ML061010264

Kansas State Univ - Update to Chapter 11, Appendix a of the Proposed Safety Analysis Report for the K-State Triga Mark II Nuclear Research Reactor

ML061010264
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Site:	Kansas State University
Issue date:	03/30/2006
From:	Whaley P Kansas State University
To:	Hughes D Division of Regulatory Improvement Programs
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{{#Wiki_filter:Department of Mechanicsl and Nuclear Engineering 302 Rothbone Hall Monhtaton, KS 66506 -5205 785.532-5610 Fax: 785.532-7057 1'. Michael W\\'haley Kansas Statc University Nuclear Reactor Facilities Manager 1 12 Ward Hall Manhattan, Kansas 66506 Daniel E. Hughes, Project Manager Rcscarch and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation DATE: 30 March 2006

SUBJECT:

Update to Chapter 11, Appendix i of the Proposed Safety Analysis Report for the K-State TRIGA Mark II Nuclear Research Reactor

Dear Mr. Hughes:

A license renewval request was submitted for the Kansas State University in 2002, and the Clcility is currently operating under "timely renewal" provisions. USNRC review has resulted in Requests for Additional Information (RAI), previously addressed. Further clarification is provided in the attached and revised Chapter I1, Appendix A, of the Safety Analysis Report. The nuclear and physical characteristics of argon and nitrogen as used in calculations were moved from A.1.1, A.2.1, A.2.2, A.2.3, and A.2.5 to A.l, Introduction. Radiological Standards was revised to more clearly identify standards. The table "General Parameters" wvas revised (1) 1,250 kW in the title, (2) added physical dimensions related to core, (3) changed ventilation rate from "changes per hour" to flow rate. In the same section as the table, paragraph "reactor core parameters" was revised to indicate 4 rods and better define core components related to water volume in the core. The order of calculations was changed in A.2.2, and calculations in A.2.3 wvere expanded aid corrected. If you have any questions or comments concerning this matter, you may contact me at 785-532-6657 or wvhalevyEksti.cdu. I verify tinder penalty of perjury that the foregoing S

correct, Executed on 30 March 2006, oceN.5.

Whaley Docket No. 50- 188 g

Enclosures:

as indicated i IVMARCIA E. CHACON NOTARYPUBLIC STATE OF0 KANAS 3 / 30/.75006

Chapter 11 Appendix A Radiological Impact of 41Ar and 16N During Normal Operations A.1 Introduction Normal operation of the KSU reactor results in two potential source terms for radioactive gaseous effluent at significant levels, 4'Ar and 16N. There are variations in experimental configuration and possible scenarios where the production of 4'Ar may be different than the routine operations; these scenarios do not produce not long term, routine radioactive effluent but are assessed to determine if the amount of radioactive effluent is so high as to impact the annual exposure that might result from routine operations. The nuclide 4'Ar is produced by thermal neutron absorption by natural 40Ar in the atmosphere and in air dissolved in the reactor cooling water. The activation product appears in the reactor room (bay) and is subsequently released to the atmosphere through the reactor bay ventilation exhaust stream. The microscopic cross section for thermal neutron absorption in 40Ar is 0.66 barns, so the macroscopic cross section for thermal neutron absorption in 40Ar in air (0.0129 weight fraction) is 1i=1.54 x 10-7 cmnf. The saturated concentrations of 40Ar in water at the coolant inlet temperature of 270 C is approximately 6.1 x 10-5 g per cm3 of water (Dorsey 1940). If it is assumed that air is saturated with water vapor above the water tank (27 mm Hg vapor pressure at 271C) and that the mole fraction of argon in dry air is 0.0094, the partial pressure of argon in air above the tank is 0.0094(760 - 27) = 6.9 mm Hg. By Henry's law, the concentration of argon in water at the inlet temperature is 6.1 x 10-5 x 6.9/760 = 5.5 x 10-7 g cm-3 (C40 = 8.3 x 1015 atoms crn 3). Therefore, macroscopic cross section for thermal neutron absorption in 40Ar in water is 5.48 x 10- cmi'. With a half-life of 109.61 minutes, the decay constant for 41Ar is X = 0.379 he, 6.32 x 103 mid, or 1.05 x 10-4 sl. The nuclide '6N is produced by fast neutron interactions with oxygen. Although a portion of the '6N produced in the core is eventually released from the top of the reactor tank into the reactor bay, the half-life of 16N is 7.14 seconds; radiological consequences from effluent release of 15N are insignificant. The only source of '6N in the reactor that needs consideration results from interactions of neutrons with oxygen in the cooling water as it passes through the reactor core. Interaction with oxygen in atmosphere is relatively insignificant and is neglected in this analysis According to the McClellan AFB SAR, the effective cross section for the '6O(np)'5 N reaction, averaged over the fast-neutron energy spectrum in the TRIGA or over the fission-neutron spectrum is a,, = 2.1 x 10-29cm2. The 16N nuclide has a half-life of 7.13 seconds, corresponding to a decay constant X; = 0.0972 s-' = 350 h-' and emits, predominantly, 6.13-MeV gamma rays. K-State Reactor 11.A-1 Original (03/06) Safety Analysis Report

CHAPTER 11 APPENDIX A A.1.1 Purpose This appendix shows methods and calculations predicting production, concentrations, and dose rates from 4"Ar/' 6N associated with normal operation of the KSU TRIGA Mk. II nuclear reactor. ShoulJ a cladding failure occur during normal operations, a fraction of the fission products would be released to the reactor tank with the noble gases and halogens evolving from the pool to the atmosphere via building ventilation. This operational occurrence, taking place in air, is addressed in Chapter 13 as the maximum hypothetical accident for the TRIGA reactor. Neutron interactions in structural and control materials and materials irradiated for experimental purpo es result in the formation of activation products. These products are in the nature of fixed sources, mainly a source of occupational radiation exposure. Administrative controls preclude the significant formation of airborne activation products, other than the aforementioned 4'Ar. A.1.2 Radiological Standards Environmental Protection Agency publication, Federal Guidance Report No. 11 (Limiting Values of Raiionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion) is based on the International Council on Radiation Protection Publication 30 and the National Council on Radiation Protection Report 22. The report provides derived guides for controlling internal exposure to radionuclides in the work place, including the Annual Limit on Intake and the Derived Air Concentration (DAC), including dose conversion factor,. The Derived Air Concentration is used in IOCFR20 and defined: Deriv.ed air concentration (DAC) means the concentration of a given radionuclide in air which, if breatbed by the reference man for a working year of 2,000 hours under conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one ALI. DAC values are given in Table 1, Column 3, of appendix B to §§ 20.1001-20.2401. Reference man means a hypothetical aggregation of human physical and physiological characteristics arrived at by international consensus. These characteristics may be used by researchers and public health workers to standardize results of experiments and to relate biological insult to a common base. In addition, IOCFR20 Appendix B specifies limits for assessment and control of dose to the public, equivalent to the radionuclide concentrations which, if inhaled or ingested continuously over the course of a year (i.e., 8760 h), would produce a total effective dose equivalent of 0.05 rem (5 0 millirem or 0.5 millisieverts). Table A.1, Limits and Terms L Source/Quantity Value IOCFR20.1 101(d) 10 mrem TEDE from stack emissions IOCFR20 Appendix B, Derived Air Concentration 4'Ar 3 x 106 pCi cm 3 IOCFR20 Appendix B, Effluent Limit 41Ar I x IO-, pCi cm-3 EFA Federal Guidance Report I IConversion factor, 2.17 x 10.0 Sv h-' aclivity to dose for 41Ar submersion 0.803 mrem h' pCi ml" K-State Reactor 11.A-2 Safety Analysis Report Original (03/06)

RADIOLOGICAL IMPACT OF 41AR AND '6N DURING NORMAL OPERATIONS A.1.3 KSU TRIGA Design Bases General System Parameters The calculations for 4tAr and 16N releases during normal operations are based on the following system parameters. Flowrate(kgfs') = 0.01 15[P(kW)] 03 6 02 Table A.2, General System Parameters for Normal Operations at 1,250 kWt Full Power. Parameter Symbol Value Reactor steady power P 1,250,000 W Core coolant mass flow rate (per element)a mh 0.150 kg s-' Core coolant density p 1.0 g cm-3 Thermal neutron flux at E ring (core ave) 2.05 x 1013 n cm-2s-Fast neutron flux at full power at E ring (core ave) 3.00 x 10'3 n cm-2 s-' Thermal neutron flux in RSR cRSR 9.00 x 1012 ncm 2S 1 Total neutron flux per watt at fast (piercing) beam port 4250 n cm%2S-' (0.5 MeV avg) Total neutron flux per watt at tangential beam port 1400 n cm-2 S% (0.1 MeV avg) Fuel element heated length L 0.381 m Flow cross sectional area per fuel elementa A4 6.2 cm2 Mass flow rate per fuel element' a 108 g s-' Reactor tank diameter 1.98 m Reactor tank depth 6.25 m Reactor tank water depth above core 4.88 m (16 ft) Coolant volume in reactor tank VI 1.92 x 107 cm3 Reflector inner diameter (HSR) df 45.72 cm. (18 in.) Core volume V,,,re 62,550 cm3 (3817 in3) Fuel element heated volume VE 417.2 cm3 Coolant volume in cored V. 27,090 cm3 Air volume in reactor bay (144,000 ft3) Vhy 4.078 x 10 9 cm3 Reactor bay ventilation flow rate' wi' 4.17 x 105 cm3 s' (884 cfm) Air volume in rotary specimen rack VRSR 3.75 x 104 cm3 aSee §4.6 of this report. bSee §5.8 of Operations Manual. cSee § 13.2.2.2 of this report d Vcore -85* VE (91 positions, 83 elements, source, rabbit and 6 water filled positions) ' See letter B.C. Ryan (KSU) to Theodore Michaels (NRC) 15 Jan 99 Reactor Core Parameters Modeling of the reactor core for radiation transport calculations is based on the following approximations. For purposes of radiation shielding calculations the TRIGA reactor core may be approximated as a right circular cylinder 0.4572 m (18 in.) 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.) high and an aluminum grid plate 0.0191 m (0.75 in.) thick. In 91 fuel locations, there are 83 fuel elements, 4 standard control rods and 1 transient control rod, 1 water filled location, I central K-State Reactor 11.A-3 Original (03/06) Safety Analysis Report

CHAPTER 11 APPENDIX A thimble (water), 1 source & 1 pneumatic transfer site (water displaced). The fuel region may be treated as a homogeneous zone, as may be the axial graphite zones and the grid plates. Fuel clements are 1.43-in. (3.6 cm) ID and 1.47-in (3.7 cm) OD, clad with type 304 stainless steel'. Fuel density is 5996 kg m-3. Fuel composition is 8.5% uranium, 91.4% ZrH,.65. The uranitm is 20% 235 U and 80% 238U. Steel density is 7900 kg m-3. Standard control rods are 0.875.in. OD, the transient rod 1.25-in. OD. Both types of rods are clad with 30-mil thick aluminum (2700 kg mn3 density). The control material may be approximated as pure graphite, with density 1700 kg m3. 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 m-3 Reflector 1147 kg m-3 Grid Plate 2700 kg m-3 Composition of the three zones, by weight fraction, are given in the following table. Table A.2, Compositions of Homogenized Core Zones. Element Mass Fraction Element Mass Fraction Fuel Zone Axial Reflector Zone 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 Grid Plate U 0.0741 Al 1.0000 Reactor Bay Parameters For purposes of radiation dose calculations within the reactor bay, the dimensions are approximated as follows: The reactor bay is approximated as a right circular cylinder 36 ft (10.973 m) high and 36.68 ft (11.1 8 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 ft (3.3528 m) radius. The free volume is 144,000 ft3 (4C78 m3). 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. ' Composition, by weight, 2% Mn, 18% Cr, 8% Ni, balance Fe. K-State Reactor 11.A-4 Original (03/06) Safety Analysis Report

RADIOLOGICAL IMPACT OF 41AR AND "6N DURING NORMAL OPERATIONS A.2 Radiological Assessment of 4'Ar Sources A.2.1 Production of 4'Ar from Beams Operation with a fully open beam port is not a routine operational condition. Beam port operations normally have shielding, collimation and beam stops that prevent a full beam from penetrating the column defined by the beam port into air volume between the reactor and the reactor bay wall. Operating experience with neutron radiography performed at 10 kW involve, a neutron flux of 2 x 107 cMn2 s-' or less. We assume here that this is the flux density along ihe beam port, which has a cross sectional area of 324 cm2 (8-in diameter). In other words, in a radiography operation S = 6.48 x 109 neutrons per second enter the atmosphere essentially in a parallel beam. The maximum distance of travel of a neutron is from the reactor tank wall to The exterior wall of the reactor bay, namely, about Lb = 1020 cm. Thus, the activity concentration of airborne 4'Ar after sustained operation with an open beam port at 10 kW is given by: S

• A *(1e-b ) =4 Bq 3

C1) Vbay '(, + 2bqy) or 8.32 x 10-5 pCi mL-' in conventional units. Operations at maximum power are not performed for radiography, and radiography is not performed long enough to achieve equilibrium 4'Ar. Therefore, scaling the calculation for sustained operations at 1,250 kW provides an extremely conservative bound on 4'Ar production. Scaling the 10 kW 4'Ar production value to 1,250 kW results in 4.28 x 10-7 pCi mL-1 which is slightly above submersion DAC for occupational exposure; however, conditions for the source term are related to a very unusual set of conditions (open beam port with no shielding) that are not continuous in two respects. Shielding :-or radiography external to the beam port limits the beam to less than l2 of the analyzed volume. Radiography configuration is implemented only for radiography operations, a small fraction of all operations. Typically radiography occurs less than I day per month. Radiography operations are inherently discontinuous as the purpose of individual operations are met when the image is obtained. Typically a day of radiography operations involves less than 4 hours of operation at full power. These conservatisms assure limits are met with no further consideration. A.2.2 Production of 4'Ar in Rotary Specimen Rack The air volume in the rotary specimen rack does not freely exchange with the air in the reactor bay; there is no motive force for circulation and the rotary specimen rack opening is routinely covered during operation. If the rotary specimen rack were to flood, water would force the air volume in the RSR into the reactor bay. After sustained operation at full power, the equilibrium 4'Ar activity (Bq) generated in the RSR volume and diluted in the reactor bay atmosphere(Al"',4') is calculated (using terms defined in the Introduction and Table A.2) by: .4 I=4 VRSR (2) Bay K-State Reactor 11.A-5 Original (03/06) Safety Analysis Report

CHAPTER 11 APPENDIX A This is the initial concentration, while limits are based on annual exposure. With radioactive decay and ventilation, the concentration would decline in time with an effective decay constant deterrained by eff = 2way + A, - Using Xeff in equation (2) equation, time dependent behavior is: 4Ag; 4' (t) = 4 r_-e ( VJZS eAe (3) 4Bay (t =ay,1 =0 *e =,q VRSR eA VBay If a worker were exposed to the full course of the decay over one year, average concentration (pCi mILl) in the reactor bay would be calculated by the integration of equation (4) over 1 year: 1y ATfA-tVS JABaytO e0 e dt AAr4 l p RSR -Ar41 0 Bayt=O VBay A Bay y A 2 eon Y Aeff Y Considering only radioactive decay (4 of 0), equation (4) indicates an average concentration of 1.03. 10-7 plCi ml-' over the year. Considering radioactive decay in combination with stack flow (removing air), equation (4) indicates an average of 5.27 x 104 [pCi ml-. over the year. Since a worker will experience less than one Derived Air Concentration for 41Ar (3 x 1046 pCi/ml) for one year i a either case, the calculated exposure of a worker in the reactor bay for 2000 hours in a year is within limits. A.2.3 Production of 4"Ar from Coolant Water The reactor tank water surface is open to the reactor bay, and 4'Ar activity in the reactor tank water results from irradiation of the naturally occurring 40Ar as a fraction of air dissolved in the water. The activated argon is transported form the core to the pool, with a removal time constant (A7) that is a combination of the radioactive decay constant (A) and the fractional removal rate from the total convection flow through the core (iw'o, the product of the flow rate per element - a function of power - and 83 - the number of elements - divided by the volume of water in the core (V core): cor =2 + core (5) Vcore The activated argon is then transported to the reactor bay atmosphere, with a removal time constEnt (2Yff ) that is a combination of the radioactive decay constant (2) and the fractional removal rate from the reactor bay (effluent flow rate avdivided by reactor bay air volume Vbay).: K-State Reactor 11.A-6 Original (03/06) Safety Analysis Report

RADIOLOGICAL IMPACT OF 41AR AND 16N DURING NORMAL OPERATIONS fb = Al + (6)y bbay The rate of change in the 4'Ar concentration (number of atoms cm 3) in the core is a function of activation of 40Ar and removal of 41Ar through a combination of radioactive decay (X) and flowi: dN 4 1re -40 .O Core -core N 4 ' (7) dt At equilibrium (dN/dt = 0), the concentration of 41Ar (atoms cm-3) in the core is calculated using equation (7) as: N 4O (c (8) And the total core equilibrium inventory of 4lAr is: C 40 a) core ore core

• core

.9 Equilibration occurs rapidly because of the large,ff and time dependent behavior l1-exet I The rate of 41Ar inventory leaving the core ('eqre ) at equilibrium is the product of 41Ar volume averaged activity in the core (Icqre1 /Vcore) and volumetric flow rate (fvCore): jeq jeq core itg I ots = f? Wore(10) core V core core V core core Substituting the equability of equation (8) into equation (10): r 40 core* 40 Core core h.r F cIore =core Wc0 e (11) 7 V core en Not all of the 41Ar activity that leaves the core reaches the surface; some 41Ar decays because of the time required to transport the nuclide from the core 16 feet below the surface of the pool. Time delay in transport is calculated as the volume flow rate exiting the core divided by the total volume (vol) through which the material is transported; where vol is 3.76 x 106 cm3, the volume of water contained in the column 6.5 ft (99.06 cm) in diameter 16 ft (4.88 x 102 c:n) between the core exit and the pool surface. Transport time is a few seconds and decay can be neglected. Assuming 100% transfer from the pool to the bay atmosphere, the rate of inventcry removal from the core through the pool is the rate of addition of 41Ar to the reactor bay atmosphere: K-State Reactor 11.A-7 Original (03/06) Safety Analysis Report

CHAPTER 11 APPENDIX A z40 cre jeq Ieq z40 c icore = Zbe7 =Wcore (12) eff The change in total radionuclide inventory (1) in the reactor bay is calculated in a similar manner as the core, using the rate of contributions to inventory released into the reactor bay atmosphere (essentially all of the inventory removed from the core) and the rate of removal from radioactive decay and transport via reactor bay exhaust: bq = lb -'< -b"(13 dtl bay ~byJ (13) Using equation (10) and equation (11): dl bay fV.(~oe 2 bay. dt Ae 4core A

• eff bay (14)

For equilibrium conditions (i.e., dIbWdt = 0), F40.qDCor-I =n S

• ¢t (1l5) bay re 2 Ibaykoe(5 Ref O-e The concentration of activity in the reactor bay is determined by the concentration in the bay (IlVbay) and the decay rate (X for 41Ar); based on equation (15), the specific activity in the reactor bay almosphere (.Abeaqy

) is: Aeq = r wc(re (16) ecoe 2 bay [I ef ~eff Va Note that if ventilation is secured, Xby= 0, and equation (15) becomes: i4 DCore Abeq = ra'th Wcore (17) 'bay

oe Vbay effbq Using equation (16), the specific activity of 41Ar in reactor bay air is 0.378 Bq ml-' (1.02 x I0-5 pCi rrl-). With a Derived Air Concentration for 41Ar of 3 x 1046 pCi/ml, if the reactor is operated continuously at full power for a year then it will be necessary to assure that an individual does not occupy the reactor bay during operations for more than 600 working hours; not limiting in a practical sense as (1) the assumed operating schedule is unrealistically conservative and (2) the reactor bay is not routinely occupied continuously during high power operations.

K-State Reactor 11.A-8 Original (03/06) Safety Analysis Report

RADIOLOGICAL IMPACT OF 41AR AND 16N DURING NORMAL OPERATIONS A more realistic approach possible within facility constraints considers a series of irradiations followed by decay until the start of the next irradiation. In this approach, the builcup to equilibrium 4'Ar is a significant consideration. Buildup to saturation activity is asymptotic: AQt) = A.> (I (- e ') In the core, the dominance of the large flow term (compared to core volume) for the effective time constant drives core activity to saturation very quickly, while the exchange rate of reactor bay air has the same order of magnitude as the radiological time constant for 4'Ar. Consequently, only the effective time constant for the reactor bay needs to be considered in determining time dependent behavior: Al (t) = Aeq * (1 - e-7 ) Reactor bay activity over the decay interval between termination of operations (Ti) and initiation of the next operation is: A2 (t)= Aeq * (1-e "' '). e~ The average activity over the irradiation and decay intervals is the integral of the two intervals over the total time: A = 1 Il(l-e )dt+(l-e f eT) Je 'ff'dtl bayq T2 0 T, A 1 2bay I T +ef (et )(.. .r) Aby T2

• eff Since the radiological time constant is the same order of magnitude as the effective time constant for 41Ar removal from the reactor bay, the results are not very sensitive to changes in reactor hay flow rate. For periodic irradiation and decay on a 24-hour cycle, T2 = 24 h. Figure A.1 shows how the fraction of saturation activity is affected by a single 24 hour irradiation cycle; for one year of daily irradiations on the order of 8 hours, the DAC is met. If the irradiation cycle is 5 days per week instead of seven (71% of available time), the DAC will be met with irradiations on the order of 12 hours each day.

Since fuel inventory, staffing, and operating practices limit (1) continuous full power operations and (2) continuous occupancy of the reactor bay, access to the reactor bay does not require additional controls to ensure individuals remain within exposure limits based on DAC values. Administrative requirements on reporting and experiment review are adequate to identify approaches to limiting conditions of exposure. K-State Reactor 11.A-9 Original (03/06) Safety Analysis Report

CHAPTER 11 APPENDIX A 24 Hour Interval Irradiation & Decay Activity vs Irradiation Time 1. I). I). 1). At.. I.) '1

zz:zzzzzz 8

i.-tii 7 6 ((-j-r--- 4:zz --V.- I-..  ii -{--- ____ III- - O.: 0., 0). 0 5 10 Irradiation Time 15 20 Figure A.1, 24 Hour Interval Irradiation & Decay Activity vs Irradiation Time Note that setting fv Xwbay to zero shows the effect on concentration with ventilation securnd. Although the exhaust fan operates to maintain a negative pressure in the reactor bay when the reactor is not secured, an upper bound for effluent contamination of 0.746 Bq ml' (2.01 x l0-5 pCi ml-) is calculated using equation (17), assuming no ventilation (i.e., Xay=O)- A.2.4 Maximum Impact of 41Ar Outside the Operations Boundary Althoigh there are three modes of 4'Ar production, only the release of radioactive argon dissolved in water occurs routinely. The 4'Ar produced in the reactor bay during normal operal ions is released to the atmosphere via an exhaust fan at approximately h = 11 meters above grade. Where Q is the rate of radionuclide release, U is the mean wind speed (m s-1), e = 2.718, and Cy and C. are diffusion parameters in the crosswind and vertical directions respectively, the maximum downwind concentration (pCi cm-3), at grade, may be computed using the Sutton formula (Slade 1968): Cr = 2.Q C. e*,r-u-h' C, (18a) In tenns of the concentration of the radionuclide in the reactor bay: K-State Reactor Safety Analysis Report 11.A-1 0 Original (03/06)

RADIOLOGICAL IMPACT OF 4MAR AND 16N DURING NORMAL OPERATIONS 2.Aeq *' C. Cmax = 2 C (13a) The ratio of the maximum concentration in specific meteorological conditions to the concentration at the point of release is therefore: C 2 -w C.

CAbq,

= e-'rhJ C u (J.9) The maximum concentration downwind occurs at distance d (m) given by: 2 d = (h / C,) 2-n, (:0) in which the parameter n is associated with the wind stability condition. The McClellan AFB SAR provides parameters associated with Pasquill stability classifications (n, C,) and data from Chapter 2 were used to infer mean wind speeds (iT) by stability class (Class A, C, E, and G). The McClellan (site-specific) data has been modified to provide data for Class B, D, and F; graphical representation of standard deviations of material in a plume from Meteorology and Atooic Energy (1968, D. H. Slade), UASAEC and Regulatory Guide 1.111 show predictable, uniform behavior across each stability class in the region of interest. Therefore the McClellan n, C. and C, data was used to derive the data for the remaining classifications by fitting to equations (assigning x=l to Class A, x=2 to Class B, etc.). n =0.00125x 3 -0.0075 x2 +0.03875-x+0.1675 ('1) C, =-0.01-x3 +0.1287-x 2 -0.465-x+0.6562 C, = =-0.00I-x 3 +0.02-x 2 -0.1465x+0.4375 (2.2) (23) The product of (1) the maximum concentration for a classification and (2) the frequency of occurrence of the classification is the average concentration at the receptor location (where the maximum concentration occurs), with other locations bounded by this calculation. n I~ Class A 1.600 0.200 0.31 0.31 53 5.OE-04 Class B 3.205 0.230 0.16 0.22 84 3.5E-04 Class C 4.000 0.250 0.15 0.15 134 2.OE-04 Class D 4.074 0.282 0.22 0.11 224 9.9E-05 Class E 3.500 0.330 0.30 0.075 393 5.8E-05 Class F 2.382 0.401 0.34 0.051 830 5.1E-05 Class G 0.770 0.500 0.28 0.035 2137 1.3E-04 freqtr 0.61% 4.50% 10.63% 53.21% 10.60% 11.82% 8.64% 3.1 H-06 I.6LE-05 2.5]3 -05 5.3] -05 6.1 E-06 6.013-06 1.11,-05 K-State Reactor Safety Analysis Report 11.A-11 Original (03/06)

CHAPTER 11 APPENDIX A The maximum product of C/Ag and the frequency of occurrence is 5.3 x 10-5; the annual dose from exposure at that location is 7.5 mrem (5.3 x 10-5*20.1 pCi*0.803 mrem h-' pCi-'*8760 h y'), less than the permitted annual dose from gaseous effluents of 10 mrem. The assumption of continuous full power operation to establish equilibrium 41Ar concentration is extremely conservative, since (as previously noted) staffing and fuel inventory will not support continuous full power operations. In addition, administrative requirements on reporting and experiment review are adequate to ensure average annual discharge meets effluent limits. A.2.5 Radiological Assessment of 16N Sources Nitrogen-16 is generated by the reaction of fast neutrons with oxygen and the only significant source results from reactions with oxygen in the liquid coolant of the reactor. The atomic density CN (c:M-3) of the nuclide as it leaves the reactor core is given in terms of the oxygen density in water., CO = 3.34 x 1022, as CN = Of ° n

• (-e

(20) 216 where time in the core is represented by t. Fast-neutron flux varies linearly with reactor power. Time in core is a function of convection flow rate, a function of reactor power (see Chapter 4). As power increases, the rate of production increase from increased neutron flux is mitigated by a reduced time in the core from the increase in core cooling flow rate. As the warmed coolant leaves the core, it passes through 1.5-in diameter (AGu = 11.4 cm2) channels in the upper grid plate, but the triflute upper end fixture of the fuel element restricts the flow. This leaves a flow area for each element of: AO = A, (IJ *sin3O 0 *cos300] = 6.69cm2 (21) Operation at power requires primary cooling; primary cooling enters the pool through a flow divert.r approximately 2 feet (61 cm) above the core exit, 14 feet (427 cm) below the pool surface. Core exit is at 16 feet (488 cm) below the pool surface. The flow diverter induces mixing and avoids the direct rise from the core to the pool surface (which could otherwise occur through a chiraney effect from core heating). A rough estimate of hydraulic diameter of the core exit (based on total flow area) is about 13 cm; calculations show the contributions to total dose rates at the pool surface are negligible at 160-200 cm below the surface of the pool, 22-25 times the hydraulic diameter of the exit into the pool. Exit flows are a small fraction of mixing flow, and under these conditions it is considered adequate to use a nuclide concentration reduced by the ratio of the total core exit surface area (approximately 555 cm2 for 83 elements) and the pool (with a total surface area of approximately 30900 cm2); mixing reduces the concentration of '6N from the core exit by 0.018. Therefore, concentration of the radionuclide used in calculation is reduced from core exit by dilution. K-State Reactor 11.A-12 Original (03/06) Safety Analysis Report

RADIOLOGICAL IMPACT OF 41AR AND 16N DURING NORMAL OPERATION4S Because of the short '/2 life, the concentration of 16N is also reduced by decay during transit. Since it is difficult to characterize flow velocity field from core exit to total mixing, flow rate from the core to the surface is conservatively assumed as core exit flow rate for dose rate calculations. Dose rate calculations were modeled as a set of disk sources, each disk containing the appropriate volume source term multiplied times the difference between the disk locations. The appropriate volume source strength for each disk source calculation was modified by exponential decay of 16N, with the time element calculated from core exit surface area, flow rate, and distance form the core exit. Does rate calculations were based on the two major emissions, 6.13 MeV (69%) and 7.11 MeV (5%). Total dose rate at each disk (where x is the distance form the disk to the pool surface) was therefore calculated as: b ZE [k(E)*E*SvAd *Ai*(E l (px))-(E(plx*sec )0 (22) Where: k(E) R MeV* cm-' *s-( _S0(flow)

• ex i
• X
• A channel (see Chapter 4 for coolant flow rate) mH
• P A, Taylor buildup factor Pi linear attenuation coefficient, mod ified byTaylor buildup factor a;
• 0= arctan Parametric variation on the distance between the disk sources showed little improvement in convergence for separations smaller than 2 cm, and essentially no improvement below 1 cm; therefore l/2 cm was used for final calculations. Locations of interest for dose calculations include 30 cm (1 ft) above the pool surface (i.e., pool surface monitor), waist high (approximately 130 cm/51 in. above the pool, 100 cm/39 in. above the bridge), and at the ceiling over the pool.(549 cm/I 8 ft above the pool).

Only a small proportion of the 16N atoms present near the tank surface are actually transferred to the air of the reactor bay. Upon its formation, the '6N recoil atom has various degrees of ionization. According to Mittl and Theys (1961) practically all '6N combines with oxygen and hydrogen atoms in high purity water, and most combines in an anion form, which has a tendeacy to remain in the water. In this consideration, and in consideration of the very short half life of the nuclide, the occupational consequences of any airborne '6N are deemed negligible in comparison to consequences from the shine from the reactor tank. Similarly, off-site radiological consequences from airborne '6N are deemed negligible in comparison to those of 4mAr. K-State Reactor 11.A-1 3 Original (03/06) Safety Analysis Report

CHAPTER 11 APPENDIX A Table A.4, Dose Rate (mR h-') Above Pool KW 30 cm 130 cm 549 cm 50 0.5 0.2 0.0 100 3.5 1.1 1.3 200 15.8 4.7 0.5 300 35.7 10.5 1.2 400 60.0 17.5 1.9 500 87.2 25.3 2.8 750 166.3 47.7 5.2 1000 255.3 72.8 7.9 1250 347.9 98.8 10.6 A.3 Bibliography EPA Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration, and Dose Conversion Factorsfor Inhalation, Submersion, and Ingestion," U.S. Envir3nmental Protection Agency, Report EPA-520/1-88-020, 1988. Kansas State University TRIGA Mark 11 Reactor Hazards Summary Report, by R.W. Clack, J.R. Fagart, W.R. Kimel, and S.Z. Mikhail, License R-88, Docket 50-188, 1961. "A na'ysis of Certain Hazards Associated with Operation of the Kansas State University TRIGA Mark lIReactor at 250 kWSteady State and with Pulsed Operation to $2.00, " by R.W. Clack, et al., arnd the Safety Evaluation by the U.S. Atomic Energy Commission Division of Reactor Licensing, License R-88, Docket 50-188, 1968. Operations Manual, KSU TRIGA Mark II Nuclear Reactor Facility, License R-88, Docket 50-188. Facility Safety Analysis Report, Rev. 2, McClellan Nuclear Radiation Center Reactor, April 1998. Mittl, R.L. and M.H. Theys, "N-16 Concentrations in EBWR," Nucleonics, March 1961, p. 81. Slade. D.H. (ed.), "Meteorology and Atomic Energy," Report TID-24190, U.S. Atomic Energy Commission, 1968). K-State Reactor SafetV Analysis Report 1 1.A-14 Original (03/06)}}