Nb3Sn Cable-in-conduit Conductors Research Articles

Metallographic investigation of first full-size high-Jc Nb3Sn cable-in-conduit conductor after cyclic loading tests

In order to verify the feasibility of applying high-Jc Nb3Sn strand in fusion magnet, a full-size cable-in-conduit conductor (CICC) with short twist pitch (STP) cable pattern was manufactured and tested in SULTAN facility at SPC, Switzerland. Three levels of cyclic electromagnetic (EM) load were applied on the sample stepwise, no visible decrease of current sharing temperature (Tcs) was observed until the EM load increased to 80 kA × 10.8 T, after that the Tcs decreased dramatically with the EM cycles, which suggested that irreversible deformation, causing a change in the strain state, or even damage has occurred in the superconducting strands. For investigating the reason which caused the conductor performance degradation, the tested conductor was dissected for metallographic observation. Eight segments which subjected to different EM loads were extracted from one of the legs, the geometric feature changes of the cable cross-sections were analyzed and compared. A good correlation was found between the decrease of the Tcs and deformation of the cable cross section. A mass of cracks were found on the sub-elements of strands in the segment which subjected to highest EM load, but the amount of crack is much lower in other segments. Combining the analyses, it is speculated that the critical EM load which causes irreversible degradation is between 850 kN/m and 870 kN/m for this conductor. The results could be a reference in high-Jc Nb3Sn CICC design.

Impact of indentation on the performance of high Jc Nb3Sn strand

To meet the peak magnetic field requirements of the central solenoid (CS) and toroidal field (TF) magnets in the China Fusion Engineering Test Reactor (CFETR), cable-in-conduit conductors (CICC) that are composed of Nb3Sn strands with high critical current density (Jc) have been developed. The Short-Twist-Pitch (STP) design is adopted to avoid performance degradation caused by the electromagnetic cycles. But the STP design and tight cabling will cause severe indentation at the contact point between strands. Previous studies on Nb3Sn CICC show that the indentation has a significant effect on the critical current (Ic) and residual resistivity ratio (RRR) of the superconducting strand. Therefore, the same study was carried out for the high Jc Nb3Sn superconducting strand produced by Oxford Superconducting Technology (OST) and Kis wire Advanced Technology CO., Ltd. (KAT). In this paper, the Ic and RRR measurements of sample at different indentation depths were made, and the results were compared with the previous Nb3Sn indentation strands and flat rolled samples. Also, the hysteresis loss of the samples with different indentation depth was measured, and the change of effective filament diameter with indentation depth was obtained.

Qualitative Study On Cable Breakage of Nb3Sn CICC Based On Direct Current Potential Drop Method

Nb3Sn cable-in-conduit conductor (CICC) technology is widely applied to the large-scale superconducting magnets. Due to the multi-level composite structure, it is easy to cause partial damage or even strand breakage of the superconducting cable in the complex deformation mode. To evaluate the cable breakage of Nb3Sn CICC, direct current potential drop (DCPD) technique is proposed in this study. Serious of benchmark samples of cable with stage 3 were prepared and different breakage rates were designed in the cable samples. Potential difference (PD) obtained from the both side of the cable breakage was measured and analyzed. The experimental results indicated that direct current potential drop (DCPD) method is effective to qualitative measurement of cable breakage.

Critical bending wavelengths associated with current sharing temperature degradation in multistage twisted Nb3Sn cable CICC

The concept of a bending wavelength has been proposed in the description of strand deformation phenomena arising from electromagnetic loads, in cable-in-conduit conductor (CICC) with multistage twisted Nb3Sn cable, that result in the degradation of the CICC current sharing temperature (Tcs). A survey of CICC Tcs performance covering Nb3Sn CICC samples from the recent past two decades shows interesting results that can help relate bending wavelength to CICC parameters such as the void fraction (Vf) and the twist pitches of the CICC cable to which the latter has attributed to Tcs degradation according to prior research. In this paper, a simple cable strand model is used to link CICC parameters especially the CICC Vf to the bending wavelength of the CICC strand. Through this treatment, a range of bending wavelengths are identified that can be associated to characteristically different Tcs performance behaviour of the CICCs. The bending wavelength values corresponding to transitions between these characteristically different CICC Tcs behaviours are identified as critical bending wavelengths. Additionally, a picture of the internal mechanism resulting in Tcs change is presented along with comparisons of bending wavelength values derived from different theories and the value from destructive examination identified in the literature for Nb3Sn CICC samples.

Thermal-hydraulic analysis of the DTT CS and PF pulsed coil performance during AC operation

The EU DEMO fusion reactor, now entering in its conceptual design phase, will have to tackle several challenges, such as the long plasma pulse duration and the exhaust of very high heat fluxes from the plasma. The Divertor Tokamak Test (DTT) facility, a compact superconducting tokamak under construction at ENEA Frascati, will address the power exhaust problem in DEMO by testing several DEMO-relevant divertor solutions and operation scenarios.The DTT superconducting coils are the subject of the analyses presented here: as the tokamak must be very flexible to face different plasma scenarios, the design of the magnet system must be proven to be robust. In particular, the thermal-hydraulic performance of the coils operated in pulsed mode, namely the Central Solenoid (CS) and the Poloidal Field (PF) coils, will be analyzed. The CS is composed of 6 modules, layer-wound with Nb3Sn Cable-in-Conduit Conductors (CICCs), while the PF coils are pancake-wound. The two PF coils closest to the machine axis are wound using Nb3Sn CICCs, while the others adopt NbTi CICCs. All the coils are cooled by supercritical He (SHe) in forced flow.The pulsed operation of these coils induces AC losses, eroding their temperature margin: a detailed thermal-hydraulic model of the DTT pulsed coils is developed here using the 4C code. The model is then applied to the simulation of the two reference plasma scenarios, a single null and a double null, computing the minimum temperature margin. Different cooling options (single- or double-pancake) are investigated for the PF coils, while the sensitivity to the coupling time constant value is assessed for the CS.

Impact of mechanical and thermal cycles at different operating conditions on the ITER toroidal field coil conductor performance

The impact of electromagnetic (EM) and thermal cyclic loading (referred as warm-up-cool-down—WUCD) on the Nb3Sn cable in conduit conductors for the toroidal field magnets of the ITER tokamak should be carefully assessed for the proper operation of the machine. The experience gained during the production phases of the TF conductors revealed that a degradation occurs when thermal and EM cyclic loading is applied at the nominal operating conditions of 10.78 T and 68 kA. Since at the beginning of the tokamak operation the nominal conditions will not be achieved, it is worth investigating to what extent lower EM loads affect the conductor performance, and the threshold of EM load that triggers the WUCD degradation. The aim of this investigation is to identify critical levels which determine the onset of this degradation, both in terms of thermal and EM cycles, and to study the performance of strands from different suppliers. For this assessment, a set of conductor samples were manufactured including all types of strands to be used in the TF coils of the ITER machine. These samples were subjected to a set of EM and thermal cycles representing in a realistic way the actual operation of the conductors in the ITER machine. The tests were performed in the SULTAN facility of the Swiss Plasma Center at working conditions ranging from 5.4 T–34 kA to 10.78 T–68 kA. This work presents the results obtained in these tests in terms of current sharing temperature (T cs), effective strain, rate of degradation at different working conditions and their implications for the prospected conductor performance in the machine.

Examination of design parameters affecting Nb3Sn CICC current sharing temperature using necessary condition analysis

The performance of Nb3Sn cable-in-conduit conductor (CICC) especially its current sharing temperature (T cs) is known to depend on the various parameters of the CICC such as the Nb3Sn strand characteristics, its cabling, the final geometry of the CICC cross section, the void fraction, the current distribution and other parameters. However, the performance has not been very predictable. The change in CICC T cs performance with repeated operation or electromagnetic loading is of particular concern. Furthermore, evaluation of T cs performance with respect to parametric variation of some CICC parameters has resulted in conflicting conclusions for different CICC designs. Thus, a more comprehensive review of published CICC T cs performance with respect to the corresponding CICC cable characteristics has been performed and is reported here for CICC samples whose designs and test results have been published during the past decade and more. Necessary condition analysis as well as linear regression is applied in the interpretation of the results which has found that the first stage cable twist pitch which has been the focus of much research is not that significant in determining the T cs performance but does affect T cs degradation. The direct effect of strand critical current on T cs performance was also found to be weak. On the other hand, void fraction was found to be significant in both determining T cs performance and limiting T cs degradation. Other correlations between parameters have also been studied. Overall, it seems that there has been appropriate targeting as well as overemphasis on some CICC parameters for improving Nb3Sn CICC T cs performance. No single parameter studied has been found to be a determining factor for high stable CICC T cs performance. Research on additional parameters is warranted.

Estimation of performance of Nb3Sn CICC with thermal strain distribution

In the last few years, the critical temperature (T c) of several ITER cable-in-conduit conductors (CICC) was determined by magnetization measurements at zero current. The distribution of the critical temperatures (T c), caused by variation of strain in the Nb3Sn strands, was found to vary with the number of load cycles. A comparison with mechanical modelling requires the strain distribution, while the T c distribution is sufficient to determine the CICC performance. The current sharing temperature (T cs) is calculated supposing that the measured distribution of T c is representative of its variation along a single strand and that the current is uniformly distributed among the strands. The T cs values, found for background and peak magnetic field, are compared with the results of the DC test in the SULTAN facility. The presented estimation techniques lead to an overestimation of T cs. The differences of the estimated and measured T cs values are discussed including the effect of different current-voltage characteristics of single strands and CIC conductors.

The effect of strain and filament fracture on the transport properties of ITER cable-in-conduit conductors

Dedicated experimental and modeling research studies on the performance of superconducting cable-in-conduit conductor (CICC) have been massively performed and are still ongoing in order to determine the operational limits of the conductors and to optimize their design. Strand strain distribution and crack formation in the filaments after cabling and compaction, and under cooling down and electromagnetic load have been considered as the main cause for the degradation of the CICC’s transport properties. In combination with the strain maps generated by the mechanical model MULTIFIL and the electromagnetic code JackPot with the basic electrical and strain properties of the superconducting strand, the current sharing temperature (Tcs) of the CICC of the ITER Central Solenoid has been simulated and analyzed. A quantitative analysis of the Tcs degradation due to strain variation and filament fracture, respectively, is still missing. Here, the approach of analyzing the performance of CICC (e.g., the short samples tested in the SULTAN facility, or the full-size CICC used in real magnets) has been presented. Consequently, the effect of filament fracture on the cable Tcs has been investigated and turns out to be limited. Instead, the dominant mechanism behind the degradation of the transport properties of ITER type Nb3Sn CICCs is shown to be the broadening and shift in the strain distribution of the superconducting filaments.

Calculation method for pulsed magnetic field energy supplied to Nb3Sn ITER CS conductors during SULTAN stability tests

Cable-In-Conduit Conductors (CICCs) for the ITER Central Solenoid (CS) magnets are designed to operate in the presence of fast changing current and magnetic field during the plasma-operating scenario. For ITER, the AC loss of several types of Nb3Sn CICCs was experimentally tested, but only very limited experimental data is available for quantitative analysis of the minimum quench energy (MQE). In the SULTAN testing facility (Swiss Plasma Centre) few CS conductors were tested on MQE, but the magnetic field amplitude and ramp rate settings are far from the actual ITER operating conditions. Nevertheless, such tests are needed as a basis to calibrate and benchmark the codes that describe the quench behavior. Moreover, during the stability tests in Sultan, the temperature measurements show severe fluctuations, which can introduce a large error for the energy calculation. An interpretation is given for the temperature fluctuation and a procedure is proposed to significantly reduce the error in the pulsed energy calculation.

Electromagnetic and thermal stability of the ITER Central Solenoid during a 15 MA plasma scenario

The combination of the electromagnetic conductor model JackPot ACDC with the thermo-hydraulic model THEA is able to reproduce and predict the behavior of cable-in-conduit conductors (CICCs) under any varying current and magnetic field. The combined JackPot-THEA model is used to simulate the turns of the CSU2 quadra-pacake of the ITER Central Solenoid (CS) that are subjected to the most demanding conditions of the 15 MA plasma scenario. The considered section is a pancake with a 150 m length of Nb3Sn CICC from helium inlet to helium outlet, from the inner towards outer turns. The simulation results in terms of temperature and local strand electric field levels are compared to those obtained from earlier minimum quench energy analysis in order to evaluate possible risk of quench. The results confirm sufficient stability of the ITER CS coil, both from electrical and thermal point of view. The temperature evolution and flow of a helium heat slug along the conductor during the 15 MA plasma scenario is also analyzed. It appeared that sequential multiple burn cycles can be carried out without accumulation of temperature in the helium heat slug, essential for continuous energy production for a fusion power plant.

Development of a Two-Dimensional Simplified Tool for the Analysis of the Cooling of the ITER TF Winding Pack

The cooling of the ITER toroidal field (TF) coils winding pack is guaranteed by the circulation of supercritical helium (He) in seven hydraulic circuits corresponding to the Nb3Sn cable in conduit conductors, and in 74 channels devoted to the cooling of the stainless steel case supporting the winding pack. A tool entirely developed inside ANSYS with the APDL language has been created with the aim of computing the temperature distribution in the TF winding pack in different poloidal locations. The tool also allows the assessment of the He temperature during plasma operation in the case cooling channels. The considered heat load is the volumetric nuclear heating computed with the MCNP code in 32 poloidal segments of a TF coil. For each segment, a two-dimensional (2-D) finite-element model is built and a thermal analysis carried out by applying the corresponding heat load, whereas a 1-D thermal-hydraulic code solves the helium transport inside the structure. Steady-state analyses have been done considering the baseline pancake wound configuration. In a second step, a preliminary transient analysis has been carried out, considering the actual ramp up and ramp down of the nuclear heating.

Coupling Loss Characteristics of Nb3Sn CIC Conductor for CFETR CS Model Coil

An Nb 3 Sn cable-in-conduit (CIC) conductor with short twist pitch structure has been developed at Institute of Plasma Physics Chinese Academy of Sciences, Hefei, China, to meet the central solenoid model coil (CSMC) project of China fusion engineering test reactor (CFETR). 12 T magnetic field and 1.5 T/s field change are the most basic design target. The coupling loss characteristics of Nb 3 Sn CIC conductor is the key design requirement for CSMS because the deposited energy in magnets from ac losses can decrease the safe margin in the nuclear fusion device. In order to evaluate the coupling loss performance of the Nb 3 Sn CIC conductor, the ac loss measurement for Nb 3 Sn CIC conductor sample was carried out at SULTAN facility. This paper presents the specific structure design of the Nb 3 Sn CIC conductor and ac losses test results first. Then, the coupling losses characteristics are discussed with Multizones PArtial Shielding model and the key fit parameters are obtained from measurement results. Finally, two typical field cases are considered to evaluate the coupling loss energy of Nb 3 Sn CIC conductor for safe operation of the CFETR CSMC and the optimization of the cryogenic system.

DC performance measurement and assessment of Nb3Sn Cable-In-Conduit Conductor for CFETR CS model coil

The Central Solenoid Model Coil (CSMC) project of China Fusion Engineering Test Reactor (CFETR) was developed from 2014 at the Institute of Plasma and Physics Chinese Academy of Sciences (ASIPP) to verify the larger-scale superconducting magnet technology. The short twist pitch (STP) option was chosen for Nb3Sn Cable-In-Conduit Conductors (CICC) due to the stronger rigidity which can withstand the higher electromagnetic (EM) cycles effectively. Performance measurement of conductor short samples in SULTAN facility is the most important assessment method. This paper introduces the Nb3Sn strand parameters and CICC design firstly; then the preparation and measurement procedure of conductor short samples for current sharing temperature (Tcs) are described; subsequently, the Tcs vs. cycles, effective strain (εeff) with B×I and transition broadness characteristics compared with strands are assessed qualitatively by power law model. The assessment result shows that the DC performance vs. cycles of the STP design for Nb3Sn CICC is very stable and no degradation. The notable result gives the confidence that life period in the CSMC can be achieved reliably although the number of the EM cycles has been limited to below 10 000 in the measurement. The successful DC performance testing of Nb3Sn CICC will be an important milestone for CSMC project. Lastly, the performance assessment in coil is performed at the typical test waveform. The minimum temperature margin (Tmag) of 0.14K and electric field of 5 uV/m are achieved respectively. Some slow EM cycles is required to decrease the coupling loss deposition before coil testing.

Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability margin of magnets. In the last decade ITER has launched a campaign for characterization of several types of NbTi and Nb3Sn CICCs comprising quench tests with a singular sine wave fast magnetic field pulse and relatively small amplitude. The stability tests, performed in the SULTAN facility, were reproduced and analyzed using two codes: JackPot-AC/DC, an electromagnetic-thermal numerical model for CICCs, developed at the University of Twente (van Lanen and Nijhuis 2010 Cryogenics 50 139–148) and multi-constant-model (MCM) (Turck and Zani 2010 Cryogenics 50 443–9), an analytical model for CICCs coupling losses. The outputs of both codes were combined with thermal, hydraulic and electric analysis of superconducting cables to predict the minimum quench energy (MQE) (Bottura et al 2000 Cryogenics 40 617–26). The experimental AC loss results were used to calibrate the JackPot and MCM models and to reproduce the energy deposited in the cable during an MQE test. The agreement between experiments and models confirm a good comprehension of the various CICCs thermal and electromagnetic phenomena. The differences between the analytical MCM and numerical JackPot approaches are discussed. The results provide a good basis for further investigation of CICC stability under plasma scenario conditions using magnetic field pulses with lower ramp rate and higher amplitude.

Modeling of ITER TF cooling system through 2D thermal analyses and enthalpy balance

The winding pack of the ITER Toroidal Field (TF) coils is composed of 134 turns of Nb3Sn Cable in Conduit Conductor (CICCs) wound in 7 double pancakes and cooled by supercritical helium (He) at cryogenic temperature. The cooling of the Stainless Steel (SS) case supporting the winding pack is guaranteed by He circulation in 74 parallel channels.A 2D approach to compute the temperature distribution in the ITER TF winding pack is here proposed. The TF is divided in 32 poloidal segments, for each segment the corresponding 2D model is built and a thermal analysis is performed applying the corresponding nuclear heating computed with MCNP code considering the latest design updates, such as thickness increase of the blanket shield module. The Heat Transfer coefficient (HTC) of the He flowing in the CICC and in the cooling channels of the SS case is computed with Dittus Boelter correlation at the nominal inlet pressure of 6bar. The He is assumed to enter the coil at 4.5K in the lower terminal junction, and then the bulk temperature in all the CICCs in each of the 32 segments is calculated by means of enthalpy balance between segments, considering the actual direction of He circulation, i.e. clockwise or counter-clockwise in neighboring pancakes. The He properties needed to compute the HTC, such as viscosity, specific heat and thermal conductivity, are also varied using the same strategy. With these assumptions, He temperatures close to 5.7K are computed, due to the high values of nuclear heating (which is estimated as high as 21.58kW in the 18TF). In the paper, the methodology is presented and the results are discussed in detail. Further parametric analyses are also presented to show the impact of the inlet temperature and of the nuclear heating on the temperature distribution.

Design of new central solenoid for SST-1

The key role of central solenoid (CS) magnet of a Tokamak is for gas breakdown, ramp up and maintaining of plasma current. The magnetic flux change in CS along with other PF coils generates magnetic null and induces electric field in toroidal direction. The induced toroidal electric field accelerates the residual electrons which collide with the neutrals and an avalanche takes place which led to the net plasma in the vacuum vessel of a Tokamak. In order to maximize the CS volt-sec capability, the higher magnetic field with a greater magnetic flux linkage is necessary. In order to facilitate all these requirements of SST-1 a new superconducting CS has been designed for SST-1. The design of new central solenoid has two bases; first one is physics and second is smart engineering in limited bore diameter of ∼ 655 mm. The physics basis of the design includes volt-sec storage capacity of ∼ 0.8 volt-sec, magnetic field null around 0.2 m over major radius of 1.1 m and toroidal electric field of ∼ 0.3 volt/m. The engineering design of new CS consists of Nb3Sn cable in conduit conductor (CICC) of operating current of 14 kA @ 4.5 K at 6 T, consolidated winding pack, smart quench detection system, protection system, housing cryostat and conductor terminations and joint design. The winding pack consists of 576 numbers of turns distributed in four layers with 0.75 mm FRP tape soaked with cyanide Easter based epoxy resin turn insulation and 3 mm of ground insulation. The interlayer low resistance (∼1 nΩ) terminal praying hand joints at 14 kA at 4.5 K has been designed for making winding pack continuous. The total height of winding pack is 2500 mm. The stored energy of this winding pack is ∼ 3 MJ at 14 kA of operating current. The expected heat load at cryogenic temperature is ∼ 10 W per layer, which requires helium mass flow rate of 1.4 g/s at 1.4 bars @ 4.5 K. The typical diameter and height of housing cryostat are 650 mm and 2563 mm with 80 K shield respectively. The protection system consists of SS310 made array of dump resistor of 20 mΩ. The detail physics and engineering design of new superconducting CS of SST-1 will be discussed in this presentation.

Design and fabrication of indigenous 30 kA Nb3Sn CICC for fusion relevant superconducting magnet

“Magnet Technology Development Group” is engaged in focused research and development of indigenous fusion relevant superconducting magnet at Institute for Plasma Research in association with various R&D organizations. The fusion relevant superconducting magnet is under development using a cable in conduit conductor (CICC) with operating current of 30 kA at 12 T and 4.22 K. The 30 kA CICC has been designed in square cross-section (30 mm × 30 mm) consisting twisted Nb3Sn strands and copper strands as superconducting cable, SS316LN tubes as jacket material and SS304L foil as wrapping around the cabled strands. It has been designed on the basis of required critical design parameters, operation requirements and mechanical consideration during its fabrication. Cabling technology required for twisting of Nb3Sn and Copper strand in required configuration of cable is discussed in this paper. The effect of heat treatment on SS316LN jacket material as well as on Nb3Sn strands is mentioned in paper. 100 m long Nb3Sn based CICC is manufactured by pulled through technology on dedicated jacketing line. The manufacturing parameters and quality procedures for development of CICCs is successfully established and have been demonstrated with fabrication of 100 m Nb3Sn based CICC without any technical difficulties.

AC Loss Before and After Cyclic Mechanical Loading in the ITER RF CICCs

AC loss of Nb3Sn cable-in-conduit conductors (CICC) decreases after cyclic mechanical loading, but remains almost the same in case of Nb-Ti ones. It was observed during previous ITER CS, TF, and PF inserts tests in JAEA. The same phenomena have been detected on numerous ITER CS, TF, and PF short samples tested in the SULTAN facility. Experimental methods of ac loss investigation in JAEA-Naka and SULTAN were different, but the final results showed a good agreement. Analysis and numerical investigation of the present work were based on RF PF&TF CICCs SULTAN test results, theory of superconductivity, and our own physical and engineering understanding of evidence. Cyclic mechanical loading in both JAEA and SULTAN was initiated by electromagnetic force. In this paper, we attempt to explain why hysteresis, coupling, and mechanical loss decrease dramatically after cyclic mechanical loading and thermo-mechanical stress caused by WUCD (warm-up and cool-down) in Nb3Sn CICC.

Correlation between degradation and broadness of the transition in CICC

Cable in conduit conductor (CICC) performance is characterized in terms of relationships involving the electric field (E), voltage (V), temperature (T), current (I), magnetic field (B) and strain. Development of the electrical field in the V–T or V–I transitions in CICCs is exponential. These transitions plotted in the coordinates log E versus T or log E versus I look like straight lines. ITER Nb3Sn CICCs show degradation of properties versus load cycles that could be attributed to plastic deformation of the Nb3Sn strands or fracture of the superconducting filaments. The degradation is expressed in terms of the reduction of the current sharing temperature Tcs or critical current Ic, respectively. It was noticed long ago that degradation is accompanied by a significant broadening of the V–T or V–I transition, that looks like a change in the slope in the semi-log coordinate plot. This paper presents some systematic observations of correlations between the critical parameters and broadness of the transition in many CICCs. In most cases, the broadness of the transition seems to be a more sensitive indicator of the conductor damage even in cases where Tcs degradation is not clearly seen. Tcs degradation typically becomes obvious later in the cycling, especially after warm-up and following cool-down and more cycling. In some cases, a CICC manifests temporary or even a permanent growth of Tcs with load cycles, especially in the latest measurements of the CS conductors with short twist pitches. A possible mechanism of degradation that allows qualitative explanation of this phenomenon is discussed and is supported by the voltage measurements on the cable in the TFUS1 sample with the voltage taps penetrating the jacket to the cable.