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Purdue University - Request for Additional Information Regarding the Purdue University Reactor License Renewal Application (TAC No. ME1594), Responses to Letter Dated August 29, 2014 (ML14115A221). Part 1 of 5
	ML15210A282
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Site:	Purdue University
Issue date:	07/24/2015
From:	Bean R S; Purdue University
To:	Montgomery C K; Office of Nuclear Reactor Regulation
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ML15210A279	List: ML15210A280; ML15210A282; ML15210A283; ML15210A285; ML15210A287; ML15210A288;
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	TAC ME1594
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Purdue University Responses to PURDUE UNIVERSITY -REQUEST FOR ADDITIONALINFORMATION REGARDING THE PURDUE UNIVERSITY REACTOR LICENSE RENEWALAPPLICATION (TAC NO. ME1594), RESPONSES TO LETTER DATED August 29, 2014.1. RAI 18 in NRC letter dated July 6, 2011, stated:TS 4.3: TS 4.3(c) should reference the minimum 13 foot depth as specified in theLCO [limiting condition for operation] (TS 3.3(c)) for primary coolant and providedin the bases for TS 4.3. Please update the TS to include the numerical minimumdepth and surveillance interval for this surveillance or justify why an alternativemeasure related to the height of the skimmer trough in TS 4.3(c) is moreappropriate for specifying the minimum performance level of TS 3.3(c).Additionally, prescribe the frequency, scope and minimum water level of thissurveillance when the reactor is secured or shutdown or justify why a minimumlevel is not required.The response to RAI 18 by letter dated January 4, 2012 (ADAMS Accession No.ML1 2006A1 93), proposed to modify TS 4.3(c) to specify that reactor pool water will be ata height of the 13 foot over the top of the core whenever the reactor is operated.However, the response to RAI 17 by letter dated January 30, 2012 (ADAMS Accession No.ML12031A223), proposed to modify TS 4.3(c) to specify the reactor pool water will be ator above the height of the skimmer. Clarify TS 4.3(c), ensuring that RAI 18 is answeredclearly and completely. Include a basis for any TS changes proposed.Response:References to the skimmer trough have been removed from the Technical Specifications.TS 4.3(c) will be modified to state:(c) The reactor pool water will be at a height of 13 feet over the top of the core wheneverthe reactor is operated. The reactor pool water height shall be visually inspectedweekly, not to exceed ten days, and water will be added as necessary to reach thespecification.And the basis for TS 4.3(c) will be revised to state:When the reactor pool water is at a height of 13 feet above the core, adequate shieldingduring operations is assured. Experience has shown that approximately 35-40 gallonsof water will evaporate weekly and weekly water make-up is sufficient to maintain thereactor pool water height.A revised copy of the Technical Specifications, Amendment 13 (draft) has been attached.2. RAl 29 in NRC letter dated July 6, 2011, stated:Page 1 of 29 TS 6.1.11, TS 6.1.14: ANSI/ANS-15.1-2007, Section 6.1.3(3) provides guidance forevents requiring the presence at the facility of the senior reactor operator. Pleaseupdate PUR-1 TS 6.1.11 and 6.1.14 for compliance with the requirements in ANSI/ANS-15.1 -2007, Section 6.1.3(3) and 10 CFR 50.54(m)(1) or provide an explanationdescribing your reason(s) for not incorporating the changes.The response to RAI 29 by letter dated January 30, 2012, proposed a modification of TSs6.1.11 and 6.1.14 that does not meet the requirements of 10 CFR 50.54(m)(1) which state:A senior operator licensed pursuant to part 55 of this chapter shall be present atthe facility or readily available on call at all times during its operation, and shall bepresent at the facility during initial start-up and approach to power, recovery froman unplanned or unscheduled shutdown or significant reduction in power, andrefueling, or as otherwise prescribed in the facility license.The response proposed a change to TS 6.1.2 (Staffing), Item (3), stating:(3) Events requiring the presence at the facility of an senior reactor operator[SRO] are:(a) Initial startup and approach to power following a core change. Thepresence of an SRO at the reactor facility is unnecessary for the initial dailystart-up, provided the core remains unchanged from the previous run;(b) All fuel or control-rod relocations within the core region;(c) Recovery from an unplanned or unscheduled shutdown except ininstances which result in the following(i) A verified electrical power failure ...;(ii) Accidental manipulation of equipment in a manner which doesnot affect the safety of the reactor;(iii) A verified practice of the evacuation of the building initiated bypersons exclusive of reactor operations personnel.Provide an explanation on how the proposed changes (a) and (c) are in compliance withthe requirements of 10 CFR 50.54(m)(1).Response:According to 10 CFR 50.54(m)(1), a senior operator is required for initial start-up and approachto power. TS 6.1 .2(3)(a) requires the SRO to be present for the initial startup and approach topower any time the core has been changed. After this, as the core has not changed and so theexpected critical rod heights are known, a senior operator is not required to be present forsubsequent operations, including start-up. If the core configuration is changed, a senioroperator is again required for the initial start-up and approach to power.As 10 CFR 50.54.(m)(1) does not recognize any exceptions for the required presence of asenior operator for recovery from an unplanned or unscheduled shutdown, TS 6.1 .2(3)(c) will berevised to state:(c) Recovery from an unplanned or unscheduled shutdown.Page 2 of 29

3. RAI 44 in NRC letter dated July 8, 2011, stated:In your RAI response concerning decommissioning cost, dated June 4, 2010, youreference an "approved cost estimate for decommissioning under PurdueUniversity's Radioactive Materials License" as the basis for the provided costestimate. Please describe and explain the relationship between the decommissioningcost estimate for Pun-1 and cost estimate for decommissioning under PurdueUniversity's Radioactive Materials License in determining the cost estimate fordecommissioning Pun-1.Provide a response to RAI 44, since we have not yet received one.Response:A rewritten Chapter 15 is attached and will be included in the Safety Analysis Report.4. RAI 45 in NRC letter dated July 8, 2011, stated:Pursuant to 10 CFR 55.59(a)(2), each licensee shall: "Pass a comprehensiverequalification written examination and an annual operating test." In your RequalificationPlan, Section B you state that "completion of the biennial requalification program willconsist of a written examination and a demonstration of operator proficiency in reactoroperation."A. Explain how the facility ensures that operator proficiency examinations areperformed annually during the biennial requalification cycle in compliance with 10CFR 55.59(a)(2) or update your plan accordingly.B. As required by 10 CFR 55.53(h), licensees are required to complete arequalification program as described by 10 CFR 55.59. The regulation in 10 CFR55.59(a) states that each license shall:(1) Successfully complete a requalification program developed by the facilitylicensee that has been approved by the Commission. This program shall beconducted for a continuous period not to exceed 24 months in duration.(2) Pass a comprehensive requalification written examination and an annualoperating test.Section F of the Pun-1 Requalification Plan states:During intervals when the licensed operations crew consists only of senior operatorswho are instructors for topics in part a.1.b., the requalification program will be modifiedto exempt those senior operators from parts A and B.1. Parts B.2, C, D, and E will remainin effect.Page 3 of 29 When the licensed operations crew increases to include those who do not instruct in theprogram, the program will revert to its initial content. Operators may place a statementinto the file stating that they have done a literature review andlor instructed the topics inSection A and B.1 in lieu of meetings and exams.During intervals when the licensed operations crew consists of only one senior operatorthis operator will be exempt from parts A and B, part C would be documented in theconsole logbook and as stated in C.3, parts D and E will remain in effect.In any of the requalification activities, exclusive of operations, additional methods maybe used to accomplish the training requirement. These may include mail, electronicclassroom or other methods may be used for training, meetings, testing or other requiredcommunication(s).The response to RAI 45 by letter dated January 31, 2012 (ADAMS Accession No.ML14234A1 09, redacted version) indicated that the Pun-1 facility previously had anexemption and intends to request one again. Either (a) explain how this section, in itscurrent form, meets the requirements of 10 CFR 55.33, "Disposition of an initialapplication" and 10 CFR 55.59; (b) delete this section of the requalification plan; or (c)submit an exemption for these requirements in accordance with 10 CFR 55.11, "Specificexemptions."Response:The requalification program has been revised to clarify that the operating test anddemonstration of proficiency is an annual requirement.Section F of the PUR-1 Requalification Plan will be deleted. Purdue University will not beseeking an exemption for the requalification program.A revised requalification program has been attached.5. RAI 63 in NRC letter dated July 14, 2011, states:Major inconsistences are noted throughout the SAR [safety analysis report]related to calculation assumptions for initial and requested maximum licensedpower under the Pun-1 license renewal. For example, SAR Section 13.2.2, p. 13-11, references current licensed power of 1 kW (kilowatt] for a reactivity insertionwith scram. Please clarify the desired maximum licensed power level requestedand ensure this power level, including any uncertainty in reactor power, isconsistently applied in the safety analyses for the license. Please provide anupdated evaluation of a safety analysis that explains all analyses, assumptionsand conclusions at the requested maximum licensed power level.Additional clarification to RAI 63 is needed. Provide responses to the questions below:(a) Provide an answer to RAI 63 or indicate if the answer to RAI 63 is provided in theresponses to RAIs 62 and 65 in your letter dated January 31, 2012.Page 4 of 29 (b) Table 4-21 in the revised PUR-1 SAR, Section 4.6 indicates that for 1 kW and 12kW, the maximum fuel temperatures for the limiting fuel plate (31.92 and 39.1degrees Celsius (C)) are lower than the maximum clad temperatures (43.42 and43.4 degrees C). Discuss the physical phenomena that would result in thesetemperature values.Response(a) All inconsistencies regarding the power level have been addressed throughout the SARand updated to 12 kW and 12kW+50% power uncertainty where appropriate.(b) The axial temperature rise along a fuel element is plotted below. Due to the high thermalconductivity of the metallic fuel, small fuel meat thickness, and the lack of a gap betweenthe fuel and the cladding of this plate design, the cladding and fuel exist at essentiallythe same temperature. The temperature of the wall never falls more than 0.02 °C belowthat of the fuel.......................... Temperature of Coolant Across Fuel ElementA AAA-.40 A_38 ------.----- 1 kW+50% -CoolantA*,3 -------- A__ *l1kW+50% -FuelS A --

12 kW+50% -Coolant34 A 12 kW+50% -Fuel30 Se0 0.1 0.2 0.3 0.4 0.5 0.61. Height (in)An updated Table 4-21 (from the SAR) is shown below. Note that when the NATOON modellingtool is moved up to 98.6 kW, the power level for the onset of nucleate boiling (ONB), themaximum coolant temperature is 44.8 °C.Page 5 of 29 Present Power Uprate Power ONB PowerPower Level 1 kW+50% 12 kW+50% 98.6 kWMax. Fuel Temp. (°C) 32.28 43.20 112.61Max. Clad Temp. (°C) 31.28 43.19 112.50Coolant Inlet Temp. (°C) 30.0 30.0 30.0Coolant Outlet Temp. (°C) 31.6 35.4 44.8Margin to incipient boiling (°C) 78.49 68.12 0Coolant Velocity (mm/~s) 5.41 19.16 56.20Coolant Mass Flow Rate (kglm2s) 5.39 19.04 55.696. RAI 70 in NRC letter dated July 14, 20)11, stated:NUREG-1537, Section 11 provides guidance for radiation protection provisions atthe facility. In Section 4.4 of the SAR, it is stated that the radiation level above thereactor pool surface is about I mrem [millirem]/hr and that the radiation levelalong the outside lateral surface of the concrete biological shield is about 0.1mrem/hr, when the core is operating at 1 kW. Please provide an updatedevaluation of a safety analysis that explains all analyses, assumptions andconclusions at the requested licensed power level for the maximum potentialradiation levels and the potential radiation effects on facility staff. As part ofevaluation, please indicate if the radiation levels bound those that would beencountered during fuel handling and maintenance operations. Additionally,include an evaluation of the safety analysis for potential dose to the facility staffand members of the public (i.e., classrooms, hallways, adjacent rooms, nearestdormitories, offices, etc.).The response to RAI 70 by letter dated January 31, 2012, did not address the expectedradiation dose levels at the requested increased licensed power level of 12 kW.-Theupdated safety analysis you indicate bounding dose levels for facility staff and membersof the public who may be located in nearby or adjacent, accessible public areas duringthe maximum operation power level for an extended period (i.e., classrooms, hallways,adjacent rooms, nearest dormitories, offices, etc.) to demonstrate compliance with 10CFR Part 20. Include all analyses, assumptions, and conclusions and indicate if thePage 6 of 29 radiation levels bound those that would be encountered during fuel handling andmaintenance operations.ResponseThe dose rate in the air above the reactor pool is given byD =i(~~iWhere ci is the flux of photons at a given energy, z is the height above the core, and P~air iS themass absorption coefficient for photons of the given energy. The flux of photons is the primaryquestion in determining the dose rate. The flux is essentially the attenuated dose at a distancer [cm] from the core, multiplied by a buildup factor.ci)(z) = f ci(z, E)dE =f 4--v2T,(E)B(lMr, E)e-gr dEWhere Sv~ is the photon source rate, fl(E) is the flux of photons with energy between E andE+dE, B is the buildup factor, p is the attenuation coefficient, r is the distance form the core, andE is the energy of the photons. The photon attenuation coefficient is a function of energy andfor water at 2 MeV, /.i = 0.0493 (Attenuation data taken from the National IsiueoStandards and Technology). For a shielding thickness of 395 [cm] (the height of the waterabove the core), the quantity/MR is 19.5 and represents the number of mean free paths thephoton must travel through to be emitted through the top of the pool. Note that this is the dosedirectly above the reactor and would be further reduced at the edges of the pool as there ismore attenuation.The buildup factor B(/Mr, E) has multiple forms one of which was reported by J.J. Taylor and hasthe form of a summed exponentialB(Mtr, E) = A(E)e-al(E)*MR + [1 -A(E)]e-a2(E)*PtRThe coefficients A(E), a1(E), and a2(E) have no physical meaning and are evaluated in tablesreported by multiple sources. For a 2 [MeV] photon,B(.MT,E) = 12.612

e-0'0532.19"5 .+/- [1 -12.612]e-°'1932.i9'5 27.7The number of fissions in the reactor is bounded by a 12 kW operating power plus 50% ofpower uncertainty. The number of fissions at full power is1000 W 1 )- 6.2415 X 1012 MeV 1 fission _0___ision18 kW = l8 kW* .*- *m * -5.617x i0x fsin1 kW 1 W l]oule 200 MeV secPage 7 of 29 As an example of the calculation, at 2 [MeV], there are 1.8 [l] h MV lxa hsurface of the pool is therefore5.617 x i0'4 [fissions.]cI'(395 [cm], 2 [MeV]) L i7 95 c2

1.8

27.7 *c(395, 2) =47.5 [cm'e]Summing this expression across available energy values will give the total photons emitted fromthe reactor pool surface. This value is 1732 over all energy groups (0-0.5 MeV, 0.5-2MeV, etc.). The equivalent dose rate in air is then 3.25 Dose RateEnergy (MeV) m tRero/hr)0.5 ,,1.338x1062 0.1414 1.6816 1.1838 0.&239""Total: 3.245 mRem/hrThese dose levels bound those that would be encountered during fuel handling operations asfuel remains at the bottom of the pool until it has cooled to levels which are acceptable withinbounds of the TS.For a dose to a member of the public in an unrestricted area, the unshielded dose reduces asthe inverse square law. The closest such a person could get to the reactor pool is at minimum 5meters. Treating the dose rate at the pool surface above the core as a point source, the sourcestrength would be reduced asS 3.245 mRI'-47r2 -4m**52 0.O010m---hrWhere S is the dose rate at the surface of the pool, and r is the distance from the pool surface.A very large person may have a frontal area of 1 m2 which would give a dose rate of 0.010 mR/hr which is less than that of the regulations in 10 CER 20. Note that this does not take intoaccount extended benefits from the shielding by air, walls, and the ceiling.Page 8 of 29

7. RAI 71 in NRC letter dated July 14, 2011 stated:The requirements of 10 CFR 20.1101 states that each licensee shall develop,document, and implement a radiation protection program commensurate with thescope and extent of licensed activities, in order to limit the total effective doseequivalent to facility workers (annual occupational dose less than 5 rem[roentgen equivalent man]) and the total effective dose equivalent to individualmembers of the public (annual public dose less than 100 mrem). Please provide...a safety analysis that explains all analyses, assumptions and conclusions at therequested licensed power level for the maximum potential estimate of the totalannual production of argon-41 from PUR-1 normal operations. In addition, pleaseevaluate and discuss the potential maximum dose to a facility worker and to amember of the public (i.e., classrooms, hallways adjacent rooms, nearestdormitories, offices, etc.) due to this bounding yearly production and release ofargon-41 from the facility.Your response to RAI 71 by letter dated April 10, 2013 (ADAMS Accession No. ML13101A044), did not provide an analysis to determine the maximum effective dose to themaximally exposed member of the public for the total annual production of argon-41from PUR-1 maximum licensed power operations. Provide a bounding safety analysis(with assumptions and conclusions) for the member of the public residing outside thefacility perimeters.ResponseThe buildup of 41Ar is due to the absorption of a thermal neutron by 4°Ar. Gasses are naturallypresent in fluids like water and the amount is dependent on temperature as well as the partialpressure of the surrounding volume. Henry's Law dictates the amount of various gasses which[ tomol 1 at standardare dissolved in fluid and carries a Henry's constant of 1.4 x 10- [m3Pajtemperature and pressure conditions for natural argon. (Sander, 2014). Because solubilitydecreases with room temperature, this is a conservative estimate as the temperature of the poolfor the rest of the analysis has been at or above 20 °C. With a pressure of 1 atm at the top of thepool, the number of Ar atoms dissolved per cubic centimeter ismmZ aosll 10 [in__3]___1.4x 1-'[m-----l]a

1035Pa

6.022 x1023 ratoms]

10 =__ 8.543 x _0___a__~c34°Ar makes up 99% of all natural Argon, so the number of 4°Ar atoms per cubic centimeterwould be corrected as8.4 07[atoms] [ 4°Ar]1 [atoms]8.53 [ 1017 1* .991 [A It 8.457 x 1017 [mL cm3A J aPage 9 of 29 The amount of time that the coolant spends within the core will be its irradiation time. TheNATOON code, used in response to RA! #5 above, predicted a mass flow rate of6.6chgamnl--se r .8 tcmanlsc " The 13 standard fuel elements have 15 channels and the3 control elements have 11 which yields 228 channels and a total volumetric flow rate of1564.08 [cm 1. The recirculation time of this portion of the pool water through the core will beL sec Jgiven by% -VPoo1cicVcoreThe pool has a radius of 4 [ft] or 121.92 [cm] and a height of 17 [ft] or 518 [cm]. The volume of2.42 x i07 [cm3] gives a recirculation period of2.42 x 107 [cm3]Tcirc =rc3 1 1.547 x 104~ [see] 4.3 [hours]1564.08 t scm3Using the MCNP6 model, the thermal neutron flux at 18 kW (12 kW + 50%) is predicted asbeing cp 2.66 x 1011 [neutrons] within the core volume. This value is conservative as theL crn2sec Jthermal neutron flux is usually considered linear with power increase which would suggest a fluxof c = 1011 [ neutrons]The saturation activity of 41Ar will be that which would be normally produced and decayed whilein the reactor volume, reduced by that which decays while circulating throughout the poolvolume.N crth d(1-- e-at )Asat =(1 -e-a(t+Tcirc))Here, N is the number of atoms found within the core at any given time, oth is the thermalneutron absorption cross section for 4°Ar (oath 6.1 x 10-25 cm2), t is the decay constant(1.05 66 x 10-4 [sec-'], and t is the transit time through the core. With a core height of60.96 [cm] and a coolant velocity of 1.916 the time in the core is60.96 [cm]t = cm=31.82 [see]1.916 Page 10 of 29 Then Asat isAsat8.47 107 #3 6. x105[c~l2.6 x101 cm~ec ( -e(0.o66xo-4 [se-'],31.82 [sec])(1 -e-(1os66x1o-4 [sec-']*(31.82 [sec] +1.547x104 [sec]))Asat 57 decays 1Asat 571Lsec

cm3lThe number of 41Ar atoms is57[ dec~ays 1][tosAsat] --57 [sec

cm3]_, .1 0 [atmsNArl--41 541 06c11.56x1o-i [ ]-cThe exchange of a gas in water with atmosphere can be modeled asS =0.93 BNAr_4lAsurfwhere B is an exchange coefficient reported as 5.7 x i0.3 [-T] and A surf is the area of thesurface of the pool as calculated from a radius of 121.92 [cm]. (Dorsey, 1940)S 0.3 5. x 1-3 [m] atoms]S=.9

5.--o~F$

5.411 x 106 3

rr(121.92)2 [cm2]S =1.34 x 100 [atoms]1Lsec JMultiplying the source value obtained above with the decay constant would yield the activityemitted from the pool surface per second.s =1.34 x i09 [atos],1.0566 x104-I =.41 x 10 *- -[ sec I sc sc 3.7x 104/LBq/I[secJThe time radioactive air remains in the reactor room is a factor of the pumping rate of ventilationsystems and the decay of the argon. The half-lives can be combined into an effective half-lifeand would be1 1 1Teff TAr-41 TairTAr_41

TairSTe~ff TAr_41 +- TairPage 11 of 29 If the reactor room has a volume of 4.24 x 108 [cm3] and the fan removes air at a rate of2 [x i0 the air has a lifetime ofVtotai 4.24 x l08 [cm3]Tai = '~eova -x o~ [cm] = 2120 [sec] = 35.3 [main]Using this as the half-life of air in the room, which accounts for mixing and parts of the airstaying over time, the effective half-life of 41Ar in the room is109.34 [min]

35.3 [mai]Teff =093[mn+353[n]= 26.7 [mai] = 1.6 02 x 103 [sec]and the decay constant. isIn(2) _ .2nx1. i2Leff -= .36X of -Considering the room now as the entire source term, the activity at saturation isAroom(t) -7poo13.81 [l'c]4.326 x 10.4 I.

4.24 x 108 [cm3]Aroom,sat [ .0 i0 iReferencing FGR-1 1, the dose conversion factor for someone submerged in argon isDCFAr_4l = 8.029 X 10 yielding a dose rate to a worker ofD) = 2.08 x 10-s

8.029 x 0s= 16.7 []rThis is an incredibly conservative estimate for the steady state derived air concentration of the41Ar effluent. It assumes the pool water has reached a saturation of 41At, which has thendiffused completely into the reactor facility, as it is continually evacuated. To simply make thesecond pass of water through the reactor core would take more than four hours. An extremelylong run time would be 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months at which time the pool would start to approach the saturationpoint, a time far less than that of the saturation time for the entire reactor facility.Page 12 of 29 The reactor facility operates at a negative pressure and air is expelled to the outside of thebuilding through an exhaust fan. This is 15 meters above the ground. Following a similaranalysis to that in RAI 12b regarding the MHA, the dispersion factor is 6.18 x 10-3[s/Trn3]. Withan exhaust rate of 0.2 [m3/s]C, = A (x/Q)V =2.08 x 10-s

6.18 x i0-3 * .2 = 2.571 x 10-8[cjMultiplying this by the dose conversion factor for someone perfectly under the plume yields15 = 8.029 x 10s

2.571 x 10-8 =0.021 [rhiThis dose rate is less than that cited in 10 CFR 2O.1301(a)(2) of 2 [mRem/hr].8. In letter dated April 10, 2013, Purdue University provided an updated Section 13 of theSAR that included responses to RAIs 74-76, 80, 83, 85-91, and 93-95. State whether theupdated SAR Section 13 also intended to provide answer to RAIs 79, 81, 82, 84, and 99or provide answers:(a) RAI 79 in NRC letter dated July 14, 2011, stated:NUREG-1537, Section 13 provides guidance to identify the limitingevent for each accident group and to perform quantitative analysis forthat event. Please identify the categories of PUR-1 experiments that areperformed and provide an evaluation of a safety analysis using theguidance of NUREG-1537, Section 13.1.6 for potential experimentmalfunctions and their consequences.ResponseThe primary usage of the PUR-1 is training of nuclear students and educating the public aboutnuclear issues. Experiments performed include out of core dose measurements, fuel irradiationsin the irradiation facilities, and other non-fueled irradiations. Each of the risks from theseexperiments has been evaluated in their respective SAR section or in responses to RAIs.(b) RAl 81 in NRC letter dated July 14, 2011, stated:NUREG-1537, Part 1, Section 13.1.1 provides guidance to identifyMaximum Hypothetical Accidents (MHA) for non-power reactors. ThePage 13 of 29 MHA is to be selected so potential consequences of the postulated MHAscenario exceed and bound all credible accidents. NUREG-1537, Part 2,Chapter 13, p. 13-5 suggests that for a low-powered MTR fueled reactor,the MHA may be one of the following two events: cladding is strippedfrom one face of a fuel plate while suspended in air, or a fueledexperiment fails in air. SAR, Section 13.1 states that "the failure of afueled experiment is designated as the maximum hypothetical accidentof the PUR-I." Please provide ... a safety analysis of an MHA thatconsiders the failure of one fuel plate in air would have lowerconsequences than the failure of a fueled experiment by justifying theMHA accident involving the fueled experiment capsule is morebounding that the failure of a fuel plate.ResponseThe maximum hypothetical accident has been clarified to be that of having the cladding strippedfrom one face of a fuel plate while suspended in air in accordance with the suggestion ofNUREG-1 537, Part 1, Section 13.1.1.(c) RAI 82 in NRC letter dated July 14, 2011, stated:NUREG-1537, Part 1, Section 13.1.1 provides guidance in identifying anacceptable MHA for non-power reactors. The PUR-1 MHA accidentanalysis for "Failure of a Fueled Experiment" is stated to be based upona 1 W power deposition in the fueled experiment as consequence of thereactor operating at 1 kW. Please provide ... a safety analysis thatprovides the details of the energy deposition determination in the fueledsample with the reactor operating at the maximum requested licensedreactor power including the power level measurement uncertainty of50% stated in SAR, Section 13.1.2.ResponseThe PUR-1 reactor is equipped with irradiation facilities that can be loaded with a fueledexperiment in order to test changes in material properties. In this section an analysis isperformed to assess the hazard associated with the failure of an experiment in which fissilematerial has been irradiated in the reactor. In the scenario of this accident it is assumed that aPage 14 of 29 capsule containing irradiated fissile material breaks and a portion of the fission productinventory becomes airborne. The consequences of the release are analyzed for both the reactorstaff and the general public.Excess reactivity of the LEU core in PUR-1 was determined to be 0.00468 (0.46%) Ak/k in a190 fuel plate core (16 dummies), including a reactivity bias of 0.32% Ak/k. The TechnicalSpecification limit of 0.6%Ak/k is used to determine the maximum fueled experiment that canbe utilized within the experimental facility. The main concern in the failure of a fueled experimentis the release of fission products. A limit of 0.5 [Ci] of radio-iodine is specified in the PUR-1 TS.This is half of the radio-iodine that is postulated to be released in the MHA analyzed. This TSlimit ensures that restrictions from 10 CER 20 are met and the complete cladding failure of oneface of a fuel plate remains the maximum hypothetical accident. Extended safety measuresmust be implemented to further ensure doses from a potential experimental failure remain farbelow those from the MHA.a. A heating analysis must be done and approved by the reactor operatingcommittee to verify that the maximum temperature experienced by thefueled experiment is less than or equal to half of the melting temperature.b. Experimental analysis must include assumptions that include a loss ofcooling within the experimental facilityThe heat production in a sample experiment was found using a combination of the F7 tally fromMCNP6. The result of an F7 tally over a cell in is units ofF7 Tlly# [ram

Source Particle]By taking the ratio of the F7 tally for the fueled experiment with that of the entire fuel inventory,the proportion of reactor heat generation in the experiment is found. This can be scaled by thelicensed reactor power to give the power generated in the fueled experiment.P ~ rector(F7)Exp

MassExpPExperiment (FP7ac )roa*19plts*2.grm(F7)roaZ *10platplateAs an example of this calculation, for the requested 12 kW license upgrade plus 50% poweruncertainty, suppose a sample of 3% enriched uranium is placed in the experimental facility witha mass of 1.19 grams.[MeV ]Pexperiment 18000 Watts

8.3x O[g*sJ*19[ra]3.74858 x 10-3 [Me-V-]

190 plates

12.5 g~g*S~jplatePExp= 1.95 WattsWhere the values for the F7 tallies came from the MCNP6 model of the PUR-1 core.Page 15 of 29 This mass would yield a radioiodine production of 0.46 [Ci].Following an identical analysis as the MHA of a fuel plate breach but with the conservativeassumption that all of the fission products are released, the dose is shown in the tables below.The first table shows the dose from intake of radioiodine by an occupational worker if he were toremain in the plume for the specified amount of time. Realistic evacuation times are on the orderof one minute which would lead to an exposure of approximately ~-0.1 morero.Total [mRem] From IodineTime [hours] Inhalation0.01 0.0550.1 0.5451 4.69710 24.704100 68.4571000 108.40310000 109.990The dose rate from submersion in the radionuclides is shown below.Submersion Dose for Noble Gasses (Kr, Xe)Time [hours] Total [mRem/hr]0.01 0.08500.1 0.08161 0.064910 0.0278100 0.00341000 2.44 x10-s10000 2.48 x 10-16Again using the conservative assumptions for dose to a member of the public in a non-restrictedarea, the dose after submersion in the exhaust plume for an hour with the fan errantly stillrunning is 9.03 [toRero]. This dose is below the limits set forth in 10 CFR 20. These dose ratesfall below those of the MHA accident of having one face of a fuel plate completely exposed,making the failure of a fuel experiment not qualify for the MHA.For other possible fueled experiments, the TS 3.5 specify that for singly encapsulatedexperimentsf. The radioactive material content, including fission products, of any singlyencapsulated experiment should be limited so that the complete release of allgaseous, particulate, or volatile components from the encapsulation will not result indoses in excess of 10% of the equivalent annual doses stated in 10 CFR 20. ThisPage 16 of 29 dose limit applies to persons occupying (1) unrestricted areas continuously for twohours starting at time of release or (2) restricted areas during the length of timerequired to evacuate the restricted area.and for doubly encapsulated experimentsg. The radioactive material content, including fission products, of any doublyencapsulated experiment or vented experiment should be limited so that thecomplete release of all gaseous, particulate, or volatile components from theencapsulation or confining boundary of the experiment could not result in (1) a doseto any person occupying an unrestricted area continuously for a period of two hoursstarting at the time of release in excess of 0.5 Rem to the whole body or 1.5 Rem tothe thyroid or (2) a dose to any person occupying a restricted area during the lengthof time required to evacuate the restricted area in excess of 5 Remn to the whole bodyor 30 Rem to the thyroid.(d) RAI 84 in NRC letter dated July 14, 2011, stated:NUREG-1537 states that the format and content of the TS followANSI/ANS 15.1. ANSIIANS-15.1-2007, Section 3.8.2 provides guidancefor double encapsulation of experiments involving fissionable,explosive, reactive, or corrosive materials. Please provide .,, a safetyanalysis for the MHA experiment of 1.1 g of U-235 with singleencapsulation is consistent with the guidance provided in ANSI/ANS-15.1-2007, Section 3.8.2.ResponsePlease reference RAI 14c) response above for a safety analysis for the failure of a fueledexperiment of 1.19 g of U-235.(e) RAI 99 in NRC letter dated July 14, 2011, stated:SAR, Section 13.2.1, page 13-9 states "This experiment corresponds tothe irradiation of 1.1 gm of U-235 in the mid-plane of the isotopeirradiation tube located in position F6." Please provide ... a safetyanalysis that establishes the basis of 1.1 gm of U-235 for failure of afueled experiment.ResponsePage 17 of 29 Piease reference RAI 14c) response above for a safety analysis for the failure of a fueledexperiment of 1.19 g of U-235.9. In a letter dated April 10, 2013, Purdue University provided an update to the SAR,Section 13 containing an analysis for the designated MHA based on the failure of afuel plate and not the malfunction of a fueled experiment. Provide an analysisdetermining whether the consequences of the failure of a potential experimentcontaining fissile material are bounded by the MHA as defined by PUR-I. In addition,discuss the basis for limiting the maximum allowable fissile content of fueledexperiments and how the limit is to be controlled either by a TS or by otheracceptable means.ResponseComments following the safety analysis of potential accidents from experiments clarify that thestripping of cladding from one face of a fuel plate is the bounding MHA.10. NUREG-1537, Part 2, Section 13, suggests that the definition of the maximumhypothetical accident (MHA) should be based on either a fuel plate or a fueledexperiment failure, whichever leads to higher consequences. In the updated PUR-1SAR, Section 13.1.1 and 13.1.6, it is stated that the failure of a fueled experimentcontaining fissile material is the designated MHA. However, in the updated PUR-1SAR, Section 13.2.1, an analysis is proved for the designated MHA that is the failureof a fuel plate. Explain how this is consistent, or correct the updated PUR-1 SAR,Section 13, with regard to the MHA.ResponseThe PUR-l SAR, Sections 13.1.1 and 13.1.6 have been updated to be consistent with PUR-1SAR Section 13.2.1.11. NUREG-1537, Part 2, Section 13, defines the MHA as the failure of the cladding of oneface of one fuel plate while suspended in air. PUR-1 SAR Section 13.2.1 defines theMHA as the failure of one face of one fuel plate submerged in the reactor poolresulting in a potentially non-conservative amount of radioactive iodine release intothe reactor air volume. Discuss whether the assumption of radioactive fission productrelease into the pool water is a conservative assumption. Discuss whether theassumption bounds a failure of the fuel plate cladding while the fuel element issuspended in air, releasing fission products directly into the reactor air volume.Page 18 of 29 ResponsePun-1 SAR Section 13.2.1 has been clarified to indicate analysis is done for the MHA of thefailure of the cladding of one face of one fuel plate while suspended in air following the guidelineof NUREG-1537, Part 2, Section 13. This complete analysis is done in the subsequent RAl 12response.12. Provide an MHA safety analysis that explains all analyses, assumptions andconclusions at the requested licensed power level for the maximum potentialestimate of the total radioactive fission product release after the failure of one side ofone fuel plate. Discuss methodological assumptions associated with the followinganalytical steps:(a) Derivation of fission product atmospheric dispersion factor, x/Q using eitherthe methodology suggested in Regulatory Guide 1.145, "AtmosphericDispersion Models for Potential Accident Consequence Assessments AtNuclear Power Plants," Revision 1, issued February 1983, or anotherequivalent method.(b) Dose conversion calculation using the Environmental Protection Agency'sFederal Guidance Report (FGR)-11 and FGR-12 dose conversion coefficients oranother equivalent methodology to account for inhalation/ingestion andsubmersion exposures.Resp~onseThis section will estimate the total amount of radioiodine released from the failed fuel plate. Nocredit is taken for the reduction in activity resulting from radioactive decay during the time of therelease, i.e. an instantaneous release of the radioiodine that can escape the fuel is assumed.Complete and perfect mixing of the available radioiodine inventory with the reactor facilityvolume is also assumed.The activity via the production rate of the ith radioiodine isotope in a plate is determined by thefollowing:At = ,tN =ftssFtwhere P is the rate of fissions in the plate of interest, Fi is the fraction fission yield for eachradioiodine, A1 is the decay constant for each radioiodine and Ni is the saturation number ofatoms of each specific radioiodine. The constants Fi, hi, and the results of calculations for AiN1and N1for plate 1348, are shown in Table 13-2 assuming 12 kW + 50% operating power.Page 19 of 29 It is assumed that not all of the iodine produced will be released from the plate. As suggested byNUREG/CR-2079, only the fission fragment gases within recoil range of the surface of the fuel(1.37 x i0-3 [cm] for aluminum matrix fuels) will escape. The thickness of the fuel meat in aPUR-1 plate is 0.0508 cm. The total volume of the fuel meat isVtt:60.01

5.96

0.0508 =18.170 [cm3]and the volume released from recoil of the fission gasses isYreiease = 60.01

5.96

1.37 X 10-3 = 0.490 [cm3]The fractional fission product gas release is thereforeVreiease _0.490Freiease --18.170 0.027Vtot 1.7The number of fissions per second in Plate 1348 with a fission power of 157 [Watts] is given by1 6.245-x1012MV1fsio____157 W =157 W* __*s *.45x MV iso 4.90 x 1012 fissions1 W 1 Joule 200 MeV secThe activity of I-135 is then the product of these number of fissions and the fractional yield perfission of this element which is 6.28%.Activity1_135 4.90 x 1012 [id]i*s0o0628[decas] 1 1 [decaY/ecn]*0.062 3.7 x 1010

0.02 7 {Fractional Release}Activitys-z35 = 0.224 [Ci]The calculated values for each of the five iodine isotopes of concern are shown in below.Radioiodines released from Plate 1348 into Facility Air.Fractional DecayIsotope Yield Half-life (sec) Constant Activity [Ci]1-131 0.0289 6.95 x l~ 9.98 x i0-7 0.1031-132 0.0431 8.21 X i03 8.44 x 10-s 0.1541-133 0.067 7.49 x i05 9.26 x 10-6 0.2391-134 0.0783 3.16 x i03 2.20 x 10-4 0.2801-135 0.0628 2.37 x i04 2.93 x i0-s 0.224Total 1 CiPage 20 of 29 The activity produced from the fuel plate breach is then dispersed throughout the immediatereactor facility air. The facility has a volume of 424 m3 giving the activity per volume of air isgiven asActivityA/-VolumeS= 0.224 [Ci] 106 [jCi] = 23 i~ [ICi 1(AVI15424 x106 [cm3]

1 [Ci] -5[31- [m]Over time, this level of activity decreases according to the laws of radioactive decay. Theactivity for each radioiodine is shown in the figure below.Activity of Radioiodines After Release fr-omn Fuel Plate 1348QQI-0.010.001-0.01 0.1 1 10 100 1000--13 1--132-1 Tt,-1 adoidiTFime IhioursiAs a facility worker breathes in the radioiodine, they will ingest some of the air borne material.rcm3Assuming that an individual breathes in 347 [s-,l the dose from continued breathing of theradioactive air is given byDi= f

C(t)i dt = {B * (A/V)j}

fte-~a't dt = B(A/V)1 * (1 -e-it)where 1iL is the dose rate from inhalation of each radioiodine, B~ is the breathing rate, (A/V)1 isthe activity as a function of time of each radioiodine, 2i is the respective decay constant, and t isthe time duration of inhalation. The final effective dose is this value multiplied by a doseconversion factor to account for the different energies of decay and other factors.Page 21 of 29 135Deff= Di1=131For this analysis, the dose is expressed in units of dose (rads) to the thyroid gland of a personbreathing the radioiodine-bearing air. Although the concentration of iodine in the building air iscontinuously reduced by various processes such as radioactive decay, purging of the buildingair by the exhaust fan, and plating out of the iodine on surfaces, we will assume only a reductionin concentration resulting from radioactive decay. This implies that the concentrations of iodinein the air outside the building, taking no credit for dissipation in the air outside the building, willbe the same as those in the building. Estimates of the thyroid dose rates are shown below.Integrated thyroid dose estimates for several exposure periods following release of Plate 1348radioiodines into the reactor air.Integral Dose in mRem for Several Exposure Periods after ReleaseIsotope1-1311-1321-1331-1341-13536 sec 6 min 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 10 hours 4.2 days 42 days1.42 x 10-2 0.142 1.42 14.0 119. 3856.47 x 10-2 0.638 5.59 20.3 21.3 21.38.26 x 10-2 0.824 8.12 70.2 239 2480.14-3 1.38 9.89 18.1 18.1 18.17.59 x 10-2 0.756 7.21 46.9 72.0 72.00.380 3.74 32.22 170 470 744 [mRem]As can be seen from these results, about two hours of continuous exposure to releasedradioiodine in the reactor room air would be required to attain a thyroid dose equivalent of~-100 oremo. Even this exposure would be extremely unlikely, since it is difficult to conceive of acredible combination of accident conditions and personnel occupancy which will result in suchdoses being achieved. The imposition of such limited cloud dispersion effects required toapproach these estimates is not realistic. It is more likely that dispersive effects will result inmuch lower doses. For example, even the building blower exhaust at 424 ft3/1mn flow rate willcause a concentration reduction.Page 22 of 29 MHA considerinq All Gaseous Fission ProductsA similar analysis can be done for all gaseous radionuclides released from a single ruptured fuelplate. The same analysis for iodine as was done in the preceding sections applies, as well as allnoble gases (fission products) available in the plate released to the building air. No credit fordecay of the radioisotopes during release and dispersion is taken. This assumption, as before,leads to conservative results in that the estimates obtained are higher than those that wouldactually occur in this postulated accident.IsotopeKr-85mKr-87Kr-88Xe-131mXe-i133mXe-I133Xe-1 35Xe-i135mTotalDose Rate in mRem/hr for Several Exposure Periods After Release36 sec 6 min i hour 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 4.2 days 42 days3.78 x 10-6 3.73 x 10-6 3.25 x 10-6 8.07 x 10.7 0 03.56 x 10-2 3.39 x 10-2 2.08 x 10-2 1.53 x i0-4 0 00.126 0.123 9.87 x 10-2 1.10 x 10-2 0 05.91 x 10-s 5.91 x 10-s 5.89 x l0-s 5.77 x 10-s 4.64 x 10-s 5.22 x 10-61 x 10-3 1 x 10-3 9.89 x 10-4 8.79 x 10-4 2.68 x 10-4 04.01 x 10-2 4.01 x 10-2 3.99 x 10-2 3.79 x 10-2 2.31 x 10-2 1.62 x 10-40.301 0.299 0.280 0.141 1.48 x 10-4 07.95 x 10-2 6.22 x 10-2 5.39 x 10-3 0 0 00.583 0.560 0.445 0.191 0.0236 1.68 X 10-4In the event of a maximum hypothetical accident such as the one discussed, reactor facilityworkers would be required to identify that an alarm has occurred, notify others in the immediatevicinity to evacuate, and disable the ventilation system to attempt to confine the fission products.This process would take approximately one minute in a very conservative setting. The dosereceived during this time due to the noble gasses listed above would be 0.01 torero. The dosecontribution from the radioiodines would be much greater at -0.75 mRem but still well below thedose limits for a facility worker.Doses to persons outside the building will come from submersion in a cloud of releasedradionuclides and from radiation emitted from the reactor building. The submersion dose resultsfrom the diluted radionuclide stream from the exhaust fan or from natural flow of air through thebuilding that exits at the roofline (if the exhaust fan was properly disabled). An analysis for theactivity concentration release from the building can be performed using the equation below,Page 23 of 29 which includes release from the exhaust fan. The concentration of activity of each radionuclideis given aswhere Ai is the fractional release of activity of the radionuclide from the fuel plate as dispersedthrough the reactor facility air [Ci/m3], (x/Q) is the atmospheric dispersion factor as discussedin NUREG/CR-2260, 1V is the volumetric release rate from the fan and t is the time after release.The dispersion factor is given by1(xlQ) = +where u1o is the velocity of the wind at a height of 10 meters above the release point, o- is thelateral plume spread based on the Pasquill-Gifford correlation, and uz is the vertical plumespread (both in meters). Both of the plume spreads are dependent on the atmospheric stabilitywhich is conservatively assumed to be highly stable (as given by Regulatory Guide 1.23).Assuming that the plume spread in the lateral and vertical direction are the same, there is awind speed of 1 m/s, an exclusion zone of 10 meters (and extrapolating values fromNUREG/CR-2260), giving values for the plume spread of ay= az= 2.38 [meters] gives adispersion factor of(xi/Q) = 1=~ ~*(.8) 8 m] 6.18 x io- [3No credit is taken in the above expression for horizontal plume meandering or the fact that airfrom the basement area is exhausted at a minimum height of 50 feet. Additionally, the averagewind speed is greater than 3 mn/s resulting in a conservative factor of at least 3.For 1-135, with an activity of 5.29 x 10-4 IIc] the downwind activity concentration would be= Ai(Q/Q)V. = Ai* 0.00618["]r~ 0.2 = 0.00124 *A1[]C, = 0.00124 * .5.29 x 10-4 [cm3 IC1 = 6.538 x i0O m3The table below shows the downwind concentration and activity for each radio nuclide.Page 24 of 29 DownwindAccident Activity ConcentrationIsotope [Ci] [uCi /m3]1-131 0.103 0.300I-132 0.154 0.4491-133 0.239 0.6981-134 0.280 0.8151-135 0.224 0.654Kr-85m 0.046 0.134Kr-87 0.091 0.267Kr-88 0.127 0.370Xe-131m 0.001 0.004Xe-i133m 0.007 0.020Xe-i133 0.239 0.698Xe-i135 0.234 0.681Xe-135m 0.039 0.115Total: 5.20 [,uCi/mr3]The submersion dose rate isfactor for submersion.this concentration of activity multiplied for the dose conversionDi,submersion = (DCF)i

Cwhere the dose conversion factor is in units of [rnci/cmrj and is given in FGR-1 1. The dose ratefor inhalation of the radioiodine is similar but includes the rate of air intake by the averageindividual which is approximately 347 cm3 /sec.Di,thyroid =(DCF)i, tfyroid

BR

CiThe total dose rate is therefore the sum of the inhalation and thyroid dose rate over all isotopes.D~total= Di,submersion-t+ Di,thyroidji i* [mRnem]Dtotai 0.69 + 18.91 =19.6 I rThis is the dose rate to a member of the public if they were standing directly in the plumedownwind of a maximum hypothetical accident when the fan in the reactor facility failed todepower as considered on a very calm day. In the worst case scenario, an individual wouldstation themselves directly outside of the reactor facility underneath the exhaust. If it took 15minutes to setup the exclusion zone, a person could receive at most 4.9 mRem which is stillbelow the dose requirements to the general public. Expanding the exclusion zone as well asPage 25 of 29 ensuring the facility ventilation is deactivated only serves to lessen the dose to the public. If it isassumed that the fan does turn off properly, and the leakage of air through the top of the facilityis reduced to 0.001 [m3/se c], the dose rate falls to 0.1 [mRem/hr].13. Provide an analysis and discuss the potential maximum radiological dose estimatedue to the MHA at the following suggested locations:(a) Facility worker -located inside the restricted area considering any evacuationprocedure and Potential residence time for staff exposed to fission productinhalation/ingestion and direct gamma ray radiation. The exhaust systemoperational status should be consistent with conservative assumptions.ResponseReference response to RAI 12 above for a discussion of the dose to a facility worker during apostulated maximum hypothetical accident.(b) Members of the public -located in adjacent, publicly accessible areas insidethe reactor building (i.e. classrooms, hallways, adjacent rooms) potentiallyexposed to fission product inhalation and/or gamma ray radiation, taking intoaccount any procedural process for evacuation, including the emergency plan.The exhaust system operational status should be consistent with conservativeassumptions.ResponseReference response to RAI 12 above for a discussion of the dose to members of the public inadjacent publically accessible areas inside the reactor building during a postulated maximumhypothetical accident.(c) Members of the public -located outside the reactor building (maximallyexposed location, nearest dormitories, offices, etc.) exposed to fission productinhalation/ingestion released from the reactor building and gamma rayradiation. The exhaust system operation status should be consistent withconservative assumptions.ResponseReference response to RAI 12 above for a discussion of the dose to member of the publicoutside the reactor building during a postulated maximum hypothetical accident.14. 10 CFR Part 20, "Standards for Protection against Radiation," provides the regulatoryframework and NUREG-1537, Part 1, Section 13.1.3 provides the guidance forlicensees to systematically analyze and discuss credible accidents in each accidentPage 26 of 29 category. Section 13.1.3 of the updated PUR-1 SAR, describes the loss of coolantaccident (LOCA) scenario. The updated PUR-1 SAR does not include an estimate forradiation levels in the reactor floor and the roof areas, due to the unshielded reactorcore, after a postulated large LOCA event. The SAR should provide the consequentmaximum dose rates at various locations on the reactor floor and outside on thereactor building roof. In accordance with 10 CFR Part 20, provide the accumulateddoses to reactor building occupants and the maximally exposed member of thepublic, considering evacuation procedure and potential residence time for staff. Inaddition, provide an estimate when facility staff may enter the reactor building to startrecovery operations.ResponseThe maximum possibie conceivable LOCA accident in the PUR-1 reactor would be that ofhaving a large hole or crack develop in the reactor pool liner and the water draining from there.The only other option would be that of pumping or evaporating water from the top of the poolwhich would not be a bounding case. Modelling the reactor pool as a suspended tank with ahole or crack in the bottom, the Bernoulli equation shows1.PgZsur; + Patm = Pot+ pgh + Patm= --zsr h)where p is the density of water, g is the gravitational constant, Patm is the atmospheric pressure,vour, is the velocity of the water going out of the hole, z is the height of the surface of the water,and h is the height of the hole. If the hole is defined to be at the bottom of the tank (boundingassumption)rot= The loss of mass through the hole is given bydmdt =-pAhtoeVand the total mass in the tank ismtot =pAtankZsurfyieldingPage 27 of 29 d(pAtankZsurf) -)~ ol~dt =-phlVudt AtankIf there was a crack at the bottom of the tank which formed and was 5 [cm] wide that ran thediameter of the pool, its area would beAhole =0.05 [in]

2.44[m] =0.122 [in2]This area of a hole is equivalent to an instantaneous hole in the bottom of the reactor pool witha diameter of --40 [cm]AholeI = -=t .) 0.12[in2]d(Zsurf~t) _ 0 (12

9.81)-0.116 dt4Zsurf (t)Integrating yieldst = 17.24 zs@urfIn order for the tank to drain half of its water, the surface of the tank would go down2.59 [meters]. The time to do this would bet = 27.7 secondsThe total time to completely drain the tank is 40 seconds. This time frame is adequate for anoperator to become aware of the situation, initiate a scram and begin an evacuation ifnecessary. In a case where the operator did not initiate the scram, the pool-top radiation monitorwill do so on a high-alarm at 50 [me]If for some reason a scram was not initiated, thereactor would become subcritical upon the loss of the water moderator. The potential sourceterm would therefore not be the reactor at full power (12 kW plus 50%), but rather the fissionproduct decays still present in the core.Following the same analysis as the shielding of the reactor at full power as in RAI 6 above,using air as the only attenuator to the top of the pool, and using a source of 1.69 x 1013 fissionsper second (assuming the source term is 3% of full reactor power after the scram), the dose rateat the original height of the pool surface is 295 [Ber] Assuming a facility worker errantly lookedPage 28 of 29 over the pool edge immediately after the scram with no shielding, he would receive a dose ofless than 5 Rem to his upper body in one minute. Upon such an event, the reactor room wouldbe immediately evacuated, reducing the worker dose further. To others not directly above thepool the ground and concrete tank would reduce the dose still further.The member of the public of potential concern in the event of a sudden LOCA is someonepresent in the classroom above the reactor room. The ceiling is 19 feet above the top of thereactor pool and the floor of the upper room is approximately 4 inches of ordinary concrete.Extending the dose calculation to include the additional air and the concrete gives a potentialdose rate to a member of the public of 6.6 [Rei!]. Again, in such an event the building would beimmediately evacuated such that the actual dose received by a member of the public would besignificantly reduced.Page 29 of 29 TECHNICAL SPECIFICATIONS,FOR THE PURDUE UNIVERSITY REACTOR,.,PUR-1DOCKET NUM!BER 50-1 82 "'FACILITY LICENSE NO. R-8gD 0 "N \}. ,;West Lafayette, IN 47907March 2015 TABLE OF CONTENTSPage1. DEFINITIONS.................................................................... 1-12. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING .............. 2-12.1 Safety Limit.................................................................. 2-12.2 Limiting Safety System Setting / ...........2-13. LIMITING CONDITIONS FOR OPERATION ....-3.1 Rectvit Lmis............< ....................... 3-13.2 Reactor Safety System 3-2......2; .............. "......3.3 Primary Coolant Conditions ...x:.. ....... .,...........-.4, ....3-43.4 Confinement ........ ................. ........................... 3-53.5 imiatins"n'Exerides. ..................................... 3-54. SURVEILLANCE ... 4-1::: ... 3;".........4.1 ReactivityyLimltS-, ';:' .... ...-.......... 4-14.2 Reactdr Saet Systen ......... ............................. 4-14. rmr olant System, \...... ............................... 4-344 Ccifn--n.'.t........................................................4.E.er.et ...... ................................................. 4-45. FEATURE....................................... 5-15.1 Sit Dscri ption..... .......................................... 5-15.2 Reactor Coolant System ................................................... 5-15.3 Reactor Core and Fuel..................................................... 5-25.4 Fuel Storage ................................................................ 5-26. ADMINISTRATIVE CONTROLS............................................... 6-16.1 Organization ................................................................ 6-1PUR-1 Technical Specifications AedetN.1Amendment No. 13 6.26.36.46.56.6Review and Audit ................................................... ........ 6-5Operating Procedures...................................................... 6-7Operating Records .......................................................... 6-8Required Actions ........................................................... 6-9Reporting Requirements.................................................. 6-10//J/\, 'A/K1KK'KNAmendment No. 11

1. DEFINITIONSThe following frequently used terms are to aid in the uniform interpretation of thesespecifications:1.1 Channel -A channel is the combination of sensor, line, amplifier, and outputdevices that are connected for the purpose of measuring the value of aparameter.1 .2 Channel Calibration -A channel calibration is an-a~djustment of the channelsuch that its output corresponds, within rainge and accuracy, toknown values of the parameter which the chann~l masures. Calibration shallencompass the entire channel, mncludingequi~pmjenit.actuation, alarm, or trip,and is deemed to include a channel test. 7> .<,:... 1.3 Channel Check -A channel check is a :?qualitative verification acceptableperformance by observation of-,chaninel behavior. This mayinclude comparison of the channetljwith other ~independe~nt'c~hannels ormethods of measuring the same variable. \,/.,1 .4 Channel Test -A introduction of a simulated signal into achannel to veerify that it is o~eabie-:-. 15 Confinement -COnfinement ~ nen~cjue-. of ~the overall facility that isto I mit ,o#\e~ffluents-.between the enclosure and itsexternal environment thro&Igh conitroll7ed or defined pathways.1.6 Containment is an eonIlosure of the facility designed to (1) beat aega!ve in-leakage, (2) control the release of.;:efflets 0(-to: th~enviroVfnment,"&and<(3') mitigate the consequences of certain-- analyZ-d-aqcdidents\orevents. ., 1 : / , .,1.7 "- Gore Configquration -The~core configuration includes the number, type, or"arr.'angement of fuel~elements, reflector elements, and regulating/control rodsoccypying the corp grid.1.8 Core iExp:erirmerit / !A core experiment is one placed in the core, in thegraphite r'efj!*ctot<'~or Within six inches (measured horizontally) of the reflector.This includes;;aniy experiment in the pool directly above or below the core.1 .9 Direct Supervision -In visual and audible contact.1.10 Excess reactivity -Excess reactivity is that amount of reactivity that wouldexist if all reactivity control devices were moved to the maximum reactivecondition from the point where the reactor is exactly critical (keff =1) atreference core conditions or at a specified set of conditions.PUR-1 Technical Specifications 11AedetN.11-1Amendment No. 13 1.11 Experiment -Any operation, hardware, or target (excluding devices such asdetectors, foils, etc.) that is designed to investigate non-routine reactorcharacteristics or that is intended for irradiation within the pooi, on or in abeam port or irradiation facility. Hardware rigidly secured to a core or shieldstructure so as to be a part of its design to carry our experiments is notnormally considered an experiment.1.12 Experimental Facility -Experimental facilities are:a. those regions specifically designated as locations :for experiments orb. systems designed to permit or passage of a beam ofradiation to another location. , : .1.13 Experiment With Movable Parts (Secured o:r Nonsecur÷ed)Y, An experimentwith movable parts is an experiment parts that are intended tobe moved while the reactor is *..\.1.14 Explosive Material -Explosive material, is' a nv' solid or-,liquidi which iscategorized as a Severe, Dangerous, or V~eryDangerous Explosion Hazard in"Dangerous Properties of Industrial Materials".: by N. I. Sax, Tenth Ed. (2000),or is given an Identification of, Ol~Reativity (Stability) index of 2, 3 or 4 by theNational Fire Protection A~sdciat{ibn f"n'its pu-biication 704, "IdentificationSystem for Fire Hazards of Materials." <.,--..-1.15 Fueled Experiment -:'A 'experiment is-any experiment planned forirradiation~of :ur'anium 2:3.3 -!,uranium 235, plutonium 239, or plutonium 241.1.16 License -. Th~e written.-aut horizatio~i[. by the responsible authority, for an.individual .ior -organization t'o- .carry out the duties and responsibilities,*:associated* K~th a~per'sonnel position., material, or facility requiring licensing.1.17:4: Lhicensed -Se'e iienrsee;¢,: ..1.18 Licensee :- An individual or organization holding a license.1.19 MeasLired ,Valu~e' -measured value is the value of a parameter as itappears atthe'o.output of a canl1.20 Movable Experiment -A movable experiment is one where it is intended thatall or part of the experiment may be moved in or near the core or into and outof the reactor while the reactor is operating.1.21 New Experiment -A new experiment is one whose nuclear characteristicshave not been experimentally determined.1.22 Non-secured Experiment -See Unsecured Experiment.PUR-1 Technical Specifications1-2Amendment No. 13 1.23 Operable -A system or component is operable when it is capable ofperforming its intended function in a normal manner.1.24 Operatingq -A system or component is operating when it is performing itsintended function.1.25 Pool Experiment -A pool experiment is one positioned within the pool morethan six inches (measured horizontally) from the graphite reflector.1 .26 Protective action -Protective action is the initiaticih\ of a signal or theoperation of equipment within the reactor safety, sy~stem in response to aparameter or condition of the reactor facility havi n'g rdached a specified limit.1.27 Reactivity worth of an experiment -The reactivity an experiment isthe value of the reactivity change, that-'results, from. the experiment beinginserted or removed from its intended pbsiti'on. -,",,.. ......; ./1.28 Reactor Facility -The reactor is .that portiniof the ground floor of theDuncan Annex of the Electrical Engidie_,'ipg ,B~uidiing occupied by the Schoolof Nuclear Engineering Usedfor activities'asgsociated with the reactor.1.29 Reactor Operatinq -Ther~eactor is o peratir whenever it is not secured orshut down. :'-1.30 Reactor wholsiice'nsed~to manipulate the controls ofthe reactor*., .. J -\ \ ,;".1.31 Reactor Safet~y System iThe rea~ctor\safety system is that combination ofmeasuring channels,'and ..associated circuitry which forms the automatic.protective, system' "of.the reactor, ')'r prvds.nomainwhc.eqieto be initiated, rvdsifrainwihrqie/ ,1.32, Reactor Securied' -, A reactorisecrdwn',\a..L Either there :i's insufficient moderator available in the reactor to attain... ; criticality o~there is insufficient fissile material present in the reactor to~attin. cn~ticalfy.,under optimum available conditions of moderation andre'flectienh b. Or the 'following conditions exist:1. Reactor shutdown2. Electrical power to the control rod circuits is switched offand the switch key is in proper custody.PUR-1 Technical specifications 1-3 Amendment No. 13PUR-1 Technical Specifications1-3Amendment No. 13

3. No work is in progress involving core fuel, core structure,installed control rods, or control rod drives unless they arephysically decoupled from the control rods;1 .33 Reactor Shutdown -That subcritical condition of the reactor where thenegative reactivity, with or without experiments in place, is equal to or greaterthan the shutdown margin.1.34 Readily Available on Call -Readily available on call shall mean the licensedsenior operator shall insure that he is within a reas~nable driving time (1/2hour) from the reactor building, and the operator on is currently informed,and can contact him by phone.

1.35 Removable Experiment -A removable -,.xp'erment, is any experiment,experimental facility, or component i,0f >an other than apermanently attached appurtenance to. the reactor <system, which canreasonably be anticipated to be -moved one or more times dcurilhg the life ofthe reactor. -. , ,.?.. -1.36 Responsible authority: A govyernmental Qt~her entity with th~e authority toissue licenses, charters, 'pb isnor certific:tes':,1.37 Secured Experiment -Any eXp~ernment, exp..enmer~tal facility, or component ofan experiment is deemed to be; securye'dor in-a secUred position, if it is helin a stationary,' 15Qsi~tion'!,relatlve t~o th~e' reactorji.:y mechanical means. Therestraining, forces m~u<st: be substantially greater than those to which theexperiment might be subjlected by, hydraulic, pneumatic, buoyant, or otherforces whi~chire norm~al to the opertiri~g environment of the experiment, or byforces which 'cani'arise6 as-a-r~esultoefcr edible malfunctions.1.38- OrIerator. An individual who is licensed to direct the activitiesr'eactor o~erators. \Si~ch'an individual is also a reactor operator.% .; , " .1.39 '$h'all,. should, arid 'may ",~The word "shall" is used to denote a requirement;the~word "should" is used to denote a recommendation; and the word "may" isused t't6denote perrnssion, neither a requirement nor a recommendation.\. NX. /:: .1.40 Shutdowni\M" :.-' The shutdown margin, relative to the cold xenon-freecondition with most reactive shim rod fully withdrawn, and the regulatingrod fully withdrawn.1.41 Surveillance and Test Intervals -These are intervals established for periodicsurveillance and test actions. Established intervals shall be maintained on theaverage. Maximum intervals are allowed to provide operational flexibility, notto reduce frequency.PUR-1 Technical Specifications 14AedetN.11-4Amendment No. 13 1.42 Reference core condition: The condition of the core when it is at ambienttemperature (cold) and the reactivity worth of xenon is negligible (<0.30dollar).1.43 Rod, control: A control rod is a device fabricated from neutron-absorbingmaterial or fuel, or both, that is used to establish neutron flux changes and tocompensate for routine reactivity losses. A control rod can be coupled to itsdrive unit allowing it to perform a safety function when the coupling isdisengaged.1.44 Rod, regqulatinq: The regulating rod is a low worth c6ntro1 rod used primarily tomaintain an intended power level that need not{ have scram capability andmay have a fueled follower. Its position may be-varie6d'mranually or by a servo-controller. ./, v .. ,,.,,./ :. / .,.:,1.45 Rod, Shim-Safety: The control rods,"use'd in PUR-,1 .as,,described in thedefinition for Rod, control. -.-' : > .- ..1.46 Tried Experiment -Atried experiment :is'-:, 7"i.' .a. An experiment preViously-performed in tthis facility, orb. An experiment of approximately the same dniclear characteristics as anexperiment pr..viously tried. /;,4.,<- .,,,1.47 True Value -The true Value of a param~eter is its-exact value at any instant.1 .48 UnscheduledK':Shutdown h An unscheduled shutdown is defined as anyunplanned shut'down,'of the reactor by actuation of the reactor safety system,operator-~error, eqtuipment malfunction, or a manual shutdown in response toconditions that could adversely safe operation, not including shutdowns,,that occur dur, rig testing, or checkout operations.1.49 "UnlsecUred Experiment -"An'y experiment, experimental facility, or componentof'an .expeniment'lis considered to be unsecured when it is not secured asdef~ined in part 1 .36 of this section.,/ .., .IPuR-1 Technical Specifications 1-5 Amendment No. 13PUR-1 Technical Specifications1-5Amendment No. 13

2. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING2.1 Safety LimitSafety limits for nuclear reactors are limits upon important process variables that arenecessary to reasonably protect the integrity of certain physical barriers that guardagainst the uncontrolled release of radioactivity. The principal physical barrier is thefuel cladding.Applicability -This specification applies to the temperatur#je,&tthe reactor fuel andcladding under any condition of operation. ..Obiective -The objective is to ensure fuel cladding integrity,..,, .Specification -The fuel and cladding temperatures shal not exceed,530oC (9860F).Basis -In the Purdue University Reactor; first and principal barmer protectingagainst release of radioactivity is the the fue! pl~ates. The.6064aluminumalloy cladding of the LEU fuel plates has an".nciPj[e~nt melting temperature of 5820C.However, measurements (NUREG-1 313) on irr-adiated fuel plates have: shown thatfission products are first the bhs~tev< emperature (-5500C) of thecladding. To ensure that the bliste.Ktemieymture is' ever reached, NUREG 1537concludes that 5300C is an acceptable, fu'e-ariU'claddil {eprtrelmtno obexceeded under any of <, ........,,2.2 Limiting Safety Systmni Settingl',.,i" .;Appcb__t_-___ssp cation/applies to the reactor power level safety systemsetting for :steady st'ate :operation:.:, ,7Obiective -assure that'the safety limit is not exceeded.Specification -The measured-valiue of the power level scram shall be no higher thanBasis -Thb.'eLSSS. has behn chosen to assure that the automatic reactor protectivesystem will be a manner as to prevent the safety limit from beingexceeded d mtSevere expected abnormal condition.The function of the LSSS is to prevent the temperature of the reactor fuel andcladding from reaching the safety limit under any condition of operation. Duringsteady-state operation, a power level of 94.2 kW is required to initiate the onset ofnucleate boiling. This is far higher than the maximum power of 18 kW, which allowsfor 50% instrument uncertainties in measuring power level.PUR-1 Technical Specifications 21AedetN.12-1Amendment No. 13 For the transients that were analyzed, the temperature of the fuel and cladding reachmaximum temperatures of 4900, assuming reactor trip at 18 kW after failure of thefirst trip. This temperature is far below the safety limit of 53000..¢< K ,N f,N ,o(3/4>3/4N k , "./ 7-,.3/4,/ / .fNNN \<A/ --/ .NtX /iVNY -PUR-1 Technical Specifications 22AedetN.12-2Amendment No. 13

3. LIMITING CONDITIONS FOR OPERATION3.1 Reactivity LimitsApplicability -These specifications apply to the reactivity conditions of the reactor,and the reactivity worths of control rods and experimentsObiective -The objective is to assure that the reactor can be shut down at all times,that the safety limits will not be exceeded, and that operation is within the limitsanalyzed in the SAR. Specification -The reactor shall not be operated unless th~e following conditionsexist:,': a. The shutdown margin, relative to the,,cold xenon-free,,condition with themost reactive shim rod fully withdraw'n, and the regulating rod fullywithdrawn shall be at least 0.01 Akde/k.,.,,b. The reactor shall be subcritical by-more thanr.3A/kdrnior odnchanges. x c. No shim-safety rod sha~ll be removed from th~e core if the shutdown marginis less than 0.01 Ak/k with,,the remaining shim-safety rod fully withdrawn.d. The reactor sh~lL~be shutd'own if-the ma~xihium positive reactivty of thecore and aniy: intalled experi'enht e~xceeds 0":006 Ak/k.e. The reactbivty worth of, each experiment shall be limited as follows:A' Reactivity Worth, .003 Alk/..'!Unse~cUred .003 Ak/k',:,:,( ,:ecu red :,,.004 Ak/kf' T,.he total worth tof all movable and unsecured experiments shall notexceed 0.003 Akg.The ttal{worth'bof all secured experiments shall not exceed 0.005 Ak/k.Bases -The shutdown margin required by Specification 3.1 .a assures that thereactor can be shut down from any operating condition and will remain shut downeven if the control rod of the highest reactivity worth should be in the fully withdrawnposition.Specifications 3.1 .b and 3.1 .c provide assurance that the core will remain subcriticalduring loading changes and shim-safety rod maintenance or inspection.PUR-1 Technical Specifications 3-1 Amendment No. 13PUR-1 Technical Specifications3-1Amendment No. 13 Specification 3.1 .d limits the allowable excess reactivity to the value assumed in theHSR. This limit assures that the consequences of reactivity transients will not beincreased relative to transients previously reviewed, and assures reactor periods ofsufficient length so that the reactor may be shutdown without exceeding the safetylimit.Specification 3.1.e limits the reactivity worth of secured experiments to values ofreactivity which, if introduced as a positive step change, are calculated not to causefuel melting. This specification also limits the reactivity worth of unsecured andmovable experiments to values of reactivity which, if introd~uc~d as a positive stepchange, would not cause the violation of a safety limit./-The manipulation of* * .z,~~/ -, /experiments worth up to 0.003 Ak/k will result in reactor' periods longer than 9seconds. These periods can be readily compensate#'dtor': bytt~e action of the safetysystem without exceeding any safety limits. N A limitation of 0.003 Ak/k for the total reactivity worth of all movab, e",and unsecuredexperiments provides assurance that a. common failure affecting all suchexperiments cannot result in an of greate&- consequences;; than themaximum credible accident analyzed in the Aiil-Specification 3.1.g along with 3.1-;a a~ssures that'the,:reactor is capable of being shutdown in the event of a positive 're~acivity insertion, caused by the flooding of an3.2 Reactor > il'ii!;/ / Applicability -applies to_, the reactor safety system and othersafety-related instrumnentation. , -' \i:Oblective J.Zhe objective,.is to" specify the' Iowest acceptable level of performance orthe 0of~ac6eptable corriponents for the reactor safety system andother, safety relatedqnstrurmertation.Speifiatin -The reactor shall ,nt be made critical unless the following conditionsare met:\ .--a. Thqe\ rieactor safety channels and safety-related instrumentation areoperabel, in agc'brdance with Tables I and II including the minimum numberof the indicated maximum or minimum set points.b. Both shim-safety rods and the regulating rod shall be operable.c. The time from the initiation of a scram condition in the scram circuit untilthe shim-safety rod reaches the rod lower limit switch shall not exceed onesecond.PUR-1 Technical Specifications 32AedetN.13-2Amendment No. 13 TABLE I. SAFETY CHANNELS REQUIRED FOR OPERATIONMinimumNumberChannel Required Setpoint Function2 cps 2 cps rod withdrawal interiockLog count rate and period 1(a) 12 sec. period Setback7 sec. period Slow Scram12 sec. period Setback/, ,Log N and period 1(b) 7 sec. period /Slow, Scram7 sec period / Fast Scram120% power" :Slow Scram110% Seti~ackx,Liea 1 1 Slow Scram%Safety 1 (b) 11:0%'/ power Setback -4..120/o.poweri, .Fast Scram " "'Manual Scram " \J/(console) \.i-,_ """,. Slow Scram(hallway) 1Y:\ ...... Scra(a) Not required after Log N-Period chann~el comrnes on scal&.,"i(b) Required to be operable-but not on scale at startup ....,"TABLE II. SAFE'TY--RELATED/CHANNELS' (AREA RADIATION MONITORS),,.'! , ' Number '---'Chann el K R? 'equired(c) Setpoint FunctionP'ool~top monitor 1 .", l. 50 mR/hr or2x full power Slow Scram" "Ci background1I 71/2 mIRlhr Slow ScramConsole Monitbr... J 1 7A1/2mR/hr Slow ScramContinuous air sampler/ 1 Stated on sampler Air sampling(c) For periods of one week or for the duration of a reactor run, a radiation monitormay be replaced by a gamma sensitive instrument which has its own alarm and isobservable by the reactor operator.Bases -The neutron flux level scrams provide redundant automatic protective actionto prevent exceeding the safety limit on reactor power, and the period scramconservatively limits the rate of rise of the reactor power to periods which aremanually controllable without reaching excessive power levels or fuel temperatures.PUR-1 Technical Specifications 33AedetN.13-3Amendment No. 13 The rod withdrawal interlock on the Log Count Rate Channel assures that theoperator has a measuring channel operating and indicating neutron flux levels duringthe approach to criticality.The manual scram button and the "reactor on" key switch provide two methods forthe reactor operator to manually shut down the reactor if an unsafe or abnormalcondition should occur and the automatic reactor protection does not function.The use of the area radiation monitors (Table II) will assure that areas of the PurdueUniversity Reactor (PUR-1) facility in which a potential high radiation area exists aremonitored. These fixed monitors initiate a scram whenever the 'preset alarm point isexceeded to avoid high radiation conditions. </ /,Specifications 3.2.b and 3.2.c assure that the saeys s~em response will beconsistent with the assumptions used in evaluatihg the ~reactor's capability towithstand the maximum credible In specification 3.2.c. the rod lower limit, switches are';positioned tbomeasure, asclose as possible, the fully inserted position.> \', < / ,3.3 Primary Coolant Applicability -This ...specification' ,ap'lies 1t the limiting conditions for reactoroperation for the primary coolant.v\ ",",Objective -The objectivd~ ~s to assure, ,< -,,>, a~compati ble, environment, adequate shielding,and a continuous coolanht path',for\ the rea=ictor core.Specification -"' ,,/<,< ',"a. Thi~im"y ooan pH s m~aintained at an average over one year>: :b- The resistivity shall be maintained at a value greater than"\.<330,000 ohm-cm. "c. The !pri!mary cool!ant shall be maintained at least 13 feet above the core.Bases ,at the PUR-1 and other facilities has shown that themaintenance of primary coolant system water quality in the ranges specified inspecification 3.3.a and 3.3.b will minimize the amount and severity of corrosion ofthe aluminum components of the primary coolant system and the fuel elementcladding.The height of water in specification 3.3.c is enough to furnish adequate shielding aswell as to guarantee a continuous coolant path.PUR-1 Technical Specifications 34AedetN.13-4Amendment No. 13 3.4 ConfinementApplicability -This specification applies to the integrity of the reactor room.Obiective -The objective is to limit and control the release of airborne radioactivematerial from the reactor room.Specification -a. During reactor operation the following conditions wiJllbe met:1. The reactor room will be maintained at a negatiwe pressure of at least0.05 inches of water with the operation of the room exhaust fan.2. All exterior doors in the reactor remain closed except asrequired for personnel, equipment, or materials b. All inlet and exhaust air ducts a'nd/the sewer~vent shall 'contain a HEPAfilter or its equivalent. ,:?: ::.: c. Dampers in the ventilaion system hlgt' outlet ducts are capable ofbeing closed. d. The air conditioner can beish~ut boffbyjithe Bases -The PUR- 1 do~s not.rely on a ~coitairimrnitb~filding to reduce the levels ofairborne radioactive Tmateri al'r, lqased teql the~ environment in the event of the designbasis accident. Hw~ever, in th~e event ofgsuch an accident, a significant fraction ofthe airborne materialwill be conifined within ithe reactor room, and the specificationsstated above will fuith~er,,eue-terlaeto ,the environment.3.5 Limitationison Expe'rimentsApl~licability- This SplpiCfcatiln, applies to experiments installed in the reactor and itsexeietlfacilities. \,,7Obiective -,The, objective isBto prevent damage to the reactor or excessive release ofradioactive m~aterials in':the event of an experiment failure, and to assure the safeoperation of the reactor.;Specification -The reactor will not be operated unless the following conditions aremet:a. All experiments shall be constructed of material which will be corrosionresistant for the duration of their residence in the pool.b. All experiments and experimental procedures must received approval bythe Committee on Reactor Operations.PUR-1 Technical Specifications 35AedetN.13-5Amendment No. 13

c. Known explosive materials shall not be placed in the reactor pool.d. Cooling shall be provided to prevent the surface temperature of anexperiment from exceeding 1 00°C*e. No experiment shall be placed in the reactor or pool that interferes withthe safe operation of the reactor.f. The radioactive material content, including fission products, of any singlyencapsulated experiment should be limited so that the complete release ofall gaseous, particulate, or volatile components from the encapsulation willnot result in doses in excess of 10% of the equivalent annual doses statedin 10 CFR 20. This dose limit applies to ii:p~ersons occupying (1)unrestricted areas continuously for two hoiirs~starting~at time of release or(2) restricted areas during the lengt~h' of time reqjuired to evacuate therestricted area. g. The radioactive material content, :in~luding fi~sson products&;,of-any doublyencapsulated experiment or vente~dexperiment should be lim~ited so thatthe complete release of all or volatile componentsfrom the encapsulation~6r-confining b~upd~ry of the experiment could notresult in (1) a dose t o afiy-.person dqc upying an unrestricted areacontinuously for a period of two.hours sta'rttingat the time of release inexcess of 0.5 Rem to body- br-- 1.5 Remn to the thyroid or (2) adose to a~restriCtedj during the length of timerequired :the res~trcte~d area in excess of 5 Rem to the wholebody ori,30O Rem to the thyroid.'i 'Bases -Specificatiqn intended to reduce the likelihood ofdamage to,-.reactor .' conmponents--ardd/or~rdioactivity releases resulting fromfailure .and seie as a for the review and approval of newexperimrerifts by th~e facility'personnel and the Committee on Reactor Operations.Specifipation 3.5.f and 3.25.g conf~qrm to the criteria set forth in Regulatory Guide 2.2issued ihnNovember, .. .. .: ' ,IN :% £ IPUR-1 Technical Specifications 36AedetN.13-6Amendment No. 13

4. SURVEILLANCE REQUIREMENTS4.1 Reactivity LimitsApplicability -This specification applies to the surveillance requirements for reactivitylimits.Obiective -The objective is to assure that the reactivity limits of Specification 3.1 arenot exceeded.S pecificatio n -,:/a. The shim-safety rod reactivity worths shall be rpeasured and the shutdowmargin calculated biennially with no inter~alto "exceed 21/2 years, whichmay be deferred with CORO approval'durnng any reactor shutdown, andwhenever a core configuration is d~for which shim-safety rod worthshave not been measured. In of a deferred mieasurement, themeasurement must be performed pii'or to res/uming reactor operations.b. The shim-safety rods shal be visua lly inspected biennially with no intervalto exceed 21/2 years, wh'ich,,may be d~eferred with CORO approval duringany reactor shutdown.\\if.th~e roQd is founhd to be deteriorated, it shall bereplaced with a rod of equiv'ale'ntor greater worth, meetingthe limiting conditions of op'peratibo specified-*,ri>-3.1. In the case of adeferred me~asdirem~ent, the m~ust be performed prior toresuming reactor operations. ' /: Sc. The reactivity worth ~of experiments placed in the PUR-1 shall bemeasured during flrst- startup ~subsequent to the experiment's insertionand shall be ~ve'fied ff"-core.b7onfiguration changes cause increases in/::-experimnert reactivity worth which may cause the experiment worth to, ..6:.-xceed the.-aluess~peciffed in Specification 3.1Nf' \.," \. \ ,BaseS. -':Specification will a~ssure that shim-safety rod reactivty worths are not',, Ndegraded~ or changed by Core manipulations which cause these rods to operate inregions where~their effectiveness is reduced.The boron shim-safety rods have been in use at the PUR-1 since1962, and over thisiperiod of time, no cracks or other evidence of deterioration havebeen observed. Based on this performance and the experience of other facilitiesusing similar shim-safety rods, the specified inspection times are consideredadequate to assure that the control rods will not fail.4.2 Reactor Safety SystemApplicability -This specification applies to the surveillance of the reactor safetysystem.PUR-1 Technical Specifications 4-1 Amendment No. 13PUR-1 Technical Specifications4-1Amendment No. 13 Objective -The objective is to assure that the reactor safety system is operable asrequired by Specification 3.2Specification -a. A channel test of each of the reactor safety system channels listed inTable Ill shall be performed prior to each reactor startup following ashutdown in excess of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or if they have been repaired or de-energized.TABLE Il1* *,/,':>:SAFETY SYSTEM CHANNELS TESTED AFTER PROLONGED ,:, SHUTDOWNLog Count Rate (startup channel) Log N-Period ./* , :.. .Linear Level .A,./ ,,Safety Channel b. A channel check of each of the reactor safety system measuring channelsin use or on scale* shalt be approximately every four hourswhen the reactor is in operation. ;<:. , , ..c. A channel calhbration of th shall be performed atthe followng'-average interval ;:7";1. An leflecronic cahibration will b4e&performed annually, with no interval toexcee'd, 1 5,moht~hs.~.TCFhis~may be deferred with CORO approval duringf <7perio~ds of,*reactor shutdown,1b~t must be performed prior to startup.s;,:2* A powerbcalIjbation by foil activation will be performed annually, with,.; ,, no intervalto excee~d 15 months. This may be deferred with CORO%:?,approval d~uring periods of reactor shutdown, but must be performed,,,,,,prior to startup.d. The operatjo'h of the radiation monitoring equipment shall be verified dailyduring',peri0d when the reactor is in operation. Calibration of thesemonitors sh1all be performed annually, with no interval to exceed 15months. This may be deferred with CORO approval during periods ofreactor shutdown, but must be performed prior to startup.e. Shim-safety rod drop times will be measured annually, with no interval toexceed 15 months. These drop times shall also be measured prior tooperation following maintenance which could affect the drop time or causemovement of the shim-safety rod control assembly. This may be deferredwith CORO approval during periods of reactor shutdown, but must beperformed prior to startup.PUR-1 Technical Specifications4-2Amendment No. 13 Bases -A test of the safety system channels prior to each startup will assure theiroperability, and annual calibration will detect any long-term drift that is not detectedby normal intercomparison of channels. The channel check of the neutron flux levelchannel will assure that changes in core-to-detector geometry or operatingconditions will not cause undetected changes in the response of the measuringchannels.Area monitors will give a clear indication when they are not operating correctly. Inaddition, the operator routinely records the readings of these monitors and will beaware of any reading which indicates loss of function. /-.The area monitoring system employed at the PUR2;I ,has exhibited very goodstability over its lifetime, and annual calibration is cdnside'red adequate to correctlong-term drift. /,:" ",, The measured drop times of the shim-safety :ro~ds have been consistent since thePUR-1 was built. An annual check of this, parameter is consider~ed xadequate todetect operation with materially changed drtop times* Binding or rubbing>L.caused byrod misalignment could result from mairteaance;,.therefore, drop tfimes will bechecked after such maintenance. /,;b,:.;,;4.3 Primary Coolant System Applicability -This specification appliiesto tln~e average surveillance schedules of theprimary coolant system.; <,; ;.. /," Oblective -The objective is to assure high, quality pooi water, adequate shielding,and to detect the~rele[ase of fis~ioh products from fuel elements.Specification -"- -... --""-a:. ,The 3H. of the primary coolant shall be recorded monthly, not to exceed six'fi j weeks. This. during reactor shutdown.b.t The conductivity ,of the coolant shall be recorded monthly, not to*"£ six weeks. This cannot be deferred during reactor shutdown.c. The r:eactop~po~ol water will be at a height of 13 feet over the top of thecore when ever the reactor is operated. The reactor pool water heightshall be viSuially inspected weekly, not to exceed ten days, and water willbe added as necessary to reach the specification.d. The primary coolant shall be sampled monthly, not to exceed six weeks,and analyzed for gross alpha and beta activity. This cannot be deferredduring reactor shutdown.Bases -Monthly surveillance of pool water quality provides assurance that pH andconductivity changes will be detected before significant corrosive damage couldoccur.PUR-1 Technical Specifications4-3Amendment No. 13 When the reactor pool water is at a height of 13 feet above the core, adequateshielding during operations is assured. Experience has shown that approximately35-40 gallons of water will evaporate weekly and weekly water make-up is sufficientto maintain the reactor pool water height.Analysis of the reactor water for gross alpha and beta activity assures againstundetected leaking fuel assemblies.4.4 ConfinementApplicability -This specification applies to the surveillance requirements formaintaining the integrity of the reactor room and fuel clad/.7Obiective -The objective is to assure that the room and thefuel clad is maintained, by specifying average sur'ei~llnce intervals.f -.J,Specfication ...-a. The negative pressure of the rea~cto&- room wifllbe recorded week~ly.b. Operation of the inlet and outlet damprfis~shall be checked semiannually,with no interval to exce'ed-7tl/2 months*".' -,c. Operation of the air condltioner shall-.b~e chegke~d semiannually, with nointerval to exc~eed,7 1/2 months ..,:;L>-,,d. Representati~ve fuel, assembhles? shall be inspected annually, with nointerval to exceed ",>.+ ,Bases -Specification at b,,,and:qch chkthe in[tegrity of the reactor room, and d theintegrity of th~e-fu~el clad.i BaSed upon--pest' experience these intervals have beenshown, to ~be 'fo,.' ensuring the" operation of the systems affecting theintegnityiof the reactor'Tronmiand ~fuel clad.4.5 >,Exp~eriments \ .,Applica bilitVf~ -,This spe6cification applies to the surveillance of limitations onexperiments. Obiective -To compliance with the provision of the utilization license, theTechnical Specifications, and 10 CFR Parts 20 and 50.Specification -No experiments will be performed unless:a. It is a tried experiment.b. The experiment has been properly reviewed and approved according toSection 6 of the technical specifications.PUR-1 Technical Specifications4-4Amendment No. 13 Bases -The basis for this specification is to ensure the safety of the reactor andassociated components, personnel, and the public by verification of proper reviewand approval of experiments as specified in Section 6 of these technicalspecifications.// \"%N "\- /NN N//<N .// ~ N3/43/4-, .~ .. .NNN-N N//Ni N~-': )>7/PUR-1 Technical Specifications 45AedetN.14-5Amendment No. 13

5. DESIGN FEATURES5.1 Site Description5.1.1 The reactor is located on the ground floor of the Duncan Annex of theElectrical Engineering Building, Purdue University, West Lafayette,Indiana.5.1.2 The School of Nuclear Engineering controls approximately 5000 squarefeet of the Duncan Annex ground floor, which includes the reactor room.Access to the Nuclear Engineering controlled reea, is restricted exceptwhen classes are held there.*//" 5.1*3 The licensed areas include the reactor room,.-.and a fuel storage room.Both of these areas are restricted to ,.personnel, or thoseescorted by authorized personnel. /- -'5.1.4 The reactor room remains locked at all'times except for the'entry or exit ofauthorized personnel or those / by authorized\ ~personnel,equipment, or materials. 'K>:-.. /: ..5.1.5 The PUR-1 reactor room is a closed room designed to restrict leakage.5.1.6 The minimum free volu~m o.the 'oom is approximately 15,000cubic feet. -P ...\5.1.7 The ventilation system is de~signed.to e~haust"-air or other gases from thereactor an exa~~et b iiu of 50 feet above the5.1 .80Openir~gs~nto the reactor room consist of the following:a. Three~per~sonnel Sbj .DOedpqor to a~storage roomi with no outside access.c. Air retake \ ,.d.. Air exhause.KS'ewer vent ,)5.2 Reactor Coolant' System5.2.1 Primary Cooling System -The PUR-1 primary cooling system is a poolcontaining approximately 6,400 gallons of water.5.2.2 Process Water System -The process water system is assembled in oneunit and contains a pump, filter, demineralizer, valves, flow meters, and aheat exchanger (see 5.2.4). The demineralizer contains a removablecartridge that is monitored continuously for radioactivity buildup. Thissystem limits, by the use of flters and ion-exchange resin, the aluminumPUR-1 Technical Specifications5-1Amendment No. 13 corrosion rate, corrosion product buildup, and neutron activation ofimpurities in the coolant.5.2.3 Primary Coolant Makeup Water System -Makeup water for the pool istaken batchwise from the Purdue University water line and is passedthrough the demineralizer enroute to the pool. A vacuum breaker excludesany possibility of siphoning pool water into the supply line. The poolmakeup water system, in addition to the demineralizer, also includes anormally closed manual shutoff and throttle valve and a check valve.5.2.4 Primary Coolant Chiller System -The chiller is designed with three loops.Polwater passes through the primary loop, a Freoni refrigerantisnthsecondary loop, and water from the building iwa&ter._+, supply is used toremove heat, which is then discharged to ,thie ibuLding sewer system. Theheat-removal capacity of the heat I"D.,5!kW. It was designedto maintain the reactor pool temp~erature at 750 continuousoperation at 10 kW* \,,'%5.3 Reactor Core and Fuel K 5.3.1 The fuel assemblies shall be MTRi typ~econslstingenriched up to 20% in the IJ-235 isotdpe. iof aluminum clad plates5.3.2 A standard fuel assembly ~ha~ll onsist of up 'to 14 fuel plates containing amaximum of 180 grams of UJ-235,. ".,"-5.3.3 A control fueJ.assemibly sh~all,,copisist, of ;iip, o '8< fuel plates containing amaximum of'103-~gr~ms of 5.3.4 Partially loaded fuel, 4ssembhies,,in which some of the fuel plates arereplaced: by aluminum conhtaining no uranium may be used.5*3.5 The core -configurahion~shall consist of 13 standard fuel assemblies as,,.describe~d in<,53:.2~ and 3 control' fuel assemblies as described in 5.3.3,./ rods and one regulating rod.5".;3.6 Represenatitv'e fuel, assemblies shall be inspected annually, with no-. ,.Interval to excbeed 15 Shonths.5.4 FuelStorage I'5.4.1 All reactor fuelassemblies shall be stored in a geometric array where keffis for all conditions of moderation and reflection.5.4.2 Irradiated 'fuel assemblies and fueled devices shall be stored in an arraywhich will permit sufficient natural convection cooling by water or air suchthat the fuel integrity is maintained per the Safety Analysis Report.PUR-1 Technicai Specifications 5-2 Amendment No. 13PUR-1 Technical Specifications5-2Amendment No. 13

6. ADMINISTRATIVE CONTROLS6.1 Organization6.1.1 StructureThe reactor facility shall be an integral part of the School of Nuclear Engineeringof the Schools of Engineering at Purdue University as shown in Figure 6.1 andlisted belowa. The Dean of the College of Engineering will be the individualresponsible for the facility's licenses or b. The Laboratory Director (Level 2) or the deSignated alternate shall beresponsible for reactor facility operation. : c. The Reactor Supervisor (Level 3) sh~all ble the day-to-daysafe operation of the PUR-1. The< Supervisor sl~iali!!be responsiblefor assuring that all operations are c#onducted in a safe man~ner and withinthe limits prescribed by the _ licenrse, including 'they technicalspecifications and other applicable reg~ujtior~s.<d. In all matters pertaini~lg to. operation of the reactor and theadminstraive apect specifications, the LaboratoryDiretor(Leel ) [r 3) in the absence ofthe aboatoy Drecor]shlll T;eport to~ahd .be~directly responsible to theLevel 1 Dean the/College6b ~Engineernn. In all mattersing to \n?,mite.safety they, '-:-, be responsible to the Radiation//PUR-1 Technical Specifications 61AedetN.16-1Amendment No. 13 rIVice Presidentfor ResearchIIRadiationSafetyCornmmitteeIRadiationSafety Officer-IPresidentPurdue UniversityProvostPurdue UniversityCollege of EngineeringLevel 1SLaboratory DirectorLevel 2----- IIIIIommittee OnReactor*IOperations*SReactor OperationsPrimarily Administration--= -Primarily SafetyFIGURE 6.1: PUR-1 Organization6.1.2 Staffing(1) The minimum staffing when the reactor is not secured shall bePUR-1 Technical Specifications 62AedetN.16-2Amendment No. 13 (a) A licensed reactor operator in the reactor room,(b) The minimum crew for operating the reactor shall consist of 2 (two)persons, one of whom must be an NRC licensed member of thePUR-1 operations staff, the second crew member must beinstructed as to how to shut down the reactor in the event of anemergency.(c) A designated senior reactor operator (unless the operations staffconsists of only one senior reactor operator, and that individual isoperating the reactor) shall be present or rea'dily available on call atany time that the reactor is operating. 'Readily Available on Callmeans an individual who 7 (i) Has been specifically designates..and the designation isknown to the operator oni.duty.V, (ii) Can be rapidly phone or othe'r-.nethod by theoperator on duty, "",:,(iii) Is capable of gettinig. to the facility' within areasonable time under, normal conditions(2) No licensed reactor or senior r:rhactor operator shall be requiredwithin the I censed fac is 'secure.(3) Events requiring the presence at the facil~ty of senior reactor operator(a) Inta and approach to power following a core change. The-/presence..-. of ,, ,SRO at the, reactor facility is unnecessary for theinitial-daily start~up, pr[ovided,!the core remains unchanged from the~..-.......prevlo~u~sruo; ....j;:/: ./7i.h 7(b)'AII fu~el or con~trol-rod relocations within the core region;\ Recovery from,. ariunplanned or unscheduled shutdown.6.1.3 'M~ihimum Qual~ifiC1ations of Reactor Personnel The minimum qualificationsshoujld. be consistent with the American National Standard for theSelection, and JTraining of Personnel for Research Reactors, ANSI/ANS15.4, include the following:V~a. At the time of appointment to the position, the Level I Licensee shallreceive briefings sufficient to provide an understanding of the generaloperational and emergency aspects of the reactor facility.b. At the time of appointment to the position, the Laboratory Director(Level 2) shall have a minimum of five years of nuclear experience.The individual shall have a recognized baccalaureate or higher degreein an engineering or scientific field. Education or experience that isjob-related may be substituted for a degree on a case-by-case basis.PUR-1 Technical Specifications6-3Amendment No. 13 The degree may fulfill four years of the six years of nuclear experiencerequired on a one-for-one time basis. The individual shall receiveappropriate facility-specific training based upon a comparison of theindividual's background and capabilities with the responsibilities andduties of the position. Because of the educational and experiencerequirements of the position, continued formal training may not berequired. The Laboratory Director shall possess a valid SeniorOperator License, and meet the certifications requirements of thelicensing agency.c. Reactor Supervisor (Level 3) -At the time of,'p~pointment to the activeposition, the reactor supervisor shall have'a minimum of five years ofnuclear experience. He shall havei a /ba6calaureate degree orequivalent experience in an engineering/or otherN scientific field. Thedegree may fulfill four years of experience on a o5ne-for-one basis. Thereactor supervisor shall posses~si7a valid SeniorKOperator License.During periods when the Reae~ctor Supervisor is theseresponsibilities may be dele'gated to a S entior Reactor Operator (Leveld. Licensed Senior .(Level 4)"-.At-,the_. time of appointment to theactive position, a seriir-. 6perator sh"all have a minimum of a highschool diploma or equivalenft arid-should ,have four years of nuclearexperience. A maximum Of two~years 0f..experence may be fulfilled byrelated academiliicor tech~nical trai'ning-cn a, one-for-one time basis. Heshall hold a valid Senior TReactor Operator's license.e. Licensed Operator -At the fime, of appointment to the active position,an orto shall havea. ploma or equivalent. He shall.-,-hold a valid NRC Reactor-Opertor's license.,÷.,f. Operator, Trainee -.. An operator trainee shall have all the qualifications..to become\a hiCernsed operator except for possessing an operator's",",. license.' ", "%NPUR-1 Technicai Specifications 6-4 Amendment No. 13PUR-1 Technical Specifications6-4Amendment No. 13 6.1.4 A Radiation Safety Officer who is organizationally independent of thePUR-1 operations group shall advise the Laboratory Director and/orReactor Supervisor in matters concerning radiological safety. Minimumqualifications for the Radiation Safety Officer (RSO) is a bachelor's degreeor the equivalent in a science or engineering subject, including someformal training in radiation protection. The RSO should have at least fiveyears of professional experience in applied radiation protection. A master'sdegree may be considered equivalent to one year of professionalexperience, and a doctor's degree equivalent to two years of professionalexperience where course work related to radiatio.n:'protection is involved.At least three years of this professional be in appliedradiation protection work in a nuclear facilt'y d'eahing with radiologicalproblems. 6.1 .5 The Reactor Supervisor or his designateci alternate shall be responsiblefor the facility retraining and replacement training prog'ramx62 Review and Audit ,6.2.1 A Committee on Reactor Operations (.CORO),shall report to thie LEVEL ILicensee on matters of administration ~and~safety. CORO will advise theLaboratory Director aiqd/I-,-the Reactor TSupervisor on those areas ofresponsibility specified' in>---Sctions 6.2:.5w ,and 6.2.6. The minimumqualifications for COROshalltbe five years of professionalwork experience in the 'discipline< or-:specifici'-field they represent. Abaccalaureate degree may fulfill four~years ,of experience.6.2.2 The CORO, shall have at least\7 (seven) members of whom no more thana be directly conderned with the administration or direct useof the readtor. These rmembers shall,,include the following:.a-:"The Chaiirman,.-- reI6iil,.'senior technical person, knowledgeable*.<- ...in-the field o~f reactor technology, who does not have line responsibility..... fordayvt-tday operation of the reactor.-,, b. A senior r'adiation safety officer.senior ,riember of the Purdue University safety and securityd. The Laboratoc Directore. The Reactor Supervisor.f. Two senior scientific staff members.PUR-1 Technical Specifications 6-5 Amendment No. 13PUR-1 Technical Specifications6-5Amendment No. 13 6.2.3 The CORO shall meet no less than once per calendar year, or morefrequently as circumstances warrant, consistent with effective monitoringof facility activities. A sub-committee may be assembled by the CORO asthe need arises.Sub committees may be formed as needed, which may consist of aminimum of 3 (three) members, only one of which may have lineresponsibility for day-to-day operations of the reactor. These sub-committees may perform the functions of the whole committee asnecessary provided the review/audit functions are maintaind Actions bythe subcommittee must be approved by the whol "CORO.whleA quorum of CORO shall consist of not less a majority of the fullCommittee and shall include the chairman 'or/hi@ alternate. Nomore than one-half of the voting meenbers present shall be members ofthe reactor operations staff. /;.,i \:;x6.2.4 The CORO shall review and approve: .; :a. Safety evaluations for 1) changes ,t6' proedre, e"pn" osystems and 2) tests or experime~nts thadt may be conducted withoutprior NRC approval under the provisionh of Section 50.59, 10 CFR, toascetainwheter sIcha5ctions woul costitute an unreviewed safetyquestion, or would require a changeln Technical Specifications.b. Proposed -cthaiges to pce1rs, ~ qj~etor systems that changethe originaLirntentf or usevor th~ose thaftmight involve an unreviewedsafety qu'estion a's defined in Section 50.59, 10 CFR.c. Proposed tests or -exPer~imerits ~which are significantly different fromprevious :a~pbroved t~ests o::r and those that might involve//i --an.:unreview~ed safety" ---questionl as defined in Section 50.59, 10 CFR.<.,d. Proposed changes ,in Technical Specifications or licenses.",e. Violations of :applicable statutes, codes, regulations, orders, technical-. specifications,' license requirements, or of internal procedures or"inst r uctions# having nuclear safety significance.f. significanti operating abnormalities or deviations from normal andexpect'e/ performance of facility equipment that might affect nuclearsafety.g. Events which have been reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the NRC.h. Audit reports.PUR-1 Technical Specifications 6-6 Amendment No. 13PUR-1 Technical Specifications6-6Amendment No. 13 6.2.5 AuditsAudits of facility activities shall be performed under the cognizance of theCORO but in no case by the personnel responsible for the item audited.Individual audits may be performed by one individual who need not be anidentified CORO member. These audits shall examine the operatingrecords and encompass:a. The conformance of facility operation to the Technical Specificationsand applicable license conditions, to be done annually with no intervalto exceed 15 months. -.b. The performance training and qualifications offthe licensed facility staff,to be done annually with no interval to,exceed15 months.c. The results of all actions taken to ~eorrect defici~enicies occurrng infacility equipment, structures,.,systems or of Vpoperation thataffect nuclear safety, to be don9e,a:nnually with no inte~val to exceed 15months. "<.>\ / i/vd. The Facility Emergency Plan and im~ier'l~enting procedures, to be donebiennially with no minierl-to e. The Facility Security plaln and implementing procedures, to be donebiennially w~itb no interval':to exceed:21I2 years.f. Any other, area,:oft facihity :operation considered appropriate by theCORO ori the Reactor Superv~isor, to be done annually with no intervalto exkceed1 15 months. \',6.2.6 .Record~s "q ..".o.. ,Y1/'i:Recordsl of-CORO activities shall be prepared and distributed as indicatedMinutes oft each C'ORO meeting shall be prepared and forwarded to"\<,ibte Reactor, S'upervisor within 30 days following each meeting.b. Re;lports..of encompassed by section 6.2.4 e, f, and g above,sh~liie pr6fepared and forwarded to the Reactor Supervisor within 30days -'foillowing completion of the review.c. Audit reports encompassed by Section 6.2.5 above, shall be forwardedto the CORO Chairman and to the management responsible for theareas audited within 30 days after completion of the audit.6.3 Operating ProceduresWritten procedures, including applicable check lists reviewed and approved by theCORO, shall be in effect and followed for the following operations:PUR-1 Technical Specifications6-7Amendment No. 13 6.3.1 Startup, operation, and shutdown of the reactor.6.3.2 Installation and removal of fuel elements and control rods.6.3.3 Actions to be taken to correct specific and foreseen potential malfunctionsof systems or components, including responses to alarms and abnormalreactivity changes.6.3.4 Emergency conditions involving potential or actual release of radioactivity,including provisions for evacuation, re-entry, recovery, and medicalsupport.6.3.5 Maintenance procedures which could have an effect on reactor safety.6.3.6 Experiment installation, operation, and 6.3.7 Implementation of the Security Plan and Emnergency, Plan.6.3.8 Calibration and preventive mainte'nanceinstruments, systems, or components:/proce~dure#on requiredNon-routine operations which require th#, sequential efraneQ~ series ofsubtasks shal be carried out with the writtenp procedure9't the consol'e. To assureadherence to the documentation of the proce'durei each step will be entered in thelog book by the operator on duty a'~'itis completed.i'iiSubstantive changes to the above'~procedures .shall be. tnade only with the approvalof the CORO. The Reactor Supervisor ,or LaboratoryDirect'or may make changes toprocedures which do .not change the "intent procedure or impactnuclear safety. All ,suich-changes to~ ,ti'6 :procedures 9shal be documented andsubsequently reviewe~d by CORd. 6.4 Operating Records /.' 6.4.1,<.Thee and Iog~sshall be prepared and retained for at least,- -five years.::.: ,,,. ,",ia. Normal facility op:eration and maintenance.Principal m~aintenance operationsc Keor~o~/ ocurnec R-~tal 9ccrrecesd. Tests, cliecks, and measurements documenting compliance withsurveillatnce requirements.e. Facility radiation and contamination surveys.f. Records of experiments performed.g. Fuel inventories, receipts and shipmentsh. Approved changes of operating procedures.PUR-1 Technical Specifications6-8Amendment No. 13

i. Records of meeting and audit reports of the Committee on ReactorOperations.6.4.2 Record of retraining and requalification of certified operations personnelshall be maintained at all times the individual is employed or until thecertification is renewed.6.4.3 The following records and logs shall be prepared and retained for the lifeof the facility:a. Gaseous and liquid waste released to the environs.b. Offsite environmental monitoring surveys. c. Radiation exposures for all PUR-1 personn~el.x,, -9...d. Updated, corrected, and as-builtJfaciity,, ,, drawings. 'qxie. Annual operating reports. ""f. Reviews of instances where the safety Iimpit'was exceeded.g. Reviews of failure ,of ,the.,automatic-,safety system that protects thelimiting safety system, set gA-LSs)'.....h. Reviews of. instances where lmiting conditions of operation were not6.5 Reqiuired Actions ,, \,The following actions 'shall,,be taken"relatirngt the types of events listed in Secs.6.6.1 and6.6:62~-*, ". 7 ;' -96.,5.1 The following actions shall be taken in the event the Safety Limit isN, vi. olated: ...,, -, , >."-(.1),\The reacto'rwill be sh'ut down immediately and reactor operation will not"".. be resumed ~without authorization by the Commission.(2) The 'S, afety >Limit Violation shall be reported to the Director of theNRC Office of Inspection and Enforcement (or designee),the Laboratory Director and to the CORO not later than the next workday.(3) A Safety Limit Violation Report shall be prepared. The report shall bereviewed by the CORO. This report shall describe (1) applicablecircumstances preceding the violation, (2) effects of the violation uponfacility components, systems or structures, and (3) corrective actiontaken to prevent recurrence.PUR-1 Technical Specifications6-9Amendment No. 13 (4) The Safety Limit Violation Report shall be submitted to the Commission,the CORO and the Reactor Supervisor within 14 days of the violation, insupport of a request to the Commission for authorization to resumeoperations.6.5.2 The following actions are to be taken in the event of reportable occurrenceas defined in 6.6.2.(1) Reactor conditions shall be returned to normal, or the reactor shall beshut down. If it is necessary to shut down the reactor to correct theoccurrence, operations shall not be resumed unles,s authorized by Level2or designated alternates;, (2) Occurrence shall be reported to Level 2/br de6"sig'nated alternates and to.. .. ';> -?-chartering or licensing authorities as required; "., "(3) Occurrence shall be reviewed b~y th,.e'rve group at ,its next scheduledmeeting..: -.\...6.6 Reporting Requirements ;:"The following information shall to the, USNRC in addition to the reportsrequired by Title 10, Code of ",'\"6.6.1 Annual Operating Reports"- 'a rep/6Yt co'ver~ing .tbe previous year shall besubmitted toPtheIBirepctor of the Ofifw5ie ofNuclear Reactor Regulation witha copy to, the" b~y March 31 of each year. Itshall incelud1e the followihg: a. Changes.!n plant design and operation7 1. changes:,n facility design-.,./ .2. perfo (e.g. equipment and fuel performance).% \ ,,.",3. changesin~ operating procedures which relate to the safety of facilityoperatiorlis \4f. results of surveilance tests and inspections required by thesetech-nicaIs specifications5. a brief summary of those changes, tests, and experiments whichrequired authorization from the Commission pursuant to 10 CFR50.59(a)b. Power Generation -A tabulation of the thermal output of the facilityduring the reporting period.c. Shutdowns -A listing of unscheduled shutdowns which have occurredduring the reporting period, tabulated according to cause, and a briefPUR-1 Technical Specifications6-10Amendment No. 13 discussion of the corrective and preventive actions taken to preventrecurrence.d. Maintenance -A discussion of corrective maintenance (excludingpreventive maintenance) performed during the reporting period onsafety-related systems and components.e. Changes, Tests, and Experiments -A brief description and a summaryof the safety analysis and evaluation for those changes, tests, andexperiments which were carried out without prior Commissionapproval, pursuant to the requirements of 10, 50.59(b).f. Radioactive Effluent Releases -A sumrny/of'the nature, amount, andmaximum concentrations of released ordischarged to the environs beyondtlhe effective~control of the licenseeas measured at or prior to the point of such release'or discharge.g. Occupational Personnel Raito Exposure -A smnrofradiationexposures greater than 25% 'of:the app~rop~riate limits of tO CFR 20received during the reporting "period: by" facility personnel (faculty,students, or experimenters). % .N. , \6.6.2 Non-Routine Reports ",, .,"\" ' .,-_ .. ;,.a. Special Reports /i, " %Special reports are< ,us~ed to ~rei~r unplanne'd events as well as plannedmajor fac5iity and adhmi~pstratie~ cbanges. The following schedule shall beincorporafed in the specifications;,:,.-*--(1)_Ther~e~shall:,be a report nr~t.'later than the following working day by..... lephohe confirmed in writing by facsimile or similar.,:/ c~onvey~ance 'to licensing authorities, to be followed by a written'. ::.,, report,,that d~esc~rbes the circumstances of the event within 14 days"--,-'\ of any 'of the following:\", , (a) opei'ation with actual safety system settings for required"a/:: less conservative than the limiting safety systemspecified in the technicalspcfato,(b) operation in violation of limiting conditions for operationestablished in the technical specifications unless promptremedial action is taken as permitted in Sec. 3,(c) a reactor safety system component malfunction that rendersor could render the reactor safety system incapable ofperforming its intended safety function. If the malfunction orcondition is caused by maintenance, then no report isrequired.Pun-1 Technical Specifications6-11Amendment No. 13 (d) an unanticipated or uncontrolled change in reactivity greaterthan 0.6% Ak/k,(e) abnormal and significant degradation in reactor fuel orcladding, or both, coolant boundary,(f) an observed inadequacy in the implementation ofadministrative or procedural controls such that theinadequacy causes or could have caused the existence ordevelopment of an unsafe condition regard to reactoroperations; / /;(2) There shall be a written report within days to the chartering orlcensing authorities of the following! // (a) Permanent changes ,irf'i th facility organization involvingLevel 1 or 2 personnel /' a (b) significant changes7 in the transipnt or accident ~analysis asdescribed in the Safety,,Analysis' Report.b. Unusual Events"<2~' Va'7* &'a "'a ", ... -...A written report shall be~ forwarded, within~t30 days to the Director,Office of Nuclear Reactor~ Reg~ulat~ion~with a ~copy to the in the event of:-- N", 2 \ , ""-t' :;1 .,Discovery of ',any sdbstantial errors in the transient or accident< ,!aialyses o~ in1 the method~s used for such analyses, as describedthe SAR o~r-the bases, for the Technical Specifications.:"2~.--Disicove r of ani'y-'- s~ubstantial variance from performance""""".specifications contained in the Technical Specifications, in the3. Discov'ery condition involving a possible single failurea', which, ,f or<, a system designed against assumed single failures,could re'sult in a loss of the capability of the system to perform itsa-i safety function.4. "'Discovery of an inadequacy in the implementation ofadministrative or procedural controls such that the inadequacycauses or could have caused the existence or development of anunusual condition with regard to reactor operations.PUR-1 Technical Specifications 6-12 Amendment No. 13PUR-1 Technical Specifications6-12Amendment No. 13 Cost Analysis for Decommissioning of Purdue University Reactor 1 (PUR-1)The following analysis for decommissioning of the Purdue University Reactor 1 (PUR- 1)is based on the analysis done by the Department of Defense (DOD) [1] for the AFRRITRIGA reactor facility and the University of Utah Safety Analysis Report (SAR) [2].The cost analysis reflects decommissioning through the decontamination of the reactorsite, referred to as DECON. DECON costs include the removal of equipment, structures,and portions of the facility that contain radioactive contaminants. The removal of spentnuclear fuels and demolition of any uncontaminated areas of the site are consideredancillary costs.The cost of decommissioning is divided into three major categories:Waste disposal costsLabor costs>" Energy costsDetailed data is provided for each of the major categories of costs based on the report byDOD [1] and taking into account differences in design. The amounts are adjusted to2015 dollars using the Bureau of Labor Statistics Consumer Price Index (CPI) taken fromreference [3].Waste Disposal CostsThe amount of structural material that has been exposed to neutron irradiation in thereactor building and the cost for transportation are provided in Table 1. The cost ofcrates and shipping are obtained from [1] which is developed based on data provided inNUREG/CR-1756 [4]. For the purposes of this report, the conservative scenario ofshipment to a destination in Washington State has been considered which cost $0.12/kgof low level waste in 1989 dollars (equivalent to $0.23/kg in 2015 dollars). The cost pervolume for disposing of radioactive waste in a depository was obtained from [ 1], which isbased on Barnwell rates of $2,825/mi3 in 1989 dollars (equivalent to $5,348/mi3 in 2015dollars). Plywood 3.5 m3 crates are used for removing the waste which cost $400 each in1981 dollars (equivalent to $1,033 in 2015 dollars).Table 1 Waste disposal costs in 2015 dollarsShippinMaterial Volume Crates Density Mass g Total Cost*m3 No. kg/rn3 kg USD USDContaminated concrete 10 3 2,400 24,000 $5,520 $62,099Contaminated sand 60 18 1,442 86,490 $19,893 $359,367Contaminated aluminum 5 2 2,700 13,500 $3,105 $31,911Contaminated stainlesssteel 5 2 8,050 40,250 $9,258 $38,064Total $37,776 $491,441Total Cost = (cost/crate)*(#* of crates) + shipping costs + disposal costs The volumes are rounded up in order to estimate the costs conservatively. The shippingcosts are adjusted to 2015 dollars based on [1] and the highest value (stainless steel) isused for all the materials to be conservative in estimating the cost.Labor CostsThe labor costs are obtained from [1] which is based on NUREG/CR-1756. The PUR-1is smaller than the AFRRI TRIGA facility; however, the numbers are unchanged toprovide a conservative estimate of the labor costs. The amounts are adjusted based onCPI from 1981 dollars to 2015 dollars.Table 2 Decommissioning labor costs (for DECON) in 2015 dollarsWork years Rate (2015 dollars) CostM~anagementand suprtstfDecomm superintendent 2 $230,100 $460,200Decomm engineer 2 $196,200 $392,400S ecre t ary 2 $62,500 $125,000Clerk 0.5 $62,500 $31,300Health physicist 2 $121,100 $242,200Radioactive shipment specialist 0.5 $101,500 $50,800Procurement specialist 0.5 $101,500 $50,800Contract and accounting specialist 0.8 $121,600 $97,300Security supervisor 0.625 $144,300 $90,200Security patrol officer 3.6 $65,600 $236,200QA engineer 0.7 $121,100 $84,800Control room operator 1 $88,600 $88,600Co n sulItan t 1 $258,200 $258,200Decomm workers .. ...: ... .... ... ..Shift engineer 1 $134,800 $134,800Crafts m an 2 $82,900 $165,800Crew leader 0.5 $114,600 $57,300Utility operator 0.342 $82,900 $28,400Lab ore r 6 $79,800 $478,800Health physics technician 3 $77,500 $232,500Total $3,305,600 Energy CostsThe energy costs are also obtained from [1] which is based on NUREG/CR-1756 and theenergy cost per kWh is obtained from the U.S. Department of Energy InformationAdministration Electric Power Monthly report [51. The average retail price of electricityin the state of Indiana for all sectors for December 2014 (YTD) was 8.97 cents per kWh.Table 3 Energy costsEquipment Energy use (kWh) Cost ($)General system 9,000 $807HV AV 20,000 $1,794Lighting 23,000 $2,063Control room 5,200 $466Fire protection 600 $54S ecu rity 5, 600 $502Communications 900 $81Domestic water 36,300 $3,256Re actor wate r 23,400 $2,099Compressed air 15,000 $1,346Building heating 302,600 $27,143Decommissioning equipment 20,000 $1,794Total (USD) $41,406Total Decommissioning Cost and Inflation Adjustment MethodologyThe total cost for the reactor decommissioning based on the costs detailed above isprovided in Table 4. The cost of spent fuel removal, shipment, and site demolition costsare also estimated in table 4.Table 4 Total cost of decommissioning PUR-1 in 2015 dollars, DECON methodologyCategory CostDECON .. ...Waste disposal $491,441Labor $3,305,600Energy $41,406Contingency fund(25% of decommissioning costs) $5,1Ancillary : ,..Spent fuel removal and shipment $387,330Site demolition $645,550Total Cost $5,830,938 The estimated cost of decommissioning the Purdue University Reactor 1 (PUR- 1) isreflected in 2015 dollars using CPI as the basis for adjustment to 2015 dollar values. Acontingency fund equal to 25% of decommissioning costs is added to the total cost asrequired by NUREG- 1756. Ancillary costs were also obtained from [ 1].References[1] M. Forsbacka, M. Moore. An Analysis of Decommissioning Costs for the AFRRITRIGA Reactor Facility. Defense Nuclear Agency, Armed Forces RadiobiologyResearch Institute. Bethesda, Maryland 20814-5 145[2] The University of Utah Reactor (UUTR) Safety Analysis Report, 2005[3] United States Department of Labor, Bureau of Labor Statistics CPI InflationCalculator [1], April 2015[4] U.S. Nuclear Regulatory Commission, NUREG/CR-1756 "Technology, Safety andCosts of Decommissioning Reference Nuclear Research and Test Reactors", 1983[5] U.S. Department of Energy Information Administration Electric Power Monthly,February 2015 Report with data for December 2014[http ://www.eia. gov/electricity/monthly/current year/february20 15 .pdf]

OPERATOR REQUALIFICATION PROGRAMfor thePUR-1 REACTOR FACILITYThis program is designed to comply with the intent of 10 CFR 55, Appendix A,concerning the continued training and requalification of operators for the PUR-1 reactor.It will be mandatory for all operators licensed on the PUR-1 reactor to participate in theprogram.The requalification program will consist of the following parts:A. INSTRUCTIONA series of eight meetings will be held over a two year period, during which alltopics listed below in part A.1 .b will be covered.1. Each meeting will consist of:a. A review of reactor operations and modifications, if any.b. A lecture of one or more of the following topics:ii.iii.iv.V.vi.vii.viii.ix.Theory and principles of operations.General and specific plant operating characteristics.Plant instrumentation and control systems.Plant protection systems.Engineered safety systems.Normal, abnormal, and emergency operating procedures.Radiation control and safety.Technical specifications.Applicable portions of Title 10, Chapter I, Code of FederalRegulations.2. The lectures will be given by the reactor operators, senior operators, universityradiation control officers, or faculty members of the School of NuclearEngineering.B. PROGRAM EVALUATIONCompletion of the biennial operator requalification program will consist of a writtenexamination and an annual demonstration of operator proficiency in reactor operation.1. Written examination:a. One of the senior operators will be exempt from taking the examination. Thissenior operator will make up and administer the examination to all otheroperators and senior operators. The senior operator may receive assistancePUR-1 Requalification ProgramPage I of 5March 29, 2015 for making up questions on the topics in part A.1 .b from the instructor for eachtopic. The senior operator exemption will rotate through the entire senioroperator roster.b. The written examination for requalifying licensees will contain representativequestions measuring the knowledge, skills and abilities needed to performlicensed duties. These will be identified from the licensed operator's dutiesperformed, information in the Safety Analysis Report, operating procedures,facility license and amendments, License Events Reports, and any otherinformation requested from the facility licensee by the NRC.c. The representative questions for the operators examination will sample thefollowing topics:i. Fundamentals of reactor theory including the fission process, neutronmultiplication, source effects, control rod effects, criticality indications,reactivity coefficients, and poison effects.ii. General design features of the core, fuel assemblies, control rods, coreinstrumentation, and coolant flow.iii. Mechanical components and design features of the reactor coolantsystem.iv. Auxiliary systems that affect the facility.v. Facility operating characteristics during steady state and transientconditions.vi. Design, components, and functions of reactivity control mechanismsand instrumentation.vii. Design, components, and functions of control and safety systemsincluding instrumentation, signals, interlocks, failure modes, andautomatic and manual features.viii. Components, capability, and functions of emergency systems.ix. Shielding, isolation, and containment design features, including accesslimitations.x. Administrative, normal, abnormal, and emergency operatingprocedures.xi. Purpose and operation of radiation monitoring systems, includingalarms and survey equipment.xii. Radiological safety principles and procedures.xiii. Procedures and equipment available for handling and disposal ofradioactive materials.d. Representative questions for the senior operators examination will sample thetopics in the operators list and in addition will sample the following list:i. Conditions and limitations in the facility license.ii. Facility operating limitations in the technical specifications and theirbases.PUR-1 Requalification ProgramPage 2 of 5March 29, 2015 iii. Licensee procedures required to obtain authority for design andoperating changes.iv. Radiation hazards that may arise during normal and abnormalsituations including maintenance activities and various contaminationconditions.v. Assessment of facility conditions and selection of appropriateprocedures during normal, abnormal, and emergency situations.vi. Procedures and limitations involved in initial core loading, alterations incore configuration, and determination of various internal and externaleffects on core reactivity.vii. Fuel handling facilities and procedures.e. Any person who scores less than 70%, overall, on the examination will berelieved from licensed duties and enrolled in an accelerated program untilsuch time as they can satisfactorily pass an examination covering thematerial. The course content and duration will depend upon the individual'sdeficiencies.2. Operator proficiency:a. The exempt senior operator will also administer an annual operatorproficiency examination to all other operators and senior operators.b. The content of the operating test will be identified from duties of thelicensed operator/senior operator and reference documents listed in PartB.l.bc. The operations test for the requalifying licensee will demonstrate anunderstanding of and the ability to accomplish a representative sample ofthe following items:i. Perform the prestartup procedures.ii. Manipulate the console controls as required to operate the facilitybetween shutdown and designed power levels.iii. Identify annunciators and condition-indicating signals and performappropriate remedial actions where appropriate.iv. Identify the instrumentation systems and the significance of facilityinstrument readings.v. Observe and safely control the operating behavior characteristics ofthe facility.vi. Perform control manipulations required to obtain desired operatingresults during normal, abnormal, and emergency situations.vii. Safely operate the facilities auxiliary and emergency systems.viii. Demonstrate or describe the use and function of the facilitiesradiation monitoring systems, including fixed radiation monitors andPUR-1 Requalification Program Page 3 of 5 March 29, 2015PUR-1 Requalification ProgramPage 3 of 5March 29, 2015 alarms, portable survey instruments, and personal monitoringequipment.ix. Demonstrate knowledge of significant radiation hazards and theability to perform procedures to reduce excessive levels of radiationand to guard against personal exposure.x. Demonstrate knowledge of the emergency plan, including, asappropriate, the operator's or senior operator's responsibility todecide whether the plan should be executed and the dutiesassigned under the plan.xi. Demonstrate the knowledge and ability, as appropriate to theassigned position to assume the responsibilities associated with thesafe operation of the facility.xii. Demonstrate the ability to act as a member of the operations crewso that all procedures, the limits to the license and its amendmentsare not violated.d. Any person who cannot demonstrate proficient operation of the reactor willbe relieved of his licensed duties until such time as proficient operationcould be demonstrated. Proficient operation may be established byperforming a minimum of six hours of supervised reactor operations anddemonstrating proficiency of section B.2.C. ON THE JOB TRAINING1. Each licensed operator in the requalification program may at the option of theexempt senior operator, be required to make 8 reactor startups, shutdowns, orpower level changes during the two year period covered by the program.2. Each licensed operator at the facility will manipulate the plant controls, and eachlicensed senior operator will either manipulate the plant controls or direct theactivities of individuals during plant control manipulations during the term of theoperators/senior operator's license. Manipulations by operators/senior operatorsmust consist of the following activities:a. Completed annually.i. Plant shutdown.ii. Significant power changes (>10%).iii. Loss of coolant. (Not considered credible)iv. Loss of electrical power.v. Loss of coolant flow. (Not considered credible)b. Completed on a two-year cycle.i. Loss of protective system channel.ii. Mispositioned control rod or rods.PUR-1 Requalification ProgramPage 4 of 5March 29, 2015 iii. Inability to drive control rods.iv. Conditions requiring use of emergency borationv. Fuel cladding failure or high activity in reactor coolant.vi. Failure of servo system.vii. Reactor trip.viii. Failure of nuclear instrumentation.Note: When the control panel of the facility is used for training, the action taken or to betaken for the emergency or abnormal condition may be discussed; actual manipulationof the controls is not required per 10 CFR 55.59 (c) (4) (iv).3. Each licensed operator at the facility will perform the function of the license held.An SRO is credited with performing the function any time the operator is on call,instructing classes, student, or student operator in training, inside the reactorroom with the key on, or maintaining custody of the key. Additionally unstructuredactivities such as participation in facility-related design and safety review groups,Emergency Plan, emergency drill, Committee on Reactor Operations (CORO)participation, experimental activities, related technical presentations, performingsecurity related functions, and performing maintenance and calibration activitiescontribute to training in all parts of the program except parts B.2, C, D and E. Astatement to the file is sufficient to document the training and/or time accounting.D. LITERATURE REVIEWEach reactor operator and senor operator will annually review the contents of theoperating manual, technical specifications, and the emergency procedures. A statementto this fact will be kept in the requalification file.E. RECORDSRecords will be maintained to document each instructor, each topic discussed, eachlicensed operator's and senior operator's participation in the requalification program.The records will contain copies of each written exam, answer sheets, results ofevaluation, and the biennial operator proficiency demonstration. Documentation ofadditional training and test required for individuals exhibiting deficiencies will also beincluded in the files. All records of the requalification program will be retained by thetraining coordinator until the licenses of the participants are renewed.In any of the above requalification, exclusive of operations, mail, electronic classroom orother methods may be used for training, meetings, testing or other requiredcommunication/s.PUR-1 Requalification Program Page 5 of 5 March 29, 2015PUR-1 Requalification ProgramPage 5 of 5March 29, 2015 SAFETY ANALYSIS REPORTfor thePURDUE UNIVERSITYPUR-1 REACTORLICENSE NUMBER R-87DOCKET NUMBER 50-1 82Prepared by:J. H. Jenkins, E. C. Merritt, June 30, 2008Additional Revisions by:C. Townsend, R. Bean July 23, 2015West Lafayette, IN 47907 TABLE OF CONTENTSTABLE OF CONTENTS ................................................................................ ILIST OF FIGURES.................................................................................... VIILIST OF TABLES ...................................................................................... X1 THE FACILITY ..................................................................................I1-11.1 Introduction................................................................................. 1-11.2 Summary and Conclusions on Principal Safety Considerations ....................... 1-11.3 General Description ....................................................................... 1-11.4 Shared Facilities and Equipment......................................................... 1-11.5 Comparison with Similar Facilities ....................................................... 1-21.6 Summary of Operations................................................................... 1-21.7 Facility Modifications and History ........................................................ 1-22 SITE CHARACTERISTICS .................................................................... 2-12.1 Geography and Demography............................................................. 2-12.1.1 Site Location and Description ........................................................... 2-12.1.2 Population Distribution ................................................................... 2-12.2 Nearby Industrial, Transportation and Military Facilities................................ 2-72.3 Climatology and Meteorology ............................................................ 2-82.3.1 General and Local Climate .............................................................. 2-82.3.2 Weather.................................................................................. 2-112.3.3 Severe Weather .......................................................................... 2-122.4 Hydrology................................................................................. 2-142.5 Geology and Seismology................................................................ 2-152.5.1 Regional Geology ....................................................................... 2-152.5.2 Seismology .............................................................................. 2-152.6 References ............................................................................... 2-22PUR-1 SAR i Rev 2, July 23, 2015 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS ..... ..................... 3-13.1 Design Criteria............................................................................. 3-13.2 Meteorological Damage................................................................... 3-13.3 Water Damage ............................................................................ 3-13.4 Seismic Damage .......................................................................... 3-13.5 Systems and Components................................................................ 3-24 REACTOR DESCRIPTION .................................................................... 4-14.1 Summary Description..................................................................... 4-14.2 Reactor Core .............................................................................. 4-44.2.1 Reactor Fuel .............................................................................. 4-54.2.2 Control Rods ............................................................................ 4-104.2.3 Neutron Moderator and Reflector...................................................... 4-114.2.4 Neutron Startup Source ................................................................ 4-114.2.5 Core Support Structure ................................................................. 4-114.3 Reactor Pool.............................................................................. 4-114.4 Biological Shield ......................................................................... 4-124.5 Nuclear Design........................................................................... 4-134.5.1 MCNP Model ............................................................................ 4-134.5.2 Normal Operating Conditions........................................................... 4-164.5.3 Reactor Core Physics Parameters..................................................... 4-264.6 Thermal-Hydraulic Design .............................................................. 4-384.6.1 NATOON Code Description ................................................... .........4-384.6.2 Fuel Element and Fuel Assembly Geometry.......................................... 4-384.6.3 Thermal Hydraulic Analysis Results ................................................... 4-434.7 References ............................................................................... 4-465 REACTOR COOLANT SYSTEMS ............................................................ 5-1PuR-1 SAR ii Rev 2, July 23, 2015 5.1 Summary Description ..................................................................... 5-15.2 Primary Coolant System .................................................................. 5-25.3 Secondary Coolant System .............................................................. 5-25.4 Primary Coolant Cleanup System.............................................. .......... 5-25.5 Primary Coolant Makeup Water System................................................. 5-35.6 Nitrogen-16 Control System .............................................................. 5-35.7 Auxiliary Systems Using Primary Coolant ............................................... 5-36 ENGINEEREDOSAFETY ....................................................... 6-16.1 Summary Description..................................................................... 6-16.2 Detailed Descriptions...................................................................... 6-16.2.1 Confinement .............................................................................. 6-16.2.2 Containment .............................................................................. 6-16.2.3 Emergency Core Cooling System ....................................................... 6-27 INSTRUMENTATION AND CONTROL SYSTEMS.......................................... 7-17.1 Summary Description..................................................................... 7-17.2 Design of Instrumentation and Control Systems........................................ 7-27.2.1 Channel 1--Start-up Channel ........................................................... 7-37.2.2 Channel 2--Log N and Period Channel ................................................ 7-37.2.3 Channel 3--Linear Power................................................................ 7-47.2.4 Log-N Period Channel.................................................................... 7-47.2.5 Channel 4--Safety Channel............................................................. 7-57.3 Reactor Control System................................................................... 7-57.4 Reactor Protection System ............................................................... 7-87.5 Control Console and Display Instruments.............................................. 7-117.6 Radiation Monitoring Systems.......................................................... 7-118 ELECTRICAL POWER SYSTEMS ........................................................... 8-1PUR-1 SAR iiRev 2, July 23, 2015 8.1 Normal Electrical Power Systems........................................................ 8-18.2 Emergency Electrical Power System .................................................... 8-19 AUXILLARY SYSTEMS........................................................................ 9-19.1 Heating, Ventilation and Air Conditioning Systems..................................... 9-19.2 Fuel Storage and Handling ............................................................... 9-19.3 Fire Protection Systems................................................................... 9-29.4 Communication Systems ................................................................. 9-39.5 Possession and Use of Byproduct, Source and Special Nuclear Material ............ 9-310 EXPERIMENT FACILITIES AND UTILIZATION ...........................................10-110.1 Summary Description....................................................................10-110.2 Experimental Facilities...................................................................10-110.3 Experiment Review ......................................................................10-111 RADIATION PROTECTION AND WASTE MANAGEMENT ................................11-211.1 Radiation Protection Program .........................................11-211.1.1 Radiation Sources.....................................................................11-211.1.2 Gaseous Effluents.....................................................................11-211.1.3 Estimated releases in the Restricted Area...........................................11-711.1.4 Radiation Protection Program ........................................................11-811.1.5 ALARA Commitment...................................................................11-811.1.6 Radiation Monitoring and Surveying .................................................11-811.1.7 Radiation Exposure Control and Dosimetry .........................................11-911.1.8 Contamination Control.................................................................11-911.1.9 Environmental Monitoring.............................................................11-911.2 Radioactive Waste Management ...................................................... 11-1011.2.1 Radioactive Waste Management Program ......................................... 11-1011.2.2 Radioactive Waste Controls ......................................................... 11-10PUR-1 SAR iv Rev 2, July 23, 2015 11.2.3 Release of Radioactive Waste ...................................................... 11-1011.3 Conclusions............................................................................. 11-1012 CONDUCT OF OPERATIONS ...............................................................12-112.1 Organization..............................................................................12-112.1.1 Structure ................................................................................12-112.1.2 Responsibility ..........................................................................12-212.1.3 Staffing, Selection and Training of Personnel .......................................12-212.1.4 Radiation Safety .......................................................................12-312.2 Review and Audit Activities..............................................................12-312.3 Procedures ...............................................................................12-312.4 Required Actions.........................................................................12-312.5 Reports ...................................................................................12-412.5.1 Annual Operating Reports ............................................................12-412.5.2 Non-Routine Reports..................................................................12-412.5.3 Unusual Events........................................................................12-512.6 Records...................................................................................12-512.7 Emergency Planning.....................................................................12-512.8 Security Planning.........................................................................12-512.9 Quality Assurance .......................................................................12-612.10 Operator Training and Requalification...............................................12-613 ACCIDENT ANALYSES......................................................................13-113.1 Accident-Initiating Events and Scenarios ..............................................13-113.1.1 Maximum Hypothetical Accident......................................................13-113.1.2 Insertion of Excess Reactivity ........................................................13-113.1.3 Loss of Coolant ........................................................................13-213.1.4 Loss of Coolant Flow.................................................................. 13-3PUR-1 SAR v Rev 2, July 23, 2015 13.1.5 Mishandling or Malfunction of Fuel...................................................13-313.1.6 Experiment Malfunction ...............................................................13-413.1.7 Loss of Normal Electrical Power......................................................13-413.1.8 External Events........................................................................13-413.2 Accident Analysis and Determination of Consequences..............................13-513.2.1 Maximum Hypothetical Accident (Mishandling or Malfunction of Fuel) ............13-513.2.2 Insertion of Maximum Allowed Excess Reactivity ................................. 13-2413.3 Summary and Conclusions ............................................................ 13-2813.4 References.............................................................................. 13-2814 TECHNICAL SPECIFICATIONS.............................................................14-114.1 Summary Description of the Document................................................14-114.2 Administrative Control of Technical Specifications ....................................14-115 FINANCIAL QUALIFICATIONS..............................................................15-215.1 Financial Ability to Operate a Non-Power Reactor.....................................15-215.2 Financial Ability to Decommission the Facility ......................................... 15-2APPENDIX 1: PUR-1 DRAWINGS................................................................... 1APPENDIX 2: NATCON INFORMATION.............................................................APPENDIX 3: FUEL SPECIFICATIONS.............................................................PUR-1 SAR vi Rev 2, July 23, 2015PUR-1 SARviRev 2, July 23, 2015 LIST OF FIGURESFigure 2-1: State of Indiana, showing location of Tippecanoe County........................... 2-2Figure 2-2: Map of West Lafayette, indiana, showing inset picture of the location of theElectrical Engineering Building on the Purdue Campus.2 ................................. 2-3Figure 2-3: Map of the Duncan Annex of the Electrical Engineering Building................... 2-4Figure 2-4: Population of area surrounding PUR-1 reactor....................................... 2-5Figure 2-5: Population projection for 2030 in area surrounding PUR-1 reactor. ............... 2-6Figure 2-6: Map showing the location of the Purdue University airport in relation to the locationof PUR-1Y2.................................................................................... 2-8Figure 2-7: Map of the 1-Hz spectral acceleration for 2% probability of exceedance in 50 yearsfor the Central and Eastern United States in standard gravity (g) ...................... 2-17Figure 2-8: Map of the 5-hertz spectral acceleration (SA) for 2% probability of exceedance in 50years in the Central and Eastern United States in standard gravity (g).'3 .............. 2-18Figure 2-9: Map of peak ground acceleration (PGA) for 2% probability of exceedance in 50years in the Central and Eastern united States in standard gravity (g).'3 .............. 2-19Figure 2-10: Map of 1-hertz spectral acceleration (SA) for 10% exceedance in 50 years in theCentral and Eastern United States in standard gravity (g).i ............................. 2-20Figure 2-11: Map of peak ground acceleration (PGA) for 10% probability in 50 years in theCentral and Eastern United States in standard gravity (g).13 ............................ 2-21Figure 4-1: PUR-1 Pool Layout..................................................................... 4-1Figure 4-2: PUR-1 Grid Plate ....................................................................... 4-5Figure 4-3: Standard assembly can detail, showing wall spacers................................ 4-7Figure 4-4: Control assembly can detail, showing wall spacers.................................. 4-8Figure 4-5: Standard fuel assembly ................................................................ 4-9Figure 4-6: Control fuel assembly................................................................... 4-9Figure 4-7: Plates are identified by unique serial numbers as shown here .................... 4-10Figure 4-8: Dummy plates are differentiated from the fuel plates by a notch machined into theend of the plate.............................................................................. 4-10Figure 4-9: PUR-1 Reactor pool temperature measurements from 1994-2007................ 4-12Figure 4-10: plate in X-Z plane .................................................................... 4-14PUR-1 SARviiRev 2, July 23, 2015 Figure 4-1 1: plate in Y-Z plane. (magnified)...................................................... 4-14Figure 4-12: plate in X-Y plane, cutaway view .............. .................................... 4-14Figure 4-13: Model representation of standard assembly plate spacing detail showing wallspacers ...................................................................................... 4-15Figure 4-14: Comparison of standard and control assemblies in the model ................... 4-15Figure 4-15: Model of regulating rod assembly................................................... 4-16Figure 4-16: Representation of LEU core load................................................... 4-16Figure 4-17: PUR-1 Core Layout and Bundle Drawings......................................... 4-18Figure 4-18: Axial Power Profiles in Plates 1348, 1228 and 1315 for Banked Rod CriticalConfiguration in PUR-1..................................................................... 4-22Figure 4-19: Radial Power Profiles in Plates 1345, 1228 and 1315 for Banked Rod CriticalConfiguration in PUR-1..................................................................... 4-22Figure 4-20: Calibration curve for SS-1 rod with calculated and measured values ........... 4-28Figure 4-21: Calibration curve for SS-2 rod with calculated and measured values ........... 4-28Figure 4-22: Calibration curve for RR with calculated and measured values.................. 4-29Figure 4-23: Figure showing water regions for perturbation models ........................... 4-33Figure 4-24: PUR-1 Core Layout with LEU Fuel ................................................. 4-34Figure 4-25: Density of sub-cooled water at 1.5 atmospheres................................... 4-36Figure 4-26: Effect of Fuel Temperature, Water Temperature and Water Density Perturbationson Core Reactivity.......................................................................... 4-37Figure 4-27: Standard LEU fuel assembly........................................................ 4-38Figure 4-28: Control LEU fuel assembly.......................................................... 4-39Figure 9-1: MCNP model of in-pool fuel storage rack............................................. 9-1Figure 9-2: k-eff Values for Flooded Condition of Fuel Storage Facility ......................... 9-2Figure 9-3: MCNP model of dry fuel storage facility....................................... .'....... 9-2Figure 12-1: Safety and Administration Responsibilities for the PUR-1 Facility................ 12-1Figure 13-1: Power and Clad Temperatures for 0.6%Ak/k step insertion with scram........ 13-25Figure 13-2: Power and Clad Temperatures for 0.6%Ak/k slow insertion with scram......... 13-26PUR-1 SARviiiRev 2, July 23, 2015 Figure 13-3: Power and Clad Temperatures for 0.6%Ak/k step insertion without scram....13-27Figure 13-4: Power and Clad Temperatures for 0.6%Ak/k slow insertion without scram. .... 13-28PUR-1 SARixPUR-1 AR IxRev 2, July 23, 2015 LIST OF TABLESTable 1-1 : Summary of amendments and changes to the PUR-1 reactor facility ............... 1-3Table 2-1: Population Data for Reactor Vicinity, centered on reactor location .................. 2-1Table 2-2: Purdue University campus population detail for 1998-2008.......................... 2-7Table 2-3: Climatography Data for West Lafayette, Indiana; Mean and Extreme Temperatures...............................................................................................2-9Table 2-4: Climatography Data for West Lafayette, Indiana; Mean Days Information6 ........ 2-10Table 2-5: Precipitation Normals for West Lafayette, Indiana (Station: West Lafayette 6 NW) 2-10Table 2-6: Average and Maximum Wind Data Measured at Purdue University Airport for 1977-2006 ......................................................................................... 2-12Table 2-7: Tornados Reported in Tippecanoe County, Indiana between 01/01/1950 and02/28/2008 .................................................................................. 2-13Table 4-1: Summary of Design Parameters for PUR-1 ........................................... 4-2Table 4-2: Summary of key reactor parameters for PUR-1 ....................................... 4-4Table 4-3: Characteristics of the PUR-1 Fuel Plates.............................................. 4-6Table 4-4: Channel Types and Thickness in PUR-1 Assemblies ................................ 4-9Table 4-5: Summary of control rod characteristics as listed in the PUR-1 Operations Manual.. 4-10Table 4-6: Representation of fuel plates........................................................... 4-14Table 4-7: Bundle Powers Predicted by f7 and f6 Tallies in MCNP ............................ 4-19Table 4-8: Plate Power (W) Computed from Heating Tallies in Bundles 4-4, 3-3 and 3-4 in PUR-1 Core with 190 Fuel Plates................................................................ 4-21Table 4-9: Axial and Radial Heating Profile for Plate 1348 of Bundle 4-4 for Banked CriticalConfiguration................................................................................ 4-23Table 4-10: Axial and Radial Heating Profile for Plate 1228 of Bundle 3-4 for Banked CriticalConfiguration................................................................................ 4-24Table 4-11: Axial and Radial Heating Profile for LEU Plate 1315 of Bundle 4-3 for BankedCritical Configuration ....................................................................... 4-25Table 4-12: Comparison of measured and calculated control rod worths...................... 4-27PUR-1 SARXPuR-1SAR xRev 2, July 23, 2015 Table 4-13: Comparison of calculated and measured maximum reactivity insertion rates. .... 4-29Table 4-14: Comparison of calculated and measured shutdown margins ...................... 4-30Table 4-15: Other core physics parameters...................................................... 4-33Table 4-16: Water~and Fuel Coefficients for the PUR-1 Core................................... 4-34Table 4-17: Channel Types and Thickness in PUR-1 Assemblies.............................. 4-39Table 4-18: Model Dimensions for the Thermal-Hydraulic Models ............................. 4-41Table 4-19: Hot Channel Factors for the Plate 1348 NATCON Analysis....................... 4-42Table 4-20: ONB Powers for the high power plates.............................................. 4-44Table 4-21: Operating Conditions for PUR-1 as Determined by NATCON for Limiting Plate1348 ......................................................................................... 4-44Table 4-22: Key Power Levels for Reactor Operation and LSSS for PUR-1................... 4-45Table 5-1: PUR-1 water process system summary................................................ 5-1Table 7-1: Summary of PUR-1 Instrumentation ................................................... 7-1Table 11-1: Estimated Neutron Flux .............................................................. 11-4Table 11-2: Ar-41 Saturation Activity.............................................................. 11-4Table 11-3: PUR-1 Personnel Exposures for 2003-2007........................................ 11-9Table 13-1 : Peak power and clad temperature for trip and no-trip insertions of 0.6% Ak/k.. 13-2Table 13-2: Values and Results for Radioiodine Production in PUR-1 Plate 1348............. 13-6Table 13-3: Radioiodines released from Plate 1348 into pool water ........................... 13-7Table 13-4: Number of Moles and Activities of Radioiodine in Reactor Room Air After Releasefrom the Failed Plate ....................................................................... 13-9Table 13-5: Thyroid Dose Rates ................................................................. 13-10Table 13-6: Integrated thyroid dose estimates for several exposure periods following release ofPlate 1348 radioiodines into the reactor pool ........................................... 13-11Table 13-7: Calculation results for gaseous fission products released from the failed fuel plate............................................................................................ 13-13Table 13-8: Associated photon information for the gaseous fission products................ 13-13Table 13-9: Integral Whole-Body Gamma Doses Inside the Reactor Room Assuming an InfiniteCloud and a Leakage Fraction of 0.005 Hr-' (Exhaust Fan Off) ....................... 13-16PUR-1 SARxiPUR-1 AR xiRev 2, July 23, 2015 Table 13-10: Integral Whole-Body Gamma Doses Inside the Reactor Room Assuming anInfinite Cloud and a Leakage Fraction of 1.87 Hr-1 (Exhaust Fan On) ................ 13-17Table 13-11: Integral Whole-Body Gamma Doses Inside the Reactor Room Assuming an FiniteCloud and a Leakage Fraction of 0.005 Hr-1 (Exhaust Fan Off) ....................... 13-18Table 13-12: Integral Whole-Body Gamma Doses Inside the Reactor Room Assuming an FiniteCloud and a Leakage Fraction of 1.87 Hr-1 (Exhaust Fan On)......................... 13-19Table 13-13: Integral Whole-Body Gamma Doses From Submersion Outside of the RestrictedArea Assuming a Leakage Fraction of 1.87 Hr1(Exhaust Fan On) ................... 13-22Table 13-14: Integral Whole-Body Gamma Doses From Direct (From the Building) DoseAssuming a Leakage Fraction of 0.005 Hr-I (Purge Fan Off)......................... 13-23Table 13-15: Peak power and clad temperature for insertions of 0.6% Ak/k with scram....13-24Table 13-16: Peak power and clad temperature for trip and no-trip insertions of 0.6% Ak/k... 13-27PUR-1 SAR xii Rev 2, July 23, 2015PUR-1 SARxiiRev 2, July 23, 2015 1 THE FACILITY1.1 IntroductionThis report is submitted in support of the application for renewal of the operating license (R-87)for Purdue University Reactor (PUR-I) for a period of 20 years, and a power increase from 1 kWto 12 kWThe reactor is located in the Nuclear Laboratories in the Duncan Annex of the ElectricalEngineering Building on the eastern edge of the campus in West Lafayette, Indiana. TheDuncan Annex is of brick and concrete block construction and was originally built as a highvoltage laboratory. In 1962 the reactor was built in half of the existing high voltage laboratory,which was a high bay area. Offices, classrooms, and laboratories had been built in theremainder of the original building.1.2 Summary and Conclusions on Principal Safety ConsiderationsThe original design power level for PUR-1 was 10 kW, and the reactor has operated safelysince its construction in 1962. The analyses presented in this SAR support the continuedoperation of PUR-1, and also support the case for a power uprate to 12 kW. Even in the unlikelycase of a failure of the reactor protective system, the reactor is self-protecting, with a calculatedmaximum power level of 2.38 MW, and a maximum clad temperature of 13300, which is still wellbelow the safety limit of 53000.1.3 General DescriptionThe PUR-l is a 10 kW design, pool type reactor, previously licensed for operation in 1962, 1968and 1988 at 1 kW, utilizing MTR type enriched fuel plates, which are graphite reflected, and lightwater moderated and cooled. It was designed and built by Lockheed Nuclear Products ofLockheed Aircraft Corp., Marietta, Georgia.The reactor is controlled by three blade-type control rods located in the core region of thereactor. There are two shim-safety rods made of solid borated stainless steel, utilizing amagnetic clutch between the blades and the lead screw operated drive mechanisms, and aregulating rod which is a screw operated direct drive and made of hollow stainless steel. Eachcontrol blade is protected by an aluminum guide plate on each side within the fuel assembly.Fuel movement is only by a fuel handling tool, which is stored securely when not in use.Security of the fuel handling tool is under administrative control of the licensed senior operators.1.4 Shared Facilities and EquipmentThe reactor facility is located within the former high voltage laboratory (Duncan Annex) in theElectrical Engineering Building. This space was converted prior to the construction of PUR-1 toalso house classrooms and laboratories.PUR-1 SAR 1-1 Rev 2, July 23, 2015PUR-1 SAR1-1Rev 2, July 23, 2015 1.5 Comparison with Similar FacilitiesSimilar research reactors are in use at the University of Missouri at Rolla, and the Ohio StateUniversity. The safe operating histories of these reactors, and PUR-1 demonstrate the reliabilityand safety of these systems. Both Missouri-Rolla and Ohio State are licensed for operation atmuch higher powers (200 kW for Rolla, and 600 kW for Ohio State), with similar reactorsystems. The safe operation of these reactors at their respective higher powers also supportsthe case for an uprate for PUR-1.1.6 Summary of OperationsThe PUR-1 reactor has been in operation since 1962. It is used for teaching and research tosupport the mission of Purdue University Nuclear Engineering, and the university as a whole.The reactor operates about 90 times per year on average, and typically has three licensedsenior operators. There have been periods, however, when there has only been one operatoron staff.1.7 Facility Modifications and HistoryTable 1-1 summarizes the facility modifications and history.PUR-1 SAR1-2PUR- SAR1-2Rev 2, July 23, 2015 Table 1-1: Summary of amendments and changes to the PUR-1 reactor facility.May 1964 Amendment 1 Permit 10 kW OperationDecember 1965 Installation of pooJ traversing mechanism______________completedJuly 1966 Amendment 2 License RenewalOctober 1968 Change 1 Installation of stainless steel linerSeptember 1969 Installation of air conditionerJanuary 1972 Change 2 Change of pH of pool waterFebruary 1974 Change 3 Regeneration of demineralizer procedure changeNovember 1978 Amendment 3 Technical SpecificationsAugust 1980 Amendment 4 Physical security planFebruary 1981 Installation of catwalk around air conditionerMarch 1981 Amendment 5 Physical security planSeptember 1982 Amendment 6 Technical specifications--revision 1 :Surveillance intervals, scram initiationOctober 1982 Amendment 7 Technical specifications--revision 2: Minortypographical modificationsApril 1983 Amendment 8 Technical specifications--revision 3:Surveillance intervals, RCO qualificationsAugust 1988 Amendment 9 License RenewalFebruary 2000 Amendment 10 Technical specifications--revision 4: COROmembersJune 2007 Amendment 11 Possession limit increaseAugust 2007 Amendment 12 HEU to LEU conversion orderPUR-1 SAR1-3PUR- SAR1-3Rev 2, July 23, 2015 2 SITE CHARACTERISTICS2.1Geography and Demography2.1.1 Site Location and DescriptionThe PUR-I reactor is located in the Duncan Annex of the Electrical Engineering Building on thecampus of Purdue University in West Lafayette, Tippecanoe County, in the State of Indiana, asshown in Figures 2.1, 2.2, and 2.3. The Lafayette-West Lafayette area is about 60 milesnorthwest of Indianapolis, the State Capitol, and about 140 miles south-southeast of Chicago,Illinois.2.1.2 Population DistributionAccording to the 2000 census summary for Tippecanoe County, the total population was148,955. The estimate of the county population for 2006 was 156,1691. Figure 2-4 shows thepopulation within 1, 2, 4 and 8 km from the reactor location (summarized in Table 2-1) andFigure 2-5 shows the projected population in 2030. It should be noted that the populationprojections, given to Purdue by the Tippecanoe Area Planning Commission, use a differentsystem for calculation, and are not as accurate as the 2000 population report.Table 2-1: Population Data for Reactor Vicinity, centered on reactor location.Projected 2030Circle radius Population (2000) Population1 km 17,156 18,3252 km 31,352 31,9924 km 60,828 68,6286 km 93,761 117,2718 km 117,285 165,374PUR-1 SAR 2-1 Rev 2, July23, 2015PUR-1 SAR2-1Rev 2, July 23, 2015

? -....i ,.,44 44442I4'---"44 ~4,4-'4". ~ .44, ~ * .4;-~, 4~;V-.,4 4'~4h '444. 44 '.~ 4~4~444 ~ ~ --4 "'4~ -~ N 4 .-j N4 ~4 44~4* 14" ~ 4-4~4-~ -4 44 4, 44~ )ii 4.'4.- ,~4 '444 bA~*-4O4. 4-~~-~-:~4 -~-44'~ I'4 r'4,2-~ f'4~ *4.'444 S ~ ~4-.4Figure 2 1 State of Indiana showing location of Tippecanoe County.2PUR-1 SAR2-2PUR- SAR2-2Rev 2, July 23, 2015 I[NI.11 .3Figure 2-2: Map of West Lafayette, Indiana, showing inset picture of the location of theElectrical Engineering Building on the Purdue Campus.2PUR-1 SAR 2-3 Rev 2, July 23,2015PUR-1 SAR2-3Rev 2, July 23, 2015 Figure 2-3: Map of the Duncan Annex of the Electrical Engineering BuildingPUR-1 SAR2-4Rev 2, July 23, 2015

-o_-n.Population SelectionS based on 2000 CensusBuffer DistanceiPopulation ........No 0 465 Northwestern Ave.-oD EJ km- Population 17156--72 km -Population 31352L4 km -Population 61828~6 km -Population 937610) l 8 km -Population117285r-roads0.Co The Area Plan Commissionc., Date: JLiy, 2008FOODat.Soxce U S Cean.nBureau2000Cams ,-01 "1CPoultin1elcton/o based on Traffic ZonesBuffer Distance N p./ !-0o 465 Northwestern Ave. F ..42km8 km-=" ..... ...Traffic Area Zonesi......CD Population .....)" BuThe Arear 2003 Com isio 203"0 Cha.n,0,3 o ipcneCut ,.--C ae uy 08", iww ,mMm5.(01 2.221 Nearby Industrial, Transportation and Military FacilitiesThe Purdue University enrollment for 2007-08 was 39,102 full and part-time students, andPurdue University employs approximately 15,304 faculty and staff members (2007-08). So, anapproximate total campus population is approximately 54,400 at peak times. The data for thelast ten years is detailed in Table 2-2.Table 2-2: Purdue University campus population detail for 1998-2008.498-99 99-00 00-01 01-02 02-03 03-04 04-05 05-06 06-07 07-08Full-Time Students 32,788 33,725 33,907 34,442 34,563 34,867 34,745 34,968 35,497 35,549Part-Time Students 4,090 4,037 3,964 3,766 4,001 3,980 3,908 3,744 3,731 3,553Total Students 36,878 37,762 37,871 38,208 38,564 38,847 38,653 38,712 39,228 39,102Total Faculty/Staff 12,888 13,144 13,411 13,831 14,052 14,329 14,636 14,966 15,217 15,304Total CampusPopulation 49,766 50,906] 51,282 52,039 52,616 53,176 53,289 53,678 54,445 54,406Purdue University owns an airport (LAF) at the southwest edge of the West Lafayette Campus,as shown in Figure 2-6 below. It has two runways, the longest of which is Runway 10/28, whichis 6600'x1 50', and the other is Runway 5/23, which is 4230'x1 00'. The Purdue airport averagesabout 130,000 aircraft operations annually (115,000 for the calendar year 2007), and it is thesecond busiest airport in Indiana.The university has a thriving Aviation Technology program, the flight instruction department ofwhich constitutes the majority of the air operations at the Purdue Airport. During the fall andspring semester, the airport is the busiest, with an average of 750 takeoffs and landings perday. The maximum daily takeoffs and landings is 1000, and the minimum is zero (0), onChristmas day. During the summer, the airport sees and average of 30 flight operations perday.5The students primarily use Runway 5/23 (outlined in red on the figure), the nearest end of whichis 6344 feet (1 93 kin) to the southwest (2280) from the Duncan Annex of EE. Only light aircraftcan use this runway, due to its length. The approach to this runway passes near the EEbuilding. Larger aircraft must use the longer runway (10/28), which does not direct aircraft eitheron approach or takeoff near the EE building.Due to the size restrictions of aircraft that use the 5/23 runway, and the heavy concrete block,concrete and brick, windowless construction of the Duncan Annex, the location of the reactorbelow ground level, and the fact that no aircraft have crashed on this section of campus in thehistory of the program, damage due to a crashed light aircraft does not pose a significant threatto the safe operation of the reactor. Should an aircraft impact the building, the location of thereactor 13 feet below the water surface (and 9 feet below floor level) in a tank of water willmitigate any potential immediate damage, and the close proximity of the Purdue FireDepartment will ensure that the public safety is maintained in the event of extensive fire damageor explosions.PuR-1 SAR 2-7 Rev 2, July 23, 2015PUR-1 SAR2-7Rev 2, July 23, 2015 Figure 2-6: Map showing the location of the Purdue University airport in relation to the locationof PUR-1.22.3 Climatology and Meteorology2.3.1 General and Local ClimateThe climate of the county is continental with hot summers and cold winters. The seasons arestrongly marked, and the weather is frequently changeable. Climatological data available fromNOAA for West Lafayette are summarized inTable 2-3, Table 2-4, and Table 2-5. The tables show the conditions are measured at the WestLafayette station (West Lafayette 6NW), where the latitude is 40°28', the longitude is 87°00',and the ground elevation is 705 feet.PUR-1 SAR 2-8 Rev 2, July 23, 2015PUR-1 SAR2-8Rev 2, July 23,2015 Table 2-3: Climatography Data for West Lafayette, Indiana; Mean and Extreme Temperatures6.Mean* (°F) Extremes (°F)Highest LowestDaily Daily Highest Month Lowest MonthMonth Max Min Mean Dailyt Year Day Mean Year Dailyt Year Day Mean* YearJan 31.5 15.0 23.3 68 1906 20 35.9 1990 -24k 1985 20 7.7 1977Feb 36.8 19.0 27.9 73 2000 26 38.9 1998 -23 1963 26 12.4 1978Mar 48.4 29.1 38.8 87 1910 24 46.5 1973 -12 1960 1 29.6 1984Apr 60.9 39.2 50.1 90 1930 11 55.7 1985 7 1982 7 44.6 1982May 72.5 50.3 61.4 96* 1911 27 68.9 1977 26÷ 1966 10 56 1997Jun 81.4 59.6 70.5 104 1934 1 75.0 1991 35 1992 22 65.8 1972Jul 84.5 63.0 73.8 111 :1936 14 77.9 1983 42+ 1972 5 70 1996Au 82.5 60.6 71.6 103 1918 5 77.7 1995 35 1965 29 66.7 1992Sep 77.0 52.9 65.0 100 !1933 9 69.4 1978 25 1995 23 59.4 1993Oct 64.8 41.6 53.2 90+ 1922 2 61.3 1971 18 1925 30 47.3 1988Nov 50.0 32.2 41.1 78÷ 1930 19 46.8 1999 -3 1930 28 32.9 1976Dec 37.0 21.1 29.1 71÷ 1982 3 38.6 1982 -22 1989 22 16.5 2000Jul Jul Jan JanAnnual 60.6 40.3 50.5 111 1936 14 77.9 1983 -24* 1985 20 7.7 1977:* Derived from 1971-2000 serially complete datat Derived from station's available digital record: 1901-2001i+ Also occurred on an earlier date(s)PUR-1 SAR2-9PUR- SAR2-9Rev 2, July 23, 2015 Table 2-4: Climatography Data for West Lafayette, Indiana; Mean Days InformationMean Number of Days*DaysMonth Max __ Days Days Days Days Days100°F Max __ 90°F Max _> 50°F Max < 32°F Min _ 32°F Min <_ 0°FJan 0.0 0.0 2.4 15.9 28.7 5.6Feb 0.0 0.0 4.6 10.7 24.5 3.5Mar 0.0 0.0 13.1 3.3 20.5 0.1Apr 0.0 0.0 24.3 0.2 7.8 0.0May 0.0 0.4 30.7 0.0 0.6 0.0Jun @4 30 0.0 0.0 0.0Jul 0.1 5.7 31 0.0 0.0 0.0Aug 0.0 3.2 31 0.0 0.0 0.0Sep 0.0 1.7 30 0.0 0.3 0.0Oct 0.0 @ 28.6 0.0 5.5 0.0Nov 0.0 0.0 14.5 1.7 16.6 0.0Dec 0.0 0.0 4.4 10 26.7 2.4Annual 0.1 15 244.6 41.8 131.2 11.6* Derived from 1971-2000 serially complete data@ Denotes mean number of days greater than 0 but less than 0.05The average annual temperature is about 50°F. The mean temperature in January, the coldestmonth, is 23°F, and in July, the warmest month, is 73.8°F. About nine days per year thetemperature falls below zero, and about 137 days per year the temperature goes below freezing(32°F).Table 2-5: Precipitation Normals for West Lafayette, Indiana (Station: West Lafayette 6 NW)PRECIPITATION NORMALS (Total in Inches)JAN FEB MAR APR MAY JUN JUL fAUG SEP OCT NOV DEC ANNUAL1.79 1.57 2.84 3.57 4.35 4.24 4.00 3.68 2.98 2.73 3.08 2.43 37.26PUR-1 SAR2-10Rev 2, July 23, 2015 The average annual precipitation is 35.68 inches. July is the wettest month with 4.74 inches,and February is the driest month with 1.41 inches of precipitation. Prevailing winds are from thewest or southwest during the winter and from the south during the summer. Wind velocity ishighest in February and lowest in August.12.3.2 WeatherWind conditions as measured at the Purdue University Airport (West Lafayette 6NW) aresummarized over the period 1977 to 2006, and are detailed inTable 2-6. These data indicate an annual mean wind speed of 8.78 miles per hour and amaximum wind speed of 72 miles per hour. According to the Unified Building Code, 1985edition, the Purdue University lies in the maximum wind zone of 80 miles per hour, whichtranslates to a wind load of 17 pounds per square foot. Buildings at Purdue University aredesigned to withstand this wind load.PUR-1 SAR 2-11 Rev 2, July 23, 2015PUR-1 SAR2-11Rev 2, July 23, 2015 Table 2-6: Average and Maximum Wind Data Measured at Purdue University Airport for 1977-2006zMonthJanuaryFebruaryMarchAprilMayJuneJulyAugustSeptemberOctoberNovemberDecemberAnnualAvg. Wind"Direction*(degrees)216.91205.39198.20195.53193.11191.52200.37191.07196.03198.89206.99215.05200.75Avg. WindSpeed (mph)10.279.8410.5110.368.737.556.906.447.098.359.589.718.78Max. WindSpeedt (mph)646262637262576057636352722.3.3 Severe WeatherThis region of the United States is subjected to tornado activity, primarily during the late springand early summer months. Table 2.4 shows the tornados occurring in Tippecanoe County forthe period from 1950 through February 2008. Thirty-eight tornados occurred over this period,Denotes the compass direction the wind is blowing from (0° North).t Max. Wind Speed is not the average maximum. It is the actual maximum that occurred during the 30years of data.PUR-1 SAR 2-12 Rev 2, July 23, 2015 which averages to less than one per year. The probability of damage due to a tornado isminimal.Table 2-7: Tornados Reported in Tippecanoe County, Indiana between 01/01/1 950 and02/28/200881234567891011121314151617181920LocationCOUNTYTIPPECANOETIPPECANOETIPPECANOETIPPECANOETI PPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETIPPECANOETI PPECANO0ETI PPECANO0ETIPPECANOEorDate Time6/1 3/1 953 21:004/3/1 956 17:003/6/1961 6:054/22/1 963 20:156/1 0/1 963 12:004/11/1 965 18:079/1 4/1 965 20:156/24/1967 12:305/1 5/1 968 21:513/19/1971 2:035/29/1973 14:206/12/1973 9:456/12/1973 10:294/1/1 974 16:323/12/1976 14:253/20/1976 15:204/10/1978 12:154/23/1 978 18:006/25/1 978 15:556/25/1 978 17:00FujitaScaleF1F2F1F2F1F4F2F2F2F2F0F1F1F2F1F4F2F1F0F3Deaths00000000000000000000Injuries000001000000000060000PropertyDamage ($)OK25KOK3K3KOK250K3K3K25KOK25KOK25KOK2.5M25K250KOKOKPUR-1 SAR 2-13 Rev 2, July 23, 2015PUR-1 SAR2-13Rev 2, July 23, 2015 212223242526272829303132333435363738TIPPECANOETIPPECANOETIPPECANOETIPPECANOELafayetteLafayetteWest LafayetteBattle GroundLafayetteWest LafayetteLafayetteRomneyDaytonDaytonRomneyCairoAmericusOdellTOTALS:7/2/1 978 12:156/7/1980 16:156/24/1981 19:473/27/1991 18:104/26/1 994 23:581/1 8/1 996 14:307/4/1998 1:309/28/1 999 18:546/11/2003 18:476/11/2003 19:007/21/2003 4:005/30/2004 19:265/30/2004 19:297/26/2005 20:004/2/2006 20:254/14/2006 18:584/14/2006 19:186/25/2006 14:10F1F2F1E0F4EQF1ElEQF0E0EQF2F0F1E0ElE000007000100000000008725KOK250KOK5.0M0200K300K00001.0M10K50K030K3K10.O004M2.4 HydrologyMost of Tippecanoe County is covered by glacial drift. The drift ranges in thickness from a thinveneer to about 425 feet and was deposited upon a bedrock surface that was eroded by apreglacial drainage system. Much of the surface drift consists of glacial till. Water-laid crossbedded sand and gravel are associated with the till. The subsurface glacial deposits alsoinclude much till with interbedded sand and gravel. Locally, clay deposits are as much as 106feet thick. Within the drift, five sheetlike water bearing units are differentiated in parts of thePuR-1 SAR 2-14 Rev 2, July23, 2015PUR-1 SAR2-14Rev 2, July 23, 2015 county. Ground water within these units occurs under artesian and water-table conditions.Locally these may occur within the same unit.9This area was repeatedly glaciated during the Pleistocene epoch. Before glacial times, a giantdrainage way, now know as the Teays River, flowed from the Appalachian Mountains acrossOhio, and passed northwestward through the present site of Lafayette-WestLafayette.1° Illinoian ice dammed the preglacial Teays River channel and ponded the relativesmall Glacial Lake Lafayette. An outlet channel, developed to drain this proglacial lake, wassubsequently perpetuated as the present Wabash River drainage line southwestward from theLafayette-West Lafayette area.1The elevation of the Purdue University campus is approximately 706 feet and the level of theWabash River is approximately 510 feet. With this difference of over 100 feet the flow of bothsurface water and ground water is in a generally easterly and southerly direction toward theWash River, which flows around two sides of the campus.Any leakage of contaminated water from the PUR-I represents no potential hazard to either theWest Lafayette or Purdue University water supply, since these flows are away from the wellfields of both. The Wabash River represents a natural barrier between the reactor and theLafayette well fields, so no potential hazard exists there.2.5 Geology and Seismology2.5.1 Regional GeologyThe county lies within the Tipton Till Plain of Indiana and is a section of the Till Plainssubprovince of the U. S. central Lowlands physiographic province; Most of the soils in this areaare derived from the glacially deposited material. Extensive upland areas are covered with a thinmantle of loose deposits. A few areas are covered with soils of alluvial, colluvial or organicorigin. Glacial drift covers the bedrock to a depth ranging from a few feet to more than 300 feet.The underlying bedrock consisting of flint, shale, sandstone, and limestone of the Mississippianperiod, is exposed as rock terraces in the Wabash Valley and on the upland in the western partof the county. Purdue University is located above an extensive glacial deposit of sand andgravel.The land surfaces of Tippecanoe County are flat to rolling, except where the major streamshave cut deeply into the surface. The entire county lies within the drainage basin of the WabashRiver and its tributaries. The land slopes generally southwestward with the streams flowingwestward. Two main tributaries, the Tippecanoe River and the Wild Car Creek enter theWabash upstream from the campus. Minor tributaries include Little Pine Creek, Indian Creek,Burnetts Creek, Mott's Creek, Sugar Creek, Buck Creek, Wea Creek, and Flint Creek.2.5.2 SeismologyThe three most significant seismic source zones which are closest to West Lafayette are:1. The New Madrid area of southeastern Missouri;2. The Wabash Valley Fault system of southwestern Indiana and southeastern Illinois;3. The Anna, Ohio area.PUR-1 SAR2-15Rev 2, July 23, 2015 Reasonable estimates of the maximum magnitude events which could occur in those areas givevalues of 7.4, 6.6 and 6.3 (body wave motion) for the seismic zones, respectively. Based on thedistance from these zones (400, 200 and 200 km respectively) and attenuation curves,estimates for peak horizontal acceleration at West Lafayette for maximum magnitude eventswhich could occur at these three seismic zones are approximately 5-15% G.12 The figures thatfollow (Figure 2-7 through Figure 2-11) show the earthquake probabilities for the area.The way, in which the reactor facility was constructed by modifying an existing building with noreinforcing bars tied into the original structure, the reactor pool can be considered a freestanding unit in the event of any seismic activity. The reactor pool consists of steel cylinderscontaining compacted magnetite sand between the cylinders and the 1/3 inch carbon steel tank.The inside of this tank was later lined with 1/16 inch stainless steel. With these barriers tocontain the reactor pool water and considering the reactor pool as a free standing unit it is highlyunlikely that any reactor water would be lost during any severe seismic activity.PuR-1 SAR 2-16 Rev 2, July 23, 2015PUR-1 SAR2-16Rev 2, July 23, 2015 SA 1ls 2%50yr PE, 200835"N30"N-)25"N100 W95"W 90"W 8"W 8"Figure 2-7:Map of the 1-Hz spectral acceleration for 2% probability of exceedance in 50 yearsfor the Central and Eastern United States in standard gravity (g).13PUR-1 SAR 2-17 Rev 2, July 23, 2015PUR-1 SAR2-17Rev 2, July 23, 2015 SA 0.2-s 2%/d50year PE, 200835N25"tboow95'w go.w 85w eo0wFigure 2-8:Map of the 5-hertz spectral acceleration (SA) for 2% probability of exceedance in 50years in the Central and Eastern United States in standard gravity (g).PUR-1 SAR 2-18 Rev 2, July 23, 2015PUR-1 SAR2-18Rev 2, July 23, 2015 PGA with 20/1o50 yr PE, 2008Figure 2-9: Map of peak ground acceleration (PGA) for 2% probability of exceedance in 50years in the Central and Eastern united States in standard gravity (g).PUR-1 SAR2-19PuR- SAR2-19Rev 2, July 23, 2015 CEUS 1 s SA 1 0%/50yr 2008O0 95"W 90"W 5W 0"W1 Aft 111712 2MB BAlm bttim InucA hI~mnitIefiFigure 2-10:Map of 1-hertz spectral acceleration (SA) for 10% exceedance in 50 years in theCentral and Eastern United States in standard gravity (g).PUR-1 SAR 2-20 Rev 2, July 23, 2015PUR-1 SAR2-20Rev 2, July 23, 2015 CEUS PGA 1 0%/50 years, 2008GO0" SS'9W 9OW 0 W 80"WFigure 2-1 1: Map of peak ground acceleration (PGA) for 10% probability in 50 years in theCentral and Eastern United States in standard gravity (g).FUR-i SAR 2-21 Rev 2, July 23, 2015PUR-1 SAR2-21Rev 2, July 23, 2015 2.6 References1 U.S. Census Bureau, http://www.census.gov2 United States Geological Survey, http://www.usgs.gov (1973).3 Tippecanoe County Area Planning Commission, (2008)4~ Purdue Data Digest, http://www.purdue.edu/datadigest/5 Private Communications with Betty Stansbury, Airport Administrator, Purdue University,(2008).6 National Climatic Data Center, NOAA-NESDIS,http://www.ncdc.noaa.gov/oa/climate/normals/usnormals. html, (2004)z Indiana State Climate Office, Purdue University, http://www.iclimate.org, (2008).8 National Climatic Data Center, NOAA-NESDIS, http://www.ncdc.noaa.gov/oa/ncdc.html,(2008)SResenshein, Joseph S., "Ground-Water Resources of Tippecanoe County, Indiana', State ofIndiana, Indiana Department of Conservation, Division of Water Resources Bulletin No. 8,1958.10 Ulrich, H.P., Barnes, T.E., and Krantz, B.A., "Soil Survery, Tippecanoe County, IndianaSeries 1940, No.22', 1959.11 Maarouf, Abdelraham M., and Melhorn, Wilton N., "Technical Report No. 61', PurdueUniversity Water Resources Research Center, June, 1975.12 Braile, L.W. Professor, Purdue University, private communications.13 Petersen, Mark D., et al., "Documentation for the 2008 Update of the United States NationalSeismic Hazard Maps", Open File Report 2008-1128, U.S. Geological Survey, (2008).PUR-1 SAR2-22PUR- SAR2-22Rev 2, July 23, 2015 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS3.1 Design CriteriaThe Duncan Annex of the Electrical Engineering Building is of brick, concrete block andreinforced concrete construction which was originally designed as a large high voltagelaboratory. It was subsequently subdivided into offices, classrooms and laboratories. Thereactor is located in the southwest corner on the ground floor in a high bay area of the building.The top of the reactor pool is approximately 10 feet below ground level.The outside air supply and exhaust are both passed through HEPA filters. The reactor room ismaintained at negative air pressure (minimum 0.05 inches of water). All doors to the reactorroom have foam rubber seals.The only floor drain to the sewers is sealed except for a vent opening. This vent is raised abouttwo feet above the floor and has a filtered inverted opening. Condensate from the air conditioneris released to this drain through an opening 12.0 feet above the floor.3.2 Meteorological DamageAccording to the Unified Building Code, 1985 edition, the Purdue University lies in the maximumwind zone of 80 miles per hour, which translates to a wind load of 17 pounds per square foot.Buildings at Purdue University are designed to withstand this wind load.With this information, and the low incidence of tornado activity in the campus area, tornadodamage to the building is very unlikely. The Nuclear Engineering Laboratory is the shelter-in-place location for tornado warnings. Additionally, due to the fact that that reactor and instrumentconsole are well below ground level, damage to the reactor or controls is also very unlikely.3.3 Water DamageThe Electrical Engineering Building, and all of the Purdue Campus, lie well above any floodplain. In 46 years of operation, there has been no standing water in the reactor room. The topfour feet of the reactor pool stand above floor level, and the sides of the pool are made of 15inches of reinforced concrete. In the unlikely event that this wall were to break, it would onlyresult in 3 inches of standing water on the reactor room floor. None of the reactorinstrumentation would be harmed by such an unlikely occurrence.3.4 Seismic DamageThe way, in which the reactor facility was constructed by modifying an existing building with noreinforcing bars tied into the original structure, the reactor pool can be considered a freestanding unit in the event of any seismic activity. The reactor pool consists of steel cylinderscontaining compacted magnetite sand between the cylinders and the 1/3 inch carbon steel tank.The inside of this tank was later lined with 1/16 inch stainless steel. With these barriers tocontain the reactor pool water and considering the reactor pool as a free standing unit it is highlyunlikely that any reactor water would be lost during any severe seismic activity.PUR-1 SAR3-1PUR- SAR3-1Rev 2, July 23, 2015 3.5 Systems and ComponentsShould the need arise for an emergency shutdown of the reactor, the 2 shim-safety control rodscan be scrammed automatically by instrumentation, or manually via two scram buttons: onelocated within easy reach of the operator on the control panel, the other outside the mainpersonnel access door to the reactor room.During emergency conditions, the ventilation systems can be shut off by a console switch, andthe sealed room will prevent the rapid spread of contamination. During an emergency, the airconditioner and the valve on the drain from the condensate holdup tank are shut off with thesame switch that shuts off the ventilation system. The condensate would then be held until it istested by Radiological Control before it is released to the sewer. If contamination is found, itwould be disposed of as radioactive liquid waste.PuR-1 SAR 3-2 Rev 2, July 23,2015PUR-1 SAR3-2Rev 2, July 23, 2015 4 REACTOR DESCRIPTION4.1Summary DescriptionThe PUR-1 reactor described herein was designed and constructed by Lockheed NuclearProducts of Lockheed Aircraft Corp. of Marietta, Georgia. It was designed for continuoussteady-state operation at 10 kW, but previously licensed for operation at 1 kW.PUR-1 is a heterogeneous, pool-type non-power reactor. The core is cooled by naturalconvection of light water, moderated by light water, and reflected by water and graphite. Thereactor is located near the bottom of a water-filled tank surrounded and supported by a concreteshielding structure as shown in Figure 4-1. An aluminum grid plate structure supports thereactor and control mechanisms at the bottom of the pool, with additional support of the controlmechanisms provided by a fixture at the top of the pool. Three detectors used for monitoringreactor conditions are located in fixed positions next to the reactor core. And the startupdetector is located in a tube affixed to a fuel element in the core, which allows the detector to beremoved from the neutron flux when the reactor is at power.RodStorage---Shim Sofeties-Fission Cham'ber/Stortup Charmer0.5 in Drop Tube-3 in Drop TubeFigure 4-1: PUR-1 Pool LayoutThe reactor core is composed of sixteen fuel elements positioned in holes in the aluminum gridplate. The grid plate contains a rectangular matrix of holes to allow the changing of fuel elementlocations and the insertion of graphite reflector elements to displace reflector water. Each fuelelement consists of several thin metal plates assembled into a unit about 7 cm by 7 cm with anactive fuel length of approximately 60 cm. Fuel elements of this general configuration were firstPUR-1 SAR4-1PUR- SAR4-1Rev 2, July 23, 2015 designed for and used in the Materials Testing Reactor (MTR) and thus are referred to as MTR-type fuel elements. Three of the fuel elements are fabricated without the four middle plates,providing space for the insertion and movement of the reactor control rods.Reactivity of the reactor core is changed by the operator moving the control rods that aresuspended from fail-safe electromagnets. The ionization chambers used for sensing neutronand gamma-ray fluxes are located near the core. The control console, from which the operatorcan observe the reactor pool and top structures, is located adjacent to the reactor, and consistsof typical read-out and control instrumentation.Heat removal is achieved by natural convection, with a general flow up through the nozzle at thebottom of the fuel assemblies. The reactor is located within a 6400 gallon cylindrical water tank,17 feet deep and 8 feet in diameter.Safety and other operational characteristics of this reactor system are similar to other reactorsusing the MTR type fuel assembly. The power and flux level of the PUR-I are of adequate rangeand the experimental facilities are sufficiently flexible to encompass a wide variety of trainingand research experiments. The reactor is designed so that a minimum of restrictions areimposed on the experimenter, and the console can be readily operated by one person.Safety is an overriding requirement in a training reactor. Self-limiting features of the PUR-1 core,coupled with carefully designed control instrumentation, assure the highest degree of safety.The safety record of this facility, demonstrated over the past 46 years give proof that the design,construction, and installation of the reactor system, coupled with the administrative control overoperation, maintenance, and utilization, are more than adequate to provide protection for thepublic health and safely.Table 4-1 and Table 4-2 summarize the key design parameters for the PUR-1.Table 4-1: Summary of Design Parameters for PUR-1DESIGN DATA ValueDesign Power Level 10 kWFuel Type MTR PlateFuel "Meat" Composition U3Si2-AIFuel Enrichment U-235 (nominal) 19.75%Mass of U-235 per plate (g, nominal) 12.5PuRl SAR 4-2 Rev 2, July 23, 2015PUR-1 SAR4-2Rev 2, July 23, 2015 Fuel Meat DimensionsWidth (mm) 59.6Thickness (mm) 0.508Height (mm) 600.1Fuel Plate DimensionsWidth (mm) 70.2Thickness (mm) 1.27Height (mm) 638.6Cladding Composition 6061 AlCladding Thickness (mm) 0.381Dummy Plate Composition 6061 AlSame asDummy Plate Dimensions FuelStandard Fuel AssembliesNumber of standard assemblies 1Number of plates per standard 1assembly 1Control Fuel AssembliesNumber of control assemblies3Number of plates per control 8assembly8Total plates in core (fuel and dummy) 206Fuel plates in core (current) 190Dummy plates in core (current, expected) 16Plate spacing in standard assemblies (mm) 3.66Plate spacing in control assemblies (mm) 4.60PUR-1 SAR4-3PUR- SAR4-3Rev 2, July 23, 2015 Table 4-2: Summary of key reactor parameters for PUR-I.REACTOR PARAMETERS CalculatedFresh core excess reactivity (%Aldk) 0.421Shutdown margin (%Ak/k) -1.80Control rod worth (%Ak~k)Shim-safety 1 3.93Shim-safety 2 2.22Regulating Rod 0.27Maximum reactivity insertion rateCk) 1 .75E-02Shim-safety 1 8.75E-03Shim-safety 2 4.66E-03Regulating RodAvg. coolant void coefficientf °/°id .'2 -1.93E-1+7%Coolant temperature coefficient(°/° Alj/ -9"05E-3+9°/Fuel temperature coefficient .k--C) -8. 05E-4+10%Effective delayed neutron fraction (%) 0.784Neutron lifetime (ps) 81.34.2 Reactor CoreThe PUR-1 core layout is a sixteen assembly (4x4 array), heterogeneous, light-watermoderated, graphite reflected, water cooled reactor fueled with LEU plate-type fuel. The core1 From the 2008 Annual Report.2 aultdorterpeettvragof0.%vo.3Calculated for the representative range of 20-30°.%vod4 Calculated for the representative range of 20-127°C.PuR-1 SAR 4-4 Rev 2, July 23, 2015PUR-1 SAR4-4Rev 2, July 23, 2015 layout is shown in Figure 4-2. Each of the thirteen standard fuel assemblies in the core can holdup to 14 fuel plates, or a mixture of fuel and dummy plates. The three control elements eachhold up to eight fueled plates.Twenty graphite reflector assemblies surround the core, 6 of which contain a cylindricalaluminum tube normally filled with graphite. These 6 elements comprise the irradiation facility.The graphite can be removed from these tubes and replaced with experiment capsules whichcan then be irradiated with normal reactor operation.©rnmmmu~K)'OOEEIIL2JETZE(QOE~_DU\(. lK/"EFacilitySW FuelElementSorce/ \Figure 4-2: PUR-1 Grid PlateThe reactor is controlled by three control rods located in the core region of the reactor. Thereare two shim-safety rods made of solid borated 304 stainless steel, utilizing a magnet clutchbetween the blades and the lead screw operated drive mechanisms, and a regulating rod, whichis a screw operated direct drive and made of hollow stainless steel. Each control blade isprotected by an aluminum guide plate on each side within the control fuel assemblies.Each of the standard assemblies and the control assemblies are contained in a 6061 aluminumcontainer. The standard graphite assemblies and the irradiation facility graphite assemblies arecontained in similar 6061 aluminum containers. The startup neutron source is located outsidethe core in a similar 6061 aluminum container.4.2.1 Reactor FuelThe reactor is fueled by standard MTR LEU plates installed in 2007 during the conversion ofPUR-1 from HEU to LEU fuel. The LEU fuel is silicide dispersion standard MTR platesmanufactured by BWX-Technologies (BWXT) of Virginia. Dummy aluminum plates identical insize to the fuel plates were also manufactured by BWXT, and these are used in place of fuel inthe reactor in some locations. The assembly cans that contain the plates were manufactured byGeneral Atomics, of California.PUR-1 SAR4-5PuR- SAR4-5Rev 2, July 23, 2015 The fuel and dummy plates are inserted into both the standard and control assembly cansindividually, each one contained within its own slot. These slots control the plate spacing toclose tolerances over the length of the plates, which is a significant improvement over the olddesign where the plates were pinned together at the four corners. The nominal fuel plateinformation is detailed in Table 4-3.Table 4-3: Characteristics of the PUR-1 Fuel PlatesDesign DetailFuel Type U3Si2-AIFuel "Meat"Com posit ion U 3S i2-AIEnrichment 19.75%Mass 235U per fuel plate 12.5 gFuel Plate DimensionsWidth (mm) 70.2Thickness (mam) 1.27Height (mm) 638.6Fuel Meat DimensionsWidth (mm) 59.62Thickness (mm) 0.508Height (mam) 600.13Cladding Type 6061 AlCladding:Along width (mm) 3.63 (min)Along thickness (mm) 0.381Figure 4-3 and Figure 4-4 show the design of the wall spacers that control the plate locationsand channel thicknesses for the standard and control elements.PUR-1 SAR4-6PUR- SAR4-6Rev 2, July 23, 2015 II JL '-I.L .I4-122222112222211222211222221122221C2221C2Z3122221C2221,CZZlC2221c'--nIIIILX.OO X 4-5LP I1~UM263628184X-2,642-2.U4'I..6-1,.4-logoI.314-.oooII*i*9<<<<<<<<JA3s°T'yP \,<P0.186 .m£701Figure 4-3: Standard assembly can detail, showing wall spacers.PUR-1 SAR 4-7 Rev 2, July23, 2015PUR-1 SAR4-7Rev 2, July 23, 2015 I II II II %lC--'I CSS3cc_ -ILL_~lh{.171 1W 11112X .063X45- C-] C-//L-'-1.949 <L l l I-1 757-T2.61CT-<____ I__* I

II I4 -,.~I---x0 119114 .25 2.0 -.1Figure 4-4: Control assembly can detail, showing wall spacers.Figure 4-5 and Figure 4-6 show the cross-sectional view of the standard and control assembliesfor the PUR-1 reactor. Figure 4-5 shows the handle assembly at the top of the assembly can,which can be removed for insertion and removal of the fuel plates. The Control assemblies haveno such handle.PUR-1 SAR 4-8 Rev 2, July 23, 2015PUR-1 SAR4-8Rev 2, July 23, 2015

-L,,! HiFigure 4-5: Standard fuel assembly.Figure 4-6: Control fuel assembly.The nominal plate-to-plate spacing, and the nominal plate-to-wall spacing is shown in Table 4-4.Table 4-4: Channel Types and Thickness in PUR-1 AssembliesPlate-to-plate (mils) Plate-to-wall (mils)Standard Control Standard ControlDimension 144+15 181+/-+15 127+/-8 127+/-8PUR-1 SAR 4-9 Rev 2, July 23, 2015PUR-1 SAR4-9Rev 2, July 23, 2015 Fuel and dummy plates are uniquely identified by serial numbers, and the dummy plates aredifferentiated by notches machined into the end of the plates, as shown in Figure 4-7 and Figure4-8.2X R.140Figure 4-8: Dummy plates are differentiatedfrom the fuel plates by a notch machined intothe end of the plate.Figure 4-7: Plates are identified by uniqueserial numbers as shown here.4.2.2 Control RodsTable 4-5: Summary of control rod characteristics as listed in the PUR-1 Operations Manual.CONTROL RODSNumber of Regulating Rods 1 -304 stainless steel, hollowNumber of Shim Safety Rods 2 -Boron-stainless steel, solidOperating RatesRegulating Rod 17.7 in/mmnShim Safety Rod 4.4 in/mmnScram Less than 1 second (from signal tocomplete insertion).SizeRegulating rod (inches) 1/2 x 2 1/4/Shim safety rods (inches) 1/2 x 2 1/4/ x 25 1/2/PUR-1 SAR4-10PuR- SAR4-10Rev 2, July 23, 2015 Maximum rate of reactivity changeRegulating rod 1"1x104 A._kkk~sShim safety rods 4.2.2.1. 9.9x10-3 Akk~sAverage rate of reactivity changeRegulating rod 5.8x10-4 AkShim safety rods k. s5.4x10-3 Ak____ ___ ____ ___ ____ ___k~s4.2.3 Neutron Moderator and ReflectorPUR-1 is moderated by the light water of the reactor pool. A graphite reflector surrounds thefour sides of the reactor, and light water reflects the top and bottom. The graphite in the reflectorassemblies is nuclear grade, manufactured by Union Carbide, and contains less than 1 part permillion boron.4.2.4 Neutron Startup SourceA 5 Curie Plutonium-Beryllium neutron startup source is located next to the core, and isremovable when the reactor reaches criticality. A drive mechanism controlled from theoperator's console raises and lowers the source as needed.4.2.5 Core Support StructureThe reactor is supported on an aluminum grid plate typical of all Lockheed MTR reactors. Thisgrid plate controls the placement of fuel within the core, and can be used to locate experimentsthat would be placed outside the core. A drawing of the grid plate is located in the Appendix 1.The grid plate is manufactured with 6061 aluminum, and has not shown any degradation overits greater than 46 years of service. It is expected that the grid plate will continue to function asdesigned.4.3Reactor PoolThe reactor pool is built below floor level except for the three foot wall that serves as a biologicalshield for the operators and experimenters. The pool is contained in a cylindrical tank 17 feet, 4inches deep and 8 feet in diameter. The core is located to one side to give additionalexperimental space. The pool has a welded stainless steel liner.PUR-I SAR4-11PuR- SAR4-11Rev 2, July 23, 2015 The supports for the drive mechanisms for the control rods, the fission chamber and the source,and the neutron detectors are fastened to the support plate at the top of the tank. A traversingmechanism was mounted on the top of the reactor pool wall after the reactor was built. A lightweight, portable aluminum bridge can be placed across the pool for maintenance and fuelhandling operations.The average pool temperature in recent PUR-1 operating history is 260C. The figure belowshows measured pool temperatures from 1993 to 2006.PUR-1 Reactor Pool Temperature Measurements35.030.0-025.0-20.0-A4.4.4.?z e.4.,4.4.4ic4;.4.V.4.4.4.% .'4~ ~ 4.I4. 4. 44.4. 4.4.4.4.4..J#4c4* c.4 4..415.0 ~Dec-93Dec-95Dec-97Dec-99Dec-01Dec-03Dec-05Dec-07DateFigure 4-9: PUR-1 Reactor pool temperature measurements from 1994-2007.4.4Biological ShieldThe biological shield consists of light water, magnetite sand, earth, and structural concrete. Thereactor tank sits in a pit 14 feet deep with only the top 3 feet 4 inches of the tank above floorlevel. The tank sits on a concrete pad and is enclosed below ground level by 2 feet of barytessand between the tank and earth. A monolithic mass of ordinary concrete 15 inches thick isprovided from the floor level to the top of the tank. The water level is normally 4 inches belowthe top of the tank.Shielding over the core is provided by 13 feet of light water. This reduces the radiation level to acalculated less than 1 mrem/hr when the core is operating at 1 kW. The concrete biologicalshield is designed for a maximum radiation level at any point along its outer lateral surface of0.1 mr/hr atl1 kW.PuR-1 SAR 4-12 Rev 2, July23, 2015PUR-1 SAR4-12Rev 2, July 23, 2015 4.5 Nuclear DesignNeutronics analysis was performed with the Monte Carlo N-Particle Code (MCNP5), publishedby Los Alamos National Laboratory. Thermal-Hydraulic analysis was performed with NATCON,a natural convection analysis code written by Argonne National Laboratory (ANL). Transientanalysis was performed using PARET, also written by ANL.4.5.1 MCNP ModelThe MCNP model of the fuel plates along various dimensions is shown in Table 4-6. Thematerials used in the model are standard for the LEU MTR plates, with the addition of 20 partsper million of a boron-equivalent to the 6061 cladding material to account for impurities in thealloy.PUR-1 SAR 4-13 Rev 2, July 23, 2015PUR-1 SAR4-13Rev 2, July 23, 2015 Table 4-6: Representation of fuel plates.Figure 4-1 1: plate in Y-Z plane. (magnified)Figure 4-10: plate inX-Z planeIIFigure 4-12: plate in X-Y plane, cutaway view.PUR-1 SAR 4-14 Rev 2, July 23, 2015PUR-1 SAR4-14Rev 2, July 23, 2015 Figure 4-13: Model representation of standard assembly plate spacing detail showing wallspacers.The representations of the complete standard and shim safety control fuel assemblies for thecore are shown in Figure 4-14 and Figure 4-15 respectively. There are one or two dummy platesin the standard assemblies, and no dummy plates in the control assemblies, which are theplates without the center fuel material in the figures. The assembly cans are identical in size andcomposition.Figure 4-14: Comparison of standard and control assemblies in the model.As stated before, the control assemblies contain no dummy plates, but do contain a guard platemade of 6061 aluminum between the fuel elements and the control rods to prevent anymechanical damage to the fuel from insertion of the control rods. The shim safety control rodsthemselves, made of borated stainless steel, are oblong shaped plates inserted down the centerof the assemblies. The regulating rod is similar to the shim-safety control assembly shown in thefigure above, but reflects the fact that the rod is hollow, and filled with water.FUR-I SAR 4-15 Rev 2, July 23, 2015PUR-1 SAR4-15Rev 2, July 23, 2015 Figure 4-15: Mode! of regulating rod assembly.The entire core model is shown in Figure 4-16. This image shows the present configuration ofthe core, the orientation of the plates, the location of the irradiation facilities, and the startupneutron source on the left side of the core.Figure 4-16: Representation of LEU core load4.5.2 Normal Operating ConditionsThe PUR-1 LEU fuel plates are fabricated with U3Si2 dispersion fuel. Each fuel plate has anominal loading of 12.5 g U-235. A fresh core with 191 fuel plates and 15 "dummy" aluminumplates was evaluated in the models presented herein. Figure 4-16 shows the core layoutPUR-1 SAR4-16PuR- SAR4-16Rev 2, July 23, 2015 modeled in MCNP, and drawings of control and 14-plate fuel assemblies. The numbers inparenthesis in the core layout drawing indicate either the number of fuel plates in the 14-platefuel assemblies (the remainder are dummy plates of aluminum) or the label of the controlassembly (shim or regulating rods).Table 4-7 compares the heating by assembly and in non-fueled components for the LEU corewith the rods banked at 53.5 cm. The banked rod critical configuration was found to have thehighest peak power density in the analysis of the HEU core during the HEU-LEU conversionanalyses. Elements 3-4, 4-3, and 4-4 are noted as having the highest average plate powers,and are therefore of interest for thermal-hydraulics analysis. Figure 4-17 provides a schematicof the core layout and the plate orientations. Table 4-8 compares the power in individual platesin elements 4-4, 3-4, and 4-3 for the reactor with the banked rods critical configuration. Thetallies were summed over the fuel meat in each fuel plate, all clad, coolant, and the bundle can.The plates are numbered from left-to-right in these elements (see the element drawings inFigure 4-17). It can be seen that plate 1348 in assembly or bundle 4-4 has the highest power(8.07 W at 1 kW). This plate is adjacent to the large water hole that the SS1 rod falls into, andnearer the center of the reactor than plate 1355 on the other side of the water hole. Plates 1228(bundle 3-4) and 1315 (bundle 4-3) face the center-line of the core, and have roughly equalpower (6.51 and 6.41 W at 1 kW).Figure 4-18 and Figure 4-19 compare the local-to-average axial power density profiles for fuelplates 1348, 1228, and 1315 for the banked rods critical configuration. Plate 1348 has thehighest peak power density of all the plates in the LEU core and was evaluated in the thermal-hydraulics analyses. The slight "pinching" of the axial power profile due to the insertion of theSS1 shim rod is evident.The local-to-average axial power density profiles in plates 1228 and 1315 are nearly identicaldue to their symmetric positioning across the core center-line. For these two plates, the highestpeak/average density (1.712) occurs in plate 1228. The peak power density in plate 1228 ofstandard element 3-4 is about 22% lower than that of plate 1348. Furthermore, the coolantchannel thickness in the standard fuel assembly is narrower than that in the control assembly(144 vs. 181 mils). This reduction in channel thickness brings the channel thickness closer tothe optimum value (= 100 mil), which will result in a higher ONB power with natural convectioncooling. Therefore, plate 1348 is the most limiting of all fuel plates.A higher water-to-fuel ratio at the edges of the fuel plates induces a power density profile alongthe width of the fuel plates. This is shown in Figure 4-19 for plates 1348, 1228, and 1315 withthe rods in the banked critical position. The radial segments are numbered from bottom-to-top inthe MCNP model (see the plate orientations indicated in Figure 4-17). It is expected that thepower density will be higher at the edge of the plate closest to the core center-line.Lastly, Table 4-9, Table 4-10, and Table 4-11 provide the axial and radial power profiles forplates 1348, 1228, and 1315, respectively. These data were used for thermal-hydraulic analysisof the LEU-fueled PUR-1.PUR-1 SAR 4-17 Rev 2, July23, 2015PUR-1 SAR4-17Rev 2, July 23, 2015 E2 F2 G2 H2(12) (13) (13) (12)I-KW7E3 H3(SS2) (1 13)E41F4IG4IH4E51F51G51H5(RR)I(13)I(13)I(13)Piudue LEU Core Layout (191 plates)Figure 4-17: PUR-1 Core Layout and Bundle Drawings.PUR-1 SAR 4-18 Rev 2, July 23, 2015PUR-1 SAR4-18Rev 2, July 23, 2015 Table 4-7: Bundle Powers Predicted by f7 and f6 Tallies in MCNP.AssemblyAssembly Average PowerPower Plate (W) Average(W) Power Plate1000 W (W) 12 kW Power (W)2-2 (RR) 38.19 4.77 458.28 57.242-3 61.85 4.76 742.2 57.122-4 (SS2) 45.65 5.71 547.8 68.522-5 45.93 3.83 551.16 45.963-2 62.56 4.81 750.72 57.723-3 78.09 6.01 937.08 72.123-4 78.42 6.03 941.04 72.363-5 61.20 4.71 734.4 56.524-2 63.52 4.89 762.24 58.684-3 80.78 6.21 969.36 74.524-4 (SS1) 60.09 7.51 721.08 90.124-5 62.29 4.79 747.48 57.485-2 52.32 4.03 627.84 48.365-3 63.97 4.92 767.64 59.045-4 63.49 4.88 761.88 58.565-5 47.64 3.97 571.68 47.64Inter-assemblywater 3.78 45.36PUR-1 SAR4-19Rev 2, July 23, 2015 Graphite reflector 9.52 114.24Grid plate 2.27 27.24Water reflector 216(pool) 18.00SS1 0.23 2.76SS2 0.17 2.04RR 0.03 0.36Total 1000.0 12000PUR-1 SAR 4-20 Rev 2, July 23, 2015PUR-1 SAR4-20Rev 2, July 23, 2015 Table 4-8: Plate Power (W) Computed from Heating Tallies in Bundles 4-4, 3-3 and 3-4 in PUR-1 Core with 190 Fuel Plates.Bundle 4-4 Power (W)Plate 1345 Meat 70Plate 1346 Meat 70.9Plate 1347 Meat 74Plate 1348 Meat 80.7Plate 1355 Meat 77.4Plate 1356 Meat 69.8Plate 12157 Meat 6589Plate 1217 Meat 54.9Plate 128ea65Plater 129Ma261.Plate 120Met5.4Platle 1222 Meatr59.1Plate 1223 Meat 57.Plate 1224 Meat 57.3Plate 1225 Meat 57.6Plate 1226 Meat 59.Plate 1227 Meat 61.2Plate 1228 Meat 659.1Clade 423Met5.8Wlater 214Met5.9Plate 1315 Meat 64.1Plate 1316 Meat 61.6Plate 1317 Meat 60.7Plate 1318 Meat 60.4Plater 139Met6.9Platle 1 -32 Meatr62.8Plate 1322 Meat 62.1Plate 1323 Meat 59.2Plate 1324 Meat 57.6Plate 1325 Meat 56.8Plate 1326 Meat 560.5Plate 1327 Meat 562.8Plate 1328 Meat 58.4Clad4.Water 2.Can 2.441Re , uy 3 21PUR-1 SAR4-21Rev 2, July 23, 2015 2.52.0 ___- _... ......1.01228 (Element 3-4).4 --Pate1315 (Element 4-3)0.01 2 3 4 5 6 7 8 9 10 11 12 13 14 1Axial SegmentFigure 4-18: Axial Power Profiles in Plates 1348, 1228 and 1315 for Banked Rod CriticalConfiguration in PUR-114513 ---- -- ---2 3 4 5 6 7 8 9 10 11Radial SegmentFigure 4-19:Radial Power Profiles in Plates 1345, 1228 andConfiguration in PUR-11315 for Banked Rod CriticalPUR-1 SAR 4-22 Rev 2, July 23, 2015PUR-1 SAR4-22Rev 2, July 23, 2015 Table 4-9: Axial and Radial Heating Profile for Plate 1348 of Bundle 4-4 for Banked CriticalConfiguration.Powerz-low1 z-high1 Power (W at 1 (W at 12Axial Segment (cm) (cm) kW) a kW)1 1.88595 5.88645 3.04 0.99% 36.482 5.88645 9.88695 4 0.86% 483 9.88695 13.88745 5.17 0.76% 62.044 13.88745 17.88795 5.95 0.70% 71.45 17.88795 21.88845 6.76 0.66% 81.126 21.88845 25.88895 7.2 0.64% 86.47 25.88895 29.88945 7.47 0.63% 89.648 29.88945 33.88995 7.71 0.62% 92.529 33.88995 37.89045 7.62 0.63% 91.4410 37.89045 41.89095 6.97 0.65% 83.6411 41.89095 45.89145 6.41 0.68% 76.9212 45.89145 49.89195 5.63 0.74% 67.5613 49.89195 53.89245 3.85 0.86% 46.214 53.89245 57.89295 1.72 1.25% 20.6415 57.89295 61.89345 1.2 1.48% 14.4Total 80.7 0.21% 968.4RadialSegmenty-low'(cm)y-high'(cm)Local/AveragePower DensityCa1 24.0838 24.6253 1.672 0.48%2 24.6253 25.1668 1.604 0.48%3 25.1668 25.7082 1.572 0.48%4 25.7082 26.2497 1.554 0.49%5 26.2497 26.7912 1.523 0.49%6 26.7912 27.3327 1.519 0.49%7 27.3327 27.8742 1.506 0.49%PUR-1 SAR 4-23 Rev 2, July 23, 2015PUR-1 SAR4-23Rev 2, July 23,2015 8 27.8742 28.4157 1.496 0.50%9 28.4157 28.9571 1.493 0.50%10 28.9571 29.4986 1.496 0.50%I 11 I29.4986 I30.04011.518I0.50%1Positions correspond to MCNP model of PUR-I.Table 4-10: Axial and Radial Heating Profile for Plate 1228 of Bundle 3-4 for Banked CriticalConfiguration.PowerAxial z-low1 z-high1 Power (W at 12Segment (cm) (cm) .(W at 1 kW) a kW)1 1.88595 5.88645 2.36 1.07% 28.322 5.88645 9.88695 3.11 0.94% 37.323 9.88695 13.88745 3.95 0.83% 47.44 13.88745 17.88795 4.65 0.77% 55.85 17.88795 21.88845 5.23 0.73% 62.760.70% 66.846 21.88845 25.88895 5.57_____0.70% 70.087 25.88895 29.88945 5.84 _____0.69% 71.768 29.88945 33.88995 5.98 _____0.69% 69.129 33.88995 37.89045 5.76 _____10 37.89045 41.89095 5.4 0.71% 64.811 41.89095 45.89145 4.98 0.74% 59.7612 45.89145 49.89195 4.36 0.80% 52.3213 49.89195 53.89245 3.42 0.89% 41.0414 53.89245 57.89295 2.55 1.05% 30.615 57.89295 61.89345 1.95 1.17% 23.4Total ____ _____ 65.11 0.22% 781.32RadialSegmenty-low'(cm)y-hig h1(cm)Local/AveragePoweraPUR-1 SAR 4-24 Rev 2, July23, 2015PUR-1 SAR4 -24Rev 2, July 23, 2015 Density1 24.0838 24.6253 1.424 0.52%2 24.6253 25.1668 1.320 0.52%3 25.1668 25.7082 1.262 0.53%4 25.7082 26.2497 1.229 0.54%5 26.2497 26.7912 1.194 0.54%6 26.7912 27.3327 1.193 0.55%7 27.3327 27.8742 1.178 0.54%8 27.8742 28.4157 1.178 0.55%9 28.4157 28.9571 1.189 0.54%10 28.9571 29.4986 1.215 0.54%11 I29.4986 I30.040 1 I 1.297 I 0.54%'Positions correspond to MCNP model of PUR-1.Table 4-11: Axial and Radial Heating Profile for LEU Plate 1315 of Bundle 4-3 for BankedCritical Configuration.Powerz-low1 z-high1 Power (W at 12Axial Segment (cm) (cm) (W at 1 kW) a kW)1 1.88595 5.88645 2.33 1.07% 27.962 5.88645 9.88695 3.03 0.96% 36.363 9.88695 13.88745 3.89 0.85% 46.684 13.88745 17.88795 4.63 0.79% 55.565 17.88795 21.88845 5.11 0.75% 61.326 21.88845 25.88895 5.43 0.72% 65.1668.047 25.88895 29.88945 5.67 0.70%68.888 29.88945 33.88995 5.74 0.70%66.969 33.88995 37.89045 5.58 0.70%62.7610 37.89045 41.89095 5.23 0.73%58.811 41.89095 45.89145 4.9 0.76%52.3212 45.89145 49.89195 4.36 0.81%41.0413 49.89195 53.89245 3.42 0.90%31.814 53.89245 57.89295 2.65 1.02% _____PUR-1 SAR4-25Rev 2, July 23, 2015 1 1 25.215 57.89295 61.89345 2.1 1.13%jTotal 64.07 10.23% 1768.84Local/AverageRadial y-low1 y-high' PowerSegment (cm) (cm) Density a1 16.3894 16.9309 1.279 0.54%2 16.9309 17.4724 1.201 0.55%3 17.4724 18.0138 1.172 0.56%4 18.0138 18.5553 1.147 0.56%5 18.5553 19.0968 1.144 0.57%6 19.0968 19.6383 1.151 0.56%7 19.6383 20.1798 1.174 0.55%8 20.1798 20.7213 1.199 0.55%9 20.7213 21.2627 1.246 0.54%10 21.2627 21.8042 1.313 0.52%11 21.8042 22.3457 1.435 0.51% _____1Positions correspond to MCNP model of PUR-1.4.5.3 Reactor Core Physics ParametersFor the model, the material composition for the plates was U3Si2-AI, and 6061 aluminumcladding. An addition of 20 parts per million of a boron-equivalent was added to the 6061cladding material to account for impurities in the alloy. The MCNP calculated maximum thermalneutron fluxes in the fuel region are 2.01 E10 n/(cm2*s) peak in the fuel region, and 1.38E10n/(cm2*s) average in the fuel region at 1 kW. Assuming a linearity of neutron flux with power,the maximum and average fluxes at 12 kW should be 2.41Ell and 1.66Ell n/(cm2*s),respectivelyEigenvalue calculations were performed with MCNP5, typically using from 25 to 50 millionneutron histories. These calculations yielded a reactor keff with a 1-c uncertainty of +/-17 pcm (0.017% Ak/k) for 25 million histories; the uncertainty was reduced to +/-12 pcm (0. 012% Ak/k)when 50 million histories were employed. The 1-c uncertainty of the eigenvalue calculationsreduces by the square root of the number of histories. The uncertainty of the reactivity feedbackcoefficients can be calculated as the square root of the sum of the squares of the eigenvaluecalculations. Based on the eigenvalue calculations with 25 to 50 million histories, theuncertainties of the water temperature and water void coefficients were found to be on the orderof 15% to 30% when calculated over the expected 10 to 20 °C perturbations of watertemperature.It was decided that the calculational uncertainty should be reduced by extending the number ofhistories in the MCNP eigenvalue calculations. Eigenvalue calculations for the nominal corestate and certain other cases which exhibit only small perturbations to the core keff wereperformed with 200 to 300 million neutron histories. This reduced the 1-c uncertainty of theeigenvalue calculations to around 5 pcm (0.005% Ak/k).PUR-1 SAR4-26PuR- SAR4-26Rev 2, July 23, 2015 4.5.3.1. Control Rod Worths and Excess ReactivityExcess reactivity of the LEU core in PUR-1 was determined to be 0.00468 (0.47%) Ak/k in a190 fuel plate core (16 dummies), including a reactivity bias of 0.32% Ak/k. This value is withinthe Technical Specification limit of 0.6% for excess reactivity.Control rod worths were calculated for the core and compared with measured data. Thecalculated and measured control rod worth values are shown in Table 4-12. The calculatedcalibration curves were done utilizing MCNP5 for the core, and Figure 4-20, Figure 4-21, andFigure 4-22 show the calculated control rod worth curves for each of the control rods, along withthe measured values for each of the rods.Table 4-12: Comparison of measured and calculated control rod worths.LEU Calculated (Ak/k) LEU Measured (Ak/k)Shim Safety 1 0.0377+0.0003 0.0393Shim Safety 2 0.0189+/-0.0003 0.0222Regulating Rod 0.0023+/-0.0003 0.0027PUR-1 SAR4-27PUR- SAR4-27Rev 2, July 23, 2015 SS-1 Rod Calibration Curve1.00%0.00%-0.50% '-1.00% jX-1.50%-2.0O0%-2.50%-3.00%-3.50% --4.00%010 20 30 40 50 60Rod Position (cm)(" SS-1 Calculated a SS-i Measured "---Poly. (SS-1 Calculated)Figure 4-20: Calibration curve for SS-1 rod with calculated and measured values.70SS-2 Rod Calibration Curve1.00%0.50%-0.00%--0.50%--1.00%--1.50%--2.00%010 20 30 40 50 60 70Rod Position (cm)(c SS-2Calculated 0 SS-2Measured ---Poly. (SS-2Calculated)Figure 4-21: Calibration curve for SS-2 rod with calculated and measured values.PUR-1 SAR4-28PUR- SAR4-28Rev 2, July 23, 2015 Reg Rod Calibration Curve0.35%0.30%-0.25%0.20%0.15%0.10%0.05%0.00%0~-. 00a070010203040506070Rod Position (cm)<" RR Calculated 0l RR Measured(RR Calculated)Figure 4-22: Calibration curve for RR with calculated and measured values.Using the rod worth curves in preceding figures, the maximum reactivity insertion rates weredetermined by finding the maximum slope, or rate of change, of the curve. The comparison ofcalculated and measured maximum reactivity insertion rates for PUR-1 are shown in Table 4-13.Table 4-13: Comparison of calculated and measured maximum reactivity insertion rates.Maximum Reactivity Insertion Rates for Control RodsCalculated MeasuredShim-safety I 1,75E-04 2.31 E-4Shim-safety 2 8.75E-05 1 .42E-4Regulating 4.66E-05 5.39E-5RodPUR-1 SAR4-29PuR- SAR4-29Rev 2, July 23, 2015 4.5.3.2. Shutdown Margqin Tesudw agnwscluae sn CP n a enmaue.Teedt rshown in Table 4-14. These values meet the technical specification (TS 3.1 .a).Table 4-14: Comparison of calculated and measured shutdown margins.Calculated MeasuredSS-2 Worth (Ak/k) -1.89% -2.22%Kexcess (Ak/k) 0.351% 0.42%Shutdown Margin (Ak/k) -1.58% -1.80%4.5.3.3. Other Core Physics ParametersReactivity coefficients and reactor kinetics parameters were calculated for the PUR-1 modelcore. These values are used to estimate the core reactivity response to changes in properties ofthe fuel temperature, coolant/moderator temperature, or coolant/moderator density. They aretherefore essential for analyses of reactivity-induced transients. These calculations wereperformed with the MCNP5 code using the same core model as for the core design and powerdistribution analyses.The reactor kinetics parameters evaluated for PUR-1 were the effective delayed neutronfraction, 3eff, and the prompt neutron lifetime, e. The effective delayed neutron fraction iscalculated using two eigenvalue calculations from MCNP5. Normal calculations of keff includeboth prompt and delayed neutrons. A additional calculation of keff is performed with delayedneutrons turned off in MCNP, yielding a keff that depends only on prompt neutrons, which isdenoted by k y~' The effective delayed neutron fraction is then defined ask promnptThe delayed neutron fraction was calculated using the formulation:keprfp ke ff p~pThe eigenvalue calculations were performed using MCNP5. The value for J3eff is calculated asshown below:keff =1.00379 + 0.00010k°' = 0.99589 + 0.00015PUR-1 SAR 4-30 Rev 2, July 23, 2015PUR-1 SAR4-30Rev 2, July 23, 2015 IDeft ' 1.00379/0.99589 = 7.870x103+/- 1.79x104 (2.3%)The bias of the PUR-1 MCNP model is APbias= 0.32% Ak/k. This was determined by comparingthe results of eigenvalue calculations for several cases with the control rods at measured criticalpositions. Accounting for the bias in the core model, the I5eff is calculated as:biaed -Ab _ -(k prompt -- Apbias) kef --' ef-f-liaedf ke- A kia kfefft -pl, ~ 1 -APb,aioThe resulting biased value is obtained then as:3 t~ed = (1.00379 -0.99589)/(1.00379 -0.0032) = 7.895x10-3 +/- 1.79x10.4 (2.3%)The reactivity insertion accident analyses were performed using the unbiased IDeft values, whichare slightly smaller (0.3%) than the effective delayed neutron fraction determined by accountingfor the bias in the MCNP model. Consequently, the reactivity insertion accident analyses wereperformed with a value of 15eff that gives more conservative results.The prompt neutron lifetime is calculated using the "1/v insertion method," where a uniformconcentration of a 1/v absorber such as 1°B is included at a very dilute concentrationeverywhere in the core and the reflector. The prompt neutron lifetime, e, is calculated by+-o N÷o[',vjwhere k1 is the keff of the system with a uniform concentration, N1, of a 1/v absorber, and 0a isthe infinitely-dilute absorption cross section of the absorber for neutrons at speed v. For thiseffort, the 10B absorption cross section is assumed to be Oa=3837 barns for a neutron speed ofv=2200 m/s.Reactivity coefficients provide an estimate of the reactivity response to changes in stateproperties, given as:Ap=a,.Axwhere ax is the reactivity coefficient due to a unit change in property x, and Ax is the valuechange for property x. The reactivity coefficients are calculated assuming that simultaneouschanges in multiple state properties are separable. These are calculated from core eigenvaluecalculations with independent perturbations to state properties, as shown here:Ap k,-ko~ 1Ax kok (x1-Xo)"FUR-I SAR 4-31 Rev 2, July 23, 2015PUR-1 SAR4-31Rev 2, July 23, 2015 PUR-1 has operated at a maximum power of 1 kW and is cooled by natural convection. Thenominal conditions for the core used in the reactor design model assumed fresh fuel (i.e., nofission products given the low burnup of the PUR-1), isothermal conditions of 2000, andimpurities in the fuel, clad, and graphite based on the best-available data.The reactivity coefficients are calculated assuming separability of the reactivity feedback effectsdue to changes in fuel temperature, water temperature, and water density. After establishing anominal state based on the conditions given above and a critical rod configuration determined inthe reactor design analysis, cases with perturbations to the temperatures or water density wereevaluated.The fuel temperature changes are assumed to occur uniformly throughout the reactor (i.e. thesame temperature perturbation in all fuel plates), while the water temperatures or densities wereperturbed in different zones of the reactor depending on the proximity to the fuel plates. In thetime immediately after the initiation of a reactivity induced transient, the water inside the fuelassembly can (shown in light blue in Figure 4-23) would heat up and have a feedback effect onthe transient. However, the water between the cans and also in the space between the controlrod and guard plates (shown in orange in Figure 4-23) would take some time to be heated as aresult of a power increase from the transient due to the slow water circulation time with naturalconvection cooling. It would take even longer for the water in the reactor tank to be heated.Consequently, cases were evaluated with the water temperature or density perturbed:* within the fuel assembly (light blue regions in Figure 4-23)* within the fuel assembly and the water between the assemblies (orange regions inFigure 4-23), and* within the fuel assembly, the water between the assemblies, and the pool or reflectorwater (all of the water in the MCNP5 model).For the evaluation of the reactivity induced transients, only the reactivity coefficients calculatedby perturbing the fuel assembly water will be used. A summary of the values determined in theanalyses of the reactor physics parameters and reactivity coefficients is presented in Tables 4-15 and 4-16.PUR-1 SAR 4-32 Rev 2, July 23, 2015PUR-1 SAR4-32Rev 2, July 23, 2015 Element 3-4ImElement 4-3NFigure 4-23: Figure showing water regions for perturbation modelsTable 4-15: Other core physics parameters.LEU(calculated)atfuel k. °C) ;8.05E-04amoderato -9.05E-03a]voidC %Ak /-1.93E-01r~e.f 0.784%e(ps) 81.3PUR.-1 SAR 4-33 Rev 2, July 23,2015PUR-1 SAR4-33Rev 2, July 23, 2015 Table 4-16: Water and Fuel Coefficients for the PUR-1 Core.PUR-1 LEU Core Design Water Temperature Coefficient (0awater)20 to 30 °C !-9.051E-05 IA P°c I+ 9%PUR-1 LEU Core Design Water Void Coefficient (aO~vod)0 to 0.60% void l-1 .933E-03 lAp/% void I+ !7%PUR-1 LEU Core Design Fuel Temperature Coefficient (afuel)20 to 127 °C !-8.053E-06 A Poc _+!10%PUR-1 LEU Core Design Effective Delayed Neutron Fraction (IBeff)S0.00784 +/- 0.00008For the LEU core, two different critical rod configurations were determined from the reactordesign analysis. A rod configuration with the SS2 rod inserted at 43 cm, and the SSl andregulating rods fully withdrawn, was found to result in the largest peak-to-average powerdensity. This critical rod configuration was modeled in the LEU core reactivity coefficients andkinetics parameters calculations.Figure 4-24: PUR-1 Core Layout with LEU Fuel.Figure 4-26 shows the effects of the fuel and water temperature, and water densityperturbations on the core reactivity. The core has a negative fuel temperature coefficientbecause of the Doppler effect on the U-238 capture resonances in the fuel. The fueltemperature coefficient is smaller than that from water temperature or density effects, but it is anon-negligible parameter for the LEU accident analyses.The behavior of the LEU core reactivity due to water temperature and density perturbations wasquite similar to that for the HEU core. The present LEU core does have a slightly harder neutronspectrum under nominal conditions, so the spectrum hardening due to the water temperatureincrease should have a smaller feedback effect on the core reactivity. On the other hand,PUR-1 SAR4-34PuR- SAR4-34Rev 2, July 23,2015 coolant voiding in the LEU core results in greater neutron leakage because of the harderspectrum, so the negative reactivity feedback effect due to the reduced water density is greaterfor the LEU core.Table 4-16 gives more detail for the reactivity feedback coefficients for watertemperature and density, and fuel temperature perturbations for the LEU core. As can be seen,the water temperature coefficient becomes more negative as the water temperature isincreased. The temperature coefficient calculated over the range from 20°C to 3000 for the "fuelassembly" water should be used for the accident analyses. It should be noted that the watertemperature coefficient over this range was actually larger than the HEU core. For the waterdensity feedbacks, the coefficient calculated over the range from 4000 to 60°C is the mostconservative value for the accident analyses.Temperature coefficients of reactivity were calculated assuming separability of the reactivityfeedback effects due to water temperature, water density (void), and fuel temperature. This is acommon practice and makes it easy to understand the inherent shutdown mechanisms that areresponsible for affecting reactivity-induced transients. Not only are the feedback coefficienttreated as separable, the reactivity coefficients were also calculated under non-isothermalconditions because different regions of the reactor will heat up at different rates during a powerincrease.Three distinct regions containing water in the PUR-1 were considered.1. The water between the fuel plates; this is called the "fuel assembly" water.2. The water between the fuel element cans and also the water in the control elementsbetween the control rod guard plates; this is called the "inter-assembly" water.3. The water in the reactor tank; this is called the "reflector" water.Water temperature coefficients were calculated by adjusting the temperature of the water ineach of these regions and calculating the impact on the core reactivity. Reactivity feedbackcoefficients due to the perturbation of the fuel assembly water temperature, the fuel assemblyplus inter-assembly water temperature, and fuel assembly, inter-assembly, and reflector watertemperature were calculated. It is important to note that only the feedback effect due to heatingof the fuel assembly water (between the fuel plates) was considered in the accident analyses,because only this water would experience an immediate heating due to power increases duringthe transient. The other regions containing water would take much longer to heat up.Heating the water increases the thermal motion of the hydrogen atoms in the water. The resultis to increase the energy of neutrons which are in "thermal equilibrium" with the hydrogenmoderator, thus hardening the neutron spectrum in the reactor. At higher neutron energies, theU-235 fission cross section is reduced. Thus, there is a negative reactivity feedback effect dueto heating of the water between the fuel plates.Increasing the temperature of the reflector (tank) water was found to have a positive reactivityfeedback effect over a small temperature range from nominal conditions. This is due to adecrease in the neutron absorption in the reflector as the temperature is increased and theneutron spectrum hardens. For the HEU core, the feedback coefficient due to heating all waterin the reactor tank from 20 to 30 00 was calculated to be 2.38x10-3 + 2.11x10-3 % Ak/k/°C.However, it would take a long time for any transient to heat the reflector water, so it is judgedPUR-t SAR4-35PuR- SAR4-35Rev 2, July 23, 2015 that there are no significant safety issues related to the positive reactivity feedback coefficientwhen all the water is heated. It should be noted that for larger increases in the temperature ofthe reflector water, the reactivity feedback effect is negative.The void coefficient was also converted to the unit of Ap/°C. The reason for expressing the voidcoefficient in the alternative units was to facilitate a comparison of the water void andtemperature reactivity feedback effects, which were treated as separable. This comparison isillustrated in Table 4-15 and Table 4-16. When the reactor coolant within the fuel element (i.e.,between the fuel plates, which is the region of interest for accident analyses in the PUR-1) isheated, the reactivity feedback from the water temperature increase is slightly larger.The unit conversion to Ap/°C was accomplished by equating the void (water density)perturbation to a corresponding water temperature perturbation at 1.5 atmospheres. This is thewater pressure in the PUR-1 core, which is at the bottom of a 15 foot tank of water (the actualtank depth is 17 feet, but the distance to the bottom of the core to the waterline is 15 feet. Thedensity of water at 1.5 atmospheres as a function of temperature is shown in Figure 4-25.1.0051.000 -0.995 __ __S0.990E0.98500.975S0.9750__0.965 -0.9600.9550 20 40 60 80 100 120Temperature (°C)Figure 4-25: Density of sub-cooled water at 1.5 atmospheres.MCNP allows for two adjustments on neutron scattering reactions based on the temperature ofthe medium. For neutron energies above 4 eV, the code adjusts the elastic scattering crosssections of nuclei in the medium using a free gas thermal treatment. The code user can specifythe temperature of each cell within the model and the cross sections are adjusted if thespecified temperature differs from the temperature of the processed nuclear data in the crosssection library.For modeling elastic and inelastic scattering events for neutrons below 4 eV, an treatment is employed. These data are available in the MCNP libraries for certain materials; lightwater and graphite are of interest for the PUR-1 analyses. S(oa,3) data for graphite at 20 °C wereemployed. Furthermore, S(oa,f3) which had previously been evaluated for light water at 20, 30,60, 100, and 150 00 were also used.PuR-1 SAR 4-36 Rev 2, July23, 2015PUR-1 SAR4-36Rev 2, July 23, 2015 Fuel Temperature PerturbationPUR-1 190-Plate LEU Core0.2%0.0% --0.2%-0.4% ______ ____ __________-0.6% _ _ ___ ________-0.8%------------ _________-1.0% --- --____-_________-1.2% --_________ __-1.4% -_____________-1.6%250 300 350 400 450 500 550 600 650Fuel Temperature (K)Water Temperature Perturbation0.2% PUR-1 190-Plate LEU Core-0.2%-0.4%ot -0.8%-1.0% Water-1.2°% -- FA+IA+REF Water ___ ___-1.6%20 40 60 80 100 120 140Water Temperature (°C)Water Density Perturbation0.2% PUR-1 190-Plate LEU Core-0.4% .... ............._ _ _-0.4% ..... .....-0.8% ---_ --- _ _-1.0% .......... -FA Water _ _ _ _ _ _ _-U--FA+IA Water-1.2% -_m --FA+IAeREF Water-1.6%20 40 60 80 100 120 140EfcieWater Temperature (eC)Figure 4-26: Effect of Fuel Temperature, Water Temperature and Water Density Perturbationson Core Reactivity.PUR-1 SAR 4-37 Rev 2, July 23,2015PUR-1 SAR4-37Rev 2, July 23, 2015 4.6 Thermal-Hydraulic DesignIn this section, the results of the thermal-hydraulic analyses are discussed in order todemonstrate that the PUR-1 core design provides the cooling capacity necessary to ensure fuelintegrity under all anticipated reactor operating conditions. Analyses for behavior underhypothetical accident scenarios are presented in Section 13.4.6.1 NATCON Code DescriptionThermal-hydraulic analyses were performed using the computer code NATCON1'2, which canbe used to analyze the steady-state thermal-hydraulics of plate type fuel in a research reactorcooled by natural convection. The reactor core is immersed in a pool of water that is assumed tobe at a constant average temperature.NATCON computes coolant flowrate, axial temperatures in the coolant and fuel plate surfaceand centerline, and the approach to onset of nucleate boiling (ONB). Other safety relatedparameters such as the Onset of Nucleate Boiling Ratio (ONBR) and Departure from NucleateBoiling Ratio (DNBR) are calculated as well. And an automatic search for the power at ONBcan be performed.Flow is driven by density differences in the coolant that are the result of coolant heating by thefuel. Resulting buoyant forces are counter-balanced by viscous forces that result from the flow.Hot channel factors may also be introduced for determining safety margins. NATCON v2.0documentation is included as Appendix 1 of this document. It includes information on thecalculation of hot channel factors, inputs, and use of the code.4.6.2 Fuel Element and Fuel Assembly GeometryIn PUR-1, each fuel plate element is loaded into its own assembly container, or can.. Crosssection views of the two different types of assemblies, standard and control, are shown in Figure4-27 and Figure 4-28.Figure 4-27: Standard LEU fuel assembly.PUR-1 SAR4-38PuR- SAR4-38Rev 2, July 23, 2015 Figure 4-28: Control LEU fuel assembly.Two types of channels are encountered in the PUR-1 fuel assemblies. One is the channelbetween plates, and the other is the channel between the last plate of an element and theassembly can wall. The plate-to-plate and plate-to-assembly wall channel thicknesses are fixedby the spacers on the wall of the assembly. It should be noted that the plate-to-wall channel isheated on only one side, so it can be conservatively assumed that half of the heat from the fuelplate associated with the fuel channel heats the coolant. Table 4-17 summarizes the channeltypes and thicknesses in PUR-I.Table 4-17: Channel Types and Thickness in PUR-1 AssembliesPlate-to-plate Plate-to-wall (mils)(mils)Standard Control Standard ControlLEU Assemblies 144+/-15 181+15 127+/-8 127+/-8In the thermal-hydraulic analyses, the peak power plates identified in Section 4.5.2 wereanalyzed using NATCON. The relative power densities in each fuel plate were obtained fromdetailed MCNP5 criticality calculations, also described in Section 4.5.2. In the NATCONanalysis, the relative axial power profiles of the individual plates were utilized in each respectivecase. Plate 1348 (see Table 4-9) was the limiting plate.Hot Channel Factors are used by NATCON to account for dimensional variations inherent in themanufacturing process, as well as variations in other parameters that affect thermal-hydraulicperformance. The geometry dimensions used in the NATCON models for the are shown inTable 4-18: Model Dimensions for the Thermal-Hydraulic Models. And the hot channel factorsare listed inPUR-1 SAR4-39Rev 2, July 23, 2015 Table 4-19. The hot channel factors were calculated using equations that include the effect oftemperature-dependent water viscosity. A conservative uncertainty of 20 mil (rather than 15 milshown in Table 4-4) on the 181 mil channel thickness (the most limiting fuel plate 1348 incontrol assembly 4-4) was used in finding the hot channel factors. To calculate the ONB power,the NATCON code was run using the hot channel factors shown inTable 4-19, for a total of 190 fuel plates, a channel thickness of 181 mil, a radial power factor of1.5414 (-the ratio of 8.07 kW power in plate to 5.2356 kW at 1 kW per average plate), and theaxial power shape for plate 1348 shown in Table 4-9.To account for the power density variation along the width of plate 1348, the FFILM inTable 4-19 (1.251) was increased by a factor of 1.085 (= 1.672/1.5412 = the maximum-to-average power density ratio variation over the width of plate 1348, shown in Table 4-9). Thisresults in FFILM =1.085x1.251 = 1.357 that was used in NATCON to calculate the ONB power.PuR-1 SAR 4-40 Rev 2, July 23, 2015PUR-1 SAR4-40Rev 2, July 23, 2015 Table 4-18: Model Dimensions for the Thermal-Hydraulic ModelsNumber of Axial Nodes 14Number of Plates 190Thermal Conductivity (W/m'K)Fuel Meat 80Clad 180Pool Temperature (°C) 30Inches mmFuel MeatHeight 23.625 600.08Width 2.345 59.563Thickness 0.00 0.508ChannelHeight 25.110 637.79Width 2.832 71 .933Channel Thickness 0.197 5.004Clad Thickness 0.015 0.381Distance assembly can extends above fuel 0.450 11.430plate ____PUR-1 SAR 4-41 Rev2, July 23, 2015PUR-1 SAR4-41Rev 2, July 23, 2015 Table 4-19: Hot Channel Factors for the Plate 1348 NATCON AnalysisCalculations were done to deterfnine the core power distribution and the ONB power for thePUR-1 based on the limiting plate, 1348. The ONB power calculation given below includes theimpact of the power density variation along the width of the plate itself. The resulting ONBpower was found to be 98.55 kW The 20 mil uncertainty used is conservative compared to the15 mil uncertainty given in Table 4-17 for the evaluated core design. These results are Table4-19. Reducing the channel thickness increases the ONB power because the new LEU designis approaching the optimum channel thickness (= 100 mil) which gives the highest ONB power.The NATCON code calculates the Darcy-Weisbach friction factor f = C/Re for fully developedlaminar flow, using a built-in table of the parameter C for different aspect ratios of therectangular channel cross section. To account for the increased pressure drop due tohydrodynamically developing laminar flow in the channel, an apparent value of the parameter Caveraged over the channel length, called Cap was calculated using Eq. (576) of Shah andLondon [Ref. 2]. The ratio Capp/C was found to be 1.0897 at a Reynolds number of 800 at theexit of the 181 mil channel in the new LEU design. Since the NATCON code multiplies the fullydeveloped friction factor by FW2, the hot channel factor FW equals 1.044 (= 1.08970.5). A highervalue of 1.048 for FW was used in the NATCON calculation to be conservative.The margin to incipient boiling shown in Table 4-21 was calculated at the present operatingpower of PUR-1 (i.e., 1 kW), and it is the smallest value of the temperature difference (ToNg -Tw) over the coolant channel length in the hottest channel where Tw is cladding surfacetemperature with all hot channel factors applied, and TONg is the local onset-of-nucleate-boilingtemperature. This basically gives an idea of how far below the onset of nucleate boilingcondition the reactor is operating. This definition can be written as an equation as follows:PUR-1 SAR4-42Rev 2, July 23, 2015 Margin to ONB =Minimum 7~lPpq(z)f -Flk{T(z)-71} +Fff1 TaiZ~TZ}whereT(z) = Bulk coolant temperature at axial position z in the channel heated bythe plate power of Pop Fr EQ/N and applying the global hot channelfactors for flow and Nusselt number of Fw and FhTwaij(Z) = Cladding surface temperature at axial position z in the channelheated by a plate power of PopFr EQ/N and applying the global hotchannel factors for flow and Nusselt number of Fw and Fhq"(z) = Heat flux at position z for the plate power of Pop~r F0/N andapplying the global hot channel factors for flow and Nusseltnumber of Fw and Ehp(z) -Absolute pressure in the channel at axial position zT1ocp(p(z), q"(z)FfluX) = Onset of nucleate boiling temperature at absolute pressure p(z)and heat flux q"(z)FfluxPop= Operating power of the reactor (e.g., 1 kW for PUR-1)N = Number of fuel plates in the core (e.g., 190 for PUR-1 LEU core)To = Coolant temperature at the channel inletFw = Hot channel factor for flow in the channelEQ = Hot channel factor for reactor powerFh = Hot channel factor for Nusselt numberEr =RPEAK = Radial power factor of the plate cooled by the channelEflrn = EEILM = Hot channel factor for temperature drop across the coolant film oncladding surfaceF~x= FFLUX = Hot channel factor for heat fluxEbulk = EBULK = Hot channel factor for bulk coolant temperature rise in the channel4.6.3 Thermal Hydraulic Analysis ResultsThe NATCON/ANL V2.0 code was used to determine the thermal-hydraulics performance of thePUR-I. Eirst, the code was used to compute the power at which the ONB is reached for theplates being examined.. This was done to identify the limiting channel. Then the limiting channelwas evaluated under nominal operating. The ONB results provide verification that the SafetyLimit (SL) and Limiting Safety System Settings (LSSS, trip points) of the TechnicalSpecifications will indeed assure safe operation of PUR-1.PUR-1 SAR4-43Rev 2, July 23, 2015 4.6.3.1. NATOON AnalysesThe reactor pool temperature varies throughout the year from about 2200 to 3000 depending onthe ambient temperature and humidity conditions in the reactor room. In all of the followingcalculations, the higher value 3000 was used.The power search function of NATCON was used to determine the power level at the Onset ofNucleate Boiling. Table 4-20 provides a summary of the ONB powers for each of the casesanalyzed. For the PUR-1 core, the limiting channel/plate was plate 1348 with the plate-to-platechannel, which had an ONB power of 98.6 kW.Table 4-20: ONB Powers for the high power plates.Plate 1348Plate 1228Plate 1315ONB Power (kW) 98.6 153.4 158.9Using NATCON, and the thermal-hydraulics parameters for the limiting plate (1348), thenominal operating conditions were also calculated. All hot channel factors are included in thesecalculations. These results are shown in Table 4-21.Table 4-21: Operating Conditions for PUR-1 as Determined by NATCON for Limiting Plate1348Present Power Uprate Power ONB PowerPower Level 1 kW 12 kW 98.6 kWMax. Fuel Temp. (0C) 31.92 39.1 112.6Max. Clad Temp. (°C) 43.42 43.4 112.5Coolant Inlet Temp. (°C) 30.0 30.0 30.0Coolant Outlet Temp. (0C) 31.7 35.8 46.3Margin to incipient boiling (°C) 78.3 67.9 0Coolant Velocity (rmins) 5.18 18.5 54.0Coolant Mass Flow Rate (kg/m2s) 5.16 18.41 53.4PUR-1 SAR 4-44 Rev 2, July 23, 2015PUR-I SAR4-44Rev 2, July 23, 2015 4.6.3.2. Safety Limits for the LEU CoreIn PUR-1, the first and principal physical barrier protecting against the release of radioactivity isthe cladding of the fuel plates. The 6061 aluminum alloy cladding has an incipient meltingtemperature of 582 °C. However, measurements (NUREG 1313) on irradiated fuel plates haveshown that fission products are first released near the blister temperature (-550 °C) of thecladding. To ensure that the blister temperature is never reached, NUREG-1 537 concludes that530 °C is an acceptable fuel and cladding temperature limit not to be exceeded under anyconditions of operation. As a result, PUR-1 has proposed a safety limit in its TechnicalSpecifications requiring that the fuel and cladding temperatures should not exceed 530 °C.4.6.3.3. Limitinq Safety System Settinqs for the LEU CoreLimiting safety system settings (LSSS) for nuclear reactors are settings for automatic protectivedevices related to those variables having significant safety functions. When a limiting safetysystem setting is specified for a variable on which a safety limit have been placed, the settingmust be chosen such that the automatic protective actions will correct the abnormal situationbefore a safety limit is reached. Table 4-22 shows the maximum power, the LSSS and operatingpower for PUR-1.Table 4-22: Key Power Levels for Reactor Operation and LSSS for PUR-11 kW 12 kW ProposedMaximum Power Level Including 50% Uncertainty 1 kW 12Limiting Safety System Settings Power Level 1.2 kW 12.0 kWOperating Power Level 1.0 kW 10.0 kWDuring steady-state operation, peak clad temperatures are maintained far below 530°C , as wellas below the temperatures required for ONB (see Table 4-21). NATCON was used to determinethe minimum power for ONB for core in the limiting channels, as well as the thermal-hydraulicparameters at these calculated powers. The results of these calculations are shown in Table4-21.The licensed operating power level of PUR-1 is 1 kW. The LSSS scram setting of 120% power(1.2 kW for 1 kW, 12.0 kW for uprate) is well below the power level of 98.6 kW at which ONBwould occur in the respective limiting channels. Thus, the present LSSS on power at 12 kW(120% normal operating power) will easily protect the reactor fuel and cladding from reachingthe Safety Limit under steady state operations.Chapter 13 (Accident Analyses) analyzes two hypothetical transients based on values of theTechnical Specifications for the LEU core. These transients are: (1) Rapid insertion of themaximum reactivity worth of 0.3% Ak/k of all moveable and non-secured experiments, and (2)Slow insertion of reactivity at the maximum allowed rate of 0.04% Ak/(k*s) due to control bladewithdrawal.PUR-1 SAR4-45Rev 2, July 23, 2015 For the case of the rapid insertion, of 0.3% Ak/k, the reactor scram was initiated based on thepower level trip, assuming failure of the period trip. For the case of the slow insertion of 0.04%Ak/(k*s), scram was initiated on the second power level trip, assuming the first power level tripfailed. The reason for this is that the period trip is never reached for the case of this slowreactivity insertion.Thus the selected LSSS is a conservative setting which ensures that the maximum fuel andcladding temperatures do not reach the safety limit of 530 °C for the range of accident scenariosthat were analyzed. In summary, the selected LSSS will protect the reactor fuel and claddingfrom reaching the safety limit of 530°C under any condition of operation.However, a NATCON thermal-hydraulic calculation for the LEU plate 1348 was performedassuming a hypothetical pool temperature of 35 00, and a hypothetical inlet loss coefficient of10.0 (increased from 0.5), while applying all six hot channel factors of the case. The ONB powerwas found to be 79.3 kW, indicating a large margin compared to the proposed PUR-1 operatingpower of 12 kW.4.7 References1R. S. Smith and W. L. Woodruff, "A Computer Code, NATCON, for the Analyses of Steady-State Thermal-Hydraulics and Safety Margins in Plate-Type Research Reactors Cooled byNatural Convection," Argonne National Laboratory (ANL), ANL/RERTR/TM-12, Dec 1988.SM. Kalimullah, "NATCON v2.0 Instructions", Argonne National Laboratory (ANL),ANL/RERTR, July 2006.PUR-1 SAR 4-46 Rev 2, July 23, 2015PUR-1 SAR4-46Rev 2, July 23, 2015 5 REACTOR COOLANT SYSTEMS5.1 Summary DescriptionThe water process system includes a 30 GPM water pump, a water filter, a demineralizer, flowmeter, a chiller, conductivity cells that measure the pool water before and after passing throughthe demineralizer and appropriate valves. The details of the reactor coolant system aresummarized inTable 5-1: PUR-1 water process system summary.WATER PROCESS SYSTEMWater Capacity, Reactor Pool 6,400 gal.Pump Type CentrifugalCapacity At 1000°F/1 10 ft Head 30 gpmIon Exchanger Mixed bed (replaceable cartridge type)Rated Capacity 15, 000 grains as CaCO3Flow Rate 0-20 gpmDesign Pressure 100 psiHeat ExchangerShell and tube (water chiller)Cooling Capacity36,000 Btu/hr (10,550 W)Design Pressure150 psiTube225 psiShellFilter Cartridge TypeMaximum Flow 80 gpmPressure Drop 2 psiPUR-1 SAR5-1PUR- SAR5-1Rev 2, July 23, 2015 5.2 Primary Coolant SystemThe process system is provided with the reactor to control the pool water quality andtemperature. The purification system is designed to limit corrosion and coolant activation by theuse of microfilters and ion exchange resins.The process system is assembled in one unit and is mounted on a base structure near one wallof the reactor room. The components include a circulating pump, a cartridge-type demineralizer,a Cuno filter, a refrigeration compressor-condenser unit, a heat exchanger, flowmeter,conductivity indicator, thermostat, wiring, piping and valves to complete the system. City wateris available through a water supply tank to provide a source of makeup pool water. City water isalso provided to the compressor-condenser unit as a secondary cooling medium.The reactor pool system is arranged in the following order: Water from the pool is drawn outfrom the scupper drain or suction line via PVC pipe leading to the circulating pump; a secondsource of water for. the pump is a water supply tank supplied with city service water andcontrolled by a float valve. Ball valves for water shutoff and a vacuum cleaning connection areprovided in, the pump supply line. From the pump a pipe with ball valve installed leads first tothe filter and then to a demineralizer. An adjustable by-pass or throttling valve is inserted in thesystem to regulate water flow through the demineralizer. A flow indicator and a conductivityindicator are installed as a check on water purity and flow rate from the demineralizer. Thewater next flows through a stainless steel heat exchanger. The water from the heat exchanger isthen returned to the reactor pool. A magnetrol water-level control is located in the reactor pool;this unit controls a solenoid valve in the line from the water supply tank to ensure that theprescribed pool water level is maintained.5.3 Secondary Coolant SystemThe water next flows through a Ross Model 302 stainless steel heat exchanger of 36,000 Btu/hrcapacity which serves as a water chiller; included with this system is a thermostaticallycontrolled Copelametic Model W300H compressor-condenser unit with a water cooledcondenser. The water from the heat exchanger is returned to the reactor pool.Experimentally, no temperature increase has been observed with the pool thermometerfollowing 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of operation at 1 kW. However, based on the mass of water as 1.85 104~ Kg,calculations indicate that the temperature increases after operating the PUR-I at power lever of10 kW would be 0.465°C/hour (this takes no credit for heat loss to the surrounding sand andgravel or loss by evaporation). The capacity of the chiller system is 10,550 Watts, which will beable to maintain pool temperature as required.The chiller is designed with three loops to prevent the spread of contamination in an emergency.The pool water passes through the primary loop while a Freon refrigerant is in the secondaryloop. The third loop uses campus water to remove the heat and is discharged into the campussewer system. The chance of contamination passing through the three loop system is small.5.4 Primary Coolant Cleanup SystemFrom the pump a pipe with ball valve installed leads first to the cartridge-type Cuno filter andthen to a Barnstead BD-1 0 demineralizer. An adjustable by-pass or throttling valve is inserted inPUR-1 SAR5-2Rev 2, July 23, 2015 the system to regulate water flow through the demineralizer. A flow indicator and a conductivityindicator are installed as a check on water purity and flow rate from the demineralizer.5.5 Primary Coolant Makeup Water SystemThe water level in the pool must be maintained at least 13 feet above the reactor core duringoperation. During periods when there are no planned operations, the water level is stillmaintained at approximately 13 feet. On average, the addition of about 40 gallons of water perweek is required to maintain that water level to make up for evaporation. The primary watermakeup system consists of a 20 gallon, gravity driven reservoir that is filled with water from theuniversity water system that is processed through deionizers consisting of each of an anion,cation and mixed-bed tank.This tank is filled as required, water is added as needed through a solenoid controlled floatswitch at the top of the reactor pool. Records are kept for the amount of water used with eachfill, and maintained at the reactor facility.5.6 Nitrogen-16 Control SystemThe main possible source of nitrogen-16 is from the fast neutron interaction with oxygen in thepool water. The nitrogen must then diffuse to the surface of the pool before it is released to theatmosphere. In normal operation, no strong currents are established in the reactor pool and withthe short half-life (7.14 seconds), the nitrogen decays before reaching the surface. No nitrogen-16 has been observed in the reactor room.5.7 Auxiliary Systems Using Primary CoolantPUR-1 has no auxiliary systems that use primary coolant.PUR-1 SAR 5-3 Rev 2, July 23, 2015PUR-1 SAR5-3Rev 2, July 23, 2015

v t e Letter Sequence Request
TAC:ME1594, (Open)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Administration Questions 24 March 2010 6 July 2011 8 July 2011 28 December 2011 29 May 2012 29 August 2014 23 December 2015 25 July 2016 Answers 4 June 2010 15 November 2011 4 January 2012 1 June 2012 15 June 2012 29 June 2012 13 July 2012 11 August 2012 29 January 2016 26 February 2016 31 March 2016 9 May 2016 19 July 2016 19 September 2016 Results Approval... Other: ML120900571, ML16265A632, ML16277A165
MONTHYEAR2012-06-29 [Table View]

ML15210A282

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