ML052590053

Kansas State University, Minor Revision Safety Analysis Report (Ch 4, 6, 7, 13) Dated 6 July 2005

ML052590053
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Site:	Kansas State University
Issue date:	07/06/2005
From:	Kansas State University
To:	Office of Nuclear Reactor Regulation
Witt K, NRC/NRR/DRIP/RNRP, 415-4075
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10 CFR 2.390(d)(1)
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KANSAS STATE UNIVERSITY TRIGA MARK II NUCLEAR REACTOR FACILITY LICENSE NO. R-88 DOCKET NO. 50-188 MINOR REVISION SAFETY ANALYSIS REPORT (CH 4,6,7, 13)

DATED 6 JULY 2005 REDACTED VERSION*

IN ACCORDANCE WITH 10 CFR 2.390(d)(1)

• Redacted text and figures blacked out or denoted by brackets

Kansas State University Minor Revision SAR (Ch 4, 6, 7, 13)

Kansas State University TRIGA Mark II Nuclear Reactor Facility License R-88 Docket 50-188 6 July 2005 Department of Mechanical and Nuclear Engineering K-State Nuclear Reactor Facility Kansas State Univcmsity 10 Ward Hall 302 Rathbonc Hall Manhattan, KS 66506 Manhattan, KS 66506

Minor and editorial changes to SAR Chapter 4 Changes:

Label added in Table 4.3, Summary ofReactor SCRAMs linking trip information to the footnote that period and fuel temperature trips are not required Missing constant incorporated in equation 8 Temperature and CHF information associated with the new power level incorporated in section 4.8.3

• Chapter 6 Changes: 500 kW changed to 1,250 kW

• Chapter 7 Changes:Editorial ("continuous energized" changed to "continuously energized")

• Chapter 13 Changes:

Table 13.4, 1,259 kW changed to 1,250 kW Section 13.2.2.a, 500 kW changed to 1,250 kW

4. Reactor Description 4.1 Summary Description The Kansas Statc University (KSU).Nuclear Reactor Facility, operated by the'Department of Mechanical and Nuclear Engineering,'is.located in Ward Hall on the campus in Manhattan, Kansas. The Department is also the home of the Tate Neutron Aciivation Analysis Laboratory.

The TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Regulatory Commission License R-88 and the regulations of Chapter 1, Title 10, Code of Federal Regulations. Chartered functions of the Nuclear Reactor Facility are to serve as: I) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United -States, 3) a facility for training nuclear reactor operators, and 4j a demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems.

The KSU TRIGA reactor is a water-moderated, water-cooled thermal reacto perated in an open pool and fueled with heterogeneous elements consisting of nominally E percent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:

• Central thimble

• Rotary specimen rack

• Thermalizing column with bulk shielding tank

• Thermal column with removable door

• Beam ports

• Radial (2)

• Piercing (fast neutron) (1)

• Tangential (thermal neutron) (I)

The reactor was licensed in 1962 to operate at a stcady-state thermal power of 100 kilowatts (kV). The reactor has been licensed since i968 to operate at a steady-state thermal power of 250 kWV and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximumn of 1,250 kW, with fuel loading to support 500 kW steady state thermal power with pulsing to $3.00 reactivity insertion. All cooling is by natural convection. The 250-kW core consists of 81 fuel elements typically (at least Jplanned for the 1,250-kW core), each containing as much astgrams of 23 U. The reactor core is in the form of a right circular cylinder about 23 cm (approxirnately 9 in.) radius and 38 cm (14.96 in.) depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.) diameter and 6.25 m (20.5 ft.) depth. Criticality is controlled and shutdown margin assured-by three control rods in the form of aluminum or stainless-steel clad boron carbide or borated graphite. A fourth control rod would be used for 1,250-kW operation. A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank.

The tank and shield are in a 4078-rn (144,000 ft. ) confinement building made of reinforced K-State Reactor 4-1 Original (5105)

Safety Analysis Report

CHAPTER 4 concrete and structural steel, with composite sheathing and aluminum siding. Sectional views of the reactor arc shown in Figures 4.1 and 4.2.

Criticality was first achieved on October 16, 1962 at 8:25 p.m. In 1968 pulsing capability was added and the maximum steady-state operating power was increased from 100 kW to 250 kW.

The original alumninum-clad ffuel elements were replaced with stainless-steel clad elements in 1973. Coolant system was replaced (and upgraded in 2000), the reactor operating console was replaced, and the control roam was enlarged and modernized in 1993, with support from the U.S.

Department of Energy. All neutronic instrumentation was replaced in 1994.

1/4 22 1 Figure 4. 1, Vcrtical Section Through the KSU TRIGA Reactor.

K-State Reactor 4-2 Original (5/05)

Safety Analysis Report II

REACTOR DESCRIPTION 4.2 Reactor Core The General Atomics TRIGA reactor design began in 1956. The original design goal was a completely and inherently safe reactor. Complete safety me -ns that all the available excess reactivity of the reactor can be instantaneously introduced without causing an accident. inherent safety means that an increase in the temperature of the fuel immediately and automatically results in decreased reactivity through a prompt negative temperature coefficient. Tbese features were accomplished by using enriched uranium fuel in a zirconium hydride matrix.

lIT, &AM .

. *. raePo X a ^%

outoo -&. t ao>CITbynA;Kug>W

~~~* *P Int*= * ** -' Frl ESn UMAIWA T- .* 4 -W M

.0, ,CA CO'C3Hm WFtA i. X East Figure 4.2, Horizontal Section Through the KSU TRIGA Reactor.

The basic parameter providing the TRIGA system with a large safety factor in steady state and transient operations is a prompt negative temperature coefficient, relatively' constant with temperature (-0.01% Ak/kc)Q. This coefficient is iafunction of the fuel composition and core geometry. As power and temperature iincrease, matrix changes cause a shift inxthe neutron energy spectrum in the fuel to higher energies. The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the power increase. Therefore, fuel and clad temperature automatically limit operation of the reactor.

K-State Reactor 4-3 Original (5105)

Safety Analysis Report

CHAPTER 4 It is more convenient to set a power level limit that is based on temperature. The design bases analysis indicates that operation at up to 1900 kW (with an 83 element core and 120VF inlet water temperature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatures which could cause loss of clad integrity could not occur. An 85-element core distributes the power over a larger volume of heat generating elements, and therefore results in a U more favorable, more conservative, thermal hydraulic response.

4.2.1 Reactor Fuel' TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature U would automatically cause the power excursion to terminate before any core damage resulted.

Zirconium hydride was found to possess-a basic mechanism to produce the desired characteristic. U Additional advantages were that ZrH has a high heat capacity, results in relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size. N TRIGA fuel is designed to assure that fuel and cladding can withstand all credible environmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state'and pulse;mode operation.

The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced' in the fuel element clad materiaL The maximum temperature limits of I 150'C (with clad < 5000C) and 950 0C (with clad > 500SC) for U-ZrH (HIZrAs) have been set to limit internal fuel cladding stresses that might challenge clad integrity (NUREG 1282). These limits are the principal design bases for the safety analysis.

a. Dimensions and Physical Properties.

The KSU TRIGA reactor is fueled by stainless steel clad Mark Ill fuel-elements. Thre' instrumented aluminum-clad Mark 11 elements are still available for use in the core.

General properties of TRIGA fuel are listed in Table'4.1. The Mark Ill elements are illustrated in Figure 43. To facilitate hydriding in the Mk 111 elements, a zirconium rod is inserted through a 0.635 cm. (114-in.) hole drilled through the center of the active fuel section.

Instrunented elements have three chromel-alumel thermocouples embedded 'to about 0.762 cm (0.3 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (I in.) above and below the center plane. Thermocouple readout wires '

pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor tank.

'Unless otherwise indicated, fuel properties are taken from the General Atomics report of Simnad [1980) and from authorities cited by Simnad.

K-State Reactor 4-4 Original (5105)

Safety Analysis Report

REACTOR DESCRIPTION Graphite dummy elements may be used to fill grid positions in the'core. The dummy ileie'nts are of the same general dimensions and construction as the -fuel-moderator elements. They are clad in alumniniu an-d have a graphite length of 55.88 cm (22 in.).

Table 4.1, Nominal Properties of Mark II and Mark III TRIGA Fuel Elements In use at the KSSU Nuclear Reactor Facility.

Dimnensions Outside diameter, D. = 2rO

• IA7 in. (3.7338 cm) IA7 in. (3.7338 cm)

Inside diameter, D, - 2r1

• IAI in (3.6322 cm) 1A3 in. (3.6322 cm)

Overall length 28A in. (72.136 cm) 28A in. (72.136 cm)

Length of fuel zone, L 14 in. (35.56 cm) 15 in. (38.10 cm)

Length of graphite axial reflectors 4 in. (10.16 cm) 3.44 in (8.738 cm)

End fixtures and cladding aluminum 304 stainless steel Cladding thickness 0.030 in. (0.0762 cm) 0.020 in. (0.0508 cm)

Burnable poisons Sm wafers None Uraniumcontent Weight percent U 23sU enrichment percent 235 U content Physicalpropertiesoffuel excluding cladding HlZr atomic ratio if 1.0 m

1.6 Thermal conductivity (W cm~l K'l) 0.18 0.18 Heat capacity [T 2Ž0C] (J cmn K') 2.04 + 0.00417T Mechantcalpropertiesofdeltaphase U-ZrHJ Elastic modulus at 200 C 9.1 x lo pSi Elastic modulus at 650 0C 6.0 x 106 psi Ultimate tensile strength (to 650 0 C) 24,000 psi Compressive strength (200 C) 60,000 psi Compressive yield (201C) 35.000 psi

'Source: Texas SAR [1991].

b. Composition and Phase Properties The Mark III TRIGA fuel element in use at Kansas State University contains nominally WU/o by weight of uranium, enriched to I/oh '35U, as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is.gQverned by the composition .and temperature. For ZrH16, the equilibrium hydrogen pressure is one atmosphere at about 760'C. The single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen. Over 25,000 pulses have been performed with the TRIGA fuel elements at General Atomic, with fuel temperatures reaching peaks of about 11501C.

K-StatWReactor 4-5

• Original (5/05)

Safety Analysis Report

` J~-

CHAPTER 4 The zirconium-hydrogen system,.whose phase diagram is illustrated in Chapter 3, is essentially a simple.eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and facc-centered tctragonal hydride phases. The two phase delta + epsilon region exists between ZrHz,. 6 and ZrHj,74 at room temperature, and closes at ZrH,.7 at 4551C. From 4551C to about 1050IC, the delta phase is supported by a broadening range of H/Zr ratios.

STAINLESS STEEL

@I cbO ACF

-STAINLESS 20 MI G2T~

STAINLESS GAHT: 3.44W STEEL BOTTOM END-fIXTURE Figure 43, TRIGA Fuel Element.

K-State Reactor 4-6 Original (5105)

Safety Analysis Report

REACTOR DESCRIPTION

c. Core Layout A typical layoui for a KSU TRIGA Ii 250-kW core (Core 11-18) is illustrated in Figure 4A. The layout for the 1,250-kW core is expected to be similar; except that the graphite elements will be replaced by fuel elements, one additional control rod will be added, and control rod positions will be adjusted..

The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with tbc additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is S2.00, the 1,250 kW core pulse rod reactivity worth is $3.00), balancing the remaining control rod's worth to meet minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge K-State Reactor 4-7 I Original (5105)

Safety Analysis Report

CHAPTER 4 4.2.2 Control Rods The pulse rod is 3.175 cm. (1.25 in.) diameter. Other rods are 2.225 cm (7/8 in.) diameter.

Control rods are 50.8 cm. (20 in.) long boron carbide or borated graphite, clad with a 0.0762 cm.

(30-mil) aluminum sheath.

The control rod drives arc connected to the control rod clutches through three extension shafts. J The clutch and upper extension shaft for standard rods extend through an assembly designed with J slots that provides a hydraulic cushion (or buffer) for the rod during a scram, and also limits the bottom position of the control rods so that they do not impact the bottom of the control rod guide "JI tube (in the core). The buffers for two standard rods arc shown in the left hand picture below J (slotted tubes on the right hand side) along with the top section of the pulse/transient rod extension. The pulse rod drive clutch connects to a solid extension shaft through a pneumatic

. Q1 i

cylinder, the dimensions of the cylinder limits bottom travel.

I,-

Is_

upper ruise, Shim & Keg Kods Keg KOU Shim Rod rulse Kod Figure 4.5, Control Rod Upper Extension Assemblies The bottom of the pulse rod is shown on the left hand side of Figure 4.5. The upper extension shaft is a hollow tube, the middle extension is solid. The upper extension shaft is connected to the middle extension shaft with iock wire or a pin and lock wire for standard rods, with a bolted collar for the pulse rod (the mechanical shock during a pulse requires a more sturdy fastener).

Securing the upper control rod extension to the middle extension at one of several holes drilled in

• the upper part of the middle extension (Figure 4.6) provides adjustment for the control rods %1-I necessary to ensure the control rod fill in position is above the bottom of the guide tube.

K-State Reactor 4-8 Original (5105)

Safety Analysis Report

REACTOR DESCRIPTION Figure 4.6, Middle Extension Rod Align'ment Holes The middle solid extension is similarly connected to the lower extension. The lower extension is hollow, the middle extension fits into the lower extension and a bole drilled in the overlap secures the lower extension to the middle extension. Typically the lower extension has a tighter fit than the upper extension liecause the lower and middle extension are not separated for inspections and because the interface with upper extension is used to set the bottom position of the control rod.

Pictures of the lower connector for the pulse rod and one standard rod are shown at the left in Figure 4.7..

Figure 4.7, Standard & Pulse Rod Lower Coupling The boitom of the lower extension attaches directly to the-control rod. Pictures of the control rods taken during the 2003 control rod inspection arc in Figure 4.8. The rods move within control rod guide tubes, shown in Figure 4.9. The guide tubes have perforated walls. The guide tubes have a small metal wire in the tip that fits into the lower grid plate; a setscrew inside the bottom of the guide tube pushes the wire against the lower grid plate to secure the guide tube.

K-State Reactor 9

• Original (5105)

Safety Analysis Report

CHAPTER 4 Pulse Rod *ShimRod Reg Rod Figure 4.8, Control Rods During 2003 Inspection

. * .I F1ull Wuidc Tube Upper Lower Lower Detail

• Position in Upper Grid Plate Figure 4.9, Control Rod Guide Tubes K-State Reactor 4-10 Original (5105)

Safety Analysis Report

REACTOR DESCRIPTION

a. Control Function While three control rods were adequate to meet Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one transientfpulsing control rod). The control-rod drives arc mounfed on a bridge at the top of the reactor tank. The control rod drives are coupled to the control rod through a connecting rod assembly that includes. a clutch. Tle standard -rod- clutch is an electromagnet; the transient rod clutch is an air-operated.shuttle. Scrais cause the clutch to release by de-energizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control, while scrams operation at limiting safety system setiings. A detailed description of the control-rod system is piovided'in Chapter 7; a summary of interlocks-and scrams is provided below in Table 4.2 and 4.3. Note that (I) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting setpoint depends on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch.

Table 4.2, Summay orControl Rod Interlocks' qr.. 4I~tOOi42; c"POTliNDQnURPOSEBO r Inhibit standard rod motion it nuclear Source Interlock 2 cps instrurent startup channel reading is Less than Snstruotent sensiekvityensure nuclear instrurnent startup channel is operating Pulse Rod Interlock Pulse rted Prevent applying power to pulse rod unless rod insertedprevent inadvertent pulse Withdraw signal, Prevent withdrawal of more than I rod/Lirnit Multiple Rod Withdrawal more than 1 rod maximum reactivity addition rate Pulse Mode Interlock Mode switch in Hi Prevent withdrawing standard control rods in Pulse pulse mode Pulse.Power Intelock

______ __Interlock___Ji 10 km_YPrevent 10 kW pulsing if power level is greater than NOTE. Pulse Power Interlock normally set at I kW

b. Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. .Te TRIGA system does not rely on speed of control as significant -for safety-of the reactor, scram times for the rods 'are measured periodically to monitor potential degradation of-the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafe excursions and the control system is-used only for the planned shutdown of the reactor and to control the power level in steady state operation.

K-State Reactor 4-11 Original (5/05)

Safety Analysis Report

CHAPTER 4 Table 4.3, Summary of Reactor SCRAMs .

Linear Channel High 110% NIA 104%

Power Power Channel 110% N/A 104%

High power DetectorHigh 90% 90% 90%

voltage __ _ _ _ _

600 0C B Ring element High Fuel 5550 C C Ring element 4500 C Temperature"? 480 0C D Ring element 380°C E Rin element350 0C Periodi I /A I NIA 3 sec NOTE [13: Period trip and temperature trip are not required The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for 1,250 kW operations for eight hours per day (including xenon override) and will assure a shutdown margin of at least S0.50.

Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per minute, with a total travel about 15 in. (38.1 cm). Maximum rate of reactivity change for standard control rods is specified in Technical Specifications.

I.-

4.2.3 Neutron Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator. Demincralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specimen rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3.4 in. in length, acting as top and bottom reflectors.

The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially. The reflector is 0.457-m (18.7 in.) inside diameter, 1.066-m (42 in.) outside diameter, and 0.559-m (20 in.) height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3.18-cm (1.25 in.) inside diameter and 27.4-cm (10.8 in.) beight. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector.

K-State Reactor 4-12 Original (5/05)

Safety Analysis Report

REACTOR DESCRIPTION' The radial'reflectoi assembly rests on an aluminum platform at the bottom of the reactor tank.

Four lugs are provided for lifting the assembly; A radial void about 6 inches (15.24 cm) in diameter is located in the reflector such that it aligns with the radial piercing beam port (NE beam port). The rifledtor supports the core grid plates, with grid plate positions set by alignment fixtures. Graphite inserts within the fuel cladding provide additional reflection. 'Inserts are placed at both ends of the fuel meat, pioviding top and bottom reflection.

4.2.4 Neutron Startup Source *.

A 2-curie americium-beryllium startup source (approximately 2 x 106 n s') is used for rea&6r startup. ..The source material is encapsulated in stainless steel and is housed in an aluminum-cylinder source holder of approximately the same dimensions as a fuel element. The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for ude in relocating the source manually from the 22-ft level (bridge level) of the reactor.

4.2.5 Core Support Structure

.The fuel elements are spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates.

The grid plates have a total of 91 spaces, up t6C*3of which arc filled with fuel-moderator elements and dummy elements, and the remaining spaces with contiol rods, the central thimble, the pneumatic transfer tube, the neutron source holdero and one 'or more voids. The bottom grid plate, which supports the wveight of the fuel elements, has holes for receiving the lower end fixtures.. Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 special holes in it. The 1.5-in. (3.8 cm) diameter holes in the upper grid plate serve t6 space the fuel elements and to allow withdrawal of the elements from the core.

Triangular-shaped spacers on the upper end fixtures allow the coolingF water to pass through the upper grid plate when the fuel elements are in position. The reflector assembly supports both grid plates.

4.3 Reactor Tank The KS)SU RIGA reactor core support structure rests on' the'base of a continuous, cylindrical aluminum tank surrounded by a reinforced, stanaard concrete strii6tuie {with a minimum thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 anid 4 2. The tank is a welded aluminum structure with 0.635 cm. (1/4-in.) thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-fl)in depth. The exterior of the tank was coated with bitumiinous material prior to pouring concrete to ietard corrosion. Each experiment facility penetration in the tank wall -(desiribed below') has a water collection pilenum at the penetration. All collection plenums are ionneclid io a kik-off volume'through individual lines with isolation'valves, with the leak-off volurnies monitored by a pressure gauge.' The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual leak-off voluim; (and :prcssure gauge) is'installed for the bulk shield tank; all other plenums drain to a corimion volume. In the event of a leak from the pool K-State Reactor *4-13 Original (5105)

Safety Analysis Report

CHAPTER 4 through an experiment facility, pressure in the volume will increase; isolating individual lines allows identification of the specific facility with the leak.

A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation. The bridge is mounted directly over the core area, and spans the tank. Aluminum grating with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating can be lowered when activities could cause objects or material to fall into the reactor pool. The grating normally remains up to reduce humidity at electro-mechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations.

Four beam tubes run from tlie reactor wall to the outside of the concrete biological shield in the outward direction. Tubes welded to the inside of the wall run toward the reactor core. Three of the tubes (NW, SW, and SE) end at the radial reflector. The NE beam tube penetrates the radial reflector, extending to -the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described iii Chapter 10).

4.4 Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield.

The shielding configuration is similar to those at other TRIGA facilities operating at power levels up to I MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cmi. (3 ft.)at the top of the tank.

The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phcnomena.

4.5 Nuclear Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fuel temperature) for TRIGA reactors. This coefficient is a function of the fuel composition, core geometry, and temperature. For fuels witht* UW]

enrichment, the value is nearly constant at 0.01% Ak/lc per IC, and varies only wely dependent on geometry and temperature.

Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures that the fuel and clad temperature limits will not be exce ded. The. design bases analysis indicates that operation at 1,250 kW thermal power with anW1ement across a broad range of core and coolant inlet temperatures with natural convective flow will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss of clad integrity.

Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety K-State Reactor 4-14 Original (5/05)

Safety Analysis Report

REACTOR DESCRIPTION limits; and fission-product release rates. Thermal hydraulic peiformnance is addressed in'Sectiofi 4.6.

4.5.1 Design Criteria - Reference Core The limiting core configuration for this analysis is a compact core defined ly the TRIGA MkII grid plates (Section 4.2.5). The grid plates have a total of*Jspaces, up toWf which are filled with fuel-modeiator eleinents and graphite dummy elements, and the remaining spaces.with coujirol rods; the central thirible, thepneumatic transfer tube, the neutron source holder, and one or more voids in the E or F (outermost two rings) as required t'osupport experiment operati6ns or limit excess reactivity. The bottom grid plate, which supports the weight of the fuel elemrents, has holes for receiving the lower end fixtures.

4.5.2 Reactor Core Physics Parameters The limiting core configuration differs from the configuration prior to.upgrade only i' the addition of 'a fourth control rod, taking the'place of a graphite dummy element 'or void experimental position. For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-rMy dose rates.

For comparison purposes, a tabulation of total rod worth for each control element from the K-State reactor from a recent rod worth measurement is provided with the values from the Ccmell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Relaied to the Renewal ofthe Operating license for the Cornell University TRIGA Research Reactor).

Table 4.4, 250 kW Core Parameters.

l (effective delayed neutron fraction) 0.007 I (effective neutron lifetime) l 43 :S

-SO.017 EC' aTf (prompt temperature coefficient) 250kW -275EC CEv (void cocfricient) .- 0.003 lemo void

-S0.006 kWl'.to-ap (power temperturecoefficient-igbtedive) So.o I Table 4.5, Comparison or Control Rod Worths.

KSU TRIGA.11 (250 kW) Cornell University (500 kW).

C-3, Shim $2.88 D-16, Shim $2.20 D-16, Relating S1.58 L D-4, Safe . . $1.99 D-10 Pulse $1.96 D-l0,Transient Sl.88

-TBD TBD E-l, Regulating SO.58 K-State Reactor 4-15 Original (5105)

Safety Analysis Report

CHAPTER 4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions. A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature.

4.53 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures.with realistic modeling of the fuel-cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures.

Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRIGA fuel and cladding temperatures associated with pulsing operations. The analysis addresses the case of a fuel element at an average temperature of 10000 C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of tinie after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 4701C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hrl fW 2OF' (2550 W rni 2 K') is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected. This section provides an independent assessment of the expected fuel and cladding thermal conditions associated with both steady-state and pulse-mode operations.

a. Heat Transfer Models I The overall heat transfer coefficient relating heat flux at the surface of the cladding to the d difference between the maximum fuel (centerline) temperature and the coolant J temperature can be calculated as the sum of the temperature changes through each clcment from the ccntcrline of the fuel rod to the water coolant, where the subscripts for each of the AT's represent changes between bulk water temperature and cladding outer xi surface, (bro), changes between cladding outer surface and cladding inner surface (roro, J cladding inner surface and fuel outer surface - gap (g), and the fuel outer surface to centerline (ricl):

Td = T, + AT, +AT, + AT +AT,; (l)

A standard beat resistance model for this system is: J

• J K-State Reactor 4-16 Original (5/05) d Safety Analysis Report

REACTOR DESCRIPTION In(2) 1

'r,h, 2k,]

and heat flux is calculated directly as:

q"=UAT= rn(iri,)) r, 3 (3) h k, r h. 2k1 in which r. and r, are cladding inner and outer radii, h, is the gap conductivity, h is the convective heat transfer coefficicnt, and kf is the fuel thermal conductivity. The gap conductivity of 2840 W m2 K7' (500 Btu hI' fl " FI) is taken from Appendix A. The convective heat transfer coefficient is mode dependent and is determined in context.

Parameters are cross-referenced to source in Table 4.6.

Table 4.6: Thermod namic Values Parameter Symbol Value Units Reference Fuel conductivity k, 18 W-miK' Table 13.3 14.9 W r'C (300~ I) Table 13.3 Clad conductivity k, 16.6 W rin' K'K (400 K) Table 13.3 19.8 W mnI' (6001K) Table 13.3 Gap resistance h, 2840 W rnK Apendix A Clad outer radius rO 0.018161 M Table 13.1 Fuel outer radius ri 0.018669 M Table 13.1 Active fuel length L 0.381 M Table 13.1 No. fuel elements N ( I NIA Chap 13 Axial peaking factor APF id22 N/A Table 13A General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that:

2L =5.1858E-04!flW (4) 2k, '

Gap resistance has been experimentally determined as indicated, so that:

- nK' .(5) t-=3.6196E-04'--

rh, W K-State Reactor 4-17 Original (5105)

Safety Analysis Report

CHAPTER 4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 K and 600 K. Under expected conditions, the value for 1271C applies so that:

(6) r.=3.10.3e-5-W k [V Table 4.7, Cladding Heat Transfer Coefficient Tcmp K)l Term (C 0 l mr Kw '

300 27 3.457e-5 400 127 3.103e-5 600 327 2.601e-5 I.-

It should be noted that, since these values arc less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small.

I'-

The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function.

Convection Heat Transfer Coefficient COD 700 900 1100 1M 10 170 ,. "De Power Level (KW)*

Figure 4.10, Convection Hear Transfer Coefficient versus Power Level I = I (7) h 0.0326P(watis) + 16985 K-State Reactor 4-18 Original (5/05)

Safety Analysis Report

REACTOR DESCRIPTION The reference core containsL" fuel clemcenIjtemperature calculations using the reference core are conservative because at least Elements are required for steady state 500 kW operations, while analysis assumes 1.25 MW operation. Actual heat productiX will be less than beat calculated in analysis, so temperatures will be lower. More th elements will distribute heat production across a larger number of fuel elements so that beat flux and temperatures will be lower than calculations based on ihe reference core.

Average heat flux per fuel rod is therefore:

, power

• P area *.83* 2nro(

With the maximum heat flux of:

to P

=q"*APF= *_= P = 0.423m-'P (9)

- ~ 83* 2wL, 2 332r.L; Therefore, core centerline temperature for the fuel rod producing the maximum heat as a function of power can be calculated as:

T =T + 0.423 16985 + 3.103 e - 5 + 3.620 e-4 + 5.186 e-4] (10)

For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 Kvalue and 600 K value) to determine expe6ted centerline temperature as a function of power level. Calculations using both vales are provided graphically, and shows the effect of thermal conductivity changes are minimal. The graph also shows that fuel temperature remains below about 750 IC at power levels up to 1900 kW with pool temperature at 27 IC (300 K), and 1700 kW with pool temperatures at 100 'C.

Table 4.8, Calculated Temperature Data for 1,250 kW Operation Fuel FuelGap Gap/Clad- CladAWater BuIk ° Centerline IC Interface "C. Interficeo C Interface "C Wt 503.2 229.0 37.7 -21.2

• 20.0 582.0 307.8 116.4 100.0 100.0

.I K-State Reactor 4-19 Original (5/05)

Safety Analysis Report

J CHAPTER 4 Hot Fuel-Rod Centerline Temperature at Power '-J (To mpo rature sevatlon over Pool Wale r Temperature)

J1

.O K-iC0l J

6W ... J J

21500 ... ..... J 400 J C

300t J I CJ:I-III rII

• T._

I TTII J

C0 J 100 J

_ , . J too 300 500 700 900 1100 1300 1500 1700 1900 Reactor Power (kW)

J

• 1 Figure 4.11, Hot Fuel-Rod Centerline Temperature J1 Finally, temperature calculations for the honest location in the core were made assuming J 1.25 MW steady state power at 20°C and 100°C with the following results: J J

For subcooled boiling, with water at ambient temperature Tb, the critical heat flux is J calculated by (Ivey and Morris 1978)

I J

(q".t).- = (qfl i1 0 .1 L)

_ Ct(Tz - - (11) J JI in which cy is the heat capacity of the coolant. At the depth of the reactor core in the J

KSU TRIGA, static pressure is 0.153 MPa and coolant ambient temperature is taken to be J 27°C. Thermodynamic and physical properties of water under these conditions are tabulated in Chapter4 Appendix B.

J J

b. Spatial Power Distribution I The following conservative approximations are made in characterizing the spatial J distribution of the power during steady-state operations. J

• The hottest fuel element delivers twice the power of the average.

• K-State Reactor 4-20 Onginal (5105)

Safety Analysis Report

REACTOR DESCRIPTION I Classically, the radial hot-channel factor for a cylindrical reactor (using R as the physical radius and R, as the physical radius and the extrapolation distance) is given2 by:

1.202* (12)

J{2A.0480(XR)]

with a radial peaking factor of 1.93 for the KSU TRIGA II geometry,. However, TRIGA fuel elements are on the order of a mean fice path of thermal neutrons, and there is a significant chanie in thermal neutron flux across a fuel clement.

Calculated thermal neutron flux data. indicates that the ratio of peak to average neutron flux (peaking factoi) for TRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection) has a small range of 136 to 1.40.

Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.

The axial distribution of power in the hottest fuel element is sinusoidal, with the peakpowera factor ofin2 times the average, and heat conduction radial only.

The axial factor for power produced within a fuel element is given by:

g(z)= 1514*co{!(.* 2*t+ e )' (13) in which t = L 12 and ,,is the extrapolation length in graphite, namely, 0.0275

m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor.

Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore cialculated temperatures will bound actual temperatures.

At full power, die thermal power ofP. = 1,250 kW is distributed overfifel elenients, with- niaximum heat fui givei by:

2 Elements of Nuclear ReactorDesign, 2" Edition (1983), J. Weisman, Section 63 3 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8114/1963), G. B. WVest K-State Reactor .'4-21 Original (5/05)

Safety Analysis Report

9)N CHAPTER 4 9 or*DP*L (14)

• The radial and axial distribution of the power within a-fuel element is given by q.'(r,Z) = q1" f(r)g(z), (15) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor followi from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core.

• 'The radial factor is given by:

a+cr+eer

-() I+br+di' (16) in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel (Ahrens 1999a). Values are: a = 0.82446, b = -0.26315, c= -021869, d = -0.01726, and e = +0.04679.

The fit is illustrated in Figure 4.11.

1.2 Li La 0.90 0.0 0.20 0.40 0.60 a.o to .2 1.4 16 1.3 2.0 r Cer)

Figure 4.12, Radial Variation of Power Within a TRGIA Fuel Rod.

(Data Points from Monte Carlo Calculations [Ahrens 1999a])

J K-State Reactor 4-22 Original (5105) Il Safety Analysis Report J

j

REACTOR DESCRIPTION

c. Steady-State Mode of Operation Table 4.9 tabulates: critical heat flux, Eqs. (3) and (4), as a function of coolant

• temperature from the assumed inlet temperature to the saturiti6n temperature. Also shown in the table is the CHFR, i.e.: the ratio of the critical heat flux and the maximum heat flux at full power.

Table 4.9: Critical Heat Flux and CHFR for 1.25 MW Operation at Selected Coolant Temperaturcs.

Tb CC) TurTb (n . q;, (MW m 2 ) CHFR 27 84.9 6.02 . 5.8 30 81.9 5.86 5.6 40 71.9 5.33 5.1

60. 51.9 4.26 4.1 100 11.9 2.13 2.0 111.9 . 0.0 1A9 1.4 It is clear from the table that there is a very wide margin between the operating heat flux and the critical beat flux even to unrealistically high pool water temperature, so that film boiling and excessive cladding temperature is not a consideration in steady-state operation. A parametric variation ofthis calculation for various power levels shows margin to DNBR up to 100IC pool temperature for power levels greater than 1.9 MW.

Critical Heat Flux Ratio I _n03 _-...-a)2 _W--two-f *225SEr _-.....-IJ MW -. 6-1 Ia 11+111 3'

V W-eU.,.kW - -

a 4.,

25 x0 25 40 45 so S e so s St es 70 go es 1o Butk Poot Temperature IC)

Figure 4.13, Ratio of Departure from Nucleate Boiling to Critical Heat Flux at Various Power Levels K-State Reactor 4-23 Original (5/05)

Safety Analysis Report

CHAPTER 4

d. Pulsed Mode of Operation Transient calculations have been performed using a custom computer code TASCOT for transient and steady state two-dimensional conduction calculations (Ahrens 1999). For these calculations, the initial axial and radial temperature distribution of fuel temperature was based on Eqs. (6) and (7), with the peak fuel temperature set to 746 C, U.., a temperature rise of 719 C above 27 0C ambient temperature. Tbe temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1% ($3.00) pulse from zero power and a 0.7% (S1.00) pulse from power operation. In the TASCOT calculations, thermal conductivity was set to 0.18. W cml K' (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cmt l K7I. The convective heat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chen 1963, Collier and Thome 1994) q"=h(7. -T,)=h(T-b)- (17) 1000 I.-

U TO 0

t.

L C

.o E

91 0 ' I ' I .'. .I. . . . . . . . I .

0.0 0.20 0.40 0.60 0.80 1.0 12 1.4 1.6 1.8 2.0 2.2 Radius (cm)

Figure 4. 14, Midplane Radial Variation of Temperature Within the Fuel Subsequent to a S3.00 Pulse.

The boiling beat transfer coefficient is given by the correlation (Forster & Zuber 1955)

K-State Reactor 4-24 Original (5105)

Safety Analysis Report

REACTOR DESCRIPTION 079

• OA4S *)

k = 0.00122$ a0:'09*o4( 0-V)°73 7 *(T. -T..) (18) in which TL isthe cladding outside temperature, To the saturation temperature (111.9 °),

and Tb the coolant ambient temperature (2rC). Fluid-property symbols and values are given in Appendix B. Subscriptsf and g refer respectively to liquid and vapor phases.

The overall heat transfer coefficient Uvaries negligibly for ambient temperatures from 20 to 60°C, and has the value 0.21 W crnml K' at Tb =270 C.

Figure 4.14 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.8 lists temperatures and heat fluxes as function of time after a 2.1% ($3.00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of IA9 MW m from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 of Appendii A, using the Ellion data, indicates a Leidenfrost temperature in excess of 5000 C. Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse.

8001- A -

V C

600 L.

• 0e 400 E

I.-4 200 020 0.40 0.60 0.80 .'LO A.2 . 14 1.6 1.8 2.0 22 -

Radius (1)

K-State'Reactor

• 4-25 Original (5105)

Safety Analysis Report

CHAPTER 4 Table 4.10, Heat Flux and Fuel Temperatures Following a S3.00 Pulse from Zero Power, with 27CC) Coolant Ambient Tcmperature.

Tm s Fuel outside Clad surface Time (s) (W mn2) CHFR Temp. (0 C) Temp. (C) 0 953 -

1 3.57 xIO' 4.2 781 224 2 7.34 xIO 2.0 683 432 4 8.52 xlO' 1.7 574 498 8 7.54 x10 5 2.0 461 443 16 5.71 xlO' 2.6 344 342 32 3A6 x10 5 4.3 224 218 64 1.04 x105 14.4 100 84 4.6 Thermal Hydraulic Design and Analysis A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure I'.

gain is given by 4An = pe/flATgL. (19) in which p. and P. are the density and volumetric expansion coefficient at core inlet conditions (27C, 0.15285 Mpa), g is the acceleration.of gravity, 9.8 c2 s7', AT is the temperature rise through the core, and L is the height of the core (between gridplates), namely, 0.556 n. The frictional pressure loss is given by ADp = 2ih2 JL (20)

'2A DA~p P in which his the coolant mass flow rate (kg se') in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 mn, and DO is the hydraulic diamneter, namely, 0.02127 m. The friction factorffor larninar flow through the annular area is given by 100 Rc ' (Shah & London 1978), in which the Reynolds number is given by D~lh / Ap, in which p. is the dynamic viscosity at core inlet conditions.

Entrance of coolant into the core is from the side, above the lower grid plate (see Section 4.2.5), %1.

and the entrance pressure loss would be expected to be negligible. The exit contraction loss is given by

,h 2K (21)

, 2p=A2 '

K-State Reactor 4-26 Original (5105)

Safety Analysis Report

REACTOR DESCRIPTION The coefficient K is calculated from geometry of an equilateral-trianilc spacer in a circular opening, for which

[Al 2 3 *A2 sin 600 cos60 1 Kg =L ;L- = 0.171, (22) where R is the radius of the opening in the upper grid plate. Equations (11) through (13), solved

• simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4.9.

Table 4.11; Coolant Flow Rate and Temperature Rise for Natural-Convecilon Cooling the TRIGA Reactor During Steady-State Operations.'

P(kWt) A (kgs") AT( 0 C) 50 0.047 3.1 100 0.061 4.7*

200 0.077 7.5 300 0.090 9.6 400 0.100 11.5 500 0.108 13.3 750 0.125 17.2 1000 0.139 20.6 1250 0.150 23.8 4.7 Safety Limit As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperatureliniits both steady-state and pulse-mode operation. The fuel temperature limit stems from potential hydrogen outgassing fr6m the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogein gas. Yield strength of cladding niaterial'decreases at a temperature of 5000 C; consequently, limits on fuel temperature change for cladding temperatures greater than 500'C. A maximum tcmperature of 1150 0 C (with clad < 500'C) and 950'C (with clad > 5000 C) for U-ZrH (H/Zrl,5) willinmit internal ftiul cladding streses that might lead toclid iniegity (NUREG 1282) challenges. .

• 4.8 Operating Limits 4.8.1 Operating Parameters The main safety consideration is to maintain ihe fuel temperature -below the value that would result in fuel damage. Setting limits on other operating parameters, that is, limiting safety system settings, controls the fuel temperature. The operating parameters established for the KSU TRIGA reactor arc:

K-State Reactor 4-27 . Original (5/05)

Safety Analysis Report

CHAPTER 4

• Steady-state power level

• Fuel temperature measured by thermocouple during pulsing operations

• Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations. As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels. As indicated in Table 4.9, the ratio.of actual to critical heat flux is at least 2.0 for temperatures less than 1000 C bulk pool water temperature for 1.25 MW operation. Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin.

Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation.

4.8.3 Safety Margins For 1,250 kV steady-state operations, the critical heat flux ratio indicated in Table 4.9 ranges from 5.8 for pool water at room temperature (271C) to 4.1 at 60 'C (pool temperatures are controlled to less than 481C for operational concerns). Even at pool water temperatures approaching boiling, the margin remains above 2. Therefore, margins to conditions that could cause excessive temperatures during steady state operations while cladding temperatures is below 5001C are extremely large.

Normal pulsed operations initiated from power levels below 10 kW with a S3.00 reactivity insertion result in maximum hot spot.temperatures of 7460 C, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of power operaiion can result in a maximum hot spot temperature of 869°C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15%

margin to the fuel temperature safety limit for cladding temperatures below 5000 C.

Analysis shows that cladding temperatures will remain below.5001C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve 5000 C (invoking a 9501C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding.

Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause cladding temperature to rise above 5001C, but the heat transfer simultaneously cools the fuel to much less than 9501C.

K-State Reactor 4-28 Original (5105)

Safety Analysis Report

REACTOR DESCRIPTION If fuel rods are placed in an-air environment immediately following long-term,-high power operation, cladding temperature can essentially equilibrate with fuel-temperature. In worst-case air-cooling scenarios, cladding temperature can exceed 5000C, but fuel temperature is significantly l6wer than the temperature limit for cladding temperatures greater than 5001C.

4.9 Bibliography "TASCOT: A 2-D. Transientand Steady State Conduction CodeforAnayphsisof a TRIGA Fuel Element, " Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C.,

"Investigationof the Radial Variationof the Fission-HeatSource in a TRIGA Mark III Fuel Element Using MCNP' " Report KSUNE-99-01, Dcpartment of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrens, C.,

"A CorrelationforBoilingHeat Transfer to SaturatedFluidsin Convective Flow,"ASME Preprint 63.HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C.,

KansasState University TRIGA MklJ ReactorHazardsSummary Report, " License R-88, Docket 50-188, 1961. Clack, RAV., J.R Fagan, W.R Kimel, and S.L Mikhail Convective Boiling and Condensation,3rd ed., Oxford Press, New York, 1994.Collier, J.G., and J.R. Thorne,

'Bubble Dynamics andBoiling Heat Transfer. "AlChE Journal 1, 532 (1955). Forster, HIK, and N. Zuber, Theory and Design ofModern PressureVessels, 2d. ed., Van Nostrand Reinhold, New York, 1974. p.3 2 . Harvey, J.F.,

"On the Relevance of the Vapour-LiquidExchange MechanismforSub-Cooled BoilingHeat Transfer at High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winfrith, 1978. Ivcy, H. J. and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457460 (1980). Lienhard, J. H. and V. K. Dhir, Thermal Migration ofHydrogen in Uranium-ZirconiumAlloys, General Dynamics, General Atomic Division Report GA-3618, November 1962. Mcrten, U., et al.,

ANRC, McClellanNuclearRadiationCenterFacilitySafetyAnalisReport, Rev. 2, April 1998.

NVUREG-1282, "Safety EvaluationReport on High-UraniumContent, Low-Enriched Uranium-Zirconium HydrideFuelsfor TRIGA Reactors," U.S. Nuclear Regulatory Commission, 1987.

K-State Reactor 4-29 Original (5/05)

Safety Analysis Report

CHAPTER 4 Ij "LaminarForcedConvection in Ducts," p. 357, Academic Press, New-York, 1978. Shah, RXK, and AL. London, "The U-Zr-HxAlloy: Its Propertiesand Use in TRIGA Fuel. 'Report E-1 17-833, General Atomics Corp., 1980. Simnad, M.T.

"Sofety Analysis Report, TRIGA ReactorFacility, NuclearEngineering Teaching Laboratory, Universityof Texas atAustin, Revision 1.01, Docket 50.602, May, 1991.

K-State Reactor 4-30 Original (5105)

Safety Analysis Report

Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the University of Texas Reactor Facility (UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998).

The following discussion relates the eleimient clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution shown in Figure 4A.I. This initial steep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the beat flux is sufficiently high, film boiling may occur and form'an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fucl temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies [Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface.

Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account.for fluid flow and temperature dependent material properties. Calculations show thit if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperatiire of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant has approached the saturation temperaturc, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [1961) for transient heating of ribbons in 100VF water, showed burnout fluxes of 0.9 to 2.0 Mbtu ft hr' for e-folding periods from S to 90 milliseconds. On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately OA Mbtu flh2 ir'. It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e, when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation.

A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A.land 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent. A boiling heat transfer model, as shown in Figure 4A.3, was used in order to obtain an upper limit for the clad temperature rise.

The model used the data of McAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boiling regime. -AXconservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760°C clad temperature for a peak initial fuel temperature of I000°C, as shown in Figure. 4A.4. Fuel tempeiature distributions for this case are shown in Figure 4A.5 and the heat flux into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-Stale Reactor 4A-1 Original (9102)

Safety Analysis Report

J J

CHAPTER 4 APPENDIX A J

J assuming a steady-state DNB correlation. Subsequently, experimental transition and film boiling data were found to have been reported by Ellion [9] for water conditions similar to those for the J TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of S. An appropriate correction was made which J resulted in a more realistic estimate of 470°C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad J

temperatures of 400°C to 500°C for TRIGA Mark F fuel elements which have been operated under film boiling conditions [Coffer et al. 1965].

J I1i*

J J

1.2 J-I.1l-

'iJ JI 1.01- J U

JU 0.9 I-j

_. a I

-0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.J 'IJ

• UIVS (IN.)

Figure 4A.1. Reprcsentativc Radial Variation of Power Within the TRI, 'A Fuel Rod IJ l.1 j U

1.0 I.

jJ 0.2 U J Il OU. J 0.7 0.6 0.S L

,' 0 1 2 3 4 5 6 7 8 AXIAL DISTAUCE F4 KiD-PUKt or rutt EL NLT (IN.)

Figure 4A.2, Representative Axial Variation of Power Within the TRIGA Fuel Rod.

K-State Reactor 4.A-2 Original (9102)

Safety Analysis Report

REACTOR DESCRIPTION

,o' w [ .

4

[

S 7

I I

. E S

A-A 4

a l . I e~lo -

I.-

ICt" S . _I . SI S I I 2

10 3o to' 1

• "0 5A; tor)

.Figure 4A3, Subcooled Boiling Heat Transfer for Water.

aoe C

tic 1300 1200 0.4 0.3 0.4 LaDIsS (10.)

Figorc4AA, Fuel Body Temperature at the AMidplanc ofa ll-Dondcd Fuel Element After Pulse.

. K-State Reactor 4A-3 Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A l'a B o i IC$

y o 3I I0 _I

1. .00i 0.01 *.1 1.0 to 100 wnItA Itp' vaat gu or Ktw( (stli Figure 4A.S, Surface Hcat Flux at the Midplane of a Wcell Bonded Fuel Element After a Pulse.

J 10.000

- 1000 t.

Ice 10 I-0.001 0.1 I.0 100 EUPIO tJs' noq tcm PSt (siCJ Figure 4A.6, Clad Temperature at bidpoint of WcIl-Bondcd Fuel Element.

Original (9/02)

K-Slate Reactor 4.A-4 4.A4 Original (9t02)

Safety Analysis Report

REACTOR DESCRIPTION The preceding analysis assessing the maximum clad temperatures associated with film boiling.

assumed no thermal resistance at fuel-clad interface: Measurements of fuel teniperatures as a function of steady-state power level provide evidence thai'after operating at high fuel temperatures, a permanent gap is produced between th fuel body and the clad by fuel expansion.

This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffcr report [1965].) The gap thickness varies 'with fue temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to wAiden the gaj and decrease-the heat transfer rate. Additional thermal resistance due to oxide and other filns on the fuel and clad surfaces is expected "Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry' 1962) which provide insight into the mechanisms involved. They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with'an assumed -value for the effective gap conductance, inorder to determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the-incipient film boiling conditions of the 1000 0 C peak fuel temperature case.

For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant floi' velocity of I f per second, which is within the range of flow velocities computed for natural convection under various steady-state conditiojns for these reactors. The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [1960]

which has been reported by Spano [1964] to give agreement with SPERT 11 burnout results within the experimental uncertainties in flow rate.

The transient thermal calculations were performed using effective gap conductances of 500,375, and 250 Btu flu hr' OF'. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the lo-cal computed surface heat flux and the DNB heat flux according to Bernath. Tbhe axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu 2 hri' 'F" is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRIGA fuel.

The surface heat flux at the midplane of the element is shown in Figure 4A.10 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu - 2 hf' IF' according to the more conservative Bemath correlation at a local water temperature approaching saturation.

K-State Reactor 4A-5

• Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A This analysis indicates that after operation of the reactor at steady-state power levels of I MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000 0 C is conservatively estimated to be 4701C.

As can be seem from Figure 4.7, the ultimate strength of the clad at a temperature of 4701C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel elementwill not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 1000TC, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is 11500 C would not produce a stress in the clad in excess of its ultimate strength. TRIGA fuel with a hydrogen to zirconium ratio of at lIast 1.65 has been pulsed to temperatures of about I 150'C without damage to the clad [Dec ct al. 1966].

7 I-U.

6 r

L'k B

0 5

I-Ca x

x IaJ Hi6 4

6-3 7 8 9 lo 11 12 13 DISTANCE FROM BOTTOM OF FUEL (IN.)

1/4..

Figure 4A.7, Surface Heat Flux Distribution for Standard Non-Gapped (h.,p=

500 Btuth fte E) Fuel Element After a Pulse.

I'-

K-State Reactor 4.A-6 Original (9102)

Safety Analysis Report

REACTOR DESCRIPTION I-Z-

a H.

1I 6.1 2

I a 5 la it 12 I$ is DlS1V1ANC rFAOM O SO1t Of er 1 L (IN.)

Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (h.,p= 375 Btulh ft2 F) After a Pulse.

6 7'. :CRITI CAL HEAT fLWX eV

-s

" 6

-I M5 T ELAPSED TIME FROM END OF PULSE IS 0.440 SEC w4 S.

t"s 3 2l ACTUAL HEAT fLUX 2

I 7 a 9 30 )1* 12 13 14 15

• DISTANCE FROM DOTTO8 OF FUEL (IN.)

Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hNpp 250 BMuMl ft *F) After a Pulse.

- K-State Reactor 4A-7 Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A

• 106

'.2 I

I

• 1 J

J IA.

S-5-

-I J

• 1 J

> 10 J

tu3 J

J d

%4

%J 0.1 4 ELAPSEO TIME FROM END Of PULSE (SEC)

\J a

Figure 4.t1O, Surfacc Heat Flux at M~Iidpoint vs. Time ror Standard Non-Gapped Fuel 4 Element After a Pulse. 4 4

4 4

4 4

K-State Reactor 4

4.A-8 Original (9102)

Safety Analysis Report .4

REACTOR DESCRIPTION Bibliography "A Theoty ofLocal Boiling BurnoutandItsApplication to Existing Data," Heat Transfer -

Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56, No.

20.Bernath, L.,

Research in Improved TRIGA ReactorPerformance,FinalReport, Gencral Dynamics, General Atornic Division Report GA-5786, October 20,1964. Coffer, C.O., ct al.,

CharacteristicsofLargeReactivity Insertions in a High Performance TRIGA U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12,1965.Coffer, C.O., et al.

Annular CorePulseReactor,General Dynamic, General Atomic Division Report GACD 6977, Supplement 2,1966Dee, J. B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Sirnad, Temperature Variation.Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water, Report SAN-1 001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, ct al.,

A Study of the Mechanism ofBoiling Heat Transfer,JPL Memorandum No. 20-88, March 1, 1954Ellion, M.E.,

Thermal Conductance ofMetallic Surfaces in Contact,USAEC NYO-2130, May, 1959.Fcnecb, H., and W. Rohsenow, An Analysis ofa Thermal ContactResistance, Trans. ANS 5, p.476, 1962.Fenech, H., and JJ.

Henry, "Tlhermal ConductanceAcross MetalJoints, "bMachine Design, Sept.15,1960, pp 166-172.

Graff, WJ.

Heat Transmission,3rd Ed., McGraw-Hill, 1954McAdams, -W.H..

AWRC, McClellan NuclearRadiation CenterFacilitySafetyAnalysis Report, Rev. 2, April 1998.

Heat, Mass andMomentum Transfer,Prentice-Hall, 1961, pp 231-232.Rohsenow, NV., and H.

Choi, "Quarterly TechnicalReport SPERTProject. April, May, June, 1964, " ISO 17030. Spano, A.

H.,

"TheEffect ofSubcooledLiquidonFilmBoilingZ"HeatTransfer 84,149-156, (1962).Sparrow, E.M. and R.D. Cess,

.K-State Reactor 4A-9

• Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A "Onset of Stable Film Boiling and the Foam Limit," Int. J. Hcat and Mass Transfer 6,987-989, (1963). Speigler, P., et al.,

UTA. University of Texas at Austin TRIGA ReactorFaciltySafety Analysis Report, Docket SO-602, Rev. 1.O1,May 1991.

"HydrodynamicAspects ofBoilingHeat Transfer,"AEC Report AECV-4439, TIS, ORNL, 1959.

Zuber, W.

K-State Reactor 4.A-10 Original (9102)

Safety Analysis Report

Appendix B Water Properties at Nominal Operating Conditions The following data are from thc NBS/NRC Steam Tables, by L. Haar, J.S. Gallagher, and G.S.

Kell, Hemisphere 1984.

Saturated 'Water at 0.153 MPa TEMPERATURE, T 111.9 _C PRESSURE, P 0.153 MPa HEAT OF VAPORIZATION, X 2.22E+03 kJ kg"'

SURFACE TENSION. a 5.66E-02 J m '72

p. density (kg m') 9.50E+02 8.78E-01 C, beat capacity (kJ kg' K') 3.71E+00 1.55E+00 C, heat capacity kJ kNW' K') 4.24E+00 2.09E+00 s, entropy (kJ kW' K'). 1.44E+00 7.22E+00 i, enthalpy (kJ kg') 4.70E+02 2.69E+03 u, internal cnergy (kJ kR ) 4.69E+02 2.52E+03 sonic speed (m s') 1.53E+03 4.79E+02 k., thermal conductivity (W m' K') 6.82E-01 2.65E-02 ui. dynamic viscosity (kg m's ') 2.50E-04 1.27E-05 v, kinematic viscosity (m2 s ') 2.64E-07 1.45E-05 cL, thennal diffusivity (m2 sl) 1.70E-07 1.44E-05 Pr, Prandtl Number 1.55E+00 1.0OE+00 1, volumetric expansion coefficient 8.14E-04 2.86E-03 Subcooled Water at 27 0 C, 0.153 MPa TEMPERATURE (T) 27 °C PRESSURE (P) 0.153 MPa DENSITY (p) 9.97E+02 kg m'3 C,.heat capacity (kJ kg'l K'1)

Cpheat capacity (kW kg7' KI) s, entropy (k kg' K')

i, enthalpy (kJ kg')

u, internal energy (kJ kg"')

sonic speed (i s')

k, thermal conductivity (W mn )

p, dynamnic viscosity (kg m'fs'l) v, kinematic viscosity (m2s'r) cc, thermal diffusivity (m2 s')

Pr, Prandtl Number

, volumetric expansion coefficient K-State Reactor 4.B-1 Revised 06128/05 Safety Analysis Report

6. Engineered Safety Features As discussed in Chapter 13, from previous analysis, and from experience at other TRIGA reactors, emergency core cooling is not required for operations at steady state thermal powers below 1900 kWV. No engineered safety features are required for the KSU reactor because the steady state power limit is 1,250 kW.

6.1 Bibliography KansasState University TRIGA Mark il ReactorHazardsSummary Report, by RAV. Clack, J.R.

Fagan, WIR Kimel, and SZ Mikhail, License R-88, Docket 50-188, 1961.

Analysis of CertainHazardsAssociated sith Operationof the KansasState University 7PJGA Markl))Reactorat250 kWSteady State and vith Pulsed Operationto 52.00, by R.W. Clack, et al., and the Safety Evaluation by the U.S. Atomic Energy Commission Division of Reactor Licensing, LicenseR-88, Docket 50-188, 1968.

NUREG-1282, "Safety EvaluationReport on High-Uranium Content, Low-Enriched Uranium-Zirconium HydrideFuelsforTRIGA Reeators," U.S. Nuclear Regulatory Commission, 1987.

K-State Reactor 6-1 Original (5105)

Safety Analysis Report

7. INSTRUMENTATION AND CONTROL SYSTEMS Much of the reactor's original instrumentation and control (I&C) systems were replaced during the control room modifications in 1993 and 1994. The original console and vacuum-tubc instruments were replaced by a surplus solid-state console obtained from U.S. Geological Survey's TRIGA Mark-] reactor. This console was then outfitted with new N-1000series neutronic channels from General Atomics.: These channels hive optically isolated outputs, allowing other devices to utilize the neutronic data.

7.1 Summary Description The bulk of the reactor I&C systems are hard-wired analog systems primarily manufactured by General Atomics and widely used at various NRC-licensed facilities. .te general layout of these systems is shown in Figurc 7.1.

.REACTOR OPERATOR INPUTS .I I*:  : CONTROL CONSOLE OUTPUTS - .

alit I Ro *1 ii ftw It UN S. S . . .T REACTOR- . - .* .:. COO .................

LING SYSTEM .. . _

......... ....... .. .. J . , "... *I . .

Figure 7.1,lnter-connectivity Diagram. .3 .

The reactor control system (RCS) consists of the instrumentation channels, the control rod drive circuitry and interlocks, and an-automatic flux coitroller. The RCS measures several key reactor parameters including power, fuel temperature, water temperature, and water conductivity. Threc neutronic instruments measure reactor power ~separately: a wide-range logarithmic channel, a multi-range linear channel, and a percent power channel. Thesc provide at least two indications K-State Reactor 7-1 Original (5105)

Safety Analysis Report

CHAPTER 7 of reactor power from source range to power range. Additionally, if a reactor pulse is performed another channel is added to the central thimble to record pulse data. Fuel temperatures can be monitored on both the console and on an auxiliary panel. Primary water temperature is displayed on the console and measured by an RTD in the water box. Titanium electrodes at the entrance and exit of the cleanup loop measure water conductivity.

The control rod drives and their associated circuitry are simple in design. A rotary switch configures the primary mode of operation, namely automatic, steady-state, or pulse mode.

Numerical indicators give drive position, with illuminated switches to manipulate the rods and to indicate rod and drive status. Several interlocks are incorporated to prevent unintentional rapid insertions of reactivity, except in pulse mode. An automatic control system links the RCS with the neutronic channels providing regulation of the power level.

The reactor protection system (RPS) is a component of the RCS instruments. The RPS will initiate a reactor scram if any of several measured parameters in the RCS are outside of their limited safety system settings. The reactor scram effectively places the reactor in a subcritical configuration by releasing the control rods from their respective drives. Since the rods are no longer physically attached to the drive, they fall into the reactor core by gravity. High reactor power, high fuel temperature, loss of detector high voltage, loss of building power, and short reactor period will automatically cause all of the control rods to be dropped into the reactor core.

A bar above the control rod drive switches allows this system to be actuated manually. Since the core is cooled by natural convection, no other engineered safety features are necessary for safe reactor shutdown.

The control console and display instruments are primarily housed in a control console, with auxiliary instruments located in a rack next to the console. At the console, the reactor operator has direct control over mode of operation, control rod drive positions, cooling system operation, opening of reactor bay doors, and manual scram of the reactor. Display instruments located in the control console provide measurements of reactor power, control rod positions, primary water temperature, and fuel temperature. Indicators in the console display scram information, low air pressure, low primary water level, high reactor sump water level, sump high water level, sump overflow water level, secondary surge tank level low, source interlock status, reactor bay upper door open, reactor bay lower door open, thermal column door open, person on stairway, and rod drive status. Secondary surge tank makeup is controlled with a backlit pushbutton that indicates surge tank low level and surge tank makeup valve operation. An intercom system on the console provides communication to numerous locations around the reactor bay and staff offices. In the auxiliary rack, the operator can control pneumatic transfer system operation, actuate timers, and add water to the secondary cooling system. Display instruments located in this rack include, primary water conductivity, water activity, remote area radiation monitors, fuel temperature, and a strip-chart output bf reactor power. Several audible alarms indicate high radiation levels in the <i primary coolant and at various locations throughout the reactor bay. A breaker-box in the control room provides control over electrical devices in the reactor facility, including ventilation systems.

Radiation protection instruments are distributed throughout the reactor bay. :AII instruments have J visual indication of radiation level, visible alarm conditions, and audible alarms. Radiation area monitors (RAM) strategically cover potential radiation areas throughout the reactor bay. A \J combined pool surface and primary.water monitor indicated water activity. A S R h l evacuation I I

K-State Reactor 7-2 Original (5105) 1 Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS alarm is located on the 22-foot level. A continuous air monitor (CAM) is :energized during reactor operation.

The huirna-machine interface principles incorporated into control room design allow the reactor to be operated by a single individual. All zionitoring instruments are visible to the reactor ojerator at the console. The instruments and controls necessary for reactor operation are within reach of the operator, including an intercom and telephone. Surveillance instruments are located nex~t to the console, with visual and audible alarms to signal the operator to abnormal conditions.

7.2 Design of Instrumentation and Control System 7.2.1 Design Criteria Reliability of essential equipment is ensured through redundancy. Multiple instruments and safcty systems perform similar functions for all modes of reactor operation. The construction and installation of instruments was performed according to applicable regulations.at the time of introduction. However, all crucial instruments are checked daily for calibration and operability.

Testing and calibration procedures exist for repair and general service.. The majority of these I&C systems were manufactured by General Atomics,. or other industrial manufacturers of nuclear equipment Crucial systems to be considered include neutronic instruments, control rod drives, radiation monitors, and control systems.

Redundancy is designed into each of these systems. During steady state operation, a minimum of two ncutronic channels provide reactor power level indication, two of which provide high power-level RPS actuation (scram). These neutronic instruments are tested prior to reactor operation for demonstration of scram capability. Two are also tested for operability by internal calibration tests. There are two fuel temperature indications. The control rod drives drop their rods into the reactor core upon loss of power or RPS actuation, providing sufficient shutdown margin with even the most reactive rod stuck out. Multiple remote area radiation monitors cover important areas, including two directly above the reactor core and two monitoring primary coolant activity.

The maximum steady state power level for KSU TRIGA Mark II reactor is proposed to be SOO kW. Similar reactors operate up to 2 MW with 2 GW.pulses. Therefore, the limited .safety systems settings associated.with reactor are extremely conservative when compared to the safety limits of the reactor. Thus the reactor has a considerable safety margin.

7.2.2 Design-Basis Requirements The primary function of the RCS is to govern the manner in which reactivity is varied in the reactor core. The RCS system should prevent the reactor operator from unintentionally inserting large amounts of reactivity, through various interlock systems. The operator should only be able to remove one rod at a time from the reactor core, preventing large insertion rates. The pulse rod must not be ableto be rapidly ejected from the core while in steady-state operation. Furthermore, the pulse rod should. be the only rod that can be withdrawn in pulse mode, preventing supercritical pulses. There should also be an interlock to prevent startup without a power level signal above the minimum instrument sensitivity, preventing unmonitored or unanticipated K-State Reactor 7-3 Original (5105)

Safety Analysis Report

CHAPTER 7 criticality. Rod position indicators should show the rod position to 0.2% or total travel for accurate reactivity calculations.

Another primary function of the RCS is to provide the reactor operator with reactor status information. Reactor power, a crucial parameter, requires at Icast two instruments to provide confirmation of reactor power from shutdown to operating levels. Three instruments are used to cover this range: a wide-range logarithmic channel, a multi-range linear channel, and a percent power channeL Accuracy of measurement at full rated power increases accordingly with the refinement of scale. The log channel provides gross reactor power indication and is accurate to 20% of scale, the linear channel is accurate to 5% of scale, and the percent power channel is accurate to 3% of scale. The percent power channel will also display pulse parameters for large pulses. These instruments arc calibrated annually and checked for operability at the start of each operating day. An additional channel is installed and calibrated in the central thimble to record pulse data. Fuel temperature must be monitored during pulsing operation.

The primary function of the RPS is to automatically insert the control rods into the reactor core when certain parameters deviate from limited safety system settings. Several scrams involve the neutronic channels in the RCS. If 110% rated power level is exceeded in steady state mode, one of two trip-points will scram the reactor. Failure of the high voltage power supplies for operating neutronic channels will also cause a scram. Manual scram will be available in all modes of reactor operation. Rod drop times for the standard rods will be measured regularly to ensure proper RPS function. No other ESF features are required in this design.

The primary function of the radiation monitoring instruments is for personnel protection measures and emergency assessment actions. The area monitors provide the reactor operator with information regarding the actual radiation environment inside the reactor bay. With this knowledge, reactor users can be informed of possible hazards. A 5 R hX monitor on the 22-foot level signals personnel to evacuate the reactor bay. A number of survey instruments (ion chambers, rem balls, G-M counters) are also available to personnel Other instruments such as the constant air monitor, pool surface monitor, and water box monitor indicate the presence of dispersible radioactive materials, an indication of possible fuel cladding failures.

The control room is designed so a single operator can manipulate all significant controls without leaving the room. The reactor operator should be able to dc-energize all equipment and experiments in the reactor bay. The control room should provide sufficient ventilation to provide cooling of the reactor instruments.

K-State Reactor 7-4 Original (5105)

Safety Analysis Report

_ INSTRUMENT AND CONTROL SYSTEMS 7.23 System Description The overall system layout is depicted in Figure 7.2. The majority of the RCS is housed ina General Atomics (GA) console originally manufactured for the USGS reactor, which is shown with modifications in Figure 7.2. A detailed description of this figure is provided in Table 7.1.

Figure 73 'shows a representative layout of the auxiliary instrumentation rack. Since the instrument racks are general use equipment, -configuration may be changed to allow better utilization of space, installation of new equipment, support specific equipment modifications, etc.

without affecting function. The functions of each piece of equipment in this configuration are discussed in following sections with additional figures showing location and layout.

7.2.4 System Performance Analysis The systen performance of the current I&C systems surpasses the original equipment. Reliability has seen high, with few unanticipated reactor shutdowns. Since daily checkouts are performed, any discrepancies would be observEd and corrected in a prompt manner..-The opto-isolated outputs of the iieutronic channels allow th'e data to be utilized by other devices without concern

• over those devices affecting the channels.. A line conditioner provides regulated power to the instrumrents, protecting the equipment from electrical disruptions.

7.2.5 Conclusion The current I&C systems outperform the original equipment supplied with the reactor, while meeting all of the necessary design bases for the facility.' The human design factors used in control room development allow the reactor to be operated by a single individual. Checkout and testing procedures ensure that all equipment is maintained in operational status.

Figure 7.2, USGS TRIGA Console with Modifications.

K-State Reactor *7-5 Original (5105)

Safety Analysis Report

CHAPTER 7 Rod drop timer NIM rack for Not used -'experiments Fuel temperature I and waler Not used  ; monitor Pneumatic system controls

.Conductivity monitor Strip chart recorder

..Area radiation monitors

• J Spare high voltage power Alarm panel supply Drawer unit I

• -Cooling fan 3 Battery charger C C

\C J

J Figure 73, Instrumentation Rack.

J J

C K-State Reactor 7-6 Original (5105) \C Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS Table 7.1, Description of Figiure 7.2.

Nu mber Function Description 1 Console Power Push Button Switch 2 Magnet Power / Scram Reset Key Switch 3 Cohtrol Rod Drive Position Push Biution Switches 4 Apply Air to Pulse Rod Push Button Switch

.5 Rod Position Indicators *LED Displays 6

• Mode Selector Rotary Switch 7 Removed *Backup Range Switch 9 Automatic Power Demand Control I0 Turn Potentiometer 10 Manual Scram Bar Bar Covering Scram Switches 11 (not used) Backup Period Channel 12 Fuel #2 & Water Temperature Display and Selector Switch 13 Scram Status, Source Interlock, Low. Indicators and Control Air Pressure, Hi & Hi-Hi sump level, Switches

• surge tank level & makeup, Upper &

• LowerDoors, and Cooling System Power

• 14 (not used) Backup Count Rate Channel 15 (not used) Backup Log Channel 16- 17 (not used) . Backup Percent Power Channel

• 18 (not used) NIA 20 Wide Range Log Power Channel GA NLWV-1000 Channel 21 Multi-Range Linear Power Channel

• GA NMP-1 000 Channel 22 Percent Power and Pulsing Channel GA NPP-1000 Channel 23 Source Interlock Override Key Switch 24

• Period Scram Override Key Switch 7.3 Reactor Control System.

The bulk of the reactor control system (RCS) is housed in the USGS console shown in Figure 7.2.

The remainder is contained in the auxiliary rack-mount panel next to the console, shown in Figure 73. The RCS consists of the instrumentation channels, the control rod drive circuitry and interlocks, and an automatic flux controller. .These are shown in Figure 7A. The RCS measures several key reactor parameters including power, fuel temperature, water temperature, and water conductivity.

7.3.1 Neutronic Instruments (Reactor Power).

Three neutronic instruments measure reactor power separately: a wide-range lcganithmic channel, a multi~iange linear channel, and a percent power channel,-as shown in Figure 7.5. Wiring diagramns and calibration procedures are found in the ifisiriument maintenance manuals listed in the bibliography.

K-State Reactor 7-7 Original (5105)

Safety Analysis Report

CHAPTER 7 The wide-range log channel uses a fission counter for detecting thermal neutrons in the range of IA to IA x 105 nv, and provides approximately 0.7 counts-nv'. The detector has an aluminum case, an aluminum electrode, a U30S Wolenriched in "M3)coating as the neutron sensitive material, and an argon-nitrogen mixture for a fill gas. A preamplifier is used to minimize noise and signal loss from the detector to the console, and it is located on the 12-foot lvel. J I . ., .- I j

,J

'i

• 11

-I Figure 7A., Instrumentation Diagram. I'.-

The remainder of the channel circuitry is located in the NLW-1000 unit in the central console.

The NLW-1000 unit supplies the high voltage for the detector and power for the preamplifier.

The instrument switches from pulse mode operation to current mode as reactor power increases out of the source range, allowing the instrument to measure reactor power in the upper ranges.

Trece displays indicate reactor power, high voltage, and reactor period. The power signal is permanently recorded via an opto-isolated output to a strip-chart recorder located in the instrumentation rack. The period meter has a scram at 3 sec and there is a high voltage scram, both of which are bypassed in pulse mode. This channel also provides a protective interlock which prevents rod withdrawal when indicated neutron flux is < 2 cps, which is also activated in pulse mode to prevent removal of the shim, safety and regulating rods. Another interlock prevents pulsing when reactor power is above 10 kW (normally set at I kW). The unit has two K-State Reactor 7-8 Original (5105)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS calibration checks in pulse mode, two in current mode, and checks for the period and high voltage scrams.

.7 ~ o . .! . - * .:. . .

Figure 7.5, N-1000 Series Instruments.

The second channel provides multi-range linear power indication. This channel uses - a compensated ion chamber for detection of thermal neutrons. The linear channel detector signal goes directly to the NMP-1000 unit in the center console, which in turn supplies high and compensation voltages. This unit features automatic or manual ranging to select the appropriate decade of power displayed. The instrument provides two indicators, power and high voltage.

The power signal is permanently recorded via an opto-isolated output to a strip-chart recorder located in the instrumentation rack. In addition there is also a high'po'wer level scram (nomnally set for 104% nominal rated power) and a high voltage scram. The signal from the detector and the high voltage scram are bypassed in the pulsing mode. -TDe unit has two calibration checks, an auto-ranging test feature, and checks for high powei level and high voltage scrams..

Power range-indication of neutron flux is provided by an un'compensated'ion chamber signal, which indicates percentige 'of power in the uippei two decades of -the power range. The uncompensated ion chamber is virtually identical in construction to the compensated ion chamber, but no gamma compensation is provided in the circuitry. The detector sends its signal to the NPP-1000 instrument in the center console, which provides a visual indication'of reactor power, high voltage, nv, and nvt measurements. The NPP-1000 supplies the high voltage for-the detector. There is a high po'Wer scram (iormally set for 104% of full power) and a high voltage scram.. In pulse mode, the -channel is designed to read off the'maximuim power'and integiral output of a reactor pulse. However, the pulse. output reiahigs are measured in reference to the 250 MW maximuL *Hence an additional channel is added to the central thimble to permit recovery of data frompulses of vari6us magnitudes. The unit has checks for high power and Migh voltage scrams.

K-State Reactor 7-9 Original (5105)

Safety Analysis Report

CHAPTER 7 An added pulsing channel consists of a small BF3 chamber, which can be inserted into the central thimble of the reactor core. A separate high voltage supply powers the instrument and a multi-range picoammetcr reads the detector current. A reference voltage output of the picoammeter is sent to a computer in the control room, which collects the pulse data. This channel is calibrated prior to pulsing operations and range selected in advance based upon the anticipated peak power.

7.3.2 Temperature Temperature indications for the primary water and specific B-Ring fuel elements are provided on the front section of the control panel and in the instrumentation rack. The instrumented fuel elements have three chromel-alumel thermocouples in the fuel element that are used for temperature indication on the console or in the instrumentation rack. The thermocouples are located 0.76 cm (0.3-in) below the fuel surface, spaced at the midpoint of the element and at i, 2.5 (1 in.) cm from the midpoint; an averaged value from all three thermocouples is typically used for instrument readings. The temperature in the primary cleanup loop is a nickel alloy thermnistor, and is displayed on a console meter, which is shared with fuel temperature via a rotary switch.

Another indication of fuel temperature is located in the instrumentation rack with the capability of initiating reactor scram if the measured fuel temperature exceeds a preset value (normally 400 CC).

Several other temperature measurements can be obtained from the computer in the control room.

The computer can read two additional fuel thermocouples from other fuel elements in various positions in the reactor core. Additionally AD590 temperature transducers are located on the inlet and exit of both the primary and secondary sides of the heat exchanger to evaluate performance. Two other transducers are located in the reactor tank for bulk pool temperature measurement and high temperature alarm.

7.33 Water Conductivity Primary water conductivity is measured at the inlet and outlet of the purification loop by titanium electrode cells that send signals to a bridge circuit in the instrumentation rack. The bridge circuit is automatically iemperature compensated and nulled to provide good conductivity measurements over all reactor conditions. The inlet and outlet conductivities provide a good indication of the overall purity of the primary water and the effectiveness of the ion exchanger.

7.3.4 Control Rod Drives Four control rods are required for reactor operations at 1,250 kW to meet reactivity control requirements: a shim rod, a regulating rod, a transient rod, and a safety rod. The shim, regulating and safety rods share identical control circuitry (Figure 7.7) and provide coarse and fine power control. Two of the rod drives arc original, analog systems. One of the rod drives uses a stepper motor. Drive position is determined by voltage drop across a potentiometer that is adjusted as the control rod drive is moved. The position indicator for the analog motors is attached to a shaft coupled to the drive motor shaft with a setscrew, while the stepper motor is connected to the K-State Reactor 7-10 Original (5105)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS position indicator with a chain drive. The pulse rod is deiigncd'so that it can be rapidly ejected from'the core to a preset height to initiate a reactor pulse. However, it still functionis as a normal control rod in steady state mode. All rods can-be individually scrammed without shutting down the reactor.

a. Standard Control Rod Drives The rod drive mechanism (see Figure .7.6) is an electric. motor actuated linear drive, equipped with a magnetic coupler. Its purpose is to adjust the reactor control rod position. In the analog drive motor, a IO-V, 60-cps, two phase motor drives a pinion gear and a 10-turn potentiometer. The potentiometer provides rod position indication.

The pinion engages a rack attached to the magnet drawtubc. An electromagnet mounted on the lower end of the drawtube engages an iron armature that screws into the end of a long connecting rod which terninates (at its lower end) in the control rod.

The magnet, armature, and upper portion of the connecting rod are housed in a tubular barrel that extends well below the reactor'water line.. Located part way down the connecting rod is a piston equipped with a stainless steel piston ring. Rotation of the motor shaft rotates the pinion, thus raising or lowering the magnet draw tube. If the magnet is energized, the armature and connecting rod will follow the draw' tube so that the control rod is.withdrawn from or inserted into the reactor core. In the event of a reactor scram, the magnet will be de-cnergized and release the armature. The connecting rod, piston, and control rod .will.then drop, thus reinserting the control rod into the reactor. Since the upper portion of the barrel is well yentilated, the piston will move freely through this range. However, when the connecting rod is within 2-in (5 cm). of the bottom of its travel, the piston is restrained by the dashpot action of the restricted ports in the lower end of the barrel. This restraint cushions bottonming impact. Control rod drop times are measured semi-annually and must be less than one second.

The analog.rod drive motor-is dynamically braked and held by an electrically locked motor. In the static condition, both windings are energized with the same phase (see Figure 7.7), electrically loikizig thi motor. Clockwise (up) or counier-clockwise (down) rotation is enabled by shifting the phase between the windings with a l-pF capacitor; motor control switches.allow.the.appropriate phase shift. The stepper motor operates using phase.switched direct currEnt power. .The motor shaft advances 200 steps per revolution (1.8 degrees per step) Since current is maintained on the motorwinding when the motor is not being stepped, high holding torque is maintained A translator module drives the stepping motor.

Three microswitches limit and control the travel of the magnet drawtube. Actuation of the magnet up limit microswitch (S901) applies line voltage to one winding therefore allowing only the phase shift, which gives counter-clockwise rotation.

Actuation of the magnet down limit microswitch (S902) applies line voltage to the other winding therefore allowing only the phase shift that gives clockwise rotation. Actuation of the rod down microswitch (S903A) causes the phase shift for counter-clockwise rotation. Therefore, if the control rod drops, the magnet drawtube drives down until the K-State Reactor 7-11 Original (5105)

Safety Analysis Report

CHAPTER 7 magnet down limit microswitch locks the rotor. Since the rod down microswitch drives the magnet draw tube down, then the rod down microswitch must be open before the magnet down microswitch during coupled withdrawal of the control rod.

Three lights indicate that, I) the magnet drawtube is full up, 2) the magnet drawtube is full down, and 3) the armature and magnet are coupled. When the magnet drawtube is full up, microswitch (S901) is actuated opening the short across the magnet up light (DS321). When the magnet draw tube is full down, microswitch (S902) is actuated opening the short across the magnet down light (DS324). When the control rod drops, the non-actuated magnet down microswitch (S902) and the actuated rod down microswitch (S903B) short the contact light (DS317) indicating separation of the magnet and the armature.

Other features of the circuit are an adjustable bias resistor (R902), a 220-ohm surge resistor, 50-ohm current limiting resistors. The adjustable bias resistor compensates for the torque applied by the weight of the control rod and the magnet drawtube. The 220-ohm surge resistor limits the capacitor current surge during-switching. The 50-ohm current limiting resistors limit the currents in the 12-volt indicating circuits when the indicating lamps are shorted.

The unconventional circuit employed in the rod-drive system minimizes the number of switch contacts required. Therefore, relays with their attendant reliability problems arc not required. It should be noted that the rod drive units are identical both mechanically and electrically (with the exception that one unit uses a stepper motor) and they are, therefore, interchangeable.

The rod position indicators are three digit, LED display indicators that receive a variable DC voltage input from 10-turn potentiometers that are driven by the respective rod drive motors. The digital display is simply' a voltmeter, since the voltage across the potentiometer is directly related to the control rod position. The position indicators have their own variable power supplies and are therefore completely independent. The indicator systems are located in the control console except for the 10-turn potentiometers on the drive and the associated wiring.

Normal rod motion speed is about 12-in. per minute. Using rod speed, rod position indication at UP and DOWN limit switch positions, and respective rod worth curves, the operator can determine the reactivity insertion rate for a given interval of rod motion.

K-State Reactor 7-12 Original (5/05)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS ET WIRNE CONDUIT Ersr 0%ftn1 MAGNET DRAW TU ROD DOWN LIMIT SWn PULL--OD APAI COVER MOTOR COV ROD PULL-ROD HOUSI

-MAGNET DRAW TUBE

-UACNET PULL R

- LOCKX ROD S6 rkd ci iD ic cii m'edan'is

Figure 7.6, Regulating, Safety and Shim Rod Drive.

a I : .

K-State Reactor :7-13 Original (5105)

Safety Analysis Report

CHAPTER 7 Its 1if n LINK

. UA'.

IIIAS tj" LIl1 RESISTOR U

Io IVZ II IMOTOR ZV id No.

I _ -L-W M901~SCCTUP no 10 w*04 C

oSizV _ma N 1 I( V-2.t,') MACtUIP *O 03 SV Aso CONT 30 Cow

- .021:' UAVCfi1

_ Z14S 5035 IINTERLOCEI 2a- I I MOTOR CONTROL NSN SWITCKES 31 0OWN NOTE:UMIT SWITCHES SHO%-N WITH ROD IN FULL DOWN POSMON.

Rod drive,. motor control. and Indieator lamp circuit Figure 7.7, Control Rod Drive Circuit.

b. Transient Rod Drive The transient rod-drive (Figure 7.8) is an electrically controlled, pneumatically operated, mechanically limited system. The transient rod and aluminum extension rod are mechanically connected to a pneumatically driven piston inside a worm gear and ball-screw assembly. The system is housed on a steel support structure mounted above the reactor tank. A three-way solenoid valve mounted below the support controls air to the piston. The throw of the piston, and hence the amount of reactivity inserted into the core during pulsing operations, is regulated by adjusting the worm gear and ball-screw assembly. The adjustment is made from the central console by actuating a reversible motor drive, which is coupled to a worm gear and a 10-turn potentiometer for position indication. The operation of the position indicator is identical to that of the shim, safety and regulating control rods. The drive circuitry is identical to the shim, safety and regulating rods (Figure 7.7), except for that the motor is not continuously energized.

Relays in the drive unit allow it act similarly to the electrically locked motors. Since air is used to support the rod, there is no compensation for rod weight. The remaining differences involve the pneumatic relay controls.

J J

K-State Reactor 7-14 Original (5105) Ji Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS The solenoid valve is actuated by means of the console mounted transient rod fire and air release (scram) switches. WThen the rod fire switch is depressed, the solenoid valve opens, admitting air to the cylinder, coupling the piston and rod to the shock absorber.

Depressing the air release switch dc-energizes the solenoid valve, which removes air from the cylinder and vents air to the atmosphere. In the event of a reactor scram, the solenoid will be dc-energized via the scram circuitry; which will allow the transient rod to drop into the core after the air is removed. Micro-switches are used to indicate the extreme positions, up or down, of the shock absorber. In steady state mode, an interlock prevents actuation of the rod fire switch if the drive is not in its fully down position.

In ihe pulse mode, a variable timer (usually six seconds) de-energizes the solenoid valve after the pulse is initiated. The shock absorber will remain in its preset position until the mode selector switch is taken to steady state. In the steady state mode of operation, the adjustable (normally six second) timer is disengaged and the cylinder remains pressurized. If the air supply for the pulse rod drive should drop below approximately 45 psig, an amber low air pressure warning light will be actuated on the control console.

Loss of air pressure will cause the rod to fall into the coic.

c. Interlocks Several interlocks are built into the control system of the reactor to prevent improper operation. These interlocks are hard-wired into the control rod drive circuitry. They are stated below:

I. No control rod withdrawal (shim, regulating and safety ixds only) is possible unless the count rate neutron chanfiel is indicating > 2 cps. This interlock preveits the possibility of a startup 'without a functional power level startup channel.

The low count rate interlock may be bypassed during fuel loading operations when core inventory is not high enough to multiply the source above 2 cps.

2. Air may not be applied to the pulse rod if the pulse rod shock absorber is above its full down position and the reactor is iAthe steady state mnode. This interlock prevents the inadvertent pulsing of a reactor in the steady state mode.

3. Tberc is no simultaneous withdrawal of two or more control rods when the reactor is in the -steady .state mode. This interlock prevents violation of the maximum reactivity insertion rate of the reactor.

4. The pulse rod is the only control rod that can be withdrawn when the reactor is in the PULSE mode (this does not prevent the scramimingof iny control rod). This interlock minimizes the p6ssibility of piulsing a supercritidal reactor. This interlock is provided by the source interlock, which is engaged When the log channel is placed in pulse mode.

K-State Reactor 7-15 Original (5/05)

Safety Analysis Report

CHAPTER 7

HOCK ABSORBER VENT HOLE EXTERNALLY THREADED CYLINDER NUT WORM

'CH-CYL UP PISTON ENT

' SUPPLY I.-

AIR SUPPLY, HOSE 'SOLENOID VALVE LIMIT SWITCI-CYL DOWN

-LIHIT SWITCH-ROD DOWN BOTTOlH LIMIT/

TH I

rodrIve K CONNECTION 10 COHNTROL ROO Transient rod drte mechanism Figure 7.8, Transient Rod Drive.

Additionally, there is an interlock that prevents reactor pulses from being fired if the reactor power is above 10 kW (normally set at I kW). There is also a key switch for bypassing the source interlock during fuel loading operations to check for criticality.

K-State Reactor 7-16 Original (5105)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS 7.4 Reactor Protection System The.reactor protection system (RPS).will initiate a reactor scram if any of several measured parameters are outside their limited safety system settings (LSSS). The reactor scram effectively shuts down the reactor by dc-cnergizing portions 6f the shim, regulating, transient, and safety rod drives, causing the control rods to drop into the reactor core by gravity. The shim, safety and regulating rod drives utilize electromagnets to hold up their respective control rods. The pulse rod drive.utilizes an electric solenoid to apply. compressed air to the pulse rod.. .High reactor power, high fuel temperature, loss of detector high voltage, or short reactor period will automatically cause the control rods to be dropped into the reactor core. The reactor operator may manually scram the reactor as well by means of a scram bar on the console. If beam ports are in use in their open configuration, an additional reactor scram may be added to reduce radiation levels ifpcrsonnel attempt to enter the beam area.

During steady state operation, the high reactor power scram, the high 'voltage power supply failure scrams for all neutionic channels; and the manual scram are requircd for operation.

Although the period scram is normally in operation, it can be bypassed provided that the reactor operator calculate reactor period for each rod movement and that the calculated reactor period is greater than one second. The fuel temperature scram is active, but it is not necessarysince the fuel temperature setpoint (usually set for 400'C) will normally not be reached during steady state mode.

In pulse mode, the mode selector switch is set to the HI PULSE position, interrupting detector signal to the linear channel. When the pulse interlock is activated (to initiate the source interlock) to prevent withdrawal of the shim, safety and regulating rods, the detector signal to the logarithmic wide range detector is interrupted. Consequently, the period scram and the linear high power scrams are disabled.' High voltake scrams for the linear and log channels are also disconnected as large pulses produce detector currents that may temporarily overload the power supplies. The percent power channel increases its range to the pulse range (the percent power channel high voltage power supply scram remains active). The uncompensated ion chamber for the percent power channel is positioned further away from the reactor core, allowing for measurements of both steady state and pulsing power levels without excessive current from the high voltage supply.

7.5 Engineered Safety Features Actuation Systems There are no engineered safety features. actuation systems. Control rod insertion is provided by gravity and core cooling is provided by natural convection in water or air. Therefore, ESF systems are not required in this design.

7.6 Control Console and Display Instruments The control console and display instruments are shown in Figures 7.2-7.5, 7.9, 7.10 and 7.12.

Their layout (Figure 7.11) provides a single operator with all relevant reactor information. All push buttons required for general operation are located on the control console, within easy reach by a seated operator. All instruments on both the console and instrumentation rack are visible K-State Reactor 7-17 Original (5105)

Safety Analysis Report

CHAPTER 7 from this seated position. In addition to the various analog displays, a computer display on top of console can also be used to show relevant reactor status information on a single screen.

There are several additional pieces of equipment in the control room.

-_ _ o -o interrupt power to electrical devices in the control room and reactor bay (see Figure 7.13a). A halon fire extinguisher is located next to the breakers for use in fighting electrical fires. Current core and facility configuration is shown in a display cabinet (Figure 7.13.b). _ _ r

_ _k A local radiation area monitor (including indicator a3d alarm) is located above the door in the control room to the reactor bay.

J 7.7 Radiation Monitoring Systems Radiation monitoring systemiis arc employed throughout the reactor facility: G-M detectors at the reactor pool surface and cleanup loop, 7 remote area monitor channels (3 general area or process monitors, 4 channels for beam ports - 11 beam port channels is currently instrumented, with the remainder scheduled for instrumentation near term), a 5 RIh' evacuation alarm, several air activity monitors, and numerous portable radiation monitors, including those for contamination monitoring. Additionally, an independent monitor with visual and audible alarms is located above the door to the reactor bay.

Figure 7.9, Control Room Overall View.

I.-

K-State Reactor 7-18 Original (5105)

Safety Analysis Report I.-

/

INSTRUMENT AND CONTROL SYSTEMS 9 .

F  : - - WA.W i -X-Figure 7.10, Control Console Operator's View.

L Figure 7.12, Instrumentation Rack Operator's View.

K-State Reactor 7-19 Original (5105)

Safety Analysis Report

CHAPTER 7 Figure 7.13a, Control Room Behind Operator; Figure 7.13b, Core Map.

The radiation monitor in the instrumentation rack provides an indication of radiation levels directly above the primary pool water surface. A meter on the right-hand section of the rack indicates a radiation dose rate of 100 mR hX full scale, utilizing a GM tube detector measures coolant activity in the cleanup loop water box. The water box monitor has an audibic alarm, with a reset button.

The remote area monitors utilize G-M detectors located throughout the reactor bay (typical unit illustrated in Figure 7.14). Permanent locations are: at the top of the reactor tank, above the bulk shield tank, and near the ion exchanger in the primary coolant system, and directly over the primary water tank. Each beam port has signal and power lines to support installing a beam port monitor. The detectors feature an analog readout in both the control room and locally with visual indicators for normal, alert, and alann conditions. The control room alarm has an audible signal as welL K-State Reactor 7-20 Original (5105) I-Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS Figure 7.14, Typical Area Monitor Instillation.

A 5 Rh l monitor on the 22-foot level serves an evacuation alarm. The alarm signals a I 00D+ dB audible alarm in the reactor bay. Both the monitor and alarm are supplied by battery power in the event of a power failure.

There are numerous ion chamber and G3-M portable survey instruments through the reactor bay and control room for gamma and beta surveys. These instruments arc calibrated semi-annually.

A rem-ball and a Bonner Sphere set are located in the reactor bay for neutron measurements. The rem-ball is calibrated annually. Low-level G-M counters in the reactor bay used for contamination monitoring. These are calibrated annually. Neutron and gamma sensitive pocket ion chambers are available in the control room for tours and personnel monitoring, and are calibrated semi-annually. Film and ring badges are issued to regular staff, which are read monthly and quarterly respectively.

An air monitoring system (Figure 7.15) samples air over the reactor pool. These monitors should sense any changes in radioactive discharge from the reactor pool to the environment. Channels are provided for monitoring particulate, noble gas and iodine activity.

An independent air monitoring system, a continuous air monitor (Figure 7.16), is stationed on the 12-foot level. This monitor should sense any changes in airborne contamination in the reactor-bay. The continuous air monitor has active background discrimination. One channel monitors radioactive contamination on a filter exposed to airflow from the reactor bay. A second channel monitors background radiation levels for a background subtraction Figure 7.15: Figure 7.16: Continuous Air Monitoring System Air Monitor K-State Reactor . 7-21 Original (5105)

Safety Analysis Report

CHAPTER 7 K-State Reactor 7-22 .Original (5105)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS 7.8 Bibliography 250-kJY TRIGA Mark il PulsingReactorMechanicalMaintenanceand OperationManualfor KSU, General Atomics, GA-3399; 1 August 1962.

NMP-1000 LinearPower Channel: OperationandMaintenanceManual Gcncral Atomics,, E-117-1017, Rev. 1; 1991.

NLiW-1000 Log Power Channel: Operationand MaintenanceManual, General Atomics, El 17-1019; 1992.

NPP-1000PercentPower Channel: OperationandMaintenanceManual. General Atomics,,

El 17-1010, Rev. 2; 1992.

TRIGA Mark]ReactorinstrumentationManualfor USGS, General Atomics, GA-9039; I November 1968. (For console acquired from USGS)

TRIGA Mark! PulsingReactor InstrumentationMaintenanceManualforKSU, General Atomics, GA-3114, no date.

I .

K-State Reactor 7-23 Original (5105)

Safety Analysis Report

13. ACCIDENTANALYSIS This chapter provides information and analysis to demonstrate that the health and safety of the public and workers are protected in the event of equipment malfunctions or other abnormalities in reactor behavior. The analysis demonstrates that facility design features, limiting safety system settings, and limiting conditions for operation ensure that no credible accideht could lead to unacceptable radiological consequences to people or the environment.

13.1 Accident Initiating Events and Scenarios This chapter deals with analysis of abnormal operating coniditions and consequent effects on safety to the reactor, the public, and operations personnel. Three conditions to be analyzed are:

• Loss of coolant

• Insertion of excess reactivity

• Fuel encapsulation failure - the maximum hypothetical accident (MHA)

These are the three conditions considered in the initial licensing of the Reactor Facility in 1962 for 100-kW steady-state operation and in the 1968 upgrade of the license permitting 250-kW steady state operation and 250-MW pulsing operation. The .analysis presented here treats the same 'conditibns, but for'steady-state operation at'1,250 kW and pulsing operation to a S3.00 reactivity insertion, estimated peak power of 1,340 MW.

The maximum hypothetical accident for a TRIGA reactor is the failure of the encapsulation of one fuel clement, in air, resultirg in the release of gaseous fission products to the atmosphere.

Failure in air could result from a fuel-handling accident or, possibly, failure in the event of a loss of reactor coolant. Failure under water, leading ultimately to atmospheric release of fission products, could possibly result from insertion of excess reactivity or operation with damaged fuel.

This chapter addresses the several scenarios potentially leading to fuel failure, and then the potential consequences, should failure occur in air.

13.2 Accident Analysis and Determination of Consequences 13.2.1 Notation and Fuel Properties Tables 13.1-13.3 identify physical characteristics of the TRlGA Mark 11.fuel. Table 13A identifies the assumptions and design basis values used in the accident analyses.

K-State Reactor 13-1 *Original (5105)

Safety Analysis Report

CHAPTER 13 Table 13.1, Dimensions ofTRIGA AMklI ZrIH,.6 Fuel Elements.

Property of Individual Element Symbol Value Length of fuel zone L 0.381 m Fuel radius r, 0.018161 m Clad outside radius r. 0.018669 m Fuel volume .r Clad volume VI 0.0000224 m3 Fuel mass Mf 0 Clad mass ,Uf 0.1845 kg WL Fraction U in fuel . x,, l WL Fraction ZrH,.6 in fuel xM Source: Training Manual, KSU TRIGA Nuclear Reactor Facility, 1998.

a Table 13.2. Ncutronic Properties of TRIGA Alkli ZrHj.6 Fucl Elements.

Property Symbol Value Effective delayed neutron fractions p 0.007 Effective neutron lifetime t 43 psec Temperature coefficient of reactivity a -0.000115 K' Source: West ct al. (1967).

Tablc 13.3, Thermal and Mcchanical Properties of TRIGA Mkll ZrHA Fuel Elements and Type 304 Stainless Steil Clidding.

Property Symbol Value Temp.

Fuel Density pl 5996 kg m7 Thermal conductivity kr 18Wm 'K All Heat capacity, cpr= 340.1 + 0.6952T(°C) cpr 340.1 J kg" K7' 0°C Cladding Density PC 7900 kg rnf 300 K Thermal conductivity 14.9 W mrnK' 300 K 16.6 400 K 19.8 600 K Heat capacity Cr 477 I kg7' K' 300 K 515 400 K Yield strength 250 Mpa 400 °C Tensile strength 455 Mpa 400 °C Source: fuel properties from Simnad (1980); cladding properties from Incropera and DeWitt (1990) and from Metals Handbook (1961).

4 is K-State Reactor 13-2 Safety Analysis Report Original (5105) 4 4

.4

ACCIDENT ANALYSIS /

Table 13A, KSU TRIGA Core-Conditions Basis for Calculations.

Steady state maximum power, P,

• 1,250 kW Fuel mass per element 2.367 kg Heat capacity per element at T('C) 805.0 + 1.646T(J K')

-Minimum number of fuel elements,N Corc radial peaking factor 2 Axial peaking factor i/2 Excess reactivity S4.00 (2.8% AkMk)

Maximum pulsing reactivity insertion S3.00 (2.1% Aklk)

Excess reactivity at 500 kW maximum power' SI.16 (0.81% Ak/k)

Fuel average temperature at 500 kW maximum power' . 285 eC

'Source: Data from GA Torrey Pines TRIGA reactor 13.2.2 Loss of Reactor Coolant Although total loss of reactor pool water is considered to be an extremely improbable event, calculations have been made to determine the imaximum fuel temperature rise that could be expected to result from such an event taking place after long-term operation at full power of 500 kW. Limiting design basis parameters and values are addressed by Simnad (1980) as follows:

Fuel-moderator temperature Is the basic limit of TRIGA reactor operation. This limit stems from the out-gassing of hydrogen from the ZrH and the subsequent stress produced In the fuel element clad material .The strength of the clad as afunction of temperature can set the upper limit on thefuel temperature. A fuel temperaturesafety limit of 11501 Cforpulsing.stainless steel U-ZrH,.gs ...fuel s used as a design valu to preclude the loss of clad Integritywhen the dadtemperatureIs below 500 C When dad temperaturescan equal thefuel temperature, thefuel temperature limit is 950 C There is also a steady-state operationalfuel temperature design limt of 750 IC based on considerationofIrradiation-andfission-product-lnducedfuelgrowth anddeformation..

As this section demonstrates, even- under extraordinarily conservative assumptions and approximations, the maximum fuel temperature reached in a loss of coolant iccident is 2901C, well below any safety limit for TRIGA reactor fuel. Conservatism notwithstanding, the margin between computed temperature and design limiti is sufficiently great to accommodate a design margin of at least a factor of two.

a. Initial Conditions, Assumptions, and Approximations The following conditions establish an extremely conservative scenario for analysis of the loss of coolant accident.

• The reactor is assumed to have been operating for infinite time at power P. 1,250 kW at the time coolant is lost.

• Coolant loss is assumed to be instantaneous.

• Reactor scram is assumed to occur simultaneously with coolant loss.

K-State Reactor 13-3 Original (5105)

Safety Analysis Report

9 CHAPTER 13

• Decay heat is from fission product gamma and x rays, beta particles, and electrons. I Effects of delayed neutrons are neglected.

• Thermal power is distributed among N =&]fuel elements, with a radial peak-to-average ratio of 2.0. In individual elements, thermal power is distributed axially according to a sinusoidal function.

• Cladding and gap resistance are assumed to be negligible, i.e., cladding temperature J is assumed to be equal to the temperature at the outside surface of the fuel matrix.

• Cooling of the fuel occurs via natural convection to air at inlet temperature To J 3001K. Radiative cooling and conduction to the grid plates arc neglected.

• Heat transfer in the fuel is one dimensional, i.e., axial conduction is neglected, and '

fuel is assumed to be uniform in thermophysical properties. x

• Heat transfer in the fuel is treated on the basis of pseudo-steady-state behavior, i.e., at any one instant, heat transfer is described by steady-state conduction and convection J equations.

b. Core Geometry 1 The following data on core geometry are derived from the KSU TRIGA Mechanical 1 Maintenance and Operating Manual (1962). The core containsi uel positions in five J circular rings (B - F), plus the central thimble (A ring). The upper grid plate is 0.495 mi in diameter and 0.0 19 m thick. Holes to position the fuel are 0.03823 m diameter and the central thimble is very slightly larger in diameter, 0.03 84 m.

Cooling water passes through the differential area between the triangular spacer block on the top of each fuel element and the round holes in the upper grid plate. The nominal diametral clearance between the tips of the spacer blocks and the grid plate is .

approximately 0.001 m.

The lower grid plate is 0.405 m diameter, with 36 holes, 0.0159 m diameter, for water l flow. However, the bulk of the water flow is through the annular space provided between the top of the lower grid plate and the bottom of the reflector. The radial reflector is D, 0.457 m inside diameter and 0.559 m height. 1l The effective hydraulic diameter for flow through the core, with an experiment in place in the central thimble, is given by AD 4*A A =

4 *(. .

= +4

-91*1r*r 2 )

= 0.02127in (Equation 13.2.2.1) ]

4 P. Jr *D, + 91 *2*z *r,<l See Todreas & Kazemi (1990) or EI-Wakil (1971) for steady-state conduction equations.

K-State Reactor 134 Original (5/05) 4 Safety Analysis Report 4

.41

ACCIDENT ANALYSIS If, in thermal-hydraulic calculations, one approximates conditions as flow through an annular section around any one fuel element, the outer. radius of the annulus, say r,, is given by D, = )2+r.

, or (Equation 13.2.2.2) r, = fD, r, I 2 i r.' =0.02 339 m, The flow arca.A4 per fuel rod is jr(r.2- r ) = rnD /2 = 0.000623.8 W.

The total length of a fuel rod is 0.7206 m (28.37 in.), of which the length of the fuel matrix, the heated length, is L1 = 0.3810 (15 in). The lengths of upper and lower axial reflectors, L. and L, are each 0.0874 mn (3.44 in). Beneath the lower reflector is a bottom end fixture of length L, about 0.0824 m (3.245 in.). Above the upper reflector is a triangular spacer of length L, about 0.0191 m (0.75 in.) and an upper end fitting of length L. about 0.0634 m (2.495 in.). Thc zone between grid plates is

= ++L 1 +L, =0.6382 m.

=L~,

C. Decay Power The time dependence of the thermal power in the core as a function of time after shutdown is based on calculations by the-CINDER code [England et al. 1976] as reported by George, LaBauve, and England (1980, 1982]. Sample results are presented in Figure 13.1 and Table 13.5 as the function R(r) defined as the ratio of the thermal power Pd(t) from gamma ray and beta particle decay at time. t after shutdown to the steady power P. prior to shutdown, based on 200 MeV energy release per fission.

For the purpose of this analysis, the time dependence function for I to Io6 s may be approximated as 0.04856 +A 0.1189x - 0.01 03x2 + 0.000228x' (Equation 13.2.23)

R(1) I +2.5481x-0.19632xi'+0.05417x'

• in which x is the natural log of the time after shutdown, in seconds. Time dependence of the thermal source in the worst-case element is given by 2P (Equation 13.2.2.4)

N in which the worst case element generates twice the power'as the core avirage, P. is the total-core thermal power prior to shutdown (1,250 kW); and N ())s the minimum number of fuel elements required for operation.

K-State Reactor 13-5 .Original (5105)

Safety Analysis Report

CHAPTER 13 MO° 30 102 nz3 14 10 106 17 roe (s)

Fi-ure 13.1 Decav Heat Function R(A). 2 Table 13.5, Decay Heat Function for Thermal Fission of IU.

Time t (s) R(t) = Pd(t)1Po 0 0.0526 1(10(? 0.0486 10 (10 0.0418 100 (10 ) 0.0282 1000 (10 ) 0.0172 10,000 (10w) 0.0087 100,000 (104 ) 0.0044 1,000.000 (10 ) 0.0025

d. Maximum Air Temperature 4 The fundamental relationships between buoyancy driven differential pressure and pressure losses from friction provide an independent verification for results of the previous calculations.

Buoyancy Driven Pressure Difference 4 The total mass flow rate w (kgls) associated with the worst-case fuel element is determined by a balance between the buoyancy driven pressure difference vertically across the core and the frictional pressure loss within the core, which is discussed in the next section. The temperature rise across the core is AT.;(t) = MO ~) _ 2.M<. (Equation 13.2.2.5) a J

K-State Reactor Safety Analysis Report 13-6 Original (5/05) a J

J

ACCIDENT ANALYSIS

.in which the heat capacity of the air evaluated at the inlet air temperature. At time zero, for example, AT,(O) =7T(0)-2T 0.6153Iw. (Equation 13.2.2.6)

Air inlet temperature To is assumed to remain constant at 270 C. Suppose pi and p" arc respectively the densities of air at the inlet and outlet temperatures. Suppose further that the effective chimney height is H. The chimney height is the distance between the center of the zone in which the air is heated and the center of the zone in which the air is cooled.

Evaluation of the latter is difficult to determine because of uncertainties in mixing of the air after it leaves the upper grid plate. Here we follow the lead of the UT SAR (1991) and choose 10 hydraulic diameters as the effective distance. Thus, H is given by Lf1 2 +

L, +1ODh,= OA22 m. and the buoyancy pressure difference is given by ap6 = (Pi . Po)*g*H (Equation 13.2.2.7) in whichg is the acceleration of gravity, 9.8 m s 2. Since

-353 *(T. - Tj)

(Equation 1322.8) and TI=300K. it follows from Eqs. (13.2.2 5) and (13.2.2-8) that 4p(t)= 0.19OR(t)Iw, (Equation 13.2.2.9)

Frictional Pressure Difference In this calculation, only frictional losses within the core, computed on the basis of the equivalent annulus model, are accounted for. Based on air inlet density and an air mass flow rate per fuel rod of wv, the frictional pressure difference is given by (Equation 132.2.10)

O.. 2p#DA, The laminar-flow (Moody) friction factor f for the equivalent annulus model, with r I r, *-1.193 is given by Sparrow and Locffler (1959) as f=100/Re, (Equation 13.2.2.11)

Re, the Reynolds number, is given by DhwlpA,, and p1 is the dynamic viscosity of the air at the inlet temperature? Equation (1322-10) nay be rewritten as Density at 1 atm, for air as an ideal gas, isgiven by p (kgrn') 353.017°K). Heat capacity, from 300 to 700 "K is 1030 J/kgK+/-: 3%(Incropera and DeWitt, 1990). .

Dynamic viscosity, over the range 250- 1000 0K is given by i0'gi (Ns/m2) --106.2941 +

16.81986[7(0K)]"1 or pt 1.85 x 10 'sNs/m at 300 'K. (Inciopera and DeWitt, 1990).

K-State Reactor 13-7 Original (5/05)

Safety Analysis Report

CHAPTER 13 0.1416Lp,5 7w = 1780iv, (Equation 13Z2.12)

Ap= D2AC Equating the frictional pressure drop with the buoyancy pressure driving force, using i Equations (13.2.2-9) and (13.2.2-12),

1j or w = 0.0l03Tfli, (Equation 13.2.2.13) aj r(t)- , =AT(t) = 1140.,J. (Equation 13.2.2.14) aj Results a For R(t) at time zero of 0.0526, maximum air temperature rise above 300 K is 261 K.

j

e. Fuel and Cladding Temperature Distribution With design power P. = 1,250 kW, a factor of two radial peak to average power, and a fuel surface area in the heated zone equal to A. = 2nrLf = 0.04469 m2, the worst-case average heat flux at post-accident time t in the heated zone is:

a q= 2*1.25e6*R(t)IAN = 3.37e5 W/m2 (13.2.2-15)

J With the conservative approximation that the axial variation of heat flux is sinusoidal, the I local value of the heat flux (W/m 2 ) is given by: j~

q"(z) = q".,, sin(= / L>), (13.2.2-16) in which z is the distance along the fuel channel, measured from the inlet and q"MU =(orf 2)q".t = 4.235 x10' R(t) W/m2 . Similarly, the local value of the air temperature in the coolant channel is given by T .(Z) =g + 0.6153R(t)[I cos(nz / L.)]

(13.2.2-17)

According to Dwyer and Berry (1970), the Nusselt number for laminar flow in a cooling channel is approximately Nu = 4.24. The corresponding heat transfer coefficient is h= k-u (13.2.2-18) hDi By using as an approximation the air thermal conductivity' of 26.3 Wn/mK at 3001K, one computes h = 5240 W/m2 K, and the cladding surface temperature 4

4.

4 For the range 200 to I 000K, data of Incropera and DeWitt (1990) is very well fit by the formula k,, =-22.055+2.8057 7(iK) in units of W/mK.

J

.J K-State Reactor 13-8 Original (5/05)

Safety Analysis Report

ACCIDENT ANALYSIS T,,(z)= T,(z)+q"(z)Ih- (13.2.2-19)

By using the fuel thermal conductivity kf = 18 W/mK, and neglecting the temperature drop across the cladding, one computes the fuel centerline temperature as TI(Z) = T,( +s (2k 1 (13.2.2-20)

Fuel and cladding temperatures arc reported in Table 13.6 and illustrated in Figure 13.2 for the case of zero time post accident. This is based on three conservative assumptions:

equilibrium fission product buildup at full power, instantaneous loss of coolant when the reactor scram occurs, and equilibrium temperature based on initial decay heat production.

Table 13.6, Post-Accidcnt Fuel and Cladding Temperatures.

ge (W/m 2 ) ITel (OK) Tw (K) Tf (°K) 0.00 0

• 300 300 300 0.10 17209 306 309 318 020 32733 325 331 348 030 45053 354 362 385

...0.40 52963 .-390 . .400 427 0.50 55688 431 441 470 0.60 52963 471 481 508 0.70 45053 ,508 516 539 0.80 32733 536 542 559 0.90 17209 . 555 558 567 1.00 0 561 561. *561 The K-State reactor does not operate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 7 days per week, so that the total inventory of fission products is significantly lower than the assumed value. During actual loss of coolant, overall heat transfer coefficient will be based on water rather than air.

There is at least. 16 feet of w~atei over the core that has t'drain before the core is uncovered. Experiments reported in GA-6596."Siniulated Loss-of-Coolant Accident for TRIGA Reactors", (General"At6inkis August 18, 1965) demonstrated that with a constant and continuous heat production; temperat ecrises to about 50%e of the equilibrium temperature in appioximately 30 iminutei (1800 seconds). At 101 seconds, R(t)'is 0.0172, 33% of the heat production following shutdown. 'Equilibrafion takes a significait amount of time; while beat productioni is decaying.

K-State Reactor 13-9 Original (5/05)

Safety Analysis Report

CHAPTER 13 000*-. I -.- I 15-Fud I-,_

N S 450 4M4 I

i ao I loo l

-. L 014I 020 0.50 040 ga &M ar a to too Figure 4.13, Axial Variation or fuel, cladding, and air temperature Immediately following a loss of coolant accident, with equilibrium fission product heating.

J

f. Radiation Levels from the Uncovered Core J Although there is only a very remote possibility that the primary coolant and reactor shielding water will be totally lost, direct and scattered dose rates from an uncovered core following 1,250 kW operations have been calculated.

This section describes calculations of on-site and off-site radiological consequences of the loss-of-coolant accident. Extremely conservative-assumptions are made in the calculations, namely, operation at 1,250 kW for one year followed by instant and simultaneous shutdown and loss of coolant. The reactor core is homogenized and the ORIGEN-2 [CCC-371] is used to define gamma-ray source strengths, by energy group.

Radiation transport calculations are performed using the MCNP code (Briesmeister 1997), with doses evaluated at the 22-ft elevation of the reactor (one meter above the operating deck), in the reactor bay at the 0-foot elevation (one meter above the floor, 30-cm from the outer wall, and at one-meter above ground level at the site boundary and at points beyond the site boundary extending to 100 m radial distance from the core.

Modeling of the reactor core for radiation transport calculations is based on the following approximations. The TRIGA reactor core is approximated as a right circular cylinder 0.4572 m diameter (OD of F ring). The fuel region is 0.381 m (15 in.) high. On each end axially is a graphite zone 0.0874 m (3.44 in.) hih and an aluminum grid plate 0.0191 m (0.75 in.) thick. Inlifuel locations, there areafiucl elements, 3 standard control rods and 1 transient control rod, I void location, 1 central thimble (void), 1 source (assume void), and I pneumatic transfer site (assume void). The fuel region may is treated as a homogeneous zone, as are the axial graphite zones and the grid plates.

J J

K-State Reactor 13-10 Original (5105) J Safety Analysis Report J

ACCIDENT ANALYSIS Fuel elements are 1.43-in. ID and I.A7-in OD, clad with type 304 stainless steels. Fuel density is 5996 kgnm 2. Fuel composition is&Juraniumft0ZVHI.6s. The uranium isl sU andM3 U. Steel density is 7900 kg/r 3: Standard control rods are 0.875-in. OD, the transient rod 1.25-in. OD. Boih types of rods are cladvwith 30-mil thick aluminum (2700 kg/rn density). The control material may be approximated as pure graphite, with density 1700 kg/r 3 .

In radiation transport calculations, the core is modeled .conservatively as a central homogenous fuel zone (air density neglected) bounded on either end by a homogeneous axial reflector zone, and by a 0.75-in: thick aluminum -grid plate, treated as a homogeneous solid. Densities of the homogenous zones'are as follow:'

Fuel 3602 kg/rn X Reflector 1147kg/m3

_ . Grid Plate 2700 kgkn' Composition of the three zones, by weight fraction, are given in Table 13.7.

I Table 13.7, Compositions orHomogenized Core Zones.'

Element* Mass Fraction Element  : Mass Fraction Fuel Zone AxialReflectorZone C 0.0617 C 0.7920 Al 0.0010 Al 0.0033 H 0.0139 . Mn 0.0041 Zr 0.7841 Cr 0.0368 Mn 0.0013 Ni 0.0164 Cr 0.0117 Fe 0.1474 Ni 0.0052 Fe 0.0469 . . GridPlate U 0.0741 Al 1.0000 The reactor bay is approximated as a right circular cylinder 36 ft (10.973 m) high and 36.68 ft (I 1.18 m) radius. The reactor vessel structure is approximated as a right circular cylinder, co-axial with the bay, 22 ft (6.706'm) high'and 11 f (3.3528 m) radius. The free volume is 144,000 ft3 (4078 mn). The site boundary, at its nearest approach to the reactor bay, is about 2 m beyond the bay boundary, that is, at a radius of 13.13 m from the center of the reactor.

Gamma-ray source strengths vs. times after shutdown are listed in Table 13.8 for an operating time of one year and a thermal power of 1,250 kW. The source strengths are gamma rays per-second, by group, for the entire reactor core. In the MCNP calculations, the source is assumed to be uniformly distributed within the core.

The roof of the reactor bay is modeled as a concrete slab 10 cm thick and with density 2.35 g/cm'. In fact, the roof is a composite structure, so reflection from the'concrete slab conservatively models gmma-mray transport and dose rates at the site boundary.

2 Composition, by weight, 2% Mn, 18% Cr, 8%Ni, balance Fe.

K-State Reactor Original (5105)

Safety Analysis Report

J CHAPTER 13 J Table 13.9 reveals substantial dose rates above the void reactor tank with long-term 500-kW operation. One day after loss of coolant, the dose rate is 17 Svlh (1700 rem/h).

However, dose rates inside the reactor bay at the 0-ft. and 12-ft. elevations are only about I 4 rezn-h immediately after coolant loss, declining to about 1.5 rem/h after I hour and 100 znrem/h after 30 days. These dose rates show that personnel could occupy the reactor bay shortly after the loss of coolant to undertake mitigating actions without exceeding NRC occupational dose limits.

Dose rates at other receptor locations are shown in Table 13.9 as indicators of areas that '1 should be checked following a loss of reactor pool water event. These dose rates are J unrealistically conservative in view of the assumption of one-year operation at full power (not possible) and scattering from a thick concrete roof. Dose rates at the site boundary (facility fence) decline from about 3 rem/h immediately after the loss of coolant to 350 J mrem/h after 1 day. Kansas State University has complete control over access to campus locations within, the zones defined by receptor locations in the analysis.

Table 13.8,. Full Core Gamma-Ray Sourecs Strengths (Number Per Second) Following Operation for One Year at 1,250 kW Thermal Power.

Time after shutdown E (McV) 0 Ih 24 h 30 days 180 days LOOE-02 8.72E+I5 2.92E+IS 1.05E+I5 2.43E+14 6.20E+13 2.SOE-02 2.09E+15 7.SOE+14 3.04E+14 5.97E+13 132E+13 J 3.7SE-02 1.74E+15 7.37E+14 3.97E+14 6.72E413 1.45E+13 S.75E-02 1.81E+IS 5.50E+14 1.77E+14 4.1SE+13 1.20E+13 1 8.SOE-02 1.50E+IS 5.20E+14 2.33E+14 3.21E+13 8.39E+12 1.25E-01 I.SOE+IS 7.65E+14 4.81E+14 8.98E+13 1.31E+13 2.25E-01 3.1 IE+15 1.03E+15 4.44E+14 2.42E+13 6.83E+12 3.75E-01 I.94E+IS 5.72E+14 2.20E+14 3.81E+13 3AIE+12 3 5.75E-01 3.25E+15 1.71E+15 7.S4E+14 1.01E414 1.08E+13 J 8.50E-01 4.16E+15 2.26E+IS 8.94E+14 4.51E+14 1.15E+14 1.25E400 2.22E+15 8.15E+14 7.83E+13 1.62E+12 6.0IE+III 1.75Et00 9.30E+14 5.14E+14 2.23E+14 4.78E+13 7.72E+10 2.25E+00 5.OOE+14 2.19E+14 6.69E+12 1.12E+12 3.72E+ I I 2.75E+00 1.99E+14 7.32E+13 8.13E+12 1.82E+12 1.55E+09 3.50Et00 1.16E+14 1.72E+13 7.12E+10 1.50E410 9.37E+07 J; S.00E400 6.23E+13 2.42E+II 6.96E+08 4.67E400 4.63E+OO 7.00E+00 S.OSE+I I 5.27E-01 5.27E-01 5.26E-01 5.22E-01 9.50E+00 9.49E407 5.98E-02 5.98E-02 5.98E-02 5.93E-02 1' Total 3.39E+16 135E+16 5.27E+15 1.20E+15 2.61E+144 MCVls 1.43E+16 6.28E+15 2.02E+1S S.77E+14 1.13E+14 J3 K-State Reactor . 13-12 Original (5105) '

Safety Analysis Report

ACCIDENT ANALYSIS Dose rates vs. post accident times are listed in.Table 13.9 for a number of receiver locations. Locations are specified at~radial distances from the center of the reactor bay.

The22-foot level provides-direct access to the reactor pool, the 12-foot level access to valves controlling water flow to the reactor pool. The 0-foot level provides access to valves and pumps, which allow water to be added to the reactor pool. The 13-metcr distance marks a zone defined by the fence surrounding the reactor bay.

g. Conclusions Although a loss of pool water is considered to be an extremely improbable event, calculations show the ma'ximum fuel temperature that could be expected to result from such an event (after long-term operation at full power of 1,250k) is 2940 C, well below any safety limit for TRIGA reactor fuel.

Maximum possible dose rates resulting from a complete loss of pool water permit mitigating actions. The area surrounding the reactor is under control of the Kansas State University, and exposures outside the. reactor bay environment.can be limited by controlling access appropriately. Kansas State University has complete'authorty to control access to campus locations. p a t Table 13.9, Gamma-Ray Ambient (Dcep) Dose Rates (Rbh) at Silected Locations for Times Following Loss or Coolant After Operation for One Year at 1,250 kW Thermal Power.

Time post accident 0 Ih 24h 30d 180d On-site (elev.)

22 t (center) 3.75E+06 138E+06 4.25E+05 1.03E+05 1.98E+04 12 R. (boundary) 1.08E+03 4.OOE+02 l.43E+02 3.25E+01 7.OOE+00*

0 fl (boundary 9.75E+02 3.75E+02 1.23E+02 3.OOE+01 6.25E+00 Off-site (radiusfrom center ofreactorbay) 13 in 6.7sE+02 2.75SE+02 8.75E+01 2.15E+01 4SOE+O0 15 mn 5.OOE+02 2.00E+02 6.75E+01

• 1.60E+01 3.25E+00 20 m 2.50E+02 1.00E+02 . 3.SOE+01 8.7SE+00 1.73E+00 30 m 8.25E+01 3.25E+01 1.10E+01 2.75E+00 50E-01 4Dm 3.75E+0l i.40E+0o 5.25.E+000 1.18E+00 2.33E-01 50m 1.93E+01 . 7.25E+00 .. 2A3E+00 6.25E-01 1.23E-01 70m . 6.50E+00 2.75E+00 9.00E1

• 2.18E201 4.00E-02 loom 2.35E+00 9.50E01 . .2.75E-01 6.75E-02. 1.33E-02

.-. . E Insero s Ri t ..

13.2.3 Insertion of Excess Reactivity K-State Reactor ,i13-13 Original (5/05)

Safety Analysis Report

CHAPTER 13 Rapid compensation of a reactivity insertion is the distinguishing design feature of the TRIGA reactor. Characteristics of a slow (ramp) reactivity insertion are less severe than a rapid transient 4 since temperature feedback will occur rapidly enough to limit the maximum power achieved during the transient. Analyses of plausible accident scenarios reveal no challenges to safety limits for the TRIGA. The fuel-integrity safety limit, according to Simnad (1980), may be stated as follows:

Fuel-moderatortemperature Is the basic limit of 7RUGA reactoroperation. This limit .

stems from the out-gassing of hydrogen from the ZrH, and the subsequent stress produced in thefuel element clad material. The strength of the clad as ofunction of temperaturecan set the upper limit on the fuel temperature. A fuel temperature safety limit of 1150 Cforpulsing. stainless steel U-Zr,.gj ... fuel is used as a design value to preclude the loss of clad integritywhen the clad temperatureis below 500 c When clad ..

temperaturescan equal thefueltemperature thefuel temperaturelimit is 950 t. ..

s1 Two reactivity accident scenarios arc presented. The first is the insertion of 2.1% reactivity at '

'3 zero power by sudden removal of a control rod. The second is the sudden removal of the same reactivity with the core operating at the maximum power level permitted by the balance of the core excess reactivity (i.e., core excess less $3.00). Movements of control rods for the first case a

are controlled, in part, administratively, while movements for the second are prevented by control circuit design. I As the analysis shows, in neither scenario does the peak fuel temperature approach the '1 temperature limit. The nearest approach is 8691C, incurred by a pulse insertion of 0.7% while the J.

reactor is operating at a steady power of 94 kWY, an action prevented both by administrative requirements and by interlocks.

a. Initial Conditions, Assumptions, and Approximations 4 The following conditions establish an extremely conservative scenario for analysis of '

insertion of excess reactivity. 4

• The reactor operates with a minimum of N4fuiel rods. '

• Reactor and coolant ambient (zero power) temperature is 27 0C. '

• Maximum reactivity insertion for pulsing or for the worth of experiments is set at Al S3.00, 6k, =2.1% orp,, 0.021.

• Reactor power equivalent to the core excess reactivity of S1.00, i.e., ok = 8k, - k,,'.

= 0.7% (p = 0.007) is P. = 107 kW and the maximum fuel temperature at that power J is T. = 1501C. Basis: Data for the Torrey Pines TRIGA, as included in the KISU '3, TRIGA Operations Manual.

• A control rod interlock preventing pulsing operations from power levels greater than 3 a maximum of 10 kW is not credited '3

• Conservative hot channel factors as calculated in 4.5.3 are used ,

4 K-State Reactor 13-14 Original (5105) 4 Safety Analysis Report

ACCIDENT ANALYSIS

b. Computational Model for Power Excursions The following relationships are for the Fuchs-Nordheim model, modified by Scalletar, for power excursions, as described for the TRIGA reactor by West et al. (1967).

If the heat capacity of the fuel is given by , cr= 340.1 + 0.69527'(Ct) (J/kg0 K), and there are N fuel elements, each of mass m1 = 2.367 kg, then the overall core heat capacity is given by K (J l K) = mNc,, = C.*+ T( C) (13.2.3-1) in which C. = 6.682 x 104 and C, - 136.6. IfP. (W) is the reactor power at the initiation of the pulse power excursion, and p. is the magnitude of an initiating step change in reactivity, then the maximum power increase (W) is given by (p-fi)2 C. (p-/J)'C,

- 2ae + 6a 2 * (132.3-2)

A maximum pulse of S3.00 would result in a power rise of approximately 1430 MW(t).

If T. is the average core temperature at the start of the excursion, the maximum temperature rise (0K) is given by T. -O,=[

T, = a' * [8* (a-1)8* 83 4a - 1)2 (13.2.3-3) in which aC.

ar= (A , (13.234) and the "+" sign applies when a> 1. Although there are nonlinear terms in the model, calculation of temperature change as a function of temperature shows a nearly linear response. Therefore the major factor in determining core-average peak temperature is the amount of reactivity available to pulse. To remain within 1000'C.at all locations, core.

average temperature cannot exceed 31 80C (based on Table 132.1.A peaking factors).

The maximum worth of the pulse rod is 53.00, therefore peak temperature following a pulse was calculated based on a maximum reactivity available-from the pulse rod at reference core temperature (270C). The core-averagc temperature change is 2291C, with a hot spot change of 716 0 C (based on Table 132.1.4 peaking factors).. Therefore, core-average peak temperature for a S3.00 pulse from 271C is 2561C, with-a hot spot temperature of 7461C.

• • With $3.00 reserved for. a maximum pulse; the reactor has remaining reactivity $1.00

• greater han ritical io support operation at power. Reactivity of Sl.0 allows operati6n 94 kW, 94 kW operation results in an average fuel temierature of 48 0 C, arid a hot-spot

• K-Sthte Reactor 13-15 Original (5/05)

Safety Analysis Report

CHAPTER 13 fuel temperature of 150'C. A $3.00 pulse increases this hot spot temperature from 1500 C to 8690 C.

Change in Peak Temperature Versus Pulsed Reactivity l1 I,- -I -- .a 1

a, 1.

.15 ISO-S ._

J I'

U U

U1 t S IJ

'SOS C..I MSO. - _ - _ ___ _ _ _ _ _ _ _ _ _ III a

j bog MO A.

ok0 4 1.0 10 too 2.50 300 150 400 45 dD00 Reactivity (S)

Figure 13.3, Change in Peak Temperature Versus Pulsed Reactivity Insertion The postulated scenarios do not result in fuel damage, but physical aspects of system prevent these scenarios from occurring. It is not possible to achieve full power operation with the pulse rod fully inserted; since the pulse rod is partially withdrawn with air applied to the pulse solenoid, it physically cannot be pulsed. Although not required to' ensure the safety of the reactor, an interlock prevents pulsing from power levels greater than a maximum of 10 kW.

c. Conclusions J1 Insertion of the maximum possible reactivity without initial temperature feedback (i.e.,

fuel temperature is too low to limit core available reactivity) results in a peak hot spot fuel temperature of 7461C, well below the safety limit.

Insertion of the maximum possible reactivity with initial temperature feedback (i.e., fuel temperature limits available) results in a peak hot spot fuel temperature of 8690 C, well below the safety limit.

s 13.2.4 Single Element Failure in Air Source quantities of radioactive noble gases and iodine are computed and tabulated for a maximum hypothetical accident involving cladding failure in a single TRIGA fuel element and the escape of the radionuclides into the environment. Two liriiting cases of operation are considered. 'For short lived radionuclides, source terms are computed for element failure subsequent to eight hours full-power operation per day for five days. For long-lived radionuclides, source terms are computed for element failure subsequent to continuous operation K-State Reactor 13-16 Original (5105)

Safety Analysis Report

ACCIDENT ANALYSIS for 40 years at the average power experienced by the reactor over its first 33 years of operation.

Also examined arc residual sources still present in fuel, but generated in reactor operations prior to local receipt of the fuel in 1973. Potential consequences of radiological releases are examined.

Even in the maximum hypothetical accident, no workers or members of the public are at risk of receiving radiation doses in excess of limits prescribed in federal regulations.

a. Assumptions and Approximations Following are assumptions and approximations applied to calculations.

1. For long-lived radionuclides, calculations of radionuclide inventory in fuel are based on continuous operation prior to fuel failure for 40 years at the average thermal power experienced by the reactor during its first 33 years of operation,

_* namely, 3.50 kW.

2. For short-lived radionuclides, calculations of radionuclide inventory in fuel arc based on operation at the full thermal power of 500 kW for eight hours per day, for five successive days prior to fuel failure, an average of 12.5 kMV-hr/day.

3. Radionuclide inventory in one "worst-case" fuel element is based onlmlements in the core for the historical period and elemcnts for full power operation, which is the case for core 1I-16,granis of 2 5U per element INUREG-2382],

_  :

• and a value of 2.0 as a very conservative value of the ratio of the maximum power in the core to the average power: Thus, for the historical priod, the worst cas clemwnt has operated at a thermal power of (3500Bx2 =LW , and for

'the one-wiek full-power operation, (500*x2 =M k%. .

4. The fraction of noble gases and iodine contained within the fuel that is actually released is 1.0 x 10-. This is a very conservative value prescribed in NULREG 2387 [Hawley and Kathren, 1982] and may be compared to the value of 1.5 x 10' 3 measured at General Atomics [Simnad et al., 1976) anid used in SARs for other reactor facilities [NUREG-1390,.1990].

5. hec fractional release of particulates (radionuclides other than noble gases and

-*iodine) is 1.0 x 10', a very conservative estimate used by Hawley and Kathiren

[1982].

b; Radionuclidc Invent6ry Buildup and Decay

-, Consider a mass of 233U yielding thermal 'power.P (kW) due to thermal-neutron induced fission. The fission ratcels related to the therrnal power bVy 'the factor k ' 3.12 x 1 0 t3 fissions per second per kW.' Consider also a fission product radionuclide, which is produced with yield Y, and which decays with rate 6onstant X. It is easily shown that the equilibrium activity A.,(Bq) of the fission product, which exists when the rate of creation by fission is equal to the rate of loss by decay, is given by A. = kPY. Here it should be

-' noted that the power must be small enough or the uranium mass large enough that the

'6Notc that the product ofk and yield Ymay be stated as 3.12 x 1013 x YBq/kW or 843 x YCi/kNV.

• K-State Reactor 13-17 Original (5105)

Safety Analysis Report

J -)

CHAPTER 13 Q depletion of the 2 "5U is negligible? Starting at time i = 0, the buildup of activity is given by J1 2(t)= A. *(I-eA) (13.2.4-1) J For times much greater than the half-life of the radionuclide, A 4A<, and for times much J less than the half-life, A(t)- A*

• t. If the fission process ceases at time ti, the specific activity at later time i is given by J A(t) =A,, * (i- e2A*) 'e~l('~') (13.4.2-2) J'.

Consider the fission product 131, which has a half-life of 8.04 days (a = 0.00359 hi) and J a chain (cumulative) fission product yield of about 0.03 1. At a thermal power of I kW, J the equilibrium activity is about A. = 9.67 x 10 Bq (26.1 Ci). After only four hours of J operation, though, the activity is only about 0.37 Ci. For equilibrium operation at 3.5 kV, distributed ovcrjfueI elements, the average activity per clement would be 26.1 x 3.5 IF81 1.13 Ci per fuel clement. The worst case element would contain twice this J activity. With a release fraction of 1.0 x 104, the activity available for release would be about 1.13 x 2 x 1.0 x 104 = 2.26 x 10'4 Ci. This type of calculation is performed by the ORIGEN code [CCC-371] for hundreds of fission products and for arbitrary times and J power levels of operation as well as arbitrary times of decay after conclusion of reactor 'l operation. The code accounts for branched decay chains. It also may account for depletion of 2"3U and ingrowth of 239Pu, although those features were not invoked in the calculations reported here because of minimal depletion in TRIGA fuel elements. J

c. Data From Origen Calculations J ORIGEN-2.1 calculation output files arc included as Appendices A and B; App. A I contains data for the buildup of long-lived radionuclides over the 40-year entire operating history of the KSU TRIGA reactor, modeled as 86.42 W continuous thermal power. App.

B contains data for the buildup ofrelatively short-lived radionuclides during a worst-casc '3, scenario modeled as 8-hours/day operations at 12.05 kW thermal power for five consecutive days. Tabulated results for Appendices A and B are pCi activities, by J

nuclide, immediately after reactor shutdown, and at 1, 2, 3, 7, and 14 days after J3, shutdown. In Appendix A, which deals with relatively long-lived radionuclides, data arc provided only for those nuclides present at activities greater than 100 mCi in a single fuel clement at I day after reactor shutdown. In Appendix B, which deals with relatively '.3 short-lived radionuclides, data are provided only for those nuclides present at activities greater than 100 mCi in a single fuel element immediately after reactor shutdown. In both '3 Appendices A and B, the activities per clement are multiplied by the release fractions previously cited, thus yielding maximum activities available for release in a maximum hypothetical accident. J 7Negligible bumup ismodeled in ORIGEN calculations by setting the fuel mass very large (I tonne) and <l, the thermal power very low (I kW or less).

• J3.

K-State Reactor 13-18 Original (5105) J Safety Analysis Report &

ACCIDENT ANALYSIS

d. Reference Case Source Terms

• Appendiccs A and B data for-worst case TRIGA fuel element arc compared;.greater

.values for any one isotope are selected as refcrence case source termns for maximum hypothetical accident. Data are presented in Table 13.10 for halogenslnoble gases, and

• ** Table 13.11 for particulate radionuclides.

e. Derived Quantities

• Thc raw data of Tables 13.10 and 13.11 are activities potentially released from a single

• worst-case fuel element that has experien6ed a cladding failure. This activitymay itself be compared to the annual limit of intake (ALI) to gauge the potential risk to an individual worker. By dividing the activity by the 144,000 R3 free volume of the reactor bay in the Nuclear Reactor Facility, one obtains an air concentration (specific activity) that may be compared to the derived air concentration (DAC) for occupational exposure as given IOCFn20 or in EPA federal guidance [Eckerman et al., 1988].

K-State Reactor 13-19 Original (5105)

Safety Analysis Report

CHAPTER 13

'I Table 13.10, Refrerencc Case Halogen & Nobic Gas Activities Potentially Released in Maximum Hypothetical Accident at KSU TRIGA Nk. 1i Nnuclear Reactor. It, Available activity (pCi) at time in days afler reactor shutdown Elcment Nuclide 0 I 2 3 7 14 28 Br 83 12110 13 0 0 0 0 Br 84 25388 0 0 0 0 0 0 I 131 8490 8245 7695 7130 5125 2810 840 I 132 23280 20943 16930 13685 5838 1318 68 I 133 69950 33530 15073 6773 278 0 0 I 134 192228 0 0 0 0 0 0 I 135 98750 7980 645 53 0 0 0 Kr 85 45 45 45 45 45 45 45 Kr 87 64498 0 0 0 0 0 0 Kr 88 79048 225 0 0 0 0 0 Kr 83M 9953 48 0 0 0 0 0 Kr 85M 23368 580 15 0 0 0 0 Xe 133 20363 24105 2i010 22390 14110 5673 895 Xe 135 58955 30518 6595 1195 0 0 0 Xe 138 158548 0 0 0 0 0 0 Xe 133M 35 33 28 20 8 0 0 Xe 135M 205 15 0 0 0 0 0 NOTE: Available activity (> 10 pC0) isfrom a single worst-casefuel element as afunction of time after reactor operation. Data are derived from ORIGEN 2.) calculations [CCC-371) as summarised in d Appendices A andB. Dataarermv computationalresults andthe number ofsignificantfiguresexceeds the precisionofthe calculation.

J-,

J J.

'j J.

3 K-State Reactor 13-20 Original (5105)

Safety Analysis Report J3

ACCIDENT ANALYSIS Table 13.11, Reference Case Particulate Activities Potentially Relcased In a Maximum HyDothetical Accident at the KSUTRIGA Mk. II Nuclear Reactor.

Availab]le activity (pCi) at time in days after reaEtorshutdov vn Element .. Nuclide 0 1 . 2 . 3* 7 14 . 28

' .. BA 139 1605 00*1 .. 0 - *. 00 0 0D. 0 BA . 140 12S .120 ... 115 IDS Is 60 2S BA 141 1480 0 0 0 0 0 0 BA 142 1470 0 0 0 0 0 CE 141 4' 53 53 50 45 40 30 CE

• 143 553 340 * :205 125 18 0 0 CS 138 1713 .0 .0. 0 0 0 0 LA 140 68 as *9S 103 95 63 33

• LA *141 1140 18 0* 0 0 ' 0 LA- 142 1455 0 . 0 0 0 0 LA 143 1493 0 0 0 0 0 0 HO 99 395 308 240 1as 68 13 0 MO 101 1273. 0 I*0 0 0 0 NBE 97 593 24S 93. 35 0 0 0 too NB 98 1463 0

• 0. . 0 o 0 0 ND .147 55 so 48 45 100 35 23 10

. ND 149 263 0 0* 0 0 0 0 ND 151 105 a 0 0 0 0 0 PH 151 40 23 13 3 0 PR 143 70 Ss 9S 90 65 33 PR 145 631 40 3 0 0 0 .0 PR 147 573 0 0 0 0 0 0 RB 88 300 0 0- 0 0 0 0~ 0 RB 89 1218 0 0 0 40 RH 105 73 65 ,- 40 25 S. 0 0 RH 107 40 0 0 0 0 0 0 RU 105 1*I 5I 0 0 0 .0 0 SE 81 53 0* 0 *0 0 0 SE

• 83

• 50 .0 *0 0 0 0 0 0 . 0 SN 128

• 83 0 SR 89 23 28 28 23 25 23 20 SR 91 795 133 25 5 0 0 0

• SR 92 1325. 3 0*. 0 0 0 0 TC 101 1273 0 00 - 0 0.

. TC .. 104 460 0 0 ~. 0 a TE 129 95 *3 0 0. 0 0 0 0 TE 131 623 -3

• 3. 0 .0 0 5E 132 250 203 165 1333 53 .. 13 0 TE 133 965 0 0 0 0 0 0 TE 134 1713 0 0 0 0 0 0 Y 92 343. *38 0 0 , 0 0 0 y 93 S63 170 33 B 0 0 0 Y 94 153 0 0 0 0 0 0 y 95 1613 . 0 0 0 0.- 0 y 91H 423 88 is 3 0 0 0 ZR 95

• 30 30. 30 25 28 25 23 ZR 97 65t ' 245 93 35 0 0 a0 Available activity (> 10 jCi) isfor astngle worstcasefuel element as afunction of time after

'reactoroperation. Daiaare dernvedfrom ORtGEN 2.1,clculationi(CCC-3 7)] asummr~ized in Appenkice.s A anctB *he table includes dnly th4Lse iruclides with activities In excess of 10 ,fiC.

Data'orerawcompuiationalresultsand the nsumbr ofsignificantfiguresexceeds the precision of the calculation. . - '

• K-State Reactor 13-21

• 13-21 Original (5105)

Safety Analysis Report

CHAPTER 13

r. Comparison with the DAC and the ALI The ALI is the activity that, if ingested or inhaled, would lead to either (a) the maximum kI permissible committed effective dose equivalent incurred annually in the workplace, J nominally 5 rem, or (b) the maximum permissible dose to any one organ. or tissue, nominally 50 rem. The DAC is the air concentration that, if breathed by reference man '3 for one work year (2000 h), would result in the intake of the ALl. ALI does not apply to noble-gas radionuclides.

Potential activity releases are compared to ALIs, and air concentrations in the reactor bay J are compared to DACs in Tables 13.12 and 13.13. Only for radioiodine does the available activity exceed the ALI. However, there is no credible scenario for accidental '

inhalation or ingestion of the undiluted radioiodine released from a fuel element. J When one compares with DACs the potential airborne concentration of radionuclides in Jo the reactor bay, only the 13'1, 1331, and 135 isotopes plus "Kr and "Kr are of potential '

consequence. However, annual dose limits could be attained only with a constant air concentration over a long period of time. The 1311 released in the failure of a single element, for example, would decay with a half-life of 8.04 days. Thus, even the undetected failure of a fuel element would not be expected to lead to violations of the I occupational dose limits expressed in IOCFR20 or in other federal guidance.

g. Comparison with the effluent concentration Effluent concentration, listed in the last columns of Tables 13.12 and 13.13, are defined in continuous exposure (8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year) rather than 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year ok occupational exposure. Exposure to a constant airborne concentration equal to the effluent concentration for one full year results in the annual dose limit of 100 mrem to members of the public. As is apparent from Tables 13.12 and 13.13, the reactor bay average concentrations immediately after fuel element failure exceed the effluent concentrations for several radionuclides. Thus, only for these radionuclides is it necessary to consider radioactive decay and atmospheric dispersal after release in estimating potential risk to members of the public. For posting purposes, concentrations relative to DACs are additive. For dosimetry purposes, products of concentrations and times, relative to DAC-hours, are additive.

h. Potential downwind dose to a member of the public In this dose assessment, it is assumed that the available activity in a failed fuel element is '3 released instantaneously and immediately after reactor shutdown. It is further assumed 3y that a member of the public is positioned directly downwind from the Nuclear Reactor Facility and remains in place during the entire passage of the airborne radioactivity. The very conservative approximations of Hawley and Kathren [1982] are adopted in the assessment, namely, that the atmospheric dispersion (X/Q) factor is 0.01 s/m3 (2.78 x 106 WiM3) and the breathing rate V is 1.2 m'/h. No credit is taken for partial containment, plateout, or other potential mitigating mechanisms, however realistic and probable.

Let the activity of nuclide i released be A (j+/-Ci) as given in Tables 13.12 and 13.13. If J one neglects radioactive decay, the activity inhaled during passage of the airborne K-State Reactor 13-22 Original (5/05)

Safety Analysis Report Uo

ACCIDENT ANALYSIS activity is A Y(x X/Q) [Faw and Shultis, 1993]. The product of the activity inhaled and the dose conversion factor 9 (mrem/pCi) [Eckerman et al., 1988] yields Di (mrem); the more critical of the organ dose.or the effective dose equivalent to the total body. Resul!ts of such calculations are.presented in Table 12.14. As is apparent from the table, individual organ doses as well as the total committed effective dose equivalent are well below any rcgulatory limits. Entries are shown only for doses of 0.001 mrem or greater.

.L Table 13.12, Comparison of Halogen and Noble Gas Avatlable Activities Immediately After Reactor Shutdown with ALls and Reactor Bay Concentrations with DACs and Effluent Concentrations.

Available Inzlto DAC Reactor '.EM~hcnt Ratio Ratio to Nuclidc Half-life activity bay co=C. colc. l D -

Elemenit 0) AU (jiCi) (pCilcmn) (pi/Ccm') (pcikm'~) DAC. Limit Br 83 2.39 h 12100 6.E+04 3.E-5 3.00E.06 9E-08 0.Q 33.3 Br 84 31.8 m 25375 6.E+04 2.E-5 62SE-06 8E-08 '03 78.1 1 131 8.04 d .8500 5.E+01 2.E-8 Zo8E-06 .2E-10 103.8 10375.0 1 132 230 h 23275 8.E+03 3.E-6 5.7SE06 2E 19 287.5 I..- 133 20.8 h 69950 3.E+02 L.E-7 1.73E.05 IE-09 172.5 M52s.o I . 134 52.6 m 192225 5.E+04 2.E-5 4.7SE.05 6E108 2. 791.7 1 135 6.61 h 98750 2.E+03 7.E-7 2.43E.05 6E-09 34.6 4041.7 Kr 83m 1.83 h 9950 1.E-2 2.45E.06 SE-05 o.o 0.0 Kr 85m 4A8 h 23375 2.E-5 5.75E.06 IE-07 03 57.5 Kr . 85 10.7 Y 50 IE-4 L_ 1.25E-08 7E-07 0.0 0.0 Kr 87 763 m 64500 . 5.E-6 1.S8E05 .2E-08 3.2 787.5 Kr 88 2.84 h 79050 2.E-6 1.95E.0S -9E-09 9.8 2166.7 Xe . 133m 2.19 d 0 IE-4 I.0oE-08 6E-07. 0.0 0.0 Xe 133 5.25 d 20350 I.E-4 S.OOE-06 5E-07 0.1 10.0 Xe 135m 15.3 m 25 9.E-6 625E-09 4E-08 0.0 0.2 Xe .135 9.09 h 58950 . 1.E-5 lA5E-05 7E-08 _I.5 207.1

'Bay concentration exceeds DAC. .

bBay concentration exceeds c~ucnt concentration.

K-State Reactor

• 13-23 Original (5105)

. . I Safety Analysis Report

i

-1 j

CHAPTER 13 i

'1 Table 13.13, Comparison of Particulate Available Activities (> 100 pCi) with ALIs and Reactor Bay Concentrations with DACs and Effluent Concentrations.

d Available I1nlation DAC Reactor bay Emuent R X ti Ii Element Nuclide Half-life activity (C) ALI (pCi) (pCilcm3)

Cilcn.

(pi/cmn)

Coc.

(PCi/cm3)

Rt DAC Limil Limit I

Ba 139 82.7 m 1600 75000 0.00001 4E-07 2E-09 4.00E-02 2.00E+02 J

Ba 140 12.7 d 125 2500 6E-07 3E-08 2E-09 5.00E-02 1.50E+01 -d Ba 141 1475 175000 0.00003 3.SE-07 1E-07 1.17E-02 3.50E+00 I Ce 141 50 1500 22-07 1.23E-08 8E-10 6.13E-02 1.53E+01 J Ce 143 33.0 h 550 5000 7E-07 1.35E-07 2E-09 1.93E-01 6.75E+01 Cs 138 32.2 m 1700 150000 0.00002 4.25E-07 8E-08 2.13E-02 5.31E+00 a La 140 75 2500 5E-07 1.85E-08 2E-09 3.70E-02 9.25E+00 3 La 141 3.93 h 1150 22500 0.000004 2.75E-07 IE-08 6.88E-02 2.75E+01 a La La 142 143 92.5 m 1450 1500 50000 225000 0.000009 0.00004 3.5E-07 3.75E-07 3E-08 1E-07 3.89E-02 9.38E-03 1.17E+01 3.75E+00 I

Mo 99 66.0 h 400 2500 6E-07 9.75E-08 2E-09 1.63E-01 4.88E+01 j

Nb 98 1475 125000 0.00002 3.5E-07 7E-08 1.75E-02 5.00E+00 I Nd 147 1.73 h 50 2000 4E-07 1.23E-08 1E-09 3.06E-02 1.23E+01 j Pm 151 50 7500 0.000001 1.23E-08 4E-09 1.232-02 3.06E+00 I Pr 143 75 1750 3E-07 1.85E-08 9E-10 6.17E-02 2.06E+01 Pr 145 5.98 h 650 20000 0.000003 1.6E-07 1E-08 5.33E-02 1.60E+01

\1 Rb 88 17.8m 800 150000 0.00003 1.95E-07 9E-08 6.50E-03 2.17E+00 '4 Rb 89 1225 250000 0.00006 3E-07 2E-07 5.OOE.03 1.50E+00 4 Ru 105 4.44 h 200 25000 0.000005 5E-08 2E-08 1.OOE-02 2.50E+00 J, Se 81 18.5m 50 500000 0.00009 1.23E-08 Sn 128 59.1 m 75 75000 0.00001 1.85E-08 32-07 4E-08 1.36E-04 1.85E403 4.08E-02 4.63E-01 I

Sr 89 25 250 6E-08 6.25E-09 2E-10 1.04E-01 3.13E+01 '4 Sr 91 9.5 h 800 10000 0.000001 1.95E-07 52E09 1.95E-01 3.90E+01 Jr Sr 92 .2.71 h 1325 17500 0.000003 3.25E-07 9E-09 1.08E-01 3.61E+01 Te Te 129 131 69.6 m 25.0 m 100 625 150000 12500 0.00003 0.000002 2.45E-08 1.53E-07 9E-08 8.17E-04 2.72E-01 "3.

Te 132 78.2 h 250 500 92-08 6.25E-08 1E-09 9E-10 7.63E-02 6.94E-01 1.53E+02 6.94E+01 I.

Te 133 12.5 m 975 50000 0.000009 2.4E 07 8E-08 2.67E-02 3.00E+00 Te 134 41.8 m 1725 50000 0.00001 4.25E-07 7E-08 4.25E-02 6.07E+00 Y 91m 49.7 m 425 500000 0.00007 1.05E-07 2E-07 1.50E-03 5.25E-01

-1j, Sr 91 9.5h 800 10000 0.000001 1.95E-07 5E209 1.952-01 3.90E+01 %4 Sr 92 2.71 h 1325 17500 0.000003 3.25E-07 9E-09 1.08E-01 3.61E+01 Te 129 69.6 m 100 150000 0.00003 2.45E-08 9E-08 8.17E-04 2.72E-01 Te 131 25.0 m 625 12500 0.000002 1.53E-07 1E-09 7.63E-02 1.53E+02 Te 132 78.2 h 250 500 9E-08 6.25E-08 9E-10 6.94E-01 6.94E+01 Te 133 12.5 m 975 50000 0.000009 2.41E407 8E-08 2.67E-02 3.00E+00 J

3 K-State Reactor 13-24 Original (5/05)

Safety Analysis Report

ACCIDENT-ANALYSIS Table 13.13, Comparison ofParticulate Availablc Activitics (>100 pCi) with ALls and Reactor Bay Concentrations with DACs and Effluent Concentrations.

Available JhIto A Reactor bay EMuent. Raotof Elmnt Nuclidc Hair-life activity.

H(pCi)

LI PI (pCi IU' (pClcm) conC. coneC. 3 (pCmcm ) DAC Limit Te 134 41.8 m 1725 50000 0.00001 4.25E-07 7E-08 4.25E-02 6.07E400 Y 91m 49.7 m 425 500000 0.00007 1.05E-07 2E-07 1.50E-03 5.25E-01 Y 92 3.54 h 850 20000 0.000003 2.08E-07 IE-08 6.92E.02'2.08E.01 Y 93 10.1 h 850 5000 0.000001 2.08E-07 3E-09 2.08E.01 6-92E+01 Zr 95 25 250 5E-08 6.25E-9 4E-10 1.25E-01 1.56E&01 Zr 97 '16.9 h 650 2500 5E-07 1.6E-07 2E-09 3.20E-01 8.O0E+D1

i. Residual Activity from Fuel Utilization Prior to Receipt All but a few instrumented Mark-Il fuel elements in the original 1962 core loading were replaced by Mark-l1l elements on July 10, 1973. The replacement elements had seen considerable use:

prior to their installation at Kansas State University. The two most heavily used elements with serial numbers 4078 and 4079, had experienced, respectively, consumption of nd Eg of 2 isU. Even after about 25 years of subsequent use, considerable ',Cs, rsr,-and aKr remain from fission during the prc-1973 use. However, the '5 Kr atmospheric concentration inside the

.reactor bay iiunediately after release would be orders of magnitude lower than the DAC.

Therefore only "Cs and "0Sr offer a potential for occupational or public risk. In the absence of knowledge about the pattmn.of early fuel utilization, itis assumed that all tfie generation of

'fission' products tookplace in 1973 and that fission product decay took placiover the period of 28 years fron 1973 until 2001.

I..

If Yis the fission yield, Ais the decay constant (si), and N. is Avogadro's number, the activityA (Bq) of any one radionuclide immediately after fissioning of mass m (g) ofr235 U is A= N mYL. . (13.2A-3) 235 Activity calculations using this formula and consequences, as computed in 13.iirc reported in Table 13.15. .. * . . .

r

.. t .

K-State Reactor 13-25 *Original (5105)

Safety Analysis Report

J

'3 J

CHAPTER 13 '3 Table 13.15. Worst Case Source Terms and Consequence Calculations for a Single TRIGA Fuel Element Experiencing 11.27 g of '"U Consumption 28 Years Prior to Element Failure.

RADIONUCLIDE FACTOR 'Sr '3'Cs Half life (y) 29.12 30.00 Decay constant A (s') 7.54 x 10.1 7.32x 10.10 Fission yield Y 0.0577 0.0615 Release fraction 1.OOx 10-' 1.O0X 10-'

Initial Bq/g contained in clement 1.12x 10*" 1.15x 10*"

Initial pCi available for release 34.0 35.2 pCi available for release in 28 y 17A 18.4 ALI (pCi) 4 200 Reactor bay concentration (ttCVcm') 4.3x 10°9 4.5x 10e DAC (pCUcm 3) 2.x 10°9 6.x 10-'

Tissue at risk Bone surface Total body Dose conversion factor (mrem/lpCi) 2690 32 Maximum downwind dose (mrem) 0.16 0.0020 Whereas the "Sr activity available for release would exceed the occupational ALI and, if dispersed within the reactor bay, would have a concentration in excess of the DAC, credible mechanisms for ingestion or inhalation of the full available activity or even its full dispersion are not apparent. Thus, neither the "'Sr nor the "'Cs would pose a

• significant occupational threat. Even if the total available activity were somehow dispersed to the free atmosphere, no person downwind of the accidental release would receive doses even approaching regulatory limits.

j. Conclusions Fission product inventories in TRIGA fuel elements were calculated with the ORIGEN code, using very conservative approximations. Then, potential radionuelide releases from worst-case fuel elements were computed, . again using very conservative approximations. Even if it were assumed that releases took place immediately after reactor operation, and that radionuclides were immediately dispersed inside the reactor bay workplace, few radionuclide concentrations would be in excess *ofoccupational derived air concentrations, and then only for a matter of hours or days. Only for certain nuclides of iodine would the potential release be in excess of the annual limit of intake.

However, there is no credible scenario for accidental inhalation or ingestion of the undiluted radioiodine that might be released from a damaged fuel element.

For the residual "'Sr and 1 7Cs remaining in fuel elements from consumption of 235U prior to receipt of the fuel at Kansas State University, only the former would pose any conceivable occupational threat. However, the total "'Sr activity available for release is K-State Reactor 13-26 Original (5105)

Safety Analysis Report

ACCIDENT ANALYSIS estimated to be ot most about 4 times the AU and there is no credible scenario for its consumption by a worker.

As far as potential consequences to thc'general public are concerned, only for the fcw radionuclides listed in Table 13.14, are maximum concentrations inside the reactor facility in excess of effluent concentrations listed in 10CFR20 and potential doses 0.001 mnrem or greater.- However, evenin the extremely unlikely event that. radionuclides released from a damaged fuel element were immediately released to the outside atmosphere, very conservative calculations reveal that radionuclides inhaled by persons downwind from the release would lead to organ doses or effective doses very far bel6w regulatory limits. As is shown in Table 13.15, the same is true for residual "Sr and 3Cs remaining in fuel elements from early operations.

1*

13.3 Bibliography ANSI/ANS-5.1 -1994. "American NatibnalStandardforDecayHeal Po)ierin Light Wyater Reactors," Arnericah Nuclear Society, 1994.

ReportLA-12625-M. Version 4B, "MCNP-A GeneralMonte CarloN-Particle TranisportCode,"

Los Alamos National Laboratory, Los Alanios, NM (1997). Briesmeister, J.F. (ed),

CCC-37L, "OJGEN2.1Isotope GenerationandDepletionCode: Matrix ExponentialMethod,"

Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1991.

Kansas State University TPRGA MIlkReactor HazardsSummary Report, License R-88, Docket 504188 , (1961) Clack, R.W., JR. Fagan, W.R. Kimel, and S2. Mildail,.

"Laminar-FlowHeat Transferforln-LineFlow Through Unbaffled Rod Bundles,"Nucl. Sci; Engg. 42,81-88(1970), Dwyer, O.E. and H.C. Berry.

"NuclearHeat Transport,"International Textbook Company, Scranton, 1971, El-Wakil, M.M.,.

• Fcderal Guidancc ReponrNo. 11, Rcport EPA-5201/1-88-020, U.S. Environmental Protection Agency, 'Limiting Values of.RadionuclideIntake andAirConcentrationandDoseConversion Factorsforlnhalation.Submersion, andIngestidi7,,WashingtoA,DC, (1988),tEckernan, K F.,

AB. Wolbarst, and A.C3. Richardson.

Report LA-5885-MS "CINDER-7:An InterimReportfor Users,,, England, T.R, et al., Los Alamos Scientific Laboratory, Los Alamos, NM, 1976.

"AadiologicalAssessment,"PrenticcHall, Englewood Cliffs, NJ., 1993, Faw, R.E., and J.K Shultis.

Report LA-9362, 'Application ofAdjusted Datain CalculatingFissionProduceDecay Energies andSpectra," Los Alamos Scicntific Laboratory, 1982.George, D.C., RJ. LaBauve, and TAR England.

NUREGICR-2387 (PNL-4028), "CredibleAccident Analysesfor TRJGA andTRIGA-Fueled Reactors." Report Pacific Northwest Laboratory, Richland Washington, 1982., Hawley, S.C., and R.L Kathrcn, K-State Reactor 13-27 Original (5105)

Safety Analysis Report

CHAPTER 13 .

"FundamentalsofHeat andMass Transfer, "3d ed., Wiley, New York, 1990, Incropera, F.P. and 3 D.P. DeWitt, GA-3399,KSU TRIGA Reactor MechanicalMaintenance and Operating.Manual,General 3 Atomics Report, 1962.

"Fission Product Analytical Source Functions," Nuclear Technology 56, 332-339 (1982).LaBauve, RJ., T.R. England, and D.C. George, See also Reports LA-9090-MS (1981) 1 and LA-UR-80-3305 (1980), Los Alamos Scientific Laboratory, Los Alamnos, NM.

Metals Handbook 8 h ed., Vol. 1,American Society for Metals, Metals Park, Ohio, 1961. J NUREG-1282, "Safety EvaluationReport on High-UraniumContent, Low-Enriched Uranium-Zirconium Hydride Fuelsfor TRIGA Reactors. - Office of Nuclear Reactor Regulation, U.S. 3 Nuclear Regulatory Commission, 1987. d NUREG-1390. "SafetyEvaluationReportRelatingtotheReneivaloftheOperatingLicensefor the TRIGA TrainingandResearch Reactor at the University ofArizona, "Report NUREG-1390, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1990.

NUREG-1537. "GuidelinesforPreparingandReviewingApplicationsforthe Licensing ofNon- 3 PowerReactors, Formatand Content."Report NUREG-1537 Part 1, Office of Nuclear Reactor '3, Regulation, U.S. Nuclear Regulatory Commission, 1996.

"7heU-Zr-HxAlloy: Its Propertiesand Use in TRIGA Fuel,"ReportE-1 17-833, Simnad, M.T, General Atomics Corp., 1980.

"FuelElementsfor Pulsed 'RJGA ResearchReactors," Nuclear Technology 28, 31-56 (1976)

Simnad, M.T., F.C. Faushec, and G.B. West "LongitudinalLaminar Flow Between Cylinders Arrangedin a RegularArray,"AlChE Journal 5,325 (1959), Sparrow, E.M., and A.L. Loeffler, and H.A. Hubbard. '3, NuclearSystems 1: Thermal HydraulicFundamentals, Todreas, N.E. and M.S. Kazimi, Hemisphere, New York, 1990. '3 Safety.Analysis Report, TRIGA ReactorFacility, NuclearEngineeringTeaching Laboratory.

University of Texas atAustin, Revision 1.01, Docket 50-602, May, 1991.

3

'Kinetic Behavior of TRIGA Reactors," Report GA-7882, West, G.B., W.L. Whittemore, J.R. '3.

Shoptaugh, Jr., J.B. Dee, and C.O. Coffer,, General Atomics Corp., 1967. s, j

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APPENDICES TO CHAPTER 13 A Origen 2.1 input file for " 5Ufission at 1 W thermal power for 40 years.

B Origen 2.1 input file for 235U fission at I W thermal power 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> perday for 5 days.

C Origen 2.1 output file extracts for 235 U fission at 1 W thermal power for 40 years.

/

D Origen 2.1 output file extracts for 7JsU fission at I W thermal power 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day for 5

-'p Days.

E Maximum activity available for release from i single T.RGA fuel element as a function of time after shutdown for a 235U-fueled thermal reactor operating at 3.5 kWV thermal power for 40 years, based on one element of 81 at 86.42 W.

-p F Maximum activity available for release from a single TRIGA fuel element as a function of

-I time after shutdown for a 235UL-fleled thermal reactor operating at 500 kWV kW thermal power for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day for 5 days, based on one element of 83 at 12.05 kW

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/

I

-I 1

-J

-p

-I

-'p K-State Reactor Original (10104)

-- p Safety Analysis Report

i I

CHAPTER 13 APPENDIX A i ORIGEN Input file for 1 tonne U-235 at 1 watt for 40 years d

-1

-1

• 1

-1 RDA ORIGEN2, VERSION 2.1 (8-1-91) TRIGA REACTOR REFERENCE PROBLEM RDA UPDATED BY: Richard E. Faw, Kansas State University BAS ONE TONNE OF U-235 RDA Continuous operation for 40 years at 1 watt RDA WARNING: VECTORS ARE OFTEN CHANGED WITH RESPECT TO THEIR CONTENT.

RDA THESE CHANGES WILL BE NOTED ON RDA CARDS.

CUT -1 RDA LIP LIBRARY PRINT (I-PRINT,0-DON'T PRINT) 0 0 0 a

RDA DECAY LIBRARY CHOICES (0-PRINT, 1 2 3 DECAY LIBRARIES; 601 ...

RDA CROSS SECTIONS; ETC, SEE P. 47)

LIB 0 1 2 3 201 202 203 9 3 0 1 38 PHO 101 102 103 10 <<< PHOTON LIBRARIES, P. 47 TIT INITIAL COMPOSITIONS OF UNIT AMOUNTS OF FUEL AND STRUCT MAT'LS RDA READ FUEL COMPOSITION INCLUDING IMPURITIES (1 G)

INP -1 1 -1 -1 1 1 TIT IRRADIATION OF ONE TONNE U-235 NOV -1 1 0 1.0 HED 1 CHARGE BUP IRP 5 .000001 1 2 5 2 1 W/Tonne FOR 5 YEARS IRP 10 .000001 2 1 5 0 1 W/Tonne FOR 5 YEARS IRP 15 .000001 1 2 5 0 1 W/Tonne FOR 5 YEARS IRP 20 .000001 2 1 5 0 1 W/Tonne FOR 5 YEARS IRP 25 .000001 1 2 5 0 1 W/Tonne FOR 5 YEARS IRP 30 .000001 2 1 5 0 1 W/Tonne FOR 5 YEARS IRP 35 .000001 1 2 5 0 1 W/Tonne FOR 5 YEARS IRP 40 .000001 2 1 5 0 1 W/Tonne FOR 5 YEARS BUP OPTL 24*8 ACTIVATION PRODUCT OUTPUT OPTS P. 56 OPTA 24*8 ACTINIDE OUTPUT OPTIONS P. 59 OPTF 6*8 5 17*8 FISSION PRODUCT OUTPUT OPTIONS P. 59 RDA DE1:AY TO 2E1 DAYS DEC 1 1 2 4 2 DEC 2 2 3 4 0 DEC 3 3 4 4 0 DEC 7 4 5 4 0 DEC 14 5 6 4 0 DEC 28 6 7 4 0 OUT -a 1 -1 0 OUT 7 1 -1 0 'J.

END 2 922340 0.0 922350 1.E06 922380 0. 0 0.0 PURE U-235 0

J-a J

K-State Reactor 13A-1 Original (10104) J Safely Analysis Report J

CHAPTER 13 APPENDIX B ORIGEN Input File for 1 tonne U-235 at I watt 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day for 5 days

-1

-1

-1 RDA ORIGEN2, VERSION 2.1 (6-1-91) TRIGA REACTOR REFERENCE PROBLEM RDA UPDATED BY: Richard E. Faw, Kansas State University BAS One tonne U-235 RDA Fuel composition CUT -1 RDA LIBRARY PRINT (1-PRINT,0-DON'T PRINT)

LIP 0 0 0 RDA DECAY LIBRARY CHOICES (0-PRINT; 1 2 3 DECAY LIBRARIES; 601 ...

RDA CROSS SECTIONS; ETC, SEE P. 47)

LIB 0 1 2 3 201 202 203 9 3

• 0 1 38 PHO 101 102 103 10 <<< PHOTON LIBRARIES, P. 47 TIT INITIAL COMPOSITIONS OF UNIT AMOUNTS OF FUEL AND STRUCT MAT'LS RDA READ FUEL COMPOSITION INCLUDING IMPURITIES INP -1 1 -1 -1 1 1 TIT One tonne U-235 8 h/d for 5 days at 1 kWt MOV 1 0 1.0 HED 1 CHARGE BUP IRP 8.0 0.001- 1 2 3 2 OPERATE FOR 8 HR AT 1 kW DEC 24.0 2 3 3 0 COOLFOR 16 HOURS IkP 32.0 0.001 3 4 3 0 OPERATE FOR 8 HR DEC 48.0 4 5 3 0 COOL FOR 16 HOURS IRP 56.0 0.001 5 6 3 0 OPERATE FOR 8 HR DEC 72.0 6 7 3 0 COOL FOR 16 HOURS IRP 80.0 0.001 7 8 3 0 OPERATE FOR 8 HR DEC 96.0 8 9 3 0 COOL FOR 16 HOURS IRP- 104.0 0.001 9 10 3 0 OPERATE FOR 8 HR OPTL 24*8 ACTIVATION PRODUCT OUTPUT OPTS P. 56 OPTA 24*8 ACTINIDE OUTPUT OPTIONS P. 59 OPTF 6*8 5 17*8 FISSION PRODUCT OUTPUT OPTIONS P. 59 RDA MOVE COMPOSITION VECTOR FROM 10 TO I MOV 10 1 0 1.0 RDA DECAY TO 0.1 UNITS (2-MINUTES) FROM COMP VEC 1 TO VEC 3 DEC 1 1 2 4 2 DEC 2 2 3 4 0 DEC 3 3 4 4 0 DEC 7 4 5 4 0 DEC 14 5 6 4 0 DEC 28 6 7 4 0 OUT -7 1 -1 0 OUT 7 1 -1 0 END 2 922340 0.0 922350 1.E6 922380 0.00 a 0.0 1 g U-235 0

K-State Reactor 13.3-1 Original (10104)

Safety Analysis Report

CHAPTER 13 APPENDIX C ORIGEN Output File Extracts for 1 tonne U-235 at 1 watt for 40 Years NUCLIDE TABLE: RADIOACTIVITY, CURIES Time post discharge 0 1.0 D 2.0 D 3.0 D 7.0 D 14.0 D 28.0 D AG 111 1.65E-04 l.SlE-04 1.37E-04 1.25E-04 8.62E-05 4.50E-05 1.22E-05 BA 140 5.23E-02 4.96E-02 4.70E-02 4.45E-02 3.58E-02 2.45E-02 1.15E-02 BA 137 2.93E-02 2.93E-02 2.93E-02 2.93E-02 2.93E-02 2.93E-02 2.92E-02 CE 141 4.93E-02 4.85E-02 4.75E-02 4.65E-02 4.27E-02 3.68E-02 2.73E-02 CE 143 4.98E-02 3.03E-02 1.83E-02 1.11E-02 1.47E-03 4.32E-05 3.72E-08 CE 144 4.56E-02 4.55E-02 4.54E-02 4.53E-02 4.48E-02 4.41E-02 4.26E-02 CS 137 3.10E-02 3.10E-02 3.09E-02 3.09E-02 3.09E-02 3.09E-02 3.09E-02 EU 155 2.75E-04 2.75E-04 2.75E-04 2.75E-04 2.75E-04 2.74E-04 2.72E-04 EU 156 1.13E-04 1.10E-04 1.06E-04 1.01E-04 8.43E-OS 6.13E-05 3.23E-o5 1 131 2.37E-02 2.20E-02 2.03E-02 1.87E-02 1.33E-02 7.28E-03 2.19E-03 1 132 3.56E-02 2.95E-02 2.39E-02 1.93E-02 8.23E-03 1.86E-03 9.45E-05

• 1 133 5.68E-02 2.62E-02 1.18E-02 5.29E-03 2.16E-04 8.00E-07 1.10E-11 1 135 5.31E-02 4.29E-03 3.46E-04 2.80E-0S 1.19E-09 2.66E-17 1.34E-32 KR 85 2.10E-03 2.10E-03 2.10E-03 2.10E-03 2.10E-03 2.10E-03 2.09E-03 KR 85H 1.07E-02 2.64E-04 6.44E-06 1.57E-07 5.58E-14 2.88E-25 0.00E+00 LA 140 5.24E-02 5.19E-.02 5.06E-02 4.89E-02 4.08E-02 2.82E-02 1.32E-02 LA 141 4.93E-02 7.77E-04 1.13E-05 1.64E-07 7.27E-15 9.89E-28 0.00E+00 HO 99 5.06E-02 3.94E-02 3.06E-02 2.38E-02 8.68E-03 1.49E-03 4.36E-OS NB 95 5.386E-02 5.38E-02 5.38E-02 5.37E-02 5.35E-02 5.28E-02 5.04E-02 NB 97 4.93E-02 1.85E-02 6.90E-03 2.58E-03 5.03E-05 5.14E-08 5.70E-14 NB 95H 3.78E-04 3.77E-04 3.76E-04 3.74E-04 3.64E-04 3.41E-04 2.95E-04 NB 97H 4.66E-02 1.74E-02 6.51E-03 2.43E-03 4.74E-05 4.84E-08 5.01E-14 ND 147 1.90E-02 1.79E-02 1.68E-02 1.58E-02 1.23E-02 7.91E-03 3.29E-03 PH 147 1.9lE-02 1.91E-02 1.91E-02 1.91E-02 1.91E-02 1.90E-02 1.89E-02 PM 149 9.11E-03 6.89E-03 5.04E-03 3.68E-03 1.05E-03 1.17E-04 1.46E-06 PH 151 3.52E-03 1.97E-03 1.10E-03 6.11E-04 5.86E-05 9.70E-07 2.65E-10 PR 143 4.98E-02 4.93E-02 4.80E-02 4.64E-02 3.86E-02 2.71E-02 1.33E-02 PR 144 4.56E-02 4.55E-02 4.54E-02 4.53E-02 4.48E-02 4.4lE-02 4.26E-02 PR 145 3.29E-02 2.06E-03 1.27E-04 7.89E-06 1.16E-10 4.06E-19 5.14E-36 PR 144 S.48E-04 5.46E-04 5.45E-04 5.43E-04 5.38E-04 5.29E-04 5.11E-04 RH 105 8.53E-03 6.09E-03 3.83E-03 2.39E-03 3.64E-04 1.35E-05 1.86E-08 RH 106 3.27E-03 3.26E-03 3.26E-03 3.25E-03 3.23E-03 3.18E-03 3.10E-03 RH 103 2.37E-02 2.33E-02 2.29E-02 2.25E-02 2.09E-02 1.85E-02 1.45E-02 RU 103 2.63E-02 2.58E-02 2.54E-02 2.49E-02 2.32E-02 2.05E-02 1.60E-02 RU 10S 8.53E-03 2.086E-04 4.90E-06 1.16E-07 3.57E-14 1.44E-25 0.00E+00 RU 106 3.27E-03 3.26E-03 3.26E-03 3.25E-03 3.23E-03 3.18E-03 3.10E-03 SB 125 2.49E-04 2.49E-04 2.49E-04 2.49E-04 2.48E-04 2.47E-04 2.45E-04 SB. 127 1.10E-03 9.28E-04 7.75E-04 6.47E-04 3.15E-04 8.93E-05 7.18E-06 SB 129 S.33E-03 1.lSE-04 2.45E-06 5.20E-08 1.06E-14 2.08E-26 0.O0E+00 SM 151 9.34E-04 9.34E-04 9.34E-04 9.34E-04 9.34E-04 9.33E-04 9.33E-04 SM 153 1.36E-03 9.55E-04 6.69E-04 4.68E-04 1.13E-04 9.30E-06 6.34E-08 SN 125 1.13E-04 1.05E-04 9.75E-05 9.07E-05 6.80E-OS 4.11E-05 1.50E-05 SR B9 4.05E-02 3.99E-02 3.94E-02 3.89E-02 3.68E-02 3.34E-02 2.76E-02 SR 90 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 SR 91 4.94E-02 8.59E-03 1.49E-03 2.59E-04 2.35E-07 1.12E-12 2.51E-23 SR 92 5.04E-02 1.09E-04 2.35E-07 5.07E-10 1.10E-20 0.OOE+0O 0.O0E+00 TC 99M 4.43E-02 3.76E-02 2.95E-02 2.29E-02 8.36E-03 1.43E-03 4.20E-05 TE 127 1.09E-03 1.02E-03 8.88E-04 7.68E-04 4.49E-04 2.27E-04 1.37E-04 TE 129 S.25E-03 6.43E-04 4.99E-04 4.86E-04 4.48E-04 3.88E-04 2.90E-04 TE 131 2.13E-02 3.95E-04 2.27E-04 1.30E-04 1.42E-05 2.92E-07 1.24E-10 TE 132 3.54E-02 2.86E-02 2.32E-02 1.87E-02 7.99E-03 1.8OE-03 9.17E TE 127 1.53E-04 1.53E-04 1.52E-04 1.52E-04 1.SOE-04 1.44E-04 1.33E-04 TE 129 7.9lE-04 7.78E-04 7.63E-04 7.47E-04 6.88E-04 5.95E-04 4.46E-04 TE 131 3.04E-03 1.75E-03 1.01E-03 5.79E-04 6.30E-05 1.30E-06 5.52E-10 XE 133 5.68E-02 S.47E-02 5.03E-02 4.52E-02 2.74E-02 1.09E-02 1.72E-03 XE 135 5.52E-02 2.02E-02 4.15E-03 7.39E-04 5.35E-07 1.47E-12 1.09E-23 XE 131 2.63E-04 2.62E-04 2.61E-04 2.58E-04 2.41E-04 1.97E-04 1.llE-04 XE 133 1.65E-03 1.51E-03 1.24E-03 9.65E-04. 2.96E-04 3.27E-05 3.90E-07 XE 135 9.52E-03 6.87E-04 5.55E-05 4.48E-06 1.91E-10 4.27E-18 2.14E-33 Y 90 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 Y 91 4.94E-02 4.91E-02 4.86E-02 4.80E-02 4.58E-02 4.21E-02 3.57E-02 Y 92 5.05E-02 1.60E-03 1.70E-05 1.60E-07 1.10E-15 5.65E-30 0.00E+00 Y 93 5.44E-02 1.06E-02 2.04E-03 3.93E-04 5.41E-07 5.32E-12 5.15E-22 Y 91M 2.86E-02 S.46E-03 9.47E-04 1.64E-04 1.49E-07 7.09E-13 1.60E-23 ZR 95 5.38E-02 S .32E-02 5.26E-02 5.21E-02 4.99E-02 4.62E-02 3.97E-02 ZR 97 4.92E-02 1.84E-02 6.87E-03 2.57E-03 5.00E-05 S.11E-08 5.29E-14 OrigInal (10104)

K-State Reactor I 3.C-1 13.C-1 Original (10/04)

Safety Analysis Report

CHAPTER 13 APPENDIX D ORIGEN Output File Extracts for 1 tonne U-235 at 1 U for 8 h/d. 5 Days NUCLIDE TABLE: RADIOACTIVITY& CURIES Time post discharge 0 1 .0 D 2.0 D 3.0 D 7.0 D 14.0 D 28.0 D AS 78 1.24E-01 2.1 7E-05 5.66E-10 1.16E-14 1.1E-33 0.00E+00 0.00E+00 BA 139 5.33E+01 3.4 7E-04 1.99E-09 1.14E-14 1.28E-35 O.OOE+00 0.00E+00 BA 140 4.22E+00 4.0 0E+00 3.79E+00 3.59E+00 2.89E+00 1.98E+00 9.27E-01 BA 141 4.91E+01 0.0 OE+00 0.OOE+00 O.OOE+00 O.OOE+OO O.OOE+00 O.OOE+00 BA 142 4.88E+01 0.0 0.00£+00 0.OOE+00 O.00E+00 0.00+00 0.00E+00 BR 83 4.02E+00 4.1 3E-03 3.92E-06 3.72E-09 3.O1E-21 O.OOE+00 0.00E+00 BR 84 8.43E+00 2.1 9E-13 5.13E-27 0.00E+00 0.00E+00 0.00E+00 0.00E+00 CE 141 1.57E+00 1.1'4E+00 1.71E+00 1.67E+00 1.54E+00 1.32E+00 9.81E-01 CE 143 1.84E+01 1.1 3E+01 6.83E+00 4.13E+00 5.49E-01 1. 61E-02 1.39E-05 CE 144 1.85E-01 2.8 4E-01 1.54E-01 1.83E-01 1.82E-01 1.79E-01 1.73E-01 CS 138 5.68E+01 3.4 4E-12 1.17E-25 0.00E+00O 0.00E+00 0.00E+00 O.OOE+00 S 131 2.82E+00 2.7 2.55E+00 2.37E+00 1.70E+00 9.33E-01 2.79E-01 I 132 7.73E+00 6.9 95E+00 5.62E+00 4.54E+00 1.94E+00 4.37E-01 2.22E-02 1 133 2.32E+01 1.1 lE+01 5.00E+00 2.25E+00 9.17E-02 3.40E-04 4.66E-09 I 134 6.38E+01 1.6 62E-06 9.36E-15 5.37E-23 O.OOE+00 0.00E+00 O.OOE+00 I 135 3.28E+01 2.6 65E+00 2.14E-01 1.73E-02 7.35E-07 1.65E-14 8.26E-30 KR 57 2.14E+01 4.5 S1E-OS 9.39E-11 1.96E-16 0.00E+00 0.00E+00 0.00E+00 KR B8 2.62E+01 7.4 49E-02 2.14E-04 6.09E-07 4.03E-17 6.19E-35 O.OOE+00 KR 83M 3.30E+00 1.5 9E-02 1.65E-05 1.58E-08 1.29E-20 0.00E+00 O.OOE+00 KR 85H 7.76E+00 1.993E-01 4.70E-03 1.15E-04 4.07E-11 2.10E-22 0.00E+00 LA 140 2.25E+00 2.0 BE+00 3.22E+00 3.38E+00 3.19E+00 2.27E+00 1.07E+00 LA 141 3.79E+01 6.1 lE-01 8.86E-03 1.29E-04 5.72E-12 7.77E-25 0.00E+00 LA 142 4.83E+01 1.1 15E-03 2.43E-08 5.12E-13 1.01E-31 O.OOE+00 O.OOE+00 LA 143 4.96E+01 0.0 OOE+00 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00 O.OOE+00

• 0o 99 1.31E+01 1.0 2E+01 7.93E+00 6.17E+00 2.25E+00 3.85E-01 1.13E-02 MO 101 4.23E+01 0.0 OOE+00 0.00E+00 O.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 NBt 97 1.98E+01 8.1 9E+00 3.06E+00 1.14E+00 2.23E-02 2.28E-05 2.53E-11 NB 98 4.86E+01 0.0 OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 ND 147 1.79E+00 1.6 69E+00 1.59E+00 1.49E+00 1.16E+00 7.50E-01 3.12E-01 ND 149 8.74E+00 5.9'6E-04 3.98E-08 2.65E-12 5.23E-29 O.OOE+00 0.00E+00 ND 151 3.50E+00 0.0 '0E+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 PD 109 1.l9E-01 3.5 SOE-02 1.02E-02 2.95E-03 2.11E-05 3.69E-09 1.13E-16 PM 151 1.36E+00 7.1 '2E-01 4.30E-01 2.39E-O1 2.30E-02 3.80E-04. 1.04E-07 PR 143 2.31E+00 2.9 2E+00 3.22E+00 3.33E+00 3.03E+00 2.16E+00 1.06E+00 PR 144 1.86E-01 1.8 4E-01 1.84E-01 1.83E-01 1.82E-0O 1.79E-01 1.73E-01 PR 145 2.12E+01 1.3 33E+00 8.25E-02 5.11E-03 7.52E-08 2.63E-16 3.22E-33 PR 147 1.90E+01 0.0 O0E+00 0.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 0.00E+00 RB 88 2.65E+01 9.337E-02 2.39E-04 6.80E-07 4.50E-17 7.40E-35 O.OOE+00 RH 105 2.57E+00 2.1 7E+00 1.37E+00 8.55E-01 1.30E-01 4.84E-03 6. 68E-06 RH 107 1.37E400 1.8 4E-20 O.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 0.00E+00 RU 103 7.45E-01 7.3 3E-0O 7.20E-01 7.08E-01 6.59E-01 5.83E-01 4.55E-01 RU 105 6.23E+00 1.5S4E-01 3.62E-03 8.54E-05 2.64E-1l 1.07E-22 O.OOE+00 SB 127 2.19E-01 1.9 96E-01 1.63E-01 1.36E-01 6.64E-02 1.88E-02 1.51E-03 SE 81 1.76E+00 2.4 9E-09 6.77E-17 1.84E-24 0.00E+00 0.00E+00 0.00+00 SE 83 1.64E+00 8.9 '8E-20 0.00E+00 0.00E400 0.00E+00 O.OOE+00 O.OOE+00 SM 153 4.27E-01 3.0 OlE-Ol 2.11E-01 1.48E-01 3.55E-02 2.93E-03 2.00E-05 SM 155 2.76E-01 8.5 53E-21 0.00E+00 O.OOE+00 0.00E+00 0.00£+00 0 .00E+00 SN 127 6.52E-01 2.3 7E-04 8.59E-08 3.12E-11 5.40E-25 O.OOE+00 0.00E+00 SN 128 2.78E+00 1.2 25E-07 5.62E-IS 2.53E-22 0.00E+00 O.OOE+00 O.OOE+00 SR 89 9.30E-01 9.2 9.15E-01 9.02E-01 8.54E-01 7.76E-01 6.40E-01 SR 91 2.64E+01 4.6 06E+00 7.99E-01 1.39E-01 1.26E-04 5.98E-10 1.35E-20 SR 92 4.40E+01 9.5SOE-02 2.05E-04 4.42E-07 9.59E-18 1.62E-36 O.OOE+00 TC 101 4.23E+01 0.0O0E+00 0.00E+00 0.006+00 0.00E+00 O.OOE+00 O.OOE+00 TC 104 1.53E+01 0.0 OE+00 o.ooe+00 O.OOE+00 0.00E+00 O.OOE+00 0.00+00 TE 127 1.61E-01 1.1 '7E-01 1.55E-01 1.31E-01 6.49E-02 1.94E-02 2.81E-03 TE 129 3.11E+00 1.1 SE-Ol 1.87E-02 1.62E-02 1.49E-02 1.29E-02 9.65E-03 TE 131 2.09E+01 1.5 lE-01 8.68E-02 4.99E-02 5.43E-03 1.12E-04 4.76E-08 tE 132 8.32E+00 6.174E+00 5.45E+00 4.41E+00 1.886E+00 4.24E-01 2.16E-02 5E 133 3.20E+01 6.4 42E-08 9.61E-16 1.44E-23 0.00E+00 0.00E+00 O.OOE+00 TE 134 5.69E+01 2.4 2E-09 1.03E-19 4.40E-30 0.00E+00 0.00E+00 0.00E+00 XE 133 6.76E+00 8.0 OOE+00 7.97E+00 7.43E+00 4.68E+00 1.88E+00 2.97E-01 XE 135 1.96E+01 1.0 O1E+01 2.19E+00 3.97E-01 2.91E-04 7.99E-10 5.94E-21

_ _ XE 138 5.26E+01 0.0 OOE+00 0.00+00 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 Y 91 7.71E-01 9.1 6E-01 9.32E-01 9.25E-01 8.84E-01 8.13E-01 6.89E-01 Y 92 2.81E+01 1.2 25E+00 1.35E-02 1.29E-04 8.81E-13 4.51E-27 O.OOE+00 Y 93 2.86E+01 5.1 64E+00 l.09E+00 2.09E-01 2.88E-04 2.83E-09 2.74E-19 Y 94 5.26E+01 0.0 OE+00 0.00E+00 0.00+00 0.00E+00 0.00E+00 O.OOE+00 Y 95 5.35E+01 0.0 O0E+00 O.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 0.00E+00 K-State Reactor 13.1-1 Original (10104)

Safety Analysis Report

CHAPTER 13l Y 91k4 1.41E401 2.93E+OO S.08E-O1 8.S1E-02 8.OOE-05 3.80Z-1O 8.56E-21 v ZR 95 9.72E-O1 9.68C-O1 9.59E-O1 9.47E-O1 9.O7E-O1 8.41E-O1 7.23E-O1 ZR 97 2.18E401 G.1SE.OO 3.O5Et~o 1.14E+OO 2.22E-02 2.27E-05 2.34E-11 J

J J

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ACCIDENT ANALYSIS CHAPTER 13 APPENDIX E Maximum Activity Available for Release One TRICA Element at 86.42 W for 40 Years (Release fractions: IE-04 for halogens and noble gases, 1E-06 for particulates)

Potential Activity Release (pCi) Inhalation Initial say Time post discharge [days) ALI DAC Concentration 2

0 1 2 3 7 14 28 pCi LCi cm pCi/cm?

AG 111 0.0 0.0 0.0 0.0 0.0 0.0 0.0 9.E+02 4.E-07 3.5E-12 BA 140 4.5 4.3 4.1 3.8 3.1 2.1 1.0 1.E+03 S.E-07 l.lE-09 BA 137 2.5 2.5 2.5 2.5 2.5 2.5 2.5 na na 6.2E-10 CE 141 4.3 4.2 4.1 4.0 3.7 3.2 2.4 6.E+02 2.E-07 l.OE-09 CE 143 4.3 2.6 1.6 1.0 0.1 0.0 0.0 2.E+03 . 7.E-07 1.1E-09 CE 144 3.9 3.9 3.9 3.9 3.9 3.8 3.7 l.E+01 6.E-09 9.7E-10 CS 137 2.7 2.7 2.7 2.7 2.7 2.7 2.7 2.E+02 6.E-08 6.6E-10 EU 155 0.0 0.0 0.0 0.0 0.0 0.0 0.0 9.E+01 4.E-08 S.SE-12 EU 156 0.0 0.0 0.0 0.0 0.0 0.0 0.0 S.E+02 2.E-07 2.4E-12 1 131 204.6 190.1 175.3 161.4 114.9 62.9 19.8 5.E+O1 2.E-08 5.OE-08 I 132 307.3 255.3 206.4 166.8 71.2 16.0 0.9 8.E+03 3.E-06 7.5E-0O 1 133 490.5 226.5 101.8 45.7 1.9 0.0 0.0 3.E+02 1.E-07 1.2E-07 1 135 458.6 37.0 3.0 0.2 0.0 0.0 0.0 2.E+03 7.E-07 l.lE-07 KR as 18.1 18.1 18.1 18.1 18.1 18.1 18.1 1.E-04 4.5E-09 KR 85H 92.2 2.3 0.1 0.0 0.0 0.0 0.0 2.E-05 2.3E-08 LA 140 4.5 4.5 4.4- 4.2 3.5 2.4 1.1 1.E+03 5.E-07 1.lE-09 LA 141 4.3 0.1 0.0 0.0 0.0 0.0 0.0 9.E+03 4.E-06 l.OE-09 MO 99 4.4 3.4 2.6 2.1 0.7 0.1 0.0 1.E+03 6.E-07 l.lE-09 KB 95 4.6 4.6 4.6 4.6 4.6 4.6 4.4 1.E+03 5.E-07 l.lE-09 NB 97 4.3 1.6 0.6 0.2 0.0 0.0 0.0 7.E+04 3.E-05 l.OE-09 KB 95M 0.0 0.0 0.0 0.0 0.0 0.0 0.0 2.E+03 9.E-07 8.OE-12 KB 97H 4.0 1.5 0.6 0.2 0.0 0.0 0.0 na na 9.9E-10 ND 147 1.6 1.5 1.4 1.4 1.1 0.7 0.3 8.E+02 4.E-07 4.OE-10 PM 147 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.E+02 5.E-08 4.OE-10 PH 149 0.8 0.6 0.4 0.3 0.1 0.0 0.0 2.E+03 8.E-07 1.9E-10 PM 251 0.3 0.2 0.1 0.1 0.0 0.0 0.0 3.E+03 1.E-06 7.5E-11 PR 143 4.3 4.3 4.2 4.0 3.3 2.3 1.1 7.E+02 3.E-07 l.lE-09 PR 144 3.9 3.9 3.9 3.9 3.9 3.8 3.7 1.E+05 5.E-05 9.7E-10 PR 145 2.8 0.2 0.0 0.0 0.0 0.0 0.0 8.E+03 3.E-06 7.0E-10 PR 144 0.0 0.0 0.0 0.0 0.0 0.0 0.0 na na 1.2E-11 RH 105 0.7 0.5 0.3 0.2 0.0 0.0 0.0 6.E+03 2.E-06 1.8E-10 RH 106 0.3 0.3 0.3 0.3 0.3 0.3 0.3 na na 6.9E-11 RH 103 2.0 2.0 2.0 1.9 1.9 1.6 1.2 1.E+06 5.E-04 5.OE-10 RU 103 2.3 2.2 2.2 2.2 2.0 1.8 1.4 6.E+02 3.E-07 5.6E-10 RU 105 0.7 0.0 0.0 0.0 0.0 0.0 0.0 1.E+04 5.E-06 1.8E-10 RU 106 0.3 0.3 0.3 0.3 0.3 0.3 0.3 l.E+0l S.E-09 6.9E-11 SS 125 0.0 0.0 0.0 0.0 0.0 0.0 0.0 S.E+02 2.E-07 5.3E-12 S8 127 0.1 0.1 0.1 0.1 0.0 0.0 0.0 9.E+02 4.E-07 2.3E-11 S8 129 0.5 0.0 0.0 0.0 0.0 0.0 0.0 9.E+03 4.E-06 1.1E-10 SKI 151 0.1 0.1 0.1 0.1 0.1 0.1 0.1 1.E+02 4.E-08 2.OE-11 SM 153 0.1 0.1 0.1 0.0 0.0 0.0 0.0 3.E+03 1.E-06 2.9E-11 SS 125 0.0 0.0 0.0 0.0 0.0 0.0 0.0 4.E+02 1.E-07 2.4E-12 SR 8g 3.5 3.5 3.4 3.4 3.2 2.9 2.4 1.E+02 6.E-08 8.6E-10 SR 90 2.6 2.6 2.6 2.6 2.6 2.6 2.6 4.E+00 2.E-09 6.3E-10 SR 91 4.3 0.7 0.1 0.0 0.0 0.0 0.0 4.E+03 1.E-06 1.OE-09 SR 92 4.4 0.0 0.0 0.0 0.0 0.0 0.0 7.E+03 3.E-06 1.lE-09 TC 99M 3.8 3.3 2.5 2.0 0.7 0.1 0.0 2.E+0S 6.E-0S 9.4E-10 TE 127 0.1 0.1 0.1 0.1 0.0 0.0 0.0 2.E+04 7.E-06 2.3E-1l TE 129 0.5 0.1 0.0 0.0 0.0 0.0 0.0 6.E+04 3.E-05 1.1E-10 TE 131 1.8 0.0 0.0 0.0 0.0 0.0 0.0 5.E+03 2.E-06 4.5E-10 TE 132 3.1 2.5 2.0 1.6 0.7 0.2 0.0 2.E+02 9.E-08 7.5E-i0 TE 127 0.0 0.0 0.0 0.0 0.0 0.0 0.0 3.E+02 1.E-07 3.2E-12 TE 129 0.1 0.1 0.1 0.1 0.1 0.1 0.0 2.E+02 1.E-07 1.7E-11 TE 131 0.3 0.2 0.1 0.1 0.0 0.0 0.0 4.E+02 2.E-07 6.4E-11 XE 133 490.7 473.1 434.5 390.3 236.9 94.5 14.9 1.E-04 1.2E-07 XE 135 477.0 174.2 35.8 6.4 0.0 0.0 0.0 l.E-OS 1.2E-07 XE 131 2.3 2.3 2.3 2.2 2.1 1.7 1.0 4.E-04 5.6E-10 XE 133 14.3 13.1 10.7 8.3 2.6 0.3 0.0 1.E-04 3.5E-09 XE 135 E2.3 5.9 0.5 0.0 0.0 0.0 0.0 9.E-06 2.0E-08 Y 90 2.6 2.6 2.6 2.6 2.6 2.6 2.6 1.E+02 S.E-0 6.3E-10 Y 91 4.3 4.2 4.2 4.1 4.0 3.6 3.1 8.E+03 3.E-06 1.OE-09 Y 92 4.4 0.1 0.0 0.0 0.0 0.0 0.0 2.E+03 1.E-06 1.1E-09 Y 93 4.7 0.9 0.2 0.0 0.0 0.0 0.0 2.E+03 1.E-06 1.2E-09 Y 91H 2.5 0.5 0.1 0.0 0.0 0.0 0.0 2.E+05 7.E-OS 6.1E-10 2R 95 4.6 4.6 4.5 4.5 4.3 4.0 3.4 1.E+02 5.E-08 1.1E-09 ZR 97 4.3 1.6 0.6 0.2 0.0 0.0 0.0 1.E+03 S.E-07 1.OE-09 K-State Reactor

• 13-3 Original (5/05)

Safety Analysis Report

CHAPTER 13 APPENDIX F Maximum Activity Available for Release One TRIGA Element at 31.125 kW, 8 h/d, 5 Days (Release fractions: lE-04 for halogens and noble gases, IE-06 for particulates)

AS 78 3.75 0 0 0 0 0 0 2.E+04 9.E-06 9.25E-10 NA BA 139 1604.3 0 0 0 0 0 0 3.E.04 1.E-05 4.0DE-07 NA BA 140 127.25 120.5 114.3 108.3 87.25 59.5 28 1.E+03 6.E-07 3.00E-OB NA BA 141 1480.3 0 0 0 0 0 0 7.E+04 3.E-05 3.75E-07 NA BA 142 1470.5 0 0 0 0 0 0 1.E+05 6.E-05 3.50E-07 NA BR 83 12110 12.5 0 0 0 0 0 6.E*04 3.E-05 3.00E-06 NA BR 84 25386 0 0 0 .0 0

• 0 6.E+04 2.E-05 6.25E-06 NA CE 141 47.25 52.5 51.5 50.25 46.25 39.75 29.5 6.E+02 2.E-07 l.15E-08 NA CE 143 553.5 340.75 205.8 124.3 16.5 0.5 0 2.E+03 7.E-07 1.35E-07 NA CE 144 5.5 5.5 5.5 5.5 5.5 5.5 5.25 l.E+01 6.E-09 1.38E-09 NA CS 138 1711.8 0 0 0 0 0 0 6.E+04 2.E-05 4.25E-07 NA 1 131 8459.3 8245.3 7694 7131 5124 2810 840.5 5.E+01 2.E-08 2.08E-06 103.8 1 132 23281 20943 16930 13686 5838 1317 67 8.E503 3.E-06 5.75E-06 1.9 1 133 69950 33529 15072 6772 276.3 1 0 3.E+02 1.E-07 1.73E-05 172.5 1 134 192228 0 0 0 0 0 0 5.E504 2.E-05 4.75E-05 2.4 1 135 98750 7980 644.3 52 0 0 0 2.E+03 7.E-07 2.43E-05 34.6 KR 87 64498 0.25 0 0 0 0 0 na S.E-06 1.58E-05 3.2 KR 88 79048 225.75 0.75 0 0 0 0 na 2.E-06 l.95E-05 9.8 KR 83H 9953.3 47.75 0 0 0 0 0 na 1.E-02 .2.45E-06 NA KR 85H 23368 580.25 14.25 0.25 0 0 0 na 2.E-05 5.75E-06 NA LA 140 67.75 86.75 97 101.8 96 68.5 32.25 1.E503 5.E-07 1.68E-08 NA LA 141 1140.3 18.5 0.25 0 0 0 0 9.E.03 4.E-06 2.75E-07 NA*

LA 142 1454.5 0 0 0 0 0 0 2.E+04 9.E-06 3.50E-07 NA LA 143 1493.3 0 0 0 0 0 0 9.E+04 4.E-05 3.75E-07 NA HO 99 395.5 307.5 239 185.8 67.75 11.5 0.25 1.E+03 6.E-07 9.75E-08 NA MO 101 1272.8 0 0 0 0 0 0 l.E+0S 6.E-05 3.00E-07 NA NB 97 596.75 246.75 92.25 34.5 0.75 0 0 7.E+04 3.E-05 1.48E-07 NA NB 98 1463.5 0 0 0 0 0 0 5.E+04 2.E-05 3.5DE-07 NA ND 147 53.75 51 48 45 35 22.5 9.5 8.E+02 4.E-07 1.33E-08 NA ND 149 263.25 0 0 0 0 0 0 2.E+04 I.E-OS 6.50E-08 NA ND 151 105.5 0 0 0 0 0 0 2.E+0S 8.E-05 2.50E-08 NA PD 109 3.5 1 0.25 0 0 0 0 5.E+03 2.E-06 8.75E-10 NA PM 151 41 23.25 23 7.25 0.75 0 0 3.E+03 1.E-06 1.00E-08 NA PR 143 69.5 88 97 100.3 91.25 - 65 31.75 7.Et02 3.E-07 1.70E-08 NA PR 144 5.5 5.5 5.5 5.5 5.5 5.5 5.25 1.E+05 S.E-05 1.38E-09 NA PR 145 638 40.25 2.5 0.25 0 0 0 8.E+03 3.E-06 1.58E-07 NA PR 147 572 0 0 0 0 0 0 2.E+05 8.E-OS 1.40E-07 NA RB 88 799 2.5 0 0 0 0 0 6.E+04 3.E-05 1.95E-07 NA RH 105 77.( 65.25 41.25 25.75 4 0.25 0 6.E+03 2.E-06 1.90E-08 NA RH 107 41 0 0 0 0 0 0 2.E+05 1.E-04 1.00E-08 NA RU 103 22.5 o22 21.75 21.25 19.75 17.5 13.75 6.E+02 3.E-07 5.50E-09 NA RU 105 187.5 4.75 0 0 0 0 0 1.E+04 5.E-06 4.50E-08 NA SB 127 6.5 6 5 4 2 0.5 0 9.E+02 4.E-07 1.63E-09 NA SE 81 53 0 0 0 0 0 0 2.E+OS 1.E-04 1.30E-08 NA SE 83 49.5 0 0 0 0 0 0 1.E+05 5.E-05 1.23E-08 NA SH 153 12.75 9 6.25 4.5 1 0 0 3.E+03 1.E-06 3.25E-09 NA SH 155 8.25 0 0 0 0 0 0 2.Et05 9.E-05 2.05E-09 NA SN 127 19.75 0 0 0 0 0 0 2.E+04 8.E-06 4.75E-09 NA SN 128 83.75 0 0 .0 0 0 0 3.E+04 l.E-05 2.05E-08 NA SR 89 28 28 27.5 27.25 25.75 23.25 19.25 1.E+02 6.E-08 6.75E-09 NA SR 91 796.25 138.75 24 4.25 0 0 0 4.x+03 1.E-06 1.95E-07 NA SR 92 1325.8 2.75 0 0 0 0 0 7.E+03 3.E-06 3.25E-07 NA

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• 1 CHAPTER 13 :1 ti TE 132 250.75 203 164.3 132.8 56.75 12.75 0.75 2.E+02 9.E-08 6.25E-08 NA SE 133 964.25 0 0 0 0 0 0 2.E+04 9.E-06 2.38E-07 NA TE 134 1712.5 0 0 0 0 0 0 2.E+04 I.E-OS 4.25E-07 NA XE 133 20362 24106 24010 22389 14111 5673 894.75 na 1.E-04 5.00E-06 NA XE 135 58955 30517 6594 1195 1 0 0 na 1.E-05 1.45E-05 1.5 XE 138 158548 0 0 0 0 0 0 na 4.E-06 4.00E-OS 10.0 Y 91 23.25 27.5 28 28 26.5 24.5 20.75 1.E+02 5.E-08 5.75E-09 NA Y 92 846.75 37.75 0.5 0 0 0 0 8.E603 3.E-06 2.08E-07 NA Y 93 861.25 169.75 32.75 6.25 0 0 0 2.E+03 1.E-06 2.10E-07 NA Y 94 1583.8 0 0 0 0 0 0 8.E+04 3.E-05 4.OOE-07 NA Y 95 1613 0 0 0 0 0 0 1.E+05 6.E-05 4.00E-07 NA Y 91M 423.5 88 15.25 2.75 0 0 0 2.E+05 7.E-05 1 .OSE-07 NA ZR 95 29.25 .29.25 28.75 28.5 27.25 25.25 21.75 1.E+02 5.E-08 7.25E-09 NA-ZR 97 657 245.5 91.75 34.25 0.75 0 0 1.E+03 5.E-07 1.60E-07 NA K-State Reactor 13.E-2 Original (1U104)

Safety Analysis Report