Determination of thermal hydraulic data of GHARR-1 under reactivity insertion

Nuclear Engineering and Design 241(2011)5203–5210Contents lists available at SciVerse ScienceDirectNuclear Engineering andDesignj o u r n a l h o m e p a g e :w w w.e l s e v i e r.c o m /l o c a t e /n u c e n g d esDetermination of thermal hydraulic data of GHARR-1under reactivity insertion transients using the PARET/ANL codeN.A.Adoo a ,∗,B.J.B.Nyarko a ,b ,E.H.K.Akaho a ,b ,E.Alhassan a ,V.Y.Agbodemegbe a ,C.Y.Bansah a ,R.Della aa School of Nuclear and Allied Sciences,University of Ghana,Legon,P.O.Box AE 1,Atomic,Accra,Ghana bNational Nuclear Research Institute,Ghana Atomic Energy Commission,P.O.Box LG 80,Accra,Ghanaa r t i c l e i n f o Article history:Received 7October 2010Received in revised form 13August 2011Accepted 12September 2011a b s t r a c tThe PARET/ANL code has been adapted by the IAEA for testing transient behaviour in research reactors since it provides a coupled thermal hydrodynamic and point kinetics capability for estimating thermal hydraulic margins.A two-channel power peaking profile of the Ghana Research Reactor-1(GHARR-1)has been developed for the PARET/ANL (Version 7.3;2007)using the Monte Carlo N-Particle code (MCNP)to determine the thermal hydraulic data for reactivity insertion transients in the range of 2.0×10−3 k/k to 5.5×10−3 k/k.Peak clad and coolant temperatures ranged from 59.18◦C to 112.36◦C and 42.95◦C to 79.42◦C respectively.Calculated safety margins (DNBR)satisfied the MNSR thermal hydraulic design criteria for which no boiling occurs in the reactor core.The generated thermal hydraulic data demon-strated a high inherent safety feature of GHARR-1for which the high negative reactivity feedback of the moderator limits power excursion and consequently the escalation of the clad temperature.©2011Elsevier B.V.All rights reserved.Contents 1.Introduction........................................................................................................................................52032.The PARET model ..................................................................................................................................52042.1.Margin of hydrodynamic crisis.............................................................................................................52043.Reactor description ................................................................................................................................52053.1.Reactor application(s)and power upgrading strategies...................................................................................52054.Model ..............................................................................................................................................52065.Results and discussion.............................................................................................................................52066.Conclusion .........................................................................................................................................5209References .........................................................................................................................................52091.IntroductionThe production of stable and sustained nuclear power is never guaranteed in the economy of neutrons.This condition is attributed to the effect of unstable fluid flow,boiling liquids or reactivity,mechanical vibrations,malfunctioning of control apparatus or a combination of any two of these situations.A necessary condition for stable reactor power is that the reactor distortion accompany-ing a temperature rise,decrease the reactivity,thus slowing any further rise in power.The assertion,that satisfying the above con-dition guarantees power stability is based on a false assumption that reactor distortion occurs simultaneously with temperature rise which caused it.This distortion lags behind the temperature∗Corresponding author.Tel.:+233244537602;fax:+23321400310.E-mail address:nanakwame10@ (N.A.Adoo).change by a time determined by the dynamics of the system.Therefore the change in reactivity also lags behind the tempera-ture change raising the possibility of unstable power oscillations Han and Thomas,1999.Reactivity and stability are affected by depletion of reactor fuels and the accumulation of “poisons”dur-ing operation (Woodruff,1982).This causes transient instability in reactor power and hence the need for the acquisition of a fair knowledge of the transients and their effects to the operation.In the analysis of reactivity accidents it is customary to consider tran-sients with and without scram event called,protected transients and unprotected or self-limited transients respectively.Unpro-tected reactivity transients have received attention because of their relevance in determining the reactivity insertion limits imposed by clad melting temperature (Woodruff,1984;Nasir et al.,1999;Mirza et al.,1998;Housiadas,2000).Recent research in nuclear reactor technology has placed a high priority on the development of efficient and accurate analytical tools to carefully predict the0029-5493/$–see front matter ©2011Elsevier B.V.All rights reserved.doi:10.1016/j.nucengdes.2011.09.0195204N.A.Adoo et al./Nuclear Engineering and Design241(2011)5203–5210 behaviour of thermal hydraulic transients.During transients,thesafety behaviour of a reactor is influenced by the design character-istics,operating conditions and event sequence.Some of the mostimportant parameters in transient analysis include:(a)peak clad and coolant temperatures(b)temperature and void coefficients of reactivity and(c)power distribution and power peaking factors(d)heatflux at the hot spot.This work considers a thermal hydraulic data build-up of unpro-tected reactivity transients for the Ghana Research Reactor-1(GHARR-1)with the use of the PARET/ANL code(Version7.3;2007).2.The PARET modelThe PARET(P rogram for the A nalysis of RE actor T ransients)code was originally developed by the US Atomic Energy Com-mission(Hainoun and Alhabit,2008)for analysis of the SPERTIII experiments for temperatures and pressures typical of powerreactors(Scott et al.,1967;Obenchain,1969).The code has nowbeen modified to include a selection offlow instability,departurefrom nucleate boiling(DNB),single and two phase heat transfercorrelations and a coolant properties library considered more appli-cable to low pressures,temperatures as well asflow rates whichare typical of research reactors.This now gives room for loss-of-flow(LOF)and reactivity insertion accident(s)(RIA)of researchreactors to be modeled.Essentially,the code provides a coupledthermal–hydraulic and point kinetics capability with continuousreactivity feedback,and an optional voiding model which estimatesthe voiding produced by subcooled boiling(Woodruff,1982,1983;Olson and Jonah,2007).The design model of the code for the reac-tor core can be represented by1–50channels–each with its ownhydraulic parameters,coolantflow rates,and power distribution(radial).Each coolant channel can be divided up into as many as21axial sections for which the heat transfer is analysed.The solutions of the one-dimensional hydrodynamic equationsused are for each of the channels at each time node and is based onthe momentum integrated model(MIM).The heat transfer couldtake place by natural or forced convection,nucleate,transition,orstablefilm boiling and coolant could range from subcooled liquidthrough two-phase regime up to and including superheated steam.It allows forflow reversal as well(Clancy et al.,1975;Woodruff andSmith,2000).The governing equations of this model are expressedas:∂¯ ∂t =−∂G∂z(1)∂G ∂t +∂∂zG2=−∂P∂z−f|G|G2D e−¯ g(2)∂E∂t+G∂E∂z=q(3)where G is the massflow rate,∂P/∂z the pressure gradient,t time,z the axial coordinate(positive upwards),D e the equivalent hydraulic diameter,f the friction factor,q heat source rate per unit volume, g the acceleration due to gravity and E average coolant enthalpy in the energy equation.The coolant may range between a subcooled liquid and a sat-urated two-phase mixture.The average coolant density in the mass conservation equation,the average coolant enthalpy E in the energy equation,and the effective densities and pertaining to the momentum equation and energy equation respectively,are defined as:¯ = l(1−˛)+ v˛Average density(4)1=(1− )2l(1−˛)+2v˛Momentum density(5) =[ l + v(1− )]∂˛∂Slipflow density(6)¯E=El(1− )+E v Average enthalpy(7) where subscripts l and v denote liquid water and water vapor respectively,˛vapour volume fraction(void fraction),and vapour weight fraction(mass quality).The two-phase friction factor and the void-to-quality relationship are determined using the Mar-tinelli method(Collier,1972).Reactor transient behaviour requires the solution of the usual point kinetic equations which represent axial or radial power dis-tributions in each region.This comprises the neutronics with six groups of delayed neutrons and nine groups of photo-neutrons and for which the solution is shown below employing a modified Runge Kutta method(Hetrick,1971;Hamidouche et al.,2004).dndt=( (t)−1)ˇeffn(t)+6i=1i C di(t)+9j=1j C pj(t)+S p(8)dC didt=dˇdin(t)− i C di(t)(9)dC pjdt=pˇpn(t)− j C pj(t)(10)ˇeff= d6i=1ˇdi+ p9j=1ˇpj(11) S p=S(12)where n is the neutron density(or power), reactivity load,ˇefftotal fraction of delayed and photo neutrons,ˇdi,ˇpj=i th and j th group of delayed and photo neutron precursors respectively, i, j,decay constants for the i th and j th precursors,C di,C pj=precursordensities of delayed neutrons and photo neutrons, di, pj= efficiency of delayed and photo neutrons,S p(Watts/s)=source term,S(neutrons/s)=neutron source strength, =energy released perfission, =number of neutrons emitted perfission, =prompt neutron generation time.2.1.Margin of hydrodynamic crisisThe heatflux in water-cooled nuclear reactor system is used to estimate the Departure from Nucleate Boiling Ratio(DNBR)of the system which is an important engineering design parameter for nuclear reactors.The DNBR signifies an operational safety limit and during normal operation,there may befluctuations in,for example, the power of the reactor,which can affect the temperature distri-bution along a coolant channel.This causes variations in the DNBR and may reduce the safety margins(Zafar,2006).Heat removal in the GHARR-1is dependent on the upper and lower orifice heights which were determined to ensure effective cooling of the core.The criteria of thermal-hydraulic design require that the burnout ratio or DNBR should be larger than the calculated burnout heatflux i.e.DNBR>2.5(SAR,1995).The DNBR for GHARR-1operating at the transient power level of99.44kW was estimated by the relation,DNBR=q dnbq norm(13)N.A.Adoo et al./Nuclear Engineering and Design241(2011)5203–52105205Fig.1.Axial cross section of GHARR-1.where qdnb [W/m2]is the heatflux at the departure from nucleateboiling and q norm is the heatflux corresponding to the prevailing power.In subcooled boiling,qdnb is a function of the coolant velocity,the degree of subcooling and pressure.For low velocityflow the Mirshak correlation was employed in this work to calculate the q dnb and is given as DNB(Garland,1997),q dnb=1.51×106(1+0.1197u)[1+9.14×10−3( T)]×[1+1.89×10−1«](14)u=1.2727˙mcool H2(15)T=(5.725+147.6H−2.674)T−0.35iP(0.59+0.0019T i)(16) where u[m/s]is thefluid velocity in the range,H is the height of the inlet orifice in millimeters(mm),T i is the inlet core coolant temperature(◦C),P is the power level of the reactor(kW),and«[MPa]is the pressure.3.Reactor descriptionGHARR-1is a commercial version of the Chinese-made minia-ture neutron source reactor(MNSR)of maximum power30kW at a thermal neutronflux of1.0×1012n/(cm2s).It has a small com-pact cubic-cylindrical core with dimensions of230mm×230mm in diameter and height shown in Figs.1and2.It is a pool type research reactor developed and manufactured by the Chinese Insti-tute of Atomic Energy of which the details have been summarized in Table1(Akaho et al.,2003).Thefission heat generated is transferred by natural convection (a phenomenon that cools the reactor)in an upward direction as shown in Fig.3and circulated within the reactor tank.Some heat is transferred by conduction through the tank wall to the pool water. Under normal operational conditions,theflow regime is single phase but nucleate boiling is expected under abnormal condition when power excursion occurs due to large reactivity insertions.3.1.Reactor application(s)and power upgrading strategiesGHARR-1is predominantly used for neutron activation anal-ysis(NAA).However there is on-going research to access the feasibility of extending its application to the production of short-lived radioisotopes for medical purposes.The reactor could also be exploited for non-destructive testing by neutrons,radioactive tracer production,radiation source to demonstrateradiography,Fig.2.Annular view of the GHARR-1.criticality experiment and control rod calibration(Akaho et al., 2003;Ahmed et al.,2006).The long term reactor operation depletes the core excess reac-tivity due to fuel burn up,Xe-poisoning and the negative feedback of coolant temperature.The GHARR-1core has a fuel burn up of1%and was designed with a lifetime core of10years if it is operated at its maximumflux for2.5h a day,five days a week (Anim-Sampong et al.,1999).To minimize neutron leakage and hence conserve the neutron economy,the core is heavily reflected on the side and underneath the fuel cage by a thick annulus and slab of beryllium alloy material(Akaho et al.,2003).Under normal operating conditions of the HEU fuel a drop in core excess reactivity within the range of2.5–2.8mk due to fuel depletion are observed (Anim-Sampong et al.,2007).Due to the aging lifetime of the core, regulated shims of beryllium have been added to the top aluminium tray to compensate for the loss of reactivity resulting from fuel burn up and accumulation offission products(Akaho et al.,2003).Studies have therefore been initiated towards the conversion of the present 90.2%enriched HEU core to a proposed12.6%Low Enriched Ura-nium(LEU)core(Anim-Sampong et al.,2007)after the end of life (EOC)of the present core.The heat removal mechanism in MNSR depends on natural con-vection which is associated with lowflow velocities and low heat transfer coefficients.To boost the heat removal from the reactor core,improving the cooling inside the reactor pool and reactor vessel were both considered by Hainoun and Alissa(2007).Table1GHARR-1core design parameters with HEU fuel loading.No.Parameter Values1Reactor type Pool-tank2Rated thermal power30kW th3Fuel element loading3444No.of dummy elements65Fuel composition U-Al dispersed in Al 6Fuel density 3.456g/cm37Dummy element material Al8U-235enrichment90.2%9Burn up∼1%10Loading of U-235in core998.12g11Coolant/moderator Deionised H2O12Coolant/moderator density995.1kg/m313Coolant inlet temperature30◦C14Coolant inlet pressure101.3kPa–1bar15Cooling mode Natural convection 16Reflector Beryllium17Control rod absorber Cadmium18Number of control element119Excess reactivity–cold,clean4mk20Total number of irradiation sites105206N.A.Adoo et al./Nuclear Engineering and Design 241(2011)5203–5210Fig.3.Schematic diagram of the coolant flow pattern.The two major concepts of power-upgrade strategies proposed were,enhancing the heat transfer coefficient between primary and secondary side of reactor vessel by pumping part of pool water through an annulus duct surrounding the vessel at reactor core level and cooling the circulated water with an external heat exchanger.The second option dealt with cooling the reactor water via a heat exchanger located outside the reactor pool.The reactor coolant in the upper part of reactor vessel was pumped out,transported to the external heat exchanger,cooled and re-pumped to the vessel.At this location the re-pumped cooled water mixed directly with the hot coolant and reduced thereafter the core inlet temperature with the consequence of restricting the increase of average core temperature.This concept as indicated by Hainoun and Alissa (2007)and shown in Fig.4can be technically realized using the already avail-able reactor water purification system which is used for controlling and ensuring the required quality of reactor water.However,some technical modifications are required before extending the use of this system for the new task.This will imply also additional safety and radiation protection measures for the case of external heat exchanger.The option of pool water cooling was found to extend the con-tinuous reactor operation time up to 10h whiles that of reactor water cooling was up to 15h,both at nominal reactor power and under the assumption of initial excess reactivity corresponding to the fresh reactor core.This was possible because the external heat exchanger in each case enhanced core cooling by eliminating the negative feedback of coolant temperature and providing an excess reactivity to compen-sate the exclusive limiting effect of the Xe poisoning.By analogy,coupling of these two scenarios as shown in Fig.5could yield better results by further prolonging the operation time of the MNSR.4.ModelSimulations for a 500,000neutron population for 500cycles were run using the MCNP-5code based on a modified input deckdesign for GHARR-1(Briesmeister,2000;Anim-Sampong,2001)for which a two-channel radial power peak profile is generated at steady state.The channels represent the hottest fuel pin and the remainder of the core respectively and each of these was sectioned into 21equally spaced nodes over the active core region to be used for transient simulations.The axial peaking profile was also com-puted using the same procedure corresponding to the MCNP input deck design for axial sectioning.The transients were initiated by external reactivity insertions which ranged from 2.0×10−3 k/k to 5.5×10−3 k/k and specified as a function of time employing a modified PARET input deck by (Olson and Jonah,2007)to suit the reactor design specifications some of which are shown in Table 2.5.Results and discussionThe reactivity insertion accident simulated in this work demon-strated the self-limiting power excursion feature of the GHARR-1upon the full insertion of cold clean excess reactivity.The MNSR control and protection system is designed with the fail-safe prin-ciple,nevertheless,should a shut-down failure occur,a limiting maximum available excess reactivity of about 4.0mk could beTable 2GHARR-1model parameters.ParameterValueFuel temperature coefficient (mk ◦C −1)˛fm =7.73×10−6to 2.9×10−7T fm +1.09T 2fmMeat radius,f mr2.13×10−3m Moderator density, mod 998.19g cm −3Moderator temperature coefficient (mk ◦C −1)˛m =2.6445×10−5to 3.4752×10−6T m Decay constant, (s −1)9.62×10−2Void coefficient (mk per %)−4.78Prompt neutron generation time, (s)8.1×10−5Effective delayed neutron fraction,ˇeff8.03×10−6N.A.Adoo et al./Nuclear Engineering and Design241(2011)5203–52105207Fig.4.Conceptual MNSR power upgrading strategies by cooling pool water and vessel water(Hainoun and Alissa,2007). inserted.This available excess reactivity is less than1/2ˇeff to avertthe possibility of prompt criticality in the reactor.The simulationscarried out ranged from500to1.50×103s in transient span afterwhich the results are presented with their respective powers andtemperatures in Figs.6–13.The analysis which was carried out atnominal power level of30kW and inlet coolant of20◦C,consid-ered reactivity feedback effect,bulk temperature as well as the peakcladding temperature output.Simulated transient events initiatedby positive reactivity insertions ranging between2.0mk and5.5mkhave also been analysed.To check the accuracy of the simulation,a2.0mk step reactivity insertion was considered for which thepeak power recorded was40.17kW as against34.4kW and34.8kWobtained by RELAP-5and TEMPFED respectively(Akaho,2002)andapproximately37kW by TPR2in PARR-2(Mirza,1997).Experimen-tal results for a2.0×10−3 k/k reactivity insertion shown in Fig.8resulted in a peak power level of37.91kW and a coolant tempera-ture difference, T of22.42◦C in518s.Results from PARET/ANL forthe temperature profile are as follows:42.95◦C for the bulk coolant,59.18◦C for the clad and88.32◦C for the meat.This is illustrated inFigs.6and7.A simulated3.6mk reactivity insertion produced a maximumpower of71.22kW in6.23min in Fig.9.For this reactivity,theFig.5.Conceptual core cooling and long-time operation design for MNSR.0.00E+005.00E-031.00E-021.50E-022.00E-022.50E-023.00E-023.50E-024.00E-024.50E-020.00E+00 2.00E+02 4.00E+02 6.00E+028.00E+02 1.00E+03 Power(MW)Time (s)Fig.6.Power plot for2.0mk reactivity insertion.0.00E+001.00E+012.00E+013.00E+014.00E+015.00E+016.00E+017.00E+018.00E+019.00E+011.00E+020.00E+00 2.00E+02 4.00E+02 6.00E+028.00E+02 1.00E+03Temperature(°C)Time (s)Tclad CTf c/l CTout C Fig.7.Temperature plot for2.0mk reactivity insertion.5101520253035400100200300400500600700 Power(kW)Time (s)Fig.8.Experimental power profile for2mk reactivity for GHARR-1.5208N.A.Adoo et al./Nuclear Engineering and Design 241(2011)5203–52100.00E+001.00E-022.00E-023.00E-024.00E-025.00E-026.00E-027.00E-028.00E-020.00E+00 2.00E+02 4.00E+02 6.00E+028.00E+02 1.00E+03 1.20E+03 1.40E+03 1.60E+03P o w e r (M W )Time (s)Fig.9.Power plot for 3.6mk reactivity insertion.0.00E+002.00E+014.00E+016.00E+018.00E+011.00E+021.20E+021.40E+021.60E+020.00E+001.00E+022.00E+023.00E+024.00E+025.00E+02T e m p e r a t u r e (°C )Time (s)Tclad C Tf c/l C Tout CFig.10.Temperatures for ramp insertion of 3.6mk.0.00E+001.00E-022.00E-023.00E-024.00E-025.00E-026.00E-027.00E-028.00E-029.00E-021.00E-010.00E+001.00E+022.00E+023.00E+024.00E+025.00E+026.00E+027.00E+02P o w e r (M W )Time (s)Fig.11.Power plot for 4.0mk reactivityinsertion.T e m p e r a t u r e (C )Time (s)Fig.12.Temperature plot for 4.0mk reactivity insertion.020406080100120050100150200250300350400450500P o w e r (k W )Time (s)Fig.13.Experimental power profile for 4mk reactivity for GHARR-1.temperature profile for the peaks as shown in Fig.10are 60.57◦C for the moderator temperature,83.66◦C for the cladding and 134.26◦C for the fuel meat.The maximum allowed coolant temperature was 70.09◦C and the change in thermal transients was 3.4–6.2s between the meat and clad and then the clad to the coolant respec-tively.The transient response for GHARR-1upon simulation with RELAP-5code for 3.6mk ramp insertion presented comparable results with a peak power of 76kW after 6min of insertion with a corresponding maximum fuel centre temperature of 91.4◦C (SAR,1995).A 4mk ramp RIA was simulated by the PARET code to study the GHARR-1system response for a fully withdrawn control rod –a case for an unprotected transient for maximum allowed core excess reactivity as shown if Figs.11and 12.This is one of the severe tran-sients for the MNSR since it considers a completely stuck control rod accident for which the passive safety mechanism of the reac-tor is paring previous work by Akaho and Maakuu (2002);for the same ramp core excess reactivity to this work,the peak power as predicted by TEMPFED was 100.8kW which agreed favourably with the experimental result of 100.2kW in Fig.13.In comparison to Pakistan Research Reactor-2(PARR-2)which shares very similar design parameters with GHARR-1,Research Reactor Simulator (TPR2)under the same initiation event,effec-tively demonstrates the passive safety feature of PARR-2with a maximum reactivity insertion of the cold clean core (Mirza,1997).This work however presents a peak power of 91.13kW in a response time of 5min 50s which is justan approximate peak obtained from the PARR-2transient over –power simulation for the same reac-tivity magnitude as shown in Fig.14.A study of Tables 3and 4which present the power and temperature data for the variousFig.14.Variation of transient reactor power of 2mk and 4mk reactivities in PARR-2using the TPR2code (Mirza,1997).N.A.Adoo et al./Nuclear Engineering and Design 241(2011)5203–52105209Table 3Peak power and transient response times.Power (kW)Time (s)Reactivity (mk)Exp RELAP-5PARET/ANL TEMPFED Exp RELAP-5PARET/ANL TEMPFED 2.037.9134.440.1734.85186405933553.6–75.571.22––240332–4.0100.22–91.13100.8294–3243535.5–163.04–––150–Exp,experimental.Table 4Peak temperature and transient response times.Temperature (◦C)Reactivity (mk)RELAP-5PARET/ANL TEMPFED T outT max T fc T clad T out T max T fc T clad T out 2.048.451.968.368.042.9558.3688.3259.1855.73.661.562.191.591.260.5770.09134.2683.66–4.0––––64.1072.12161.2686.9272.65.5––––79.4283.23212.63112.36–reactivities reveal differences in time trends with the two codes.For example in the case of the power excursion for the 3.6mk inser-tion,whiles RELAP5recorded a response time of 240s,PARET/ANL comes up with a 332s time which is attributed to the different dis-cretization schemes/numerical schemes employed by these codes in solving the hydrodynamic equations.Whiles it is well established that PARET/ANL uses a modified Runge Kutta scheme,it is not clear whether the implicit or semi-implicit numerical solution scheme is employed by RELAP5in this case.Besides RELAP5is not the best code to simulate atmospheric pressure conditions as may be the case in this system.The differences in the experimental and ana-lytical results may be due to uncertainties in the correlations used in the models.The maximum reactivity load limit for the PARET simulation in this work was 5.5mk as illustrated in Figs.15and 16.The peak power calculated by PARET was 163.04kW which produced a large power pulse within a relatively short period of 150s.The fluctuat-ing power and thermal transient is due to the reactivity feedback caused by heating in the moderator (negative temperature coeffi-cient of reactivity),hence the reactor power is forced to drop until a stable power level of 102.12kW.This power excursion lasts for a period of 110s.Similar effects are experienced in the bulk coolant temperature and clad for the thermal hydraulic profile.The calcu-lated DNBRs for the heat fluxes generated during the 2mk,4mk and 5.5mk RIAs were found to be 13.42,6.63and 4.15respectively.0.00E+002.00E-024.00E-026.00E-028.00E-021.00E-011.20E-011.40E-011.60E-011.80E-010.00E+001.00E+022.00E+023.00E+024.00E+025.00E+02P o w e r (M W )Time (s)Fig.15.Power plot for 5.5mk reactivity insertion.0.00E+002.00E+014.00E+016.00E+018.00E+011.00E+021.20E+021.40E+021.60E+021.80E+020.00E+001.00E+022.00E+023.00E+024.00E+025.00E+02T e m p e r a t u r e (C )Time (s)Fig.16.Temperature plot for 5.5mk reactivity insertion.6.ConclusionThe PARET/ANL code has successfully been used under the MNSR operating conditions to simulate reactivity transients ranging from 2.0×10−3 k/k to 5.5×10−3 k/k.The code solved neutron and thermal hydraulic dynamic models under nominal reactor power and the results gave a basis for good comparison with earlier works on similar reactivity magnitudes by “Akaho et al”and analysis in the SAR.The code has been used to find the thermal hydraulic margins ofGHARR-1and has demonstrated that the high negative reactivity feedback of moderator temperature limits power excursion in the reactor.The maximum clad temperatures did not reach the condi-tion of onset of subcooled boiling and the fuel meat temperature remained below 640◦C –the melting point of U-Al.From the results of the safety margins,it can thus be said that no boiling occurred within the core since the MNSR thermal hydraulic design criteria for DNBR >2.5and also all measured temperatures were below the saturation temperature of light water at 100◦C at 101.3kPa.ReferencesAhmed,Y.A.,Ewa,I.O.B.,Umar,I.M.,Bezboruah,T.,Johri,M.,Akaho,E.H.K.,2006.TheLow Power Miniature Neutron Source Reactors:Design,Safety and Applications,ICTP Miramare –Trieste IC/2006/020.Akaho,E.H.K.,Maakuu,B.T.,2002.Simulation of reactivity transients in a miniatureneutron source reactor core.Nuclear Engineering and Design 213,31–42.Akaho,E.H.K.,Anim-Sampong,S.,Dodoo-Amoo,D.N.A.,Maakuu,B.T.,Emi-Reynolds,G.,Osae,E.K.,Boadu,H.O.,Bamford,S.A.,2003.Ghana Research Reactor-1Final。