER G AS FOR N N2 -M NM N 7M N -M - Nf HOLOUPTANKS H: Ht H[ Nt H: H: LY ( ( ( ( ( GAS N AN AL Y2E R y y L10010 RA0 WASTE N SYSTEM TYPICAL PWR GASEOUS WASTE PROCESSING SYSTEM (STORAGE AND RELEASE) 4
2.2 STORAGE AND RELEASE SYSTEM 2.2.1 Introduction The gaseous waste processing system in use at most present-generation pressurized water reactor facilities is designed to collect and store gaseous waste for radioactive decay prior to controlled release to the environment. The system consists of a collection header into which the various sources of waste gas discharge, compressors to reduce the volume of the gas, and tanks in which the pres-surized gas is stored for decay prior to release. Provisions are also made to allow use of the stored gas as a cover gas for selected liquid holding tanks to prevent aeration of the fluids contained in these tanks. During plant operations, gaseous wastes will originate from:
- 1. Degassing the reactor coolant discharged to the Chemical and Volume Control System,
- 2. Displacement of cover gases as liquids accumulate in various tanks.
- 3. Miscellaneous equipment vents and relief valves, and
- 4. Sampling operations and automatic gas analysis for hyorogen and oxygen in cover gases.
The waste gases consist primarily of hydrogen stripped from reactor coolant which is processed by the boron recycle system as a result of boron dilution, nitrogen and hydrogen i 2.2-1
gases purged from the CVCS volume control tank when degas-sing the reactor coolant, and nitrogen from the closed gas blanketing system. Fission gases make up enly a small fraction of the total volume processed by the gaseous waste system. The gas decay tanks which provide for waste gas holdup and control have sufficient capacity to permit a minimum of 45 days for decay of waste gas before discharge. The annual noble gas activity release which the system is able to handle as a minimum is contained in Table 2.2-1 and is based on the following assumptions: For Xe-133:
- 1. The quantity of Xe-133 removed from the plant over a core cycle is determined assuming all gaseous waste is initially at peak reactor coolant activity concentration based. on 1% failed fuel, and 3250 Mwt power with daily load reduction to 15% power.
- 2. Using the same reactor coolant activity concentrations as in (1), the total annual Xe-133 removed to the Wast 9 Disposal System by degassing the Reactor Coolant System for anticipated cold shutdowns are combined. The cold shutdowns are assumed to occur at the following times:
(a) during the second week of operation, (b) at the peak xenon level and (c) during refueling.
- 3. Using the same reactor coolant activity concentrations as in (1), the total Xe-133 removed from the reactor coolant ,
J 2.2-2
' to the Waste Disposal System as a result of 4 hot shut-downs occurring at equal intervals in the core cycle is established.
1 Items 1, 2, and 3 above are summed to obtain the total Xe-133 removed to the Waste Disposal System and allowance is made for 45 days decay to obtain the total estimated annual release of Xe-133. For Kr-85:
- 1. Since there is not significant decay of Kr-85 during the operating periods involved, the total Kr-85 that enters the reactor coolant during the core cycle based on three col'd shutdowns, and four hot shutdowns, is deter-mined. It is assumed that this total eventually is released through the Waste Disposal System. The basic assumption of 1% failed fuel is retained for this determination.
1 In comparison to Kr-85 and Xe-133, there will be no signifi-cant activity release after 45 days of decay from the i remaining gaseous wastes since the half lives of the isotopes are short and/or the quantities present in the reactor coolant are small. The rigid gas control, system flexibility, and system capability, which provides long decay times allow releases of radioactive waste gases to be kept as low as practicable and in any event within 10 CFR 20 limits. Typically, PWR gas releases have been far below MPC. 2.2-3
Most of the gas received by the Waste Disposal System during normal operation is cover gas displaced from the Chemical and Volume Control System holdup tanks as they fill with liquid. Since this gas must be replaced when the tanks are emptied during processing, facilities are provided to return gas from the decay tanks to the holdup tanks. A backup supply from the plant nitrogen header is provided for makeup if return flow from the gas decay tanks is not available. To prevent hydrogen concentration from exceeding the combustible limit in the waste gas system, components discharging to the vent header system ara restricted to those containing no air or aerated liquids and the vent header itself is designed to operate at a , slight positive pressure (0.5 psig minimum to 2.0 psig maximum) to prevent in-leakage. Out-leakage from the system is minimized by using Saunders patent diaphragm valves, bellows seals, self-contained pressure regulators and soft-seated packless valves throughout the radioactive portions of the system. 2.2.2 System Design and Operation The process flow diagrams for the Waste Gas System are shown on Figures 2.2-1 and 2.2-2. Gases vented to the vent header flow to the waste gas com-pressor suction header. One of the two compressors is in continous operation with the second unit instrumented to
)
\
2.2-4
U 1 act as backup for peak load conditions or failure of the first unit. From the compressors, gas flows to one of I the six (typical) gas decay tanks. The control arrangement on the gas decay tank inlet header allows the operato- to place one tank in service and to select another tank for backup if the tank in operation becomes fully pressurized. When the tank in service becomes pressurized to 110 psig, a pressure transmitter automatically closes the inlet valve to that tank, opens the inlet valve to the backup tank and sounds an alarm tu alcrt the operator so that he may select a new backup tank. Pressure indicators are supplied to aid the operator in selecting the backup tank. Gas held in the decay tanks may either be returned to the Chemical and Volume Control System holdup tanks as cover gas, or discharged to the atmosphere if it has decayed sufficiently for release. Generally, the last tank to receive gas will be the first tank used as c)ver gas to permit the maximum decay time before releasing gas to the environment. The system is designed with sufficient flexibility to allow the operator to align separate tanks for storage, reuse and/or discharge to the environment simultaneously, without restric-ting operation of the other tanks. During degassing of the reactor coolant prior to a refueling shutdown, it may be l l 2.2-5 l
P desirable to transfer the high t.ctivity gas purged from the volume control tank into a particular tank and isolate that tank for decay rather than re-use the gas in it. This is done by aligning the controls to open the inlet valve to the desired tank and closing the outlet valve to the cover gas header. One of the other tanks can be opened to the cover gas header at this time, while still another might be discharged to atmosphere. Before a tank can be discharged to the environment, it must be sampled and analyzed to determine the activity to be released, and only then discharged to the plant vent at a controlled rate through a radiation monitor. During release a trip valve in the discharge line is closed automatically by a high activity level indication in the plant vent. 4 Dur ig operation, gas samples are taken periodically from tanks discharging to the waste gas vent header and from the gas decay tank being filled at the time. These sam-ples are automatically analyzed to determine their hydrogen and oxygen content. The hydrogen analysis is for surveil-lance purposes since the concentration range will vary considerably from tank to tank. Because only deaerated l sources input to the gaseous waste system, there should be no significant oxygen content in any_ of the tanks. An alann will warn the operator if any sample shows 2% by volume of oxygen. This allows time to take the required
)
2.2-6
action before the combustible limit is reached. Another tank is placed in service while the operator locates and eliminates the source of oxygen. 2.2.3 , Components Waste Gas Compressors Two compressors are provided for removal of gases to the gas decay tanks from all equipment that contains or can contain radioactive gases. These compressors are usually of the water-sealed centrifugal displacement type. The operation of the compressors is automatically controlled I by the gas manifold pressure. While one unit is in oper-ation, the other serves as a standby for unusually high flow or failure of the first unit. Number 2 Locati.on Auxiliary Building Operating Design Temperature 70-130 F Suction, N at 140 UF 2 psig maximum 2 Design Discharge Pressure 110 psig Design Flow (N 2 at 140 F, 2 psig) 40 cfm Gas Decay Tanks Six welded, carbon steel tanks are provided (number may i vary) to contain compressed waste gases (hydrogen, nitrogen, 1 and fission gases). After a period of radioactive decay, these gases may b( released at a controlled rate to the atmosphere through the plant vent. Ali discharges to the atmosphere are monitored. 2.2-7
Number 6 (typical) Location Auxiliary Building Design Temperature 130 F Design Pressure 150 psig Volume 600 ft 3 1 Material Carbon Steel Type Vertical Cylindrical 2.2.4 Summary The storage and release gaseous radioactive waste processing system is the most common design utilized at present-gener-ation pressurized water reactors and it is anticipated that many future facilities will also be equipped with this type system. The system collects and stores hydrogenated and nitrogenated radioactive gases in gas decay tanks. After an appropriate decay period, the gas is sampled and released through a monitored release point. Provisions are made to recycle the stored gas to the tank cover gas system which minimizes aeration of the liquids contained in the various tanks. The system is not designed to withstand a hydrogen explosion. An automatic gas analyzer continuously monitors the system and its inputs to ensure that explosive mixtures are not present.
]
2.2-8
TABLE 2.2-1 ESTIMATED ANNUAL GASEOUS RELEASE BY ISOTOPE'
- ACTIVITY RELEASED
]
TO THE ENVIRONMENT
." ISOTOPE CURIES /YE AR Y
e Kr 85 5400 Kr 85m,87,88 NEGLIGIBLE Xe 133 574(2) 1 Xe leem, 135,135m,138 NEGLIGIBLE TOTAL 5974 IIIBASED ON 1% FAILED FUEL AND 3250 Mwt LOAD FOLLOW OPERATION (2)45 DAY HOLDUP B
p i' l'
^
2.3 VOLUME REDUCTION SYSTEMS 2.3.1 Introduction There are two gaseous waste management system designs applicable to pressurized water reactors which utilize volume reduction to minimize or eliminate routine dis-charges to the environment. This volume reduction is accomplished in both systems through the use of a recom-biner system to remove the hydrogen in the waste gas stream, and by re-using the nitrogen collected in waste gas. Since hydrogen and nitrogen comprise the major portion of the volume of waste gas, this will result in a much smaller volume of gas to be stored. Although the systems provide the capability for discharge to the environ-ment after decay, it is anticipated that no scheduled releases will be necessary over the forty year life of the plant. The two systems differ in that one is designed for use with l l continuous CVCS volume control tank purge to remove fission gases, while the other is designed for facilities which employ a gas stripper in the CVCS letdown purification system. The' system employing a continuous VCT purge mixes { the hydrogen-fission gas mixture with a continually cir-culating nitrogen stream where the' hydrogen is comb'ined with oxygen to produce water and the remaining nitrogen-fission gas .nixture is stored in gas decay tanks. The 2.3-1 L.
system employing a CVCS gas stripper collects the hydro-gen-fission gas mixture in a gas decay tank and processes it, on a batch basis, through a recombiner. The resultant nitrogen-fission gas mixture is then stored in additional gas decay tanks. Both systems provide a means of re-using nitrogen for tank cover gases to minimize the volume of nitrogen present in the waste gas system. 2.3.2 System for Use with Continuous VCT Purge i 2.3.2.1 Introduction
! This gaseous waste system is designed to process the fission product gases removed from the reactor coolant sprayed into the volume control tank. The system is also designed to collect gases from the boron recycle evaporator, reactor coolant drain tank and other miscellaneous sources. The system has the capacity for long term storage which eliminates the need for scheduled discharge of any quantities of radio-active gases.
Under normal operation, the gaseous leakage from the system is approximately 100 scf yr. The associated gaseous activity is given in Table 2.3-1. This leakage is mixed with lenti-lation exhaust and diluted further in the atmosphere, resulting in site boundary doses which are a small fraction of regulations. J 4 2.3-2
Component redundancy and diversity of instrumentation and controls enables the system to operate during faults of moderate frequency in combination with fuel defects. The system is designed to preclude the posr ibility of an , internal explosion. However, the system volume is dis-tributed so that the dose in the unlikely event of an explosion is approximately the same as the dose due to gas decay tank rupture as analyzed in the Safety Analysis Report. The Gaseous Waste Processing System consists mainly of a closed loop comprised of two waste gas compressors, two catalytic hydrogen recombiners, and gas decay tanks to accumulate the fission product gases. The process flow diagram is shown on Figures 2.3-1 and 2.3-2. 2.3.2.2 System Design and Operation Prior to the system being put into operation, the Gaseous Waste Processing System is flushed free of air and filled with nitrogen. During norval power operation, nitrogen gas is continuously circulated around the process loop by 4 one of the two compressors. Fresh hydrogen gas is charged to the volume control tank where it mixes with fission gases which have been stripped from the reactor coolant 2.3-3 ,
in the tank gas space. The contaminated hydrogen gas is 4 then vented from the volume control tank into the circulating nitrogen stream of the Gaseous Was.te Processing System. The resulting mixture of nitrogen-hydrogen-fission gas is
- pumped by the compressor to the recombiner where enough oxygen is added to reduce the hydrogen to a low residual 1
level by oxidation to water vapor on a catalytic surface.
- After the water vapor is removed, the resulting gas stream is circulated to the on-line gas decay tank (s) and back to the compressor suction to complete the loop. A radiation monitor is pro ~vided in the process stream to detect any large increase in fission product activity which would
; indicate a possible fuel failure.
Each gas decay tank is capable of being isolated and the number of tanks valved into operation at any time is restric- 'l ted to limit the amount of radioactive gases which could be released as a consequence of any single failure, such as the rupture of any single tank or connected piping. By alternating use of these tanks, the accumulated activity is distributed among the tanks. When the residual fission gases and the hydrogen contained in the reactor coolant must be reduced in preparation for a cold shutdown and RCS maintenance, operation of the
; Gaseous Waste Processing System remains unchanged until the 2.3-4
~
coolant fission gas concentration is reduced to the desired levels. At that time the gas decay tanks are valved out of oparation, hydrogen addition to the volume control tank is terminated, and one of the ~wo c shutdown tanks placed in service. This tank, however, is inserted in the process loop directly at the compressor discharge so that the gas mixture returning to the compressor flows through the shut-down tank and then the recombiner before returning to the compressor suction to complete the loop. During the first plant cold shutdown, fresh nitrogen is charged to the f volume control tank where it mixes with the hydrogen com- [ ing out of solution. The gas mixture is vented to the com-pressor suction, flows through the shutdown decay tank to
'the recombiner where hydrogen is removed, and then returns to the compressor suction. During the initial unit shutdown, there is an accumulation of nitrogen in the shutdown tank i
which is accommodated by allowing the tank pressure to increase. During subsequent shutdowns, however, additional nitrogen is not required since the gas from the first shut-down is re-used. 2.3.2.3 Estimated Releases The Gaseous Waste Processing System removes fission product gases from the volume control tank and recycle evaporator i and has the capacity to contain them, eliminating the need for regularly scheduled discharge of these radioactive gases to the environment. Since the system reduces fission product 2.3-5 i
)
gas concentrations in the reactor coolant during unit operation, it significantly reduces the escape of radio-active gases arising from reactor coolant leakage. Design is based upon continuous operation with reactor coolant a system activities associated with cladding defects in fuel rods generating one percent of the rated core thermal power. , 1 1 Table 2.3-2 and 2.3-3 show the maximum and expected fission product inventory in the Gaseous Waste Processing Systems i over the forty year plant life, i Figures 2.3-3 and 2.3-4 shows that for a given power rating the quantity of fission gas activity accoulated after forty l continuous years of operation is only twice the activity , i accumulated after thirty days operation. This is because most of the accumulated activity arises from short lived isotopes reaching equilibrium in one month or less. t , The difference between the thirty day and forty year accu-i
- mulations is essentially all Krypton-85. This accumulation of Krypton-85 is not a hazard to the plant operator because
j 1. Radiation background levels in the plant are not noticeably affected by the accumulation of Krypton-85 which is a beta emitter, for which the tanks themselves provide adequate shielding. t
- 2. The system activity inventory is distributed in several
- tanks so that the maximum permissible inventory in any
] single tank is actually less than that of earlier Gaseous Waste Processing System designs. 2.3-6 )
1 M
- 3. Since this system permits fission gas removal from the j reactor coolant during nonnal operation, it is expected to reduce plant activity levels caused by leakage of reactor coolant.
i With operation of this system, it is possible to collect virtually all of the Kr-85 released to the reactor coolant and to achieve a reduction in the fission product gas inventory in the Reactor Coolant System as shown in Table 2.3-4. Provisions are also made to collect any residual ' i gases stripped out of solution by the boron recycle evaporators and gases from the reactor coolant drain tank. Although the system is designed to preclude explosions, activity released due to system rupture is approximately the same as for a gas decay tank rupture as analyzed in l Chapter 15 of the Safety Analysis Report. This is due to the fact that more than 95% of the total system volume of l fission gaser are contained in the on-line gas decay tank. Further, continued flow from the volume control tank. for i about one hour after system rupture increases the release 1 of activity by only about 1%.
- Hence, by controlling the accumulation of fission gases in l
the on-line gas decay tank, site boundary dose requirements are fully met whether complete system rupture or gas decay 1 tank rupture only is used as a base for the accident
. evaluation.
i 2.3-7
"]
- 2.3.2.4 Components
- l Waste Gas Compressors ;
Two waste gas compressor packages are provided. One unit is nomally used with the other on a standby basis. 1 1
; The units are water-sealed centrifugal displacement machines which are skid-mounted in a self contained package. Construc-i tion is primarily of carbon steel. Mechanical seals are provided to minimize the out leakage of seal water.
Type Centrifugal Quantity 2 1 j Design Suction Pressure, N2 at 1400F, psig 0.5 Design Discharge Pressure, psig 100 P j Design Flow (N 2 at 140 F), scfm 40 Recombiners Two catalytic hydrogen recombiners are provided. One of the two recombiners is normally used to remove hydrogen i from the hydrogen-nitrogen-fission gas mixture by oxidation to water vapor, which is removed by condensation. The other , i ! l recombiner is available on a standby basis. Both units
! are self-contained and designed for continuous operation.
The recombiner is located in the system where the hydrogen concentration and pressure are optimum with respect to , hydrogen removal. j Type Catalytic Quantity 2 ) l 2.3-8
Design Flow Rate, scfm 50 Design Hydrogen Recombination Rate, scfm 1.4 Design Discharge Pressure, psig 25 Design Discharge Temperature, OF 140 Material of construction Stainless Steel Gas Decay Tanks Gas decay tanks are provided to collect and store the fission gases stripped from the reactor coolant in the volume control tank. The actual number of tanks provided may vary from plant to plant. Type Vertical Quantity (Single Unit) 8 (typical) Design Pressure, psig 150 Volume (each), ft 3 600 Material of Construction Carbon Steel 2.3.3 Systens for Use with CVCS Gas Strippers (Figure 2.3-3)
- 2. 3. 3.1 Introduction This Gaseous Waste System is designed to process the fission product gases removed from the reactor coolant by the gas stripper located in the CVCS letdown line.
The system also collects gases produced by the boron recycle and waste evaporators, the VCT vent and the fuel sipping equioment. 2.3-9
The system is divided into two subsystems: The retention subsystem and the recyc1r. subsystem. The retention sub-l system is similar to the system described in Section 2.3.2 except that the hydrogen-fission gas mixture is collected in gas decay tanks and is then processed on a batch basis. The recycle system allows for total recycle of the cover gas used for the Reactor Drain Tank and the Equipment Drain Tank. As in the volume reduction system described previously, this system is designed to eliminate any scheduled discharges of radioactive gases. Sufficient storage capacity is pro-vided to allow on-site storage of all fission gases collected over the forty year design life of the facility. . Similar reactor coolant activity reduction, accumulation of fission gases and system leakage and release rates can be expected for this system as. compared to the volume reduction system designed for use with VCT gas stripping. 2.3.3.2 System Design and Operation Retention Subsystem Waste gases which are routed to the retention subsystem are mainly hydrogenated or nitrogenated radioactive gases from various sources throughout the plant. Gaseous waste's are ] generated from reactor coolant degassing, fuel sipping,
)
2.3-10
l 1 and volume control tank venting o(erations. The nitrogen gas from the fuel sipping operatid will be vented to the gas collection header. However, if a predetermined activity i d leve.1 is detected in -}he nitrogen gas (by the fuel sipping monitor), it will be'c'irected to the gas surge header. + Therefore, the refueling failed fuel detector vent is not considered a normal input collected by the gas surge header ; (GSH). Waste gases enter the retention subsystem via the gas surge header. The gas surge header collects the hydro-i genated or nitrogenated radioactive gases with negligible oxygen from the volume control tank and the gas stripper. L 2 The nitrogen effluent from the gas recombiner is also col-lected and returned to the system via the gas surge header.
! Sources, volumes, and flow rates to the gas surge header are given in Table 2.3-5. The specific activities of these 1 l sources are given in Table 2.3-6. The bases for these j specific activities are listed below.
i
- 1. All noble gases entering the gas stripper in RCS liquid are assumed to be removed with the vented gas. One j tenth percent of the iodines entering the gas stripper [
are assumed to be removed with the vented gas. The 4 I stripping rate is 84 gpm. : i , i 2. A partition factor of ten was assumed for release of t noble gases from the RCS liquid in the volume control . tank. One tenth percent of the iodines in the liquid l was assumed to be vented with the purge gas, four purges per year were assumed. I 1 2.3-11 i l
- _ _ . _ _ _ _ . _ , - - _ , _ . - , . . - . _ . _ _ . . _ _ _ _ . . ~ , . _ _ _ _ _ . _
- 3. A partition factor of ten was assumed for release of 4
noble gases from RCS liquid during fuel sipping of 1/3 of the core. One tenth percent of the iodines in the J liquid was also assumed to be removed with the vented gas. Gases from the gas surge header flow into the gas surge tank (GST) where they are collected. The gases remain in the gas surge tank until the pressure builds to a point which actuates a waste gas compressor (WGC). The waste gas compressor feeds a preselected gas decay tank (GOT) until the pressure in the Jas sLrge tank drops to a point where the waste gas compressor stops. A second waste gas compressor starts automatically if the pressure in the gas surge tank increases due to a surge of the inputs. Tnis automatic operation of the waste gas compressors will continue until the on-line gas decay tank is observed to approach its upper operating pressure. At this point another gas decay tank will be manually aligned to receive the waste gas compressor's discharge. The just filled tank is analyzed by the gas analyzer for hydrogen and oxygen content and grab samples can be taken for a radio-activity analysis. After a gas decay tank has been sampled and analyzed, it is aligned to the gas recombiner system for processing on a batch basis. The gas flows through a regulator valve into the gas recombiner system. Processing is essentially a controlled reaction between hydrogen and oxygen to 2.3-12
I j produce water. The influent hydrogen gas is diluted with 4 1 j nitrogen to maintain a 3-6 percent hydrogen mixture. This ' 1 J mixture is then pre-heated and o;;ygen is added to produce
- a stoichiometric mixture of hydrogen and oxygen. The
~
I addition of oxygen is controlled by analysis of either the
- influent or effluent hydrogen content. The entire gas j stream is then passed over a catalyst bed. Recombination [
j of hydrogen and oxygen occurs on the surface of the catalyst. i The effluent gas stream is a mixture of nitrogen, steam : j and noble gases. The steam is condensed and separated- l i
- out as water. The remaining gas is comprised essentially of nitrogen and noble gases. The gas recodiner will con-l tinue processing until the selected gas decay tank pressure i is reduced to a predetemined value. [
i i - The gas recodiner system effluent is returned to the gas j surge header where it re-enters the system through the gas l surge tank and waste gas compressors. ' l The gas decay tank which is currently aligned to the waste j gas compressors will collect the normal influents plus the I hydrogen free gas recombiner effluent. When this gas decay , j tank is filled, the process is repeated. By operating in
- this manner gases can be stored for the projected 40 year
- life of the plant. Provisions are available to release -
1 i waste gases via the gas discharge header and radiation i i i i
~onitor. The capability is also provided to place the ,
3
- recombiner effluent fission gases in shipping cylinders for i disposal or storage. i l 2.3-13 l l r
1 I The retention subsystem is primarily a manually controlled { system. When a gas decay tank is to be filled or processed by the recombiner, the valves controlling this operation l must be operated manually. These valves are provided with i J extension stems to allow operation to take place from i inside a valve gallery or from behind a shield wall. This ) feature will minimize radiation exposure to operating per-sonnel. The operation of the waste gas compressors is automatically controlled by gas surge tank pressure instru-mentation. l , ) i ?
) Recycle Subsystem i i t The gases which are routed to the recycle subsystem are the l' ; nitrogen cover gases in the equipment drain tank (EDT) and t
the reactor drain tank (ROT). The recycle subsystem pro- , vides a means to collect, store and return nitrogen gas ! l " {, from these two tanks. ' s These tanks contain an initial nitrogen cover at a preset I positive pressure. When liquid leakage enters, either or i both tanks it will raise the pressure of the cover gas, i When the pressure reaches a specified upper limit, the f t ] recyle compressor is actuated. The compressor discharges i t , to the nitrogen recycle tank until the pressure in the ( equipment drain tank and reactor drain tank reduces to the ! nonnal operating pressure. Conversely when itquid is J j I E i i i 2.3-14 i i :' i
l l removed from either or both tanks, the cover gas pressure will drop to a lower limit. A pressure regulating valve then opens allowing nitrogen to flow into either or both tanks from the nitrogen recycle tank. The nitrogen recycle tank is periodically sampled by the gas analyzer. In the event of hydrogen or oxygen intrusion into the cover gas, ! the nitrogen recycle tank can be manually aligned with the j gas recombiner system. The nitrogen recycle tank gas flows through a regulator valve into the gas recombiner system. The effluent (essentially nitrogen) from the gas recombiner system is returned by the gas recombiner compressor into the nitrogen recycle tank, l i 2.3.3.3 Components Gas Surge Tank
, Seismic Category I Quantity 1 Type Vertical, Cylindrical Internal Volume 20 ft3 l Design Pressure 40 psig Design Temperature 200 F Normal Operating Pressure 1.5 to 3.0 psig Normal Operating Temperature 120'F l Material Carbon Steel l
2.3-15 B
Gas Decay Tank Seismic Category I Quanti ty 9 Type Vertical, Cylindrical Internal Volume 700 ft 3 Design Pressure 380 psig Design Temperature 200*F Nomal Operating Pressure O to 345 psig Normal Operating Temperature 110'F Material Carbon Steel Nitrogen Recycle Tank Seismic Category I Quanti ty 1 Type Vertical, Cylindrical Internal Volume 700 ft Design Pressure 380 psig Design Temperature 200'F Nomal Operating Pressure O to 180 psig Material Carbon Steel Waste Gas Compressor and Cooler Seismic Category I Quantity 2 Type Positive Displacement Design Inlet Pressure 40 psig F 2.3-16
4 I . Design Inlet Temperature 200'F i
- Design Discharge Pressure 380 psig Maximum Operating Discharge 345 psig
- Pressure !
Maximum Aftercooler Outiet 110'F Temperature 2 Rated Flow (Normal Operation) 2 SCFM i EDT/RDT Recycle Compressor l Quanti ty 1 1 Type Positive Displacement Design Inlet Pressure 50 psig Design Inlet Temperature 150'F Design Discharge Pressure 380 psig
! Normal Operating Discharge 3-180 psig ! Pressure 1 , Maximum Aftercooler Outlet 110 F Temperature i Flow 20 scfm i Cooling Water Requirements 2 gpm l l Seismic Category I l
i I 2.3-17
Gas Recombiner Compressor Quanti ty 1 Type Positive Displacement Design Inlet Pressure 100 psig ! I Design Inlet Temperature 120"F Design Discharge Pressure 380 psig Normal Operating Discharge 3-180 psig Pressure Maximum Aftercooler Outlet 110'F Temperature Flow 2 scfm Cooling Water Requirements 1 gpm Seismic Category 1 Gas Recombiner Package 1 Quantity 1 Type Catalyti: Design Inlet Pressure 180 psig f Design Inlet Temperature 150 F Design Discharge Pressure 3-15 psig Normal Inlet Pressure 10-25 psig i t Normal Discharge Pressure 3-15 psig Flow 2 scfm Design Discharge Temperature 120 F
! Normal Discharge Temperature 100 F Cooling Water 1 gpm 2.3-18 ,
i i
- 2.3.4 Sumary l l l
]i There are two typical gaseous radwaste system designs i utilizing volume reduction to reduce or eliminate the . . I
! need for discharges to the environment. Both systems I 1 .-.
! incorporate hydrogen recombiners and the capability of nitrogen recycle to minimize the volume of waste gas > 3 which must be stored or released.
- The first system design (Section 2.3-2) is a recombiner i .
I system designed for operation with continuous purging l from the volume control tank. During operation hydrogen j purged through the volume control tank carries the radio- ) active gases to the closed loop gas processing facility. ! During shutdown, the purge may be continued to reduce ! l the primary coolant activity concentrations. Prior to a a
! plant shutdown a nitrogen purge may be used to remove hydrogen from the coolant. The system includes circulation j of nitrogen in a closed loop totassure dilution of the I
I J mixture of hydrogen and added oxygen to less than flammable i concentrations, removal of hydrogen to reduce the flamma-i ! bility hazard and to reduce the waste gas volume by recom- ! a bination with oxygen,and compressed gas storage for decay , ! of the radioactive gases. ! The second system design is a recombiner- volume reduction ; i ! system for use with a reactor system employing a CVCS , ! Letdown Stripper. This system nomally is operated batch-i l wise. Gas decay tanks are successively filled and the i 2.3-19 i I I
~
contents are individually processed through the recombiner (internal nitrogen recirculation type) and the effluent is routed to the gas surge tank. This syste.r. also inclodes compressors and recombiners to reduce the volume storage requirements and reduce the flammability hazard. During batch operation, gas from a gas decay tank is con-tinuously analyzed for hydrogen and oxygen content, and the gas is then processed through a recombiner system. The influent to the recombiner is diluted with internally recirculating nitrogen to maintain appropriate hydrogen concentrations. The mixture may be pre-heated and oxygen 1 added to produce a stoichiometric mixture. The gas stream is then passed over a catalyst bed for combination of the hydrogen and oxygen. The water vapor is condensed and separated out leaving an effluent composed pr~ narily of nitrogen and noble gases which is returned to the gas surge tank. This operation continues until the decay tank supplying the feed gas reaches a predetermined low pressure. The recombiner effluent is sent to another gas decay tank via the gas surge tank and compressors.
)
l 2.3-20 i
TABLE 2.3-1 . GASEOUS ACTIVITY RELEASED TO VENT STACK FROM LEAKS IN THE GASEOUS WASTE PROCESSING SYSTEM ACTIVITY DISCHARGED ISOTOPE CURIES / YEAR Kr-85m 2.4 Kr-85 170 Kr-81 0.2 Kr-88 2.4 - Xe-133m 9 Xe-133 1 50
.] $ Xe-135 21 1-131 3.45 x 10-3 1-132 11.1 x 10-4 l-133 4.75 x 10-3 1-134 2.3 x 10-5 l-135 8.23 x 10-4 TRITIUM NEGLIGIBLE
, THIS IS BASED ON OPER ATION OF ONE UNIT WITH CLADDING DEFECTS IN FUEL HODS GENERATING 0.2 PERCENT OF THE RATED CORE THERMAL q l POWER AND A LEAKAGE RATE OF ONE HUNDRED STANDARD CUBIC FEET PER YEAR FROM THE GASEOUS WASTE PROCESSING SYSTEM. THE STRIPPING EFFICIENCY USED FOR NOBLE GASES IN THE VOLUME TANK IS 100%. SEPARATION FACTOR FOR IODINES IN THE VOLUME CONTROL TANK IS 100. j
1 I i TABLE 2.3-2 4
- ACCUMULATED RADIOACTIVITY PER UNIT IN THE GASEOUS WASTE i PROCESSING SYSTEM AFTER FORTY YEARS OPERATION i
ACTIVITY (CURIES) FOLLOWING PLANT SHUTDOWN ! y ISOTOPE ZERO DECAY 30 DAYS 50 DAYS'
~
[ Kr-85 61,270 60,950 60,730 i w i ALL OTHER NOBLE GASES a Kr-85 37 ~0 ~0 Kr-87 3.6 ~0 ~0 . Kr 88 36 ~0 ~0 ) Xe-131: 514 91 29 I ) Xe-133 65,650 1270 92 Xe-133m 568 0.06 ~0 Xe-135 297 0 ~0 j Xe-135m 0.14 ~0 ~0 Xe-138 0.16 ~0 ~3 l THE TABLE IS BASED ON FORTY YEARS CONTINUOUS OPERATION WITH 1% FUEL DEFECT. POWER ASSUMED TO BE 3565 Mwt. ; t I
~ _- ._
TABLE 2.3-3 EXPECTED ACCUMULATED RADIOACTIVITY PER UNIT IN THE GASEOUS WASTE PROCESSING SYSTEM AFTER FORTY YEARS OPERATION Activity (Curies) Following Plant Shutdown Isotope Zero Decay 30 Days 50 Days Kr-85 7480 7444 7407 All other noble gases Kr-85m 1.31 s0 .0 Kr-87 0.096 s0 s0 Kr-88 1.19 s0 s0 Xe-131m 32.0 98 31 Xe-133 3650 70.5 5.1 Xe-133m 29.3 s0 so l Xe-135 10.4 so s0 Xe-135m 0.001 so so Xe-138 0.004 so so The table is based on forty years continuous operation. Power assumed to be 3565 MWt. The data are based on a volume control tank purge rate of 0.7 scfm, a 40 percent stripping efficiency. , 2.3-25
TABLE 2.3-4 REDUCTION IN REACTOR COOLANT SYSTEM GASEOUS FISSION PRODUCTS PER UNIT RESULTING FROM NORMAL OPERATION OF THE GASEOUS WASTE PROCESSING SYSTEM
- REACTOR COOLANT GASEOUS FISSION PRODUCT ACTIVITIES - pc/gm ISOTOPE GWPS OPER ATIONS GWPS NOT OPER ATING Kr-85 0.15 8.8 i
Y co Kr 85m 2.1 2.1 Kr-87 1.2 1.2 I Kr 88 3.7 3.7 Xe-131m 0.3 1.9 Xe-133 83 280 Xe-133m 1.6 3.1 Xe 135 5.9 6.3 Xe-135m 0.7 0.7 Xe-138 0.7 0.7 !
- BASED ON OPERATING Wi1H CLADDING EFFECTS IN FUEL GENERATING l 1 PERCENT OF THE RATED CORE THERMAL POWER. POWER IS 3565 Mwt.
PURIFICATION LETDOWN RATE IS 75 spm. I
l l l TABLE 2.3-5 SOURCES, VOLUMES AND FLOW RATES TO THE GAS SURGE IIEADER(I} Annual Average Source Gas Type Annual Volume Flow Rate Basis (SCF) ($FMT Volume Control Tank Nitrogen 1624 4.0 (-3)* Four tank volumes purged per year l l Gas Stripper Hydrogen 145,000 2.3 (-1) Thirty cc/kg of ga; i stripped at 84 gpm ro continuous letdown i, flow. hl Refueling failed Nitrogen 1670** 4.0 (-3)** Based on sipping 1/3 Fuel Detector of the core during refueling. (1) See Table 11.3-3 for specific activities of sources.
* ( ) denotes power of ten.
** Values are presented for GDT sizing considerations only.
- __~
TABLE 2.3-6 SPECIFIC. ACTIVITIES OF SOURCES TO THE GSH DURING NORMAL OPERATION (uCi/cc 0 STP) Volume Gas Refueling Failed ** Nuclide Control Tank Stripper Fuel Detector Kr-35m 2.8 (-4)* 1.1 (1) 2.9 (-1) Kr-85 4.7 (-6) 1.7 (-1) 4.9 (-3) Kr-87 1.7 (-4) 7.0 1.8 (-1) Kr-88 5.0 (-5) 1.9 (1) 5.3 (-1) Xe-131m 1.6 (-6) 6.3 (-1) 1.7 (-2) Xe-133 5.3 (-3) 2.1 (+2) 5.6 Xe-135 1.0 (-3) 4.2 (+1) 1.1 Xe-138 1.2 (-4) 5.0 1.3 (-1) I-129 7.7 (-14) 2.7 (-12) 7.7 (-9) 1-131 6.7 (-6) 2.3(-4) 6.7 (-4) 1-132 1.9 (-6) 6.3 (-5) 1.9 (-4) f lit 9.9 (-6) 3.5(-4) 9.9 (-4) 1-134 1.3 (-6) 4.5 (-5) 1.3 (-4) 1-135 5.5 (-6) 1.9 (-4) 5.5(-4)
* ( ) denotes power of ten. ** Values are presented for GDT design considerations only.
I 2.3-31
PWR STORAGE AND RELEASE - NITROGEN RECYCLE
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c . N m TYPICAL PWR GASEOUS WASTE PROCESSING SYSTEM (STORAGE AND RELEASE WITH N2 RECYCLE AND CVCS GAS STRIPPING)
1 140 TOTAL - ALL GASE0US ISOTOPES 130 120 7
$ 100 -
E u 3 2 g 80 o x
- 68.3 5
ALL GASES EXCEPT KR-85 L' o 60 - E a" E . d g 40 1 20 I I I O 10 20 30 40 TIME (YEARS) Waste Gas Processing System Fission Gas Accumulation Based on Continuous Core Operation at 3565 MWt with 1% Fuel Defects, Stripping Efficiency 100% Figure 2.3-4 2.3-39
I 12 TOTAL - ALL GASEOUS ISOTOPES 10 G W E v S 8 R E S 5 E 6 - E ' o 5 s e C 4 -
- ALL GASES EXCEPT KR-85 5
2 0 0 10 20 30 40 TIME (YEARS) Estimated Waste Gas Processing System Fission Gas Accumulation Based on Full Power Operation of 3565 MWt. Figure 2.3-5 2.3-41 i I
w-2.4 CHARC0AL ADSORBER SYSTEMS 2.4.1 Introduction Another type of gaseous waste management system which may be interfaced with pressurized water reactors is the charcoal adsorber design. This design accomplishes holdup for decay of fission gases by adsorption on a charcoal bed with subsequent release to the environment. There are several variations of this system with the major difference between designs being the temperature maintained in the adsorber beds. Refrigerated or even cryogenic (Figure 2,4-3) systems are available to increase the efficiency and holdup times of the adsorber. The charcoal systems are not often used on present-generation pressurized water reactors and will probably comprise only a small percentage of the systems supplied with future facilities. 2.4.2 System Design and Operation (Figures 2.4-1 and 2.4-2) 2.4.2.1 Ambient Charcoal i
,, Prior to operation, the process gas portion of the system is flushed free of air and filled with nitrogen.
The gas from the CVCS letdown gas stripper (primarily hydro-gen with small amounts of nitrogen, xenon, krypton and
+ iodine) and hydrogenated gases from other sources feeding a
t' 2.4-1
't
. 1
t the waste gas header are collected in the waste gas surge i tank. These gases are dehumidified (dew point 35 F) in
; the process gas refrigerant dryers, and passed through and filtered by the ambient temperature charcoal adsorbers.
The hydrogen or nitrogen in the gas stream will pass , through the adsorber beds while the xenon, krypton and any iodine present will be adsorbed. The charcoal-beds are designed to delay xenon isotopes for a minimum of 60 5 days and Krypton isotopes for a minimum of 4 days. In addition an iodine decontamination factor of 100 is obtained during passage through the charcoal beds.
, After passing through the adsorber beds, the hydrogen, ,
nitrogen and delayed fission products are processed through the waste gas compressors to the waste gas receiver tank
]
l from which they are discharged to the environment through a radiation monitor. Provisions are also included for the recycle of the purified hydrogen stream to the volume control tank. The process gas receiver located after the gas compressors ensures an adequate supply of hydrogen to the VCT under changing operating conditions while in the recycle mode. The normal mode of operations, however, is to discharge this - gas stream to the plant vent thus limiting the buildup of long-lived. fission products (mainly Krypton 85) in the reactor coolant system. i I 2.4-2
-,_ , ~ , . - . _ . , . _ _ _ _ . _ -, _ _ , _ , , _ _ - _ - - .
2.4.2.2 Cryogenic Charcoal (Figure 2.4-3) Removal of essentic.lly all fission product gases from the off gas stream can be accomplished by a cryogenic charcoal system. Gases are removed by adsorption onto a charcoal bed which is maintained at approximately -275 F. Fission product gases, removed from the primary coolant, are tem-porarily stored in a surge tank to allow batch operation of the system. The process stream is regulated to about 10 psig and filtered to remove particulates. Trace quantities of oxygen and ozone are removed in the recombiner by the reaction of those gases with excess hydrogen. Removal of these gases prevents ozone from building up in the cryogenic portion of the system. The gas is then cooled and passed through a moisture separator and desiccant dryer to prevent plugging of the charcoal bed by ice crystals . An aftercooler is used to remove the heat of adsorption from the system prior to entry into a regener-ative heat exchanger which lowers the temperature to -275UF. Effluent gas from the charcoal bed, which is essentially free of noble gases, is cooled in a liquid nitrogen bath to approximately -300 F to achieve the necessary refriger-ation effect prior to return through the regenerative heat exchanger. The gas is warmed to ambient temperatures by the regenerative heat exchanger, monitored for radioactivity, and released to the environment. 2.4-3
l When the charcoal adsorber bed reaches saturction as indi-i cated by break through of krypton detected by radiation monitors, the bed is isolated from the system and regen-eration is initiated. Using a controlled heat source, the noble gases and any coadsorbed carrier gas are desorbed and transferred to storage bottles by a diaphragm com-pressor. An accumulator tank is used to temporarily store the gases in the event that the desorption rate exceeds the capacity of the compressor. When the desorption is l complete, as indicated by a bed temperature of 300 F, nitrogen gas is used to purge the system. The nitrogen gas is stored in the bottles with the noble gases. At this time the desiccant dryer is also regenerated by heating and purging with clean nitrogen. A cryogenic charcoal system for a PWR plant is extremely [ compact. The entire system, except for the surge tank, ) i control panel, and liquid nitrogen storage tank, is con-l tained on a skid 9' x 7' x 10' high. 2.4.3 Summary The adsorption type waste gas processing system may process feed froin a CVCS letdown stripper, volume control tank
- purge, or vents from deaerated tanks. It dehumidifies i
the gas stream (mainly hydrogen) prior to passage through charcoal adsorption tanks for holdup and decay. The gas l I 2.4-4 i J m ..%. - _ , . - --
._.--_r... .,-.3.__....--.~,-,.n,. ,.4.-- - -. . - e. ..._, .
1 4 4t l is normally released to the environment with holdup times (
- for xenon quite similar to the storage and release gas I
decay tank design. The holdup time for krypton is only i about four days but there is insignificant decay of krypton
~
i even during a sixty day holdup time. 1 l Cryogenic systems could be utilized to separate the : i ! fission product gases from the hydrogen stream and then I transfer,them to storage bottles for decay but it is not t anticipatt:d that this system will be much used. t
)
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2.5 MISCELLANE0US GASE0US WASTE SYSTEM FUNCTIONS (FIGURE 2.5-1) 2.5.1 Introduction Several systems which are not involved in processing the . ! hydrogen-nitrogen-fission gas mixture produced by degassing j the reactor coolant or other hydrogenated fission gas sources are included as part of the Gaseous Radwaste Management System. These include: The plant nitrogen system which provides nitrogen gas for purging gaseous radwaste system components and for cover gas for various liquid radwaste and boron recovery system tanks. The plant hydrogen system which provides a source of hydro-i ! gen for the VCT gas space to maintain sufficient RCS hydro-I gen inventory. The Hydraulic Baler Package vent which filters any particulates i I produced by the baling of solid wute. L i The ccadenser off-gas system which will (on some plants) route condenser off-gas through a HEPA and charcoal filter ! if a high radiation level is sensed. The gas analyzer package which monitors various components
- in the waste gas system for potentially hazardous mixtures I
of hydrogen and oxygen. t 1 2.5-1
2.5.2 Plant Nitrogen Systems (Figure 2.5-1) As an auxiliary function, the Waste Disposal System supplies nitrogen from standard cylinders to primary plant components. The nitrogen is utilized mainly as a cover gas for tanks in the liquid waste disposal system and as a purge source for the gaseous waste system prior to initial operation or after maintenance. Two headers are provided, one for normal supply and one for backup. The pressure regulator in the operating header is set for 100 psig discharge and that in the backup header for 90 psig discharge. When the operating header is exhausted, its discharge pressure will fall below 100 psig and an alarm will alert the operator. The second header will come into service automatically at 90 psig to ensure a continuous supply of gas. After the exhausted
~
header has been recharged, the operator manually sets the operating pressure back to 100 psig and backup pressure at 90 psig. This operation is similar for both the nitrogen supply and the hydrogen supply.
~
2.5.3 Plant Hydrogen System (Figure 2.5-1) The plant hydrogen system provides hydrogen from standard cylinders to the Volume Control Tank to maintain the hydro-gen concentration in the reactor coolant for oxygen con-trol. An alarm is provided.to alert the operator when the tank contents reach a preselected minimum value.
]'
2.5-2
?
2.5.4 Air E'ector Off-Gas System (Figure 2.5-1) The Air. Ejector Off-Gas System monitors the off-gas from i the main condenser and upon senting a high radiation aligns l the off-gas through a HEPA filter-charcoal adsorber assem-l bly. A high radiation in the condenser off-gas would be caused by a steam generator tube leak which would allow 4 f ! reactor coolant flow into the steam generator. This contaminated coolant would then be converted to steam in the steam generator and fission product gases and iodine would be carried with the steam to the condenser where the non-condensable fission products would be removed by the air removal system. ; ) i Note that not all plants are equipped with. the capability to filter condenser off-gas although all are provided with a condenser off-gas monitor. 2.5.5 Hydraulic Baler Package Vent System This system is actually a part of the plant baler package or ventilation system but is included here for the sake of convenience. i ! Whenever the hydraulic baler package is in use, any parti-culates dislodged by the compression process are collected in a filter assembly to preclude their release to the
- envi ronment. The fan maintains a slight negative pressure j t on the baler to ensure all particulates are trapped by i the filter.
i 2.5-3 i
, c - - . , , _ _ _ _ _ . _ - . _ _ , _ , , _ _ _ , , . _ . . - , . . . , _ _ _ . , , , - . , . _ _ , . . . . _ . _ _ _ , , _ , , , _ _
__,I
I 2.5.6 Gas Analyzer (Figure 2.5-2) An automatic gas analyzer is provided to monitor the con-centrations of oxygen and hydrogen in the various systems I i and tanks where an explosive mixture of oxygen and hydro- ; gen might occur. Upon indication of a high oxygen level, ( provisions are made to purge the equipment to the gaseous waste. system with an inert gas. l t The oxygen analyzer utilizes the fact that oxygen is para- ' I magnetic. This type of detector was chosen rather than a [ thermal conductivity detector because of the varying con-l centrations of hydrogen in the waste disposal system. The , detector for the oxygen analyzer is unaffected by any other, , ] a property of the gas stream except magnetic susceptibility. ' i The hydrogen analyzer utilizes a thermal conductivity tech- l 1 l nique for determining hydrogen concentration. Hydrogen concentration is detemined primarily for information only I l as the concentration of hydrogen will vary considerably , dependent upon plant operating conditions. l 3 A schematic flow diagram of the waste gas analyzer is shown ! on Figure 2.5-3. I : , Sample Points:
, Boric Acid Evaporator Gas Decay Tanks Spent Resin Storage Tank f I
t
- 2.5-4 !
Reactor Coolant Drain Tank Boron Recycle Holdup Tanks Pressurizer Relief Tank Volume Control Tank , If oxygen concentration exceeds a preselected limit (s2%), an alarm is initiated to alert the plant operators. This will provide adequate time to locate and isolate the air in-leakage before an explosive concentration is reached. The gas analyzer normally sequences through each of the listed points but may be selected to skip any combination of points or to continuously sample a point of interest. Provisions are made to periodically calibrate the analyzer using a known zero gas and a known span gas assuring accurate analysis of the locations of interest. i } l 2.5-5
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1 1 PRESSURIZED !!ATER REACTOR RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM CHAPTER 3.0 SOLID WASTE PROCESSING SYSTEf1 e L
TABLE OF CONTENTS Section Title Page 3.0 SOLID WASTE PROCESSING SYSTEM . . . ..... 3.1-1
3.1 INTRODUCTION
. . . . . . ........... 3.1-1 4 3.2 GENERAL DESCRIPTION . . . .......... 3.2-1 3.2.1 Wet Wastes. . . . . . . . . . . . . . . . . . 3.2-1 3.2.2 Col l ec ti on. . . . . . . . . . . . . . . . . . 3.2-2 3.2.3 Pretreatment (Volume Reduction or Dewatering) 3.2-2 3.2.4 Mixing / Packaging. . . . . . . . . . . . . . . 3.2-3 3.2.5 Asphalt Extruder / Evaporator System. ..... 3.2-7 3.2.6 D ry Wa s t e s . . . . . . . . . . . . . . . . . . 3.2-8 3.3 PRETREATMENT SUBSYSTEMS . . . . . . . . . . . 3.3-1 3.3.1 Decanti ng Tanks . . . . . . . . . . . . . . . 3.3-1 3.3.2 Re ve rs e Os mos i s . . . . . . . . . . . . . . . 3.3-1 3.3.3 Pre-Coa t Fi l ters . . . . . . . . . . . . . . . 3.3-3 3.3.4 Centri fuge Separation . . . . . . . . . . . . 3.3-3 3.3.5 Forced Circulation Evaporators. . . . . . . . 3.3-4 3.3.6 Thin Film Evaporators . . . . . . . . . . . . 3.3-5 3.3.7 Calciner Systems. . . . . . . . . . . . . . . 3.3-6 3.4 SOLIDIFICATION AGENT SYSTEMS. . . . . . . . . 3.4-1 i 3.4.1 Ceme n t Sys ten.s . . . . . . . . . . . . . . . . 3.4-1 3.4.2 Urea Formaldehyde Systems . . . . . . . . . . 3.4-2 3.4.3 Asphalt Extrusion-Evaporatcr System . . . . . 3.4-4 3.4.4 D ry Wa s te Compac to r . . . . . . . . . . . . . 3.4-7 3.4.5 Filter Trans fer System. . . . . . . . . . . . 3.4-9 3.5
SUMMARY
. .................. 3.5-1 1
LIST OF FIGURES Ficures Title
- Page 3.1 Solid Wastes Processing Flow Diagram. . . . . 3.5-3 3.2 Decanting Tank System . . . . . . . . . . . . 3.5-5 3.3 P rec oa t Fi l te r. . . . . . . . . . . . . . . . 3.5-7 3.4 Forced Circulation Evaporator . . . . . . . . 3.5-9 3.5 Typical Wiped-Film Evaporator . . . . . . . . 3.5-11 3.6 Fluidized Bed Calciner. . . . . . . . . . . . 3.5-13 3.7 In-Line Mixing Portland Cement System . . . . 3.5-15 e
iii
l l
-3.0 SOLID WASTE PROCESSING SYSTEMS
3.1 INTRODUCTION
The most rapidly changing area in reactor radwaste manage-ment today is probably that of solid waste management. New solidification processes and agents are being used and much of the emphasis is being placed on volume reduction. In-creasing concern for volume reduction is a result of the lack of adequate space in commercial shallow land burial grounds, and the spiraling cost of shipping and disposal. A recent study projects that the existing comercial burial i grounds will be filled to capacity by 1988 and burial faci-lities will experience difficulty handling the large number [ of waste shipments they will be receiving by the year 1980. This study predicts annual volumes of LWR solid wastes that ' are well in excess of previous estimates and suggests that the capacities of most plant solid waste systems are under-designed by approximately a factor of two. It is this realization on the part of the facility managements that has. caused the recent rush to back-fit solid waste handling systems. The materials destined for the Solid Waste Systems are grouped into two broad categories: Wet wastes which are the subjects of dewatering or volume reduction processes, and. dry wastes whose volumes are reduced primsrily by' baling or compaction. See Figure 3'.l. t 3.1-1 i . . ~ . . _ _ - __- _ _ , ,. , ,-.. _ .. _ ,- .-_ .._ ,- _. _ , _ .- . ,_-.. . _ . . . _ , - . . . . . _ _ . _ _ _ _ _ _ _ _ .
. ~ ._ - . - - - . . - . . .-- . - . - . - . . - .. ..
j The wet wastes, which consist primarily of spent resins,
; filter sludges and evaporator concentrates are generally in the form of slurries containing as much as 85% water.
Ob'viously, if this water can be removed and either re-cycled or re-used in the plant, the demands on the Solid Waste Disposal Systems is greatly reduced. Several methods and options are available to the facilities and will be r
- discussed later in this chapter.
The table below illustrates the need for volume reduction: 55 gal. drums /yr.
- 1. Drum packaging with cement, no dewatering or volume reduction = 4500
- 2. Drum packaging with cenent and
, volume reduction = 500
- 3. Drum packaging with Urea formaldehyde polymer, no dewatering or Vol. Red. = 2500
- 4. Drum packaging with Urea formaldehyde polymer and volume reduction = 200
- 5. Drum packaging with Asphalt extrusion /
evaporator process = 280 l This chapter will detail the most common techniques an'd j processes in use or planned for solid waste handling systems l-at light water reactors. It should be understood that 1 these are strictly individual facility options and various ; I canbinations may be encountered by the inspectors. In addi- ; tion, some volume reduction systems may not be described here because of manual space restrictions. E i l 3.1-2
l l 1- 3.2 GENERAL DESCRIPTION j 4 3.2.1 Wet Wastes (Fioure 3.1) The term " wet wastes" designates contaminated wastes having j sufficient water content to allow them to be pumped to ' 4 collecting tanks to be further processed. Since all commercial I burial sites prohibit the burial of liquids, these wet wastes must be solidified or immobilized in a freestanding i solid form. i
~
} Wet solid wastes are classified into four basic types:
- spent resins, filter sludges, evaporator concentrates and
- - miscellaneous liquids. Spent resins result from liquid
- radwaste demineralization, reactor coolant cleanup, BWR R 1 condensate p'olishing and fuel storage pool coolant cleanup.
Filter sludges consist of spent filter aid material from precoat filters and powdered resin from precoat filter / demineralizers. Concentrated salt solutions (evaporator concentrates or evaporator bottoms) result from the treat-ment of liquid radwaste by evaporators. Evaporator con- ' i centrates consist primarily of concentrated borie acid i
- solutions from reactor coolant adjustment in PWRs and !
J > ! concentrated sodium sulfate solutions from the regeneration { of condensate polishers in BWRs. These concentrates may also contain other salts and chemicals from the simultaneous ;
't
- evaporation of floor drain, laboratory and decontamination waste.
i ! 3.2-1 i >
l l l Although the immobilization of wet solid wastes is primarily concerned with the incorporation of the waste with a soli-dification agent, there are a number of other discrete : I
., operations or subsystems involved in the treatment of these wastes that may affect the imobilized waste product. The imobilization process may be broken down into five basic operations: waste collection, waste pretreatment, solidi-fication agent handling, mixing / packaging, and waste package
- handling. The properties of the waste forms that are t j
] ultimately shipped from the reactor site are primarily in- ) fluenced by the methods utilized during the waste collec-tion, waste pretreatment and mixing / packaging operations. 3.2.2 Collection , Demineralizer resins and filter sludges are routinely
- collected in separate tanks which facilitates further a
- treatment as it can be directed towards a specific waste. ,
Liquid wastes, including regenerant solutions, boric acid i j solutions, floor drain waste, laboratory waste and decon- ] tamination solutions may be segregat'd virtually coinpletely i or not at all during waste collection. Increasing the j degree of waste segregation generally facilitates subsequent I imobilization treatment. + i 3.2.3 Pretreatment (Volume Reduction or Dewatering)
- Pretreatment is primarily directed towards reducing waste
- volume which results in decreased transportation and burial 1
l / [ ( 3.2-2 ,
costs for imobilized wastes. Wet solid wastes may or may ' not receive pretreatment prior to the mixing / packaging opera-tion. Spent bead resins wastes and filter sludges are nor-mally dewatered. This may be accomplished in the shipping container for bead resins which are the easiest to dewater or may rely on decant tanks, centrifugation or dewatering filters for bead resins and filter sludges. Evaporator bottoms may be reduced by any of several filtering systems including reverse osmosis, pre-coated cake filters, etc. They may also be re-evaporated in special high efficiency evaporators. Another method is called calcination that uses intense heat to flash evaporate the water off, leaving a dry powder which is collected and then mixed with the imobilizing agent. . 3.2.4 Mixing / Packaging The mixing / packaging (solidification) operation is pernaps the most important stage of the imobilization process. In this operation, the wet solid wastes are incorporated with a solidification agent to form a monolithic free-standing solid. The basic solidification agent types are:
- 1) absorbants
- 2) hydraulic cement
- 3) urea-formaldehyde
- 4) bitumen (asphalt)
- 5) other polymer systems 1
3.2-3 i
I h Absorbants such as diatomaceous earth and vermiculite when mixed with liquid wastes will hold free liquids physically
- within its porosity. Absorbants do not react chemically with the incorporated liquids and are not strictly solidi-fication agents since they do not produce monolithic solids.
When absorbants are used, they are normally placed into
- the shipping container after which the liquid waste is i added. Some burial sites will not accept liquid wastes im-
)
![ immobilized in this manner. Compared with other imobili-zation agents, the absorbant product is highly leachable , , and dispersible. Caremust be taken to avoid oversaturation i ; of the absorbant.
J
- j. Solidification of wet solid wastes using hydraulic cement, ;
i l either with or without additives, has been practiced for many years. Water in the waste reacts chemically with the ] cement to form hydrated silicate and aluminate compounds.
- Solids in the waste act as an aggregate to form concrete, i
1 although the types of solids encountered in wet solid wastes l . may produce a concrete of low strength. The optimum pro-portions of waste and cement and the type of cement chosen will vary witn the waste type and its composition. Cement requires a minimum amount of water to obtain workability. ' This minimum watar to cement ratio is approximately 0 5 by weight for Portland ' Cement, but will also depend on the t waste itself as some waste solids may absorb large amounts l 4 , 3.2-4 l
\ '
l
of water. The addition of too much water may result in a layer of free standing water on the surface of solidified product. Optimum formulations should consider each waste individually because of possible interactions between the waste constituents and the cement. One such interaction is the effect on the setting of the cement matrix. Accel-eration of the set can result in cement hardening in the processing equipment. Boric acid and borate salts retard setting in Portland Cement; if sufficient quantities are added the set may be retarded such that the cement never hardens. Modifications of the cement system include cement-sodium silicate and cement-clay mixtures as the solidification agent. Concentrated sodium silicate added at approximately ten percent of the waste volume is used to improve the setting properties of the cement and its volumetric effi-ciency. The addition of clays to cement 15 used to improve the leachability of cement products. Wet solid wastes can be mixed with cement either in the storage container or in-line. In-container mixing can take one of three forms, gravity mixing, tumbling / rolling or external agitation. With the gravity mixing method, liquid waste is added to a premixed blend of cement and a light weight absorbant such as vermiculite which absorbs the liquid and disperses it throughout the mixture, in the 3.2-5
i tumbling / rolling method, mixing weights are added to the drum along with dry cement. A predetermined quantity of waste is then added to the drum which is capped and trans-ferred to a tumbling or rolling station where the contents are mixed. A mixing blade is lowered into the drum during or after waste addition in the external agitation process. Urea-formaldehyde (UF) resin is an aqueous emulsion of urea and formaldehyde chemically combined to form linear polymeric chains. The resulting liquid is a viscous emul-sion with water containing approximately 65 weight percent solids which is completely miscible with water but immi-scible with nonpolar solvents. Upon the addition of an acid catalyst, cross-linking polymerization by a condensa-tion mechanism occurs and a solid is obtained. During this polymerization, water added to the resin before
~
the addition of the acid catalyst is physically entrapped'
~
in the polymer matrix. The UF emulsion may generally be mixed with aqueous waste in volume ratios of from 1:1 to 1:3, with 1:2 typically recommended as a corapromise between . s product properties and cost. The optimum formulation should be determined for each waste to be solidified. A ' saturated solution.of sodium bisulfate is usually employed ' x as the catalyst and is typically added to 2 to 3 volume percent of the waste-UF mixture to init' tate polymerization. The polymerization is pH dependent and the amount of catalyst . F 3.2-6
1 necessary to produce a waste-UF mixture pH of 1.5+0.5 should be determined for each waste. The formulation will begin to gel within several minutes after addition of the required cuantity of catalyst and will generally form a free standing solid within thirty minutes. The polymeri-zation reaction may continue for several hours, however, during which small quantities of acidic free standing water may be released. Certain wastes such as concentrated sodium sulfate and soap solutions are difficult to entrap . in UF, although such difficulties can be minimized by proper selection of the UF/ waste ratio. Various clays can be added to UF to decrease product leachability. As in the case with cement, the UF mixing process can be carried out either in an in-line_ mixer or directly in the burial container. In both cases, the catalyst is added last to avoid the problems associated with premature resin g setting. With an in-line mixer, the catalyst is added to a g_ the premixed UF/ waste mixture as it enters the solidification n i container. i S Ih% 3.2.5 Asphalt Extruder / Evaporator System i Asphalt and either liquid or sludge wastes are continuously
- i. pumped into one end of a screw extruder which may contain one or multiple screws. The design and operation of the l- extruder is such that the asphalt and waste are intimately 4
4 mixed and spread into.a thin film on the heated surface of i . 3.2-7 [ n e
t I .f 4 N , the extruder barrel. This mechanical processing together with the maintenance of a temperature of appioximately 200 C affects the almost complete (99.S[) evaporation of the water contained in the waste and provides a homogeneous a product. The evaporated water is ventedithrough large disengaging seccions called steam domes and passed through an oil separator before being condensed. Carryover of activity to the condensate is reported to be-less than 0.1%. i The asphalt waste mixture is discharged directly into solidification containers at the end of the extruder and i allowed to cool. Average residence time in the extruder is a few minutes, s e
- 3.2.6 Dry Wastes Contaminated articles such as rags, disposable clothing, cleaning equipment, tools and in some cases, plant equip-ment will be disposed of as dry wastes. This type material will be placed in a hydraulic compactor to reduce the volume and either baled or compacted into 55 gallen dhumo.
Disposable filter cartridges present a different problem-due to abnormally high radiation levels. They are placed , in.special shielded casks for off-site disposal. The compaction, baling or casking procedures that dry ~
. wastes undergo are.usually fully remote operations. '
i i 3.2-8 o ,
. , , . . . . . . , , -m., .r. -c-- ._-..-m~ .- - - , - , . , - y , , . - , , . . , , , . - - - - - . - - - _ . . . . -
l l
- 7 3.3 PRETREATMENT SUBSYSTEMS 3.3.1 Decanting Tanks (Figure 3.2)
The process of decanting has been used for several years, primarily to dewater resin slurries. The spent resin is flushed out of its vessel by a manually controlled back-wash of demineralized water to the spent resin tank. As required, the spent resin tank is pumped into the decant tank. This thin slurry is allowed to settle, with the resin beads dropping to the bottone of the decant tank (step 1). After settling, the decant arm, which is l' actually a movable pump suction tube, is lowered to q just above the resin level. The position of the decant arm is determined by a sonic sensor, which can discern the resin-water interface. The pump will pump the water l out of the tank back to the Liquid Waste System (step 2). ! i l The pump will then switch suction and discharge paths to : take suction on the bottom of the tank and discharge the resins to mixing-packaging system for solidification ! (step 3). When the tank is empty, the pump, decant arm ] and tank will be back-flushed to the spent resin holding i tank. < 3.3.2 Reverse Osmosis Filtration i > ~ Reverse osmosis (RO), as a process, has progressed rapidly
'from the initial laboratory studies in the mid-1950's to a technically practical method for removal of salts from '
brackish and polluted surface waters. With respect to J i
- 3. 3-l '
_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - _ - _ _ - - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' ' - - - - - - - ' - - - - - - ~ ^ ^ - -
nuclear radwaste, R0 is a relative newcomer and only a few plants currently employ RO membrane devices. They are used in processing solution of laundry waste and floor drains. In reverse osmosis (or hyperfiltration) pressurized feed is passed through a selective membrane that retains salts and other low molecular-weight solutes. The difference in concentration between feed and product varies from zero at negligible product flow rates up to a limit given by the characteristic of the feed and properties of membrane at high pressures and product flow rates. The pure water
. will be collected by a holding tank where its ultimate disposal will be determined. The impurities deposited on the wall of the membrane will be disposed of along with a the membrane element when it becomes saturated.
The three types of R0 units currently marketed are the t spiral wound, hollow fiber, and tubular design -- internal and external. Interest in R0 for waste treatment has been principally attributed to the low energy requirements and comparatively low capital costs. The operation of the process is extremely simple, requiring a pressurization of feed through the special permeable membrane. Packaged R0 plants are currently available fror a number of manu-facturers. f , 3.3-2 l
i
/
- 3.3.3 Pre-Coat Filters i t j In a pre-coated filter system, the process fluid is passed
- through a filter media which is in the form of a cake de-posited on a screen or grid. (See Figure 3.3.) The filter 1 medium (usually diatomaceous earth or Solka-Floc powder) is precoated on as a slurry. When the filter is put on j line, solids in the process stream are trapped in and on the precoat cake until the maximum design pressure drop is reached. Then the filter medium with its entrained i
i
, solids are backwashed to the mixing and packaging system for solidification. Variations of pre-coated filters , include traveling bed systems and disc-type elements' i
i with electro-magnetically attached steel spheres as the
~
retaining elements for the precoat media. i.
! 3.3.4 Centrifuge Separation l
Centrifugation is a method for removal of suspended solids
! from liquid waste streams. Centrifuges are mechanically i i
j driven rotating devices which use centrifugal forces to j separate solids from liquids. ~Three types commonly found are tube, disk, and basket. The tube and disk type use rotational force to cause solids to migrate to a wall where they can be collected and the clear liquid overflows. Basket or filtering centrifuges force liquid through a ro- i i { tating screen which collects the solids on a filtering aid , which is; supported t'y the screen. Units were used in early i 3.3-3 i
. . . _ _ - . _ _ , - . _ - . _ . _ _ _ . _ _ _ _ , _ . _ , , _ , . . , _ , _ _ . _ _ - _ . _ _ . _ - . , ~ _ , _ . . . _ . _ , . _ , . _ _ _ . . _ _ . _ . _ .
or free-flowing solid. In addition to concentrate redistri-bution and novement, the rotating blades also serve to ac-complish removal of foam and/or entrainment from the central vapor space. 3.3.7 CalcinerSystems(Figure 3.6[ Calcination is a high temperature process in which liquid wastes are dried and thermally decomposed to form stable, non-fused compounds such as oxides. To date, much of the developmental work in calcination has been directed toward treatment of high-level wastes, but the technology is also applicable to low- and intermediate-level wastes. In re-processing plants, some of the intermediate-level waste strea,ns might be combined with high-level wastes as feed to a calcination process. Various calciner designs are available such as the spray, fluidized bed, pot, and rotary kiln calciners. Except for void spaces, the product form , represents " minimum volume" and can range from powdery dusts to free-flowing granular material to a porous friable cake. In the calcination systems, evaporator concentrates are sprayed into a direct-fired fluidized bed calciner vessel.
- Here water flashes off as the liquid droplets contact hot fluidized hed particles. Heat for this process is supplied by the burning of a combustible fluid (e.g., kerosene) with air directly in the lower part of the vessel. The dry ,,
l 3.3-6 .
I l significant reduction in personnel exposures. Since the heating section is well below the liquid level and the recirculation rate is high, boiling on heat exchanger tubes is usually nominal and the associated fouling.and scale deposition is minimized. This type of evaporator is usually employed in the continuous flow mode of operation whereby feed flow is balanced by the condensate and bottoms output flow. 3.3.6 Thin Film Evaporators (Figure 3.5) The most significant trait of the thin film evaporator (or wiped film dryer as it may be called) is its capability to evaporate to high product concentrations. The heating sur-face consists of a cylinder which contains a rotating agi-tating blade or series of wipers, either maintaini.g a fixed close clearance from the cylinder inner wall or riding on the liquid film on the wall. The cylinder is externally i heated by a steam jacket or other means. Liquid waste is fed into one end of the cylinder and agitated vigorously by the rotating blades, which wipe the waste on the heated surface in a thin turbulent film. Partial evaporation from the film occurs quickly, and subsequent blade travel con-tinuously redistributes the concentrate-film as it progresses toward the discharge end of the cylinder, from which the product is discharged through a nozzle as a liquid, slurry, a 3.3-5
calcined product is washed from the top of the vessel and collected by a cyclone separator for storage. Spent ion exchange resins are received in a holding / dewatering tank and, after being dewatered, the " dry" solids are fed into the fluidized bed calciner/ incinerator vessel for in-cineration. a l 3.3-7 I i
. . , _ . . __,y _. . - , - . . . . - - _ _
.m.c . , . . _ , . _ , - . .. . ., . _
- The two basic types of cement solidification systems are ;
5 the drum-tumbling mixing system and the in-line mixing system. The prime differences are in where the mixing of !
- j. the cement and waste occurs. Advantages of cement as a
] solidification agent are its availability, low cost, fire i resistance and well-known characteristics. Disadvantages in- , clude post leachability and problems with solidification if j the correct waste / cement ratios are not maintained, i i Sodium Silicate is used as an additive in some cement 1 l systems to increase the waste / cement ratio to lower the ! total number of containers used. It is especially effi-
- cient when tha waste is boric acid solutions. In the case of waste re. sins or filter sludges and powders, liquid I
; wastes or pure water may have to be added to form an i j acceptable waste for good setting properties.
1 i All operations are carried out with remote handling equip-3 ment located in shielded areas. Automatic sequences sim-
- plify individual processes and reduce the possiblity of a
operator errors. Special shielding arrangements have been : i established to improve access to equipment requiring routine i maintenance. ' [ 3.4.2 Urea Formaldehyde Systems a Urea-formaldehyde is physically in the form of a white con-j centrated liquid which may be diluted with equal volumes of water to form a solidified mass with strengths ranging I 3.4-2 ,
k 3.4 SOLIDIFICATION AGENT SYSTEMS ; i j - 3.4.1 Cement Systems
}
Cement is used to solidify free liquids by chemically binding ! i' the heavy liquid waste into a crystaline structure. Differ- ! e l ent' mixtures and ratios of cement and additives are used i 1 depending upon the type of waste to be solidified. Cement has been commonly used as a solidification agent and is generally acceptable to all burial sites.
- Cement is usually handled in bulk quantities to reduce costs. *
!L Cement trucks generally blow dry cement into storage hoppers. ! I The exhaust air from this operation is vented through filters i . tc prevent cement dusting in the solidification area. 1 The dry cement is metered to the waste container or waste i mixer using rotary feeders or gravity with a system of scales { a and/or timers. - 1 b 4 An operator prepares the drums for waste filling by ini-l tiating an automatic dry cement fill procedu're. An overhead i ,
- crane loads the empty drums onto the conveyer. At the dry L l cement loading station the cap is removed, the drum is j
loaded with the proper amount of dry cement, a mixing weight [ 1 } is added and the cap is replaced. These drums are now
- ready to be filled with waste materials (Figure 3.7).
i [ j ' i
- 3.4-1
._, . . . _ .,_-~ . . . _ - . _ _ - , - _ - - , _ . _ _ _ _ . . - - . _ . _ _ _ . _ _ , . _ _ _ . . _ . . . _ , _
3.4.3 Asphalt Extrusion-Evaporator System Bitumen or asphalt is a mixture of high molecular weight hydrocarbons obtained as a residue in petroleum or coal-tar refining. Several types of bitumen are available, but the direct-distillation product is the one most widely suggested for radwaste solidification. There are no power reactors in this country that are using a bitumen solidifi-cation process , though it is used in Europe. Bitumen sys-tems appear to be able to handle most reactor waste streams and permit a wide latitude of waste proportions. Substances which decompose at the working temperature of the bitumen process should not be added. Bitumen does burn and there is some evidence that the incorporation of oxidizing agent's increases the fire risk. In this process, (Figure 3.8) bitumen and either liquid or sludge wastes are continuously pumped into one end of a screw extruder which may contain one or multiple screws. The design and operation of the extruder is such that the bitumen and waste are intimately mixed and spread into a . thin film on the heated surface of the extruder barrel. This mechanical processing together withthe maintenance of a temperature of approximately 2000 C affects the almost complete (99.5%) evaporation of the water contained in the waste and provides a homogeneous product. The evaporated water is vented through large disengaging sections called steam domes and passed through an oil separator before 3.4-4
i I
- ( from 400 to 500 psi. Gel times can be controlled by the 1
amount of catalyst used. Several different acids may be l used as the catalyst, and PPI uses sodium bisulfate. This ' i i material can be diluted in water or other liquids to make ) distribution into the system easy. Increasing the concen-l tration of the catalyst shortens the gel time as well as the resultant full strength time. Temperature increases i of the polymer also work for this same end. l Figure 3.8 shows a simplified schematic of the U-F process j flow. Urea-formaldehyde is pumped and mixed with radwaste )s from the radwaste pump through a mixer to the disposable liner where the catalyst is pumped in separately for mixing within the liner. The system uses progressive cavity-type i pumps throughout its system. It can be used to meter fluids
- since it is a self-priming positive action pump. It can be used on corrosive chemicals, abrasive slurries, plus !
resins, powdered resins, or filter aid material such as j solka floc. i The process module has the four pumps located on a standard 4 4' x 8' skid. The skid also includes the static mixer, two l 3-way stainless valves with electric actuators, miscellaneous
- piping, and three thermally actuated flow switches. The j catalyst is introduced inside the disposable container and 4
4
. mixed therein with the incoming radwaste and U-F mixture. '
. Several skids can be provided to keep the two nonradioactive i pumps separate from the radioactive pumps.
4
- 3.4-3 L .
Discharge Handling The extruder-evaporator normally discharges the asphalt / l salts mix into standard DOT 55-gallon drums at a rate from l one-fourth of a gallon to 30 gallons per hour, depending upon the size and speed of the extruder and the concentra-tion of the feed stream. However, other sizes and types of containers can be used if desired. Normally, a filled 4 drum is allowed to cool and settle for a period up to 24 hours before it is closed for storage prior to transport to the disposal site. If radiation levels require addi-tional shielding around a drum, permanent or reusable casks of concrete or other material can be used. The radioacti-vity level in the final product is governed by the ratio of asphalt to radioactive salts fed into the extruder as well as the type and concentration of activity in the salts. Remote Control System 1 The entire process, complete with interlocks, is controlled remotely from a central control panel, with local controls included where needed or desired to meet individual site installation requirements. the control system allows the solidification process to be operated in total or in part, either continuously or in an on-off mode. This versatility provides the operator a number of advantages over a wide range of operating conditions. For example, the materials input can be stopped while the extruder-evaporator continues 3.4-6
l i i / being condensed. Carryover of activity to the condensate is reported to be less than 0.1%. The bitumen waste mix-ture is discharged directly into solidification containers at the end of the extruder and allowed to cool. Average residence time in the extruder is a few minutes. All of the processing takes place in the extruder-evaporator at near ambient pressure, a safety feature. All of the feed materials enter the extruder in the first and/or second barrel section where they are imediately enmeshed in the mixing action of the self-cleaning screws. Simultaneous with the mixing, the excess water is evaporated through steam domes located downstream from the feed barrel as shown in Figure 3.8. The number of domes needed varies up to six, depending on the size of the extruder needed to evaporate 99.5 percent of the water out. Evaporation rates for the various size extruders range from one up to 200 liters of water per hour. The evaporation rate needed to achieve a 99.5 percent moisture removal is the prime factor in determining what size of extruder is re-quired in any given situation. The extruder-evaporator is connected to a variable speed motor capable of turning the screws up to 300 revolutions per minute. The design of the extruder provides for a shielding wall to be erected between the motor drive and the feed barrel. 3.4-5
4 the compactor can be set into a shallow pit so that the bottom of the drum is level with the floor in order to make hand truck handling easier, i Most compactors have been designed with a 20,000 pound 1 i maximum force. The compactor is designed to comptet up to a 60-inch stack of waste. It permits the operator to roll up paper used during refueling, place in end-wise into the i drum, close the doors, and compact the rolls. A hinged 1 , loading table also provides a means of positioning the drum i i for loading. Both the loading table and the enclosure door must be closed before compacting. i l The drum extention space is evacuated by a built-in fan which draws the air through a roughing filter and a HEpA filter. Pressure gages are for indication of filter fouling.
~
Used filters can be dropped into a drum by means of a hook, f hence avoiding direct handling. The various operations required prior to actual compacting are interlocked with limit switches. These interlocks pre-I vent the ran from operating unless the drum is all the way into the compacting position and unless both the loadtbg table and the enclosure door are closed and latched. Further , safety is provided by deadman-type controls for each of the two hydraulic cylinders. The hydraulic fluid is usually phosphate ester to minimize fire hazaro. ) 3.4-8 l _ _ . _ , . - _ _ . , , _ _ _ . _ _ . . _ . _ _ ~ _ . . ~ _ . _ - . , _ . _ . , _ _ _ _ . , __ _ . _ . . . . . , _ . _ _ _ . . - .- .m.,.-__. _ ._. __,
] l/ ,- i to operate. Beacuse the in-process inventory is small and the residence time is short, it will take only a minute { for the extruder to empty and self-clean. In the event I of a power and steam heat failure or other incidents that would cause the entire system to stop while filled with 1 asphalt product, it is a comparatively simple procedure to restart. When heat and power are restored, simply allow l j the asphalt to heat up to its flow temperature, turn the ( f j f process controls on, and normal operation will be resumed ' l ,
; automatically, i ;
i 3.4.4 Dry Waste Compactor i j About all nuclear plants have a dry waste baler or compactor, to reduce the volume of loose dry waste requiring shipment : to a licensed burial site. l ! The compactor is designed specifically to compact paper, cloth, glass, floor sweepings, and other low-level dry waste 1 into standard 55-gallon drums. It consists essentially of j a hydraulic system with a ram operating vertically downward,
- a contoured support plate, frame, and safety enclosure with loading table, vent, and filter system with fan, and gages i and controls.
l The unit has been designed so it receives the drum out in front of the compactor and can be placed into position and i taken away with an overhead crane or lif t truck. Furthermore, < 1 i ,- 3.4-7
i l 3.4.5 Filter Transfer System ( t Pressurized water reactors usually have cartridge filters ! i i i in the system which need replacement af ter the filter has { reached high radiation or high pressure drop. The radia- : i l l tion level on these filters may be high enough such that ' q a shield transfer cask is required to move the spent car- l 4 l 2 tridge to the solid radwaste packaging system. The cask l l incorporates an internal hoist arrangement with grapple which is operated oy a simple pendent switch. A drip pan , and check valve assembly are provided to insure a drip-1
- proof trar,sfer to the fill station. The assembly has an 8-inch nominal inside diameter and has shielding equivalent ;
to 5 inches lead. The filter can be solidified if necessary i l with Urea-formaldehyde and other waste in a 50 cu. f t. I i . cask and shipped in a shielded transportation cask. t a ! t i i
?
i f i ! 1 l I - i I f ' I 1 I I l i 3.4-9 1 I
. )
3.5
SUMMARY
Final disposition of radioactive wastes has recently been receiving more attention from utility management. Not least among these is economic. Operating costs have risen to the level of several millions of dollars per year with no indication of leveling off. Much of the development energy is now directed toward reducing the total volume of packaged waste being shipped off-site. Increasing con-cern for the detection and elimination of free water in the final solidified mass is also a factor in this selection of a new or back-fitted Solid Waste Disposal System. Early plants had simple systems predicated on meeting 10 CFR 20 release limits. Earlier p'hllosophy was to rely heavily on the regulatory limits since these limits had large safety margins and were considered safe. Later philosophy has tended to enhance the "as low as resson-ably achievable" goal. Consequently, new approaches have been taken, new equipment designed and purchased, and greater emphasis placed on literally attaining zero releases. 3.5-1
1 WET WASTES COLLECTION MIXING 1ACR AGING (DEW ATERING/ VOLUME REDUCTION) HANDLING SYSTIMS CONI AINE R H ANDllNG
* ' +
SPENT RESIN gg h NTANK I
/ / a
+
MINING & PACKAGING
- CEMENT :
FULL
+ CE Niettf DGE , SYSTE MS O DRUMS FILTE RS LUDGE TANK URE A-h l
,' DEWATERING FCRMALDE HVDE -*- I
+ (PRE-CO AT E D) , STORE L FILTERS l
k EVAPORATOR _ _ B Y P ASS _ _ _ _ CONCENTRATE TRUCK TANK EVAPORAT OR $V ST E MS LOAD
, . (THIN TILM. FORCED ,
i CIRCULATION ETC.) L-
" SPE CIALtoiHE R FLUIDS T ANK , CALCINER ,
i SYSTEMS i
; FILTR ATION SYSTEMS i + (REVERSE OSMOSIS, +
BED DRVE RS.ETC )
! g ASPHALT EXTRUDER / EVAPORATOR ,
I i DRV WAST E S CARTRIDGE FILTER FILTE RS CASKING
- y DRY WASTES
- R AGS. CL OT HING. BALER Q ETC.
2 CONT AMINATE D F TOOLS &
- CDMPONENTS SOLID WASTES PROCESSING FLOW DIAGRAM lI ll l
1 i TT _ 4 % MIXER R LEVEL SENSOR C 5 A. IM r- iil .'" m Step 1. The thin slurry is allowed to settle, with rn.... 3.. . . . . j; .. the heavier resin beads ffhllll DECANT ARM 11 I - Al' N ll .,
.[.
'.'{.:.
C- ..f. , and other solids settling
~.
to the bottom of the tank I l[ i(h j'd<f
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TABLE OF CONTENTS (CONTINUED) Section Ti tle Page 4.4 DUAL CONTAINMENT AIR PURIFICATION AND CLEANUP SYSTEM. . . . . . . . . . . . . . . . . . . 4.4-1 4.4.1 I n t rod uc ti o n . . . . . . . . . . . . . . . . . 4.4-1 4.4.2 System Design and Operation . . . . . . . . . 4.4-3 4.4.2.1 Emergency Gas Treatment System. . . . . . . 4.4-3 4.4.2.2 Auxiliary Building Gas Treatment System . . 4.4-12 4.4.2.3 Ice Condenser . . . . . . . . . . . . . . . 4.4-16 4.4.3 Suma ry De s c ri p ti on . . . . . . . . . . . . . 4.4-16 e 11
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i , i j TABLE OF CONTENTS } Section Title Page i 4.0 VENTILATION SYSTEMS . . . . . . . . . . . . . 4.1-1 4.1 AUXILIARY BUILDING VENTILATION SYSTEMS. . . . 4.1-1 4.1.1 In t roducti on . . . . . . . . . . . . . . . . . 4.1-1 4.1.2 System Design and Operation . . . . . . . . . 4.1-1 ! 4.1.3 Components. . . . . . . . . . . . . . . . . . 4.1-7 - 4 4.1. 3.1 Supply Sys tem . . . . . . . . . . . . . . . 4.1-7
! 4.1. 3.1.1 Supply Fil te rs . . . . . . . . . . . . . . 4.1-7 4.1.3.1.2 Heating Coils . . . . . . . . . . . . . . 4.1-8 :
I 4.1. 3.1. 3 Supply Fans . . . . . . . . . . . . . . . 4.1-8 4.1.3.1.4 Cooling Coils . . . . . . . . . . . . . . 4.1-8 l 4.1.3.2 Exhaust System. . . . . . . . . . . . . . . 4.1-8 4.1.3.2.1 Exhaust Fans. . . . . . . . . . . . . . . 4.1-8 , 4.1.3.2.2 Charcoal Booster Fans . . . . . . . . . . 4.1-9 '
- 4.1.3.2.3 Fuel Handling Building Exhaust Filters. . 4.1-9 4.1.3.2.4 Cubicles Exhaust Filters. . . . . . . . . 4.1-9 3 4.1. 3. 2. 5 Main Exhaust Filters. . . . . . . . . . . 4.1-10 4.1.3.2.6 Charcoal Fil ters . . . . . . . . . . . . . 4.1-10
- 4.1.3.2.7 Cubicle Unit Coolers. . . . . . . . . . . 4.1-10 1
4.13.2.8 Drumming Station Exhaust Filters. . . . . 4.1-11 4.1.4 Summary . . . . . . . . . . . . . . . . . . . 4.1-11 4.2 CONVENTIONAL CONTAINMENT VENTILATION SYSTEMS. 4.2-1 4.2.1 I n t rod uc ti on . . . . . . . . . . . . . . . . . 4.2-1 4.2.2 System Design and Operation . . . . . . . . . 4.2-3 4.2.2.1 Containment Purge System. . . . . . . . . . 4.2-3 4.2.2.2 Pressure and Vacuum Relief System . . . . . 4.2-9 4.2.2.3 Hydrogen Removal Systems. . . . . . . . . . 4.2-11 4.2.2.3.1 Hydrogen Purge System . . . . . . . . . . 4.2-11 4.2.2.3.2 Hydrogen Recombiner System. . . . . . . . 4.2-12 4.2.2.4 Containment Fan Coolers . . . . . . . . . . 4.2-13 4 4.2.2.5 Containment Activated Charcoal Filter Units 4.2-16 4.2.2.6 Manipulator Crane Ventilating System. . . . 4.2-17
- 4.2.2.7 Other Containment Ventilation Systems . . . 4.2-18 4.2.3 S umma ry . . . . . . . . . . . . . . . . . . . 4.2-18 4.3 DUAL CONTAINMENT VENTILATION SYSTEMS. . . . . 4.3-1 4.3.1 Introduction. . . . . . . , . . . . . . . . . 4.3-1 4.3.2 System Design and Operation . . . . . . . . . 4.3-3 4.3.2.1 Containment Air Cooling Systems . . . . . . 4.3-3 4.3 2.2 Containment Purge System. . . . . . . . . . 4.3-9 i 4.3.2.3 Containment Vacuum Relief System. . . . . . 4.3-14 4.3.2.4 Containment Air Return Systems. . . . . . . 4.3-17 4.3.2.5 Containment Combustible Gas Control System. 4.3-20 j 4.3.3 Summary Description . . . . . . . . . ... 4.3-26 i
l i j 4
LIST OF TABLES Tables Title Page 4.3-1 Data Table for the Vacuum Relief System . . . 4.3-33 4.3-2 ' Electric Hydrogen Recombiner Typical Para-
- meters. . . . . . . . . . . . . . . . . . . 4.3-35 4.4-1 Dual Containment Characteristics. . . . . . . 4.4-23 LIST OF FIGURES Figures Title Page 4.1-1 Auxiliary Building V,entilation Supply System 4.1-13 4.1 -2 Auxiliary Building Exhaust System . . . . . . 4.1-15 ,
I 4.2-1 Conventional Containment Ventilation System . 4.2-25 4.2-2 Containment Purge and Relief System . . . . . 4.2-27 4.2-3 T'hermal Recombiner Piping and Instrumentation Fl ow Di a g ram. . . . . . . . . . . . . . . . 4.2-29 4.2-4 Reactor Containment Fan Cooler. . . . . . . . 4.2-31 4.3-1 Containment Ventilation Systems . . . . . . . 4.3-37 4.3-2 Schematic of Air Return Fan System. . . . . . 4.3-39 4.3-3 Electric Hydrogen Recombiner. . . . . . . . . 4.3-41 4.4-1 Emergency Gas Treatment System Schematic , Diagram . . . . . . . . . . . . . . . . . . 4.4-25 . ~~s 4.4-2 Auxiliary Building Gas Treatment System Exha us t Ne two rk . . . . . . . . . . . s. . . 4.4-27 4 I s
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as required to maintain a nominal supply air temperature of 75 F + 100F. The system incorporates three, 50% capacity supply fans; two which normally operate and one which is standby. During normal operation (full power or shutdown), two supply fans are in operation. In the event of a loss of off-site power, the system is designed to operate with one ventilation fan which is connected to the emergency power supply. During normal operation, the two operating supply fans are controlled to maintain a constant supply volume in the main supply duct. Supply air is ducted to various areas in the Auxiliary and Fuel Handling Buildings and in general is delivered to clean areas which are normally accessible. The volume of air delivered in each area is based on the quantity of heat to be dissipated and/or to provide sufficient air change for personnel occupancy. All of the ventilation air ficws to areas of' progressively greater potential contamination where it is returned through a duct system to the Auxiliary Building Ventilation System equipment room. ~ Pressure differential control dampers are located as required to maintain potentially contaminated areas at a - negative pressure. All exhaust air which is returned from the auxiliary building and fuel handling building is filtered through HEPA filters which will be tested on site for a nominal bank efficiency of 99.95% based on 0.3 micron D0P tests. 4.1-2
4.0 VENTILATION SYSTEMS 4.1 AUXILIARY BUILDING VENTILATION SYSTEMS l 4.1.1 Introduction The Auxiliary Building Ventilation System serves all plant areas of the Auxiliary Building and Fuel Handling Building. The system consists of supply and exhaust subsystens designed to maintain negative pressures in potentially contaminated areas thus resulting in minimal unintentional activity spread or release. These subsystems normally condition and filter all air entering and leaving the above mentioned buildings. In addition, a special charcoal adsorption subsystem is automatically utilized to process exhaust air from critical portions of these buildings in the event of high radiation. All exhaust air is discharged from the plant through tall stacks thereby neximizing the dispersion of potential activity upon release to the environment. 4.1.2 System Design and Operation The Auxiliary Building Ventilation System is shown on Figures 4.1-1 and 4.1-2. The supply system filters 100% outdoor air in two stages where the final stage has a nominal efficiency of 85% based upon the NBS atmospheric dust spot test. The filtered air is heated or cooled I 4.1-1 l l
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a) Boric acid evaporator feed pumps i b) Hold-up tank recirculating pump c) Primary equipment sump tank and pump d) Auxiliary building sump A e) Residual heat removal pumps f) Diesel residual heat removal pumps (future) g) Cavity fill pumps (future) h) Gas decay tank
- 1) Auxiliary building sump B i j) Containment spray pumps k) Residual heat exchangers
- 1) Safety injection pumps m) Charging pumps l n) Boric acid and radwaste evaporators A local filter system is provided for the drumming station exhaust air. Nonnally, the exhaust air from the drumming station bypasses the local filter. Whenever the drumming station is in operation, the exhaust air is passed through a HEPA filter and charcoal filter before being routed to i
the main ventilation exhaust system. Six auxiliary building exhaust fans are provided; four which nonnally operate and two which are standby. The exhaust fans are arranged such that two out of three exhaust fans discharge the air flow to each of two stacks. In the event of a loss of off-site power, the system is designed i to operate with two exhat st fans. I 4.1-4 w, n- - - - - - - - --,.-,,m-- , , - ,- -
d The HEPA filters are arranged in three separate banks. The l main bank, which filters return air from general areas in the auxiliary building, consists of six filter modules; five which operate normally and one module is standby. A 2econd bank of HEPA filters for filtering exhaust air ) from the auxiliary building equipment cubicles consists of three filter modules; two which operate normally and one { which is standby. A third HEPA filter bank filters exhaust air from the fuel handling building and consists of two filter modules; one normally operates and one is in standby. l The spare exhaust filter modules are provided to pemit the replacement of expended filter elements without interrupting j the normal exhaust air capacity. The filtered exhaust air
, is then routed up the ventilation stacks.
l Two banks of charcoal filters have been provided and on l detection of high radiation, air from potentially con-taminated equipment cubicles or the pipe tunnels will auto-matically be routed through the charcoal prior to exhaust. In addition, exhaust air from the fuel handling building is routed through the charcoal during the refueling oper-
- ati on. A maximum of three out of four charcoal booster
! fans will automatically start to account for the increase in system resistance in accordance with the required flow through the charcoal. The exhaust air from cubicles of . potentially high radiation which can be passed through-the charcoal filters include the following: 4.1-3
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cooling coil housed in a common cabinet. One of the fans on each cubicle unit cooler is a spare. However, whenever cooling is required, all fans will automatically start. The design allows for the loss of one fan without the loss of design cooling capacity. The auxiliary building cubicle unit coolers are connected to the emergency power supply and are designed to operate under any normal or abncrmal plant condition. Cubicle unit coolers have been .i i provided for the following equipment. a) Residual heat removal pumps b) Safety injection pumps c) Containment spray pumps 4 d) Charging pumps All essential operating functions are monitored and con-trolled from the control room. Each supply and exhaust fan may be manually started and stopped from the control Provision has been made for the automatic starting room. of standby fans in the event of a failure or trip of an operating fan. Operation of the auxiliary building cubicle unit coolers is completely automatic ar.d these units start whenever the equipment in the respective cubicle operates. 1 _. All system temperature control is maintained from a local panel and the supply temperature is maintained between 0 the limits of 65 F and 85 F.
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The auxiliary building ventilation stacks have been compart-mentalized to handle, independently, the exhaust from dif-ferent systens. The compartmentalized design will maintain the design discharge velocity for any of the exhaust systems regardless of operation of the other systems. The
- ventilation stack for unit one has three compartments with one compartment each for the auxiliary building ventilation exhaust, the containment purge exhaust for unit 1, and the gas vent for the gaseous radwaste system. The ventilation stack for unit 2 has three compartments with one compart-ment each for the auxiliary building ventilation exhaust air, containment purge exhaust for unit 2, and a niiscel-laneous exhaust from the hot and cold laboratories, decon-tamination room, battery rooms, hot instrument room, kitchen and toilet.
Equipment located in the safeguard equipment cubicles does not nornally operate. These cubicles are connected to the ventilation system in such a manner so as to pro-vide a normal amount of ventilation during operation. Whenever the equipment within an individual cubicle is operating, supplementary cooling is required. To meet this function, auxiliary building cubicle unit coolers are installed. The unit coolers are each designed to limit the maximum ambient temperature to 105U F. Each cooler is equipped with either two, three, or four ventilating fans arranged to operate in parallel with a single section 4.1-5
. _~
- ~ - _ _ - .. .. ._- -
l 4.1.3.1.2 Heating Coils The auxiliary building ventilation heating coils are com-posed of twelve (12) coil sections arranged in 2 sets in parallel, each of six (6) coils. The heating coils are designed to heat 300,000 cfm from -10 F to 65 F when supplied with hot water at 270 UF. The total coil capacity is 24.3 x 106 Btuh. 4.1.3.1.3 Supply Fans The auxiliary building ventilation supply fans are of the direct-driven vane axial type located down stream of the heating coils. Each fan is rated at 150,000 cfm at a total pressure of 6.5" H O and is driven by a nominal 200 2 HP motor. 4.1.3.1.4 Cooling coils The auxiliary building cooling coils are composed of twelve (12) sections arranged in 2 sets in parallel, each of six (6) coils. The cooling coils are designed to cool 300,000 cfm from 98.5 F to 81.5 0F when supplied with water temper-ature at 78 F. This guarantees the minimum required capacity to supply air at 85 F in case of a loss of one coil section. 6 The total coil capacities is 5.53 x 10 Btuh. i 4.1.3.2 Exhaust System 4.1.3.2.1 Ex;.aust Fans The auxiliary building vent system exhaust fans are of the direct-driven vane axial type. Each fan is rated at 75,000 cfm at a total pressure of 9" H 20 and is driven by a nominal J 150 HP motor. 4.1-3
. - .. ~ - .
All volume control is from the local control panel. The supply fans are controlled to maintain constant air supply to the auxiliary building. The exhaust fans are controlled to maintain the auxiliary building at a nominal 1/4" negative
- pressure with respect to the outdoors. The pressure in potentially contaminated areas is controlled for approxi- l mately 1/4" negative pressure with respect to adjacent clear areas in the auxiliary building.
System variables pertaining to normal operation are indicated on the main control room panel. Abnormal conditions, such as high temperature, low temperature, low building differ-ential pressure, high pressure drop a' dross filters, and high radiation are annunciated both locally and on the main con-trol panel. 4.1.3 Components 4.1. 3.1 Supply System 4.1. 3.1.1 Supply Filters j The auxiliary building ventilation supply filters are com-posed of banks of prefilters and high efficiency filters installed in series. Each filter unit has a rated flow of 300,000 cfm. Each prefilter bank contains 168 indi-vidual filter elements rated at 10% efficiency based on the NBS atmospheric dust spot test. Each high efficiency filter bank contains 168 individual filter elements rated at 85% efficiency based on the NBS atmospheric dust spot test. , 4.1-7 [' _ . . _ _ . . . . _ - ._. _ _
test. Each HEPA filter bank contains 24 individual filter elements each having a nominal efficiency of 99.7% based on the D0P test. 4.1.3.2.5 Main Exhaust Filters The auxiliary building vent exhaust filters are composed of banks of prefilters and HEPA filters installed in series. Each filter unit has a rated flow of 48,000 cfm. Each pre-filter bank contains 48 individual filter elements rated at 85%. efficiency based on the NBS atmospheric dust spot test. Each HEPA filter bank contains 48 individual filter elements each having a nominal efficiency of 99.7% based on the D0P test. 4.1. 3. 2. 6 Charcoal Filters The auxiliary building charcoal exhaust filters are composed of banks of charcoal filters rated at 32,000 cfm. Each filter unit contains 91 individual filter elements. 4.1.3.2.7 Cubicle Unit Coolers The auxiliary building cubicle unit coolers are designed to limit the maximum ambient temperature to 105 F. Each cooler is equipped with ei.ther two, three, or four venti-lating fans arranged to operate in parallel with a single section cooling coil housed in a common cabinet. One of the fans on each cubicle unit cooler . a spare. However, whenever cooling is required, all fans will automatically 4.1-10 6
4.1.3.2.2 Charcoal Booster Fans The auxiliary building vent system charcoal booster fans are of the direct-driven vane axial 'vpe. Each fan is rated at 22,000 cfm at ,a total pressure of 3.5" H O and is 2 driven by a nominal 15 HP motor. These fans are designed to overcome the additional resistance of the charcoal filters when the air from the cubicles, fuel handling building or the pipe tunnel is routed through the charcoal filters as a result of high radiation or during refueling operations . 4.1.3.2.3 Fuel Handling Building Exhaust Filters
)
The fuel handling building exhaust filters are composed of banks of prefilters and HEPA filters installed in series. Each f'ilter unit has a rated flow of 22,000 cfm. Each prefilter bank contains 24 individual filter elements rated i at 85% efficiency based on the NBS atmospheric dust spot test. Each HEPA filter bank contains 24 individual filter elements each having a nominal efficiency of 99.7% based on the D0P test. 4.1.3.2.4 Cubicles Exhaust Filters The auxiliary building cubicles exhaust filters are composed of banks of prefilters and HEPA filters installed in series. Each filter unit has a rated flow of 20,000 cfm. Each pre-filter bank contains 24 individual filter elements rated at 85% efficiency based on the NBS atmospheric . dust spot l l 4.1-9 l
, .. -. . ._.. -_ __ - _._. _ ~ . . .- _ -. _ . ..
d i system as described above. During normal operation, the temperature in equipment areas is limited to 105 F. In the event of a loss of outside power, this temperature is limited to 115 F, except in cubicles with unit coolers I where the temperature is limited to 1050F. Conditioned air is supplied to clean areas and is routed to areas of progressively greater potential contamination. Pressure differential control dampers are employed as required to maintain a nominal 1/4" negative pressure in potentially contaminated cubicles or pipe chases. All exhaust air is routed through a return duct system, passed through HEPA filters and discharged to two ventilat' ion stacks which are directed up the side of each containment building. In addi- .
- tion, normally bypassed charcoal filters are provided and, j subject to high radiation, exhaust air from the fuel h handling building, potentially contaminated equipment cubicles, or pipe tunnel can be routed through the charcoal filters prior to discharge to the ventilation stacks.
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4.1-12
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l start, and consequently, the design allows for the loss of one fan without the loss of design cooling capacity. 4.1.3.2.8 Drumming Station Exhaust Filters The drumming station exhaust filters are composed of banks of prefilters, HEPA filters and charcoal filters. Each has a rated flow of 4,000 cfm. The prefilter bank contains 4 individual filter elements having a nominal efficiency of 35% based on the NBS atmospheric dust spot test. The HEPA filter bank contains 4 individual filter elements having a nominal efficiency of 99.7% based on the DOP test. The 1 l charcoal filter bank contains 12 individual filter elements. 4 4.1.4 Summa ry The Auxiliary Building Ventilation System serves all plant areas of the Auxiliary Building including the Fuel Handling Building. The Auxiliary Building Ventilation System also incorporates individual cubicle cooler units to provide
- supplementary cooling to specific safeguard equipment 3 cubicles.
The Auxiliary Building Ventilation System is designed to provide a continuous source of filtered, temperature con-ditioned outdoor air to maintain a thermal environment in I accordance with the maximum ambient temperature for the operating equipnent in the various areas served by this 4.1-11
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f 4 i The Pressure and Vacuum Relief System is provided to compen-sate for the nomal containment pressure changes resulting ) from containment temperature changes, barometric pressure changes, instrument air bleed, and inleakage from the pene- , r tration pressurization system. l ! The Hydrogen Purge System is designed to be utilized follow- i l ing a loss of coolant accident to purge hydrogen from the containment if it reaches 3% by volume. The Hydrogen Recombiner System is provided for essentially I the same reason as the hydrogen purge system; however, with ', the recombiner system, there is no necessity for off-site 4 release. . The fan cooler units provide cooling so as to limit the air 1 temperature in the closed containment to a maximum temper-ature of 1200F during all normal modes of operation and a minimum of 65 F during shutdown conditions. The system also is designed to remove heat and particulate radioactivity } from the containment as required following a loss-of-coolant i I accident. The Containment Activated Charcoal Filter Unit System is
- provided to pemit cleanup of the containment atmosphere prior to limited personnel access at power and prior to personnel access for refueling.
i !. 'J l
- 4.2-2 4
, i 4.2 CONVENTIONAL CONTAINMENT VENTILATION SYSTEMS 4.2.1 Introduction The Containment Ventilation System is comprised of several subsystems each of which has separate design objectives. .
- The systems are designed around the objective of dealing with normal operation, personnel entry, containment pro-tection, and accident conditions. The following subsystems l are included in the total ventilation system
I 1. Containment Purge System
- 2. Pressure and Vacuum Relief System
- 3. Hydrogen Removal Systems
- 4. Containment Fan Coolers
- 5. Containment Activated Charcoal Filter Units
- 6. Manipulator Crane Ventilating System
- 7. Other Containment Ventilation Systems '
i j The Containment Purge System is designed to assure safe,
- continuous access (40 hours / week) to the containment within three hours after a planned or unplanned reactor shutdown !
by reducing the airborne particulates of the containment atmosphere. 3 The Containment Purge' System provides filtered, heated as required, outside air to the reactor containment. The purge exhaust is routed through HEPA filters prior to dis-charge to the ventilation stacks. 4 f 4.2-1 l l
- - - .___ --- -. . . -. - = .
E concentration. This purge rate provides approximately 1-1/2 air changes per hour and will permit safe access to the containment three hours after a planned or unplanned shutdown. The purge exhaust is routed through HEPA filters prior to discharge to the ventilation stacks. The purging system exnaust and air supply connections through the containment are provided with two tight-seating,125 psig butterfly valves with one located inside and one located out-side the containment. These valves are closed during plant operation with the space between the valves pressurized by the penetration pressurization system. Interlocks to the radiation monitoring system and containment pressure sensors automatically close the butterfly valves upon a high containment activity or high containment pressure < signal. Closure of any of the butterfly valves trips both the purge supply and exhaust fans. All actuators a.re remotely controlled from the control room I by the operator, except where automatic interlocks are in-volved as described above. All actuators are designed to fail in the position required for post accident operation j upon loss of electric or pneumatic power. ! The Containment Ventilation System is shown on Figures 4.2-1 and 4.2-2.
.)
4.2-4 1
- ._- . - - , , _ _ _ . , , _ _ y , . ~ .,_.- , _.- . ~ . _ _ . _. ,_. _.- .._.
The Manipulator Crane Ventilating System is designed to i minimize the airborne hazard to operators working on the manipulator crane during refueling operations. Other Containment Ventilation Systems are provided to cool various equipment located inside containment. 4.2.2 System Design and Operation 4.2.2.1 Containment Purge System The Containment Purge System is designed to assure safe, continuous access (40 hours / week) to the containment within i three hours after a planned or unplanned reactor shutdown by reducing the airborne particulates of the containment atmosphere. Prior to activatir.g the purge system, the particulate and gas monitor will indicate the system activity i levels inside the containment and is used as a guide for routing release from the building. A manual sample is taken and analyzed and a release permit is completed before purging begins. The Containment Purge System provides 40,000 cfm of filtered, heated as required, outside air which is delive. red to the , reactor containment around the periphery of the reactor refueling pool. The air is discharged through linear grills to create a fluid Faundary, or air curtain, between access areas surrounding the pool and air with a potential tritium i i 4.2-3 i i e r ~ y .c _ -- .,,. , --. - .m.. . - . ~-. . . .. - . ~y ,-. . , - ,- . - - - - - - - - , - - - - - . . _-..
purge exhaust particulate, gas or iodine monitor. The isolation valves will also trip shut upon receipt of a containment isolation signal. The flow then passes through a pre-filter and a HEPA filter to the suction of the purge exhaust fans. The two 100% capacity purge exhaust fans are each rated at 40,000 cfm. If the purge supply temperature decreases to 60 F or if any of the four containment isolation valves closes, the purge exhaust fans will automatically trip. The purge exhaust flow is then routed to the plant vent stack for discharge to the environment. Purge Supply Filters The Containment Purge System supply filters are located in the supply air inlet, and are composed of banks of prefilters and high. efficiency filters. The filters are arranged in two parallel independent modules and each module has a rated air flow of 20,000 cfm. The prefilter bank in each module contains 10 individual filter , elements having a ' nominal efficiency of 35% based on the NBS atmospheric dust spot test. The high efficiency filter bank in each module contains 10 filter elements each naving a nominal efficiency of 85% based on the NBS atmospheric dust spot test.
/
4.2-6
Outside air enters the containment purge system through an inlet damper and a prefilter. The flow then passes through a heating coil which maintains the purge supply temperature at 65 F. The two 100% capacity purge supply fans are each rated at 40,000 cfm. If the purge supply temperature decreases to 60U F or if any of the four containment isolation valses closes, the purge supply fans will automatically trip. The two containment isolation butterfly valves are normally closed and the space between them pressurized by the penetra-tion pressurization system. During purging operations, the 2 isolation valves will be tripped shut on a high radiation signal from the containment particulate or gas monitor, or from the purge exhaust particulate, gas or iodine monitor. The isolation valves will also trip shut upon receipt of a containment isolation signal. The purge supply flow is then ducted to the periphery of the refueling pool. The purge exhaust exits the containment through two series containment isolation valves. These valves are normal?y closed with the space between them pressurized by the pene-tration pressurization system. During purging operations, the valves will trip shut on a high radiation signal from l the containment particulate or gas monitor, or from the l 4.2-5
(fail safa operation) of the isolation valve. The air reservoir is sized to assure closure (fail safe operation) 4 of the isolation valve on failure of exhaust air pressure. Purge Exhaust Filters ! The containment purge exhaust air filters are composed of banks of prefilters and HEPA filters installed in series. The filters are arranged in two parallel independent modules and each module nas a rated flow of 20,000 cfm. The pre-filter bank in each module contains 21 individual prefilter elements having a nominal efficiency of 85% based on the NBS atmospheric dust spot test. The HEPA filter bank in each module contains 21 individual filter elements having a rated efficiency of 99.97% based on the D0P test. The arrangement of the two filter modules in parallel permits replacement of elements in the module without interruption of the purge exhaust ventilation. Purge Exhaust Fans The Containment Purge System exhaust air fans are of the direct-driven, vane axial type and are mounted on the outlet of the respective exhaust air filter and coil plenum. Each fan is rated at 40,000 cfm at a total pressure of 7.5" H2 0 and is driven by a nominal 75 hp motor. Two exhaust i air fans are provided for each containment unit, each of 100% design capacity with one fan as spare. 4 4.2-8 4 d
Purge Supply Preheating Coil The Containment Purge System supply air preheating coil is
- designed to heat 40,000 cfm of air from -10 F to 68 F0 when supplied with 67 gpm of water at 2700F. The total coil 6
capacity is 3.37 x 10 Btu /hr and consists of 3 finned tube sections supported and arranged for plenum mounting. ('arge Supply Fans The Containment Purge System supply air fans are of the
. direct-driven, vane axial type and are mounted on the outlet of the respective supply air filter and coil plenum. Each fan is rated at 40,000 cfm at a total pressure of 6.5" H O 2
and is driven by a nominal 60 hp motor. Two (2) supply air fans are provided for each containment unit, each of 100% design capacity with one fan as spare. Containment Isolation Valves Each containment purge supply air duct is equipped with i two isolation valves in series. An air reservoir is con- < nected to the respective air operator on each isolation valve. The air reservoir is sized to assure closure (fail safe operation) of the isolation valve on failure of supply air pressure. Each containment purge exhaust air duct is equipped with two isolation' valves in series. An air reservoir is con-nected to the respective air operator on each isolation valve. The air reservoir is sized to assure closure 4.2-7
is automatically routed through the purge supply system high efficiency filters and heating coil to assure a source of filtered and heated (in winter) air delivery to the containment. ' Containment samples must be analyzed and a release permit completed before any release to the environment is permitted. Containment Isolation Valves The 10 inch pressure and vacuum relief line contains two series, air operated isolation valves. An isolation valve air reservoir is connected to the air operator for each
! isolation valve. Each reservoir is sized to provide suf-ficient air to assure closure (fail-safe operation) of the isolation valve upon loss of operating air. Both valves trip shut upon receipt of a containment isolation signal.
Control Valves Two af,* operated control valves are provided to provide . either makeup air in the case of containment vacuum or a vent path in the event of a containment pressure. Both valves 7.re controlled by a containment pressure signal. The pressure relief valve begins to open when containment pressure exceeds 0.3 psig. The vacuum relief line begins to open at -0.1 psig. This -0.1 psig signal also I opens the outside air intake damper in the purge supply system to provide an air supply through the purge supply inlet filter and heating system. / 4.2-10 i
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i ) 4.2.2.2 . ressure and Vacuum Relief Svstem This system is provided to compensate for the normal contain-ment pressure changes resulting from containment temperature changes, barometric pressure changes, instrument air bleed, and inleakage from the penetration pressurization system. i . The Pressure and Vacuum Relief System is shown on Figures 4.2-1 and 4.2-2.
- This system consists of one 10" line penetration through the '
containment with fast-acting, double isolation, gate valves located outside the containment. These valves are normally closed and are designed to fail closed in the event of an incident or failure of control power. If the pressure in the containment is greater than +0.3 psig or less than -0.1 psig, an alarm in the main control room annunciates to alert , the operators. The two isolation valves are designed to be manually opened to allow the pressure to equalize with the outdoors. Control valves, which are operated by a contain-ment pressure signal, will automatically align the flowpath as follows: In the case of high containment pressure, the flow from the isolation valves is discharged to the venti-lation stack. In the case of vacuum in the contain-ment, the isolation valves will be opened to allow outside air to equalize the pressure. Outside air J 4.2-9 l l
The Hydrogen Purge System is required by Technical Speci-fications to be operable for reactor operation above cold
' shutdown.
4.2.2.3.2 Hydrogen Recombiner System I The Hydrogen Recombiner System is provided for essentially the same reason as the Hydrogen Purge System; however, with the recombiner system, there is no necessity for off-site l' release. l,.. i This system consists of two portable skid mounted electric recombiners, one acting as a backup for the other. In the event of a loss of coolant accident, one recombiner is connected outside the affected containment with no vent to the atmosphere (See Figure 4.2-2). The Reconbiner System is illustrated by the piping and instru-mentation diagram, Figure 4.2-3. The process gas is drawn from the containment into the recombiner by a blower with the flow fixed at 50 cfm by an orifice built into the inlet
- flange. The gas flows from the blower to the tube side of an economizer where it reaches a temperature of about 0
700 F. The gas then flows into an electric heater where it is heated to a preselected, automatically controlled temper-0 ature of 1100 to 1300 F. As the gas leaves the heater, it enters the reaction chamber where, at the temperatures I present, the hydrogen reacts with the oxygen in the process 4.2-12
i 4.2.2.3 Hydrogen Removal Systems
- 4. 2. 2. 3.1 Hydrogen Purge System The Hydrogen Purge System is designed to be utilized follow-ing a loss of coolant accident to purge hydrogen from the containment if it reaches 3% by volume.
The purge rate was selected to match the rate of production of hydrogen at the time of purge initiation so that the hydrogen concentration will slowly decrease as the produc-1 tion rate decreases. Operating in this manner allows a i lower purge rate thus. limiting the off-site dose. The Hydrogen Purge System is shown on Figures 4.2-1 and 4.2-2. The Hydrogen Purge System takes suction from the containment purge exhaust system between the HEPA filters and the con-tainment purge exhaust fans. Flow then passes through a prefilter, HEPA filter, and charcoal filter to the hydrogen purge fans. The hydrogen purge fans are 100% capacity , units rated at 350 cfm each. An interlock is provided to preclude running the hydrogen purge fans and the containment purge exhaust fans simultaneously. An alternate suction is provided from the containment pres-sure and vacuum relief .line.
-4.2-11
.. - - _ - - . -_- __ _ _ _ . . ~ _ - _
x v The main portion of this system consists of five air handling units (RCFC units) located in the space between the containment wall and the secondary shielding (crane support wall). Each unit draws air from the containment atmosphere utilizing a return air riser extending approxia mately 50 feet above the operating floor. Each unit discharges ventilation air inside the periphery of the secondary shield wall through concrete shafts designed for missile and radiation shield protection. The ventilation air is circulated first to reactor coolant pump and the steam generator area and then flows upward above the operating floor and is mixed with air in the upper containment atmosphere. i j Each RCFC unit is provided with the following components: -
- a. Demister or moisture separator
- b. High efficiency particulate air filters l c. Cooling coils
- d. Two speed fan The air flow inside each unit follows either one of two paths:
- a. Normal Operation: During normal reactor operation and after reactor safe shut down, the air is routed from the return air risers directly to the cooling coils bypassing the demister and the HEPA (high efficiency particulate air) filters, and to the fan suction plenum. The fan is operating at approximately 1200 )
rpm during normal operation. 4.2-14 l L
i l l l gas. The exothermic reaction raises the gas temperature in direct proportion to the hydrogen concentration. As the gas exiting froe the reaction chamber reaches 1300 F, the heater power is automatically reduced. When the hydrogen concentration reaches 3-1/2%, the heater exit temperature is reduced to about 900 F. If the reaction gas chamber temperature were to reach 1400 F, the heaters would shut off automatically until the overtemperature was corrected. From the reaction chamber, the gas travels through the economizer and back into the containment. At the rated flow, the recombiner has the following ratio of uncombined hydrogen in effluent to uncombined hydrogen in the feed: . 3% hydrogen in feed less than 1/30 2% hydrogen in feed less than 1/20 1% hydrogen in feed less than 1/10 4.2.2.4 Containment Fan Coolers The fan cooler units provide cooling so as to limit the air temperature in the closed containment to a maximum temperature of 120 F during all normal modes of operation 0 and a minimum of 65 F during shutdown conditions. The system also is designed to remove heat and particulate radioactivity from the containment as required following a loss-of-coolant accident. The flowpath for the Con-tainment Fan Cooler is shown on Figure 4.2-4. 4.2-13
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- c. All components are capable of withstanding differential pressures which may occur during the rapid pressure rise to 47 psig in ten (10) seconds.
- d. All components and their supports are designed to meet the requirements for Class I (seismic) structures.
- e. Each fan is provided with isolators to isolate the fan vibration from the other components.
4.2.2.5 Containment Activated Charcoal Filter Units This system is provided to permit cleanup of the contain-ment atmosphere prior to limited personnel access at power and prior to personnel access for refueling. The system consists of two charcoal filter units located on the refueling floor in the space between the contain-ment wall and the crane support wall equally spaced around the containment perimeter. Each unit is provided with the following components:
- a. High efficiency particulate air (HEPA) filters
- b. Activated charcoal filters
- c. Circulating fan All components and their supports meet the requirement for seismic Class II structures--. Each unit is capable of circulating 8,000 cfm in the normal containment atmosphere conditions . Operation of these two units for approximately
]
4.2-16
-. ., .- -.. _ -- .- - . ~ . - . - .-..
i
- b. Accident Operation: Immediately following a LOCA,
, ventilation air is automatically rerouted to flow from the return air risers to the demister, the high efficiency particulate. filters, cooling coils, and
! into the fan suction plenum. During accident oper-
, ation, the fan speed is automatically reduced to approximately 900 rpm. -
[ The reactor containment fan coolers are required to operate after a loss-of-coolant accident, and consequently they are designed to operate in a 47 psig containment pressure result-ing from the accident. In addition, every component of each unit is capable of withstanding, without impairing oper- , ability, a pressure of 1.5 times the design pressure and the associated temperature of the air-vapor mixture (298 F) for t a period of one hour. The following design criteria are conmon and applicable to filter assemblies, moisture eliminators, cooling coils, fans, RCFC unit housing and connecting ductwork for each of the 4 five air handling assemblies,
- a. Nonnal design air flow rate is 85,000 cfm/ unit and
, corresponding to accident operation the flow is reduced to 53,000 cfm/ unit.
- b. The normal maximum environment is a dry bulb temper-
- i. ature of 120 Uh and dew point temperature of 80UF.
The design maximum accident environment corresponds to a saturated steam air mixture of 2710 F at 47 psig, and density of 0.175 lb/cu ft. 4.2-15
- = _ . _. . .
4 inhaling air mixed with water vapor from the refueling water where the possibility of higher concentration of
- tritium exists in the immediate vicinity of the refueling water.
The design of this ventilation system directs a curtain of air downward over the operators standing on the manipulator crane bridge. 4.2.2.7 Other Containment Ventilation Systems Ventilation Systems are provided to cool various equipment inside the containment including: The control rod drive mechanisms. The reactor cavity and out-of-core nuclear instrumentation. 4.2.3 Summary The Containment Purge System is. designed to assure safe, continuous access (40 hours / week) to the containment within i three hours after a planned or unplanned reactor shutdown by rcducing the airborne particulates of the containment atmosphere. This system provides 40,000 cfm of filtered, heated as required, outside air which is delivered to the reactor i containment around the periphery of the reactor refueling pool. This purge rate provides approximately 1-1/2 air changes per hour and will permit safe access to the con-tainment three hours after a planned or unplanned shutdown.
- )
t 4.2-18 i
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32 hours will permit two hours access to the containment at full power under normal operating conditions. These units are not part of the engineered safeguards system and are not designed to operate after a loss-of-coolant accident. I The containment charcoal filter unit circulating fans are of the direct-driven, vane axial type and are mounted on the outlet of the respective containment activated charcoal filter housing. Each fan is rated at 8,000 cfm at a total i pressure of 2.5" H2O and is driven by a nominal 5 hp motor. Two fans are required for maximum clean-up with no spares. The containment activated charcoal filter units are composed of banks of HEPA filters and charcoal absorbers installed in i series. Each filter unit has a rated flow of 8,000 cfm. Each filter bank contains nine individual HEPA elements rated at 99.97% efficiency based on the 00P test. The charcoal absorbing bank has 24 drawer-type elements each nominally rated at 8,000 cfm and having a nominal efficiency of 99% on the removal of methyl iodine at 75 F and 90% R.H. Two filter units are required for maximum clean-up with no spares. 4.2.2.6 Manipulator Crane Ventilating System . This system is provided to minidize the risk of the oper-ators working on the manipulator (refueling) crane from l 4.2-17
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1 ! l of one 10" line penetration through the containment with two fast-acting, isolation gate valves located outside i the containment. These valves are normally closed and are designed to fail closed in the event of an incident or failure of control power. The two isolation valves are designed to be manually opened to allow the pressure to equalize with the outdoors if the pressure in the con-tainment becomes greater than +0.3 psig or less than -0.1 psig. Two air operated control valves, which are operated by a containment pressure signal, will automatically align the flowpath to provide either makeup air in the case of con-tainment vacuum or a vent path in the event of a contain-ment pressure. The Hydrogen Purge System is designed to be utilized following a loss of coolant accident to purge hydrogen j from the containment if it reaches 3% by volume. This system takes suction from the containment purge exhaust system between the HEPA filters and the containment purge i exhaust fans. Flow then passes through a prefilter, HEPA filter, and charcoal filter to the hydrogen purge fans. The hydrogen purge fans are 100% capacity units rated at 350 cfm each. An interlock is provided to preclude run-ning the hydrogen purge fans and the containment purge exhaust fans simultaneously. An alternate suction is provided from the containment pressure and vacuum relief line. . 4.2-20
--) ~ , - -,. . . , - . . .
i l l i Outside air enters the containment purge system through l an inlet damper, a prefilter, a heating coil, and is then routed to the suction of the purge supply fans. The two 100% capacity purge supply fans are each rated at 40,000 cfm. i The Purge Supply and Exhaust Systems are each provided with two containment isolation butterfly valves with one located inside and one located outside the containment. These valves are closed during plant operation. The butterfly l valves automatically close upon a high containment activity or high containment pressure signal. Closure of any of , the butterfly valves trips both the purge supply and exhaust fans. The purge exhaust exits th'e containment, passes through a pre-filter and a HEPA filter, and is routed to the suction of the purge exhaust fans. The two 100% capacity purge exhaust fans are each rated at 40,000 cfm. The purge exhaust flow is then routed to the plant vent stack for discharge to the environment. The Pressure and Vacuum Relief System is provided to com-pensate for the normal containment pressure changes result-ing from containment temperature changes, barometric pressure changes, instrument air bleed, and inleakage from the penetration pressurization system. This system consists i 4.2-19 '
k of the secondary shield wall. The ventilation air is cir-culated first to reactor coolant pump and the steam gen-erator areas and then flows upward above the operating floor and is mixed with air in the upper containment atmosphere. Each RCFC unit is provided with the following components:
- a. Demister or moisture separator
- b. High efficiency particulate air filters
- c. Cooling coils
/.
- d. Two speed fan t
l The Containment Activated Charcoal Filter Units System is provided to permit cleanup of the containment atmosphere prior to limited personnel access at power and prior to personnel access for refueling. i- , The system consists of two charcoal filter units located '
- on the refueling floor in the space between the contain-3 - ment wall and the crane support wall equally spaced around the containment perimeter. Each unit is provided with the following components
j a. High efficiency particulate air (HEPA) filters )
- b. Activated charcoal filters
- j. c. Circulating fan Operation of these two units for approximately 32 hours i
will permit two hours access to the' containment at full
),
4.2-22 l- _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ ___ ._____.____________..__.____._____.______m __ _____m-_____.____ _ _ _ _ _ . _ ___
l The Hydrogen Recombiner System is provided for essentially the same reason as the hydrogen purge system; however, with the recombiner system, there is no recessity for off-site release. This system consists of two portable skid mounted electric recombiners, one acting as a backup for the other. In the event of a loss of coolant accident, one recombiner is connected outside the affected contain-ment with no vent to the atmosphere. The process gas is drawn from the containment into the recombiner by a blower with the flow fixed at 50 cfm by an orifice built into the inlet flange. The gas flows through the recombiner, where hydrogen is removed, t'ien directly back into containment. The containment fan cooler units provide cooling so as to limit the air temperature in the closed containment to a maximum temperature of 120 F during all normal modes of operation and a minimum of 65 F during shutdown conditions. The system also is designed to remove heat and particulate radioactivity from the containnent as required following a loss-of-coolant accident. The main portion of this system consists of five air handling units (RCFC units) located in the space between the containment wall and the secondary shielding (crane support wall). Each unit draws air f rom the containment atmosphere utilizing a return air riser extending approximately 50 feet above the oper-ating floor and discharges the air inside the periphery 4.2-21
I f power under normal operating conditions. These units are not part of the engineered safeguards system and are not designed to operate after a loss-of-coolant accident. The Manipulator Crane Ventilating System is provided to minimize the risk of the operators working on the manipulator (refueling) crane from inhaling air mixed with water vapor from the refueling water where the possibility of higher concentration of tritium exists in the immediate vicinity of the refueling water. The design of this ventilation system directs a curtain of air downward over the operators standing on the manipulator crane bridge. Other Containment Ventilation Systems are provided to cool various equipment inside the containment including: The control rod drive ns.hanisms The reactor cavity and out-of-core nuclear instrumentation. 4.2-23
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THERMAL RECOMBINER PIPING AND INSTRUMENTATION FLOW DIAGRAM LEGEND A = ALARM (ENUNCI ATOR) C = CONTROL E = VOLTAGE MM Ts THA H= 5 = HIGH INDICATE h 4 4
~ . L= LOW T M ~~~~--~~
N S = SWITCH T = TEMPE R ATURE (THE RMOCOUPLE) 5 (4,j (4) X= HE ATER CUTOFF v = AUCTIONuR iSEuc10R, 15 { p RE ACTION CHAMUE R l RETURN 9 yw I i I m ' rh m 1. w -- @1 > c q w w-
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$$$ i 8 LOWER E CONOMl2E R L _ _ _ _ __ __ _ __ _ _ _ _ _ _ _ _ m _ _ l 8 LOWE R "ON" IN TE R LOCK FIGURE 4.2-3
REACTOR CONTAINMENT FAN COOLER o 1
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l access during inspection, testing, maintenance, and refueling l operations, and to limit the release of radioactivity to the environment. l The purpose of the Vacu'um Relief System is u, protect the vessel from an excessive external force. The Vacuum Relief System does not serve accident mitigating functions or serve to limit the spread of radioactivity. It is a self-activated system that limits external pressure in the event of maloper-ation or inadvertent operation that results in additional external forces on the containment vessel. The primary purpose of the Air Return Fan System is to enhance the ice condenser and containment spray, heat removal operation by circulating air from the upper containment to the lower compartment, through the ice condenser, and then back to the upper compartment. The operation will take place at the appropriate time following the accident. The secondary purpose of the systemis to limit hydrogen concentration in potentially stagnant regions by ensuring a flow of air from these regions. Electric hydrogen recombiners are provided to remove hydrogen from the containment atmosphere and prevent the formation of combustible gas mixtures. The functional arrangement of these systems is depicted on Figure 4.3-1. l ) 4.3-2 l
l 1 l 7
. i 4.3 DUAL CONTAINMENT VENTILATION SYSTEMS i
4.3.1 - Introduction , l : The Containment Ventilation System is comprised of several l 4 t subsystems each of which has separate design objectives. I The systems are designed around the objective of dealing l 1 ' i with normal operation, personnel entry, containment pro-tection, and accident conditions. The following subsystems , are included in the total ventilation system: ! J . l 1. Containment Air Cooling Systems ' t i
- 2. Containment Purge System l!:
- 3. Containment Vacuum Relief System i
- 4. Containment Air Return System l
] ) 5. Containment Combustible Gas Control System , i i i t
- The Containment Air Cooling Systems are designed to maintain
l ] an acceptable temperature within the containment upper and i
- lower compartments, reactor well, control rod drive mechanism i i j (CRDM) shroud, and instrument room for the protection of equipment and controls during nonnal reactor operation and L t
I normal shutdown. The instrument room is mechanically cooled i to permit personnel access during nonnal reactor operations. The Containment Purge System is designed to maintain the environment in the primary and secondary containment within i acceptable limits for equipment operation and for personnel
- l r
i 4.3-1 I l
The four lower compartment air cooling fan-coil assemblies are located in two annular concrete chambers around the ; periphery of the lower compartment. Each fan-coil assembly consists of plenum, eight air cooling coils, vaneaxial fan, assembly isolating backdraft damper, instruments, and con-trols. Each fan-coil assembly is designed to cool 65,000 < cfm of 1200F air to 90 F0 or lower when supplied with 200 gpm of 83 0F water from the plant Essential Raw Cooling Water
! (ERCW) System. A cooling water throttling valve for each assembly is automatically controlled by a temperature indi-cating controller which utilizes an input from a thermo-couple mounted in the assembly's return air supply and set j to control the containment air temperature at 120 F.
1 Lower compartment air passes directly to each active fan-coil assembly where it is cooled and supplied through a comon duct distribution systen to the lower compartment spaces. The system is designed for three of the four fan-coil assemblies to operate together, with one on standby,
- for a. total of 195,000 cfm of air and 6.078,700 BTU per hour of cooling. The cooled air is supplied directly to each steam-generator compartment, pressurizer compartment, letdown heat exchanger room, main lower compartment space. I and to the space below the reactor vessel.
Control Rod Drive Mechanisms Air Cooling System r During nonnal reactor operation, the CROM Air Cooling System is designed to operate in conjunction with the lower compart-l ment air cooling system to maintain a maximum air temperature i 4,3-4
. _ _ _ - __ - _ _ . _ . _ _ ___..._____ _ __. _ _ . _ .m . - . __ _
4 4.3.2 System Design and Operation i ! 4.3.2.1 Containment Air Cooling Systems 1 ! The Containment Air Cooling Systems are shown in Figure i i 1 4.3-1. It consists of four subsystems as follows:
- 1. Lower Compartment Air Cooling -
l 2. Control Rod Drive Mechanisms (CRDM) Air Cooling 1 1 3. Upper Compartment Air Cooling
- 4. Reactor Building Instrument Room Air Cooling I
l Lower Compartment Air Cooling System j The Lower Compartment Air Cooling System, together with operation of the CRDM air cooling system, is designed to supply air at a maximum temperature of 900 F to maintain a : i maximum air temperature of 120 F in the following lower j . l compartment spaces during normal reactor operation. These i spaces include the steam generator and pressurizer compart-ments, the space below the reactor vessel, the generator and - 1 r ! the pressurizer compartments, the space below the reactor l i i ! vessel, the space around the reactor vessel, the reactor l
\
I vessel nozzle and support openings, and the reactor well i l space around the CRDM shroud. Four 33.34 percent capacity } fan-coil assemblies are provided to allow three or less to i operate during reactor nonnal operation with one or more on standby. i i 4 t i 4.3-3
damper, instruments, and controls. Each fan-coil assembly 0 is designed to cool 35,000 cfm of 164 F air to 120 F or I lower when supplied with 84 gpm of 830 F water from the plant Essential Raw Cooling Water System. A cooling water throttling valve for each assembly is automatically con-trolled by a temperature indicating controller which utilizes an input from the thermocouple mounted in the cooled air discharge to the lower compartment, and set to control the air outlet temperature at 120 F.0 The four CRDM air cooling fan-coil assemblies are divided 4 into two pairs with either one of each pair, for a total j of two, designed to operate together to exhaust a total of 70,000 cfm of air from the CRDM shroud during nomal l reactor operation. Reactor well air enters the shroud , at a maximum temperature of ll70F and is exhausted from the shroud at a maximum temperature of 1643. The air is cooled to 120 F or lower by the fan-coil assemblies and-is discharged to the lower compartment spaces. A total l of 3,080,000 BTU per hour of cooling capacity is provided. The lower compartments and control rod drive mechanism air :
- cooling fan-coil assemblies are energized from the Emergency Power System; however, they are not required to operate during LOCA conditions.
I } l 4.3-6
. ._. ~ ~ . . .- -- . - - -
within the reactor well of 117 F, and to route a portion of the reactor well air through the CRDM shroud to maintain a ' maximum air temperature of 164UF. The CRDM Air Cooling System consists of four 50 percent capacity fan-coil assemblies combined into two 50 percent capacity subsystems. Of these four fan-coil assemblies, l one for each subsystem is normally operated for a total of i two, and one for each system normally remains on standby for a total of two. Air drawn through the CRDM shroud is cooled by the active fan-coil assemblies to 120 F and dis-charged into the lower compartment, Lower Compartment Air Cooling System manual dampers are l j , adjusted to provide sufficient air flow through the reactor well to maintain a maximum air temperature of 117 F. ! Upon the requirement for additional cooling in the lower compartment, an arrangement of dampers will allow either j or both standby control rod drive mechanism fan-coil assem-blies to operate to recirculate cool an additional portion , of the' lower compartment air. The four control rod drive mechanisms' air cooling fan-coil i assemblies are located in the main lower compartment space, f 4 L Each fan-coil assembly consists of plenum, two air cooling coils, vaneaxial fan, assembly isolating motor-operated i l l l 4.3-5
Containment Instrument Room Air Cooling System The containment instrument room is cooled during normal i reactor operation or shutdown by either of two 100 percent capacity air-conditioning systems. Each system is designed to automatically maintain the room air temperature at a 0 maximum of 75 F 08 and 62.5 F WB. Each system consists of a fan-coil unit located within the instrument room, a water-chilling condensing unit and chilled water pumo located in the auxiliary building, and the connecting chilled water piping including containment penetration valves. 1 Each water chilling unit is rated at 10.4 tons of refriger-ation at a condensing temperature of 105 F and a suction 0 l temperature of 40 F when supplied with 30 gpm of condenser cooling water at 83 F. The unit compressor is driven by a 10-hp motor and the unit is designed to cool not less than 24 gpm of water from 580 F to 48 F. The fan-coil unit is designed to supply not less than 6200 i cfm of air at an external static pressure of 0.25 inch of water and is driven by a 5-hp motor. The coil is designed i 0 to cool 6200 cfm of air at 75 F DB and 62.5 F WB to 57 F DB and 55.5 F WB when supplied with 24 gpm of water of 80F. The chilled water pump is driven by a 1.5-hp motor. 1 The chilled water penetrations through containment are each provided with two isolation ball valves with one located 4 a 4.3-8 i l 1
- , . , . . , . - -. . - - . , . . - - _ , _ . . - - , - _ , . ~ _ , . - - - . , . - . . _ , . . - . - , - . ,
4
- Upper Compartment Air Cooling System
.The Upper Compartment Air Cooling System is designed to 0
supply air at a maximum temperature of 98 F air to maintain the upper compartment at a maximum temperature of 110UF f during nonnal reactor operation. Four 33.34 percent capa- , city fan-coil assemblies are provided to allow three or less to operate with one or more on standby during normal reactor operation. j j The four upper compartment air cooling fan-coil assemblies are located within the upper compartment at elevation 778.69. Each fan-coil assembly consists of plenum, three air cooling ] coils, vaneaxial fan, instruments, and controls. Each fan- ] coil assembly is designed to cool 16,000 cfm of 110 F air 0 to 97 F or lower when supplied with 23 gpm of 83 F water from j the plant Essential Raw Cooling Water System. A cooling j water throttling valve for each assembly is automatically ' controlled by a temperature indicating controller which I utilizes an input from the thermocouple mounted in the h i return air supply and set to control the containment air l temperature at 110 F. A portion of the upper compartment air is continuously recir-culated and cooled by the upper compartment fan-coil assem-blies. The system is designed for three of the four assem- I j blies to operate together, with one on standby, for a total j of 48,000 cfm of air and 576,000 BTU per hour of cooling { capacity, i ,
- 4.3-7 i ;
. _ . . , , , . ,s-,_ , .m.- , ~,,..-.-_-.._..___.._,__m.-_,--.._-,,_.:~y_,._.,,-.,...-_,.r,-___.- , , - , . . - . , . , _ . . _
.I r i The containment upper and/or lower compartments, are purged with fresh air by the reactor building purge system before - occupancy. The annulus can be purged with fresh air during reactor shutdown or at times when the annulus vacuum con-trol system of the emergency gas treatment system is shut down. The instrument room is purged with fresh air during ) operation of the reactor building purge system or can be l separately purged by the instrument room purge subsystem. i j Each purge system consists of two 50 percent capacity air i j supply fans, two 50 percent capacity air exhaust fans, two , 50 percent capacity cleanup filter trains, instrument room supply fan, instrument room exhaust fan, air supply distri-bution system, air exhaust collection sys' tem, and containment isolation valves, system airflow control valves, and con- , tamination control leakoff valves. I i 4 l i J i The purge air supply fans are located in the penetration ! room at El. 714. Filtered fresh air, heated when required, l is taken from the auxiliary building air supply systems located in the mechanical equipment rooms at El.714. These fans are designed to supply a total of 28,000 cfm. The purge air exhaust fans and air cleanup filter trains are located within the penetration room at El. 690. The_ cleaned air is discharged to the outdoors by means of the shield 4 1 I 4.3-10
inside and one located outside containment for each pene- i tration. These 2-inch valves are pneumatic motor operated 3 and designed to fail closed. ! 4.3.2.2 Containment Purge System j The Containment Purge System is shown schematically in Figure 4.3-1. One complete and independent Purge System is provided
; for each unit.
i l The Containment Purge System provides for mechanical ventila-
- tion of the primary containment, the instrument room located '
I within tne containment, and the annulus or secondary con-i tainment located between the containment and shield building. I
- The system is designed to supply fresh air for breathing, and contamination control to allow personnel access for j maintenance and refueling operations. The exhaust air is filtered to limit the release of radioactivity to the environ- ,
j ment. f j During power operation, cooling of the reactor building , upper compartment, lower compartment, and control rod drive i . ! mechanism is accomplished by the air cooling systems. The i 1 annulus normally maintained at a negative pressure by the l annulus vacuum control subsystem of the emergency gas I treatment system, i i i l l l l i j r . 4.3-9
- , ~ _ _ . . . . . . . ~ . - _ . _ , , _ _ _.
_ - - - .- . . . - =. .- . . _-~ - -. - . - . i l The systen air supply and exhaust ducts are routes through j the secondary containment to several primary containment l 1 penetrations . Two air supply locations are provided for each of the upper and lower compartments and one for the 4 instrument room. Several air pickup points are located to !, exhaust air from the lower compartment and instrument room and to provide an air sweep across the surface at the
- refueling canal, j
1 ; l Annulus purging air is taken from system ducts which is j routed through the annulus. These air supply and exhaust j duct openings are, located approximately 180 degrees apart I for maximum ventilation. i The primary containment penetrations for the ventilation l supply and exhaust subsystems are designed to primary . containment integrity requirements. 1 l j Each purge system containment penetration is provided with
! both inboard and outboard motor-operated isolation butterfly
! valves designed for minimum leakage in their closed position.
- A similar type of valve is mounted in each purge supply and exhaust air opening for the annulus, and in each of the systemt main supply and exhaust duct located exterior to the i j shield building. Each of the above butterfly valves is designed to fail closed and to normally close during purge j system shutdown. ,
i j. l i 4.3-12
-. . _. =. . . -
building and extending through the roof of the reactor building. These fans are designed to exhaust a total of 28,000 cfm. The purge air supply fans are centrifugal type, each rated at 14,000 cfm against 9.25-inch water gauge static pressure and belt driven by a 30-hp motor. The purge air exhaust fans are centrifugal type, each rated at 14,000 cfm against 10.50-inch water gauge static pressure and belt driven by a 50-hp motor. These fans are not connected to emergency 1 power. The supply fans, exhaust fans, and air cleanup filter assers blies for each unit are connected and controlled in two 50 .. percent capacity trains. The controls are designed to have simultaneous starting and stopping of the matching supply and exhaust equipment. The controls are also designed to give an automatic shutdown and isolation upon receipt of ! containment ventilation isolation signal. Upon failure of , a fan, system redundancy will be 50 percent capacity. I Each air cleanup filter plenum contains a bank of prefilters, a bank of HEPA filters, and a bank of charcoal adsorbers. Each plenum is provided with static pressure differential indicators, thermometers, connection for inplace testing of filters and adsorbers, and access doors for filtef and adsorber maintenance. 4 4.3-11
4.3.2.3 Containment Vacuum Re_ lief System The containment vessel is fitted with a Vacuum Relief System. The purpose of the Vacuum Relief System is to protect the vessel from an excessive external force. The Vacuum Relief System does not serve accident mitigating functions or serve to limit the spread of radioactivity. It is a self-activated system that limits external pressure in the event of maloperation or inadve.rtent operation that results in additional external forces on the containment. The Containment Vessel Vacuum Relief System assures that the external pressure differential of the containment vessel does not exceed the design external pressure of 0.5 psig. When the external pressure exceeds the valve set pressure, air flows from the annulus space through the containment vacuum relief units into the containment vessel. The operation of the Vacuum Relief System results in a pres-sure reduction in the annulus space between the containment vessel and the shield building. The shield building is designed to withstand this pressure reduction in the annulus. The Vacuum Relief System is not required to mitigate acci-dents such as a LOCA. Rather, it is a system designed to protect the containment vessel in the event of excessive cooling and subsequent external pressure on the containment vessel. The system is designed to mitigate the following abnormal occurrences:
)
4.3-14
To permit personnel a'ccess to the instrument room during reactor operation or during purge system shutdown, the room can be purged by the instrument room purge subsystem fans. These subsystem supply and exhaust fans are located , alongside the main system supply and exhaust fans and will use the main system ducts and filter train. Approximately 900 cfm is mechanically supplied and 800 cfm is mechanically exhausted, and butterfly valves are positioned to allow only the instrument room to be served. The instrument room purge air supply and exhaust fans are centrifugal type. Each supply fan is rated at 900 cfm against 1.75-inch water gauge static pressure and belt driven by 0.75-hp motor. Each exhaust fan is rated at 800 cfm against 1.25-inch water gauge static press'ure and belt-driven by a 0.50-hp motor. The containment purge system is an engineered safety feature. All penetration isolation valves and piping between these valves have a Nuclear Safety Class designation in accordance with ANS Safety Class 2A. Other portions of the exhaust systerr are designated ANS Safety Class 2B. The instrument room purge system is not an engineered safety feature and credit for LOCA mitigation is not claimed. Containment ventilation isolation signals automatically shut down the fan systems and isolate the purge systems by closing their respective dampers and butterfly valves. 4.3-13
Each containment vessel vacuum relief valve is a 24-inch, self-actuated, horizontally installed, swing-disc valve, with an elastomer seat. The seat material will withstand post-LOCA temperature, pressure and radiation conditionc. Each unit has a design airflow rate of 28 pounds per second at a pressure differential of 0.5 psig across the entire unit. Each nonnally closed vacuum relief valve is equipped with limit switches so that the fully open and fully closed positions of the valves are indicated in the main control room. The opening of any of these valves is annunciated in the main control room. The valves begin opening at a contain:nent external pressure differential of 0.1 psig and will be fully open in less than 3 seconds for a vacuum relief valve design basis event. Each containment vessel vacuum relief isolation valve is a pneumatically operated butterfly valve with an elastomer seat. The valve, including seat material, will withstand the post-LOCA temperature, pressure, and radiation conditions. Two separate trains of control air supplies are available to the two independent solenoid valves which power the isolation valve. The isolation valve, which is normally open, fails open, and will close when containment high pressure signal is developed from either of two independent sets of three pressure sensors and is completely '<idependent of other
]
4.3-16
)
l
- 1. Inadvertent Containment Spray Actuation Operation
- 2. Inadvertent Containment Air Return System Operation
- 3. Simultaneous Occurrence of both of the above.
Other abnonnal occurrences such as heating and ventilation equipment malfunction may result in external pressures. However, the effect is always less than for those occorrences listed above. The design bases, parameters, and resultant design for the Vacuum Relief System are summarized in Table 4.3-1. The Vacuum Relief System is designed in accordance with the containment system general design criteria. The system is designed to withstand the safe shutdown earthquake without failure. The Containment Vessel Vacuum Relief System has three iden-tical units, all located on the dome, at the same elevation, and 120 degrees apart. One of the three units is redundant. In essence, each unit contains a vacuum relief valve in series with a containment isolation valve, the vacuum relief valve being outside of the isolation valve, as shown in Figure 4.3-1. The units are installed such that there is sufficient space between the Vacuun Relief System and the shield building to prevent contact during seismic or pressure transient motion and to allow for an adequate airflow path. l l 4.3-15
flow into the upper compartment through doors at the top of the ice condenser. Each main duct contains a non-return damper which prevents flow from the lower compartments to the upper compartment during the initial stages of a lon-of-coolant accident. Both fans will start 10 minutes after receipt of a phase B isolation signal. In addition, either fan may be controlled manually from the control room. Each fan can develop suf-ficient head to keep the non-return dampers and ice con-denser inlet doors open after blowdown is complete. A flow indicator upstre6m of each fan and a pressure differential indicator across each fan are provided. In addition, the discharge flow rate is indicated in the control room. The design life of the air return system under normal (standby) conditions is 40 years. Under normal conditions the components of the system will withs' 'nd a temperature of 120 F, a rela-tive humidity of 100 percent, and a radiation dose of 2 rads per hour without loss of ability to function. Materials of the system are essentially steel, coated to prevent corrosion. Gaskets are elastomer-coated fabric. During accident conditions, the system is capable of oper-ating continuously with temperatures ranging up to 250 F for the first hour, and at 170 F Uand 100 percent relative humidity for a year, with a total radiation dose of 108 rads.
/
4.3-18
containment isolation signals for other systems. E6ch isolation valve is equipped with a limit switch so that the fully open and fully closed positions are indicated in the main control room. If the valve is not in the fully open position, an alarm is annunciated in the main control room.
- 4. 3. 2.4 Containment Air Return Systems The primary purpose of the Air Return Fan System is to enhance the ice condenser and containment spray heat removal operation by circulating air from the upper compartment to the lower compartment, through the ice condenser, and then back to the upper compartment. The operation will take place at the appropriate time following the accident. The secondary purpose of the system is to limit hydrogen concentrati;n in potentially stagnant regions by ensuring a flow of air from these regions.
There are two 100 percent capacity air return fans. Each will remove air at the rate of 40,000 cfm from the upper compartment.through a main duct to an accumulator room of the lower compartment (See Figure 4.3-2). The discharged air will flow from each accumulator room through the annular equipment areas into the lower compartment. Any steam produced by residual heat will mix with the air and flow through the lower inlet doors of the ice condenser. The steam portion of the mixture will condense as long as ice remains in the ice condenser and the air will continue to l l 4.3-17 1 1 1
I Simultaneously with the return of air from the upper compart-ment to the lower compartment, post-LOCA hydrogen reixing capability is provided by the Air Return Fan System in the following regions of the containment: containment dome, each of the four accumulator' rooms, and the instrument room. These regions are served by sever hydrogen collection headers which tenninate on the suction side of either of the two air return fans. The minimum design flow from each region is sufficient to limit the local oncentration of hydrogen to not more than 3 percent when the containment average is 2 percent. Minimum detign flow rates are shown in Figure 4.3-2. The header systems will be adjusted prior to initial plant operation to assure that the actual flows are at least equal to the minimum design flow when only one fan is in oper-ation. 4.3.2.5 Containment Combustible Gas Control Syste_m Following a loss-of-coolant accident, hydrogen may accumulate within the containment as a result of: l
- 1. Zirconium-water reaction involving the fuel cladding and the reactor coolant.
- 2. Radiolytic decomposition of water in the core.
- 3. Radiolytic decomposition of water in the containment sump.
]
l 4.3 20
The Air Return Fan System is an engineered safety feature and meets the qualification requirements of the seismic. Category I. The main duct through the divider deck between the upper and lower compartments including the non-return
~
dampers meet the requirements of TVA Class B (ANSI Safety Class 2A). The remainder of the system meets the require-ments of TVA Class C (ANSI Safety Clast. 28). The design of the fans and controls of each 100 percent capacity system meets the intent of Regulatory Guides 1.29
! and 1.53.
i Each air return fan is direct drive, vaneaxial, with a capacity of not less than 40,000 cfm against a static pres-sure of 5 inches water gage; each is driven by a 460-volt, 3-phase electric motor which develops 50 horsepower at 1,170 rpm. The non-return dampers are heavy duty and are designed to prevent flow from the low compartment to the upper com-partment under a differential pressure of 12 psi. The dampers are controlled to open when the differential pressure across the operating fan assures flow from the upper to lower compartment. The gravity-loaded damper will fail in the closed position upon loss of necessary flow head, and has a lerkage at 12 psi differential pressure of not more than 2 square inches. The position of the damper is moni-tored in the control room. 4.3-19
--y . = - - , . . . . ,. -,e.e,- , --- , -- -- - - -.
for each recombiner unit, located outside the containment in an area that is accessible following a loss-of-coolant accident. The recombiner systems are completely redundant. Figure 4.3-3 is a sketch of the recombiner unit. The recom-biner unit consists of a preheater section, a heater-recom-bination section, and an exhaust section. Containment air is drawn into the unit by natural convection, passing first through the preheater section. This section consists of the annular space between the heater-recombination i
- i section duct and the external housing. The temperature of l
the incoming air is increased by heat transferred from the I heater-recombination section. This results in a reduction of heat losses from the unit. The preheated air passes through an orifice plate and enters the heater-recombination section. This section consists of a thermally insulated vertical metal duct enclosing five assemblies of metal-sheathed electrical heaters. Each heater assembly contains individual heating elements, and the operation of the unit is virtually unaffec-ted by the failure of a few individual heating elements. The incoming air is heated to a temperature in the range of 1150 to 14000F, where recombination of hydrogen and oxygen occurs. Finally, the air from the heater-recombination section enters the exhaust section where it is mixed with cooler containment air and discharged from the unit.
/
4.3-22
i l l
- 4. Corrosion of materials within the containment.
Based on the AEC-TID release model, as described in NRC Regulatory Guide 1.7, each recombiner has the capacity to maintain the concentration of hydrogen in the containment
- atmosphere at less than 4 volume percent following a loss-of-coolant accident.
The Containment Air Purification and Cleanup System, des-cribed in Section 4.4, provides the capability of a con-trolled purge of the containment and functions as a backup for the Combustible Gas Control System. The Combustible Gas Control System is designed to sustain all norum1 loads as well as accident loads including seismic loads and temperature and pressure transients from a loss-of-coolant accident. Mixing of containment atmosphere following a loss-of-coolant accident is provided by the Air Return Fan System. A redundant hydrogen sampling system is designed to detect and give MCR indication of the presence and concentration of I hydrogen in the primary containment atmosphere subsequent to a LOCA. 4 The Combustible Gas Control System consists of two electric hydrogen recombiner units, located in the upper containment < compartment, and separate control panels and power supplies 4.3-21
e ._ . The basic design parameters for the electric hydrogen recombiners are given in Table 4.3-2. The post accident purge system required by Regulatory Guide 1.7, for use as a backup to the redundant hydrogen recom-biner system is functionally designed to exhaust containment air into the annulus and reple11sh this discharge with a dilution air supply. The hydrogen purge exhaust subsystem consists of a single penetration in the primary containment wall equipped with two normally closed, remote manually operated isolation valves, one on either side of the containment wall; one pneumatically operated annulus purge exhaust valve located within the annulus; and two 1/2-inch leakoff nipples located 4 between the outboard isolation valve and the annulus purge j exhaust valve. With the containment isolation valves open, 1 and the annulus purge exhaust valve closed, a flow path is established from the primary containment through the leak-r offs and into the annulus, which will permit purging of the containment for hydrogen control subsequent to a LOCA. The impetus for flow will be provided by the differential pressure between the primary containment and annulus. If the concentration cannot be maintained below 4% through the leakoff path, the annulus ; urge valve will be opened to supply dilutant air for minimum time sufficient to maintain i 4.3-24 r c .y,,, _ , , , --,-. - - n
Tests have verified that the recombination of hydrogen and oxygen in the unit is not the result of a catalytic surface effect but occurs as a result of the increased temperature of the process gases. The performance of the unit is therefore unaffected by fission products or other impurities which might poison a catalyst. The power panel for each recombiner unit is located outside the reactor containment in an area accessible after a loss-of-coolant accident. The panel contains an isolation transformer plus a controller to regulate power input to { the recombiner. For equipment test and periodic checkout, a thermocouple readout instrument is also provided in the control panel for monitoring temperatures in the recombiner. To control the recombination process, the correct power input to bring the recombiner above the threshold temper-ature for recombination is set on the controller. The con-troller setting is accomplished at the control panel, and power input is monitored by a wattmeter. This predeter-mined power setting covers variations in containment tem-perature, pressure, and hydrogen concentration in the post- , loss-of-coolant accident environment. Results of testing a prototype of the electric hydrogen recombiner are given in WCAP-7709-L and WCAP-7820,
" Electrical Hydrogen Recombiner for Water Reactor Contain-ments," dated liay,1972 and December,1972 respectively.
l l I l 4.3-23
. . _ . ~ , , , - - . - . ,, , . . . . . - - m. -,~ ..,, , , ,. , , c, - - , -
4.3.3 Summary Description The following subsystems are included in the total ventila-tion system:
- 1. Containment Air Cooling Systems
- 2. Containment Purge System
- 3. Containment Vacuum Relief System l
- 4. Containment Air Return System
- 5. Containment Combustible Gas Control System The Containment Air Cooling Systems are shown in Figure 4.3-1. It consists of four subsystems as follows:
- 1. Lower Compartment Air Cooling
- 2. Control Rod Drive Fechanisms (CRDM) Air Cooling
- 3. Upper Compartment Air Cool.ing
- 4. Reactor Building Instrument Room Air Cooling The Lower Compartmen+. Air Cooling System, together with oper-ation of the CRDM Air Cooling System, is designed to supply air at a maximum temperature of 90U F to maintain air temper-ature of 120 F in the following lower compartment spaces during normal reactor operation. These spaces include the steam generator and pressurizer compartments, the space ,
below the reactor vessel, the space around the reactor vessel, the reactor vessel nozzle, and support openings, and the reactor well space around the CRDM shroud. Four 33.34 percent capacity fan-coil assemblies are provided to allow three or less to operate during reactor normal operation with one or more on standby. / 4.3-26
t the hydrogen concentration below 4%. The containment effluent purged for hydrogen will flow directly to the annulus where it will mix with the annulus atmosphere and be filtered by the air cleanup system prior to dis-charge to the outside environment. The calculated radio-4 logical consequences of the LOCA, including the hydrogen purge, will not exceed the guidelines of 10 CFR Part 100 under the most severe condition in which the annulus purge valve is opened at a pressure differential of 1.0 psid. Redundancy is not required for this system since it is a backup system to the redundant hydrogen recombiners. The dilution air flow is introduced into the containment from the service air system. This service air system has provisions enabling it to receive diesel power. The dilution supply subsystem consists of a single 2-inch penetration in the primary containment wall equipped with provisions for containment isolation. The inboard containment isolation feature is a check valve located in the primary containment. The outboard containment isolation feature is a double 0-ring sealed flange located in the auxiliary building. A pressure hose will be required to provide a flow path from a service 2 air flange. Operation of this subsystem is accomplished by removing the flange and coupling a service air hose to the pipe penetration. The system will be sized to provide 60 scfm of dilution air at a service air pressure of 60 psig. 4 1 4.3-25
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_ - _ . - _ _ = _ _ _ l of a fan-coil unit located within the instrument room, a ' water-chilling condensing unit and chilled water pump located in the auxiliary building, and the connecting chilled water piping including containment penetration valves. The Containment Purge System is shown schematically in Figure 4.3-1. One complete and independent Containment Purge System is provided for each unit. The Containment Purge System provides for mechanical ventila-tion of the primary containment, the instrument room located within the containment, and the annulus or secondary con-i tainment located between the containment and shield building. The system is designed to supply fresh air for breathing, ' and contamination control to allow personnel access for main-tenance and refueling operations. The exhaust air is filtered to limit the release of radioactivity to the environment. Each purge system consists of two 50 percent capacity air supply fans, two 50 percent capacity air exhaust fans, two . 50 percent capacity cleanup filter trains, instrument room supply fan, instrument room exhaust fan, air supply distri-bution system, air exhaust collection system, containment isolation valves, system airflow control valves, and con-tamination control leakoff valves. The containment vessel is fitted with a Vacuum Relief System. The purpose of the Vacuum Relief System is to protect the vessel from an excessive external force. The Vacuum Relief / l 4.3-28 l
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During nonnal reactor operation, the CRDM Air Cooling System is designed to operate in conjunction with the lower compart-ment air cooling system to maintain a maximum air temperature 0 within the reactor well of ll7 F, and to route a portion of the reactor well air through the CRDM shroud to maintain a maximum temperature of 164 F. The CRDM Ai: Coo?ing System consists of four 50 percent capacity fan-coil assemblies combined into two 50 percent capac-ity subsystems. Of these four fan-co.il assemblies, one for each subsystem is normally operated for a total of two, and one for each system normally remains on standby for a total of two. Air drawn through the CRDM shroud is cooled by the i ac'ive fan-coil assemblies to 120 F and discharged into the lower compartQnt. The Upper Compartment Air Cooling System is designed to 0 supply air at a maximum temperature of 98 F air to maintain the upper compartment at a maximum temperature of 1100F during normal reactor operation. Four 33.34 percent capa-city fan-coil assemblies are provided to allow three or less to operate with one or mo;a on standby during normal reactor operation. The reactor building instrument room is cooled during normal reactor operation or shutdown by either of .+? 100 percent' capacity air-conditioning systems. Each system is designed to automatically maintain the room air temperature at a maximum of 75 0 F DB and 62.50 F WB. Each system consists 4.3-27 1
There are two 100 percent capacity air return fans. Each will remove air at the rate of 40,000 cfm from the upper compartment through a main duct to an accumulator room of the lower compartment (See Figure 4.3-2). The discharged air will flow from each accumulator room through the annular equipment areas into the lower compartment. Any steam produced by residual heat will mix with the air and flow through the lower inlet doors of the ice condenser. The steam portion of the mixture will condense as long as ice remains in the ice condenser and the air will continue to flow into the upper compartment through doors at the top of the ice condenser. Each main duct contains a non-return damper which prevents flow from the lower compartments to the upper compartment during the initial stages of a loss-of-coolant accident. Both fans will start 10 minutes after receipt of a phase B isolation signal. In addition, either fan may be controlled manually from the control room. Each fan can develop suf-ficient head to keep the non'-return dampers and ice condenser inlet doors open after blowdown is complete. Following a loss-of-coolant accident, hydrogen may accumulate within the containment as a result of:
- l. Zirconium-water reaction involving the fuel cladding and the reactor coolant.
)
4.3-30
System does not serve accident mitigating functions or serve to limit the spread of radioactivity. It is a self-activated system that limits external pressure in the event of maloper-ation or inadvertent operation that results in additional external forces on the containment vessel. The design bases, parameters, and resultant design for the Vacuum Relief System are summarized in Table 4.3-1. The Containment Vessel Vacuum Relief System has three identical units, all located on the dome, at the same elevation, and 120 degrees apart. One of the three units is redundant. In essence, each unit contains a vacuum relief valve in series with a containment isolation valve, the vacuum relief valve being outside of the isolation valve, as shown in Figure 4.3-1. The primary purpose of the Containment Air Return Fan System is to enhance the ice condenser and containment spray heat removal operation by circulating air from the upper compart-ment to the lower compartment, through the ice condenser, and then back to the upper compartment. The operation will take place at the appropriate time following the accident. The secondary purpose of the system is to limit hydrogen concentration in potentially stagnant regions by ensuring a flow of air from these regions. l l 4.3-29
. - _.= - -.- .. . . - . . - - -. . - _- - ._
1 1 4
- 2. Radiolytic decomposition of water in the core.
l , 3. Radiolytic decomposition of water in the containcent sump. 4 i 4. Corrosion of materials within the containment. 1 The Combustible Gas Control System consists of two electric hydrogen recombiner units, located in the upper containment compartnent, and separate control panels and power supplies for each recombiner unit, located outside the containment in an area that is accessible following a loss-of-coolant l accident. The recombiner systems are completely redundant. i Figure 4.3-3 is a sketch of the recombiner unit. The recom- . biner unit consists of a preheater section, a heater-recom-bination section, and an exhaust section. t l The basic design parameters for the electric hydrogen recom-s biners are given in Table 4.3-2. I ! The post accident purge system required by Regulatory Guide 1.7, for use as a backup to the redundant hydrogen recom-tiner system,is functionally designed to exhaust containment air into the annulus and replenish this discharge with a dilution air supply. i [ 4.3-31 L
TABLE 4.3-1 DATA TABLE FOR THE VACUUM RELIEF SYSTEM Design basis: Maximum containment external pressure . differential 0.5 psig Maximum shield building external pressure differential 2.0 psig Design parameters: Upper compartment free volume 716,000 ft 3 Lower compartment free volume 400,180 ft 3 Ice condenser free volume 126,940 ft 3 Annulus space free volume 375,000 ft 3 Number of containment spray headers 2 Flow rate for each containment spray header 4.750 gpm Distance between spray headers and upper deck 152 feet Number of air return fans 2 Flow rate for each air return fan 40,000 cfm Maximum initial upper compartment dry bulb temperature 1100F Maximum initial lower compartment dry bulb temperature 1200F Minimum initial upper compartment relative humidity 30 percent Minimum containment spray water temperature 600F Ice condenser temperature (dry air) 150F Set pressure of ice condenser doors connected to the upper and lower compartment 1 psf Resultant design: Number of steel containment vacuum relief units 3 (1 redundant) Maximum initial external pressure. differential on the containment (containment vacuum relief system setpressure) 0.1 psig Design flow rate of each containment vacuum relief unit at 0.5 psig 28 lb/sec Maximum response time for any unit to be fully open for a design basis event 3 seconds 4.3-33
TABLE 4.3-2 ELECTRIC HYDROGEN RECOMBINER TYPICAL PARAMETERS Power (Maximum) 75kW II) Capacity (Minimum) 100 scfm Heaters
- Number 5 - Heater Surface Area / Heater 35 ft 2 2 - Maximum Heat Flux 2850 BUT/hr ft 2 or 5.8 Watts /in - Maximum Sheath Temperature 15500F Gas Temperature - Inlet 80 to 155 F - In Heater Section 1150 to 1400 F Materials - Outer Structure 300-Series S.S. - Inner Structure Inconel-600 - Heater Element Sheath Incoloy-800 Dimensions - Height 9 ft - Width 4.5 ft - Depth 5.5 ft Weight 4500 lb.
II) Power can be controlled by SCR. Normal operating power for typical PWR containments is 48.9 kW. 4.3-35
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l l l 4.4 DUAL CONTAINMENT AIR PURIFICATION AND CLEANUP SYSTEM i
, 4.4.1 Introduction Four engineered safety feature systems are provided for air purification and cleanup. One of these is the Containment Spray System discussed in Subsection 4.5. Two others are the air cleanup systems used in the two secondary contain-ment buildings. The one serving the reactor secondary 1
containment enclosure is the Emergency Gas Treatment System and the one serving the Auxiliary Building Gas Treatment System. The Ice Condenser System is the fourth engineered safety feature designed to serve as a Containment Air Purification i and Cleanup System. The ice condenser serves primarily as a large heat sink to readily reduce the containment temper-ature and pressure and condense the steam. For this purpose,
- ice is stored in a closed compartment between the lower and upper compartments of the containment. The containment is i
- designed such that the only significant flow path from the lower to the upper compartment is through the ice bed. I Immediately following a LOCA, a large pressure differential exists between the lower and upper compartment; therefore providing flow through the ice bed. Later in the transient, flow is provided by two 40,000 cfm fans which circulate l
upper containment air into the lower compartment. Since all flow between the lower and upper compartments must pass B i 4.4-1
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through the bed, the ice bed also serves as a removal mech-anism for fission products postulated to be dispersed in the containment atmosphere. Radiciodine in its various forms is the fission product of primary concern in the evaluation of , fission product transport and removal following a LOCA. The major benefit of the ice bed is its capacity to absorb molec-ular iodine from the containment atmosphere. To enhance this
; iodine absorption capacity of the ice, the ice solution is adjusted to an alkaline pH which promotes iodine hydrolysis to non-volatile forms.
There are no formal design bases established for air cleanup by the Containment Spray System. This was done with the knowledge that water from the Containment Spray System will remove halogens and particulates from the containment atmos-phere following a LOCA. No credit, however, was taken for 2 this removal process in accident analyses. In such cir-cumstances, no design bases are needed for this air puri-fication action. The design bases for the Emergency Gas Treatment System are:
- 1. To keep the air pressure within each Shield Building annulus below atmospheric at all times in which the integrity of that particular containment is required.
4 2. To reduce the concentration of radioactive nuclides in annulus air that is released to the environs during a LOCA in either reactor unit to levels sufficiently low to keep the site boundary dose rate below the 10 CFR 100 guideline value. 4.4-2
The design bases for the Auxiliary Building Gas Treatment System are:
- 1. To establish and keep an air pressure that is below atmospheric within the portion of the Auxiliary Building serving as' a secondary containment enclosure during accidents.
- 2. To reduce the concentration of radioactive nuclides in air releases from the Auxiliary Building secondary con-tainment enclosures to the environs during accidents to levels sufficiently low to keep the site boundary dose rate below the 10 CFR 100 guideline value.
- 3. To minimize the spreading of airborne radioactivit~y ll within the Auxiliary Building following an accidental release in the fuel handling and waste packaging areas.
4.4.2 System Design and Operation 4.4.2.1 Emergency Gas Treatment System The Emergency Gas Treatment System is shown schematically , in Figure 4.4-1. This system has two subsystems. One of these is called the Annulus Vacuum Control Subsystem and the other is called the Air Cleanup subsystem. Annulus Vacuum Control Subsystem The Annulus Vacuum Control Subsystem is a fan and duct net-work used to establish and keep a negative pressure level within the annular space between the two reactor containment structures. It is utilized during all normal operations in 4.4-3
4 which containment integrity is required. In emergencies in which containment isolation is required, this subsystem is isolated and shutdown. Under such an operating schedule, this subsystem performs no safety related function after the need for containment isolation has been established. Because of this, the Annulus Vacuum Control Subsystem was not classified as an engineered safety feature. This subsystem has two independently controlled branches. i~ Each branch serves one reactor unit. These branches draw air from their assigned annulus and release it into the auxiliary building exhaust duct system. The air inlet for each branch is centrally located in the secondary containment volume above the steel containment dome. Air pressure control in each secondary containment annulus is achieved with a redundant fan, differential pressure sensor, motor operated damper and control circuitry installation incorporated into each branch. This equipment provides a capability to vary the volumetric flow rate drawn from the annulus to keep the pressure at a predetermined negative pres-sure level. This control function is accomplished with a modulating damper under control of differential pressure sensor that adjusts the amount of outside air introduced up-stream of a constant capacity fan in the proper manner to keep the annulus pressure within a designated narrow range. Two independent installations of these items are provided to
)
4.4-4 .
promote operational efficiency. One of the two is utilized as a standby redundant unit that starts automatically in the event the operating control unit fails to function in the proper manner. The fans and flow control dampers serving both reactor second-ary containment annuli are installed in an Auxiliary Building room adjacent to the Unit 2 Shield Building. The nominal setpoint for each annulus vacuum control equip-ment installation is five inches of water below atmospheric. The fans employed to create such a negative pressure are described in Table 4.4-1, Dual Containment Characteristics. Air Cleanup Subsystem The Air Cleanup Subsystem is a redundant, shared airflow net-work having the capability to perform two functions for the . affected reactor secondary containment during a LOCA. One of these is to keep the secondary containment annulus air volume below atmospheric pressure. The second function is to remove airborne particulates and vapors from air drawn from the annulus that may contain radioactive nuclides. Each of these is accomplished by this subsystem without disturbing the unaffected reactor unit operations. Both of these functions are performed by processing and con-trolling a stream of air taken from the affected reactor unit secondary containment annulus. The air cleanup operation is 4.4-5 {
conducted by drawing the air stream through a series of filters and adsorbers. Annulus air pressure control is accomplished by adjusting the fraction of the airstream that is returned to the annulus air space. The rated capacity of each redundant air cleanup unit in the subsystem is 4000 cfm. These were designed in accordance with engineered safety feature standards. The air flow network for the Air Cleanup Subsystem was designed to provide the redundant services needed for either reactor secondary containment annulus. The intakes and ducting in this network used to bring annulus air to the Emergency Gas Treatment System room in the Auxiliary Building are those ' also used by the Annulus Yacuum Control Subsystem. The intake is centrally located within each Shield Building above the steel containment dome. Within the Emergency Gas Treatment System room the network branches out in a manner to supply two air cleanup unit installations that can be aligned with flow control dampers to serve either annulus air volume. After the air is processed, the Air Cleanup Subsystem air flow network directs the air to redundant damper controlled flow dividers in each reactor unit annulus. At these points; the flow networt contains two air flow paths leading to the
. reactor sh 4t, < lt and two air flow paths to a manifold that -
distrw ce!, ; a releases the air uniformly around the bottom of,the annulus. l l l l 4.4-6 l- . . . . , . . - _ . __
l t l The vertical seoaration between the intake above the dome and the exhaust ports in the manifold is 168-3/4 feet. Butterfly valves, rather than dampers, are installed in the ducts just above the flow distribution manifold to blinimize the outside air in-leakage from the reactor unit vents into the annulus. Another feature incorporated into the Air Cleanup Subsystem air flow network is the capability to cool the filters and adsorbers in an inactive air cleanup unit that is located with radioactive material. This is accomplished with two cross-over air flow ducts that can draw air at 200 cfm from the active air cleanup unit through the inactive air cleanup unit. (Such an air flow is sufficient to keep the temper-ature rise through a fully loaded inactive air cleanup unit to less than 75 F). Two butterfly valves in series are installed in*each cross-over air flow path to assure sufficient isolation to perform accurate removal efficiency tests on the HEPA filter and carbon adsorber banks. The two air cleanup units in the Air Cleanup Subsystem are steel housings containing air treatment equipment, samples, heaters, a drain, test fittings, and access Ocilities for maintenance. The air treatment equipment within the housing includes a demister, relative humidity heater, prefilter , bank, HEPA filter bank, two banks of carbon adsorbers in series and another HEPA filter bank. This equipment is 4.4-7
installed in the order listed. Between the two carbon adsorber banks is a thermostatically controlled electric heater installation designed to keep both carbon adsorber banks above the dew point of any air that could be drawn through the adsorber banks when the air purification oper-ation is initiated. A drain is incorporated into the housing adjacent to the demister installation to allow moisture sep-arated from the air stream to flow by gravity to a water collection tank in the Auxiliary Building. Integral to this housing are test fittings properly sized and positioned to permit orderly and efficient testing of the HEPA filter , and carbon adsorber banks. The relative humidity heater installed in the air cleanup units is an electric heater designed to heat the incoming air sufficiently to reduce the relative humidity of saturated air to 70 percent. Included in this installation is a tem-perature limiting controller that will shut the heater off 1 if excessive temperatures are detected. The HEPA filters, carbon adsorbers and carbon adsorber samples installed in the air cleanup units are standard items widely used in the nuclear power industry. The HEPA filters are 1000 cfm units designed to remove at least 99.97 percent of l the particulates greater than 0.3 micron in diameter. These ! filters are water and fire resistant units fabricated in 1 j accordance with MIL-F-51068C. The carbon adsorbers are Type ? I l l 4.4-8 l
l II unit trays, fabricated in accordance with AACC Standard l i CS-8. These trays contain two inch thick impregnated carbon beds. These trays, rated at 333 cfm, are installed in banks in which the face velocity is less than 40 feet per minute. Under such circumstances the residence time for air in the carbon bed is about 0.25 seconds. The carbon adsorber samples installed in the air cleanup units are made from the same carbon batches utilized to fabricate the adsorber trays. Six of the samples are mounted on the inlet side of the first adsorber bank. The total number of filters and
, adsorber unit trays provided in each air cleanup unit are listed in Table 4.4-1 Dual Containment Characteristics.
Two fully housed direct driven centrifugal fans are provided in the Air Cleanup Subsystem. Each of these is associated with a specific air cleanup unit. These fans were designed to function in' process air flow streams at temperature up to 200UF. See Table 4.4-1, Dual Containment Characteristics, for additional information on these fans. & Two air flow control modules are also included in the Air Cleanup Subsystem. Each of these modules are also assigned to a particular af r cleanup unit. These consist of a dif-ferential pressure sensor and transmitter, control circuitry, a damper actuator and two modulatirg dampers. The single damper actuator adjusts the dampers simultaneously in opposite 4 directions - one is closed when the other is opened. 4.4-9 i _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ . _ . _ _ . _ _ _ _ _ . _ __m-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ - _ _ - _ _ _ _ _ _ _.____m
i This air flow control equipment, installed in the secondary containment annuli, provides the capability to adjust the amount of air returned to the affected reactor unit annulus. The control circuitry provides four differential pressure I sensor setpoints for accomplishing this task. Two of these setpoint values are at pressures above the nominal negative pressure level needed in the annulus and two are below this nominal negative pressure level. The setpoint immediately above and immediately below the nominal annulus pressure level are utilized to initiate position changes in the modulating dampers installed in the air flow ducting accom- I modating the flow from the assigned air cleanup unit. Annulus pressures above the less negative pressure setpoint prodcce l a signal causing the damper actuator to begin to start opening the damper controlling the air flow to the reactor unit vent. Annulus pressures below the more negative setpoint initiate the opposite kind of damper motions. No damper position changes will occur when the annulus pressure level is between these two setpoint values. The controls for the Air Cleanup Subsystem were designed for two basic control modes. One mode of control has both air cleanup units in operation simultaneously. The second mode of control has either one of the units in operation and the other in a state in which it can automatically come into d.4-10
operation in the event that the operating unit fails. Annulus pressures beyond the dead bank defined by the set-points farthest from the nominal annulus negative pressure level are utilized to make this failure determination. This operating redundancy is achieved with spatially separated power and control circuitry having different independent power sources to prevent a loss of function from any single subsystem component. The tenn " train A" is used to identify one complete set of full capacity equipment and the term
" train B" is used to identify the other r.et of full capacity equipment. Power for both trains of equipment is supplied by the Emergency Power System.
Operation of the Air Cleanup Subsystem during accidents is initiated by the Phase A Containment Isolation Signal. Both the A and the B trains will be started by this signal coming from either reactor unit. A capability is also provided to start both trains with a hand switch in the main control room. Damper alignment is also initiated by th'e same signal, however, just those associated with the affected reactor unit will be activated. Another adjustment of a hand switch in the main control room will change the operating mode to the single train operation with the redundant train in a standby status. Employment of this operating mode is expected after the first 30 minutes of operation. The control room operator can select either train to remain in operation. 4.4-11 I
4.4.2.2 Auxiliary Building Gas Treatment System The Auxiliary Building Gas Treatment System is a fully redundant air cleanup network provided to reduce radioactive nuclide releases from the Auxiliary Building Secondary Con-tainment Enclosure during accidents. It does this by draw-ing air from the fuel handling and waste packaging areas through ducting normally used for ventilation purposes to air cleanup equipment and then directing this air to the reactor unit vent. In doing so, this system draws air from all parts of the Auxiliary Buildir.g to establish a negative pressure region in which virtually no unprocessed air passes from this secondary containment enclosure to the atmosphere. The rated capacity of each redundant air cleanup unit in this gas treatment system is 9000 cfm. These were designed in accordance with engineered safety feature standards. The unique portions of the Auxiliary Building Gas Treatment System are shown schematically in Figure 4.4-2. The airflow network for this system consists of two parallel duct instal-lations originating from exhaust ducting that normally serves the fuel handling and waste packaging areas in the building. Each of these ducts lead directly to an air cleanuo uriit, to the fan associated with the air cleanup unit and then directly to the reactor unit vent. - 4.4-12 1
The air flow network that is not unique to this system con-sists of most of the normal ventilation ducting installed in the Auxiliary Building Secondary Containment enclosure. When the Secondary Containment enclosure is isolated,this duct network provides a flow path for reducing the air pres-sure level in all parts of this enclosure. In some instances, air is drawn in the opposite direction to the normal air flow i pattern during operation of the Auxiliary Building Gas Treat-ment System. Two air cleanup units are utilized in the Auxiliary Building Gas Treatment System. Heaters located just upstream of the air cleanup units are designed to reduce the relative humidity of incoming saturated air to 70 percent. The air cleanup units are galvanized steel housings equipped with air treat-ment components, samples, a heater, test fittings, and access facilities for maintenance. The air treatment components within the housing include a pre-filter bank, HEPA filter bank and a carbon absorber bank. This equipment is installed in the order listed. Adjacent to the carbon absorber tank is a thermostatically controlled electric heater rated at 1500 watts. This is positioned to keep the carbon absorber bank above the dew point of any air that could be drawn through the absorber banks when the air purification operation is started. Integral. to this housing are test fittings properly sized and positioned to allow HEPA filter and carbon absorber
- bank leakage tests to be conducted in an orderly and efficient manner.
i ! 4.4-13 L i
Air is drawn through each of these air cleanup units by a belt driven centrifugal fan. The drive for the fan is an electric motor rated at 20 HP. Additional information on these fans is given in Table 4.4-1 Dual Containment Charac-teristics. l Two air flow control modules, each assigned to a particular air cleanup unit are utilized in the Auxiliary Building Gas Treatment System. These contain a differential pressure sensor and transmitter, control circuitry and a motor oper-ated modulating damper. These two air flow control modules provide the capability for keeping the pressure within the Auxiliary Building Secondary Containment Enclosure at 1.4 inch of water below atmospheric. These modules do this by varying the amount of air drawn from . this enclosed volume in a manner to keep the pressure 'at this desired negative value. This is done with a modulating damper that is controlled by the differential pressure transmitter to adjust the amount of outside air introduced into the duct network just upstream of the constant capacity fan described i above. Such action wil1 bring in sufficient outside air to keep the fan flow rate at its rated flow at all times. It will also draw enough air from the Auxiliary Building Second-ary Containment Enclosure to establish and keep the desired i t negative pressure level. i l 4.4-14 l
l The controls for the Auxiliary Buildir.g Gas Treatment System were designed to provide two basic control modes. One con-i trol mode has both air cleanup units in operation simultan- , eously. The second control mode has either one of the air cleanup units in operation and the other in a state in which it can automatically come into operation ir. the event the operating unit fails. A low flow signal from the operating unit is utilized in this control mode to make this failure determination. This operational redundancy is achieved with spatially separated power and control circuitry having dif-i ! erent independent power sources to prevent a loss of function from any single system component failure. The term " train A" is used to identify one complete set full capacity equipment and the term " train B" is used to identify the other set of full capacity equipment. Power for both equipment trains is a supplied by the Emergency Power System. Operation of the Auxiliary Building Gas Treatment System begins automatically upon receipt of a:
- 1. Phase A containment isolation signal from either reactor
- unit, or a
{ 2. High radiation signal from the fuel handling area radi-ation monitors, or a
- 3. High radiation signal from the, auxiliary building exhaust
, vent monitors. e 4.4-15 ' f
+
1
.- .-._v- . , - _ . - - - - .y,- -- -- . - - - -
.---r . . - -. ,-, _ . , , - - - ---..
C 4 A capability is also provided to start both trains with a hand switch in the main control room. Another ad,justment capability provided in the hand switch in the main control room will change the operating mode to the single train operation with the redundant train in a standby status. Employment of this operating mode is expected after the first 30 minutes of operation. In this instance the main control room operation has the capability to select either unit to remain in operation. 4.4.2.3 Ice Condenser The function of the post LOCA iodine removal served by the Ice Condenser is accomplished by chemically controlling the ! alkaline ice to a pH range of 8.5 to 9.5. This is accomplished
- by adding sodium tetraborate to the Grade A feedwater in the solution of Na247 00. 10H O with 2000 + 100 ppm of Boron 2
prior to ice basket loading. During the accident, the melting i ice provides a medium for removal of iodine from the contain-ment atmosphere and fixation in solution. t 4.4.3 Summary Description Four engineered safety feature systems are provided for air purification and cleanup. One of these is the Containment Spray System discussed in Subsection 4.5. Two others are the air cleanup systems used in the two secondary containment ! buildings. The one serving the reactor secondary containment enclosure is the Emergency Gas Treatment System and the one serving the Auxiliary Building secondary containment enclosure is the Auxiliary Building Gas Treatment System. 4.4-16
The design bases for the Emergency Gas Treatment System are:
- 1. To keep the air pressure within each Shield Building annulus below atmospheric at all times in which the integrity of that particular containment is required.
- 2. To reduce the concentration of radioactive nuclides in annulus air that is released to the environs during a LOCA in either reactor unit to levels sufficiently low to keep the site boundary dose rate below the 10 CFR i
100 guideline value. The design bases for the Auxiliary Building Gas Treatment System are:
- 1. Tor establish and keep an air pressure that is below atmospheric within the portion of the Auxiliary Building i serving as a secondary containment enclosure during accidents.
j 2. To reduce the concentration of radioactive nuclides in , air releases from the Auxiliary Building secondary con-tainment enclosures to the environs during accidents to j levels sufficiently low to keep the site boundary dose rate below the 10 CFR 100 guideline value.
- 3. To minimize the spreading of airborne radioactivity within the Auxiliary Building following an accidental release
- in the fuel handling and waste packaging areas.
t 4.4-17
+ The ice condenser serves primarily as a large heat sink to ! readily reduce the containment temperature and pressure and condense the steam. Since all flow between the lower and upper compartments must pass through the ice bed, the ice bed also serves as a removal mechanism for fission products postulated to be dispersed in the containment atmosphere. Radioiodine in its various forms is the fission product of primary concern. The Emergency Gas Treatment System is shown schematically in Figure 4.4-1. This system has two subsystems. One of these is called the Annulus Vacuum Control Subsystem and the other is called the Air Cleanup Subsystem. 1 The Annulus Vacuum Control Subsystem is a fan and duct network
- used to establish and keep a negative pressure level within the annular space between the two reactor containment structures.
l. It is utilized during all normal operations in which contain-ment integrity is required. The nominal setpoint for each annulus vacuum control equipment installation is five inches of water below atmospheric. The fans employed to create such negative pressure are described 4 in Table 4.4-1, Dual Containment Characteristics. 4 The Air Cleanup Subsystem is a redundant, shared airflow net-work having the capability to perform two functions for the
/
- l 4.4-10
affected reactor secondary containment during a LOCA. One of these is to keep the secondary containment annulus air volume Delow atmospheric pressure. The second function is to remove airborne particulates and vapors from air drawn from the annulus that may contain radioactive nuclides. Each of these is accomplished by this subsystem without disturbing the unaffected reactor unit operations. Within the Emergency Gas Treatment System room the network branches out in a manner to supply two air cleanup unit installations that can be aligned with flow control dampers to serve either annulus air volume. The two air cleanup units in the Air Cleanup Subsystem are steel housings containing air treatment equipment, samples, heaters, a drain, test fittings, and access facilities for maintenance. The air treatment equipment within the housing includes a derr; ster, relative humidity heater, pre-filter bank, HEPA filter bank, two banks of carbon absorbers in series and another HEPA filter bank. This equipment is installed in the order listed. Two fully housed direct driven centrifugal fans are provided in the Air Cleanup Subsystem. Each of these is associated with a specific air cleanup unit. Operation of the Air Cleanup Subsystem during accidents is initiated by the Phase A Containment Isolation Signal. 4.4-19
The Auxiliary Building Gas Treatment System is a fully redun-dant air cleanup network provided to reduce radioactive nuclide releases from the Auxiliary Building Secondary Con-tainment Enclosure during accidents. It does this by drawing
- air from the fuel handling and waste packaging areas through ducting normally used for ventilation purposes to air cleanup equipment and then directing this air to the reactor unit
- vent. In doing so, this system draws air from all parts of the Auxiliary Building to establish a negative pressure region in which virtually no unprocessed air passes from this second-ary containment enclosure to the atmosphere.
, The unique portions of the Auxiliary Building Gas Treatment System are shown schematically in Figure 4.4-2. The airflow network for this system consists of two parallel duct in-
'stallations originating from exhaust ducting that normally serves the fuel handling and waste packaging areas in the building. Each of these ducts lead directly to an air cleanup i unit, to the fan associated with the air cleanup unit and a
then directly to the reactor unit vent. The air cleanup units are galvanized steel housings equipped j with air treatment components, samples, a heater, test fit-tings, and access facilities for maintenance. The air treat- _. ment components within the housing include a pre-filter bank, i HEPA filter bank and a carbon absorber bank. This equipment is installed in the order listed. 4.4-20 .I
- - ~ . __ , , _ .
_ , . , - - , , - . - , _ . _ - .r-., . . - . _ . . , , , ,c __,y ,% _--3,79 w ,, _w,,,mm._ .
. . _ ._ -._. . _ _ ~ . _ . . . ___ __ .- _
i Air is drawn through each of these air cleanup units by a belt driven centrifugal fan.
}
Operation of the Auxiliary Building Gas Treatment System begins automatically upon receipt of a:
- l. Phase A containment isolation signal from either reactor f
unit, or a
- 2. High radiation signal from the fuel handling area radi-i ation monitors, or a
- 3. High radiation signal from the auxiliary building exhaust i
vent monitors. The function of the oost LOCA iodine removal served by the f Ice Condenser is accomplished by chemically controlling the alkaline ice to a pH range of 8.5 to 9.5. This is accomplished by adding sodium tetraborate to the Grade A feedwater in the ! solution of Na 247 0 0 . 10H O 2 with 2000 + 100 ppm of Boron prior ; to ice basket loading. During the accident, the nelting ice 4 provides a medium for removal of iodine from the containment atmosphere and fixation in solution. There are no formal design bases established for air cleanup by the Containment Spray System. This was done with the knowledge that water from the Containment Spray System will , remove halogens and particulates from the containment atmos-phere following a LOCA. No credit, however, was taken for this renoval process in accident analyses, i i s 4.4-21 t i _ . . _ . _ . _ . _ _____.________________m_ _ . _ _ _ _ _ . _ _ _ _ _ _ . _ . _ _ _ _ . . _ . _ _ _ _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _
TABLE 4.4-1 . DUAL CONTAINMENT CHARACTERISTICS
- 1. Shield Building A. Air annulus volume: 375,000 Ft 3 B. Design inleakage rate at 0.5 in, wg: 38.5%/ day C. Recirculation fans *
- 1. Number: 2
- 2. Type: Centrifugal
- 3. Air flow rate: 4000 cfm each at 11 in, wg (exhaust 100 cfm. recirculate 3900 cfm)'
T Ae Banks / Train Number / Bank Number / Train Total Number Prefilter 1 2 2 4 HEPA 2 4 8 16 Carbon 2 12 24 48 D. Exhaust fans **
- 1. Number: 4 (2 for each reactor unit)
- 2. Type: Centrifugal
- 3. Air flow rate: 1000 cfm each at 5 in, wg II. Auxiliary Building A. Free volume: 3,480,000 Ft 3 B. Exhaust fans (Auxiliary Building Gas Treatment System)
- 1. Number: 2
- 2. Type: Centrifugal
- 3. Air flow rate: 9000 cfm at 8 in, wg C. Filters (in the Auxiliary Building Gas Treatment System)
Iype. Banks / Train Number / Train Total Number Prefilter 1 9 18 HEPA 1 9 18 Carbon 1 27 54
*The fans described are the air cleanup subsystem fans in the Emergency Gas Treatment System that operate only in the postaccident period. **The fans described are the annulus vacuum control fans in the Emergency Gas Treatment System that operate only during nonaccident operations.
4.4-23
TO AUX BLOG EXHAUST VENT O TRAIN A JL 3 TRAIN B
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- c. ""_______q ' ""I l ==,l l H l r hO u O@ o UNIT NO. I
=' = w--y = =
1 UNIT NO. 2 l : l I ! ANNULUS (f) 3 l A N N U LU S I 1 g 1 I A(?)O A 1 31 l I+i Gl u u r ;;;5,TA,li "se g1 -av i i I-ce c , a I "I i l " *ikg "*-D$l ' ' r -HTG ColLS l i XM g *dl 7 r i i g gl 1 i jg j i e PR E FILTER l g gl y g l I iI a-HEPA Fit TER y y l '
$2 CHARCOAL g g l FILTERS IN SERIES l l
g O '
--HEpA riLTEn g
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l u o i n T: n I u u I l bSC "NIEl v i S TREAT. [ v MENT SYS Q . 0 0 j I l EXHAUST l -- -- I "4 g "." l FA N S l f--0@ _ l
/-{) " " I I 1.L 1 1 1 1 1 1 l l
l fg U@ U@f7g l i 11111111l l l l L________J G O L________J M = MODULATING DAMPER T = TWO POSITION DAMPER I = ISOLATION DAMPER OR BUTTERFLY VALVE FIs FLOW INDICATOR Figure 4.4-1 Emergency Gas Treatment System SLhematic Diaoram 4.4-25
SIDE SHIELD BLDG , PREFILTERS O EXHAUST VENT ! WASTE PACKAGING AREA O RH HEATER a e
~h f ,. a 7.g ~;' _J:a x T Q :7: , I I - *- F CHARCOAL FILTER AUX BLDG GAS TREATMENT EXHAUST FANS OUTSIDE l MODULATED DAMPERS O FOR AUX BLDG g PRESSURE CONTROL l
.- a N_ a a I a b g __, 6
-J: -
- x C3 :7: T' V L -
(SAME AS ABOVE) FUEL HANDLING AREA O INDICATES A TRAIN W INDICATES B TRAIN m TO FUEL HANDLING AREA EXHAUST FANS Figure 4.4-2 AUXILIARY BUILDING CAS TRGTMENT SYSTEM EXIIAUST NETWORK
/ TABLE OF CONTENTS Section Title Pace 5.C RADIATION MONITORING SYSTEM. . . . . 5-1
5.1 INTRODUCTION
. . . . . . . . . . . . 5-1 5.2 PROCESS RADIATION MONITORING SYSTEM . . . . . . . . . . . . . . 5-2 5.2.1 Containment Air Particulate Monitor. . . . . . . . . . . . . . 5-3 5.2.2 Containment Noble Gas Monitor. . . . 54 5.2.3 Containment Purge Exhaust Monitor. . 5-6 5.2.4 Gas Decay Tank Effluent Gas Monitor. . . . . . . . . . . . . . 5-7 5.2.5 Auxiliary Building Cubicle Ventilation System Monitor . . . . 5-7 5.2.6 Seismic Class I Pipe Tunnel Monitor. 5-8 5.2.7 Stack Gas Monitor. . . . . . . . . . 5-9 5.2.8 Miscellaneous Ventilation Exhaust Monitor. . . . . . . . . . . . . . 5-10 5.2.9 Condenser Air Ejector Gas Monitor. . 5-11 5.2.10 Steam Generator Blowdown Liquid Monitor. .. . . . . . . . . . . . . 5-12 5.2.11 Waste Disposal System Liould Effluent Monitor. . . . . . . . . . . . . . 5-13 5.2.12 Component Cooling Loop Liouid Monitor. . . . . . . . . . . . . . 5-14 5.2.13 Reactor Containment Fan Cooler Service Water Monitor. . . . . . . 5-15 5.2.14 Radwaste Evaporator Service Water Monitor. . . . . . . . . . . . . . 5-16 i
TABLE OF CONTENTS CONT'D Section Title Pm 5.2.15 Evaporator Steam Condensate Return Liquid Monitor . . . . . . . . . . 5-16 5.2.16 Blowdown Condenser Service water Monitor. . . . . . . . . . . . . . 5-17 i 1 5.2.17 Residual Heat Removal Pump, RHR i Heat Exchanger, and Centrifugal Charging Pump Cubicle Monitors . . 5-18 5.2.18 Passive Failure Gas Monitor. . . . . 5-19
! 5.2.19 Control Room Intake Air Monitor. . . 5-20 5.2.20 Control Room Intake Fro.n Turbine Building Monitor . . . . . . . . . 5-20 5.2.21 Reactor Vessel Leakage Monitor . . . 5-21 1
5.3 FAILED FUEL MONITORING SYSTEM , . . 5-23 5.4 CONTAINMENT AREA SAMPLING SYSTEM . . 5-28 5.5 AREA RADIATION MONITORING SYSTEM . . 5-25 t 11
l t i
- i i
4 LIST OF TABLES l i Table Title Page 4 5-1 Area Monitors. . . . . . . . . . . . 5-29 P r l h i > I i e 6 I I r 1 a t i r i I i i i h ,i e I e I i e , E 1 l l t I 4 ) k 1 I I i I E I i f f I f I .i i t ) l 111 i i l I l
l I f } q. 5.0 RADIATION MONITORING SYSTEM ; 4
5.1 INTRODUCTION
The Radiation Monitoring System is designed to continuously 1 monitor the containment atmosphere, all plant effluents, and 1 various in-plant locations .
! In addition, this system will also serve to give an immediate ] alarm in case of any significant change in the level or radio-activity in the containment, the spent fuel storage area, or the auxiliary building. Recorders provide a permanent record of the radioactivity level at the various in-plant locations.
The monitoring system is divided into the following subsystems: i l The Process Radiation Monitoring System, consisting of ! channels which primarily give early warning of an equipment 1 or system malfunction and also warn operating personnel l i of increasing radiation which might result in a radiation health hazard. ! The Area Radiation Monitoring System, consisting of channels
- i
! which primarily inform operating personnel of existing radiation levels in various areas of the plant. L i Table 5-1 identifies those areas of the plant in which radiation monitors are located and summarizes the important features of ( l l .each monitor. ! l l l l p 5-1 l l
- - -= .
If any of the process or area monitors listed in plant Technical Specifications is out of service, manual grab samples or surveys
~
shall be taken once per shift. 5.2 Process Radiation Monitoring System This system consists of nucerous channels with detectors located } throughout the plant at the measurement location. The computing and readout equipment is located in cabinets in the control room. 5 I The process radiation monitoring equipment has been designed i
- for long term operation under the following environment conditions.
j _ Reactor Containment Temperatur1: +50 F to +120 F i j Relative Humidity: 0 to 100% Pressure: 14.7 1 psia normal Radiation: 10 mR/hr gamma background typical, at detector I locations. Control Rocm and Auxiliary Buildir.3 Temperature: + 75 F t 10 F normal operating range (+400 F to +120 F maximum operating range). Belative Humidity: 15% to 95% Pressure: Atmospheric Radiation: 2.5 mR/hr background Stack Monitors i Temperature: -30 F to +110 F Relative Humidity: 0 to 100% l 5-2
f Pressure: Atmospheric Radiation: Background 5 2.1 Containment Air Particulate Monitor The-containment air particulate monitor is sufficiently sensitive for detection of reactor coolant leakage into the containment within several minutes after leakage occurs. This instrument is capable of continuously detecting particulate radioactivity in. concentrations as low as 10~9 uc/cc of containment air based on I-131. It is anticipated that leakage rates of the order of 5 cc/ min. will be detectable within minutes after the leak occurs. Continuous air samples are taken from the containment atmosphere near the reactor containment fan cooler inlet, drawn outside the containment in closed sealed systems and monitored by a scintillation counter and movable filter paper detector assembly. The air sample is passed through a filter paper which collects 99% of all particulate matter greater than 0.3 micron in size. The constantly moving filter paper is viewed by a shielded photomultiplier-scintillation crystal combination. The air samples are returned to the containment after passing through the gas monitor. The amplified detector output is transmitted to a rate meter in the Radiation Monitoring System cabinets in the control room. The activity is indicated on a meter and recorded. High-activity alarm indication is displayed on the 5-3
control board annunciator in addition to the Radiation Monitoring cabinets. Receipt of a high activity alarm will automatically close the containment purge supply and exhaust dampers, the pressure and vacuum relief isolation valves, and shift the containment ran coolers dampers to the accident mode (the fans, however, remain in fast speed). The photomultiplier-scintillation monitor is beta-gamma sensi-tive and is used principally to detect the containment air I-131, I-133, Cs-134, Cs-137 activities, and has a sensitivity range of 10-9 to 10-6 microcuries per cubic centimeter. Displayed range is 10 cpm to 10 6 cpm. 5 2.2 Containment Noble Gas Monitor The containment gas monitor is provided in order to supply the operator with information pertaining to the noble gas activity in the containment. This activity is due to neutron activation of the primary shield cooling air and from leaks in the reactcr coolant system when operating with cladding defects in the fuel. The radio gas detector will supplement the information obtained from the air particulate monitor regarding the occurrence of leakage from the primary system. Based on a reactor coolant gas activity of 0.3 ue/cc a leakage rate of approximately 20 cc/sec would produce a containment 5-4
- _ _ - . - . _ - - _ . _ - = _ _ _ . __ - -
noble gas activity of approximately 5.0 x 10-6 ue/ce at equilibrium. This instrument is capable of detecting gaseous radioactivity at concentrations as low as 10-6 uc/cc based on i Kr-35 Continuous samples are taken from the containment atmosphere after they pass through the previously described' air particu-late monitor through a closed, sealed system to the gas monitor assembly. The samples flow continuously to the fixed, shielded volume, where the activity is measured by a Geiger-Mueller tube. The samples are then returned to the containment. The detector output is transmitted to the Radiation Monitoring System cabinets in the control room. The radioactivity is indicated by meters and recorded. High-radioactivity indications are displayed on the control board annunciator in addition to l the Radiation Monitoring cabinets. i Receipt of a high activity alarm will automatically close the containment purge supply and exhaust dampers, the pressure and vacuum relief line isolation valves, and shift the containment fan coolers' dampers to the accident mode (the fans, however, remain in fast speed). This monitor will detect noble gases, principally Kr-85, Ar-41, Xe-133 and Xe-135, and has a sensitivity range of 10-6 to 10-3 microcuries per cubic centimeter. 5-5
-~ - - , , - . , - - - - , . . , ~ - . - -- , - . , , - - . .
_. - . . - - - . = - . . - - _-. - - . _. . _ _- - - _ - - 5.2.3 Containment Purge Exhaust Monitor , This channel monitors the effluent from the containment purge for gaseous activity, iodine and particulate activity whenever the purge system is in operation. The system consists of three separate channels, one of which is a fixed filter air parti-culate monitor with a beta scintillation detector, the second is a Geiger-Mueller detector for monitoring gaseous activity and the third is a spectrometer grade gamma scintillation detector for iodine monitoring. Detector outputs are transmitted to the Radiation Monitoring i System cabi'ets n in the control room. High radioactivity during 4 l containment purge operations automatically initiates closure i of the butterfly valves to prevent any radioactive release to ! the atmosphere. The radioactivity levels are indicated by meters and recorded. High radioactivity-alarm indications
- are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets.
1 i A high radioactivity alarm will automatically trip shut the l containment purge supply and exhaust dampers.
- The sensitivity range of these monitors is 10-6 to 10-2 uc/cc for noble gases,'lc'11 to 10-5 uc/cc for particulates an. ;
~11 10 to 10-5 uc/cc for lodine.
The containment purge exhaust monitor is listed in the plant j Technical Specifications. 5-6 r k t- _ , _ ~ _ , . , . , - , _ - _ _ , , . _ , _ - - - . - _ . . . . _ . _ . - . - _ - . _ -
5 2.4 Gas Decay Tank Effluent Gas Monitor The detector is a Geiger-Mueller type (high sensitivity beta-
~~~
gamma detector) operated inside a shielded chamber with three inches of lead in all directions. The detector output is transmitted to the Radiation Monitoring System cabinets in the control room. The radioactivity released through the plant vent is indicated by a meter and recorded. A high radio-activity alarm will automatically close the isolation valve in the vent line, thus terminating the release and vill initiate operator action to establish and correct the cause of the alarm. The alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. Technical Specifications require this monitor to be operating during any gaseous release. The monitor will detect noble gases, principally Kr-85, Xe-133 and Xe-135, and has a sensitivity of 10-6 microcuries per cubic < centimeter based on Kr-85 5 2.5 Auxiliary Building Cubicle Ventilation System Monitor This channel continuously monitors the ventilation system ex-haust air.from all the potentially contaminated equipment cubicles. One moving filter monitor is used for particulate and a fixed filter monitor is used for iodine. Detection of. a high radioactivity level from the iodine monitor will. auto-matically route the ventilation exhaust from the equipment 5-7 l l
cubicles through the charcoal filter banks prior to exhausting the air to the atmosphere. The detector fer particulate is a plastic phosphor beta scintillation unit mounted in a shielded assembly. For iodine monitoring, a spectrometer grade gamma scintillation detector consisting of a sodium iodide crystal in a shielded assembly is used. The detector outputs are transmitted to the Radiation Monitor-ing System cabinets in the control room where the radioactivity levels are indicated by meters and recorded. High radioactivity-alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. The sensitivity range of these monitors is 10-11 to 10-5 ue/cc for both iodine and particulates. The auxiliary building cubicle ventilation system monitor is listed in plant Technical Specifications. 5.2.6 Seismic Class I Pipe Tunnel Monitor - This channel monitors the ventilation system air discharging from the pipe tunnel connecting the containment with the auxil-iary building for any indication of leakage from piping systems in the tunnel. The system consists of a dual channel iodine and air particulate monitor. On detection of a high Iodine level, the ventilating air from the tunnel is automatically 5-8
- - - - = -
t . routed through the charcoal filter bariks prior to exhaust. This system serves to identify any leakage from passive failures within the tunnel during post accident recirculation operation. The detectors used are of the moving filter beta scintillation type for particulate and fixed filter gamma scintillation type for iodine. Both detectors are enclosed in shielded assemblies for protection from background fluctuations. The detector outputs are transmitted to the Radiction Monitoring System cabinets in the control room where the radioactivity levels j are indicated by meters and recorded. High radioactivity-1 alarm indications are displayed on the control board annun-clator in addition to the Radiation Monitoring cabinets. The sensitivity range of the monitors is 10-11 to 10-5 ue/cc for both iodine and particulate. The seismic class I pipe tunnel monitor is lised in the plant Technical Specifications. 5.2.7 Stack Gas Monitor This channel monitors the ventilation system air discharging i from the auxiliary building ventilation system to the plant ventilation stack. .The sample gas is returned to the suction of the auxiliary building exhaust fans. The channel utilizes four Geigel-Mueller tubes connected in parallel. The radioactivity is indicated-by a meter and 5-9
1 recorded. High radioactivity alarm indications are displayed on the control board annunciator in addition to the radiation monitoring cabinets. There are no automatic functions assoc-lated with this channel. This monitor has a sensitivity of 10~ micorcuries per cubic centimeter for radioactive noble gases, principally Kr-85, An-41, Xe-13 3 and Xe-135 The atack gas monitor is listed in the Plant Technical Speci-fications. 5.2.8 Miscellaneous Venti'.ation Exhaust Monitor This enannel continuously monitors the ventilation exhaust from the auxiliary electrical equipment room, computer room, laboratories, decontamination rooms and other miscellaneous areas. One mo"ing filter monitor is provided for collecting particulate activity. No automatic action is initiated in the system by an indication of a high activity level. The operator will determine the contaminated area by means of local inspection, isolate the area, and take the necessary corrective action upon system alarm. l The gas monitor utilizes a Geiger-Mueller detector in a shielded assembly while the particulate monitor employs a beta scintillation detector. l l 5-10
l The detector outputs are transmitted to the Radiation Monitor-ing System cabinets in the control room where the radioactivity levels are indicated by meters and recorded. High radio-activity-alarm indications are displayed on the control board annunciator in addition to the Radiation MonitorinF cabinets. The sensitivity range of these monitors is 10-11 to 10-5 ue/cc
-2 for particulates and 10- to 10 uc/cc for gases.
The miscellaneous ventilation exhaust monitor is listed in the plant Technical Specifications. 4 5.2.9 Condenser Air Ejector Gas Monitor _ This channel receives a continuous air sample from the air ejector exhaust header and monitors it for gtseous. radio-activity and provides the plant operator >:ith a rapid indication of a primary to secondary system leak. The sample gas is re-turned to the gas effluent. The Geiger-Mueller detector output is transmitted to the Radiation Monitoring System cabinets in the control room. The radioactivity is indicated by a meter and recorded. High radioactivity-alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. This monitor has a sensitivity of 10-6 microcuries per cubic centineter for radioactive noble gases, principally Kr-85, Xe-133 and Xe-135 5-11 w - - w rw +
This monitor is listed in Technical Specifications. 5 2.10 Steam Generator Blowdown Liould Monitor This channel monitors the liquid phase of the secondary side of the steam generator for radioactivity, which would indicate a primary to secondary system leak, providing backup informa-tion to that of the condenser air ejector gas monitor. Samples from each of the four steam generator bottoms are mixed in a common header and the common sample is continuously monitored by a scintillation counter and holdup tank assembly. Upon indication of a high radioactivity, each steam generator is individually sampled in order to determine which unit is leaking. This sampling sequence is achieved by manually selecting the desired unit to be monitored and allotting suf-ficient time for sample equilibrium to be established. The detector output is amplified and transmitted to the Radiation Lionitoring System cabinets in the control room. High radioactivity alarm indications are displayed on the control board annunciator and the Radiation Monitoring System cabinets. The radioactivity of the sample being monitored is both indicated and recorded. The sensitivity range of thic _c s monitor is 10 ' to 10 microcuries per cubic centimerer based en Co-60. l This monitor is listed in Technical Specifications. l 5-12 l
5.2.11 Waste Disposal System Liauid Effluent Monitor This channel continuously monitors all Waste Disposal System liquid releases from the Lake Discharge Tanks. Automatic valve closure action is initiated by this monitor to orevent further release after a high radioactivity condition is indi-cated or alarmed. The valve is located over 250 feet down-stream of the monitor to allow closure prior to any radio-active release. A Geiger-Mueller detector in a shielded assembly monitors all effluent discharges. The detector assembly output is amplified and transmitted to the Radiation 4 Monitoring System cabinets in the control room. The activity level is indicated on a meter and recorded. High radioactivity alarm indications are displayed on the control board annun-ciator in addition to the Radiation Monitoring cabinets. The instrument is beta-gamma sensitive and has a sensitivity range of 10~ to 10 -2 microcuries per cubic centimeter based on Co-60. The accuracy of these monitors will be maintained at +5% of set point to provide a highly reliable backup to the multiple sample analyses prior to discharge. A single monitor is pro-vided on each discharge line and is considered adequate since the tank sample analyses are the primary method for deter-mination of allowable discharge volume and flow. The release l of liquid waste is under administrative control and the ; 1 monitor is provided to maintain surveillance over the release. 5-13
By Technical Specifications, the waste disposal system liquid effluent monitor must be in operation to make a liquid release. , 5.2.12 component cooling Loop Liquid Monitor This channel continuously monitors the component coolant loop of the Auxiliary Coolant System for activity indicative of a leak in any heat exchanger tube in the Chemical and Volume
. Control System, Residual Heat Removal System, Sampling or the Spent Fuel Pit Cooling System, or a cooling coil for the thermal barrier cooler on a reactor coolant pump.
Upon indication of high radioactivity the operator can estab-lish the source of leakage to the component cooling loop by selectively controlling the flow to each system. The component cooling surge tank vent valves will automatically close upon a high activity condition. A scintillation counter inserted in an in-line well is used as the detector. The output is preamplified and transmitted to the Radiation Monitoring System cabinets in the control rooms. The activity level is indicated on a meter and recorded. High-activity alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. The sensitivity range of this monitor is 10 -5 to 10 -2 m1c,o_ curies per cubic centimeter based on Co-60. ' 5-14
l I The component cooling liquid monitor is listed in the plant Technical Specifications. 5.2.13 Reactor Containment Fan Cooler Service Water Monitor This channel is used to monitor the activity of the service water return from the fan coolers in order to detect in-leakage into the cooling water. The detector consists of a Geiger-Mueller detector mounted in a shielded assembly. The - a outlet flow from each of the heat exchangers is mixed in a common header and monitored. Upon indication of a high radio-activity, each heat exchanger is individually sampled to determine which unit is leaki.ig. This sampling sequence is achieved by manually selecting the desired unit to be moni-tored and allotting sufficient time for sample equilibrium to be established. The amplified detector output is transmitted to the radiation monitoring cabinets in the control room. The radioactivity of the sample being monitored in indicated at the monitor location and indicated and recorded is the control room. t High radioactivity alarm indications are displayed on the con-trol board annunciator and the radiation monitoring system cabinets. The instrument is beta-gamma sensitive and has a sensitivity range of 10
~
to 10-2 microcuries per cubic centimeter based en Co-60. 5-15
5.3.14 Radwaste Evaporator Service Water Monitor This channel continuously monitors the service water return header from the radwaste evaporator in order to detect any leakage in the cooling lines. A common service water return header from the evaporator, concentrates cooler, distillate cooler, and vapor condenser, allows monitoring of all com-ponents of the radwaste evaporator equipment. Manual valve closure is required to isolate the unit after a high radio-activity condition is indicated or alarmed. A Geiger-Aueller detector in a shielded assembly monitors the service water return. The detector assembly output is transmitted to the Radiation Monitoring System cabinets in the control room. The activity level is indicated on a meter and recorded. High radioactivity alarm indications are dis-played on the control board annunciator in addition to the Radiation Monitoring cabinets. The instrument is beta-gamma sensitive and has a sensitivity range of 10- to 10-2 microcuries per cubic centimeter based . on Co-60. 5.2.15 Evaporator Steam Condensate Return Liquid Monitor This channel continuously monitors the steam condensate returns from the boric acid and radwaste evaporators for any steam line leakage. The monitor is located downstream of a common collecting tank and allows the condensate t.o be checked for 5-16 , . . - _- . . _ . ._ _ _ _ ~ _ . _ - . .
act4.vity prior to being returned to the auxiliary steam system. Manual valve closure is required to isolate the units after a high radioactivity condition is indicated or alarmed. A scintillation detector is used to monitor the flow and the detector is enclosed in a shielded assembly. The detector assembly output is transmitted to the Radiation Monitoring System cabinets in the control room. The activity level is indicated on a meter and recorded. High radioactivity alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. The instrument is beta-gamma sensitive and has a sensitivity range of 10 -6 to 10 -2 microcuries per cubic centimeter based ' on Co-60. 5.2.16 Blowdown Condenser Service Water Monitor This channel continuously monitors the common service water return header from both blowdown condensers to, detect any leakage in the cooling lines. Manual valve closure is re-quired to isolate the units after a high radioactivity condition i is indicated or alarmed. A Geiger-Mueller detector in a shielded assembly monitors the service water return. The detector assembly output is transmitted to the Radiation Monitoring System cabinets in l 5-17 l
i the control room. The activity level is indicated on a meter and recorded. High radioactivity alarm indications are displayed on the control board annunciator in addition 4 to the Radiation Monitoring cabinets. The instrument is beta-gamma sensitive and has a sensitivity range of 10- -2 microcuries per cubic centimeter based to 10 on Co-60. 5.2.17 Residual Heat Removal Pump, RHR Heat Exchanger and Centrifugal Charging-Pump Cubicle Monitors. These channels monitor the air returns from the RHR Pump and Heat Exchanger and Centrifugal Charging Pump equipment cubicles for any evidence of leakage. An air sample is taken from the ventilation air leaving each cubicle and passed through separate moving filter air particulate monitors. Any increase in acti-vity above normal background levels would be indicative of a possible leak and appropriate operator action could be initiated. The detectors used with these monitors are of the beta scintil-lation type in a shielded assembly. The detector output of each monitor is transmitted to the Radiation Monitoring System cabinets in the control room where the radioactivity level is indicated by a meter and recorded. High radioactivity-alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring 5-18' .l
cabinets. Local alarms annunciate the supporting equipment's operational status.
~11 The sensitivity range of each monitor is 10 to 10 -5 ue/cc.
5.2.18 Passive Failure Gas Monitor This system monitors the ventilation system return air from the RHR Pumr nd Heat Exchanger and Centrifugal Charging Pump Cubicles and Seismic Class I Pipe Tunnels for gaseous activity which would be indicative of a leak. These locations, as previously discussed, are being monitored for air particulate activity and also for iodine in the case of the Seismic Class I Pipe Tunnel. After passing through the air particulate monitors the sample al: is circulated through gas monitors. This pro-vides a complete spectrum of leak detection capability for passive failures in these important areas. The detectors are of the Geiger-Mueller type in a shielded assembly. The detector outputs are transmitted to the Radiation Monitoring System cabinets in the control room where the radio-activity levels are indicated by meters and recorded. High
- radioactivity-alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets.
-6 -2 The sensitivity range of these monitors is 10 to 10 uc/cc.
The passive failure gas monitor is listed in the plant Technical Specifications. 5-19
i 5.2.19 Control Room Intake Air Monitor This channel continuously monitors the outside air intake to the contr61 room. Two monitors, one for particulates and one for iodine, are provided. Upon an increase in activity above the alarm point in either channel, the ventilation system air in-let is closed add make-up air for maintaining a pressurized con-trol room is introduced from the turbine room. The particulate monitor uses moving filter paper and a beta scintillation detector and the iodine monitor employs a fixed filter and gamma scintillation detector. Both of the detectors are housed within shielded assemblies. The detector outputs are transmitted to the Radiation Monitoring i System cabinets in the contro.1 room where the radioactivity levels are indicated by meters and recorded. High radioactivity-alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. The sensitivity range of these monitors is 10-1 to 10 -5 uc/cc for both particulate activity and iodine. The control room intake air monitor is listed in the plant Tech-nical Specifications. 5 2.20 control Room Intake From Turbine Building Monitor , This channel monitors the intake air to the control room when l t the normal air intake is closed and make-up air for pressurizing I l 5-20 9- . - - ,
the control room is being taken from the turbine building. The system consists of a spectrometer grade gamma scintillation de-tector in a shielded assembly for iodine detection. In the event of high radiation detection in the make-up from the turbine building, the emergency make-up filter fan is automatically started and make-up air is introduced through a HEPA filter and charcoal filter for the removal of potential radioactive contamination. The detector output is transmitted to the Radiation Monitoring System cabinets in the control room where the radioactivity level is indicated by a meter and recorded. High radioactivity-alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. to 10 -5
-11 The sensitivity range of this monitor is 10 uc/ce.
The control room intake from turbine building monitor is listed in the plant Technical Specifications. 5.2.21 Reactor Vessel Leakage Monitor This channel monitors the air stream between the reactor vessel and reactor vessel insulation. This system has been designed to indicate activity levels above the background level with consideration for a continually changing background level. The sample air is taken from four locations, 90 degrees apart, around the reactor vessel-at the elevation of the nozzles. Air 5-21 l
is Introduced between the vessel and its insulation from in-leakage around the in-core instrumentation penetrations. This air is part of the ventilating air introduced into the reactor cavity An air sample, for reference purposes, is drawn from the reactor cavity as the' ventilating air enters. The activity level of this sample will vary with the background level in the containment. Each of the four sample locations from the vessel is sequentially compared with the reference sample on a ratio basis. The alarm point is set such that it is always a constant value above the actual background level. The system employs two detectors of the Geiger-Mueller type in shielded assemblies. The detector outputs are transmitted to the Radiation Monitoring System cabjnets in the control room where the radioactivity levels are indicated by meters and recorded. High radioactivity alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets. This monitor has a sensitivity range of 10~ to 10-2microcuries i per cubic centimeter for radioactive noble gases, i No automatic action is initiated by high activity alarms on this j channel. l 5-22
l l The reactor vessel leakage monitor is listed in the plant Tech-nical Specifications. 5.3 FAILED FUEL MONITORING SYSTEM Two different methods are used to detect failed fuel. One channel monitors the radiation level in the chemical and volume control system volume control tank room. In the event of a fuel element failure s the radioacitive noble gas inventory in the volume control tank will increase. This will result in a higher radiation level inside the volume control tank room. The range of this instrument is 1 mR/HR to 1000 R/HR. Predicted levels are as follows: At shutdown 1 mR/HR Operating 100 mR/HR 1% failed fuel 1000 R/HR The second channel monitors the chemical and volume control system letdown flow for activity. This system is based on de-tecting the 1.7 mev gamma from Iodine 135 With 1% failed fuel, 1.84 uc/cc of I-135 should be seen. Either monitor will detect a fuel defect within one or two minutes. Both fa.iled fuel monitors are listed in the, plant Technical Specifications. If both r.onitors are out of service, a reactor coolant iodine analysis must be performed once per chift. 5-23
5.4 CONTAINMENT AREA SAMPLING SYSTEM Activity levels within the containment areas are continuously monitored by means of air particulate and gaseous radioactivity monitors previously described. The presence of a small leak in any of the reactor coolant systems within the containce.nt can readily be detected by means of this monitcring system. The system is equipped with an alarm to alert the operator to any unusual radioactivit,y condition with the containment areas. This alarm system, however, will give no indication of the source of the radicactivity other than the fact that it is from within the containment. In order for the operator to have some in-dication of the source of the leakage, additional sample lines nre installed from v-etous areas of the containment. Four sampling locations inside the missile barrier at approximate elevation 590' are used for this purpose. The sample air is withdrawn through four fixed filter col 19ctors located outside the containment. Air is drawn from the four containment area sampling locations through the filter capsules and discharged through the normal discharge of the containment air monitor. When air has been drawn through the filter capsules for a short period of time, the pump is shut off, all isolation valves are shut and the filter paper removed from the filter capsules and taken to the radiochemistry lab for counting. The filter papers are counted in a beta counter and the one indicating the highest activity level will indicate the approximate location of the l 5-24 - l
.. . . _ = -
E leak so that a search can be made in a more limited area. The sampling technique is continued on a daily basis in order to establish an average air activity for each area. i Bottled air samples of the containment atmosphere can also be taken with this system. For this operation the filters are.by-passed and air is collected from any of the four sampling loca-4 tions. The principal use of the bottled air sample is to anal' for hydrogen and iodine after an accident involving loss of reactor coolant. The same four locations are used since studies indicate that the greatest hydrogen concentration will be at the floor level af the containment near the water surface rather than at the dome level. Operation of either the fixed filter or bottled sample collection is totally manual and can be initiated whenever the operator considers it necessary. . 5.5 AREA RADIATION MONITORING SYSTEM This system consists of channels which primarily monitor and in-dicate radiation levels in various physical areas of the plant. Several channels continuously monitor areas such as the control room, containment, radiochemistry laboratory, and auxiliary building for gamma radiation with a fixed position G-M tube de-tector. The detector output is amplified and the log count-rate is determined by the integral amplifier at the detector. i 5-25
Since the G-M detector system counts pulses, it could saturate in a very high field causing the output meter to read zero. To compensate for this, a special circuit is added to the G-M area monitor which will hold the meter at full scale until manually reset. The radiation level is shown at the detector and is transmitted to tne Radiation Monitoring System cabinets in the control room where it is indicated on a meter and recorded. Radiation alarms are displayed in the control room and locally. Some channels utilize a gamma scintillation type detector with an integral amplifier at the detector. Since this type unit is a current integrating device rather than a pulse system, it is not affected by stray electro-static or electro-magnetic fields. In addition, this detector will not saturate in high radiation fields. When the level exceeds the range of the instrument, it merely reads full scale until such time as the level recedes to a point within the instrument's range. The radiation level is shown at the detector and is transmitted to the Radiation Monitoring System cabinets in the control rcom where it is indicated on a meter and recorded. Radiation alarms are displayed in the control room and locally. The remaining area monitors are of the air particulate type with fixed filter collectors. These units employ a beta scintillation 5-26 .
type detector. The detector output is transmitted to the radi-ation monitoring system cabinets where the radioactivity level is indicated and recorded. The control board annunciator provides a single window which alarms for any channel being alarmed. Verification of which channels has alarmed is done at the Radiation Monitoring System cabinets. Table 5-1 lists the area radiation monitors and gives the type of detector used for each including the instrument range. The only channel which has an automatic function is the fuel handling building pool area monitor which causes the fuel han-dling ventilation exhaust to be routed through the charcoal filters and booster fans in the auxiliary building ventilation system. All radiation monitors are tested for proper operation daily by use of the installed check source. 5-27
l l l Table 5-1 AREA RADIATION MONITORS Total Location Type Rance Number
- Control Room Geiger-Mueller 0.1 mr/hr-10 R/hr 1 Radiochemistry 0.1 mr/hr-10 Laboratory Geiger-Mueller R/hr 1 Primary Sampling 0.1 mr/hr-10 Room Geiger-Mueller R/hr 1
- Fuel Handling Bldg. 0.1 mr/hr-10 Decontamination Area Geiger-Mueller R/hr 1
- Fuel Handling Bldg. 0.1 mr/hr-10 Pool Area Gamma-Scintillation R/hr 1 In-Core Seal Table Geiger-Mueller 0.1 mr/hr-10 R/hr 2 Boric Acid Evaporator Air Sample 10 cpm-10 Cubicles Beta-Scintillation epm 2
- Containment Geiger-Mueller 0.1 mr/hr-10 R/br 2 Gamma-Scintillation 0.1 mr/hr-1 R/hr 2 Waste Drumming 0eiger-Mueller 0.1.mr/hr-10 Station R/hr 1 Gamma-Scintillation 0.1 mr/hr-1 R/hr 1 6
Dry. Active Waste Air Paiticulate 10 cpm-10 Storage Area Beta-Scintillation epm 1
- Gas Decay Tank Air Sample 10 cpm-10 6 Cubicles Beta-Scintillation cpm 1
- Auxiliary Bldg. 0.1 mr/hr-1 HVAC Filter Area Gamma-Scintillation R/hr 1 Radwaste Evaporator Air Sample 10 cpm-10 6 Cubicle Beta-Scintillation cpm 1
- Miscellaneous 0.1 mr/hr-1 Auxiliary Bldg. Area Gamma-Scintillation R/hr 4
" Technical Specification Listing 5-29
e > M N 4 1 ff i i j '. PRESSURIZED WATER REACTOR RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM i i 1 t i i I CHAPTER 6.0 1 t CHEMICAL AND VOLUME CONTROL SYSTEM i s . 1 l 1 1 J I e ) ) i k-t
. _ . , _ . . _ _ . . _ _ . _ _ - _ . . _ , , - _ , ~ _ . . - . _ - . _ , _ . _ _ _ - . --
CHAPTER 6 TABLE OF CONTENTS Secticr Title Pa~e d.0 CHEMICAL AND VOLUME CONTROL SYSTEM . . 6-1
6.1 INTRODUCTION
. . . . . . . . . . . . . 6-1 6.2 GENERAL DESCRIPTION . . . . . . . . . 6-1 6.2.1 Charcing, Letdown and Seal Water . . . 6-1 6.2.2 Chemical Centrol, Purification and Makeup . . . . . . . . . . . . . . . . 6-8 pH Control . . . . . . . . . . . . . 6-3 0xygen Control . . . . . . . . . . . 60 Purification . . . . . . . . . . . . 6-10 Chemical Shim and Makeup . . . . . . 6-12 6.3 DESIGN BASIS AND EQUIPMENT i DESCRIPTION . . . . . . . . . . . . . 6-14
- 6.3.1 Charging, Letdown and Seal Water
. Equipment . . . . . . . . . . . . . . 6-13 Regenerative Heat Exchanger . . . . 6-14 Letdown Orifices . . . . . . . . . . 6-14 Letdown Heat Exchanger . . . . . . . 6-15 Reactor Coolant Filter . . . . . . . 6-16 Volume Control Tank . . . . . . . . 6-16 Pccitive Displacement Charging Pump 6-17 Centrifugal Charging Pumps . . . . . 6-19 3eal Injection Filters . . . . . . . 6-19 Seal Water Filter . . . . . . . . . 6-20 Excess Letdown Heat Exchanger . . . 6-20 Seal Water Heat Exchanger . . . . . 6-21 6.3.2 Chemical Control, Purification and Makeup Equipment . . . . . . . . . . . 6-21 Chemical Mixing Tank and Orifice . . 6-21 Mixed-bed Demineralizers . . . . . . 6-22 Cation Bed Demineralizer . . . . . . 6-23 Batching Tank and Agitator . . . . . 6-21 i
TABLE OF CONTENTS (CONT'D) Section Title Page Boric Acid Transfer Pumps . . . . . 6-24 Boric Acid Tanks . . . . . . . . . . 6-26 Boric Acid Filter . . . . . . . . . 6-26 Boric Acid Blender . . . . . . . . . 6-26 Valves . . . . . . . . . . . . . . . 6-27 6.4 SYSTEM AND EQUIPMENT DESIGN PARAMETER: . . . . . . . . . . . . . . 6-27 6.4.1 Charging, Letdown and Seal Water Ecuipment . . . . . . . . . . . . . . 6-28 Regenerative Heat Exchanger . . . . 6-28 Shell Side . . . . . . . . . . . . 6-28 Tube Side . . . . . . . . . . . . 6-28 Operating Parameters . . . . . . . . 6-28 Letdown Orifices . . . . . . . . . . 6-29 Letdown Heat Exchanger . . . . . . . 6-29 Shell Side . . . . . . . . . . . . 6-29 Tube Side . . . . . . . . . . . . 6-30 Operating Parameters . . . . . . . . 6-30 Shell Side . . . . . . . . . . . . 6-30 Tube Side . . . . . . . . . . . . 6-30 Volume Control Tank . . . . . . . . 6-30 Reciprocating Charging Pump . . . . 6-31 Centrifugal Charging Pump . . . . . 6-31 Excess Letdown Heat Exchanger . . . 6-32 General . . . . . . . . . . . . . 6-32 Shell Side . . . . . . . . . . . . 6-32
. Tube Side . . . . . . . . . . . . 6-32 Seal Water Heat Exchanger . . . . . 6-33 Shell Side . . . . . . . . . . . . 6-33 Tube Side . . . . . . . . . . . . 6-33 6.4.2 Chemical Control, Purification and Makeup Equipment . . . . . . . . . . . 6-33 Mixed-bed demineralizers . . . . . . 6-33 Cation Bed Demineralizer . . . . . . 6-34 Boric Acid Tanks . . . . . . . . . . 6-35 Boric Acid Transfer Pump . . . . . . 6-35 l 11 I
J LIST OF FIGURES a Figure Title Page 6-1 Chemical and Volume Control System . . 6-37 1 i + 4 i 4 iii .
l l i 6.0 CHEMICAL AND VOLUME CONTROL SYSTEM
6.1 INTRODUCTION
The basic functions of the Chemical and Volume Control System (CVCS) follows:
- 1. Maintain programmed water level in the pres-surizer; i.e., maintain required water inventory in Reactor Coolant System (RCS).
- 2. Maintain seal-water injection flow to the reactor
-O coolant pumps.
3 Control reactor coolant water chemistry con-ditions, activity level, soluble chemical neutron absorber concentration and makeup. I
- 4. Provide means of filling, draining, and pressure testing of the RCS.
5 Provide injection flow to the RCS following actuation of the Safety Injection System (SIS). 6.2 GENERAL DESCRIPTION The CVCS is shown on Figure 6-1. 6.2.1 Charging, Letdown and Seal Water The charging and letdown functions of the system are employed to maintain a programmed water level in the 6-1 7-
reactor coolant system pressurizer, thus maintaining proper reactor coolant inventory during all phases of plant operation. This is achieved by means of a continuous feed and bleed process during which time the feed rate is automatically controlled by pres-surizer water level. The bleed rate can be chosen to suit various plant operational requirements by selecting the proper combination of letdown orifices in the letdown flowpath. Reactor coolant is dis-charged to the CVCS from a cold leg of the RCS; it then flows through the shell side of the regenerative heat exchanger where, during normal operation, its temperature is reduced to approximately 290 F. The coolant then experiences a large pressure reduction in passing through a letdown orifice (a = 1900 psi) p and after passing through the containment boundary it flows through the tube side of the nonregenerative letdown heat exchanger where its temperature is further reduced to about 115 F. Downstream of the letdown heat exchanger a second pressure reduction occurs as the coolant flows to the purification system demineralizers. This pressure reduction is performed by the low-pressure letdown valve, the function of which is to maintain an upstream pressure of about 300 - 350 psig which prevents flashing counstream of the letdown orifices. i 6-2
The coolant then flows normally through one of the mixed-bed demineralizers through the reactor coolant filter, and into the volume control .t.ank via a di-version valve and finally a spray nozzle in the gas space of the tank. The volume control tank normally operates with a gas-to-liquid volumetric ratio of about 2:1. The gas space in the volume control tank is filled with hydrogen, which is regulated to a pressure of 15-20 psig during normal plant operation. The partial pressure of hydrogen in the volume con-trol tank determines the concentration of hydrogen dissolved in the reactor coolant. An alternate letdown path is provided which allows part of all of the letdown flow to pass through the Boron Thermal Regeneration System (BTRS) when boron concentration changes are desired to follow plant load. The alternate letdown flow path is directed to the BTRS downstream of the mixed bed demineralizers. After processing by the BTRS, the flow is returned to the CVCS at a point upstream of the reactor coolant filter. The charging pumps normally take suction from the volume control tank and return the cooled, purified reactor coolant to the reactor coolant system via 6-3
the charging system. The charging pumps discharge at a pressure dictated by the prevailing reactor coolant system pressure, the resistance of the charging line, and the pressure drop impressed by the positioning of an air-operated control valve situated in the charging line (normally the pressure will-be about 2350 psis). Normal charging flow is handled by the single positive disple- .nent recip-rocating charging pump. The flow rate will be dependent upon the speed of the positive displace-ment charging pump, controlled either by pressurizer level requirements or by operator choice. If the reciprocating charging pump reaches the high speed limit, it becomes necessary to place a centrifugal pump in operation to provide the higher flow capacity and to remove the reciprocating pump from service. The flow rate for the centrifugal charging pump is controlled by a modulating valve in the pump dis-charge line. The charging flow controller maintains the preset charging flow, which is reset by the pressurizer level requirements. A minimum flow for the centrifugal charging pump protection is continuously diverted from the charging pump discharge bcck to the volume control tank through the seal water heat exchanger.
)
- 6-4
6 l The bulk of the charging is pumped back to the re-actor coolant system via the tube side of the regen-erative heat exchanger where the outlet temperature approaches the reactor coolant temperature. The flow is then injected into a cold leg of the RCS. Two redundant injection paths are provided for rapid boration of the system. A flow path is also pro-vided from the regenerative heat exchanger outlet to the pressurizer spray line. An air operated valve in the line is employed to provide auxiliary spray to the vapor space of the pressurizer. It is employed during cooldown to supplement the spray to the vapor space of the pressurizer. It is employed during cooldown to supplement the spray from the reactor coolant system, and thus provides a rapid means of cooling the pressurizer near the end of plant cool-down, when the reactor coolant pumps ase not operating. The remainder of the charging flow is directed to the reactor coolant pumps via the seal-water-injection filters. It enters the pumps at a point between the labyrinth seal and the No. 1 seal. Here the flow splits and a portion enters the RCS via the labyrinth seals and thermal-barrier-cooler cavity. The re-mainder of the flow flows up the pump shaft (cooling 7 6-5
i . . the lower bearing) and leaves the pump via the No. 1 seal. The labyrinth flows are removed from the RCS as a portion of the letdown flow. The No. 1 seal discharges flow to a common manifold, exits the containment, and then passes through the seal-water-filter and the seal-water heat exchanger to the suction side of the charging pumps,. or by alternate path to the-volume controlLtank. (Note that the
-back-pressure on the No. 1 seal is the summation of flow resistance in the seal-water return lines and the charging pumps suction pressure.)
An alternate letdown path from the RCS is provided l in the event that the normal letdown path is inoperable. Reactor coolant can be discharged-from a cold leg
- and flows through the tube side of the excess let- '
down heat exchanger, where it is cooled to about 165 F. Downstream of the heat exchanger.a remote-manual control valve controls the excess letdown flow. The flow then normally joins the No. 1 seal discharge manifold and passes through the seal-j water-filter and heat exchanger to the volume control i tank. This excess letdown can also be directed to the reh'ctor coolant drain tank so that it bypasses the No. 1 seal return manifold. The excess letdown 4 6-6 L
. . . _ . _ , . _ _ _ _,_.._.a. . , . . _ _ _ _ __. _ _ _ , _ . ~
i i i flow path can also be used to maintain normal heatup rate of the plant, by providing additional letdown capability during the final stages of heatup. This path removes some of the excess reactor coolant due to expansion of the system as a result of the reactor coolant system temperature increase. Surges in reactor coolant system volume due to- load changes are accommodated for the most part in the pressurizer; however, the volume control tank is designed to accommodate programmed pressurizer level mismatches which may occur due to a + 4 F temperature error. High water level in the volume control tank actuates a hi-level alarm and causes the letdown flow normally entering the tank to be c'iverted to the recycle holdup tanks (Boron Recycle System) (BRS). Low level in the velume control tank initiates make-up from the Reactor Makeup Control System (RMCS). If the RMCS does not supply sufficient makeup to keep the volume control tank level from falling to a lower level, a low level alarm is actuated. A low-low level signal causes the suction of the charging pumps to be transferred to the refueling water storage tank. 6-7
l l 6.2.2 Chemical Control, Purification and Makeup The water chemistry, chemical shim and makeup re-quirements of the RCS are such that the following functions must be provided:
- 1. Means of addition and removal of pH control chemicals for startup and normal operation.
- 2. Control of oxygen concentration following venting and that due to.radiolysis in the core during normal operation.
3 Means of purification to remove corrosion and fission products.
- 4. Means of addition and removal of soluble chemical
- neutron absorber (Boron) and makeup water at concentrations and rates compatible with all phases of plant operation including emergency situations.
pH Control The chemical control element employed for pH contrdl is lithium hydroxide (Li 0H). This chemical is chosen 7 for its compatibility with the materials and water chemistry of borated water / stainless steel / irconium systems; in addition, Li-7 is produced in the core l
)
6-8 l
i region due to irradiation of the dissolved boron in the coolant. The Li 0H is introduced to the reactor 7 coolant system via the charging flow. A chemical mixing tank is provided to introduce the solution to the suction of the charging pumps. The solution is prepared in the laboratory and poured into the chemical mixing tank. Reactor makeup water is then used to flush the solution to the suction manifold 4 of the charging pump.
! The concentration of Li 0H in the reactor coolant 7
system is maintained between 0.2 - 2.2 ppm Li-7. If the concentration exceeds this value, as it may well do during the early stages of core life, the cation bed demineralizer is employed in the let-down in series operation with a mixed-bed demineralizer. When the BRS is being utilized, the cation bed demineralizer should pass all the letdown to remove as much LI-7 and cesium as possible before the :ater , is diverted to the BRS. Oxygen Control During initial plant startup from the cold condition hydrazine is employed as an oxygen scavenging agent. l The hydrazine solution is introduced to the reactor 6-9
coolant system in the same manner as described above for the pH from the cold shutdown state. A second mechanism is employed to control and scavenge oxygen produced due to radiolysis of water in the core region. This mechanism is the provision of hydrogen in the reactor coolant. Sufficient partial pressure of hydrogen is maintained-in the volume control tank, such that the equilibrium concentration of between 25 .35 cc of hydrogen per kg of reactor coolant is maintained in the reactor coolant system. Hydrogen is supplied from the hydrogen manifold in the Waste Processing System (WPS) and a pressure control valve maintains a minimum pressure of about 15 to 20 psig in the ' vapor space of the volume control tank. This regulator can be adjusted to provide the correct 4 equilibrium hydrogen concentration. Purification Mixed-bed demineralizers are provided in the letdown system to provide cleanup of the letdown flow. The demineralizers remove ionic corrosion products, certain fission products, and act as filters. One demineralizer is usually in continuous service for normal letdown flow and can be supplemented inter-l mittently by the cation bed demineralizer for 1
/
6-10
additional purification in the event of fuel defects. In this case, the cation resin will remove principally cesium isotopes. The demineralizer has a sufficient capacity to maintain the cesium concentration in the reactor coolant below 1.0 pc/cc with 1 percent de-fective fuel. The design life of the mixed-bed resins is one core cycle. Each mixed-bed demineralizer is sized to accept maximum letdown purification flow (120 gpm). , A further cleanup feature is provided for use during cold shutdown and residual heat removal. A remote-operated valve admits a bypass flow from the Residual Heat Removal System (RHRS) into the letdown line up-stream of the letdown heat exchanger. The flow passes through the heat exchanger, through a mixed-bed demineralizer and a reactor coolant filter to the volume control tank. The volume control tank atmosphere at this time will be nitrogen. The fluid is then returned to the reactor coolant system via the normal charging route. Spent resins (changes for reasons of pressure drop or activity level) are initially fluidized by back-flushing with water and are then flushed to the spent resin storage tank in the Waste Processing System. d 6-11 l I
Filters are provided at various locations to ensure filtration of particulate and resin fines and to protect the seals on the reactor coolant pumps. Chemical Shim and Makeup The function of soluble neutron absorber (boron) concentration control and makeup is provided by the RMCS employing 4 wt. percent boric acid solution and reactor makeup water from the Reactor Makeup Storage Tank. In addition; for emerge.cy beraticn and makeup the capability exists to provide ret'ueling water or 4 wt. percent boric acid to the suction of the charging pump. Initial filling and makeup quantities of 3 ut. percent boric acid solution are prepared in the boric acid batching tank where boric acid crystals are dissolved in het water and pumped to the boric acid storage tank. The batching tank is steam-heated and if re-quired, it heats the contents to the desired te.7-perature (85 ?) at which the 4 wt. parcent eciation is prepared. The batch is transferred to the boric acid tan: e; the transfer pumps. Soric acid ic stared 5: rec -. boric acid tanks. The tanks , which are in 'cc ta:' . ten c 1 l l L 6-12 I I
l l l that are maintained at a temperature greater than or equal to 65 F, share the boric acid transfer pumps. A pump can be periodically run to recirculate the i tanks contents through the beric acid filter back to the tank. On a demand signal by the RMCS, the 4 pamp starts and delivers boric acid to the boric I acid blender. A standby boric acid nump is prcvided. The reactor makeup water pumps, taking suction from P the reactor makeup water storage tank, are emplcyed for various makeup and flushing operations throughout the systems. One of those pumps also starts on demand from the RMCS. The flow of boric acid from the boric acid transfer
- pump and the reactor makeup water from the reactor I
makeup water pump is directed to a blending device. The flow is then directed to either the suction manifold of the charging pumps or is spraved into the volume control tank through the spray nozzle. i The normal flow path will be the line to the volume control tank where hydrogen pickup will be assured during long dilution processes. In the event that xenon transients require rapid boration, the direct line'to the charging pumps suction can be used. i 6-13 e -
-- -. , ., , ----,.-,,-..n - . , - , ,, .-w,...,-,-. c, .nn,-.
6.3 DESIGN BASIS AND EQUIPMENT DESCRIPTION 6.3.1 Charginz, Letdown and Seal Water Equioment Regenerative Heat Exchanger The regenerative heat exchanger is designed to re-cover heat from the letdown flow by reheating the charging flow, to eliminate reactivity effects due to insertion of cold water, and to reduce thermal shock on the charging line penetrations into the reactor-coolant-loop piping. The design also considers the limitation due to the temperature difference which occurs when the letdown flow exceeds the charging flow. This case increases the possibility of flashing downstream of the let-down orifices. The letdown stream flows through the shell side of the unit and the charging stream flows through the tubes; this places the lower design pressure limit on the shell. The unit is constructed entirely from austenitic stainless steel and is of all-welded construction. Letdown Orifices Three letdown orifices reduce the coolant pressure from reactor conditions and control the flow of i 6-14 i i
reactor coolant leaving the RCS. The orifices are placed into or out of service by remote operation , of their respective isolation valves. One orifice is designed for normal operating flow with the other two serving as standby. One or both of.the standby orifices may be used in parallel with the normally operating orifice for either flow control when the RCS pressure is less than normal, or greater letdown flow during maximum purification or heatup. Each orifice consists of an assembly which provides for permanent pressure loss without recovery, and is made of austenitic stainless steel or other adequate corrosion resistant material. Letdown Heat Exchanger The letdown heat exchanger cools the letdown flow to 115 F, such that the water discharged from a, centri-fugal charging pump will be cool enough to inject into the reactor coolant pump #1 seals. The letdown flows through the tube side of the single shell multi-pass heat exchanger while component-cooling water flows through the shell cide. The letdown stream outlet temperature is automatically controlled by a temperature-control valve in the component-
~
cooling-water outlet stream. 6-15
The surfaces of the unit in contact with the reactor coolant are austenitic stainles steel and the shell side is carbon steel. The letdown operating pressure is maintained between 300-350 psig by a pressure-control valve located downstream of the unit. Reactor Coolar.t Filter A reactor coolant filter is designed to collect resin fines and particulate matter larger than 25 microns from the letdown stream. The filter is located be-tween the demineralizers and the volume control tank diversion valve. Disposable synthetic cartridge forms the filtering media, and the vessel is constru ted of austenitic stainless steel. The unit designed to accept maximum design letdown flow. Volume Control Tank The volume control tank provides surge capacity for part of the reactor coolant not accommodated by the _pressuriner following load transients (only due to RCS temperature errors). When the volume control tank reaches the high level setpoint, the letdown stream is diverted to the recycle holdup tanks. In addition, the volume control tank provides a means for introducing hydrogen into the coolant, to main-l tain the required equilibrium concentration, the i l 6-16
volume control tank is used for degassing the re-actor coolant during shutdown, and it serves as a head tank for the charging pumps. An inlet not:le assures intimate contact to equi-librate the incoming fluid and the hydrogen atmosphere in the tank. The tank is provided with level con-trols which are described in the section on Instru-mentation and Control. During normal plant operation, the hydrogen and fission gases in the tank are con-tinuously purged to the WPS, to limit the release of radioactive gases through leakage by maintaining the radioactive gas level in the reactor coolant several times lower than the equilibrium level. Also during degassing,~the normally locked closed valve from the seal water exchanger may be opened to allow miniflow-from the charging pumps to pass through the auxiliary spray no le on the volume control tank. This increases the stripping efficiency in the tank. Relief protection, sampling and nitrogen purge con-nections are also provided. The tank is constructed of austenitic stainless steel. Positive Displacement Charging Pump This charging pump takes suction from either the - volume control tank, the RMCS, the refueling water 6-17
._ _. - .- _ . .. .-_ . _ - . ~ . - = _ - -. . - _- -._
stor' age tr ., or the emergency boration line. Normally, this rump takes suction from the volume control tank and delivers charging flow to the RCS and reactor coolant pump seals. The pump is a positive displacement, variable-speed pump; its speed ratio is compatible with minimum flow requirements. A low-speed stop is provided to prevent the flow rate from falling below the required flow to the reactor coolant pump seals. The speed control feature is described in the Section on Instrumentation and Control. All parts of the pump in contact with reactor coolant are constructed of austenitic stainless steel. i Special, low-chloride content packing is used in the pump glands. The capacity of this pump is sufficient to provide normal charging (55 spm), seal water flow (32 gpm), plus a margin for leakage or increased seal-water requirements. The pump design pressure is chosen
- to assure rated flow for all ccnceivable reactor coolant system pressures up to the safety valve settings, plus the calculated
- charging system pressure drop at maximum flow, and hydrotest requirements, i
s ! 6-18 i
Centrifugal Charging Pumps The two centrifugal charging pumps take suction from either the volume control tank, the RMCS, the refuel-ing water storage tank, or the emergency boration line. Normally these pumps are not operating, but in the event that the charging flow requirements exceed the capacity of the positive displacement charging pump one of the centrifugal pumps is put on the line and the positive displacement charging pump is shut down. During a loss-of-coolant accident, both centrifugal charging pumps operate as part of the Safety Injection System (SIS), and therefore take suction from the refueling water storage tank. This charges 2000 ppm borated water to the boron injection tank. All parts of the pump in contact with the reactor coolant are constructed of austenitic stainless steel. Seal Injection Filters Two seal injection filters (Figure 6-1) are located in parallel in a common line to the reactor coolant pump seals; they collect particulate matter that could be harmful to the seal faces. Each filter is s* ed to accept in excess of the normal seal water-flow .eosirements. 6-19
__ . . . . _ _ _ . ~ _ _ _ _ _ . _ . - _ _ _ . _ . . _ . _ _ . _ _ _ _ _ . _ . _ _ . _ _ _ _ _ _ _ _ . . _ . _ . . . 4 1 Disposable synthetic filter elements are employed and the units are constructed from austenitic stain-less steel. Seal Water Filter The seal water filter collects particulate from the reactor coolant pump seal water return and the excess , j letdown flow. The filter is designed to pass the sum of excess letdown flow and the maximum design leakage from the reactor coolant pumps. The vessel is constructed of austenitic stainless steel. A 4 disposable synthetic filter element is used. Excess Letdown Heat Exchanger The excess letdown heat exchanger cools letddwn flow equivalent to that portion of the nominal seal in- ! jection flow which flo'ws into the RCS through the reactor coolant pump labyrinth seals. It may be i
! employed either when the normal letdown path is i temporarily out of service or it can be used to sup-i i
plement maximum letdown during the final stages of heatup. The unit is designed to reduce the letdown 1 { stream temperature from the cold-leg temperature to i 165 F. The letdown flows through the tube side of the unit and component-cooling water is circulated through the shell. All surfaces in contact with j l ! i } 6-20
reactor coolant are austenitic stainless steel and the shell is carbon steel. Seal Water Heat Exchanger The seal water heat exchanger is designed to cool fluid from three sources: reactor coolant pump seal water returning to the volume control tank, re-actor coolant discharged from the excess letdown heat exchanger and miniflow from a centrifugal charging pump. Reactor coolant flows through the tube side of the heat exchanger and component-cooling water is circulated through the shell. The design flow rate is equal to the sum of the nominal excess letdown flow, the maximum design reactor coolant pump seal leakage and miniflow from one centrifugal charging pump. The unit is designed to cool the above flow to the temperature normally maintained in the volume control tank. All surfaces in contact with reactor coolant are austenitic stainless steel and the shell is carbon steel. 6.3. Chemical Control, Purification and Makeup Eauipment Chemical Mixing Tank and Orifice The chemical mixing tank is sized to ensure sur-ficient capacity of injecting a solution of 35 percent 6-21
p _ __ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ . . _ _ _ _ _ _ - _ _ _ _ _ _ _ ._ i 1 hydrazine necessary to provide a concentration of 10 ppm in the cold RCS for oxygen scavenging. This capacity far exceeds that requ-ired for addition of caustic solution (Li 0H) for pH control. The tank 7 is constructed from austenitic stainless steel and is provided with fill, vent and drain connections. An orifice is provided in the piping upstream of the mixing tank. This orifice limits the flow rate through the tank to 2 gpm to avoid slugging the pump seals with concentrated chemicals. i Mixed-bed Demineralizers Two flushable, mixed-bed demineralizers maintain reactor coolant purity. A lithium-type cation resin and hydroxl type anion resin are initially charged into the demineralizers. Both forms of resin remove fission and corrosion products. Initially the anion 4 resin is converted to the borate form. The resin-bed size is chosen to provide a decontamination J factor of ten for most fission products (exceptionc i are cesium, yttrium, and molybdenum). Each demin-eralizer is sized to accept maximum letdown flow
, and has a minimum design life of one core cycle.
The demineralizer vessels are constructed entirely from austenitic stainless steel and are provided I l 1 l 6-22 I i
. - _. - - - ~. -- - _ - . . . - _.
with connections for resin addition, replacement, flushing, and draining. The vessel incorporates retention screens, and inflow deflector and mesh , screens on drain connections; the screens are de-signed to retain the resin, but to minil.1:e pressure drop. Cation Bed Demineralizer A flushable cation resin bed in the hydrogen form is located downstream of tne mixed-bed deminera3.ners and is used intermittently to control the concentration of Li-7 (pH control) in the reactor coolant system Its site is based upon the estimated production of Li-7 in the core reg. ion due to the B10 (y,,),L1 7 reaction during base-load operation. The resin also ! has sufficient capacity to maintain the cesium con-centration in the coolant below 1.0 pc/cc with one percent defective fuel. The demineralizer is sized to accommodate maximum letdown flow when in service, which is more than adequate to control Li-7 and/or cesium concentration in the reactor coolant system. Its construction is identical to that of the mixed-bed demineralizer.. l Batching Tank and Agitator The batching tank is used in the preparation of fresh 4 wt. percent boric acid solution. The tank is l 6-23 l
provided with a steam-heated jacket, whose area is , sufficient to heat the tank from ambient tempersture to 90 F in less than three hours when employing process steam at 15 psig saturated. The tank is provided with drain connections, a fill line for de-mineralized water, and a screen-protected filling no le for introduction of boric acid crystal. An electrically-driven, locally controlled agitator to ensure complete mixing is included with the tank. A local sample point is provided in the discharge piping. The tank is constructed from austenitic stainless steel. Boric Acid Trahsfer Pumps ,Two canned motor pumps are supplied per unit. A twin unit plant has three. Normally one pump is aligned with one boric acid tank, and manually or automatically starts on demand from the RMCS ac described under Instrumentation and Control. Mini-flow from this pump flows back to the boric acia tank and helps maintain thermal equilibrium. The second pump is then considered as a standby pump, with service being transferred as operation requires. This second pump can intermittently circulate toric acid solution through the tank to maintain thermal equilibrium in this part of the system. Emergency B 6-24
beration, the supplying of 4 wt. percent boric acid solution to the suction of the charging pumps, can be accomplished by manually choosing either or both pumps. The transfer pumps also function to transfer 4 wt. percent boric acid solution from the batching tank to the boric acid tanks. The design capacity of each pump is equal to the normal letdown flow, with the capacity of two pumps being equivalent to the normal design capacity of ; one centrifugal charging pump. The design discharge pressure is sufficient to overcome any pressures which may exist in the suction manifold of the charging pumps (volume control tank relief valve . setting). In addition to the automatic actuation by the makeup control system, and manual actuation from the main control board, these pumps may also be controlled locally at a local control center. The pumps are placed in a heated room (65 F) to prevent crystallization of the 4 wt. percent boric ' acid solution. All parts in contact with the solution are of austenitic stainless steel. Con- , nections are provided to enable the use of these pumps to flush the equipment and piping with re-actor makeup water. i 6-25
i Boric Acid Tanks
~
Two boric acid tanks are provided. The combined capacity of the tanks contains sufficient boric acid to provide for each unit sufficient boric acid for refueling plus enough boric acid for one cold shutdown immediately following refueling with the most reactive control rod withdrawn. > The tanks are placed in heated roems (65 F) . The tanks are constructed from austenitic stainless steel and include a diaphragm to exclude oxygen. Boric Acid Filter-The boric acid filter collects particulates from the boric acid solution being pumped to the blender in the RMCS, the emergency boration line, and the boric acid tank. The filter is designed to pass the design flow of two boric acid transfer pumps operating simultaneously. The vessel is constructed of auster-itic stainless steel and the filter elements are disposable synthetic fiber cartridges. Boric Acid Blender The boric acid blender, made of austenitic stainless steel, is provided to ensure thorough mixing of the 4 wt. percent solution of boric acid and reactor make-up water when required. The blender consists of a 1 l 6-26 l l i L
conventional piece of pipe fitted with a perforated tube insert. A sample point is provided in the piping just downstream of the blender, j Valves All Chemical and Volume Control System (CVCs) valves I
- which are larger than 2 inches and which are desig-nated for radioactive service are provided with a
{ stuffing box and lantern leakoff connections. All l control (modulating) and three-way valves are either 5 provided with stuffing box and leakoff connections or are totally enclosed. Leakage to the atmosphere is essentially zero for these valves. Basic material i
; of construction is stainless steel for all valves except the batching eteam jacket valves which are l carbon steel.
Isolation valves are provided at all connections to the RCS. Lines entering the reactor containment also normally have check valves inside the containment i to prevent reverse flow from the contain:..ent. 6.4 SYSTEM AND EQUIPMENT DESIGN PARAMETERS
, System Parameters:
- Seal' water supply flow rate, spm 32 Seal water return flow rate, spm 12 i
6-27 i I
---m _ _m._A_____._____________.___.____.___...___._ _ - _ . . _ _ _
Normal letdown flow rate, gpm 75
!ormal charging flow rate, spm 55 4 Maximum letdown flow rate, gpm 120 Temperaturg of reactor coolant entering system, F <560 Normal coolant discharge temperature to recycle holdup tanks, UF =115 0 6.4.1 Cnarging, Letdown and Seal Water Eauipment Regenerative Heat Exchinger Heat transfer rate at normal conditions, g Stu/hr 11.0 x 10" Shell Side Design pressure, psis ,
2485 Design temperature, F 650 Material of construction Austenitic stain-less steel Tube Side Design pressure, psig 3100 Design temperature, F 650 Material of construction Austenitic stain-less steel Onerating Parameters Shell Side (Letdown) Normal Flow, lb/hr 37,300 Inlet temperature, OF 560 6-28
Outlet temperature, F 290.0 Tube Side (Charging) Flow, lb/hr 22,400 Inlet temperature, F 130.0 Outlet temperature, F 516 Letdown Orifices Design pressure, psig 2485 Design temperature, F 650 Normal operating inlet pressure, psig 2200 Normal operating temperature, F 290 Materials of construction Austenitic stain-less steel 75 gpm orifice 1 Design flow, lb/hr 37,300 Differential pressure at design flow, psi 1900 45 gpm orifice 2 Design flow, 1b/hr 22,200 Differential pressure at design flow, psi 1900 Letdown Heat Exchanger Heat transfer rate at design conditions (Heatup), Btu /hr 16.1 x 10 6 Shell Side Design pressure, psig 150 Design temperature, F 250 Material of construction Carbon steel 6-29
Tube Side Design temperature, F 150 Design pressure, psig 6CO Material of construction Austenitic stain-less steel Operatine Parameters Shell Side Flow, lb/hr Normal Inlet temperature, F 105 Outlet temperature, F Tube Side Flow, lb/hr 37,300 Inlet temperature, F 290 Outlet temperature, F 115 Volume Control Tank Number 1 Internal volume, ft 3 400 Design pressure, internal, psig 75 Design pressure, external, psig 15 Design temperature, F 250 Operation pressure range, psig 15-60 Normal operating temperature, F 115 Spray nozzle pressure drop, psi 15
)
6-30 _ . . . . . .~-_,--. ,_ , _ . - - ,_
)
Letdown nozzle at 120 gpm 8 Seal return no::le at 90 gpm Austenitic stain-less steel Reciorocating Charging Pump Number 1 Positive displace-ment with variable speed
- peed reduction ratio 6: 1 Design pressure, psig 3200 Design temperature, F 300 Normal operating temperature, F 115 Design flow rate, each, spm 98 Design head, ft. 5800 Material of construction Austenitic stain-less steel Hydrostatic test pressure, psis 3125 Centrifugal Charging Pump Number 2 Type Horizontal centrifugal Design pressure, psig 2800 Design temperature, F 300 Normal suction temperature, F 115 Material Austenitic stain-less steel Normal fluid Borated r^ actor coolant 6-31
Fluid during recirculation after LOCA Radioactive borated water with H, and NaOH in solution Design flow rate, gpm 150 Design head, ft 5800 Required HPSH, ft 15 Excess Letdown Heat Exchancer General Heat transfer rate at design 6
% conditions, Btu /hr 5.2 x 10 Shell Side Design pressure, psig 150 Design temperature, F 250 Design flow rate, lb/hr 129,000 Design inlet temperature, F 105 Design outlet temperature, F 145 Material of construction Carbon steel Tube Side Design pressure, psig 2485 Design temperature, F 650 Design flow rate, lb/hr 12,400 Design inlet temperature, F 560 Design outlet temperatur5, F 165 Material of construction Austenitic stain-less steel 6-32
l Seal Water Heat Exchanger Heat transfer rate at design conditions, Btu /hr 1.88 x 10 6 Shell Side Design pressure, psig 150 Design temperature, F '250 Design flow, lb/hr 157,000 Design operating inlet temperature, F 105 Design operating outlet temperature, OF Material of construction Carbon steel Tube Side Design pressure, psig 150 Design temperature, F 250 Design flow, lb/hr 51,900 Design opgrating inlet temper-ature, F 151.2 Design operating outlet temper-ature, OF 115 Material of construction Austenitic stain-less steel 6.h.2 Chemical Control, Purification and Makeup Ecuipment Mixed-bed Demineralizers Number 2 Type Flushable 6-33 h -.
Vessel design pressure, ext. psig 300 Vessel design pressure, int., psig 15 Vessel design temperature, F 350 Resin volume, each, ft 3 33,9 Vessel volume, each, ft 3 a3,9 Bed depth, ft 5.5 Bed diameter, inc. 31 5 Design flow rate, gpm :. Resin bed and vessel pressure drop for 120 gpm flow (fouled condition), psi 12.8 Minimum decontamination factor for ions removed 10 Upper and lower retention screen U.S., mexh 140 (150 micron) Normal operating temperature, F 115 Normal operating temperature, psig <l50 Resin type Rohm & Hass Amberlite IRN-217 or equivalent Material of construction Austenitic stain-less steel (I-Cation Bed Demineralizer Number 1 Type Flushable Vessel design pressure, internal, psig 300 Vessel design pressure, external, psig 15 l l ',', / 6-34 U: i,1_ ._ _ , -
l l l Vessel design temperature, F 250 t Resin volume, ft' 30 Vessel volume, ft 3 43 Bed depth, ft 5.5 Bed diameter, in. 31 5 ' Design flow rate, gpm 120 Resin bed and vessel pressure drop at design flow, ps$ 12.8 Upper and lower retention screen U.S., mesh 140 (105 micron) Normal operating temperature, F 115 Normal operating pressure, psig <150 Resin type Rohm & Hass Amber 11te IRM-77 or eauivalent Material of construction Austenitic stain-less steel Boric Acid Tanks Number per unit 2 Capacity, each, gallons usable 24,000 (46,000 twin units) Design pressure Atmospheric Design temperature, F 200 Normal operating temperature, F 65 Material of construction Austenitic stain-less steel Boric Acid Transfer Pump Number per unit 2 single (3 twin unit) Type Centrifugal canned motor j 6-35
,- , , .1
Design flo's rate, gpm 75 Design hr.ad, ft 235 Design pressure, psig 150 Design temperature, F 250 Temperature of pumped fluid, F 65 Material of construction Austenitic stain-less steel 6-36 g . _ _ .
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c PRESSURIZED WATER REACTOR
-RADIDACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM CHAPTER 7.0
. BORON THERMAL REGENERATION SYSTEM s f I I, . , , . , , -,-.,..-,,-c~. , ~ . - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - ~ ~ ~ , - - ~ - - - - - - - - - - - ' ~ - - - " "
CHAPTER 7 TABLE OF CONTENTS Section Title Page 7.0 BORON THERMAL REGENERATION SYSTEM . . 7-1
7.1 INTRODUCTION
. . . . . . . . . . . . 7-1 7.2 GENERAL DESCRIPTION . . . . . . . . . 7-1 73 DESIGN BASIS AND EQUIPMENT DESCRIPTION . . . . . . . . . . . . . 7-6 Chiller Pumps . . . . . . . . . . . 7-6 Heat Exchanger . . . . . . . . . . 7-7 Moderating Heat Exchanger . . . . . 7-7 Letdown Chiller Heat Exchanger . . 7-7 Letdown Reheat Heat Exchanger . . . 7-8 Chiller Surge Tank . . . . . . . . 7-10 Thermal Regeneration Demineralizars 7-10 Chiller Package . . . . . . . . . . 7-11 7.4 CYSTEM AND EQUIPMENT DESIGN PARAMETERS . . . . . . . . . . . . . 7-13 ! System Parameters . . . . . . . . . 7-13 Chiller Pump . . . . . . . . . . . 7-13 Moderating Heat Exchanger . . . . . 7-13 Letdown Chiller Heat Exchanger . . 7-14 Letdown Reheat Heat Exchanger . . . 7-15 Chiller Surge Tank . . . . . . . . 7-15 Thermal Regeneration Demineralj ers 7-15 Chiller . . . . . . . . . . . . . . 7-16 i
i e 1 I
LIST OF FIGURES Figures Title Page 7-1 Dilution Boron Thermal Regeneration System . . . . . . . . . . . . . . . . 7-17 7-21 Borate Boron Thermal Regeneration System . . . . . . . . . . . . . . . . 7-19 11
7.0 BORON THERMAL REG'ENERATION SYSTEM
7.1 INTRODUCTION
The Boron Thermal Regeneration System (BTRS) varies the Reactor Coolant. System (RCS) boron concentration to ecmpensate for xenon transients and other re-activity changes which occur when the reactor power is changed. The standard load follow cycle is 12 hours at full power, a 3 hour ramp to 50 percent power, 6 haurs at 50 percent power and a 3 hour ramp re-turning to 100 percent power. Immediate return to full power capability is maintained to 85 percent of core life. 7.2 GENERAL DESCRIPTION The BTRS is shown on Figure 7-1 and 7-2. The system, which operates in conjunction with the Chemical and Volume Control System (CVCS), consists mainly of demineralizers, one chiller package, one chiller pump, three heat exchangers, valves, and associated piping for each unit. This equipment controls the flow rate through the BTRS and also controls the temper-ature of the flow stream entering the demineralizers and the volume control tank. During normal operation of the CVCS, the letdown flow from the RCS passes through the regenerative heat 7-1
exchanger, letdown he'at exchanger, mixed bed (puri-fication) demineralizers, reactor coolant filterc and volume control tank. The charging pumps then
~
take suction from the volume control tank and return the purified reactor coolant to the RCS. An alternate letdown path is provided which allows part or all of the letdown flow to pass through the BTRS resin when boron concentration changes are required to follow plant load. This alternate let-down flow path is directed to the BTRS downstream of the mixed bed demineralizers. After processing
. by the BTRS, the flow is returned to the CVCS at a point upstream of the reactor coolant filter.
Storage and release of boron during load follow oper-ation is determined by the temperature of the fluid entering the BTRS demineralizers. A group of heat exchangers and chiller units is employed to provide the desired coclant temperature at the demineralizer (ion exchanger) inlets for either storage or release operations of the system. t The flow path through the BTRS is different for storage and release. During storage, the letdown stream enters the moderating heat exchanger and from there it passes through the letdown chiller heat 7-2
exchanger. These two heat exchangers cool the let-down steam prior to entering the demineralizers. The letdown reheat heat exchanger is valved out on the tube side and performs no function during storage. The temperature of the letdown stream at the point of entry to the demineralizers is con-trolled automatically by the temperature control valve which controls the shell side flow to the let-down chiller heat exchanger. This valve is control-led by a thermocouple located between the letdown reheat heat exchanger and the demineralizers. After passing through the demineralizers, the letdown enters the moderating heat exchanger shell side, where it is heated by the incoming letdown stream before going to the volume control tank. Therefore, for storage, a decrease in the boric acid concentration in the reactor coolant is accomplished by sending the letdbwn flow at relatively low tem-perature (50 ) to the thermally regenerable demin-eralizers. The resin which was depleted of boron at high temperature during a prior boron release operation, is now capable of storing boric acid from the low temperature letdown stream. Reactor coolant with a decreased concentration of boric acid leaves the demineralizers and is directed to the CVCS. 7-3
During release the letdown stream is entering the moderating heat exchanger tube side, bypassing the letdown chiller heat exchanger and passing through the shell side of the letdown heat exchanger. The moderating and letdown reheat heat exchangers are heating the letdown stream prior to entering the resin beds. The temperature of the letdown at the point of entry to the demineralizers is controlled
'oy adjusting the flow rate on the tube side of the letdown reheat heat exchanger; the temperature con-trol valve receives its signal from a thermocouple located between the demineralizers and the letdown
~ reheat heat exchanger. After passing through the demineralizers, the letdown stream enters the shell side of the moderating heat exchanger, passes through the tube side of the letdown chiller heat exchanger and then goes to the volume control tank. The tem-perature of the letdown stream entering the volume control tank is controlled automatically by adjust-ing the shell side flow rate on the letdown chiller heat exchanger; the control valve for this flow receives its signal from a thermocouple located in the line leading to the volume control tank. Thus, for release, an increase in the boric acid con-centration in the reactor coolant is accomplished by s
~
s 7-4
h-8 sending the letdown flow at relatively high temper-atures (1400) to the thermally regenerable demineral-izers. The water flowing through the demineralizers i now releases boron which was storea on the resin at low temperature during a previous boron storage operation. The boron enriched reactor coolant is
. returned to the RCS vis the CVCS.
The letdown reheat heat exchanger heats the letdown j flow to the desired temperature for the boron re- - q 1 ease operation. This heat exchanger is also a re-l generative type, with the heat supplied by flow taken i
! from the letdown line upstream of the letdown heat exchanger. During the boron storage operation, the ,
letdown reheat heat exchanger. tube-side flow is valved off. i. 4 i Downstream from the letdown reheat heat exchanger, - the' letdown flow passes in parallel through the thermally regenerable demineralizers for boron storage j or release, as described above. The capacity of the ! demineralizers is based on the total boron storage
. required for the daily load follow operation.
] After passing through the demineralizers, the let-7 down flows through the shell side of the moderating l 7-5
, , - . . _ _ _ _ _ - . _ _ . - ~ . - . _ . _ _ _ . . . . . . _ , _ . . . - . . _ . _ _ . ~ . _ . _ _ _ . . . _ _ _
heat exchanger and is routed to the CVCS at a point upstream of the reactor coolant filter. During the boron release cperation, the flow is cooled by the letdown chiller heat exchanger bef\re being directed to the CVCS. The capability of the BTRS to change the boron con-centration by an equal amount in the RCS decreases with core life (and boron concentration in the RCS). Consequently, the design peint for the BTRS is the latest point in core life for which a certain load cycle is warranted. 73 DESIGN BASIS AND EQUIPMENT DESCRIPTION Chiller Pumps The capacity of the chiller pumps is determined by the chilled water flowrate, which is dictated by the chiller package. Two pumps per chiller package are supplied. The required pump head is determined by line pressure drops and elevation head. The chilled water loop is connected to atmosphere via the surge tank and is a 150 psig system. The pumps are made of carbon steel. 7-6
Heat Exchangers Moderatine Heat Exchanger The moderating heat exchanger reduces the load on
-~
the chiller package by operating as a regenerative heat exchanger between incoming and outgoing BTRS streams. The size of this heat exchanger is directly affected by the size of the letdown chiller heat exchanger and the chiller package. Also, its size is chosen such that it has the same UA as the let-down chiller heat exchanger. The fluid entering the BTRS enters the tube side of the moderating heat exchanger at the normal letdown stream temperature of 115 F. The shell side fluid, which comes directly from the demineralizers, enters at 50 F during storage and 140 F during release. The design flowrate for the tube and shell side equals the maximum letdown rate since the letdown stream passes through both sides of the heat exchanger. Letdown Chiller Heat Exchanger The letdown stream enters the tube side of the let-down chiller heat exc. anger after leaving the tube side of the moderating heat exchanger. For the design case, the outlet temperature on the tube side of the letdbwn chiller heat exchanger is 40 7. This 7-7 7 , , ,---.
t will enable the chiller loop to cool down the ion exchangers, associated piping and heat exchangers from 140 F to 50 F. The shell side flow rate is dictated by the ch'111er package. The inlet temper-ature on the shell side to the letdown chiller heat exchanger is 39 F, which is the lowest water temper-ature the chiller can deliver without special controls. In addition to the design case described above, the letdown chiller heat exchanger is also used during release of boron from the demineralizers. For this case it is utilized to cool the liquid going to the volume control tank to insure that the temperature does not exceed 115 F. The temperature of the water returning to the chiller will be higher during re-lease operations than it would be during storage operations. A chiller requirement prevents the tem-perature of the water entering the chiller from ex-ceeding 105 F. Letdown Reheat Heat Exchanger The letdown reheat heat exchanger is used only during i release operations. During release, it heats the letdown stream from the tube side outlet of the moderating heat exchanger to 140 F. Water used for i i 1 . 7-8 i
heating this letdown stream is diverted from the CVCS letdown line (upstream of the letdown heat exchanger). From there, it passes through the let-down reheat heat exchanger shell and is returned to the letdown stream upstream of the letdown heat ex-changer. The temperature upstream of the letdown heat exchanger depends upon the letdown rate, charging line temperature and RCS temperature. The design case for the letdown reheat heat exchanger occurs when the shell side inlet temperature is 115 F and the tube side heat temperature is between 280 - 380 F. The upper limit of 380 F equals the maximum outlet temperature on the shell side of the regenerative heat exchanger during maximum letdown. For the design case, the tube side flow rate is arbitrarily assumed to be 75 percent of the maximum letdown rate. It should also be noted that the normal inlet temper-ature on the shell side is around 130 F rather than the design value of 115 F. This conservatism provides a faster heat up transient when the BTRS is switched to the release mode than would otherwise be obtained. l 7-9
Chiller Surce Tank The chiller surge tank fulfills the following functions:
- 1. Handles thermal expansion and contraction of the water in the chiller circuit.
- 2. Provides holdup for a leak in the chiller heat exchanger. Assuming the tank is half full, it can accommodate a 1 gpm leak for 4 hours.
3 Provides a thermal buffer for the chiller package. Thermal Regeneration Demineralizers . The function of the thermal regeneration demineralizers is to store and release boron on the resin following changes in plant load. This must be done at a suf-ficient rate so t 'it the total amount of boron that must be removed from the RCS to compensate for xenon transients and other reactivity changes can be ac-complished during a load cycle. Furthermore, the
'demineralisers must be able to release, at a suf-ficient rate, the previously stored boron.
The xenon transient for the applicable load cycle must be determined in order to calculate the total amount of boron that must be handled by the demin-eralizers. The amount of boron concentration change
/
7-10
I l l i in the RCS depends on many. things; some of which are i the plant's specific load cycle, minimum power level, and return to power capability. The demineralizers utilized in the BTRS can accept flow in two directions. The flow direction through the demineralizers during storage is always opposite to that during release. This provides much faster response when the beds are switched from storage 1 to release and vice versa, than would be the case if the demineralizers could accept flow in only one direction. The flow through the demineralizers is maintained be-low 7 gpm/ft 2 for design purposes. Chiller Package The purpose of the chiller package is twofold.
- 1. To cool down and keep the demineralizers at 50 F during storage of boron.
- 2. To maintain an outlet temperature from the BTR3 l
at or below 115 F during release of boron. Case 1 above imposes the greatest load on the chiller package and is therefore used for design. 7-11 u_ .
Included in the chiller package is a centrifugal type compressor. It vill deliver a chilled water temperature of 39 F, without any special controls for freeze protection. It requires cooling water at a temperature that does not exceed 95 ? which indicates that service water is normally used for cooling. The chiller is controlled by a thermo-couple in the chiller package. During release of boron, the temperature of the chilled water is allcwed to increase but the return water to the chiller package must not exceed 105 F. The chiller package is located in a chilled water ! loop containing a surge tank, chiller pumps, the let- .
- down chiller heat exchanger, piping, valves and con-trols. The chiller heat exchanger has a bypass so that the flow through this heat exchanger can be varied from no flow to the full chilled water flow.
This bypass will also allow the chilled water loop to be cooled down to 39 F prior to cooldown of the BTRS which will insure a mere rapid' cooldown rate of this system. The desired cooling capacity is ad-just..ed by controlling the chilled water flow rate r passed through the shell side of the letdown chiller heat exchanger. J l 7-12
i 7.4 SYSTEM AND EQUIPMENT DESIGN PARAMETERS System Parameters Maximum letdown flow rate, spm 120 Temperature of resins during storage, F 50 Temperature of resins during release, OF 140 Normal temperature of letdown flow entering system, OF 115 Temperature of cooling water required, F 95 Chiller Pump Number, per unit 2 Type Centrifugal Design prescure, psig 150 Design temperature, F 200 Design flow rate, gpm 400 Design head, feet 150 Material of construction Carbon Steel Moderating Heat Exchanger , Number, per unit 1 Type U-tube, four tube passes, 2 shell passes 6 Design Heat Transfer BTU /hr. 2 53 x 10 7-13
.- _ . . . . .~ . . . .
3 hell Tube Design Pressure, psig 300 300 Design temperature, F 200 200 Design flow rate, lb/hr 59,640 59,640 Design inlet temperature, F 50 115 Design outlet tempera:ure, OF 92 73 Fluid circulated Reactor Reactor Coolant Coolant Material Stain- Stain-less less Steel Steel Letdown Chiller Heat Exchanger Number, per unit 1 Type U-tube, 4-tute pass and 2-shell pass 6 Design Heat Transfer, BTU /hr 1.65 x 10 Shell Tube Design pressure, psig 150 300 l Desig.h temperature, F 200 200 I Design flow rate, lb/hr 175,000 59,620 Degign inlet temperature, e as ,; Design outlet temperature, OF US 25 Fluid circulated' Chromated React:r
- ' dater C
- clant t
l Material Carbon .' t ain-Steel lecc Steel , 7-14 e' ~ "- - - * - - , . , . _, __ , _ . , , , , _
i Letdown Reheat Heat Exchanger Number, per unit 1 Type U-tube, 2-tube pass and 1 shell pass Design Heat Transfer, BTU /hr 1.49 x 106 Shell Tube Design pressure, psig 300 600 Design temperature, F 200 h00 Design flow rate, lb/hr . 59,640 44,730 Degign inlet temperature, "F 115 280 Degign outlet temperature, F 140 247 Fluid circulated Reactor Reactor Cociant Coolant Material , Stain- Stainless less Steel Steel Chiller Surge Tank Number, per unit 1 Type Horizontal Volume, gal. 500 Design pressure, psig Atmosphere Design temperature, F 2000 Material Carbon Steel Thermal Regeneration Demineralizers Number, per unit 5
. Type Flushable, Reversible flow 7-15
Design Pressure, psig 300 Design temperature, F 250 Resin volume, ft 3 79 Resin type IRN 78 Material of construction Stainless Steel Chiller Number, per unit 1 Capacity Ice ton 138 BTU /hr 1.66 x 106 Evaporator Flow, gpm 352 Inlet temperature, F 48.4 Outlet temperature, F 39 Condenser Flow, gpm 414 Inlet temperature, F 95 Outlet temperature, F By manufacturer of chiller package 4 7-16
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l PRESSURIZED WATER REACTOR RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM CHAPTER 8.0 7 BORON RECYCLE SYSTEM 't w w< ~v- m v'
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CHAPTER 8 r TABLE OF CONTENTS Section Title Page i 8.0 BORON RECYCLE SYSTEM . . . . . . . . . 8-1
8.1 INTRODUCTION
. . . . . . . . . . . . . 8-1 8.2 GENERAL DESCRIPTION . . . . . . . . . 8-1 ;
I Collection Requirements . . . . . . 8-1 Load Follow . . . . . . . . . . . . 8-3 Cold Shutdown and Startups . . . . . 8-3 Hot Shutdowns . . . . . . . . . . . 8-5 System Operation . . . . . . . . . . 8-6 i Capacity Requirements . . . . . . . 8-7 Design Core Cycle . . . . . . . . 8-7 Design' Surge . . . . . . . . . . . 8-7 8.3 DESIGN BASIS AND EQUIPMENT DESCRIPTION. . . . . . . . . . . . . . 8-8 Recycle Evaporator Feed Pumps . . . 8-8 Tanks . . . . . . . . . . . . . . . 8-9 Recycle Holdup Tanks . . . . .- . . - 8-9
- j. Recycle Evaporator Reagent Tank . 8-10 Demineralizers . . . . . . . . . . . 8-10
,~ - Recycle Evaporator Feed Demineralizers . . . . . . . . . . 8-10 Recycle Evaporator Condensate Demineralizers . . . . . . . . . . 8-11 Filters . . . . . . . . . . . . . . 8-11 Recycle Evaporator Feed Filters . 8-11 Recycle Evaporator Condensate Filter .. . . . . . . . . . . . . . 8-11 Recycle Evaporator Concentrate Filter . . . . . . . . . . . . . . 8-12 Recycle Evaporator . . . . . . . . 8-12 Recycle Holdup Tank Vent Eductor . 8-14 Electrical Heat Tracing and Build-l ing Temperature Control . . . . . 8-15 8.4 SYSTEM AND EQUIPMENT DESIGN PARAMETERS . . . . . . . . . . . . . . 8-17
) .
I s Y r 1 =
i TABLE OF CONTENTS (CONT'D) Section Title Page
> Pumps . . . . . . . . . . . . . . . . 8-17 Recycle Evaporator Feed Pumps . . 8-17 Tanks . . . . . . . . . . . . . . . 8-17 Recycle Holdup Tanks . . . . . . . 8-17 Recycle Evaporator Reagent Tank . 8-17 Demineralizers . . . . . . . . . . . 8-17 Recycle Evaporator Feed Demineralizers . . . . . . . . . . 8-17 Recycle Evaporator Condensate Demineralizer . . . . . . . . . . 8-19 Recycle Evaporators . . . . . . . 8-19 a
11
- ~
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LIST OF FIGURES Figures Title Page 8-1 Boron Recycle System . . . . . . . . . 8-21 8-2 Hot Shutdown Xenon Transients . . . . 8-23 8-3 Boric Acid Evaporator . . . . . . . . 8-25 e.' l iii
- .- . _ . . = ~ . .- . . _ _ . . . - . - = . . . -
8.0 BORON RECYCLE SYSTEM [
8.1 INTRODUCTION
I The basic functions of the Boron Recycle System ___ (BRS) are as follows:
- - a. Collect borated radioactive effluent from the Reactor Coo'lant System (RCS) and from other i
4 miscellaneous sources during normal plant i
! operations.
I 1 ] b. Process the volume of water collected during a core cycle, and in addition, accommodate short term surges.
- c. Process the recycled water to meet the chemical requirements for reactor makeup water and 4 wt.
percent boric acid with 1 percent fuel defects in the RCS. 4 8.2 GENERAL DESCRIPTION The Boron Recycle System (BRS) is shown on Figure 8-1. i Collection Requirements 4 The BRS accepts and processes all effluent which can 'l be readily re-used as makeup to the RCS. The recycle holdup tanks, basically, accept deaerated, tritiated,
- borate 4 radioactive letdown and' drains.
I l o 8-1
The BRS normally collects most of its water from the RCS via the chemical and volume control system let-down line and the reactor coolant drain tank in the Waste Processing System (WPS). Letdown is diverted to the BRS as a result of changes made to the RCS boron concentration by the Reactor Makeup System (RMS). This occurs during such operations as plant shutdown and startup, refueling, and dilution for core burn-uo. The reactor coolant drain tank is a collection point for reactor coolant from equipment inside the containment. The BRS also collects water from the following sources:
- 1. Volume control tank pressure relief (CVCS) - pro-vides a volume that can contain the radioactive water and gas which can be processed and recycled.
- 2. Boric acid blender (CVCS) - allows draining BAT for maintenance.
3 Safety injection system (SIS) flush operation - accepts flush water when boron injection tank valves are being tested or flushed,.
- 4. Waste evaporator condensate tank (WPS) - provides flexibility for recycling water from the WPS.
8-2 _ ~ _
5 Spent fuel pit pumps (SFPCS) - provides a means of storing the fuel transfer canal water in case maintenance is required on the transfer equipment.
- 6. Valve leakoffs and equipment drains - the recycle holdup tanks collect deaerated, tritiated water from valve leakoffs and equipment drains.
Load Follow Load follow is the capability of the RCS to respond to power demand fluctuations. To compensate for the Xenon transient as a result of increased or decreased power requirement on the reactor, the reactor coolant boron concentration is increased or decreased respec-tively. Normally, the Boron Thermal Regeneration System (BTRS) is used for load follow. However, the BRS does have some limited capability to load follow. This operation produces effluent which must be processed. Cold Shutdown and Startups During a cold shutdown and startup, letdown is di-verted to the BRS as a consequence of changes made to the reactor coolant system boron concentration and density. During a shutdown, the reactor is first brought to hot, no-power conditions (with rod insertion) and at that point borated to the shutdown concentration. 8-3
l 1 Boration is accomplished by using the Chemical and Volume Control System Reactor Makeup System (CVCSRMS) to feed concentrated boric acid through the boric acid blender to'the charging pumps and into the RCS while bleeding' reactor coolant from the RCS via the letdown line to the BRS. The reactor is then cooled down, during which time the coolant contracts. In - order to maintain a constant volume in the RCS, i.e., constant pressurizer water level, the contraction is compensated for by feeding a blend of concentrated boric acid and reactor makeup water to the RCS, while no letdown effluent is diverted to the BRS. When the reactor is cooled down and maintenance is required on the RCS, this system can be drained to
, the BRS. This water can be stored in the BRS and then returned to the RCS with only demineralizer and filter processing required.
During plant startup the reactor coolant is first heated to a hot, no-power condition with heat from the reactor coolant pumps and the pressurizer heaters. Effluent produced as a result of reactor coolant expansion occurring during heatup and pressurizer steam bubble formation is diverted via the letdown line to the BRS. The reactor coolant is then diluted J 8-4 i
from the shutdown concentration to the hot, just critical concentration. After the reactor is brought to power, additional dilution is required as xenon builds up to equi'.ibrium. The BTRS may be used to reduce the amount of effluent sent to the recycle holdup tanks. Hot Shutdowns During a hot shutdown, boron is used to follow the reactivity changes caused by the xenon transient (Figure 8-2). Two types of shutdowns are considered:
- 1. Short shutdown for one day or less, where xenon peak is compensated for by dilution and the xenon burnout (after reactor startup) is compen-sated for by boration. Both dilution and boration are accomplished by feed and bleed operations resulting in effluents for the BRS.
- 2. Long shutdown for greater than one day, where the xenon peak is not compensated for, but the decay below equilibrium is compensated for by boration. After startup, xenon builds up to equilibrium which is compensated for by dilution.
Both dilution and boration are accomplished by feed and bleed operations also resulting in effluents for the BRS. 8-5
System Operation The BRS effectively processes the reactor coolant by means of demineralization and gas stripping prior to evaporation which separates and reclaims the boric acid and the reactor makeup water. The water col-lected by the system is initially processed through the recycle evaporator feed demineralizers, and the
; recycle evaporator feed filters, and then stored in the recycle holdup tanks. The water is then pumped to the evaporator package by the recycle evaporator feed pumps. In the evaporator package, fission and other gases are stripped from the water and the water is then separated into boric acid (concentrate) and makeup water (condensate). The makeup water is passed through the condensate demineralizer to limit the boron and anion concentration, then filtered by the recycle evaporator condensate filter, and flows to the reactor makeup water storage tank. If the activity level of the condensate is too high, the water is automatically diverted back to the recycle evaporator feed demineralizers and recycle holdup .
tank. The concentration is 4 wt. percent boric acid. The boric acid flows through the recycle evaporator concentrates filter prior to storage in the boric acid tanks (CVCS), Before transferring the boric 1 l l l l 8-6 1
acid from the evaporator to the boric acid tanks, it is analyzed, and, if it does not meet the required chemical standards, it can be diverted back to the recycle evaporator feed demineralizers and recycle holdup tanks. If the boric acid deviates severely from the required standards, it can be diverted to the WPS and drummed for disposal. Capacity Reauirements l Design Core Cycle During the course of a core cycle,the volume of water collected from the reactor coolant drain tank is primarily a function of the No. 2 and 3 seal leak-off flows from the reactor coolant pumps. The volume of letdown water collected is a function of plant operation and time in core cycle life. The assumed design core cycle operations which constitute the total volume of water processed by the BRS are: reactor coolant pumps' seal leakoff, core burnup dilution operations, two short hot shutdowns and startups, two long hot shutdowns and startups, three cold shutdowns and startups, and refueling and startup. Design Surge The design surge per unit for the BRS is a cold shutdown at approximately 80 percent core life, 200 8-7
ppm boron in the RCS with no rods and equilibrium xenon, followed by a startup. 8.3 DESIGN BASIS AND EQUIPMENT DESCRIPTION Recycle Evaporator Feed Pumps The sise of the recycle evaporator feed pumps is determined by the rated capacity of the recycle evaporator and minimum flow requirements. A runout condition is specified for recirculation and for returning maintenance drains to the RCS or fuel transfer canal. Based on the design flow, RCS drain can be returned in less than 12 hours. Two pumps are supplied for redundancy. The required head is deter-mined by the line pressure drops and elevation head when the evaporator is being fed via the evaporator feed demineralisers and filters. The recycle holdup tanks serve as head tanks. The' pump is a canned motor pump with all wetted surfaces constructed of austenitic stainless steel. An auxiliary discharge connection is provided to return water to the re-fueling transfer canal from the recycle holdup tanks, if those tanks were used for storage of refueling transfer canal water during refueling equipment maintenance. Another auxiliary discharge connection 8-8
is provided to supply water to the suction of the charging pumps (CVCS) for refilling the RCS after loop or system drain. Tanks Reevele Holduo Tanks Two recycle holdup tanks provide storage of radio-active fluid which is discharged from the RCS during startup, shutdown, load changes and boron dilution. The sizing criteria for the tanks is based on the design surge cold shutdown and startup. The tanks are constructed of austenitic stainless steel. Each tank has a diaphragm which prevents air f rom dissolving in the water and prevents the hydrogen and fission gases in the water from mixing with the air. The volume in the tank above the diaphragm is continuously ventilated with building supply air, and the volume of gas below the diaphragm is vented to the WPS (Gas). In addition to the collection of effluent, the re-cycle holdup tanks provide the following functions:
- 1. Serve as a head tank for the recycle evaporator feed pumps.
8-9
- 2. Provide holdup for RCS draining to the center-line of the reactor vessel nozzles, including the pressurizer and steam generators.
3 Provide storage for refueling transfer canal water during refueling equipment maintenance. 4 Collect discharge from the various relief valves. Recycle Evaoorator Reacent Tank The recycle evaporator reagent tank provides a means of adding chemicals to the evaporator. The tank is constructed of austenitic stainless steel.
. Demineralizers Recycle Evaporator Feed Demineralizers Two flushable mixed bed resin demineralizers are used to remove anions and cations frcm the water entering the recycle holdup tanks. The demineralizers are aligned in parallel. The total capacity of the resin is sufficient for one equilibrium core cycle, assuming load-follow operation and one percent fuel defects. Each bed will accept maximum unit letdown flow. The vessels are constructed of austenitic stainless steel.
B-10
Recycle Evaporator Condensate Demineralicer A flushable anion resin limits the boron and anion concentration of the water sent to the reactor makeup water storage tank. The resin bed will accept the maximum evaporator condensate flow. The vessel is constructed of austenitic stainless steel. Filters Recycle Evaporator Feed Filters Two cage-assembly-type recycle evaporator feed filters (one per unit) are aligned in parallel to collect resin fines and particulate matter 5 microns or larger from water entering the recycle holdup tanks. Each filter is designed to accept maximum letdown flow. A disposable, synthetic cartridge is used. Recycle Evaporator Condensate Filter The cage-assembly-type recycle evaporator condensate filter is designed to filter the recycle evaporator condensate, prior to storage in the reactor makeup water storage tank. This filter is designed for maximum condensate flow. A disposable, synthetic cartridge is used. 8-11
Recycle Evacorator Concentrate Filter The recycle evaporator concentrate filter is a cage-assembly type. It filters 4 percent boric acid at the design rate of the evaporator concentrates pump, as it is transferred to the boric acid tanks (CVCS). A disposable, synthetic cartridge is used. Recycle Evapcrator The recycle evaporator (Figure 8-3), consists of an evaporator, absorption tower, evaporator condenser, distillate cooler, feed preheater, stripping column,
. vent condenser, eductor, concentrate pump, distillate pump, valves, piping, and instrumentation. The recycle evaporator removes hydrogen, fission gases, and any other gases from the evaporator feed. The condensate produced is of reactor makeup grade, and f
the concentrate is 4 wt. percent boric acid. With the exception of heat tracing, which is not installed on the recycle evaporator, it is identical to, and interchangeable with, the. waste evaporator (WPS). The borated water from the recycle holdup tanks flows into the package to the feed preheater. The preheater heats the feed stream, using process steam. The steam flow is controlled by temperature instru-mentation at the feed exit of the preheater. The 8-12
i s preheater feed flows into the stripping column where hydrogen, fission gases, and any other gases are stripped from the borated water. The feed flow is controlled by the evaporator liquid level. The stripping medium is a portion of the evaporator over-heads (steam) which enters the bottom of the packed stripping column. The stripper gases and an equi-1 librium quantity of stripping steam leave the top of the stripping column and flow to the vent con-denser which cools the gases and condenses the major portion of the stripping steam. Component cooling water provides the source of cooling water. The i j gases are' vented via the vent header to the WPS and i the liquid condensate is returned to the stripping column by means of an eductor in the feed line. The evaporator concentrates the borated liquid to 4 wt.
- percent boric acid. The evaporator bottoms are con-I tinuously recirculated by the boric acid concentrate
- pump, and when a batch is completed it is transferred to the boric acid tanks. The unit is designed such that dirtillate and concentrate samples can be taken when the unit is operating, thus the unit does not need to be cooled dcwn for sampling prior to trans-ferring a batch of concentrates.
( 8-13
The major portion of vapors leaving the evaporator flow through the absorption tower and then are con-densed in the evaporator ccndenser. The distillate is pumped through a distillate cooler and out of the unit at a maximum of 120 F. A portion of the distillate is recycled to the absorption tower to serve as the absorption medium. The evaporator condenser and distillate cooler are serviced by component cooling water. In the evapora-tor condenser, cooling water is controlled by the distillate flow. The evaporator is tube side steam heated and steam flow is controlled by evaporator pressure. All equipment in the unit in contact with the pro-cess fluids is constructed of austenitic stainless i steel. Recycle Holdup Tank Vent Eductor The recycle holdup tank vent eductor is used period-ically (about once every other month) to remove gases which accumulate under'the recycle holdup tank i diaphragm. The motive force is provided by the stanc-by waste gas compressor. 8-lh .
,,. - - - -.---,y- ---- . r- - .,.r-
Electrical Heat Tracing and Building Temperature Control Electrical heat tracing is not required on any BRS components, provided those containing 4 wt. percent boric acid are located in a building having redundant temperature control. This insures that the 4 wt. percent boric acid solution will not precipitate. 8-15
8.4 SYSTEM AND EQUIPMENT DESIGN PARAMETERS The Boron Recycle System (bas) is designed to accept 150 gpm per unit peak effluent and batch process at a 15 gpm rate. Pumps Recycle Evaporator Feed Pumps Number 2 Type Canned Design flow, gpm 30 Design head, ft 320 Design pressure, psig 150 Design temperature, F 200 Operating temperature, F 115 Material of construction Austenitic SS i Tanks Recycle Holdup Tanks Number 2 Type Vertical with Diaphragm (1) Usable tank capacity, gallons each (3) Tank design pressure Atmospheric (2) Tank design temperature, F 200 l Tank operating temperature 115
- Material of construction Austenitic SS i
1 8-17 l l ! l l
1 l i (1) The diaphragm is designed for 180 F and atmospheric pressure.
, (2) Not including hydrostatic head.
(3) 4 loop = 56,000 single unit, 112,00J twin unit, 3 loop = 42,000 single unit, 84,000 twin unit, 2 loop = 31,000 single unit, 02,000 twin unit. Recycle Evaoorator Reagent Tank Number 1 Type Vertical Tank capacity, gal . 5 Tank design pressure, psig 150 Tank design temperature, F 200 Tank operating temperature Ambient Material of construction Austenitic SS Demineralizers Recycle Evacorator Feed Demineralizera Number 2 Type Flushable Design temperature, F 250 Design pressure, internal, psig 300 Design pressure, external, psig 15 Resin volume, ft 3 30 8-18 L _
Vessel volume, ft 3 43 Bed depth, ft 55 Resin type Rohm and Hass Amberlite I.R.N. -150 or Equivalent Design flow rate, gpm 120 Pressure drop at design flow, psi 12.8 Material of construction Austenitic SS Recycle Evaporator Condensate Demineralizer Number 1 Type Flushable Design temperature, F 250 Design pressure, internal, psig 300 Design pressure, external, psig 15 Resin volume, ft 3 3g Vessel volume, ft 3 43 Bed depth, ft 55 Resin Type Rohn and Hass Amberlite I.R.N. -78 or Equivalent Design flow rate, gpm 35 Pressure drop at design flow, psi 3 Material of construction Austenitic SS Recycle Evaporator:3 . Number 1 Capacity 15 spm 8-19
Steam design pressure 50 psig Concentrated hold-up volume 500 gallons Component cooling water flow 780 gpm 6 DF liquid 10 DF gas 10 5 l l 8-20.
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PRESSURIZED WATER REACTOR , RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM f
- CHAPTER 9.0 i RADIATION SOURCES 4
t y i i i b i i i t -_.,._,_._, . _ - - , ,__. , _ .__- - , ,_._ _-- _ _ . _ _ _ _~ _ .-- - - - ..__ _ - . _ . . _ . _ _ _
CHAPTER 9 TABLE OF CONTENTS Section Title Page 9.0 RADIATION SOURCES . . . . . . . . . . 9-1
9.1 INTRODUCTION
. . . . . . . . . . . . 9-1 9.2 ACTIVITIES IN THE CORE . . . . . . . 9-1 9 *. 3 ACTIVITIES IN THE FUEL ROD GAP . . . 9-1 9.3 FUEL HA! IDLING SOURCES . . . . . . . . 9-3 95 REACTOR COOLANT FISSION PRODUCT ACTIVITIES . . . . . . . . . . . . . 94 9.6 REACTOR COOLANT TRITIUM SOURCES . . . 9-6 General Discussion . . . . . . . . 9-6 9.7 VOLUME CONTROL TANK ACTIVITIES . . . 9-11 9.8 GAS DECAY TANK ACTIVITIES . . . . . . 9-11 9.9 ACTIVITY IN RECIRCULATED SUMP WATER . 9-11 l
l i
LIST OF TABLES Tables Title Pa5e2 9-1 Core and Gap Activities . . . . . . . 9-15 9-2 Core Temperature Distribution . . 3. 9-15 9-3 Nuclear Characteristics of Highest Rated Discharged Assembly . . . . . . 9-17 9-4 Activities in Highest Rated Discharged Assembly Curies at Time of Reactor - Shutdown . . . . . . . . . . . . . . ,9-21 9-5 Parameters used in the Calculation of Reactor Coolant Fission Product Activities . . . . . . . . . . . . . 9-23 9-6 Reactor Coolant Maximum Activities , Downstream of Regenerative Heat Exchanger . . . . . . . . . . . . . . 9-27 I 9-7 Tritium Production in the Reactor Coolant . . . . . . . . . . . . . . . 9-31 9-8 Volume control Tank Activities . . . 9-33 9-9 Gas Decay Tank Activities . . . . . . 9-35 9-10 concentration of Iodine Isotopes in the Recirculation Loop . . . . . . . 9-37 9-11 Radiation Sources Circulating in Residual Heat Removal Loop and Associ-
.ated Equipment - MEV/CC - SEC . . . . 9-37 11
_-2 -_._._._.-__ _ _ _ _ _ _ _ . _ _____ _ . - _ _ - _ _ _ . _ _ _ _ . _ _ _ ______ .-.-__-_ -_. --_ _ __________ ____- -_____ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - - _ _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
9.0 RADIATION SOURCES
9.1 INTRODUCTION
This section presents the quantities of radicactive isotopes present in the core, fuel rod gap, coolant, volume control tank, gas decay tank and recirculatpf sump water. A brief discussion of tne derivations is also p.rovided. 9.2 ACTIVITIES IN THE CORE , The total core activity calculation is consistent with TID 14844(1) and data from ORNL-2127.(2) Numerical values for isotopes which are important as health hacards are given in Table 9-1. 93 ACTIVITIES IN THE FUEL ROD GAP The computed gap activities are based on buildup in the fuel from the fission process and diffusion to the fuel rod gap at rates dependent on the operating temperature. For the purposes of this analysis, the fuel pellets were divided into five concentric rings each with release rate dependent on the main fuel temperature within that ring. The diffusing isotope is assumed present in the gas gap when it has dif-fused to the boundary of its rings. The diffusion coefficient, D', for Xe and Er in UO 2' varies with te?.perature through the following expression: 9-1
D' (T) = D' (1673) exp -k (k 10h3) where E = ictivation energy D'(1673)=diffus{oncoefficientat sec 1673 K = 1 x 10-11 og T = temperature in R = gas constant The above expression is valid for temperatures above 1100 C. Below 1100 U C fission gas release cccurs mainly by two temperature independent phencmena, re-coil and knock-out, and is predicted by using D' at 1100 C. The value used for D' (1673 K), based on data at burnups greater than 10 19 fissions /ce, accounts for possible fission gas release by other mechanisms and pellet cracking during irradiation. The diffusion coefficient for iodine isotopes is assumed to be the same as for Xe and Kr. Toner and Scott (3) observed that iodine diffuses in Uo, a: about the same rate' as Xe and Kr and has about the same activation energy. Data surveyed and reported by Belle ( ) indicater that iodine diffuses at slightly slower rates than do Xe and Kr. For a full core cycle at 3391 MWt, the above analysis results in a pellet-cald gap activity of less than 9-2
_ - . .- ._ . - - - . . - . ~ - . . -- ... - -. - . . - . ._.
/
i 3% of the dose equivalent equilibrium core iodine l inventory. The noble gas activity present in the pellet-clad gap and assumed release to the contain-ment is about 2 5% of the core inventory. ~~ j The percentage of the total core activity present in the gap for each isotope is also listed in Table
, 9-1.
i j The core temperature distribution used in this analysis, i j based on hot, channel factors, F3d, = 1 70 and F q = 2.82, is presented in Table 9-2. 9.4 FUEL HANDLING SOURCES . The inventory of fission products in a fuel assembly i is dependent on the rating of the assembly. The parameters used for the calculations of the highest rated assembly in the case to be discharged are summarized in Table 9-3, while the associated l activities at the time of shutdown are given in Table 9-4 ] The expected end-of-life temperature and power dis-1- .tributions were calculated by using the radial and axial power peaking factors of 1.47 and 1 37 l respectively. The conservative end-of-life temper- , ature and power distributions were calculated by f 9-3 i i
using the same radial power peaking factor as in the expected case, but with a higher axial power peaking factor of 1.72. Thus, the temperature / volume distribution in the fuel is changed and the maximum temperature is increased (Table 9-3), re-sulting in an increased fraction of fission products in the fuel-cladding gap (Table 9-4). 9.5 REACTOR COOLANT FISSION PRODUCT ACTIVITIES The parameters used in the calculation of the re-actor coolant fission product inventories together with the pertinent information concerning the expected coolant cleanup flow rate and decineralizer effectiveness, are summarized in Table 9-5; while the results of the calculations are presented in j Table 9-6. In these calculations, the defective fuel rods were assumed to be present at the initial core loading and were uniformly distributed through-out the core. Thus, the fission product escape rate coefficients were based upon the average fuel temperature. The calculations were performed for the prevailing temperature downstream of the regen-erative heat exchanger. The coolant density correction of 0.733 should be made in order to obtain the correct activities of the reactor operating temperature. 9-4
, a 7 ._.-- -
The fission product activities in the reactor coolant during operation with small cladding defects (fuel rods containing pinholes or fine cracks) in the 1% of the fuel rods were computed using the fol-lowing differential equations: For parent nuclides in the coolant, dN wi
- B' dt "1"C i -(i* Ui+B o - tB') "wi for daughter nuclides in the coolant, dN wj dt "j NC j - (^j
- U j +B o B'
- tB,) Nw) + AN g y1 where:
N = population of nuclide D = fraction of fuel rods having defective cladding R = purification flow, coolant system volumes per sec. Bg = initial boron concentration, ppm B' = boron concentration reduction rate by feed and bleed, ppm per see n = removal efficiency of purification cycle for nuclide A = radioactive decay constant v = escape rate coefficient for diffusion into coolant l l l l 9-5 ! l L !
i l Subscript C refers to core < I Subscript w refers to coolant . Subscript i refers to parent nuclide Subscript j refers to daughter nuclide 9.6 REACTOR COOLANT TRITIUM SOURCES General Discussion During the fissioning of uranium, tritium atoms are generated in the fuel at a rate of approximately
-2 8 x 10-5 atoms per fission (1.05 x 10 curies /mwt -
i day). Other sources of tritium-include neutron reactions with boron (in the coolant for shim , control), neutron reactions with lithium (utilized in the coolant fcr pH control, and produced in the coolant neutron reactions with boron), and by i neutron reactions with naturally occurring deuterium j in light water (see Table 9-7). A. Release of Ternary-Produced Tritium The tritium formed by ternary fission in uranium fueled reactors, can be reta ned in the fuel, accumulate in the void between the fuel and clad- . ding, react with cladding material (zirconium tittide), or' diffuse _through the cladding into
- the coolant. Operating experienced at the Ship-pingport reactor (zirconium clad) indicated that i
9-6
. _ _ . . , . , , _ _ _ _ . _ . . __s - _ _ _ _ _ . . _ . . . , . _ , _ . . - _ _ . , - _ . . . , _ , , , , . . , . _ _ , . . . , . . ~ _ . - . , , _ _ .
less than 1% of the ternary produced tritium is released to the reactor coolant. In order to insure adequate sizing of liquid waste treatment facilities, WNES conservatively assumes that 30% of the ternary produced tritium is released to coolant. This assumption then requires that the waste treatment system be sized to process approximately 4 reactor coolant system volumes in addition to normal reactor plant liquid wastes. Anticipated ternary tritium losses to the reactor coolant is 1%. B. Tritium Produced from Boron Reactions The neutron reactions with boron resulting in the production of tritium are: B10 (n, 2a) T 10 (n, a), L17 (n, na) T B 1 B (n, T) Be 9 B10 (n, d) Be9 (n,a)T Of the above reactions, only the first two con-tribute significantly to the tritium productior.. The B11 (n, T) Be9 reaction has a threshold of 14 Mev and a cross section of % 5 mb, since the number of neutrons produced at this energy are i 9-7
, ,1
less than 10 9 n/cm 2 -see the tritium produced 0 from this reaction is negligible. The B reaction may be neglected, since Be9 has been found to be unstable. C. Tritrium Produced from Lithium Reactions The neutron reactions with lithium resulting in the production of tritium are: Li (n, na) T Li (n, a) T In the WNES designed reactors, lithium is used to maintain the reactor coolant pH at s 9.5 The reactor coolant is maintained at a maximum level of 2.2 ppm lithium. A cation demineralizer is included in the Chemical and Volume Control System to remove the excess lithium produced in the B10 (n, a) Li I reactions. The Li (n, a) T reaction is controlled by limit-ing the Li 6 impurity in the lithium used in the reactor coolant and in lithiating the demineralin-ers to less than 0.001 parts of Li 6 . This limitation has been in effect on WAPD designed reactors since 1962. l 9-8
1 i D. Tritium Production from Deuterium Reactions Since the amount of naturally occurring deuterium , is less than 0.00015, the tritium produced from this reaction is negligible at less thar 1 curie per year. E. Tritium Sources from the Reactor Emoloying Ag-In-Cd Absorber Rods Basic Assumptions and Plant Parameters:
- 1. Core thermal power 3391 MWt
- 2. Plant load factor 0.E 3 Core volume 1153 ft 3 4 Core volume fractions
- a. UO 2 0 3023
- b. Zr + SS 0.1035
- c. H iO 2
O.5942 5 Initial reactor coolant boron level
- a. Initial cycle 840 ppm
- b. Equilibrium cycle 1200 ppm
- 6. Reactor coolant volume 12,760 ft 3-7 Reactor coolant transport
__ times d e 9-9 , _ m - , ,
- a. In-core 0.77 sec
- b. Out-of-core 10.87 sec
- 8. Reactor coolant peak lithium level (99% pure L17) 2.2 ppm 2
9 Core averaged neutron fluxes: n/cm -sec
- a. E > 6 MeV 12 2.91 x 10 12
- b. E > 5 Mev 7.90 x 10
- c. 3 MeV 1 E 1 6 MeV 2.26 x 1013
- d. 1 MeV 1 E 1 5 Mev 5.31 x 10 13
- e. E < 0.625 ev 2.26 x 1013
- 10. Neutron reaction cross-sections
- a. B 10 (n, 2a) T:
a(1 MeV 1 E 1 Mev) = 31.6 mb (spectrum weighted) c(E > 5 Mev) = 75 mb
- b. L17 (n, na) T:
a(3 MeV 1E1 6 Mev) = 39.1 mb (spectrum
, weighted) c(E > 6 Mev) = 400 mb
- 11. F,-action of ternary tritium diffusing through zirconium cladding
- a. Design value 0.30
- b. Expected value 0.01 9-10 '
9
i 1 l 9.7 VOLUME CONTROL TANK ACTIVITIES The 400 ft 3 volume control tank is assumed to contain 160 ft 3 of liarid and 240 ft 3 of vapor. Table 9-8 lists the acti' ties in the volume control tank with clad defects in _% of the fuel rods. f 9.8 GAS DECAY TANK ACTIVITIES The activity in one gas decay tank was calculated assuming that one tank, initially void of activity, is filled with all the gaseous activity that could be stripped off from the entire reactor coolant system operating with 1% fuel defects. The specific activities presented in Table 9-5 were multiplied by the reactor coolant system volume of 12,760 ft'. Table 9-9 lists the total activity in one gas decay tank which is also the total activity in the reactor system. 9.9 ACTIVITY IN RECIRCULATED SUMP WATER Table 9-10 lists the concentration of iodine and noble gas isotopes in the recirculation loop at time cerc after the design basis loss-of-coolant accident based on the following assumptions: Power level 3391 MWT Reactor coolant volume 12,760 ft 3 f ? - l 9-11
Emergency cooling water volume 46,900 ft 3 lodine gap activity picked up by containment sump water 100% Percent of core iodine activity present in fuel
- rod gaps 3%
The radiation sources circulating in the res$ dual heat removal loop are shown in Table 9-11 and are used for whole body radiation doses in the auxiliary building. The radioactivity in the containment also would be additional source of radiation to the auxiliary building following a loss-of-coolant accident. References (1) DiNunno, J. J., et. a., " Calculation of Distance Factors for Power and Test Reactor Sites", TID 14844, March 1962. (2) Bloneke, J. O. and Todd, Mary F., " Uranium-235 Fission Product Production as a Function of Thermal Neutron Flux, Irradiation Time, and Decay Time", (4 volumes) August and December 1957).
- Under TID 14844 assumptions, 50% of the core iodine is assumed.to be released from the fuel rods and picked up by the recirculated water.
9-12
(3) D. F. Toner and J. L. Scott, " Fission Product Release for UO '" Nuclear Safety Volume III 2 No. 2, December 1961. (4) J. Belle, Uranium Dioxide: Properties and Nuclear Applications Naval Reactors Division of Reactor Development, USAEC, 1961. 4 A l 9-13
i TABLE 9-1 CORE AND GAP ACTIVIT'IES i l Assumptions: Operation at 3391 MWt for 50) days l
! Temperature Distribution Spe:ified in Table A.2-2 l Curies Percent j in the of Core Curies Core Activity in the 6"E Isotope (X 107) in the Gap (X 105)
I i I-131 8.35 2.3 19.2 i I-132 12.75 0.26 33 I-133 19.09 0.79 15 1
- I-134 23 01 0.16 3.8 3
.I-135 17.05 0.43 75 a' ^
Xe-133 19.02 1.85 35 2 Xe-133m 5.16 1.27 6.55 Xe-135 7.23 0.54 3.90 Xe-135m 5.25 0.086 0.45 Kr-85 0.1228 21 57 2.64 4
.Kr-85m 3.76 0.29 1.09 4
Kr-87 7.22 0.20 1.44 Kr-88 _ 10.27 0.29 2.98 i 1. l
. TABLE 9-2 l CORE TEMPERATURE DISTEIBUTION i
, % of Core Fuel Volume Local Temperature,UF i Above the Given Temperature i
- 0.0 4100
- i. 0.2 3700 4
1.8 3300 7.0 2900 14.5 *2500 1 c 15
, _ _ - . ....:.--.,__a-,.__..__-_.._.__..-..-.-.--....,...
4 !~ 4 TABLE 9-3 NUCLEAR CHARACTERISTICS OF HIGHEST RATED DISCHARGED ASSEMBLY e Conservative Reactor Power Expected Case Case Rating, MWt- 3391 3391 102% Rating 3459 3459 i
! Number'of Assemblies Array 193 193 15 x 15 15 x 15 Core Average Assembly Power i- at 102% Rating, MWt 18.93 18.93 ; Discharged Assembly (highest power)
Axial P'eak to Average Ratio 1.37 , 1.72 e Peak Power, KW/ft 13 9 17.4 Maximum Temperature, F 3550 4000 Radial Peak to Average, Ratio 1.30 1.30 l-Temperature / Power . Distribution % Fuel Power, MWt- % Fuel Power, MWt [ Local Temperature, F Volume in Volume Volume in Volume
> 3900 0 0 0.67 0.17 3700 .3900 0 0 2.00 0.49 3500 - 3700 0.33 0.08 3.33 0.83 3300.- 3500 1.67 0.41 4.67 1.16 l
9-17. A
TABLE 9-3 (Cont'd) NUCLEAR CHARACTERISTICS OF HIGHEST RATED DISCHARGED ASSEMBLY Temperature / Power Distribution % Fuel Power, MWt % Fuel Power, MWt Local Temperature, F Volume in Volume Volume in Volume 3100 - 3300 3.00 0.75 6.00 1.49 2900 - 3100 4.33 1.07 7 33 1.82 2700 - 2900 5.67 1.41 8.67 2.15 2500 - 2700 7.00 1.73 10.00 2.48
< 2500 78.00 19.35 57 33 14.22 100.00 24.60 100.00 24.61 1
1 1 1 9-19 .
l l I 1 TABLE 9-4 ACTIVITIES IN HIGHEST RATED DISCHARGED ASSEMBLY CURIES AT TIME OF REACTOR SHUTDOWN Expected Case Conservative Case Fraction Fraction in Fuel- Fuel- in Fuel- Fuel-Cladding Cladding Cladding Cladding Isotope Total Curies Gap Gap Curies Gap Gap Curies I-131 5 90 x 10 5 0.0186 1.10 x 10 0.0476 2.81 x 10' I-132 8'.95 x 105 0.00207 1.85 x 10 3 0.00552 4.94 x 10 3 I-133 1 36 x 10 6 0.00677 9.21 x 10 3 0.016S 2.28 x lo" 0 I-134 1.55 x 10 0.00129 2.00 x 10 3 0.00341 5 29 x 10 3 6 I-135 1.20 x 10 0.00383 4.60 x 103 0.00946 1.14 x 10 3 Kr-85m 2.61 x 10 5 0.00533 1.39 x 10 3 0.00601 1.57 x 10 Kr-85 8.50 x 10 3 0.273 2.32 x 10 3 0.447 3.80 x 10 3 2 Kr-87 5 00 x 10 5 0.00154 7.70 x 10 0.00414 2.07 x 10 3 Kr-88 7.12 x 10 5 0.00533 3.79 x 10 3 0.00601 4.28 x 103 4 2 2 Xe-133m 3 38 x 10 0.0102 3.45 x 10 0.0263 8.8*9 x 10 6 Xe-133 1 33 x 10 0.0153 2.03 x 10 0.0383 5 09 x 10 Xe-135 3.63 x 10 5 0.00416 1 51 x 10 3 0.0111 4.03 x 10 3 9-21 l
l l TABLE 9-5 PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT FISSION PRODUCT ACTIVITIES
- 1. Core thermal power, max. calculated, MWt 3391
- 2. Fraction of fuel containing clad defects 0.01 3 Reactor coolant liquid volume, ft 3 12,760 4
Regetor coolant average temperature, F 570 5 Purification flow rate (normal), gpm 75
- 6. Effective cation demineralizer flow, gpm 7
- 7. Volume control tank volumes
- a. Vapor, ft 3 270
- b. Liquid, ft 3 130 S. Fission product escape rate coefficients:
_ a. Noble gas isotopes, sec -1 6.5 x 10 -8
- b. Br, I and Cs isotopes, sec -1 1.3 x 10-8
- c. Te isotopes, sec -1 1.0 x 10 -9
- d. Mo isotopes, sec -1 2.0 x 10 -9
- e. Sr and Ba isotopes, sec -1 1.0 x 10
-11
- f. Y, La, Ce, Pr isotopes, sec -1 1.6 x 10
-12 9 Mixed bed demineralizer decontamin-ation factors:
- a. Noble gases and Cs-134, 136, 137, Y-90, 91 and Mo-99 1.0
- b. All other isotopes 10.0 9-23
TABLE 9-5 (Cont'd) PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT FISSION PRODUCT ACTIVITIES
- 10. Cation bed demineralizer decontamin-ation factor for Cs-134, 136, 137, Y-90, 91 and Mo-99 10.0
- 11. Volume control tank noble gas stripping fraction (closed system):
Isotope Stripping Fraction
-5 Kr-85 2.3 x 10
-1 Kr-85m 2.7 x 10
-1 Kr-87 6.0 x 10 Kr-88 4.3 x 10 -1 Xe-133 1.6 x 10 -2
-2 Xe-133m 3 7 x 10 Xe-135 1.8 x 10 -1
-1 Xe-135m 8.0 x 10 Xe-138 1.0 i
9-25
6 TABLE 9-6 REACTOR COOLANT MAXIMUM ACTIVITIES DOWNSTREAM OF REGENERATIVE HEAT EXCHANGER Activatior? Products uc/cc (150 F) Mn-54 5.6 x 10 Mn-56 2.1 x 10-2
-2 Co-58 1.8 x 10 Fe-59 7.5 x 10-0 0
Co-60 5.4 x 10 Non-Volatile Fission Products (Continuous Full Power Ooeration) uc/cc (150 F) uc/cc(150 F) Br-84 -2 4.24 x 10 I-133 3.59 Rb-88 -2 3.36 Te-134 2 59 x 10 Rb-89 8.53 x 10 -2 I-134 4.75 x 10 -1 Sr-89 3 73 x 10 -3 Cs-134 2.21 x 10 -1
-4 Sr-90 1.15 x 10 I-135 1.84 Y-90 2.27 x 10- cs-136 1.22 x 10 -1 Sr-91 1.77 x 10 -3 cs-137 1.11 Y-91 6.29 x 10 -3 Cs-138 7.33 x 10-1 Mo-99 4.64 Ba-140 h.05 x 10-3 1 58 x 10 -3 I-131 2.34 La-140 Te-132 2 54 x 10 -1 Ce-144 2 97 x 10
-4 I-132 8.28 x 10 -1 Pr-144 2 97 x 10
-4 I
9-27
i i TABLE 9-6 (Cont'd) REACTOR COOLANT MAXIMUM ACTIVITIES DOWNSTREAM OF REGENERATIVE HEAT EXCHANGER Gaseous Fission Products uc/cc (150 F) Kr-85 4.4h Kr-85m 2.145 Kr-87 1.19 Kr-88 3.36 2 Xe-133 2.40 x 10 Xe-135 6.87
-1 Xe-138 5.52 x 10 9-29
TABLE 9-7 TRITIUM PRODUCTION IN THE REACTOR COOLANT Released to the Coolant Total Design Expected Tritium Source Produced Value Value Ternary Fissions 10,420 3126 104 Burnacle Poison Rods (Initial Cycle) 922 277 9 Soluble Poison Boron (Initial Cycle) 378- 378 378 (Equilibrium Cycle) 525 525 525 Li-7 Reaction 11 11 11 Li-6 Reaction 6 6 6 Deuterium Reaction 1 1 1 Totals Initial Cycle 11,738 3799 '509 Totals Equilibrium Cycle 10,963 3669 647 t 9-31
TABLE 9-8 VOLUME CONTROL TANK ACTIVITIES Assumptions are given previous under reactor coolant activity. I 1 Isctope Vapor ) i Kr-85 1 56 x lo l , 1 Kr-85m 6.34 x 10 1 Kr-87 2.32 x 10 2 Kr-88 1.01 x 10 Xe-133 1.23 x 10 Xe-133m 1.34 x 10 2 Xe-135 2.80 x 10 2 8.71 x 10 -1 Xe-135m 0 Xe-138 3 90 x 10 9-33
m -- TABLE 9-9 GAS DECAY TANK ACTIVITIES Assumptions: Volume of the tank immaterial to this calculation. Clad Defects in 1% of fuel rods. Operation at 3391 MWt for 280 days. Tank contains entire gaseous activity stripped off from the reactor coolant system. t Reactor Coolant System Volume is 12,760 ft". Isotooe Total Activity Curies Kr-85 5 52 x 10 3 2 Kr-85m 5 53 x 10 Kr-87 2 3 15 x 10 Kr-88 9 76 x 10 2 0 Xe-133 8.09 x 10 Xe-133m 8.43 x 103 Xe-135 2.47 x 103 Xe-135m 4.6 x 10 1 Xe-138 1.62 x 10 2 9-35
TABLE 9-10 CONCENTRATION OF IODINE ISOTOPES IN THE RECIRCULATION LOOP Recirculation Loop Isotope . Concentration (uc/ce) I-131 1 3 x loa 2 I-132 3 3 x 10 I-133 ~.5 x 10 3 2 I-134 3 9 x 10 I-135 7.6 x 10" TABLE 9-11 RADIATION SOURCES CIRCULATING IN RESIDUAL HEAT REMOVAL LOOP AND ASSOCIATED EQUIPMENT - MEV/CC - SEC Gamma Time After Reltase Energy Mev/y 0 1 Hour 2 Hours 8 Hours 1 Day 1 Week 0.4 1.8 x 107 1.6 x 10 7 1 5 x 10 7 1 3 x 10 7 1.2 x 10 7 1.1 x 10 7 0 O.8 1.3 x 10 0 8 1.5 x 10 1.2 x 10 1.2 x 10 7 8.0 x 10 7 5.6 x 10 7 6 0 i 13 9.70 x 10 7.0 x 10 5.2 x lo 8.2 x 105 5 5 x 10 1 3 x 10 2 1.7 5.6 x 10 6 3.9 x 10 6 2 9 x 10 6 4.7 x 105 2.1 x 10 4 3.3 x 10 2 2.2 5.3 x 10 6 4.2 x 10 0 3.4 x 10 6 6.h x 105 2.0 x 10 5 __ 6 6 25 1.9 x 10 1.4 x 10 9.4 x 10 5 1 5 x 10 5 8.1 x 103 __ 35 5.1 x 10 5 - 3.2 x 10 5 2 3 x 10 5 29xlo N 1.1 x 10 2 __
- 9-37 l
a ., - -
e Aa ..w .w. L. - am,.. ,e d*- e A,_ - 4 w % , r I
' PRESSURIZED WATER REACTOR RADI0 ACTIVE WASTE (RADWASTE) MANAGEMENT SYSTEM CHAPTER 10.0 CHEMISTRY 4
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TABLE OF CONTENTS Section Title Page 10.0 PLANT CHEMISTRY . . . . . . . . . . . . . . . 10.1-1 i 10.1 PRIMARY SYSTEM CHEMISTRY , . . . . . . . . . 10.1-1 10.1.1 Introduction . . . . . . . . . . . . . . . . 10.1-1 10.1.2 Basic Principles . . . . . . . . . . . . . . 10.1-2 10.1.2.1 Corrosion of Materials . . . . . . . . . . 10.1-2 General Corrosion . . . . . . . . . . . . 10.1-3 Local Corrosion . . . . . . . . . . . . . 10.1-5 10.1.2.2 Radiation Chemistry . . . . . . . . . . . . 10.1-6 Hydrogen-Oxygen Reactions . . . . . . . . 10.1-7 Nitrogen Reactions . . . . . . . . . . . 10.1-8 Physical Effects of Irradiation . . . . . 10.1-10 10.1.2.3 Nuclear Reactions . . . . . . . . . . . . . 10.1-12 Neutron Reactions . . . . . . . . . . . . 10.1-12 Fission . . . . . . . . . . . . . . . . . 10.1-22 10.1.2.4 Ion Exchange and Filtration . . . . . . . . 10.1-25 10.1.2.5 System Chemical Processes . . . . . . . . . 10.1-30 Chemical Addition . . . . . . . . . . . . 10.1-30 Gas Stripping . . . . . . . . . . . . . . 10.1-31 Waste Disposal . . . . . . . . . . . . . 10.1-31 Decontamination . . . . . . . . . . . . 10.1-32 10.1.2.6 Chemical and Radiochemical Analysis . . . . 10.1-34 Analysis Methods Summary . . . . . . . . 10.1-34 Chemical Specifications . . . . . . . . . 10.1-39 Corrective Actions . . . . . . . . . . . 10.1-39 l i
i LIST OF FIGURES a Figure Title Page 1 10.1-1 Corrosion of Iron . . . . . . . . . . . . . . 10.1-43 10.1-2 Total Corrosion of Steels . . . . . . . . . . 10.1-45 10.1-3 Release Rate of Corrosion Products . . . . . 10.1-47 10.1-4 Reactivity Change with pH . . . . . . . . . . 10.1-49 10.1-5 Reactor Plant Crud Buildup . . . . . . . . . 10.1-51 10.1-6 Water Chemistry Typical Operation Log . . . . 10.1-53 10.1-7 Plant Startup With Air in Loop . . . . . . . 10.1-55 ] i 10.1-8 Predicted Sources of Tritium . . . . . . . . 10.1-57 10.1-9 Reactor Coolant Mixing . . . . . . . . . . . 10.1-59 10.1-10 Reactor Coolant Tritium Buildup . . . . . . . 10.1-61 10.1-11 Containment Air Tritium Activity . . . . . . 10.1-63 10.1-12 Introduction of Fission Products Into Reactor Coolant . . . . . . . . . . . . . . 10.1-65 10.1-13 Fission Product Activity Versus-Core Life . . 10.1.67 10.1-14 Chemistry of an Ion Exchange Process . . . . 10.1-69 i 10.1-15 Typical Ion Exchanger System . . . . . . . . 10.1-71 4 4 l ii ' 1
i. 10.0 _ PLANT CHEMISTRY 10.1 PRIMARY SYSTEM CHEMISTRY
~~-10.1.1 Introduction All thermal process plants have three functional requirements which 4
must be met:
- a. The structural integrity and. operational status of the plant, including its mechanical and hear transfer components, must be assured at all times.
- b. The design of the plant must permit continuous operation and convenient maintenance,
- c. Process wastes must be controlled and disposed of safely in compliance with established standards.
In meeting these requirements in closed-cycle water-cooled reactors, the water chemist is confronted with a number of factors which.com-plicate conventional problems and add many new ones. These are;
- a. Ordinary chemical corrosion reactions.
! b. Chemical reactions resulting from the ionizing effects of the reactor radiation energy,
- c. Nuclear reactions, including fission.
Chemical reactions resulting from the high radiation energy in the j -reactors directly; influence-the chemistry of the primary coolant and the reactions between the coolant and container, and coolant and fuel. These reactions are of direct concern to the structural ~ integrity and o erational status of the plant. The creation of l. l; 10.1-1 l
radioactive elements by neutron capture of the plant sterials and corrosion products and by fission of the fuel material largely determines the convenience of operation, maintenance and waste disposal ~. The resultant task of water chemistry is two-fold: the identi-fication and evaluation of the effects of the nuclear variables
-and the development of means of control consistent with functional requirements .
10.1.2 Basic Principles Before considering the problems of water chemistry in terms of system processes or design problems, it is essential to have a good understanding of the principles of the fundamental chemical
, and nuclear processes which take place. Such an understanding is particularly important in nuclear applications because of the complex interactions which occur in the plant. Accordingly, we will briefly review the theories of corrosion as they apply to water technology, radiation chemistry, nuclear reactions, and ion-exchange and filtration.
10.1.2.1 Corrosion of Materials Most of the materials employed in nuclear power plants are thermo-dynamically unstable to pure water at the temperatures of opera-tion in these plants. Their practical utility depends in each case on the development of protective films of corrosion products which slow down the-rate of attack-to values which give useful lifetimes for the material. In addition to the general or uniform 10.1-1
Y type of attack, most metals are susceptible to one or more of a variety of local forms of corrosion, such as pitting, cracking, and galvanic or crevice corrosion, which are usually due to a specific component or impurity in the water or the mechanical arrangement of the component. General Corrosion The general or uniform corrosion of the most, prevalent materials in a closed-cycle water-cooled reactor, carbon or stainless steel, is illustrated by the corrosion of iron in pure water. At ele-vated temperatures iron corrodes to form magnetite Fe34* # "E to the following overall equation 3Fe + 4H O --1 2 Fe3 4 + 4H2 The detailed mechanism, which has been the subject of much in-vestigation is considered to be as follows. The iron atoms being thermodynamically unstable to water tend to form ferrous ions, Fe+ and electrons. The ferrous ions, in contact witl. water, fr cm ferrous hydroxide FE(OH)2 The electrons discharge hydrogen ions in the water to form hydrogen. This reaction is then Fe + 2 H2 O 4 Fe(OH)2 + H2 Ferrous hydroxide decomposes to form magnetite and hydrogen and t water, thus,
-- 0.
3Fe(OH) 2 Fe 034+H2+ 2 The physical picture of the growing film is represented by Figure 10.1-1. l i 10.1-3 4 i i
)
-~ - - , _ _ . _ _ .
t The rate is controlled by the diffusion of ferrous ions through the magnetite film, and is inversely proportional to the film thickness. From this a parabolic rate law is deduced which is in reasonable agreement with experiment. Typical corrosion i i stes of carbon and stainless steel are shown in Figure 10.1-2. Of considerable importance in. nuclear technology is the fact E that some of the corrosion product gets into the water, either Ebv erosion of the film or due to the fact that the reaction of
]
J i the ferrous ions to magnetite is not complete at the oxide-water f interface. If the base material is radioactive, such as core I cladding, the radioactive corrosion products are transported around the primary system. Evidence from nuclear plant operation indicates that some of the transported corrosion products are e i.
; incorporated'into the adherent corrosion films of the plant sur-f faces. This evidence indicates that the release of corrosion i
products to the system is not entirely due to erosion. A .l.
! significant amount of corrosion product is found as loose, finely divided material, called crud.
In considering the release of radioactivity to the system by corrosion of' core cladding, what . is significant then 'is the rate f at which material leaves the metal and enters the water. Cor-rosion product release rates are usually measured by obtaining the weight of exposed specimens before ano after descaling.
; This permits-the determination of the total amount of metal which has reacted with water, and the metal in the retained oxide
,]-
10.1-4 1-
- L
. . . - . g _.; ._ - , _ , . . , . _ . . , _., . , , _ _ . _ _ . , _ . . _ _ _ . -- . , ~ . _ , , . - _ . _ . _ . . _ - , . , . . . .
~
'l p.. film. The difference is taken as the metal released to water.
- If, as is now believed, some of the retained oxide film was deposited from the water, the value thus obtained is too low to properly reflect the problem of contamination of plants by corrosion of fuel cladding. The difference is at present not known and may be quite small.
4 I Figure 10.1-3 shows the effect of pH on the corrosion product release rates for carbon steel at low velocity. In this test water was heated to temperature, passed over the specimens, and the transported corrosion products collected. Similar measurements for stainless steel require tracer techniques be-cause of the low rates involved. The ideal method of measure-ment is to study a system consisting of a single material, say 4 stainless steel, part of which is made radioactive. The total release of activity, corrected for losses in purification and sampling, if any, and that carried back to the source, is a direct reasure of the total release rate. Measurements of this j type have been attempted for complete reactor systems, based on corrosion products and on induced activity. f Local Corrosion . Local corrosion effects lend themselves to limiting service life on metal components, rather than contribution to radioactive release as in the case of general corrosions. The various types of local corrosion, the susceptible materials, , and their causative agents, are listed below. i 10.1-5
l r ec . --- sten. Nucle:Ir ?cwer ?lar.ts Susceptible Recuired Corrosien Materials Causative Acent
- 1. Chloride stress Austenitic stainless Cl". 0, corresion cracking steels
+
- 2. Caustic corrosion Zircaloy, austenitic and Concentrated caustic ferritic steels at point of attack
- 3. Crevice corrosion Most materialc, rate de Crevice, 0,, salts
~
pends on general in water corrosion resistance
- 4. Galvanic corrosion Most materials, rate de- Coupling of dissim-pends on general cor- ilar materials, rosion resistance and conducting solution position in electro-chemical series
- 5. Pitting Ferritic alloys Chloride, 0 2
~
- 6. Fluoride' Stress Zircalloy, in'conel Fl , 0 2
General conclusions for primary water conditions for nuclear power plants: From the above discussion, the general requirements for primary coolant chemistry appear to be: low chloride (pure water), low oxygen, neutral or high pH. The problem of caustic corrosion does'not necessarily preclude high pH chemistry since high local caustic concentrations arc required, and this can be prevented j by designs eliminating concentration from the dilute bases employed. 10.1.2.2 Radiation Chemistry 1 The slowing down of fission neutrons by collision with water molecules in water moderated reactors, results in the disruption 10.1-6
,wy- ~-.. -- w , , , ,r --9
1 l of the - primary bonds of the water molecules. The coolant in the reactor thus has an enormous population of protons and radicals resulting from neutron collision processes. The gamma rays and some fission betas also contribute to vater dissociation. This combined rate is so large that all of the water of the plant undergoes dissociation in a few days. For-tunately the radicals' formed mostly recombine with each other , to form water again, so that the net effect is actually small. Nevertheless, the net dissociation of water under irradiation 4 can be significant and must be controlled. The significant parameters in the radiation reactions of water + are the intensity and nature of the flux, the temperature and the composition of the coolant, particularly with respect to excess H fU. Reactions are not necessarily limited to 2 2 hydrogen and oxygen. Nitrogen, which is frequently present also
; participates in radiation chemistry reactions. The radiation ; chemistry of water is too complex to be considered in detail, and the discussion here will be limited to_overall reactions, and the~ practical considerations affecting reactor operation under two headings: 1) hydrogen and . oxygen, and 2) nitrogen.
1 Hydrogen-oxygen reactions: At high temperature (500 F) the
-hydrogen-oxygen equilibrium j 1 9
- 2H O 2 2H2+O2 .
Y in a typical closed-cycle reactor it such that oxygen is no; , detected if H concentration exceeds See H /kg. The kinetics 7 7 i l 10.1-7 g--- ,--e - - -cw^, w- w - g- d g +-r*7 u- w--p y --m-+ e--e y p- g- =- Mry=
N. 4
- j. of the reaction are quite rapid in either direction with the reactor operatinc. Thus, if an excess of hydrogen is maintained in-the coolant, oxygen trapped in or added to the system can be u
rapidly converted to water. Dissociation is increased at lower temperatures but even at low temperatures the residual gamma flux in a shut down reactor can be used to recombine oxygen with i an excess of hydrogen, but at a much slower rate than with the i reactor operating at full ~ power. Since power reactors operate at high power levels only at high temperature, water dissociation j~ does not represent a large problem since it can be controlled by i maintaining a slight excess of H in the water, of the order of 2 25 cc/kg. i L Nitrogen reactions: Under reactor irradiation, nitrogen reacts 4 l with excess hydrogen or oxygen according to the following 1 reactions:
! Formation of ammonia --
= 2NH 3H2+N2 3 l Formation of nitric acid --
4 s.
, .g-- - 4HNO 2N + 502 + 2H20 3 depending on whether H or 0 is in excess. In the absence of an 2
excess of either, ammonium nitrate is formed. If conditions are 4
- changed, nitric acid can be converted back to nitrogen as shown by the following reactions
2HNO3 + SH2 - T N1 + 6H 20 4NH, + 30, ;7 2Ny+6HO 2 10.1-8 1
,4 , - --m ,~~,.7,..-en m---cy_,. ,,.n n . m ew.- ey g,e---my.ne.,,.,.,,,--.vepl,,- 4p.m.,----mn., .,
The nitrogen can react further according to the appropriate equations above~. Although at full reactor power the hydrogen-oxygen recombination
^
rates are too rapid for convenient observation, the ammonia synthesis and decomposition are much slower. In the range of f h'ydrogen concentration studied, the forward' reaction does not appear to be dependent on the hydrogen concentration. The nitric
acid reactions have not been observed quantitatively in operating
' plants because of the pote'r.tial damage to plant materials, but they appear to be more rapid than the ammonia reactions and ,
slower than the hydrogen-oxygen reaction. The effect of these reactions on the chemistry of closed-cycle water reactors is illustrated in Figures 10.1-6 and 10.1-7. Figure 10.1-6 shows the chemical conditions with high hydrogen and some nitrogen in the water. The conductivity and pH of the water are largely the result of ammonia, the concentration of which is determined by the rate.of synthesis and the rate of removal by a mixed-bed ion exchange purification system. Hydrogen losses are made up by additions. The concentration of suspended solids, or crud, is quite low. Figure 10.1-7 shows experimental results with high oxygen and nitrogen in the system, in the form of air added at startup. As the power level and temperature were increased, nitric acid formed and neutralized the ammonia "' present- f rota previous operation. Conductivity and pH decreased until pH 7 was reached; when the water became acid the conduc-tivity showed a corresponding increase. With the formation of 10.1-9
nir .;= t corresponding increase in chromate ion was observed, resulting from chemical attack on the materials. The whole process was readily reversed by adding hydrogen, which combined with the oxygen and reduced the nitrate and chromate ions. Physical effects of irradiation - deposition of solids: During the irradiation of fuel elements in in-pile loops, fouling has been observed at times in the form of more or less uniform deposibsoffinelydividedmagnetite,acorrosionproductofthe stainless steel of the 1 cop. The high heat Itransfer coefficients in closed cycle reactors require that fouling on the fuel elements be significantly lower than in conventional heat transfer appara-
, tus. Moreover, deposits of corrosion products would become radioactive and if later released would contaminate the system in the same way as corrosion' products from stainless steel clad-ding.
An additional disadvantage of deposition on the fuel cladding is that these deposits may contain boron. If this boron is removed frcm the core in a crud burst, during a plant startup or shutdown, a small positive reactivity insertion would occur. Associated with the deposition of solids is the pH - reactivity effect. If the pH of the coolant is increased there will be a i gainIin reactivity. The present theory of this is that as the pH increases, the structure of the deposit changes to one which has less neutron absorption properties as well as less resistance l .) to heat flow. . Consequently, as more neutrons are available for 1 l l 10.1-10 l I t
.-. , m. , , -- -- - --. -- -. -,
l fission and as the heat transfer increases, the fuel temperature will drop and a gain in reactivity will be accomplished due to the doppler and non-leakage effects. Figure 10.1-4 illustrates the, relationship experienced in the Saxton core of reactivity with respect to pH. The factors contributing to the formation of such deposits ar'e:
- 1) deposition as a function of the specific effect of irradiation.
Deposits have been formed at high ionization densities with electron beam irradiation; 2) deposition is heavier on Zircalloy than on stainless steel surfaces; 3) deposition is' lower at
- high velocities, but velocities up to 20 fps will not prevent
$ it; 4) operation at high pH, above 9.5, leads to less deposition than operation at neutral pH. Tests indicate that this ic probably due to decrease in corrosion product release rate, 4 rather than effect on deposition per se. General Conclusions j From the above considerations, it follows that the possible i adverse effects of radiation chemistry on primary coolant re-quirements can be controlled by maintenance of a small excess l of dissolved hydrogen in the water. Dissociation is suppressed, oxygen introduced into the system is converted to water, and nitrogen is converted to amnonis rather than undesirable nitric acid. High pH operation also assists in limiting deposition on irradiated surfaces. Figure 10.1-5 illustrates the mechanism i cf corrosion product deposition in the primary system. 1 l 10.1-11 ____ ___ -_____-_-__=_- - -_----- - - ____ _ ..
10.1.2.3 Nuclear Reactions _ Reactor ope ration is uniquely characterized by the nuclear reactions which take place. Of concern here are the chemical transformations resulting from nuclear reactions and the radio-logical consequences of the presence of the reaction products. The discussion c'onsiders separately neutron interactions other than fission, and the fission process. t Neutron Reactions-
> w Although protons, deuterons, and alpha particles are capable of producting nuclear reactions, the neutron is the most preva-lent important particle in reactors in this connection. Nuclei of different isotopes interact with neutrons with a unique
. probability dependent on the energy of the neutron. Capture probabilities (cross-sections) are generally higher for low energy or thermal neutrons, but some elements have high proba-bilities for capture of neutrons in narrow ranges of energy (resonance) generally at high energy levels, and some reactions leading to emission of protons or other particles require a high minimum energy of the incident neutron, so-called threshold reactions. The most significant neutron reactions for our considerations are the n,' reactions and the n,p reaction which are discussed in detail below.
(n.I/) Reactions The radioactive capture reaction referred to as a neutron gamma (n,)I) reaction in which a neutron is captured and a j 10.1-12
i gamma ray emitted, takes place with nearly all elements. The i reaction products ara sometimes stable, but most decay subse- l quently by emission of S praticles and additional gammas. Typical reactions occurring on elenents in the materials of construction are: 58 59 45 - 59 . Pe + n- + Fe days Co + + r l 4 Co + n
- 7+ Co
- years Ni +h+7 Mn 55
+ n* 7 + Mn 56 2.6 - 56 hours Fe + +7 1
These equations represent' the formation of radioactive isotopes 4 with the half-lives indicated, from naturally occurring parents. The initial gamma is emitted essentially instantaneously and thus occurs in the reactor. The delayed decay of active atoms after these have been transported from the reactor to the 1 system, is of major significance. Another significant capture reaction is that involving Li 6leading to the formation of I tritium. (n,p) neutron-proton reactions l Reactions of this type involving hi(;. energy neutrons also lead i to significant radioactivities in water-cooled and moderated 16 17 l reactors. First, O and 0 , occurring naturally, undergo I n,p reactions as follows: 0 + n --* p + N = see 0
- f.$ + Y i
i
- .O 17
+ne p+N 17 4.14 z 16
+
0 + sec i
- l. 10.1-13
.- - , -_. . ~ ,- , . , . . _ _ , . , _ _ _ _ . . . - . . . _ . - , _ , _ . _ , ,____ _-__ _ _ ,
These two reactions, involving highly energetic gammas and neutrons establish the requirements for shielding of the primary system of water cooled, closed-cycle reactors. Se-cause of the short half-life a decay of 1 to 2 minutes, effectively eliminates this activity. This consideration is important in the design of sampling and other secondary systems. Another class of n,p reactions is that involving structural S8 54 materials as targets. Thus, Ni and Fe , ,7 g elements, undergo n,p reactions as follows to form radiologic-ally significant radionuclides: 58 71 58 Ni 58 + n-= p + Co days r Ni + + ")/ Fe 54 + n-= p + Mn54 300 - Cr 54 ' e days + 7 +1 (n,a) reactions - poisons A few materials of low atomic number, such as Li and B , have very high capture cross-sections for thermal neutrons, with reactions as follows Li +n=H +a B + n = Li +a The latter reaction is of particular,importance because it pro-vides the potential of controlling reactor reactivity by incor-poration of B in the fuel, or in the coolant. Specifically, boron as boric acid in the coolant, is used for shim control of the reactor system. 10.1-14
. --- _ . . - . . - _ . . - _ _ . . _ _ = _ _ _ . .. . . . . . .-
f In view of the increasing significance of tritium in closed system wa :er inventory, for present generation pressurized water reactors, a discussion considering the implications of tritium follows. Tritium production results from, a) tritium yield from thermal fissioning of uranium, b) neutron reactions with natural lithium to be used as a reactor coolant corrosion inhibitor, c) neutron reactions with naturally occurring deuterium in the reactor coolant, and d) neutron reactions with the soluble-boron poison, j With resp,ect to item b, it has been recommended that all Westinghouse-designed reactors use 99.9 atom percent Li-7 both in the reactor coolant and in the lithium form demineralizer. Test results indicate that all the tritium generated in a i pressurized water reactor could be discharged into the plant 1 circulating water without exceeding 1/100 of the permissible discharge concentration. A graph depicting contribution from 1 the varicus sources is shown in Figure 10.1-8. Information gained as a result of recent operating experience with pressurized light water reactor power plants has success-fully demonstrated the "gettering" characteristic of zirconium clad fuel for ternary-produced tritium as compared to the in-f ability of stainless clad fuel to limit the release of this source of tritium. The comparison of tritium release data from two plants operating at approximately the same power level and 10.1-15
' l l
4 4 I i boron concentration, but with zircalloy clad fuel versus stainless clad, indicated that the tritium release from the plant with the zircalloy clad fuel is less than 10 percent of that released from the plant with the stainless clad fuel. 4 Tritium is a radioactive isotope of hydrogen which decays to stable He by emission of a sof t beta particle and without I attendant' gamma emission. The half-life of tritium (12.3 years) is relatively long. Because it emits a sof t bota (18.6 kev maximum) it has a short biological half-life (about 12 days), and does not appear to selectively concentrate in individual I body organs. Tritium is therefore of small consequence as com-j pared to other radioactive isotopes. This is evidenced by the l relatively high permissible discharge concentration in plant I liquid and gaseous effluents. For example, general public ex-posure from an intake of tritiated water.at a concentration of
-3 l 3 x 10 pc/cc is permissible, as defined by the International 4
Commission on Radiation Protection. Thus, the permissible f l discharge concentrations for release to aquatic environments, I as defined in Part 20 of the Code of Federal Regulations, are
. essentially those acceptable to the ICRP for drinking water.
In pressurized water reactors, where the reactor coolant is i j continually recycled through the core in a closed system, 4 l tritium exists primarily as water in both liquid releases, and as water vapor in gaseous releases. [ ' 10.1-16 i i I I -s __-__._._.__.-__________...._m. ___ _ _ _ - . _ _ . _ _ _ _ . _ _ _ . _ . _ _ _ . _
- _ - - - - __ - . ,~
s Ternary Produced Tritium One tritium atom is produced for every 1.25 x 10 thermal neutron fissions of U . Since approximately 90 percent of the power produced in current Westinghouse PWR reactors results from the fissioning of U , the tritium production from ternary fission translates to a tritium production rate of 0.0105 curies , of tritium per thermal megawatt day of power. Predictions for the expected release of ternary fissio- tritium are that one percent will be released from zircalloy clad cores, and 80 percent will be released from stainless clad cores. It should be noted that even if all (100 percent) of the tritium
~
produced in a typical PWR were released, the discharge concen-l tration in the plant liquid effluent would remain at a small fraction of MPC. To date, PWR operating experience with zirc- ) l alloy clad fuel indicates a ternary-tritium release close to the i predicted one percent. l l Absorber Rod Sources of Tritium l 4 The burnable poison (B 0 ) r ds, used in the initial cycle to 23 maintain a negative moderator temperature coefficient, contribute to tritium production from neutron reactions with boron. The neutron reactions with boron resulting in the production of tritium are: 1 , B 10 (n, 2a)T B (n, a) Li (n, na)T . 10,1-17 __ , .. _ , ~ _ _ _ . . . _ - . _ _ . _ . .-...__ _ . _ _ _ . . , _ . . ~ . - . _ _ _ _
I
- B (n, T) Be B (n, d) Be (n, a) Li (n, a)T
'1
, Of the above reactions, only the first two contribute signi-ficantly to tritium oroduction in a PWR. The B (n, 1) Be reaction has a thresho.'d of 14 Mev and a cross section of N 5 mb. Because the number of neutrons produced at this energy is less than 10' n/cm -sec, the amount of tritium produced from
- this reaction is. negligible. The B (n, d) reaction also may be neglected since Bc9* has been found to be' unstable.
The burnable poison rods are removed af ter the first cycle and their contribution therefore is not present throughout the life of the plant. Sitch metals as Ag-In-Cd are used for the rod cluster control rods, and these metals are not a source of tritium. Current Westinghouse estimates on tritium release from stainless clad burnable poison rods range from 30 percent to 80 percent of that produced. Reactor Coolant System Tritium Sources Tritium sources in the reactor coolant include neutron reactions with the soluble poison baron, the lithium used for pH control, and naturally occurring deuterium. Westinghouse-designed PWR's employ lithium for pH adjustment of the reactor coolant. A maximum level of 2.2 ppm lithium is maintained by the addition of Li , OH and by a cation demineralizer that will remove any excess of lithium produced in the B (n.a) Li reaction. 10.1-18 f
l a i The Li (n, a) T reaction is controlled by limiting the Li impurity in the Li OH used in the reactor coolant and in the lithium form demineralizers with 99.9 atom percent Li . Since the amount of naturally occurring deuterium in water is less than 0.015 atom percent the tritium produced from this reaction ,
- in PWRs is less than one curie per year.
I Measurement Experience Zircallov Clad Fuel Reactor
, To develop the desired data on tritium release from zircalloy-clad fuel, a program was formulated to follow the buildup of tritium I ; at a new PWR plant. The objective was te characterize the various
- 1 l tritium sources and activity levels in the reactor coolant system
! and to provide verification of predictions and plant design criteria.
The Ginna Station of the Rochester Gas and Electric Co. (RGE) was
- chosen for this study because it was the first large Westinghouse-j designed PWR with zircalloy-clad fuel to be placed into operation in the United States. A scheme for the regular monitoring of tritium l in various plant systems was developed with Rochester Gas and Electric.
i Stainless Steel Clad Fuel Reactors i i Review of the H data reported by Yankee Rowe reveals that following
- refueling shutdowns, an uncoma;n increase occurs in the level of H3
- in the reactor coolant. As Yankee Rowe was the only available PWR '
1 , operating with recycled fuel, it was decided that H behavior in 10.1-19
_ _ . _ _ _ . . _ _ -_ . . . . _ . . _ _ _ _ . _ _ _ m.. _ . - _ _ _ _ _ _ _ . _ _ _ _ _ . . . _ _ - . _ _ - i 4- - the Connecticut-Yankee reactor coolant would be studied during i-4 the' initial operations of Core II. Since both Yankee Rowe and Connecticut-Yankee employ stainless-steel clad fuel, it was believed that the Connecticut-Yankee refueling experience could
, confirm the phenomenon observed at Yankee Rowe.
i !- An inventory indicated tnat during operation 2/3 of the tritium 1 produc'ed.in the fuel of Core I-had been released through the i stainless steel cladding. ! It has been confirmed that relatively large amounts of tritium J i 1 were released from the Connecticut-Yankee core within the first l few weeks of operation. The release rate was seen to respond I i sharply to reactor operation at power levels of close to 100 per-cent, and would then tapar off after about 2 weeks. During this
. entire period, the reactor coolant boron concentration remained i
i essentially constant. The large release of tritium experienced following a refueling shutdown is attributed to a redistribution of power and temperature in the recycled fuel assemblies. I l- Recently, a partial refueling was accomplished at the RGE plant, i which employs zircalloy-clad fuel. During the first month of
- reactor operation after refueling no increase in tritium release t from the core was observed, j
j Long Term Bui? lup in Plant Retaining Tritium i Based on the assumption tiat all of the tritium in the plant liquid will remain within the plant during its entire operating j life, an analysis was performed to establish plant operating i t 10.1-20 i
. _ _ _ . . _ . , _. , . _ , . _ , - .- _ _ _ - - _ , , ~ , _ _ . . ,-,..-,-,,.,........~_....,_,_..,_,.,,.-..-._-._,,m .,_,. .
~ . . . .. . - , - .. _. .- -- . - . - . . . .. - - . . - - -
e-i i conditions. The distribution of the predicted tritium sources during normal operation and during refueling are shown in Figure a .; 10.1-9. No mixing between the spent fuel pit water and refueling i water was assumed for this study. However, actual plant operating j experience indicates that some mixing does occur, thus providing i i . additional dilution volume. 1 Figure 10.1-10 illustrates the buildup of tritium in the reactor coolant over a 32 year period.(equilibrium), assuming a plant i ! operating power of 1875 MWt and no discharge of tritium. The maxi- l mum expected reactor coolant tritium activity is 3.3 uc/cc i (4.5 uc/gm) of coolant, which occurs after approximately 12 yearr
- of operation. Currently, PWR's with stainless steel clad cores have operateo satisfactorily with coolant tritium levels in excess i ; of 5 uc/gm of coolant, t ,
a Figure 10.1-11 illustrates the tritium activity in the containment air during refueling. l l l The results of these studies clearly indicate that a PWP employing i chenical shim, silver-indium-cadmium control rods and zircalley j - clad fuel, has the potential to be operated for its entire life I without any need for intentional tritium duuping. From a practical peint of. view, it is anticipated that actual plant tritium lesses ! vill occur during containment purging and from unrecoverable riant i leakage. These losses (expected to be <100 curies / year) should 4
- remove any plant operating restrictions imposed by recycling tritium.
) l 10.1-21
, . ' e , . . , - .,,-rme u-..,.~-,.---,,..-.,,--n_- ,,g- ,,, ,,.---,,,-,.,--,m.
.,,,...,.-,.,,w-, ..,,.,n--. me ng
.. . .- . ._ - . ~ . - . . -
1 Fission The fission process leads to a characteristic distribution of radio-active elements. Most of the fission products lead to long chains. [ Thus, the fission products and their daughters cover essentially all elements in the mass range 72 to 161 with peaks at the noble gases krypton and xenon. The quantities of the radioactivity formed in power reactors is, from the point of view of radiological hazards, astonomical, and a reactor design consists of the development of safe means of containing the fission products first within the fuel, i
; then within the system, and finally within the container of the plant.
The fission product problem varies greatly with the nature of the fuel employed. For closed-cycle water cooled power reactors UO 2 is the ideal fuel because of its chemical and radiation stability. Experien'ce has indicated that it presents little or no problem of gross fuel element failures at the design heat release duties now j employed or contemplated. The major problems are the build-up of pressure of fission gases within undefected fuel elements, and the l escape to the coolant of fission products from fuel elements with defects. The latter problem is the one with which we are enncerned. The number of-cladding tailures to be expected in a reactor core cannot be readily predicted; considerable design and fabrication .t effort is devoted to prevent such defects. An estimate of this ] number and a knowledge of the escape of fission products to the i coolant from defected fuel elements is essential for design of i -e i 10.1-22
purification and waste disposal systems. For the Shippingport PWR it was estimated that not more than 1 percent of the fuel elements would develop defects. (Operation so far has shown a i' considerably smaller number, certainly not more than ten.) It was also assumed that the most probable defect was a small hole of the order of 5 mils (such as a porosity in a weld). A program was carried out in which the escape of fission products to the water was measured under irradiation for samples containing defects. Tests were also made with cracks and with multiple defects per rod under steady and cycling conditions. It was found that activities of medium to long half-lives escape by a diffusion mechanism; the
~
rate of escape is proportional to the amount present in the fuel rod. Figure 10.1-12 illustrates the method of dif fusion. Shorter lived activities must initially come from the surface of the fuel by recoil since they do not live long enough to diffuse out of the 4 fuel. The results of the study are suenarized in the following table which presents the escape rate coefficients, v, for fission products for rods similar to those employed in most UO2 - ueled, l water cooled power reactors. The escape rate coefficient, y, is the fraction of accumulated fission products escaping per unit time. ESCAPE RATE COEFFICIENT, v, FOR A NORMAL ELEMENT Relative ~ Orders Elements v, sec 8 - 1 Cs , I , Xe , Xr , Rb , B r 1.3 x 10 2 Mo 2 x 10-3 Te 1 x 10 -9 4 Sr, Ba 1 x 10 ~I
~
5 Zr, Ce, other rare earths 1.6 x 10 10.1-23 ,
4 The escape rate coefficients of the elements appear to vary relatively in the same way as the volatility and solubility of the chemical species. They are highest for elements in Groups I, VI, VII and VIII of the periodic table, and are lower for Group's e II, III and IV. Absorption characteristics and resulting con-tamination follow the opposite trend. They are essentially zero l ! for Groups I, VII and VIII, and highest for Groups II and IV. Tellurium, Group VI, is also strongl;' absorbed. It should be noted that to be significant for contamination, an isotope must have an appreciable half-life and be an energetic gamma emitter. l These factors and the nature of the decay chains combine to favor the use of UO as a fuel, as illustrated by consideration of three 2 2 95 91 141 1 significant contaminant isotopes: Zr ,Y and Ce . The isotope Zr has very short half-lived precursors, and a low escape rate coefficient, v, and therefore, by both indirect and direct j paths, has a low escape to the system. Y' has short-lived Kr and I 91 1 Rb precursors, and although its immediate precursor Sr has a significant half-life (9.7 hours). The escape coefficient of Sr i 141 + is low. The considerations for Ce are analogous to those of Y . A similar situation applies to other isotopes that are signi- ! ficant' radiological hazards; for example, I and Sr . Although i I escapes readily, it is not absorbed in the system and is readily removed by an ion exchanger. With the information thus obtained on the release of fission products, 1 analytical studies have been made of the fission products which could 1 [ be expected in the reactor for a given number of defects, and of the 4 l 10.1-24 l
r a requirements and expected performance of the purification and waste disposal systems. Figure 10.1-13 projects fission product activity in coolant with respect to core burnup. 10.1.2.4 Ion-Exchange and Filtration It will be evident from the previous discussion that many of the problems of transport of activity in closed-cycle, water cooled reactors would be minimized by the employment of means for separa-tion from the water of dissolved and suspended impurities. From the very early days of the development of such reactors, therefore, consideration was given to the application of ion-exchangers and filters for this purpose. Ideally, the separation process should be carried out at the normal temperature of operation so as to eliminate the necessity for cooling the process stream. (See Figure 10.1-14) Most dissolved solids can be removed by the ion exchange process. Ion exchange material has the ability to exchange one ion for another, hold it temporarily in chemical combination, and give it up to a strong regenerating solution. Impurities that dissolve in water dissociate to form positivcly and negatively charged particles. The positive ions are named cations. Examples of cations are hydrogen (H+) and the various metals such as ferrous iron (Fe ) and lithium (Li+). The negative ions are named anions. Examples
~
of anions are hydroxide (OH ) and the various acid radicals such as borate (B e). Various chemical compounts can be formed by matching up the cations and anions: l 10.1-25 i
r i Hydrogen and Hydroxide form Water 4
- . H0H (H,0)
H+ + OH r . i Ferrous Iron and Hydroxide form Ferrous Hydroxide j I y FE(OH)2
; Fe + 20H-As we saw before this is part of the corrosion reaction. The ferrous hydroxide will break down into magnetite and hydrogen and water. < Hydrogen and Borate form Boric Acid 3H+ + Bo H Bo 3 7 3 3 J
Lithium and Hydroxide form Lithium Hydroxide
~ 10H Li+ + OH 5 1 i l The Cation Demineralizers in the CVCS are used to remove lithium and 37 control cesium activity in the coolant from fuel defects. These
~ ~
are charged with hydrogen base cation exchange resin (H+R ) . The R i
- represents the resin bead to which the hydrogen is attached. (This l
is only symbolic.- Actually there will be many hydrogen ions attached to each resin bead.)
~
When a litnium hydroxide molecule (Li+0H ) contacts the cation resin, an exchange takes place. The resin picks up the lithium and releases j the hydrogen which was attached to it. The hydrogen combines with the. hydroxide to form water. 4
~ ~ ~
Li CH + H R > Li R + H OH" A similar reaction will take placa for cesium (Cs+). , l i 10.1-26 l I
The Deboratine Demineralizers in the CVCS are anion exchangers and are used to remove boric acid at the end of core life. These are
+
in the hydroxide form (OH- R). When a boric acid molecule contacts the anion resin, the ions ex-change places. The resin picks up the borate ion and releases the hydroxide. The hydroxide then combines with the hydrogen from the beric acid to form water. I 1
~
H+3bo! + (OH )3R---tBo a e
+ 3H'~ i OH~
The deborating demineralizers of course have the ability to remove other anions, besides the borate, such as chloride and iodine. As was noted earlier, resin has the ability te release the ions it has removed from the water to a strong regeneration sclution. The deborating and evaporator condensate demineralizer are regenerated in this manner with sodium hydroxide (Na' OH'). After backwashing to remove insoluble particles and to loosen the resin bed, sodium hydroxide from the ca:stic tant; is pumped through the demineralicer. When the sodium hydroxide contacts the anion resin to which the borate ions are attached, the exchange takes place. The resin picks up the hydroxide and gives of f the borate. The borate then combines with the sodium to form sodium borate which is sent to waste.
* ~
+ 3Na OH~ R +
Be} R > (OH )3 (Na )3 Bo} Af ter rinsing to remove any remaining chemicals, t h- tesin is returned to its hydroxide form again ready to remove more anions. 10.1-27
1 Anion exchangers are also referred to as base removal ion exchangers. The mixe1 bed demineralizers contain an intimate mixture of cation , and anion resin. In the CVCS they are used to remove any ionic impurities except lithium hydroxide and boric acid. The cation
+-
resin in the mixed bed is of the lithium form (Li R ) ar.d thus I cannot remove another lithium ion. The lithium can be replaced by other cations which are more strongly attracted to the resin. The anion resin in the mixed bed will originally be in the ,h,ydroxide form but will soon be transformed to the borate form (Bo 3 R") by its removal of boric acid from the coolant. After the anion resin is completely saturated with borate, it will no longer remove boric acid but will remove other anions such as iodine. In addition to being chemically bonded to the resin some of the boric acid will be absorbed by the resin beads. This reaction will continue until equilibrium is established. This boric acid " concentration" in the resin will attempt to match the concentration in the coolant. If the concentration af boric acid in the coolant is increased, the anion resin will absorb more boric acid. If the concentration in the coolant is decreased, boric acid will Jeak out of the resin l nntil equilibrium is again established. Unlike boric acid, lithium 1 will not be absorbed by the resin. The mixed bed, being a mixture of cation and anion resin, will remove both cation and anion impurities from the coolant. If a salt (metal and an acid radical), such as ferrous iron chloride t - j Fe (Cl )2, e ters the mixed bed, it will come in contact with a 10.1-28 _ . _ . ~ _ . _ _ - . _
f cation and an anion resin bead at the same time. The ferrous ion will be picked up by the cation resin which releases lithium. The f
- chloride will be picked up by the anion which releases borate.
The lithium and borate will thea leave the demineralizer and enter I the coolant. . i The mixed bed demineralizers do not contain equal quantities of , anion and cation resin. The actual mixture is approximately 2 anion for 1 cation (66% anion, 33% cation). This mixture will ensure that the mixed bed effluent contains neutral molecules j rather than charged ions. Tne ef ficiency of the mixed bed depends on the cation and anion
; exchange occurring at the same time. This in turn depends on the resins being in an intir. ate mixture. Since backwashing would cause the lighter anion to separate from the ivavier cation, the efficiency would be reduced by backwashing.
An exa # 6 6# typical ion exchange system is depict d on Figure j 10.1-15. When an ion-exch . ige resin is about 9 C *. t o - C '~ sxhausted,
- impurities will begin te leak through the resin bed. vill then be necessary to replace the resin or regenerate in the cis- of the i
anion exchangers. Experience with ion-exchange beds alsc iceo,rtrated that they are quite ef fective as filters for the solids found 4 reactor coolant water. This factor also contributed t3 ene .. - - tion of filters. i Because of the reduced effectiveness and possible da- m to reatn. ! high temperature water must not be passed through the ien exchancer 10.1-29
}
i The resins are organic and if operated above 145 F they will break down and cause fouling of the reactor coolant system. The ability of an ion exchanger to remove a given ion is measured by its decontamination factor (DF). Activity into ion exchanger DF = Activity out of ion exchanger 10.1.2.5 System Chemical Processes Chemical Addition Hydrogen is introduced into the coolant at the volume control tank. It is used to reduce corrosion, suppress the dissociation of water into hydrogen and oxygen, convert any oxygen introduced into the l system to water, and prevent the formation of nitric acid from , i l I nitrogen. I Lithium Hydroxide is introduced at the chemical mixing tank. It is used to control the pH of the coolant in order to reduce corrosion and control the deposition of corrosion products on heat transfer , surfaces. I
- Hydrazine is also introduced at the chemical mixing tank. It is i
used to remove dissolved oxygen from the coolant during startup. Hydrazine reacts with any oxygen present to form ditrogen and water. NH + v2 2H O +N 24 2 2 , ! It will react slowly below 100 F and will decompose rapidly to ammonia above 400 F. It will be injected into the coolant in the temperature range 200 - 250 F. l l I 10.1-30 i
Boric Acid is introduced at the volume control tank or charging pump suction. It is used for reactivity control. das Stripping Some gases ionize when they are dissolved in water. The gases which ionize are removed by the ion-exchanger, but those which do not ionize pass through unaffected even though they are in solution. Oxygen, nitrogen, argon and hydrogen and the noble fission gases, xenon and krypton, are not removed by ion exchange. These gases must be stripped from the coolant. The gas stripper in the recycle stream of the chemical and volume control system will remove any dissolved gases from the letdown before the evaporation process is carried out. The principle of operation involves breaking up the incoming feed into small droplets so the water may be heated by the rising steam vapor and the gases can leave quickly. As the water is heated, the gases come out of solution and are vented off. The gas free liquid is collected at the bottom of the stripper and is pumped to the evaporator. The volume control tank in the CVCS will be used to strip hydrogen and fission gases from the coolant prior to a cold or refueling shutdown. This is done by increasing the letdown flow rate and venting the tank. Waste Disposal The means employed for waste processing and disposal will depend on the coolant cheeistry. First, because a large part of fission product 10.1-31
activity is gaseous, a stripper is employed, and the gases are collected and stored until they have decayed enough to be released to the atmosphere at allowable levels. Because of the hydrogen content of the coolant, there is a danger of explosion and the J hydrogen must be removed by combination with oxygen in a blanket gas. Soluble and insoluble activity must be reduced in quantity to levels such that the purified water can be conveniently re-used i in the plant or dumped. If the total solids content is low, ion-exchange can be used. If the solids content is high, on the
- average, such as in plants utilizing chemical control for shut-4 l down or shim, evaporation is preferable. Some evaporator capacity j
is always required for non-primary coolant wastes such as inhibited cooling water, laboratory wastes and so forth. j Decontamination l i l Direct cycle water-cooled reactor systems may become highly active i from one or more of three sources. These are: 4 59 y $4
- a. Induced activities -- Co . Co , Fe
- b. Fission products separated from UO such as -- Ba, Sr. Ce, La, -
2 Ru, Yt, Zr, Nb.
- c. to, or uranium corrosion products containing fission products.
~
In UO 2 ueled react rs, the first two sources are most likely (in i the order stated) to be the sources of activity of such levels as to ' merit consideration of methods for activity removal or decontamination. In these two cases, the activity will be essentially chemically
, 10.1-32 '
i l
- . ~ . - - - . . .
i combined in the transport corrosion products and the adherent corrosion film on the system surfaces. Accordingly, the objective of a decontamination procedure for this situation is to remove the corrosion products from the system with the combined radio-activity, without impairing the structural and operational integrity of the plant, and in as simple and convenient a manner as possible. The technique employs successive treatments with alkaline per-i nanganate and annonium citrate solutions (APAC). The solutions and their conditions of application are as stated below. APAC DECONTAMINATION Step Component and Concentration Temperature Time 1 10% NaOH-3% KMnO 4 108-115 C 90 min. I 2 Drain and rinse with water 65- 95 C 15 min. 3 10% ammonium citrate 90-100 C 120 min. ; i 4 Drain and rinse with water 65- 95 C 15 min. i ) ! Tests indicate that the first step essentially breaks up the cor-l rosion ' fin without appreciably suspending or dissolving it. This is inferred from Inw iron and activity pick-up. The bulk of the j material and activity pick-up is achieved in the third step. Velo-city is an important factor and for good results should not be under 7 fps. Under ideal laboratory conditions, decontamination farters j of 1000 have been achieved. Tests in small systems have not been a= j successful, but verv useful results have been obtained in all appli- [ . cations to test s y s t r.as . There has so far been no application to'a j closed-cycle reacter plant , but several are being planned. i 10.1-33
~ _ . _ _ _ _ _ _ - _ _ . - _ _ - _ . - . . . ~ . _ . _ - - _ _ _ _ _ - - _ _ _ _ _ . _ _ _ _ _ _ . - _ _ . . _ - _ _ - _
i i i I e 10.1.2.6 chemical and Radiochemical Analysis Methods Chemical analysis involves the determination of the composition of a substance and/or the separation of a substance into its con-
- stituents. Radiochemical analysis deals with the separation and i
purification of radioactive isotopes and their measurements. The description which follows will briefly cover the analysis methods for the important chemical and radiochemical constituents of the reactor coolant. 4 L Analvsis Methods Summary a Dissolved Oxvgen (ppm) Amount is determined by manual titration or technicon (automatic titration). Lithium (ppm) Determination is made by flame photometer. Boric Acid (ppm)
; Amount is determined by acid-base titration.
i
! Conductivity (umhos)
Determination is made electrically by probes in a sample. The measured conductivity must be corrected to 25 C. f 1 Total Suspended Solids or Filterable Crud (ppn) Amount is determined by weighing a filter paper through which a known volume of reactor coolant has passed. i 1 l 10.1-34 t
i { Total Dissolved Cases (cc/kg) Determination is made by depressurizing a sample bomb into a column ] of water, where the water level rise is proportional to the amount i of total gas. 4 1 Dissolved Hvdrogen (cc/kg) Amount is determined by stripping the gases from the above sample and passing it through a hydrogen analyzer. pH (no units) { Determination is made electrically by probes in a sample. Since pH affects both corrosion and radiation chemistry, it will be beneficial at this point to go into a more detailed discussion concerning pH. pH is a measure of the hydrogen ion concentration in a liquid. In addition to molecules of H 0, pure water contains separated parts 2
~
of colecules called hydrogen ions (H+) and hydroxyl ions (OH ).
~
One liter of pure water at 25 C (77 F) always has 10 moles of
~
hydrogen ions. There are also 10 moles of hydroxyle ions in the same liter. By definition, pH is the negative logarithm of the hydrogen ion concentration in moles / liter. The pH scale goes from j 1 to 14 with 7 being neutral. A pH below 7 is said to be acidic, while a pH above 7 is said to be basic or alkaline. A number of factors will affect the pit of the coolant l When the temperature of pure water is increased more molecules of H2 O will dissociate into hydrogen ions and hydroxyl ions. If we measure the pit of the pure water at this higher temperature, the increased number of hydrogen ions would indicate a lower pil. This l 10.1-35 I
. 1 in turn would falsely indicate the pure water as being acidic. Because >f this ef fect. it is important to always measure pH at the reference temperature of 25 C. 1 I Boric acid in the coolant at room temperature will add hydrogen ions and thus lower the pH, But as the coolant temperature is increased to its operating value the boric acid changes to a less ionized form and therefore adds iewer hydrogen ions to the coolant. Due to this change, the pH of the coolant at operating temperature will not be greatly affected by varying the boric acid concentra-tien. Lithium hydroxide adds hydroxyle ions to the coolant which raises its pH. The main control of the coolant pH will be the lithium hydroxide concentration. Activation and Fission Products The activity of a specific isotope is determined by first chemically separating and purifying the element under investigation. These radioactive isotopes are present in such minute amounts that the ordinary means of chemical analysis are not adequate for their re-covery. Therefore car.iers are added in sufficient amount to mke the chemical procedures applicable. The carrier will usually be a non-radioactive isotope of the element to be measured. After separation, the isotope is weighed and counted. The counting pro-cedure used will depend on the particular isotope being measured. The counting procedure can be demonstrated using iron 59 as an example. The purified iron is first counted by gross gamma. This 10.1-36
j i l 1 reading is given in counts per minute which can be converted to activity in micro curies. The specific activity of the sample is i , i then calculated by dividing the activity by the sample weight to ! i i obtain uc/mg or dividing by the sample volume to obtain uc/ml. A f correction factor, the yield, is applied to the calcuiations to ' take into account the addition of the carrier and the amount of iron 59 not recovered fr'm o the sample. i
; The iron 59 is also subjected to a gamma spectrum count. This is 1
1 a plot of energy of the gamma radiations versus counts per minute. j Since iron 59 releases gamma photons of specific energy, the plot , I r
! will show peaks at these energy levels. This plot can then be !
j compared to a standard plot to insure the isotope being measured l I j is iron 59 and that no impurities are present. 1 i j i Other isotopes are subjected to similar counting procedures. ;
; Where two isotopes of the same element may be found in a sample i
i j (cobalt 56 and 60), more complicated counting procedures are i ! required to find their separate activities. [ r
- r i Measurement of tritium concentration is perferned by a consultiny.
i
- labo ra t o ry . j l
t j rhemical Specifications 1 I 1 ! j parameter I.imi t Renmon for I.inft r 1 j A - Oxygen, max. 0.1 ppm Reduce correnton '
. It - Chloride, nax. 0.15 ppe Prevent strese corresten
- C - r!uoride, nax. 0.1 ppn Sun.e es chlorian. Tha
i zircalloy cladding 16 pnt- , ticularlv endreptible t' r (; attach by fluoride . f 10.1-37 l 1 1
l I 4 1 l
- Parameter Limit Reason for Limit ,
i , D - Hydrogen, range 25-35 cc (stp) Min requirement to reduce l per KG H O corrosion, suppress dis-2 ' sociation of water into
; H, and 03 , convert any oRygen introduced into the system to water and j prevent the formation of i nitric acid from nitrogen. !
j Max. requirement to pre- , i vent waste of hydrogen. ! ] E - Total suspended 1.0 ppm Reduce deposition i solids, max. ; i i l F - pH 9ontrolagent .22-2.2 ppm Min requirement to reduce i l (11 ), range corrosion and control de- .' position of corrosion pro-I ' ducts. Max. to prevent
. possibility of caustic embrittlement.
I { G - Boric acid, range 0-4000 ppm Max. to prevent poritive I j boron moderator coefficient. t ' H - Electrical conducti- 1-40 umhos/cm Measure of amount of vity, expected range @ 25 C deter- dissolved solids.
, mined by con- ! centration of l boric acid and
,I lithium hydroxide ,
! - pH, expected range 10.5 0 Min, requirement to reduce I
4.j 25 C deter- corrosion products. Max. , j mined by con- requirement to prevent possi- l 1 centration of bility of caustic embrittle- " i boric acid and ment. ! j lithium hydrox- l j ide. See note i below. [ Note: The pH range of 4.2 - 10.5 may be confusing because it was i previously stated that a low pH, like 4.2, would cause a . higher corrosion rate than a pH above 7.0. It must be ! i remembered that this pH will only exist when the reactor !
; coolant is at room temperature. In addition, the corrosion rate is much smaller at lov temperatures. Because of this, l
the low pH at low temperature will not greatly increase cor- 1
- rosion. l I
l I i 10.1-38 { i t 1 I i 4 i
=
J
f As the coolant temperature is increased to its operating value the corrosion rate would also tend to increase. Due to the temperature increase the boric acid will change te a less ionized form and the pH will increase until at operating temperature the actual pH will be above neutral. Make-l'p Water Chemistry - Reactor Coolant System Parameter Limit Reason for Limit A - Oxygen, max. 0.1 ppm Reduce corrosion B - Chloride, max. 0.15 ppe Prevent stress corrosion C - Fluoride, max. 0.1 pp= Same as chloride D - Total solids, max. 0.5 ppm Reduce deposition E - Carbon Dioxide, max. 2.0 ppm Reduce corrosion from carbonic acid (H 2Co)) F - Particulates Filtered to less Keep suspended solids than 25 microns down to reduce the amount of activation products and prevent large material from entering the reactor coolant system. C - Electrical conduc- 2.0gmhos/cm Indicates high dissolved civity, max. @ 25 C solids H - pH. range 8.0 0 Min. requirement to reduce 6.gC 25 corrosion and control de-position of corrosion pro-ducts. Max. requirements to prevent possibility of caustic embrittlement. Corrective Action List System Troubl,e Possible Cause Corrective Action Reactor High Oxygen 1. Low hydrogen 1. Increase hydrogen pressure Coolant Concentration concentration in in volume control tank to System the reactor cool- remove oxygen.
""E "I**
- 2. For startup from a condition
- 2. Oxygen in where the reactor coolant makeup system has been opened hydrazine may be required.
Add to chemical mixing tank.
- 3. Check oxygen concentration of makeup.
i 10.1-39 i i 1
- _ _ _ . _ _ _ _ . _ _ . . _ . - _ _ . ~ _ _ _ _ _ _ _ _ .__.-~->m.__.
l 4 l j evstem Traubte PossiSte Cause Correctiv- Action 2 Keactor 91gh chtcride 1. hpurities in 1. Check makeur eSentatrv Coolant or fluoride makeup ,p gg , ; Systsm concentration ,y g ,, I chlorides or fluorid.n. from i j the reactor coolant ' I 3. Increase letdewn flow ( rate as required. l Low hydrogen 1. Low hydrogen 1. Increase hydrogen pren- l 1 concentration pressure in the sure in volume control tank '
" I"** # "E# 1 2. Check oxygen concentra- ,
) tion of makeup. j 2. Excessive r oxygen added in { j makeup High hydrogen 1. Excessive hy- 1. Decrease hydrogen pres- ! c)ncentration drogen pressure sure in volume control tank. l in volume control J
! tank i High suspended 1. Crud burst 1. Check makeup chemistry. !
solida concen 2. Impurities in 2. Check for possible source [ E#"EI " makeup of increased corrosion. {
- 3. Increased 3. Reduce concentration in I l corrosion in reactor coolant by filtration E j reactor coolant using mixed bed demineralizers l 1 system. as filters. :
< r i 4. Increase letdewn flow r j rate as required. i tow 11thit.m 1. Long period 1. Add lithium hydroxide at j concentratien of dilution the chemical mixinr, tank. l liigh 11thiun 1. Lithium build- 1. Reduce concentration using j concentration up from baron-10 cation bed denineralizer. i teaction. j 2. Addition of } 1 excessive amount ! j of lithium I hydroxide. [ ] t 1 Low pil 1. Long period 1. Check lithium concentration ! of dilution, and if low add lithiu- hy.trovile ! { at the chemical mixing tank. l 2. Impurities in ; i nakeup l l l liigh pH 1. Lithium build- 1. Chee'r lithium cencontration j
- up irom borot, 10 and if high reduce 11thium !
1 reaction concentratten using cation l } bed d.imineraliser. l
- r I
j 10.1-40 i l
- 1 1
i .. . L
l i i I i i Svstem Trouble Possible Cause Corrective Action ; Reactor High pH 2. Addition of 2. Check makeup chemistry Coolant excessive amount 3. Remove impurities f rom System of lithium hydro reactor coolant using mixing i xide bed demineralizers. (
- 3. Impurities in i makeup [
High radio- 1. Fuel element 1. Reduce activity using f activity Icvel cladding failure, mixed bed demineralizers. ,
- 2. Crud burst. 2. Place cation bed deminer-alizers in service if neces- [
sarv. ,
- 3. Increase letdown flow ;
rate as required. (
)
I l 5 f i I f t t f I I t I i I i 10.1 41 i e
Fe 0 34
* =
e +n+ =H
; Fe *+20H" 2 Te 034*N2
/
\\
Tigure 10.1-1 Corrosion of Iron 10.1-4)
f 1 I i l 1' , i j 1 1 10000 i i i 1000 i
! 500 -
I I
\ !
200 A CARBON STEEL PH 5-1C l N
, 103 s' - m m WATER i v i
l N E
- N \ q w
j
\ N N PEAN LINE CARB0', LTEi. !
(EASED ON pH 10 12 M T/ l b 20 \ u l ) i
\ !?AINLESS STEEL Pd 6 10
,t. 3. e. .
WATER ! N t*A!NLESS STEEL w!Gu , I 9 a I n .< i
/ ': ? 50 100 1: y,
] Tr A; tyr.ns H . H";URi 4 l 4 I t t t i F MJPI 10.1-2 -
T/L 00kPN : 'r ; ![,s ir 1 I
L f 10.1-45 I ) ' a
B i r l 15 ! I t } 10 1 i 1
)
8 *a i ! t i ; i i 6 . n ! E
! a
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- 4 a e W '
i, a e M i i E \ W 2 i i f % i
! O e
- i E
0 O m d d s I l l 6 7 8 9 10 11 l LOOP WATER pH ! I i i l FIGURE 10.1-3 . RELEASE RATE Or CARB0'd STEEL CORROS!0N PRODUCTS I i
! i
! L i I t i l l 10.1 47
I 0
-- 6 G
V
-- A 3
e 1 I 0 8 4
._ =
H H p p T
- _ D t N E / 1 A - _ D o 7 L . , D t 1 O A O m1 0 C H N O = 1 6 H0 i 3 TP L o
- IO
- WB Nt S R
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- E F EN gR
- I T
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,_ gU p E AX HA CS 1
4 2 - 3~ 1 1 1 0 E 1 N G O N E I A R T H U GCE X G I F F
- - _ _ _ - 0 5 0 5 0 5 5 4 3 2 i 0
- 3 3 3 3 3 3 7 7 6 6 5 5 5 5 5 5 5 m *4 eE n
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'9l
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to w 2 Oru > OO v. -
- CD 2
! D O OC O O,
;: g O cC g d w- ,
U c,- c.: Or* "- r C - eq s .J <- i 1 p 2 a - C{ 3W'! 738 g g ,,a U c.. J C pgyM u
< - s '- ,
- e. . o ,0 6 J .; < m.,
- s. L. -
c L r_ n *- < G 7 cE CC C g O r G or
;2 <
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W W J Q w C
=.~;mr: ~ ::.:..: . :::: :-:: .
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- : c. .: . -
=~=. :: :,:
2 .~'.~..,..?.=. 2: ~::::.-
,, i.e .,=
- Cwm Q . ' ~;. .-: *. ; ;.
,~ .
- QQH W
g *W" . . ; M N M O u. o
. w, ; . w 2 . - . .- - .
Ob Ua .S'b : . ~:,h:j? !l :Y,:- Y UL
~
. 1 m. .. W DW 4 r:.:-:r:. . _:- e:.::- .,:a-. m.~* *. :
CmO C : ?. * : ' -J. . ~ *, ';T '
-::. ': * . n ! . ::.7%:6. M ' ' ~
=? -: e ~- :. : u.:
- c : -.w-:
m: , ~:.: Figure 10.1-5 10.1-51
CONDUCTIVITY
=
u mhos/cm A 0.2 ~ pH
=W X _
7 - HYOR0 GEN cc/kg 120 - l I I 100 - l ~ ~ - ,I 80 " - "lh HYDROGEN ADDED CRUD ppm 08 - 07 . , . I I I ' I I i i i i'1 I ! 8 16 24 8 16 24 Figure 10.1-6 WATER CHEMISTRY TYPICAL OPERATING LOG 0 TEMP. 450 F POWER 8 TO 30% OF FULL POWER FILTER AND DEMINERALIZER IN SERVICE 10.1-53
I 400 = O TEMP. F j 300 - 200 -
/\
OXYGEN l 60 - cc/K9 [ HYDROGEN cc/Kg f
=
/
/
20 = / H ADDED AT ARROWS l 2 t 49 I 10 "
, CONDUCTIVITY 8 = u mhos/cm l
6 , 4 , 2 , i 9 l .
- l
- 7 .
5 pH i , l 4 . SOLUBLE Cr 3 - ppm % NITRATE 2 . ppm i 1 . 0 .n.. ... n...n...n...a...n... 20 24 4 8 12 16 20 24 HOUR OF DAY
- PRESSURIZED WATER REACTOR PLANT START-UP WITH AIR IN LOOP WATER CHEMISTRY (LOWPOWER)
Figure 10.1-7 10.1-55 I I- --. - .- - - - - - . - . . - . - - . - - , . - - , - - - . - - - -
.( .1 400 LEGEhD: p A - TERN ARY FI S$10N TRITIUM (If 0F TOT AL PRODUCE 0)
B - BURNABLE POIS0h TRITIUM SOURCES (80! 0F TOTAL PRODUCED) C - SOLUBLE Pols 0h 1RITIUM SOURCES 350 - o . TOT AL PREDICTED REACTOR COOLANT TRITILH 300 - 2
$ 250 -
5 S C ' { MO - C C E I 150 G e-100 60 A 20 1 i i I I o 0 30 60 90 120 150 180 210 240 EFFECTIVE FULL POWER (DAYS) Predicted Reactor Coolant Sources of Tritium for o PWR Operating at 1300 Mwt Figure 10,1-8 10.1-57
NORMAL OPERATION REACTOR COOLANT RECYCLE HOLDUP SYSTEM ; TANK 3 3 (VOLUME - 6000 FT ) (VOLUME-7500FT) n REACTOR MAKEUP WATER STORAGE TANK 3 (V0LUME - 7500 FT ) l REFUELING OPERATIONS REACTOR COOLANT REFUELING WATER SYSTEM y STORAGE TANK (VOLUME - 6000 FT ) 3 (VOLUME-46,800FT) 3 l n f i SPENT FUEL PIT (VOLUME - 80,000 FT3 ) Flow Poths for Mixing Reactor Coolont During Normal Operation and During Refueling Figure 10.1-9 10.1-59 l
f u.o . l I 1
; 32
_ i 10 7 2.0 - a e i s r , O II eu s - c
=
0 2 '; e i.o l c -
) *p 0.8 l u -
! 0.6 4 - '
i 2 o.n i d 8 i = l W
- o. 2 -
J i 1
! l I I I I , 0.1
- 0 2 4 6 8 10 12 14 i
i CYCLE OPERATING TIME (MONTH 5) , i h 1 Reactor Coolant Tritium Buildup - Zero Losses 1300 Mwt Reactor Figure 10.1-10 10.1-61
c.- 10'5 8 - 6 40 h00R PER WEEK MPC (10 CFR 20) 4 - c 5 - 3 2 - s a E E 10
= 8 E 6 -
'E 5 _
O 34 - 2 io-7 ; I I I I I i 0 5 10 15 20 25 30 35 REFUELING OPERATION Containment Air Tritium Activity During Refueling Operations
- Zero Losses 2 Loop Reactor Figure 10.1-11 10.1-63
Introduction of Fission Products Into the Reactor Coolant from Fissioning of Uranium in Core Cladding Recoil Range , (7-11 microns) Fission Fragment Escapes o
} O ,W 1 0
/
Zircaloy Cladding ' / Fission Fragments Do Not Escape Fuel < Figure 10.1-12 10.1-65 y
l l CHANGES IN FISSION PTIODUCTS ACTIVITY WITH CORE LIFE 100 90 - 80 - 70 - f x 60 -
- I E 50 -
f 40 -
/
p FUEL COD CLADDING ! 2 FAILURE AT 2200 / g 33 _ r.nvD / r.uU f 8 z I 20
/ I \
h NO DEFECTIVE / NO DEFECTIVE P FUEL RODS FUEL RODS N (FIFIST CORE) (AFTER REFUELING S o O A
$ 10 - @ 9 -
E 8' - 7 - 6 5 4 I I ' ' ' ' ^ v 0 1000 2000 3000 4000 5000 6c00 1900 BURNUP MWD / MTU T Figure 10.1-13 10.1-67
FIGURE 10.1-14 - CllE!!!STRY OF AN ION I:XCilASGl'. PROCI:SS Sodium-cation exchangers swap sodium ions for other metal ions
- f. .; :
7 7 f .. :7 / f.:; cy / f. : :/ / - w,o
-llCO - - !
3 / 2# #
-11C0 3 / -
/ if
/ -
-IICO 3 j Na3E Ca E 2 Na llCO Ca (It' 03) 2 - 3 Calcium bicarbonate Sodium-cation Calcium-cation Sodium bicarbonate exchanger exchanger 4 Regeneration restores capacity by replacing calcium ions with sodium E. .:.y /
A.y of / E6 _ h t?
-Ci
/ 3 -
h -Cl E OY # -Cl
/
l l C} f Ca E & Na2E CaCl o 2Na Cl 2 L Calcium-cation exchanger Sodium ch.oride Sodlum-cation exchanger Calcium chloride h Ilydrogen-cation exchangers swap hydrogen io.. for metal cations
/
fi ;:7 / D
- f. ._ :s / sii :s / fii 5.75 - 0 11 Co
/
-IICO 3 /
h M V / h
-IICO 3 / -
/ ,
/ -nll ,
CO 2 , Ca (liCO 2 32 211 0 2C0 2 Calcium bicarbonate flydrogen-cation exchanger Calcium-cation exch. 2 Water Carbon dioxide Cation exchanger turns sodium chloride into hydrochloric acid
& -f / , _ , 7, , ,_ _,y , L_ -f /
-Ci -L1 E E f.:;. / , f: if /
._:f _
-Cl
{} -Cl f g ya y f 211C1 2 Nacl Sodium chloride liydrogen-cation axchanger Sodiun-cation exchanger liydrochloric acid Then, regeneration with acid restores hydrogen exchange capacity
/
, s .. :7
/
4-(:;. ::/ / fii ./
/
+ f;: /
E SO E SO 0 ~ 4
) .
/ /
I
' '3 k' ash water eclie:ter o i Ion exch. g u unit Pressure bd3 water bEEhrial Back-wash Inlet "4 outlet N O ,=::- ,._., A go cc ,,
j::ll:.
. :R:! :::1 hh hk Backwash inlet Ej ec-Rinse u: :: ,
~;
outle' N~~ p "U
< -. )
Regenerant tank To waste j; Supporting bed Figure 10.1-15 Typical Ion Exchanger System e 10.1-71
RE :.ESHER COURSE - WELDING TECHNOLOGY AND CODES N'JCLEAR REGULATORY COMMISSION 25 - 29 October 1982 Course Schedule .
'enca:. 25 Cccober 5:30 - 10:00 Introduction & Code Review: ASME I Green 10:00 - 10:15 Break 10:15 - 11:45 Review of Welding Metallurgy Albright 11:45 - 1:15 Lunch 1:15 - 2:45 Frocedures for Welding Carbon &
Low Alloy Steels Howden 2:45 - 3:00 Break 3:00 4:30 Frocedures for Welding Stainless Steels McCauley Tuesday. 25 October E:^0 .an:00 Code Review: ASME III Green 10:00 - 10:15 Break 10:15 - 11:45 ;rocedures for Welding Punched
& Tempered Steels Howden 11:45 - 1:15 Lunch 1:15 - 2:45 Fracture Toughness Albright 2:45 - 3:00 Break 3:00 - 4:30 Residual Stresses Tsai Wednesday, 27 October E:30 4: 30 To Welding Engineering Labs for demonstrations and return to Fawcett Center (SEE DETAILED SCHEDULE)
AB
Tnursday, 28 October S:30 - 10:00 Code Review: ASME IX Green 10:00 - 10:15 Break 10:15 - 11:45 Corrosion of Materials McCauley 11:45 - 1:15 Lunch
- 1:15 - 2:45 helding Controls Richardson 2:45 - 3:00 Break 3:00 4: 30 Welding Processes Richardson Friday. 29 October 8:30 - 10:00 Code Review: AWS D1.1 Green 10:00 - 10:15 Break 10:15 - 11:45 Plastic Analysis and Design Tsai 11:45 - 1:15 Lunch 1:15 - 4:15 Examination 8
0 I _ _ - ~ _ . _ . . . . . _ _. _
Refresher Course - Welding Technology and Codes NUCLEA:. REGULATORY C0!"':SSION Lab Demonstrations - Wednesday, 27 October 1982 5:30 - 2;A5 Fawcett.. Center to Welding Eng. Labs Laser Weldinc Room 106 Diebold E::: c.:a--' Laser Heat Testing Room 106 Jones c., , ) ~~ Resistance Welding Room 111 Lee
~~~~
Pulsed GTAW Room 1381 Barone 4 10:15 - 10:30 Ereak Room 131 Puddle Vibrations Room 111 Renwick yn.:n
"' , ));):" Gutow TV Imaging Room 111
?notoelasticity Room 255 Mistry 11:)5 - 12:00 12:00 - 1:00 Lunch 1:00 - 1:45 U.T. of Welds Room 267 Fitting 1:45 - 2:30 Robot Room 100 Spence 2:30 - 2:45 Ereak Room 131 Metallography R m 270 Campbell 2-45 .,"
i Microjoining Room 264 Ling 3:30 4:15 Residual Stress Analysis Room Hwang 4:15 4: 30 Welding Eng. Labs to Fawcett Center
CODE REVIEW - ASME SECTION I (1974) Section Page Para. Subject Preamtie 1 Scope General Recuirements . 3 PG-2 Service Limitations 7 PG-ll.3 Standard Welded Parts 9 PG-25 Quality Factors For Steel Castings 10 PG-25.2.1.4,.5 Repairs by Welding - 4i" Wall or Less 10 PG-25.2.2.3 .6 Repairs by Welding - 41" Wall 27 PG-39.2 Welded Connections to Vessel Walls 28-30 PG-42 Welded Connections - Pipe, Valves, Flanges
& Fittings 53-54 PG-76 Cutting l 54 PG-78 Material Repairs 57-58 PG-105.4 Quality Control System Soilers Fabricated by Welding 63 PW-1.1 Scope 63 PW-1.2 Responsibility for Welding 63 PW-1.3 Welding Definitions 63 PW-5 Approved Materials 63-64 PW-9 Welding Joint Design 64 PW-10 Heat Treatment 65 PW-12 Joint Efficiency Factors 65 PW-15 Welded Connections 65-68 PW-16 Attachment Welds 68 PW-19 Welded 'n Stays 69 PW-27 Welding Processes 69-70 PW-28 Welding Qualifications & Records 70 PW-29 Base Metal Preparation 70 PW-31 Assembly 70 PW-3.? Alignment Tolerance 70-71 PW-35 Weld Finishing 71 PW-36 Misc. Welding Requirements 71-72 PW-38 Preheating
l i l 5ection Pace Para. Subject 72 74 PW-39 Postweld Heat Treatment 74 PW-40 Reoair of Welds 74-75 PW 41 Circumferential Joints in Pipes, Tubes & Headers 75 PW-41.5 Socket or Sleeve Type Joints 75 PW-42 Joints in Yalves and Appurtenances 76 PW-47 Check of Welding Procedure 76 PW-48 Check of Performance Qualifications 76 OW d9 Check of Heat Treatment 7E-81 PW-53 Test Plates Watertube Boilers 85-86 PWT-11 Tube Connections 86-87 PWT-12 Welding of Staybolting Box Type Headers Firetube Boilers 89 PFT-10 Welded Shell Joints i 90 PFT-11 Attachment of Heads & Tube Sheets 90-91 PFT-12.2 Attachment of Tubes 98-99 P FT-20 Attachment of Furnaces 99 PFT-21 Fireboxes & Waterlegs 100 PFT-25 Attachment of Stays & Staybolts 106 P FT-40 Welded Door Openings Minature Boilers 114 PMB-9 Welding Requirements Quality Control System 114 A-301 General 171-172 A-302 Outline of Quality Control System f
CODE REVIEW ANI: E 2',.7 The conditions of heating and cooling during post weld heat treatment
- lass 1 are restricted as follows:
() Above 600* F, the rate of heating is limited to 400* F per hour divided by the material thickness in inches-with a 400* F per hour maximum () During heating the furnace temperature cannot vary *more than 205 F within any 10 foot length ( At tne noidinc tem:erature, the maximum variation through-out the piping being treated is limited to 150* F () Above 600* F, the rate of cooling is limited to 500* F per hour divided by the material thickness in inches with a 500* F per hour maximum ASME I The external surface of a butt weld shall: Power Boilers () not have undercut in excess of 1/8" or 10% of the wall thickness whichever is less () have the reinforcement on a longitudinal joint removed if 1001 joint efficiency is used in the design () have a reinforcement not in excess of 5/32" on 3" thick base metal () have a minimum of 1/8" reinforcement if creep is con-sidered in the design stress ASME III Tne Standard Test recuirements may be used in lieu of the General Nuclear Test requirements for weld metal produced under specifications: , N5 () SFA - 5.1 () SFA - 5.3 () SFA - 5.5 () SFA - 5.7 AS'iE IX A welder, failing to meet the mechanical test requirements for one or Welding more test specimens, may be retested:
;ualifica-tions () by an immediate retest requiring two consecutive satis-factory test coupons for each position failed
! () after additional training or practice by a complete retest in all positions for which qualification is desired 1
2-(} by the radiographic inspection method only if the cualification is to be used on welds requiring full radiographic inspection () only with the concurrence of the Inspector AWS Recuirements of the AWS Structural Code Dl.1 include: 5;ructural Ccde Dl.1 () a restriction that low hydrogen covered arc welding electrodes not be dried more than twice () a limitation in the n0mber of submerged arc welding electrodes feeding the same molten pool to two () limitations on electrode diameters with shielded retal arc welding only () provisions for the use of copper, flux, glass tape
. or iron powder for backing O
l
CODE F.EVIEW - ANSI B31.7 (1969) Cnanter ' ace Para Subject
! 3 700 General Introduction 4 700.1 Scope 4 700.1.1 Definition of Nuclear Piping 4 700.1.2 ' Classification of Nuclear Piping 4 700.1.4 Jurisdictional Soundarfes
- 4-8 700.2 Definition of Terms I-!! 25-29 I-711.1 .5 Weld Types 25-29 I-711-718 Piping Joint Types 30 1-719.9 Code Springing I-III 40 I-724.1.7 Repair of Base Me'tal Defects 1970-1971 43 I-724.5.6 Repair of Base Metal Defects 1970 43 I-725.5 1971 I-V 45 I-727.2.1 . Welding Filler Metals 45 I-727.2.2 Backing Rings 45 06 I-727.3.1 Preparation for Butt Welds 47 I-727.3.2 Preparation for Fillet Welds 47 I-727.4.1 General Welding Procedures 47 I-727.4.2 Girth Butt Welds 1970-1971 47-49 I-727.4.3 Longitudinal Butt Welds 49 I-727.4.4 Fillet Welds 49 I-727.4.5 Seal Welds I-727.4.6 Branch Connections 1971 49-51 51 I-727.4.7 Attachment Welds 51 I-727.4.8 Heat Treatment I-727.4.9 Temporary Attachments 1970 I-727.5.1 Welding Qualifications 1970 51 51 I-727.5.3 Welding Responsibility I-727.5.7 Additional Welding Qualifications 1970 51 I-727.6 Records 1-727.7 Repair of Welds 1970 51
2-Chapter Face Para Subject 53 I-731 Heat Treatmert i 53 I-731.1 Heating & Cooling Methods 53 I-731.2 Preheating 53-54 I-731.3 Postweld Heat Treatment 55 I-731.4 Reheat Treatment 2-:: 62 2-711 Welded Joints 2-:II 64 2-725.5 Repair of Material Defects 1970 65 2-725.5 Welding & Brazing tiaterials I 2 'I 67 2-727 Welding Class II Piping 1970 1 3-II 72 3-711 Welded Joints 3-V 77 Welding Class III Piping 1971 i 4 [ [ I b
> n , _ . - .. , - , -
-- e ,--n , , , . - . ,
-.,,c-- - - - - - - - - - ,,-,-,,c- - - ----a ,--m-,m,---, _ - - - - - - - - , -r-,g-w,-w,
. . . _ _ _ - - - - _ _ _ _ - . _ - _ - - - - . _ - _ _ - _ _ - = _ . - .- . _ - . - .
CODE REVIEW - ASME III (1974) I 1 Section Pace Para Subject NA 4 NA-1220 Materials Restrictions on Subcontracting Welding 574, W75 17 NA-3130 NA-4450
- Control of Fabrication Proc, esses S75
^ 40 Nonmandatory Preheat W75 469-471 App D S74 NE 10 NS-2150 Material Identification Welding Material Tests $75 16 NS-2420 575, W75, 57f 17 NS-2430 Weld Metal Tests '
W76 18 NS-2440 Storage of Welding Materials 22 NS-2539 Material Repairs by Welding-Plate Base Metal Repairs by Welding W74 160 NS-4132 4 165 NB-4230 Fitting & Aligning Permitted Welding Processes 575 173 NS-4311 174 NB-4320 Welding Qualifications ) W74
! 174 NB-4330 Procedure Qualification W74 175 NS-4335 Impact Testing of Weld Metal 175 NB-4350 Qualifications for Tube to Tube l Sheet Welds 176 NB-4360 Qualifications for Special Seal Welds 178 NS-4380 Hard Surfacing 179 NG-4411 Identification, Storage & Handling Welding Materials 179 NB-4412 Cleanliness of Welding Surfaces 179 NB-4421 Backing Rings 179 NB-4422 Peening 179 NB-4423 Double Welded Jcints 179 NB-4424 Weld Surfaces 179 NB-4425 Components of Different Diameters 180 NB-4426 Weld Reinforcements 180 NB-4427 Fillet & Socket Welds 180 NB-4428 Seal Welds of Threaded Joints 180 NB-4429 Welding of Clad Parts
_ , _ . , -. --,.-..---_.-_--,_.4 - , _= . . , - _ . _ - - - . . - , _ _ _ - . - , .-.,_,m ,_,,
2-I Section Page Para Subject W74, S76 NE 180 NB 4430 Welding of Attachments Weld Metal Repairs S76 182 NB-4450 185 NB-4610 Welding Preheat 185 NS-4613 Interpass Temperatures 575, W75, 185 NS-4620 Postweld Heat Treatments S76 W76 18S NB-4623. PWHT Heating & Cooling Rates ISE NS-4624 PWHT Methods 1Eo NS-4630 Intermediate PWHT ' 185 NB-4640 Heat Treatment After Welding Repairs 191 NS-4643 Repair Welds to Cladding After PWHT NC (NB-2539.7) Repair of Cladding Base Metal Repairs by Welding W74 177 NC-4130 188 NC-4246 Atmospheric Storage Tank Special Joints 193 NC-4247 0-15 psi Storage Tank Special Joints l 197 NC-4250 Weld Joints in Vessels Designed to NC-3200 211 NC-4421 Backing Rings (NB-4429) Welding of Clad Parts (NB-4434) Welding Supports to Clad Components 576 215 NC-4437 Attachments Weld Repairs Without PWHT 576 231 NC-4642 (NB-4643) Repair of Cladding Af ter PWHT 1 W74 165 ND-4130 Base Metal Repairs by Welding ND 1 91 ND-4312 Production Test Plates 193 ND-4337 Alloy Welds in Base Metal j Qualifications for Special Seal Welds (NC-4360) 195 ND-4423.3 Plug Welds 199 ND-4460 Welded Test Plates 199 ND-4470 Welded Stayed Construction Base Metal Repairs by Welding NE (ND-4130) (ND-4312) Production Test Plates
3 Section Paae Para Subject NE (ND-4246) Atmospheric Sto age Tank Special Joints (ND-4247) 0-15 psi Storage Tanks Special Joints NE-4340 Performance Qualifications fer Clad Materials NE-4413 Procedure Qualification Tests W74. W76 NE-4423 Miscellaneous Welding Requirements NE-4429 Welding of Clad Parts - NE 4434 ' Welding Supports to Clad Comoonents NE-4454 Inserted Straps in Clad Materials
%~ 20-23 NF-2500 Examination and Repair of Materials 54 NF-4240 Requirements for Welded Joints l
l N3 22 NG-2539.6 Report of Repairs 68 NG-4132 Base Metal Repairs (NF-4240) Weided Joints 72 NG-4380 Qualification for Hard Surfacing 74 NG-4429 Welding of Clad Parts , (NE-4430) Welding of Attachments 80 NG-4640 Heat Treatment Af ter Repsirs 576 d O
. . - . . . . . . . - - . = _ _ - - _ _ - . _ _ . _ , . . . - _ _ . -. .- -..- -. . _. . _ _ . _ _
I PROCEDURES FOR WELDING CARBON & LOW ALLOY STEELS Types of Defects Weldment Thermal Cycles Transformantion of Austenite . Time-Temperature Transformation Curves Martensite Hydrogen 1 Hydrogen Induced Cracking , Weldability of Carbon Steels Low Alloy Steels - Types Hardenability i Weldability of Alloy Steels
! Carbon Equivalence
) Stress Relief I i f i ! 1 i . I I I } i
1 PROCEDURES FOR WELDING OUENCHED AND TEM?ERED STEELS Need for Hardenability Properties of 0 & T Steels
- Thermal Cycles and Metallurgical Effects !
Low hydrogen Procedures Control of Properties Control of Hydrogen Filler Metals Stress Relief Microalloyed Steels i i 4 1 b I e
. , . , . _ _ . . _ . . _ _ - . _ _ _ _ _ . , - . s _, _, ,_ . _-,4- . . -_.- ._, ,-_ , , ., ._ ...
y l l l PROCEDURES FOR WELDING CARBON & LOW ALLOY STEELS Types of Defects Weldment Themal Cycles Transfomantion of Austenite , Time-Temperature Transformation Curves Martensite Hydrogen Hydrogen Induced Cracking Weldability of Carbon Steels l Low Alloy Steels - Types Hardenability Weldability of Alloy Steels Carbon Equivalence , Stress Relief i i e I l l i T k
i l
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FRACTURE & FRACTURE TOUGHNESS TESTING I. CU:T!LE RU:TURE OF METALS A. Cross Material Deformation : E. Nucleation at Inclusions C. State of Stress on Fracture Appearance
!!. : ...u rRACTURE OF METALS A. Cleavage E. Fatigue C. Intergranular III. MATERIAL RESTRAINT !
A. Notched Tensile Bar B. Lateral Contraction . C. Plane Stress & Plane Strain l , IV. STRESS CONCENTRATION A. Gesnetrics i E. Crack Tip i V. CHARPY IMPACT TESTING A. Test Configuration ,
- B. Temperature Effects C. Crack Initiation - Crack Propagation
, C. Other Impact Tests j t j VI. Kg- FRACTURE TOUGHNESS
- A. Stress Concentration at Crack Tip B. Calculation of K C. K - Critical Value For Crack Propagation c
. D. Speciman Thickness
~
VI!. FRACTURE TOUGHNESS TESTING A. K Testing Methods B. Speciman Geometry i
\.II. CRACK OPENING DISPLACEMENT (COD) 4
..e 14 u i t, v2 at=n2 i
- . Pure Iror.
A. Ti=e-Te=perature Cooling Curve
- 1. Solidification 1535 C ( 2795 F) I Delta Iron - BCC
- 2. Allotropic Transfor=ation 1390 C ( 2535 F)
Camma Iron - FCC
- 3. Allotropic Transformation 910 C ( 1670 F)
Alpha Iron - BCC o Curie Temperature 768 C ( 1414 F) II . Iron-Iron ' Carbide Phase Diagram * '
; A. Inte r=e tallic Co= pound i
- 1. Iron Carbide - Cementite - Fe)C 6.67 % Carbon Orthogonal Lattice '
) High Tensile Strength - Low Ductility B. Solid Solutions
- 1. Austenite 2.0% Carbon Maximum at 1130 C ( 2066 F) i FCC Lattice High Ductility l
- 2. Ferrite i
Delta Ferrite - 0.10% Carbon Eazimum at 1492 C ( 2718 o F( 1 Alpha Ferrite - 0.0255 Carbon Maximum at 723 C ( 1333 P BCC Lattice High Ductility C. Eutectic Reaction - 1130 C ( 2066 7)
- 1. Ledeburite Austenite + Cementite Mixture
- 2. Eut5ctic Co position - 4 3% Carbon D. Eutectoid Reaction - 723 C ( 1333 F)
- 1. Fearlite Ferrite + Ce=entite i
{ 2. Eutectoid Co= position - 0.8% Carbon E. Crain Size Changes
- 1. Reduction (Refinement)
Transformation to Austenite i Critical Temperatures : ! 2. Growth Ti=e + Temperature Above Critical Temperatures (i i l _ . . . . . . _ . _ . .. I e
Hent Treating
- 1. Critical Tooperatures 9 A. Equilibrium Critical Temperatures l
- 1. Lower Critical - A t
- 2. Upper Critical - A) and A ca B. Nonequilibrium Critical Temperatures
- 1. Heating Lower Critical - A e
. Upper Critical - A and A e e3 cm
- 2. Cooling .
Lower Critical - A rg Upper Critical - A 7 and A II. Carbon Steel Classes r* A. Hypoeutectoid Steels Approx. 0 75% Carbon Eaximum E. Eutectoid Steels Approx. 0 755- 0.85% Carbon C. Hypercutectoid Steels Approx. 0.85% Carbon Minimum III. Heat Treating Procedures - A. Full Annealing - Maxi =um Ductility + Kinimum Hardness
- 1. Heat to 100 F above A) for Hypoeutectoid Steels Heat to 100 F above A g for Hypercutectoid Steels
- 2. Seat one hour per inch of thickness
- 3. Cool very, very slowly B. Nor=alizing - Refine Crain Size & I= prove Toughness
- 1. Heat to 150 F above A) for Hypoeutectoid Steels Heat to 150 F above A ca for Hypereutectoid Steels
- 2. Seat one hour per inch of thickness
- 3. Cool in 'still air' C. Hardening (Quenching) - Increase Tensile Strength
- 1. Heat to 100 F above A) for Hypoeutectoid Steels Heat to 100 F above A g for Hyparcutoctoid Steels
- 2. Sc:k one hour par inch of thicknees
- 3. Cool rapidly in water or oil D. Te= paring (Drawing) - Increase Ductility e.nd TouGhne ss
- 1. Heat 'Mardened' Steel to te=parature below lover critical
- 2. Hold for celected tice
- 3. Cool in air
Heat Treating (Continued) IY. Time -7caperature-Transformation Diag::ams A. Isothermal Transformations I 1. Ferrite or Cementite Pro-eutectoid Phase s
- 2. Pearlite Coarse Medium Fine
~
- 3. Sainite Feathery -
Acicular
- 4. Y.r.r+a nsite E. Continuous Cooling fransforn tions
- 1. Full Armaaling Propertie s
- 2. Morealized Properties
- 3. Harde ne d Prope rtie s
. 4. Austempered Properties
- 5. Eartempered Properties C. Transfor=ation Rate Factors
- 1. Co= position Carbon Content
' Alloy Content
- 2. Crain Site ,
- 3. Austenitization Temperature D. Hardenability
- 1. Jo:iny End Quench .
- 2. Critical Dia=s +ar l
l C
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3000 2795 _ _ __._ _ _ _ _ __ _ _ __ ___ ___ _ _ _ _ B.C.C. DELTA IRON
- 2414 .
2500 - F.C.C. GAVy.A IRON 2000 _ 1670 _ _ _ _ _ _ _ _ _ _ _ B.C.C. ALPHA IRC: c P^, r_.n. L. u,. h r ,. I c .
- 1500 -
1414 --_---
.- _. CURIE __ - - TEKFE_RATURE t ,
i i 1000 - B.C.C. FERROMAGhTTIC ALPHA IRON I f 500 - e 4 TIME
- y. - , , _ ~ - - . - , x . - - _ - , .- -_._ --., ,-. ,r_ , . - . - , - - , _ . - . - . - -
g-i IRON-IRON CARBIDE ; EOUILIBRIUM DIAGRAM GAS 2900l - 2ss (
- 2 3 2) 2800 -
3400 y _f.D 37 ( / (3339) g 3200 ^ l 1700 -
' / - 3000
-[gg7 _ c u0uio. delit re LIOUID , / ,
au ! f j 2800 ! t2sN: ! i}N $j
- eWi *, i,'" n LIOUC].
6400 ,Lt? - - 2600 ruoui:ss . re3C l
!300 - 'so.tous 2400 4 U LIOU!D + feJ v2eD 's u. ! e g ::,w,,,,, J27 ,
l 12% "AUSTErs i tTE AUSTENITE " ~~ ~ ,j 12240: , 4 W ( G:mm: Fe) . , ,c e , ~r,~ W
, g" itOO Fece Centerec ' 2 n *4 (20ss; -
2000 C D Cuce Q 1O')O 5:2 1800 ke t AUSTENITE + Fe3C i W g r LFE) W i' 1 1 F E R r::TE - 1600 2 2 800 J -A3 - W - y :, AUST E titTE
& Y7c -
l400 W j 700 727 T')' ' ' ' EC 77% IAs " 3d " f I-oc2:es - l200 ; i 600 - FERRITE + CEMENTITE 3 403 _ ( Alphc Fe + Fe3 C) - 800
' Body Centered Cubic i
300 - 600 . i FERRITE CEMENTITE 1 l . 200 - ( Alphc Fe ) ( Fe:C ) - 400 100 - 200
! O "'''''''''""'"I"'""**"""I'"
O I 2 3 4 5 6 6.67 100% re I O*4 Fe3C 1 C A R BON - PERCENT i t v 1
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% CMSON COMPOSTTON Fig.34.-Halmos of an,asessiso as a f,.csy of wy m
.-em e
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Transformation at J3L0 'T 14Ca'-
.l ites..n rature 1.nds
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12t $ 1 ' ,g g , c;;- I earnte formmt . 32
, from susternte Pearlae
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57
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- g. .lly temperature . g, Martensite . g, ,
1 2 4 5 15 30 1 2 4 S 15 30 1 2 4 5 15 Seconds . .T!tnute , Houn 7tme of Transformation Fug. 31 -The , a e d,agram for she doiomposition of ausse iso i. a esseaoJ sarbon sserl s
*n m-r b *ar m- - ,
j aar - Mg { fan
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( 200 = 4 g A2 a.4 ed Ao LO L2 7 carbon Fic. 7 17. Variation of Af, and AffTem-gratura ' eth Carbon Conieni of Nin-Carbon Siceh. (Tempratura are approu-maic only.)
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o- o o co 200 soo soo too 600 700 L-@ :..g Tu.~ .^.:.ra (* C) T see 2so ses *:s son cm 7:s T 21e 390 $1e 730. 930 tii8 1290 f; L13 - K.vq c.rotrA.3CQf~" W M =? d
Common Heat Treatments mi Hot Rolled Alloy Steel Bars and Plates lower critical (A,) to produce the I#2 TIM Ie*#2T/v/F
- desired combination of tirenFth and h
6 .hk ductihty (tempenng); and then cochng y, J in air to room temperature. Primary J ._ purpose: enhance and optimize CM*tohef 5W W mechanical properties. Tempenn; at
'l 41 relatively low temperatures will n a nimrze hardness and wear
## M #NI resistance. Tempenng at relatively high temperatures will masimize toughness and ductihty.
- Temer /ure Tee hme (
Annealing - Heaung to a temperature Normalizing - Heating to approx. ^J above the upper enucal (A,) or to a imately 100 F (55 C) above the .. point within the entical temperature upper critical temperature (A,), Md- - . . - - range, then furnace o>ohng through holding long enough to insure a 1 the entical range at a controlied, uniform temperature, and then coohng specified rate. Pnmary purposc soften to room temperature in still air. the steel, making it more workable for Primaty purpose: ref ne tbe hev/ SW C00' subsequent forming or machining nonuniform internal structure of hot "P*'**'"* rolled steel produced by the variations in hot finishing temperature 7g u ;fyff Tane encountered dunng rolling or forging. [ , The normahzed structure will usually Thermal Stress Relieeing - Heating to 43 eshibit more uniform mechanical a temperature below the lower critical properties, better ductility, and higher ( A,1, holding for a predetermined time, resistanec to impact icading. and then slow cooling to room
.... .. M *..... . .
A/ temperature.' Primary purpose reduce Tempera /vre residual stresses set up as a result of n 31 , fabricating operations such as cold __,,,,_,5 4 working, machining, and welding.The Nes/ Cool _ __ pg treatment restores mechanical properties (particularly ductility) and
- g f.....
i helps prevent distortion. Tot Source R. Joe Kasten, senior product Spheroidize Anneating - Prolonged ggf specialist, Western Steel Div., Armco heating at an apptcpriate temperature g,,f g,gf Inc., Houston. (February 1980.) near the lower critical ( A ,), followed by slow cooling to produce a
=
microstructure containing globuln carbides. Primary purpose produce a I*f i structure which may be desirable for Quenching and Tempering - Heating l machining. cold forming, or cold at a temperature above the upper l dra w ing. critical ( A,) until the structure transforms to austenite (austenitizing), quenching the steel to room l temperature in oil or water to produce a martensitic structure (hardening), reheating to a temperature below the
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t STRENGTH OF STEEL cMSoN o ebb ! e ADB flu.0YING EdfsfM6 l ANNEAL
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GU6NcI4 de TEMPER l 1
i ! CMSON . EQUIVAlGNT I e C + El
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TihE - SECONDS p666666 iiiiiii; .e i s i . . i ... : . . ii i . 6 e i. ii = 4.0 Ni-0.6 C 1
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1 l In special In spectat 1r Dissolved Combined nonmetallic intermetallac elementa!
! Element in femte an carbane inclusions compounca sta te l Nic ael N ... .... Ni.5s compound ( ?) ....
, Ni. sal Silicon Si .... SiOs.M,0, . . . . ....
Aiuminum A! .... Al:On. etc A!,N, ....
, Zarcoruum Zr .... ,ZrO: Ir,N, ...
i 1 Manganese M@Ma
- Mr.S .... ....
, , MnO.SiO:
Csronuum Cr *--- . Cr Cr 0, . . .t
- Tunesiet % *--+-W .... .... ..
- - Mosvboenum Mo -+-Me .... .... ....
- Vanadium V
- Y V,0, V,N,
- Ltanive T.
- Ti Ti,0, j Ti,N,C, ..
l , Ti,N, j Columbaum CbMCb .... .... .... i Phosphorus P .... .... .... .... Sullur 5( F) .... ' (MnTe) 5 .... .... Zr$ Cepper C. .... '.... .... Cu when >
, about 0.8%
1.444 .... .... .... .... Pb (?) 4 i I i 1 l
!' A. Decrease haroenabilits -
factors that hasten' nucleation: i 1. Fane grams of austerute . 2. Undamolved ina.lusions
- a. Carbides (or nitrides)
- b. Nonmetallic inclusions
- 3. Inhomogeneity of austenite B. Increase hardenabihty -
l factors that retard nucleation or growth of Ar products from
. nuclen
- l. Dissobed elernents in austenite (except cobalt)
- 2. Ccarse grains of austenite
- 3. Homogeneity of austenite Consolidating these items without respect to the direction of the inSuence, we have 6ve principal (but interrelated) factors: ,
i
- 1. Mean composition of the austenste
- 2. Homoceneity of the austenite
- 3. Grain site of the austenite
- 4. Nonmeta!be inciusions in the austenite
- 5. Undissolved carbides (and nitrides).in the austenite 4
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, 1 Af cah fm A Acicular Ferrite.
)L Rose Line Toughness e I
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S P I'***""* 'I i f 4 Segregation of Martensitie- Potential blocky latti Constituents to R*Pl acement of Martensite solidification and
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(a) AS WELDED (b) STRESS RELIEVED
Nb TRIE 1200
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I f I t i 3 10 10 10 10' 1 TIME (SECS.) The ef f ect of niobium on the C.C.T. diagram of weld metal with a baseline composition of 0.1%C, 1.35: Ma.
HEAT TREATMENT CRACKING
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i testraint i weld shape rnatorial composition I i i { i i I 1 i P 1
I l l I POTENTIAL WELDING PROBLEMS Problens wisch can arise dunng wekkng of steels may be dnnded into two categories: 1 Problenes of wwold metal or heet effected The first category covers: , mone cracking directly attriletable to the weld metal hot cracking
% 2 5 76 Process. heat affected zone burning and hot tearing lamellar teenng heat affected zone hydrogen induced cracking 2 Proidems wwhich nomy ocesse et wwelded The second categwy:
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Y QL, Mak Corrosion Enginorring 32 Table 3-2 Gele,anic Series of Some Commernal Mesala and pys in Seewater Platinum
- Gold Noble or Graphire cathodic Titanium Silver
'Chlorimes 3 (62 Ni, is Ct.1e Mo) '
'.Hastellor C (62 Ni,17 Cs,15 Mo)
'1s.8 Mo seminless aseel (paastve) 1s.s stainless sisel (passive)
, Chromium stainless steel 1130s Cr (passive)
'Inconel (passive) (so Ni,13 Cr,7 Fe)
, Nickel (passive)
Silver solder
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Cupronickels (60 90 Cu,4010 Ni) Broones (Cu.Sa) COPPER
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'Incomel (actige)
, Nickel (active)
TiIn Lead Lend. tin solders a
'IS.s Mo stainless steel (active)
,16.s stainless steel (s'ctive)
Ni Resisi (hish Ni cast iron) Chromium siainless steel,13s Ct (active)
' Cast iron
,$icel or ima 4 2024 aluminum (4.5 Cu,1.5 Mg,0.6 Mn) ,
Active or Cadmium anodic Commercially pure aluminum (1100) Zinc Msgriesium and magnesium alloys In general, the positions of metah and alloys in the galvanic se closely with their constituent elements in the emf series. Passivit gahanic corrosion behavior. Note in Table 3 2 the more noble positi by the stainless steels in the passive state as compared with the of these materials when in the setise condition. Similar b Inconel, which can be considered as a stainless nicke! Another interesting (cature of the galvanic series is the brackets shout , Table 3 2. The alloys grouped in these brackets are somewhat sim
)
CLA99IGT6lNLE99 GTEEL9 BA20WJM-MATENSOC-RARDMBLE Type 440 A
. "T0r(.70 C) .
Type 440B 17 Gr (.85C) Type 4400 - 440F 17Crt.0C) 17Cr(1%Se , Type 420 , Type 920F 13 cr(350) '3tr(350)ts L I Type 410 Type 416 Type 416Se M2Cr 12Cr+S 3Cr+ hse 1 i~,x-w.-:=- ,1,e .= ~--.ag_,_ i Type 403 Type 405 410 eb 12Cr+ AL . 12Cr+03 l 12CrTQ 1 ( ' 2-2Type 414 a 414Ti 12-2Ti
,, Type 431 '
16-2
Hardonability Calculated Frcm Comp sitism By blarcus A. Grossmann w w ma Ao /dv A / atl,Oryp'ot?/ 04 , im I / l l l l/ ca j j G* s'"*yL l l/
- t
+
y- ix s- w -
+me~'=o q l ,Vii -
cm
>l ) J.,ml s &- 4tm I ; l
, .i l
cx 9'
~. i I l / / )/
Ekto'Unnow },%. E " ~l " r/Y/ ,j0 k HEF46% H Ei is . , ll I 1/7 HU "O {MygB % L l l : l J i w (ic-$' /A( W/ 0,. 3 / ///Vi .
- l. v/ , ,,I,,,,in -,nem I i ;
I,
/// : \
/
s 8 . CN
- i. l , 70 !
l L' ' Ctt -
// .J ! __ w wh a&9e l ' ell ',.N I i I
/MA%.NWsr,a 6 , i \ .
W~,\%pss, %l ArQwti Wus#:n
////+=se=s; I l l l N !he bv' l 0 18
,V i OM -
i t O C2 004 OM OM 00 02 0 14 02 02 GM 0?? QN s
; Coctit, % .
g,, CN 04 060 QA: 4 400 0 1 s o L'D M 2D
. l . [ " l 1 l ! l il l[f((((((M M'y' >
l
!1 l i ,
T Vf//f M/// V/f f///// M -
-j sx , ! .
ix I ,, i I i I i /fdMZ//r%& q exic ' y zh ex% ' ggg., ,o 3* \ l l \ i VM/////M/4 q
;;~- o, l
VM///4%YMMk e
' & ; *sb,w,6) Os I l V//H/f///////////JRE
'VMAVH F f f f f ;'
d@ i A, h ! mr o, 0? .. A$ - g? m.o m es p T7
~ : s i
o - O O# 00 IS IS SO E# SO l l T67,7B f
'/ M i Esample: To compute hardenabil6ty of a commercial steet of the analysis shown:
in , - , .- lesu A mow w? Fac,oa Point
,,, Ideal hardesability of given steel = D, =
fy , tgAg 3syf l IU fy s),97anr4 mag)"" 014 u 4 = 1.1 x 1.05 x 0 98 x 1.1 x I.7 x l.16 a l A2 = , , IM r 2.40. Such figures for D, will occurately coan. Cam wtest om ggg fm. GJe 4.00 8
/N pare one steel with another. useassene M 000 22 0 26U Jealey egulvalent of the given steel at sutsee 4.10 s.le C O Phos,bame 8.620 9.85 D point K that is, hardness of center of 6deelly h a 14 31 54 7 fg fgp fu f quenemed 2 40-tr. bar. is the same as that P,"*' f
. b 5 spee n. ,[ . " , , ,
y Mro) Eict,tt'n ./S/>y. Jet.at hard.e co,,e, o.es J e
'g E W O g .t ce nie r, is b en from lowest curve, is about C,45 for -. i j,,, *Ealamated oa the Dae for alcheL Que lag pe e H e commercial batbs may be deterrained by methods shown l ,JT g, la the data sheet. McTat Paocatas, Oct.1941. page 520. or estimated from LM table gg 4 g y
st right, below. 45 0 i l l "e , J
)j lit f f Q
dn , taa m pie: w ier agitated; H a quench, mddir I CS: Steen D, == 2 40. c,,, n,,,,,
- o. Ao,. vie. %
ws oe a , w ,,, g,,,, gl< Actual crit 6 cal diameter D (from Q' g 2 po6nt af abose D, == 2 40) is 16 in h*",' 8 ' I g ,, gy (( l f f pt y3 , 3, I g , gg Q cg staa of bar of given steel that will harden in gives quench to half mar. Wedernie Good 443 ta s.40 tJto J 6.4 to gJ 3.4 to IJ l Leasttu strueture et center; hardness Sirent SJ to GJ I J te 3.0 . . . . ( 0* vi*et oJ
- t. 4 s C a/0 040 ca7 030 10 '57 c45.
b ) O m 9 9 u O \ s M U *
% U O i
i
~ 44 '
c
- O
.M -
v .s 4 O (\ c1 ~ 9 u 9N e@ g\ 't N tm A \k -
- C C *O k ' h'N'E Ns e, . % .
. .3 e
) . u ~4
- b. g %
- L g 's *
, O, u -- om g4 .. ..
'g c
, a
\
- h. N- - h w
f.) 3% 9 N O N%\ %
- 4.\ eg y W% " . .-. 23
--O 0
his s 8"-t , %:._ \ .i.(i
" '4 ,0 e ==m u.
9 s. n' 3 % N 1 %
% s ---.N. J J -4
-Q \ ,: g Q'n o CN 2 s g ,f O = =
- O \. % 3 ~\ ** t e am-0 d
\\ s,*
e *;A N ' 3 s .:7 - 4 t U .J .C f . s. -* JO M f./3 d
\* '\ j/
ti-4o
-= )
W O k4
.of W *t = c.
C
.O
/
u=
'Ie J *-4 *. **
-4 /+ E ~. ~
"J J
. { g \
d, N l s d C OCN u OO *
' i it
- 5 0 O bl R - S i3 u
O 3
- o ) . t 1
' -f' * ' . . , . . 1 3 COG f, s")1.L tf T/ C */o 'O
.4 4
4
- 4) /3 a
U O
**4 *
=E O 'J 5 O yQ
-.4 W '".e c U O c.: cc c,e
=
3 h0-r- 4
% f3 %,
e
-- tc 6 :r ~3 g = E J g ; i t :9 , :f:
U
)
1 3
.ppW/ ..
u
- d C O 2 4 . . ', r 9N
~ \ ,
/ .
s n N -
.e
\i s** 9C %. ~O
.c
. pu be m M *H U
/-/ g -
oe1S W
*J U "O 4~ c On i L7 <
**
- T c nN v *W, w . -O
< E (
- Q UAC e
g l o O 9 g W $c*3 unsm 9 ,ft oc 4 D ~ q 46 --e . . . a
~
o a tr. ~O
= u 4 *-=
& y ' 4 y -J 4 *O
% k O +
C C-k-
" $ =
- ' 9, * - w .1 lC CLe
- p g4
- . !- M c
,T 2, -
% g 6 .u .aco. . -4..
- 4. G 3N U . ~ 6 .C GO
- q uu .O O a 8.
l} e3 3 e 1 % l I Iu - 4UU
. o o % O = D'
- > ~' O M
0* % O
*d 5 , Y 3
,__.,,a n ,- - ---
55Ec)De x>iseeesc s p *A C r e=(fe s' c) Pwes 4 5 000 L. i Q u ( C) ff + L F4L+A l S.Foo 4 A 19 L nw , A+C
/S** F+A*C g47L
(
.- i
- c se 1
12%Cg o.'s t.o I.s
% C ,
..~n. -. .
INPUT TIME TH IC'K NES S -_ tt it.o w a.e s Frog FOR c Ap nou PER tNc H I472 f ro A"* A '* Y
*131 F sTeet.s NOMOGRAPil FOR . 14 J I I tac H es C41.C U L47 sM &-
20.-. seco Aj os _ i. Ceot lM c- TIME **** 3 [ (,tyJ2 cj3gp} I~If " ' S**
. - A 3 1.
gSO- ~U## '
~4 llo tol4 tc t4 LAPST.5 (2C- - ITT
~ ~3** -
-p so H E h! THe S 14As~E. coets _
150 t QT'
~
se, I" A v r e lt XRC 1
~ # - h -
WEl D /A/6 ? y,
- ree . . t, o
- - So ~
. So ~ ~
'4 * -. ._I
&c- ~ 30 g4 9C -
sa- - lo
.t f Ac l NST - Lo . . gg ~
~q 40 - . gy . .
T-3 r .-- ro " '
. . 1 4r G sco 30 - .
q, g-4 - ? .LL a s .- ar - - 3 - -? g4
- 3e -- 4 I
no - -- 2
- 2.C A FTER IntA G Akt- SEKr&uca 15 -- _20 C8' 8 E T G EA D otv Pt. ATE W 6 L. D YEE .J Q t M T
#** ' O , 3 S- Lf " THt c k A.!F.s .5 CARBOM - C . I 'l */e 6
y4 rc - S M AW - S AW GMAw ( Ar) o G- M A W (,,, Co t o g)
' voo -
coo, l ' a R E t iq rio u scrweewRAMGE l are . Tl M G ( IL4'12,- 9 3 2. F-k AWD M AX tM vm t-(A R vA26 3cc IN THG H EAT AVPECTED ] T
- are -
)
.te o -
c hW P ~T Su ce No Ca AL o.eq a.si a.ozy 041 f.sy ,c2Y.aty e.g.t ..as, i tro l ' # si , e ie
' f ! t '
.2 0 Ll o 90 Hoo too .
l .'2 'i f= 8 10 t, a loo 400 doo S B cgAt, COOLING T I N t G. f:ACM /47AP Tc 932 5 /At
I Weidmt Characteristics / M.17 dwA W w Pr,n w new g mme m rums.. l,
- i
{ .I . - ---------Al - { .- f. _ - _ - - _ _ - - A; E j k i E
- ,i r
.2
--_- ----_-____% _- _____ -_---__-M Lee im WC 0.
Prener Prarwt --lee we i Ca4 bete % Dut het Cood Deme M
#wtner PwG. nas anetteve 5 . -3 J p-I -__---------Al 3 -_. _ _ _ _ __ _ _ _ A; L L E E 1
DN Tm (cend Qt? f b thM Ag ClltCSI le46. latlOA IMI f - oc w enn.~ % % **rtemk For setan leae tanout 8GF)
%Q, -
,n e,,,
' - Ierrne, Cam
-J ----- ---__-___-AI =A tes amo P.*rtemh i
E ++- 1empersc Mortemte I
--.. - _ - _ _ _ _ _ _ _ - at hrw l ,
- (
y.; u * . ? *.::. * *
. ren N est ter%
- duiG 5tructure ef .4 9 1
fi.v =ng M e .. .-
'. s,...
d iO
i CLh29 H STIINLE29 CTEEL9 CW10W1W;?WC-NoMMMi Type 446
- 2.7 0 r .;
I-g Typc 442. m i 2A Cr Type 430 i Typet30F
" " 17 Cr+ 9 r y Type 43OF S 17 Cr ,
17 Er+P + St 3 Type15 Gr 429 g i Type 434 f 17 8t+ 1Mo Type A36 .
. ll 17Cr+1 Moi Cb t
1h. 430Ti a 17 Cr+Ti
!., 4-30 Gb l; 17Br&Bb E
409 11 cr+ Ti
CLA29 M STAINLE99 STEEL 9 CHROMlVM-110Ka- AUSTENm0-NONRARDENABLE
, TypcSDZB l
18-S Si Type 302 i Type 304 Type 304 L ga gg.S LC " 1B-8 El.C, - 18-8 f8 Typc 321 18-8 Ti Type 347 Type 348 347 FSe f 18400b+Ta 1B-10 Cb 18-10C19Ta.tSe f Typc303 m Type 303Se(
!; IB-B+S 18+P+ Se j l
l Type.305 i Type 384 Type.2B5 C P 18 -12 P 16-18 12-15 L fL Type 301 jf17-7 i Type 316 Type 316 L f TB-l2 Mo 1B-12 MoEl.C. , i Type 317 31 j9 1913-3Mo Ne713-3 L MeRt l 1 1 Typt308
? 20-10 l
S hPN09
=3 gpE _]g
.lj Type S10 Type.310S ! Type 314 2 #
f 25-20 2.5-20 LC, 25-20 Si hcg;;.m.g g Type.201 g Type 20Z 18-5-9+N 17-4-7+ M l
CLA 5 5 TTT STA tNL fsS S TKE/ S (ADD) CH R om i um - Ni CKE L- Md^/6AMET[ ~ M ITK OG E A v57~EN f r / C. ~ NON H A R PEN A Bl F f 1 T1PE .R0/ Y PE OR i AO3 mem ,
/7-4~7 ?N / E - 6~- 9 + n' I S -f- T +h'+.S l
2 0.6~
\ s 17-/ -/C+H : ::-i id '~~
; .2. /(o
,a l
l 8-6-91-Wf M C 7 . 4
' l li / S-2 N 5
4 18-2-/2. + N 1 -, { / 8-3-/34 NJS-3 4 . l [I 2/-6-9 [1 l !2/-6-9 + # I 1 22 -/3 -S*
- f22-/3,5+N -
l l . 4- C 6 + V T - fa 22-4-1 2 4 -V- 9 4H1 C ' (... I 20-45-5 20 - VS-S'+ W
+ mo+ch .
' s
, i ; l
! l
' i ;
28 28 - , THE MAUREE DI AG 11M
!\
24 l\\~ .
- 24
.) I
- i p .
! - 20 I -
i . . .a . i r 20 .
- i '
1 ul I V -- 16
' m i= = i=
o 16 - - o . J W *****Dit' - 12 _ ~ ~ ' - - - - ^ ~ ~ ~ " * - I 12 s E ferrite Z austenite
. ~8 0_ - . martensite martensite f:k.ne yearli sustenite &martensite &, ,
_4 4 f erri te ' Pearlite a teMte fernte i 0 I l l l l l 1 20 24 28 32 8 12 16 0 4 ' CEROMIDM CONTEET - % . l
-r -, - ~ .- , , - - , - - - - - - , , ~ - - v -, ,- e - . - , - w - - -
g -
Constitutiona Dingrana for Stainless Steel Weld Medal si.. . ,a.... i.. s.i ..g,7 7 ,,- 30 ~
- -- / /
/
28 .. / 310 ~
--- d - - - - -/ ~--
26 - - - - - - i' / /
- c. ~
N' 2 + A u s tenite -
' - pt --
22
- / -
- 50 20 x
gg g -
' /- e
.A+M
/ / 3' X e3Sg\Y/ '
5
$6 N N f 7
,g _ _ _ _ . _ ,
y w eMI 7 /
- /
+f e '
' /
W /~},/f E - . _ - K
%go /
0 Mar fen sile / 1(ef. A A+ M+F [ f.~._ /s x g
. ._ f - . /
/ \ 1
-- .- - - - - A' - - - ).. .. . _. _ ._
f\ 4 ._. ._.
/ Ferrite p
1+ l y 1 \ g. { 305 -2lSO 4-42 0 M 2
\SLI 4 6 8 IO I2 I4 I6 I8 20 22 24 26 28 30 32 34 36 38 40 Ghromium Equiv0 lent */,, Cr + % MO + 1.5 x % Si + 05 x % Gb multiplied by the " potency factnr" indicated for fi% ferrite; magnetic analynis of the nrtual E s a u rs.r. : l'oint X on the stiegram indientes the element in question along the nues of the pample revealed an average ferrite content of 2%.
the equivalent enmposition of a Type 3RR l'or mustenite -plys- ferrite st ructu res, the 1316 Ch) weld deposit containing 0.07% cartmn, diagram, in order t< determine the chromium dingram predicts the percentasre ferrite within I.M% manganese, 0.57% silicon, 1H.02% chro- equivalent and the nickel equivalent. When these 4% for the following stainless atecle 30R, 309, mium, 11.87% nickel. 2.16% molybdenum and were plotted, as point X, the constitutinn of the 309Ch, 310, 3I2, 316, 317 and 3lR ('tir.Cb). $, 3 4*7 wel<l was indicated n= austenite plus from 0 to O RO% calumbium. Each of these percentages was
,1letal l*rogress llata Shers; Naurmher 1919; l' age blio-It i:1:s mton]
1 0 .
/ i f\/
/
/
2i I WRC I
/'
/ // / / i' M
$ ,o
. 5
+
1, ,
//
/,o/ / / / /// / /
/ -
/ / / / /,x 5' ,,
/ 'Y /7/ / / b / /
/ ////
/ / / /
n 17
,/
/ /$ / / / / /
+
1 5 '
/ / / //fn.l ///
to
/j /,/,/,/j/f//
\ n
- ea'oa MAGNET lC l // / j j 7 7 7 J bj g
+ ,, FERCENT
,lg y ,s l
$u 'cie
/ / / /fflf
/op,l l/l ,/
f ll A. h4 ,/ / /
/ /
- E b 13
*/ -
i .._
/ l -
J-4
.b
[i[ p' / e / ,f
/
,N
=>
z- jy ' / -
. AUSTENITE a Q / / ./
$9f
,' 7 '/ PLUS FERRITE C N / / / / / / .
@9%9 m
jj j/ / / / / /.
/ / ,f//
ls'/ /l/!//l ! U_ 2' / / '
$ ~
ll if 26 27 10 l 22 23 I 24 25 i E - 16 17 18 W 20 21 m CHROMlUM EQUIYALENT = %Cr + */.Mo + }.5 X */.51 + 0.5 X %Cb.,-i.. i $ 3. o.. .., su. .i. .. ora , ! o 6,.- . . ... .. . . v n. ..id sir i. m... p5. . -.ie -.. 6......i....-.
.n... i .i.. ,.. ,i.i . .
. 4 , . s.
c.., ... .i m in. s n..m ! ; c.i..i. n. .i.i.
...ir.i.. n ...ir.i. . n. -.is ..'. s. ..
- t. A .i.i. .g.i..i... . so x v.H 6.
6... 4 4 4. 4. i s. s.m ..= -
.i.4 4 m . i ..e-.
.a...=. . . ....i
? K
....i .ti.....
...... o.n v. i osa ..a ... 4 .ia'. 4.. 3. r. i.. H.-6-. a t ... 4 .ini .4.. ..
j m
...n.6i..
o ne y, s. oma -.e4 . er en. .>.-s.*r i. . .. . en. 16. *ies i .son
.it.r seet so,. si..aa . 4 s i r i.-ia 6... .
- n. w. 6 -ini , t.. i.. . ... .i n j
z . .,..- . i... n. wee r .....r.,... s u. i n. >i,..e
. ... : . a u.6- tw iai. i. . .
-d .. ........ ... ,,,v . . - . ........u_.....
i e d3 Z 28 - I I I I O.10 - l I I I I l l A L 26 - Austenite 0 09 - Austenite O - (*'/ ' - p 24 -
' 0 08 t*
i + - <j@ e,,
~~) 22 -
0.07 - '
< g .
A+F O goSs. - 0.06 h4 o x 20 i.
$ h6 19 21 235 26 29 30 2D 5O %Cr "<<"'
- i6 -
crg A+M ,o W y 14 i
- i> x s2 --
p o"' -
.2 d iO l
*+ Martensite soo % " ",,.
~
4*
~
A+M+F N o 6 - i o4 F+M ~ m Martensite + perrite i
+
.__ 2 / Ferrite -
- z i i i i i i i i i i i i 30 i i i 34 36 i
38 i 40 20 22 24 26 28 32 [ O 2 4 6 8 10 12 14 16 18 Chromium Equivalent i ! % C r+% M o + 1.5 x % Si + 0.5 x % N a + 2 x %W+2.5x%Ti +3x% Al i i l
- _ - - _ _ . _ - . . - - . - - ~ _ _ . - - - - . - - _- _ _ ,-. . - - . . _ .
g :- .
<A 2*- 6* B 2. . e*
s s* ( 2** -. ~ T: s., ( 2:. r. ,
.e P. 4;.
60 $*. '8
- 8 a'8 /, 43, - 6; *, Srec .t
( .c,
'f t . ". ~ .
s
'}.
f.( 2: . j, - - -
% s: ~ . j, s ,j.. n, . ..: -
f ! ! -
%,s ' ' .. 5 . u '.,
3 $ 6: d .c -l '
-4~ 't . _ _ $ s =, .. '.j'..j
. .. . > W. * .u%
} v . . . ' . y.- A
/* ' ',
,,./"
-. s-v
, , ' , ,Y
/ ,/
(s e
.e
' , f ,"
s
\
N
'\ m. '/
'q ,
j r. -
~*
. ' I" e' 'N -* -
I c ~ // , . . -
- + x- . - ,.; . __. . h /
- .. -- - o-- p IC 907- 4 s e
.;:- r s Ng.- w np u -- ..- ..~ , y .... . , . - , s l
te ': 2: n 4; 1: 6: 7 e; 90 M Fe i; Z: s: 4: 5 4" 6* M - 6e r s e :e :;e to:.e .se r Len e**:a *. :**
;g _~e, , . , - - y .~ .r,,5. , u.or.o r, . ste:r ' a-hqu.aa surrare. E souc - rar:st e i
1 a.4tenite w it - territe it ans d 5- inic gamless steet welomen:s w hich tem (Fe N tc. ine tainar. e.ne : u
- , 1. suled aiong tne cellular oencritiC esneD
- t Cuples mg ec' structures. (Ment at J'*Cr a3% Ne a .l!c :.:e -
4
.g cou ncarJe> Honeser as ahos compo- 2 To deselop a rapic so. dihcation Fig 1 A Ahos c omoc-.iion. n ing o-
,g. n.m. are evienee m ternte torming tecnnique capable ot preserung soke- the Cr rich sice of tm> i.ne w a uacm e* men- o c-pieted m austenste- ihration substructures at room 1em- predommanth a* BCC oct:. w4 l, ,, . :.n,zm;; e.er ents celta territe re- perat ure while those on the N" c' w << :r -
.,.,, 2. re- as 'eno- as in.- primats phase 3. To irwestigate tne murosegrega- kne will sohd t turcu . .an* a ' U l n.,w m.- :. apor. Lpon coohng to , tion accompanung w.hdihcation bs ausienste i t .- w.., eture trom ine ohdnica- means o1 microprobe anaiws Tne sobr:u- sur: ace' an > .e*
[ * * , a .; . % ie rete s coa umec bs sen:eg: m a .must ma .ner ,. N-
+ -. .- a e - > u men torm, ir the cen- traieci m Fic 1E Tne o'herit - .c :N
- c. .-a-%r e . unta utiimateh rinh a Theory deiia te. roe anc .u ien te icy.. . - .e i
seN o.a netwon of sernte temaint seca ated bi Inree-r a e ...an e . 1 The Iron Chromium Whei ht rawn1 hne regirsn. V'w..1 r a.onc Ine iterme- cenorste Lore 5 ittnars $ssiem
. .ti,r:s e.a:e., .
- i. ::it entormatio ,
e- nean rurvec woortu- ime- a- - d . a.:a bie son.eming eitn- the me- The Fe-Cr N ternary smem pro- conse ge to a poin' at the t r' n a - a,n.c o :ne ohoo. cation proces o- udes tne ba ts ior precirting phase eutectic temperature o 1 & s- (. i S*. N nature et ine >ohc-state terreie- enuihtrium in aust enit u sta miess O The er.aumun -ena anon b a ce' 4
. stence transiormation sierA w e most c ommerc.a! stainless the hem tu and o! r:u urt e- n %
-l eve- 4:e c ompo ed of 1N5 wt % iron-ris h -pr.e a c t he s .~ ~ -
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1 COMTRUCTIOA E Knife line corrosion in stainless steel structures
.i V Cibal' and R A Jarmant J.1 Stab.laec austerunc staanless steels are regularly used in welced of locahmed intergranular cursuwon hes sta it.c ba.se n. eta:
1 construcuora to comoat the problem of antergranular corrcaion meetuely adyscent to the weld deposit so netames occurr C icommonly referred to as weld cecayi, an insidious form of m weldmenu of the niobium - or, more prid-e Q. corrosac attack that can occur m the weisment beat anected titaruurn-stabilued stamlcm steels. e .2 zone that nas czoenenced thermal evelmg within a temperature There are at lea; tw. difiering phenomena which may acco P rance ofSoc4 o*C.3 Trie steelin such condition u referred to as for the susceptibilary to corrosive attack at the fusion ime g: Deme 'sensiused'. causec bv the precipitanon of chrortuum stabihsed stainless steel and the present purpose is to atter carbde at tne austenue grain bouncanes which lowen the a clanscauon. Tne locanon of the environment senuine ar h* ; enectne' cnrorruum content and corrosion resistance near the relauve to the weld metal is illustrated m Fig. :.
-,4 gra:nbouncanet Sens2tssation may be prevented bv stab:hsation
- of the steel which is achieved by the addauon of titan 2 urn or
'f racc.um, recogrdsed as forrrung innocuous carbices, leanng no Knife-line attack associated with chromium iree caroon m the steel to combine wit.n enromium to procuee carbides (Type I)
+:1e :. arm?.G oc unca. v cnromium carbice. However. it is pesnoie ' Th:s is recueec corrosion resistance that occurs m zone
- wh.
to enco nter otner corrouon probiems assoc 4atec with weicec curme weline. have expenenece a double heat misuen:e Lencanans : stabihsed stairdess steel. Important among these . Luring tne 6rst thermal evcic, temperatures are reached t is a sensiuvnv to intergranular corrosion in a zone immeciatelv enable dassolution of the stabie carbides tic and NbC. 7
. a: acent to tne weid at its boundary with the parent plate and area then cools quickly throuch the temperature rance of 7
*" . ccmmoniv referrec to as 'inife-line' attaci. This seems to re- or NbC precipitation and, after coolmg to room temperati
.l preser:t a certain type of corrosion often unrelated to the sus- this small recion is high in carbon with the titanium or n ob.
,. cesubilirv of aussentuc steels to corrosion as a resuh of the pre- in solid solution in the austerute.8 If this narrow mone is ti captanon of enromium carbide. reheated into the range 3oo-6 3o*C. the kmenes of chroma s
; Cme cf the carbes: reporu of this mode of corrosion was re- carnice precipitation will be much faster than those of T C
; i corcec ano cescribed in i95t1 but, subsequently, cases of knife. NbC precipitauori, and the reF on i becomes sensitised" aim.
l hne attack have been iscovered more and more regularly in a to the reacuon that occurs in unstabilised steels' althourn
; number of environments of which nitnc acid appears to be possibihry of tne presence and inftuence of other phases, i predominant. Suen corrosion cevelops very rapidly and leads, sigma phase. has been considered.
,.' in man? cases, to premature scrapping of expensive apparatus. The sensitisation of this narrow zone may be brought at, bv one of a number of processes: foi the weldment may be p+
, _! a stress rehef anneal in ne sermitsstion temperature rance.
- 1 A cottfusion of terrns
! Tr.e hierature contains a vanery of terms to describe inter. a muitipass weid may be mace on thick plaie and the potentu
{. eranda currouon of stabilmed steels ir. a localised region susceptibic zone created bv one pass may be sensmsec b-4 arnmeci.tei3 (.acent to the weid deposit. In a Weldmg Research suosequent pass' or (t) sinese pan weldmc of thick p: ate n
*. ! Coun:i1 Report the tem weld hne attack' is referred to as res. alt in a slow cool throuch the sensitisation rance suficien:
, l aneetmg a relatively narrow tone separated from the weld cause susceptibihty. Regardless of the mode of sensiusanen.
pI ceposit bv a small increment of parent metal which has been zone is now liable to interfranular corrosion for reasons sa . soluticn treatec by the heat of weldinf.' It is pointed out that to those associated with weld decay. After relauvely lon this should not be confused with knife Ime attack where the zone penods, possibly at higher teroperatures, carbides of stabilising elements are also precipitated.' This form of fus Ime corrosion susceptibiliev should be amenable to reversion
} *C l' Avnn 5:nar AtuanA Assuurfer Nauna s p,.1,am, Pragw.
- tbrp sf Mr.urgy & Maunals, Cap of Lendae p )=chnic, solution treatment at so5o*C in which the chromiurn carbid are s o' m te.t.ne eno o 12ones o/intercrys. re<iiuohed in the austenite to be followed by the reformat 4,
,W rarkne and we.Ane cf stabihser carbide, thereby fixinE the carbon.
p corresson. l s
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d 2 Titanium cerbiae Knife-line attack associated with stabiliser carb:
, \A 4 preciporate structures g3W, gg f# X e)in the of reason superheere,nt wei,,
Tne principal conclusion that is drawTi from the precec e t hdy / F#0s.SNi.Tisteel er. ser double beer rieer. section is that knife.line attack is caused by the precipitation chromium carbide. However, it is regularly oberved that
'"'"Ol "' '8"i"* titanium stabilised steels are more troublesome than those c r"' '^ M ,[i nt t t $ taining niobium. This is dif5 cult to reconcile with a the C I con rephces x 3r00. based on grain boundary depletion of chro.nium, althour i, C, m $ '-
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