Preliminary Safety Analysis Research Articles

Abstract P5-22-20: Oncologic safety of endoscopic assisted breast surgery compared with conventional breast surgery: An analysis of 1295 primary operable breast cancer patients from single institute

Abstract Background:Endoscopy-assisted breast surgery (EABS) performed through minimal axillary and/or periareolar incisions is a possible alternative to open surgery for certain patients with breast cancer. In this study, we report the oncologic safety results of EABS compared with conventional breast surgery (CBS). Methods: The medical records of patients who underwent EABS for breast cancer during the period June 2010 to April 2017 were collected from the EABS database at Changhua Christian Hospital, a tertiary medical center in central Taiwan. Data on clinicopathologic characteristics, type of surgery, method of breast reconstruction, margin involvement, locoregional recurrence, distant metastasis and overall survival were collected and compared to another cohort of patients, who received CBS at the same hospital, to determine the effectiveness and oncologic safety of EABS. Results: A total of 1295 patients were enrolled in current study, including 214 patients receiving EABS and 1081 patients underwent CBS. The mean age of them were 50.4 ± 9.4 years-old in EABS group, and 52.5 ± 11.3 in CBS group (P=0.01). Patients who received EABS were associated with more early stage breast cancer than patients with CBS group (P<0.01). The margin involved rate was 3.3% in EABS group, and 6.1% in CBS group (P=0.11). During a median follow-up of 57.7 ± 24.7 months, the locoregional recurrence was 3.7% in EABS group, and 7.3% in CBS group (P=0.06). The distant metastasis rate was 2.8% in EABS group and 10.3% in CBS group (P=0.001). The overall survival rate was 99.5% in EABS group, and 93.8% in CBS group. Comparison of endoscopic assisted breast surgery (EABS) and conventaional breast surgery (CBS) EABSCBSP value N=214 (%)N=1081 (%) Pathology tumor size (cm)2.2 ± 1.62.3 ± 1.70.414Operation method Mastectomy141 (65.9)530 (49.0)<0.001Breast conserving surgery73 (34.1)551 (51.0) Lymph node surgery SLNB only149 (69.6)534 (49.4)<0.001SLNB then ALND40 (18.7)244 (22.6) ALND7 (3.3)259 (24.0) Not done18 (8.4)44 (4.1) Breast reconstruction Yes108 (50.5)930 (86.0)<0.001No106 (49.5)151 (14.0) Lymph node metastasis Yes49 (24.7)375 (35.0).0.005No149 (75.3)797 (65.0) Stage 055 (25.7)160 (14.8)<0.001I67 (31.3)340 (31.5) II84(39.2)439(40.7) III8(3.8)136(12.6) Grade I35 (17.6)178 (17.2)0.365II118 (59.3)566 (54.8) III46 (23.1)288 (27.9) Margin status Involved7 (3.3)66 (6.1)0.110Not-involved207 (96.7)1015 (93.9) Locoregional recurrence Yes8 (3.7)79 (7.3)0.057No206 (96.3)1002 (92.7) Distance metastasis Yes6 (2.8)111 (10.3)0.001No208 (97.2)970 (89.7) Survival Yes213 (99.5)1014 (93.8)0.001No1 (0.5)67 (6.2) SLNB: sentinel lymph node biopsy, ALND: axillary lymph node dissection. Conclusion: The preliminary oncologic safety analysis from our institution showed that EABS is a safe procedure and results in low margin involved rate, and no increase of locoregional recurrence, distant metastasis or mortality compared with conventional breast surgery. However, this study might be biased due to its retrospective nature and possible selection bias.Ongoing case controlled comparison study will be performed to further consolidate the oncologic safety of endoscopic assisted breast surgery. Citation Format: Lai H-W, Chen S-T, Chen D-R, Kuo S-J. Oncologic safety of endoscopic assisted breast surgery compared with conventional breast surgery: An analysis of 1295 primary operable breast cancer patients from single institute [abstract]. In: Proceedings of the 2017 San Antonio Breast Cancer Symposium; 2017 Dec 5-9; San Antonio, TX. Philadelphia (PA): AACR; Cancer Res 2018;78(4 Suppl):Abstract nr P5-22-20.

Analysis of hydrogen and dust explosion after vacuum vessel rupture: Preliminary safety analysis of Korean fusion demonstration reactor using MELCOR

Hydrogen explosion is one of the most dangerous accident phenomena in both nuclear fusion and fission reactor. In the typical fusion reactor, tritium and deuterium are fuel elements for fusion reaction, and they exist in condensed state around the cryo-pump in the normal operation condition. In the accident situation, increasing heat transfer system temperature can mobilize lots of tritium in the vacuum vessel. And then, coolant can react with high-temperature structures generating hydrogen gas inside vacuum vessel. In addition, air ingress into plasma terminates fusion reaction producing activated dust on the first wall plasma facing surfaces. Through these series of process, mobilized tritium and dust aerosol can explode in the vacuum vessel and damage the containment building integrity, making a sort of source term leakage to environment through various system volumes. In this paper, preliminary safety analysis for hydrogen and dust explosion accident for assumed design of Korean fusion demonstration reactor is conducted. An MELCOR system analysis code is used to simulate this accident and parametric analysis to investigate effect on the accident scale and safety systems like detritiation system such as heating, ventilation, and air conditioning isolation system and pressure suppression system. As a result, aerosol distribution of each control volume and its release route are estimated, and final aerosol release to the environment is compared with release guideline for ITER. The amount of mobilized aerosol in demonstration reactor is evaluated conservatively using ITER normal operation condition.

Three-Dimensional (3D) Printers in Libraries: Perspective and Preliminary Safety Analysis

As an emerging technology, three-dimensional (3D) printing has gained much attention as a rapid prototyping and small-scale manufacturing technology around the world. In the changing scenario of library inclusion, Makerspaces are becoming a part of most public and academic libraries, and 3D printing is one of the technologies included in Makerspaces. Owing to the ease of availability and cost effectiveness, most libraries use fused-deposition-modeling-based 3D printers compatible with plastic printing materials, such as polylactic acid (PLA) and acrylonitrile butadiene styrene (ABS). During the printing, PLA and ABS emit ultrafine particles (UFPs) and volatile organic compounds (VOCs) that may deteriorate the indoor air quality. In this article, first, we have discussed the background of 3D printing, the most common technologies used for 3D printing and printing materials, its applications in chemical education and sciences, as well as 3D printing health and safety concerns. Second, we measured and analyz...

Preliminary safety analysis of LOCAs in one EU DEMO HCPB blanket module

Safety analysis for the design basis accident (DBA) is essential to support DEMO blanket concept design. It is necessary to study the pressure behaviour in the blanket and the connected systems during the loss of coolant accident (LOCA) in a blanket module, as well as the temperature evolution in the coolant flow and the associated structures. For the Helium Cooled Pebble Bed (HCPB) blanket concept (version 2014) three representative accidental sequences of LOCA have been simulated using system code MELCOR 1.8.6 for fusion. The LOCA is identified to be the failure of cooling channels in the stiffening grid, in the FW or in the breeder unit. Simulation results are discussed in this paper.

Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Core design and safety analysis

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor II (EBR-II) for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor technologies. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable.Core design characteristics, fuel cycle performance, and safety analysis of the FHR DR preconcept have been evaluated. The FHR DR core design features a negative or negligible void coefficient throughout a reactor operating cycle. Both single-batch (cartridge) and multiple-batch fuel cycles can be demonstrated in the FHR DR. The single-batch cycle length of the FHR DR core is estimated at between 12 and 18months, assuming the successful qualification of composite carbon (C/C) or silicon carbide (SiC/SiC) structural fuel block tie rod material. Fuel cycle performance of the FHR DR is similar to a high temperature gas-cooled reactor. Preliminary safety analysis of the FHR DR indicates that the reactor could be used to demonstrate the inherent safety characteristics of FHR designs.

Preliminary safety analysis of high-intensity interval training (HIIT) in persons with chronic stroke.

The purpose of this study was to assess safety via electrocardiographic (ECG), blood pressure (BP), heart rate (HR), and orthopedic responses to 3 different high-intensity interval training (HIIT) protocols in persons with stroke. Eighteen participants (10 male; 61.9 + 8.3 years of age; 5.8 + 4.2 years poststroke) completed a symptom-limited graded exercise test (GXT) with ECG monitoring to screen for eligibility and determine HR peak. The 3 HIIT protocols involved repeated 30 s bursts of treadmill walking at maximum speed alternated with rest periods of 30 s (P30), 1 min (P60), or 2 min (P120). Sessions were performed in random order and included 5 min warm up, 20 min HIIT, and 5 min cool down. Variables measured included ECG activity, BP, HR, signs and symptoms of cardiovascular intolerance, and orthopedic concerns. Generalized linear mixed models and Tukey-Kramer adjustment were used to compare protocols using p < 0.05. No signs or symptoms of cardiovascular intolerance, significant arrhythmias, ST segment changes, or orthopedic responses resulted in early termination of any HIIT session. HIIT elicited HRs in excess of 88% of measured HRpeak including 6 (P30), 8 (P60), and 2 (P120) participants eliciting a HR response above their GXT HRpeak. Both maximum BP and HR were significantly higher in P30 and P60 relative to P120. Preliminary data indicate that persons with chronic stroke who have been prescreened with an ECG stress test, a symptom-limited GXT, and a harness for fall protection may safely participate in HIIT, generating substantially higher HRs than what is seen in traditional moderate intensity training.

A Preliminary Safety Analysis for the Prototype Gen IV Sodium-Cooled Fast Reactor

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

A Core Design Study for a Small Modular Boiling Water Reactor with Long-Life Core

This paper presents the core design and performance characteristics of the Novel Modular Reactor (NMR-50), a 50-MW(electric) small modular reactor. NMR-50 is a boiling water reactor with natural-circulation cooling and two layers of passive safety systems that enable the reactor to withstand prolonged station blackout and loss of ultimate heat sink accidents. The main goal in the core design is to achieve a long-life core (~10 years) without refueling for deployment in remote sites. Through assembly design studies with the CASMO-4 lattice code and coupled neutronics and thermal-hydraulic core analyses with the PARCS and RELAP5 codes, a preliminary NMR-50 core design has been developed to meet the 10-year cycle length with an average fuel enrichment of 4.75 wt% and a maximum enrichment of 5.0 wt%. The calculated fuel temperature coefficient and coolant void coefficient provide adequate negative reactivity feedbacks. The maximum fuel linear power density throughout the 10-year burn cycle is 18.7 kW/m, and the minimum critical power ratio is 2.07, both of which meet the selected design limits with significant margins. Preliminary safety analyses using the RELAP5 code show that the core will remain covered during the entire transient procedure of two design-basis loss-of-coolant accidents. These results indicate that the targeted 10-year cycle length is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

Feasibility of weekly intraperitoneal versus intravenous paclitaxel therapy delivered from the day of radical surgery for gastric cancer: a preliminary safety analysis of the INPACT study, a randomized controlled trial

Peritoneal carcinomatosis is common after curative resection of gastric cancer. Intraperitoneal administration of paclitaxel (PTX) is known to control ovarian peritoneal metastases. Patients with either linitis plastica or T4 cancer with high risk of peritoneal metastasis or recurrence but whose cancer was considered resectable were preregistered. After their cancer had been confirmed intraoperatively as resectable, the patients were randomized into either group A (PTX at 60mg/m2 intraperitoneally on the day of surgery and on days 14, 21, 28, 42, 49, and 56) or group B (PTX at 80mg/m2 administered intravenously by the identical schedule) before being treated by evidence-based chemotherapy. The primary end point was the 2-year survival rate. Safety, the secondary end point, was also analyzed. The study has been registered as UMIN000002957. Of 177 preregistered patients, 83 underwent treatment (39 by intraperitoneal administration and 44 by intravenous administration). There was no difference in patient demographics between the two groups. The incidences of surgical complications were similar between the groups, except for transient bowel obstruction observed exclusively in group A. The relative dose intensity of PTX was 81.4% for group A and 76.3% for group B. There was one death due to pulmonary thrombosis and a case of anaphylaxis that led to termination of the protocol treatment (group B). Other adverse events were mild and manageable. Intraperitoneal administration of PTX from the day of gastrectomy did not result in a higher incidence of surgical complications and adverse reactions when compared with intravenous administration of PTX.

Using FRAPTRAN 1.5 code for safety evaluation of TVS-2006 fuel rod under transient conditions of VVER-1200 (AES-2006) reactor

This article shows the results of important thermal-mechanical parameters related to the TVS-2006 fuel rod design which were analyzed and evaluated by using FRAPTRAN1.5 code. Based on the data given in the Preliminary Safety Analysis Report (PSAR) of AES-2006 (Novovoronezh NPP-2 Power Unit No.1)and FRAP-T (Fuel Rod Analysis Program-Transient), FRAPTRAN1.5 calculations of TVS-2006 fuel rod behaviors in Loss of Coolant Accidents and Reactivity-Initiated Accidents conditions have been made. The calculated results related to safety criteria of fuel element and cladding temperatures, cladding stress and strain, fuel enthalpy, local oxide thickness, gap gas pressure, and elongation of fuel rod have been compared with the ones given in AES-2006 PSAR. A good agreement has been observed between AES-2006 PSAR andFRAPTRAN1.5 calculations.

2226 Feasibility of weekly intraperitoneal versus intravenous paclitaxel delivered from the day of radical surgery for gastric cancer: A preliminary safety analysis of INPACT study, a randomized controlled trial

2226 Feasibility of weekly intraperitoneal versus intravenous paclitaxel delivered from the day of radical surgery for gastric cancer: A preliminary safety analysis of INPACT study, a randomized controlled trial

Verification of TVS-2006 fuel rod design of VVER-AES2006 reactor under steady-state operating condition Using FRAPCON-3.5 code

The purpose of this paper is to discuss the independent verification of TVS-2006 fuel rod design used in VVER-AES2006 reactor (Novovoronezh NPP-2 Power, Unit 1), based on the acceptance criteria and the reference data given in the Preliminary Safety Analysis Report of the State Research, Design, Construction and Survey Institute “Atomenergoproekt” (PSAR) and the operation of VVER-1000 reactor. The calculations were performed using FRAPCON-3.5 code, including fuel temperature, cladding temperature, fission gas release, internal gas pressure, cladding stress and strain, fuel extension, fuel rod elongation, cladding creep rate, fuel swelling rate, cladding oxide thickness and hydrogen concentration. The results are compared with the calculated data using START-3 code in PSAR and the acceptance criteria required by Russian nuclear regulatory body. Despite some discrepancies, the results showed conformance with the calculated data given in the PSAR and meet the acceptance criteria.

Preliminary studies of a new accelerator-driven minor actinide burner in industrial scale

Pursuing high transmutation rate of minor actinide (MA), a preliminary conceptual design of a lead-bismuth (LBE) cooled accelerator-driven system (ADS) is proposed in this study. Parametric studies are performed to optimize the neutronics and thermal–hydraulics performances. The proton energy and axial position of the proton beam impact is investigated to obtain high neutron source efficiency and spallation neutron yield. The influences of MA/Pu mixing ratio and the ratio of pin pitch to diameter (P/D) are also optimized to control the burnup reactivity swing and the minimum coolant velocity for adequate cooling. To reduce the power peak, three kinds of power flattening techniques are adopted and compared. The results show that the inert matrix ratio zone loading method seems more versatile. Based on the analyses, an optimized three zone loading pattern is proposed for the 800MWth subcritical core. The total transmutation rate of MA is 328.8kg per effective full power year. Preliminary safety analyses based on the balance of power method (BOP) are performed and the results show that in the reactivity insertion and beam overpower transients, the core shows inherent safety, but the scram is necessary by cutting off the beam current to protect the core from possible damages caused by the loss of flow.

Preliminary Safety Analysis of Mixed Oxide Sphere-pac Driver Fuels in the CP-ESFR WH Core

Abstract Future sodium cooled fast reactors (SFRs) with Minor Actinide (MA) burning potential have to demonstrate an enhanced safety, minimal waste production and increased proliferation resistance in addition to their high economical potential. In this context, the PELGRIMM project, part of the European Framework Program 7 (FP7), investigates new MA-bearing oxide fuel developments for SFRs. Two fuel designs are studied: the standard pellet and the innovative sphere-pac fuel. Due to its favorable swelling behavior expected under irradiation and its dustless fabrication, the latter one appears to be a promising option. A special work package in PELGRIMM addresses the safety performance of this fuel under operational, design basis (DBC) and design extension conditions (DEC). For the preliminary safety assessment, the Working Horse (WH) core, designed and analyzed in the former CP-ESFR project was chosen. The SIMMER-III code was applied after implementing key features as the specific thermal conductivity of sphere-pac fuels. The feasibility of loading the core with MOX sphere-pac instead of pellet fuel, regarding operation and safety limitations, was first proven. Both fuel designs for the WH core at beginning of life and after irradiation were then compared at nominal and under unprotected loss of flow conditions. Results show a similar behavior of these fuels under irradiation, opening a perspective to the practical application of sphere-pac fuels in SFR systems.

Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO2–Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF.

Methodological approach to a safety analysis of radioactive waste disposal in rock salt: An example

A methodological approach was applied in preparing of the project “Preliminary Safety Analysis of the Gorleben Site” according to the German requirements for final disposal of heat generating radioactive waste. The German requirements and the approach are based on recent international documents from IAEA and OECD/NEA on safety assessment and the safety case. The approach included the description of the geological site and its future evolution coupled with a range of possible waste emplacement scenarios using site specific repository designs. The repository designs were developed to provide operational safety, long-term safety and retrievability/recoverability of the waste to comply with the German safety requirements. The German requirements require ongoing optimisation of repository designs. Analysis of the integrity, fluid flows and radiological consequences was done and the performance of the system was assessed. If the present assumptions are confirmed by future research and development, the designed repository system for drift disposal was assessed to be robust. Further optimization is possible in relation to gaseous radionuclide species for the drift disposal design and in relation to the technology of retrieval for the borehole disposal design. The site specific safety analysis identified important tasks for research and development which would have not been possible by a generic safety analysis. The methodology and safety concept can be applied to other salt rock sites and can be transferred partially to clay sites.

Development of a mouse-adapted live attenuated influenza virus that permits in vivo analysis of enhancements to the safety of live attenuated influenza virus vaccine.

The live attenuated influenza virus vaccine (LAIV) is preferentially recommended for use in persons 2 through 49 years of age but has not been approved for children under 2 or asthmatics due to safety concerns. Therefore, increasing safety is desirable. Here we describe a murine LAIV with reduced pathogenicity that retains lethality at high doses and further demonstrate that we can enhance safety in vivo through mutations within NS1. This model may permit preliminary safety analysis of improved LAIVs.

Preliminary thermal-hydraulic safety analysis of Tehran research reactor during fuel irradiation experiment

Following domestic fabrication of nuclear fuels in Iran, it is necessary to investigate fuel material behavior, fission gas release, fuel swelling, cladding material behavior and fuel integrity of domestic fuels at different burnup in a research reactor during irradiation. Currently, Tehran research reactor is the sole operating research reactor which can be used for fuel irradiation experiments in the country. In this regard, standard codes as well as developed complementary computer programs are applied to verify thermal-hydraulic performance of irradiating a domestic rod-type fuel assembly of natural UO2 pellets in Tehran research reactor core, which itself contains 20% enriched plate-type U3O8–Al fuels. Maximum temperatures of fuel, clad and coolant, onset of nucleate boiling, onset of flow instability and departure from nucleate boiling during irradiation experiment are investigated by subchannel analysis as indicators to verify the reactor core safe operation during the experiment. The results give the confidence that during this irradiation experiment, thermal-hydraulic steady state safety criteria of the mixed-core are satisfied and the fuel irradiation experiment does not induce any significant operational change.

Abstract DDT01-04: SGN-CD19A: A novel Anti-CD19 antibody drug conjugate

Abstract CD19, a member of the immunoglobulin superfamily, is a B-cell specific marker that is found on B cells as early as the pro-B cell stage. CD19 is maintained upon malignant transformation and is expressed in the majority of patients with B-lineage acute lymphocytic leukemia (ALL) and non-Hodgkin lymphoma (NHL). SGN CD19A is a novel antibody-drug conjugate (ADC) composed of a humanized anti-CD19 monoclonal antibody conjugated to the microtubule-disrupting agent monomethyl auristatin F (MMAF) via a maleimidocaproyl (mc) linker. Upon binding to CD19, SGN-CD19A internalizes and releases cys-mcMMAF, which binds to tubulin and induces G2/M arrest and apoptosis in the targeted cells. In mice bearing B-lineage ALL and lymphoma xenografts, SGN-CD19A induced dose-dependent tumor growth delay and extended survival. Two first-in-human, phase 1, dose-escalation studies are ongoing to investigate the safety, tolerability, pharmacokinetics (PK), and antitumor activity of SGN CD19A; one trial in adult and pediatric patients with relapsed or refractory (R/R) B cell ALL or highly aggressive B-cell lymphoma (CT.gov NCT01786096) and the second trial in adult patients with R/R B-cell NHL (CT.gov NCT01786135). Both trials employ a modified continual reassessment method (CRM) to estimate the dose-toxicity relationship and to estimate the maximum tolerated dose. Exploratory analyses include correlation of target (CD19) expression, target saturation, and apoptotic cellular markers with clinical response and clearance of leukemic blasts; dose and PK of SGN-CD19A will also be correlated with target saturation and intracellular levels of cys-mcMMAF. Early PK analyses indicate that plasma SGN-CD19A ADC generally increases with higher doses. Rapid clearance of ADC was observed at low doses in patients with leukemia, suggesting target-mediated drug disposition. Dose escalation of SGN-CD19A is ongoing in both clinical trials. In preliminary safety analyses, SGN-CD19A was generally tolerable at doses examined to date. Encouraging antitumor activity has been observed, with multiple patients achieving complete remissions in both trials. Citation Format: Tina M. Albertson, Lali Sandalic, Baiteng Zhao, Ana Kostic, Che-Leung Law. SGN-CD19A: A novel Anti-CD19 antibody drug conjugate. [abstract]. In: Proceedings of the 105th Annual Meeting of the American Association for Cancer Research; 2014 Apr 5-9; San Diego, CA. Philadelphia (PA): AACR; Cancer Res 2014;74(19 Suppl):Abstract nr DDT01-04. doi:10.1158/1538-7445.AM2014-DDT01-04