Criticality safety of fuel debris, particularly MCCI(Molten-Core-Concrete-Interaction) products, is one of the major safety issues for decommissioning of Fukushima Daiichi Nuclear Power Station. Criticality or subcriticality condition of the fuel debris is still uncertain; its composition, location, neutron moderation, etc. are not yet confirmed.
Progress in Nuclear Energy 101 (2017) 321e328 Contents lists available at ScienceDirect Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene Study of experimental core configuration of the modified STACY for measurement of criticality characteristics of fuel debris Satoshi Gunji a, *, Kotaro Tonoike a, Kazuhiko Izawa b, Hiroki Sono b a b Japan Atomic Energy Agency, Nuclear Safety Research Center, Shirakata 2-4, Tokai-mura, Ibaraki, Japan Japan Atomic Energy Agency, Department of Fukushima Technology, Shirakata 2-4, Tokai-mura, Ibaraki, Japan a r t i c l e i n f o a b s t r a c t Article history: Received 26 August 2016 Received in revised form 10 February 2017 Accepted March 2017 Available online 13 April 2017 Criticality safety of fuel debris, particularly MCCI(Molten-Core-Concrete-Interaction) products, is one of the major safety issues for decommissioning of Fukushima Daiichi Nuclear Power Station Criticality or subcriticality condition of the fuel debris is still uncertain; its composition, location, neutron moderation, etc are not yet confirmed The effectiveness of neutron poison in cooling water is also uncertain for use as a criticality control of fuel debris A database of computational models is being built by Japan Atomic Energy Agency (JAEA), covering a wide range of possible conditions of such composition, neutron moderation, etc., to facilitate assessing criticality characteristics once fuel debris samples are taken and their conditions are known The computational models also include uncertainties which are to be clarified by critical experiments These experiments are planned and will be conducted by JAEA with the modified STACY(STAtic experiment Critical facilitY) and samples to simulate fuel debris compositions Each of the samples will be cladded by a zircalloy tube whose outer shape is compatible with the fuel rod of STACY and loaded into an array of the fuel rods This report introduces a study of experimental core configurations to measure the reactivity worth of samples simulating MCCI products Parameters to be varied in the computation models for the experimental series are: It is concluded that the measurement is feasible in both under- and over-moderated conditions Additionally, the required amount of samples was estimated © 2017 The Authors Published by Elsevier Ltd This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/) Keywords: Fuel debris MCCI product Modified STACY Fukushima Daiichi nuclear power station Critical experiment Uranium dioxide with 235U enrichments of 3, 4, and wt.%, Concrete volume fraction in the samples of 0, 20, 40, 60, and 80%, and Porosity of the samples filled from to 80% where the sample void is filled with water Introduction Criticality safety is one of the major safety issues for defueling of the damaged reactors in Fukushima Daiichi Nuclear Power Station (1F-NPS) (Tonoike et al., 2013) Criticality control method of fuel debris must be established in the mid- or long-term process of defueling and decommissioning A significant difference, from the view point of criticality control, between situations in the 1F-NPS reactors and the Three Mile Island Unit reactor (TMI-2) is that * Corresponding author E-mail address: gunji.satoshi74@jaea.go.jp (S Gunji) cooling water for fuel debris in the 1F-NPS reactors cannot be poisoned continuously as was done in the TMI-2 Because the cooling water is not flowing in a closed loop, the destination of injected water is not known Therefore, it is difficult to manage concentrations of the poison in the cooling water for the purpose of criticality control It would be necessary that a mitigation-based criticality control method is adopted for decommissioning of 1F-NPS For this purpose, it is necessary to get the criticality characteristics of fuel debris However, the actual fuel debris in the reactors has not yet been observed and it is difficult to obtain accurate information on its composition, location, neutron moderation, etc (Tonoike et al., 2015) This situation leads to large uncertainty in estimation of criticality characteristics, and criticality or subcriticality condition of the fuel debris Therefore, a database of computational models for possible criticality characteristics of the fuel debris is being built which will help to predict in which condition critical events may occur (Tonoike et al., 2015) Most of criticality characteristics of fuel debris have not been http://dx.doi.org/10.1016/j.pnucene.2017.03.002 0149-1970/© 2017 The Authors Published by Elsevier Ltd This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/) 322 S Gunji et al / Progress in Nuclear Energy 101 (2017) 321e328 evaluated before the accident of 1F-NPS, especially, that of moltencore-concrete-interaction (MCCI) product Molten core might drop from the pressure vessels of the reactors on the concrete floors in the containment vessels, where MCCI products might be produced MCCI product has a small neutron absorption cross-section, and may be porous and contain water when submerged In fact, it has been confirmed that the containment vessel floors are submerged (Status of Fukushima Daiichi Nuclear Power Station (2015)) In past studies, criticality characteristics of MCCI products had been evaluated only by computations Some of them suggest small critical mass and high boron concentration in cooling water that guarantee subcriticality (Izawa et al., 2012) Detail study of criticality characteristics of MCCI products, therefore, should be conducted, including criticality experiments, for establishment of criticality control or criticality risk assessments Critical experiments are being planned to validate such highaccuracy computations to support criticality safety or criticality risk evaluation of defueling in 1F-NPS that will change a volume ratio of fuel debris and water It will be conducted at the modified Static Experiment Critical Facility (STACY) with samples simulating fuel debris compositions (Tonoike et al., 2015) In this paper, experimental core configurations with samples of MCCI products were considered based on recent knowledge of criticality characteristics of fuel debris The amount of samples to be prepared was determined as function of their reactivity worth and the differences of the core critical water heights This is important because there are limitations on the insertion reactivity and the critical water height in particular specifications of the critical facility Analysis and experimental conditions 2.1 Submerged MCCI product The moderation condition of critical experiments should be varied, however, more widely because hydrogen in the concrete would contribute to neutron moderation and because the amount of hydrogen in actual MCCI products is still unknown A series of analyses of criticality characteristics of MCCI products was shown in Ref 5, where infinite multiplication factors of MCCI products with 235U enrichments of 3, 4, and wt.% were computed in homogeneous and heterogeneous conditions The results indicate that optimum moderation conditions of those MCCI products would be at the volume ratios of moderator to fuel (Vm/Vf) of 0.2e4 Vm/Vf is expressed by following equation (1), this is specifically the modified STACY design based This value means ratio of water volume to fuel pellet volume in an active core Therefore, it does not consider water contents in samples Volume of moderator water Vm V ¼ Volume of UO2 f (1) 2.2 Outline of the modified STACY The modification of the STACY is now underway at Japan Atomic Energy Agency (JAEA) in order to accumulate fundamental experimental data relating to the criticality control for fuel debris handling in 1F-NPS The modified STACY is designed to be a tanktype light-water-moderated critical assembly, whose first criticality is expected in FY2018 (Sono et al., 2015; Miyoshi et al., 2015) An overview of the modified STACY is shown in Fig Each fuel rod will consist of UO2 pellets with a diameter of 8.2 mm and a 235U enrichment of wt.%, and a zircalloy cladding with an outer- Fig Concept of modified STACY diameter of 9.5 mm The stack height of the pellets will be 1420 mm The modified STACY will be operated by means of filling the tank with water, which works as neutron moderator and reflector, from the bottom of the core tank Reactivity will be adjusted by adjusting the water height Vm/Vf of the core will be varied by changing fuel rod interval For example, Vm/Vf is 1.2 when a fuel rod interval in a square lattice is 11.5 mm (Izawa et al., 2015; Sakon et al., 2015), which was selected as the experimental core configuration in this study 2.3 Experimental core configurations Two experimental core configurations were studied by using the MCNP 5.1 code system (X-5 Monte Carlo Team, 2003; Brown et al., 2009) and the nuclear data library JENDL-4.0 (Shibata et al., 2011) In order to have statistical error of less than 0.015 %Dk (~2 cent1), Â 107 effective neutron histories was used for each calculation The effective critical water heights of the two core configurations were approximately 1000 mm One configuration was designed to have a hard neutron spectrum and an under-moderation configuration, which is shown in Fig The configuration consists of 701 fuel rods arrayed in a uniform square lattice with the interval of 11.5 mm Its critical water height is estimated to be 990 mm The other configuration has a soft neutron spectrum and an over-moderation configuration The configuration consists of 400 fuel rods in total and is divided into two regions as shown in Fig The “driver region” is an array of fuel rods with the same interval of 11.5 mm that surrounds the “test region” The “test region” consists of 85 fuel rods arrayed more sparsely with 84 positions left vacant (filled with water), an effective interval of 16.3 mm, and Vm/Vf of about 3.7 Its critical water height is estimated to be 935 mm The neutron spectra at the center of each experimental core The effective delayed neutron fraction was calculated by SRAC/TWODANT; beff ¼ 0.0075 S Gunji et al / Progress in Nuclear Energy 101 (2017) 321e328 323 Fig Neutron spectra at the center of each core configuration Fig “Under-moderation” experimental core configurations in square lattice of the modified STACY array and a water height of approximately 1000 mm To determine the reactivity of the samples, each sample was modeled in the base array and the reactivity worth was obtained from the change in keff Effective multiplication factors (keff) were 1.00035 ± 0.00011 and 1.00357 ± 0.00011, respectively, for the under-moderation and the over-moderation configurations The relation between the water height and keff of each configuration are shown in Fig The reactivity worth per water level change of each configuration was estimated to be about 0.62 ¢/mm or 0.63 ¢/mm at a water height of 1000 mm The accuracy of the water height gauge of the modified STACY will be ±0.2 mm Therefore, the reactivity worth derived from a water height difference will have an accuracy of ±0.1 ¢, which will be acceptable as the experimental precision Thus, the reactivity worth of pseudo fuel debris samples in each experimental configuration should be greater than 0.3 ¢ to be distinguished from zero Geometrical buckling of each experimental configuration should be minimally changed The limitation of the changes was determined to be less than 1% This limitation was set to determine the experimental limit, and there is no reason based on quantitative consideration Under these experimental conditions, the change of the water heights should be less than 100 mm Therefore, favorable change of water height in reactivity worth measurement should be from 0.5 to 100 mm, which corresponds to reactivity worth of pseudo fuel debris samples from 0.3 to 62 ¢ Fig “Over-moderation” experimental core configurations in square lattice of the modified STACY condition are shown in Fig The thermal neutron flux of the overmoderation configuration was about times as large as that of the under-moderation configuration These are considered the base arrays with only fuel rods in the Fig Relations the effective water height and keff of each core 324 S Gunji et al / Progress in Nuclear Energy 101 (2017) 321e328 Reactivity worth of samples 3.1 Samples for reactivity worth measurements The samples of pseudo fuel debris simulating MCCI products were modeled using several compositions which were made of uranium dioxide with 235U enrichments of 3, 4, and wt.% fuel; and a concrete A list of sample types in this study is shown in Table The composition of the concrete in this study is shown in Table and its density is 2.3 g/cm3 Concrete volume fractions in the samples were 0, 20, 40, 60, and 80% Additionally, porosities of these samples were varied from to 80%, which were filled with water Concrete volume fraction is expressed by following equation (2), and porosity is expressed by following equation (3), respectively; Concrete volume fraction %ị ẳ Volume of concrete Volume of MCCI product 100 (2) Porosity %ị ẳ Volume of moderator water Volume of MCCI product ỵ moderator waterị 100 (3) Ideal sample loading conditions, based on reactivity worths for each loading, were evaluated for five arrays with 1, 5, 5, 9, and 13 MCCI products samples loaded in the patterns illustrated in Fig The in the test region of the under-moderation configuration fuel rods were replaced with sample rods For the over-moderation configuration, the samples were inserted into vacant positions in the test region Values of keff were computed for arrays of fuel rods and the samples Reactivity worths were estimated by comparing the keff values and those of the base arrays The estimated relative reactivity worth of the pseudo fuel debris whose 235U enrichment is wt.% are shown in the following sub-sections 3.2 Relative reactivity worths by changing the concrete volume fraction Fig and Fig show the computation results of relative reactivity worth dependency of the concrete volume fraction in each configuration They are the results of the samples based on using the 235U enrichment of wt.% fuels and their porosities are 0% Fig shows that the changing of the concrete volume fraction has a big impact on the reactivity worths in the under-moderation configuration The reactivity worth of the samples with no concrete, shown in the figure, was negative because the 235U enrichment of wt.% was lower than the wt.% enrichment of fuel rods The absolute value was, however, small and the worth turned to positive if the concrete volume fraction is beyond 40% There was a tendency that reactivity worths increase into the positive according to increase the concrete volume fraction It is considered that the water in the concrete contributed to moderate of neutron in these configuration Fig shows that for the over-moderation configuration, the reactivity worths are negative for all of the patterns because the samples excluded the moderator water There was a tendency that reactivity worths increase into the negative according to the concrete volume fraction increase The positive reactivity worths should have been inserted, because the moderation conditions of Table A list of the reactivity worth samples and their specifications Sample Materials Length Loading Patterns MCCI product with zircalloy cladding Parameters; 235 U enrichment (3, 4, and wt.%) Concrete volume fraction (0, 20, 40, 60, and 80%) Porosity (0, 20, 40, 60, and 80%) 1420 mm 1, 5a, 5b, 9, and 13 Table The composition of the concrete in this study Element Number density [atoms/b cm] Element Number density [atoms/b cm] Element Number density [atoms/b cm] H O C Na 1.374E-02 4.592E-02 1.153E-04 9.640E-04 Mg Al Si K 1.239E-04 1.741E-03 1.662E-02 4.606E-04 Ca Fe 1.503E-03 3.451E-04 Fig Loading patterns of the reactivity worth samples into the test regions S Gunji et al / Progress in Nuclear Energy 101 (2017) 321e328 Fig Relative reactivity of the pseudo fuel debris samples by changing the concrete volume fraction in the under-moderation configuration 325 change in the critical water height is too much for a larger number of samples Table and Table summarize the reactivity worth per each sample in each moderator condition In addition, samples of 100% concrete and water with zircalloy cladding are shown as references In the under-moderation configuration, positive reactivity worths were inserted by increase of the concrete volume fraction, and more reactivity worth was inserted by insertion of the 100% concrete sample Table shows the effect of the replacement of the fuel rod of the water sample is approximately 12 ¢, and that of the 100% concrete sample is approximately 3.3 ¢ in each insertion pattern And wt% fuel rods (see Concrete volume 0%) have negative reactivity worths in each insertion pattern Moreover, the maximum reactivity worth was inserted by swapping the fuel rods for water holes In this configuration, the reactivity worths per rod were almost the same for sample types in each insertion pattern On the other hand, in the over-moderation configuration, small negative reactivity worths were inserted by increase of the concrete volume fraction Table shows the effect of the replacement of the fuel rod of the water sample is approximately À2.5 ¢, and that of the 100% concrete sample is approximately À10¢ in each insertion pattern Sensitivities of both the insertion pattern and the concrete volume fraction for the reactivity worths were small 3.3 Relative reactivity worths by changing the porosity Fig Relative reactivity of the pseudo fuel debris samples by changing the concrete volume fraction in the over-moderation configuration this case were close to the suitable moderation condition by removing water However, contrary to expectations, the reactivity worths remained negative Maybe, these results show that dry condition is nearly optimum moderation condition Furthermore, in this configuration, it is also concluded that loading of up to samples will be suitable to measure their reactivity worth because Fig and Fig 10 show the computation results of relative reactivity worth depend on changing the porosities of the sample in “Pattern 5a” for several concrete volume fraction in each configuration They are the results of the samples based on using the 235U enrichment of wt.% fuels The relative reactivity worths in each configuration has proportional relations to porosities Fig shows that the increasing of the porosities have moderation effects, therefore, the samples has a positive reactivity worth About 40 ¢ positive reactivities occurred by the porosity increased from to 80% in the under-moderation configuration The effect of porosity changing is dominant than that of the concrete volume fraction changing, because the amount of hydrogen differ by one order of magnitude between two parameters Fig 10 shows that about 25 ¢ positive reactivities occurred by the porosity increased from to 80% in the over-moderation configuration This results show that this experimental core configuration is not enough “over-moderation”, because the positive reactivity worths were inserted by increasing of water content 3.4 Additional analysis for over-moderation core configuration In section 3.3, it has turned out that “over-moderation” core configuration was not have enough moderation ability Therefore, a new over-moderation experimental core configuration which Table The reactivity worth per sample rod in the under-moderation configuration (Unit: cent/rod) Samplea Concrete Concrete Concrete Concrete Concrete Concrete Water a b Volume Volume Volume Volume Volume Volume 0% 20% 40% 60% 80% 100% b All sample has zircalloy cladding References Pattern Pattern 5a Pattern 5b Pattern Pattern 13 1.1 0.8 ỵ0.7 þ0.7 þ2.0 þ3.4 þ11.9 À0.5 À0.2 À0.1 þ0.6 þ1.4 þ3.4 þ11.8 À0.6 À0.4 þ0.2 þ0.6 þ1.4 þ3.2 þ12.0 À0.8 À0.4 þ0.1 þ0.8 þ1.4 þ3.3 þ11.5 À0.7 À0.3 þ0.1 þ0.7 þ1.3 þ3.0 þ10.9 326 S Gunji et al / Progress in Nuclear Energy 101 (2017) 321e328 Table The reactivity worth per sample rod in the over-moderation configuration (Unit: cent/rod) Samplea Concrete Concrete Concrete Concrete Concrete Concrete Waterb a b Volume Volume Volume Volume Volume Volume 0% 20% 40% 60% 80% 100 %b Pattern Pattern 5a Pattern 5b Pattern Pattern 13 À7.4 À8.9 À9.5 À10.7 À10.2 À9.8 À2.4 À9.7 À10.0 À10.3 À10.8 À11.2 À10.5 À2.5 À8.3 À8.9 À9.5 À9.7 À10.5 À10.1 À2.6 À9.6 À10.0 À10.1 À10.6 À11.0 À10.2 À2.6 À11.0 À11.1 À11.1 À11.3 À11.4 À10.4 À2.7 All sample has zircalloy cladding References Fig Relative reactivity of the samples in the under-moderated “Pattern 5a” configuration Fig 11 A new “Over-moderation” experimental core configurations in square lattice of the modified STACY Fig 10 Relative reactivity of the samples in the over-moderated “Pattern 5a” configuration shifted array of fuel rods was considered This configuration is shown in Fig 11 In this configuration, local Vm/Vf (¼3.7) at the test region not change by insertion of the reactivity worth samples The relative reactivity worths of the samples by changing the concrete volume fraction and the porosities are shown in Figs 12 and 13, respectively As considered in section 3.3, some features of the over-moderation were seen in this core configuration The increase of moderator water by increasing of the concrete volume Fig 12 Relative reactivity of the pseudo fuel debris samples by changing the concrete volume fraction in the new over-moderated configuration S Gunji et al / Progress in Nuclear Energy 101 (2017) 321e328 327 prepared It was revealed that the experimental “over-moderation” core conditions in this study was not enough over-moderation condition for the sample of pseudo fuel debris Therefore the “new” over-moderation core configuration was analyzed in this paper This configuration was good to evaluate of the criticality characteristics for high concrete volume fraction samples Further studies Fig 13 Relative reactivity of the samples in the new over-moderated “Pattern 5a” configuration The experiment plans drafting in the modified STACY is carried out continuously Further discussion is necessary on criticality characteristics of the 1F-NPS fuel debris For example, water content of fuel debris, MCCI products, and usage of neutron absorber materials should be studied before the experiment using the modified STACY In this paper, a combination of enriched uranium fuel, concrete and water was considered as a first plan, other combinations (burnup, cladding, steel construction, control rod, and so on) should be studied in near future It is scheduled to conduct the actual measurements of reactivity worth for those materials using the modified STACY after FY 2020 Table The reactivity worth per sample rod in the new over-moderated configuration (Unit: cent/rod) Samplea Concrete Concrete Concrete Concrete Concrete Concrete Waterb a b Volume Volume Volume Volume Volume Volume 0% 20% 40% 60% 80% 100%b Pattern Pattern 5a Pattern 5b Pattern Pattern 13 À4.2 À4.7 À7.4 À9.4 À11.5 À13.1 À8.1 À3.7 À5.5 À7.8 À11.0 À15.5 À20.6 À18.5 À3.4 À5.1 À6.7 À8.9 À11.7 À13.7 À8.8 À3.3 À4.9 À7.2 À9.9 À13.6 À17.3 À14.0 À3.5 À5.5 À8.2 À12.3 À18.8 À29.4 À31.2 All sample has zircalloy cladding References fraction or the porosities caused insertion of negative reactivity worths Especially, in these graphs, the relations of the concrete volume fraction and the reactivity worth or the porosity and the reactivity worth are characterized by not being linear Table shows the reactivity worth per each sample in each moderator condition with references In this configuration, negative reactivity worths were inserted by increasing of the concrete volume fraction, and more negative reactivity worth was inserted by insertion of the 100% concrete sample Moreover, minimum negative reactivity worths were inserted by swapping the fuel rods for the 100% uranium fuel without water These features in the over moderation configuration have not seen in the past “over-moderation” configuration described in section 3.3 Conclusions As a part of design works of critical experiments, core configurations to measured reactivity worth of MCCI products were studied It was found that the measurements using the modified STACY in under- and over-moderation configurations with pseudo fuel debris simulating MCCI products are feasible because the worth can be estimated with enough accuracy from change of the critical water height The suitable loading numbers of the samples were estimated From these results, it is possible to determine the amount of the pseudo fuel debris sample which should be Acknowledgments This report includes results of the contract work funded by the Nuclear Regulation Authority (NRA)/the Secretariat of NRA of Japan References Brown, F.B., et al., 2009 MCNP5e1.51 Release Notes 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facility for experimental study on fuel debris criticality control Chap In: Nuclear Back-end and Transmutation Technology for Waste Disposal, vol 22 Springer, pp 261e268 Status of Fukushima Daiichi Nuclear Power Station, 2015 available online URL http://www.tepco.co.jp/en/nu/fukushima-np/index-e Tonoike, K., et al., 2013 Major Safety and Operational Concerns for Fuel Debris Criticality Control Proceeding of GLOBAL 2013, Salt Lake City, Utah, USA, September 29-October 3, pp 729e735 328 S Gunji et al / Progress in Nuclear Energy 101 (2017) 321e328 Tonoike, K., et al., 2015a Options of principles of fuel debris criticality control in Fukushima Daiichi reactors Chap In: Nuclear Back-end and Transmutation Technology for Waste Disposal, vol 21 Springer, pp 251e259 Tonoike, K., et al., 2015b Study on Criticality Control of Fuel Debris by Japan Atomic Energy Agency to Support Nuclear Regulation Authority of Japan Proceeding of ICNC 2015, Charlotte, North Carolina, USA, September 13e17, pp 20e27 Tonoike, K., et al., 2015c Criticality Characteristics of MCCI Products Possibly Produced in Reactors of Fukushima Daiichi Nuclear Power Station Proceeding of ICNC 2015, Charlotte, North Carolina, USA, September 13e17, pp 292e300 X-5 Monte Carlo Team, 2003 MCNP e a General Monte Carlo N-particle Transport Code, Version LA-UR-03-1987, LANL, USA ... recent knowledge of criticality characteristics of fuel debris The amount of samples to be prepared was determined as function of their reactivity worth and the differences of the core critical... drafting in the modified STACY is carried out continuously Further discussion is necessary on criticality characteristics of the 1F-NPS fuel debris For example, water content of fuel debris, MCCI... were inserted by increase of the concrete volume fraction Table shows the effect of the replacement of the fuel rod of the water sample is approximately À2.5 ¢, and that of the 100% concrete sample