[en] LANL portal monitor was a modification of a previously installed (permanent) unattended monitoring system (UMS). Modifications to the UMS to make the portal were sometimes based on mistaken assumptions about exercise-specific installation and access. Philosophical approach to real-time portal differs in some areas from UMS.

[en] The Dual Slab Verification Detector (DSVD) has been developed, built, and characterized by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of 3He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. By performing DSVD measurements at several different locations around the outer surface of the DUC, a signature 'fingerprint' can be established for each DUC based on the neutron flux emanating from inside the dry storage cask. The neutron fingerprint for each individual DUC will be dependent upon the spatial distribution of nuclear material within the cask, thus making it sensitive to the removal of a certain amount of material from the cask. An initial set of DSVD measurements have been performed on the first set of dry storage casks that have been loaded with canisters of spent fuel and moved onto the dry storage pad to both establish an initial fingerprint for these casks as well as to quantify systematic uncertainties associated with these measurements. The results from these measurements will be presented and compared with the expected results that were determined based on MCNPX simulations of the dry storage facility. The ability to safeguard spent nuclear fuel is strongly dependent on the technical capabilities of establishing and maintaining continuity of knowledge (COK) of the spent fuel as it is released from the reactor core and either reprocessed or packaged and stored at a storage facility. While the maintenance of COK is often done using continuous containment and surveillance (C/S) on the spent fuel, it is important that the measurement capabilities exist to re-establish the COK in the event of a significant gap in the continuous CIS by performing measurements that independently confirm the presence and content of Plutonium (Pu) in the spent fuel. The types of non-destructive assay (NDA) measurements that can be performed on the spent fuel are strongly dependent on the type of spent fuel that is being safeguarded as well as the location in which the spent fuel is being stored. The BN-350 Spent Fuel Disposition Project was initiated to improve the safeguards and security of the spent nuclear fuel from the BN-350 fast-breeder reactor and was developed cooperatively to meet the requirements of the International Atomic Energy Agency (IAEA) as well as the terms of the 1993 CTR and MPC and A Implementing Agreements. The unique characteristics of fuel from the BN-350 fast-breeder reactor have allowed for the development of an integrated safeguards measurement program to inventory, monitor, and if necessary, re-verify Pu content of the spent fuel throughout the lifetime of the project. This approach includes the development of a safeguards measurement program to establish and maintain the COK on the spent fuel during the repackaging and eventual relocation of the spent-fuel assemblies to a long-term storage site. As part of the safeguards measurement program, the Pu content of every spent-fuel assembly from the BN-350 reactor was directly measured and characterized while the spent-fuel assemblies were being stored in the spent-fuel pond at the BN-350 facility using the Spent Fuel Coincidence Counter (SFCC). Upon completion of the initial inventory of the Pu content of the individual spent-fuel assemblies, the assemblies were repackaged into welded steel canisters that were filled with inert argon gas and held either four or six individual spent-fuel assemblies depending on the type of assembly that was being packaged. This repackaging of the spent-fuel assemblies was performed in order to improve the stability of the spent-fuel assemblies for long-term storage and increase the proliferation resistance of the spent fuel. To maintain the capability of verifying the presence of the spent-fuel assemblies inside the welded steel canisters, measurements were performed on the canisters using the Spent Fuel Attribute Monitor (SPAM), which was a neutron coincidence counter very similar in design to the SFCC, but with a larger operational volume to accommodate the canisters. The analysis of the neutron coincidence data from the measurements of the welded steel canisters using SP AM yielded a Canister Attribute (CA) that was shown to be proportional to the total amount of Pu mass that was present inside the canister and comparable to the CA values that were calculated based on the previous SFCC measurements. The next step in securing the spent fuel from the BN-350 reactor is the packaging of the steel canisters into Dual-Use Casks (DUCs) and the transportation of the DUCs to a remote dry storage site.

[en] The filling of reactor spent fuel pools, the need for interim storage, the desire on the part of many nations to handle long-term storage of nuclear wastes, including uranium and plutonium from spent fuel, and the advent of international fuel cycle plans that promote international movement of spent fuel demand more comprehensive and thorough safeguards combined with upgraded physical protection of spent fuel. The bodies that regulate, license, and inspect spent fuel while stored in a reactor spent fuel pool, in interim storage, in transit, and at receipt at a reprocessing plant will need to develop these new safeguards and physical protection methods. The International Atomic Energy Agency (IAEA) needs better tools and approaches to safeguarding the spent fuel accumulating globally at nuclear power plants. The IAEA also needs to be better able to verify the isotopic composition of spent fuel arriving at a reprocessing plant. The international fuel cycle plans propose the transport of spent fuel across international boundaries. Hence, there needs to be improvements in safeguarding and especially physical protection and security of spent fuel as it moves across State borders and over the ocean from reactors to international fuel cycle centers. (authors)

[en] Japan Atomic Energy Agency (JAEA) and Los Alamos National Laboratory (LANL) jointly developed the Epithermal Neutron Multiplicity Counter (ENMC). A measurement test was performed using the standard samples and its results showed that ENMC achieves high measurement accuracy (approx. 0.4%) under the optimum conditions which is the same level with destructive assay. However, in the practical measurement for nuclear material accountancy or safeguards, bias is observed due to the variation of the sample properties. With this recognition, JAEA jointly with LANL conducted simulations for identifying the causes of this bias and its correction method. The results showed that the dominant cause of the bias is difference in sample density and this bias can be mitigated by correcting neutron counting efficiency. JAEA and LANL evaluated the applicability of the two ways for correcting the counting efficiency; either applying simulation data or real measurement data. It was concluded that as application of real measurement data has certain difficulties, the simulation data is more reasonable to be applied for the correction of the counting efficiency. (author)

[en] The Unattended and Remote Monitoring (UNARM) system at the BN-350 fast breeder reactor facility in Aktau, Kazakhstan continues to provide safeguards monitoring data as the spent fuel disposition project transitions from wet fuel storage to dry storage casks. Qualitative data from the initial cask loading procedures has been released by the International Atomic Energy Agency (IAEA) and is presented here for the first time. The BN-350 fast breeder reactor in Aktau, Kazakhstan, operated as a plutonium-producing facility from 1973 W1til 1999. Kazakhstan signed the Nonproliferation Treaty (NPT) in February 1994, and shortly afterwards the IAEA began safeguarding the reactor facility and its nuclear material. Slnce the cessation of reactor operations ten years ago, the chief proliferation concern has been the spent fuel assemblies stored in the pond on-site. By 2002, all fuel assemblies in wet storage had been repackaged into proliferation-resistant canisters. From the beginning, the IAEA's safeguards campaign at the BN-350 included a constant unattended sensor presence in the form of UNARM which monitors nuclear material activities at the facility in the absence of inspector presence. The UNARM equipment at the BN-350 was designed to be modular and extensible, allowing the system to adapt as the safeguards requirements change. This has been particularly important at the BN-350 due to the prolonged wet storage phase of the project. The primary function of the BN-350 UNARM system is to provide the IAEA with an independent, radiation-centric Containment and Surveillance (C and S) layer in addition to the standard seals and video systems. The UNARM system has provided continuous Continuity of Knowledge (COK) data for the BN-350's nuclear material storage areas in order to ensure the validity of the attended measurements during the lifetime of the project. The first of these attended measurements was characterization of the spent fuel assemblies. This characterization utilized the Spent Fuel Coincidence Counter (SFCC) instrument (ref) to measure neutron multiplicity and calculate Pu mass. These calculated masses were then compared to modeling simulation of the assemblies as well as declarations from the facility in order to baseline the amount of material under IAEA safeguards (ref). Once the baseline was established, bundles of four or six assemblies were repackaged into proliferati n-resistant canisters. This provided an additional physical barrier to material diversion and provided further protection by choosing assemblies for each canister so that the overall dose rate met self-protection requirements. Each of the canisters were then characterized using a similar technique to the SFCC, but with the Spent Fuel Attribute Monitor (SPAM) instnunent (ref). The data from these measurements were then used to calculate an attribute proportional to the total Pu mass in each canister. This attribute was then compared to the know Pu mass of each assembly in order to verify the accuracy of SPAM. In the event that COK is lost, the SPAM detector remains positioned to reverify Pu content of individual canisters without requiring the canister to be opened.

[en] The Dual Slab Verification Detector (DSVD) has been developed and built by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of ³He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. The DSVD will be used to perform measurements of the neutron flux emanating from inside the dry storage cask at several locations around each cask to establish a neutron 'fingerprint' that is sensitive to the contents of the cask. The sensitivity of the fingerprinting technique to the removal of specific amount of nuclear material from the cask is determined by the characteristics of the detector that is used to perform the measurements, the characteristics of the spent fuel being measured, and systematic uncertainties that are associated with the dry storage scenario. MCNPX calculations of the BN-350 dry storage asks and layout have shown that the neutron fingerprint verification technique using measurements from the DSVD would be sensitive to both the amount and location of material that is present within an individual cask. To confirm the performance of the neutron fingerprint technique in verifying the presence of BN-350 spent fuel in dry storage, an initial series of measurements have been performed to test the performance and characteristics of the DSVD. Results of these measurements will be presented and compared with MCNPX results.

[en] Summary of this presentation: (1) EFC instrument design for ²³⁵U verification measurements issued to EURATOM to issue a call for commercial tender; (2) Achieved a fast (Cd mode) measurement with less than 2% relative uncertainty in the doubles neutron counting rate in 10 minutes using a standard source strength; (3) Assay time in fast mode consistent with the needs of an inspector; (4) Extended to realistic calibration range for modern fuel designs - Relatively insensitive to gadolinia content for fuel designs with up to 32 burnable poison rods and 15 wt % gadolinia concentration, which is a realistic maximum for modern PWR fuel; (5) Improved performance over the standard thermal neutron collar with greater than twice the efficiency of the original design; (6) Novel tube pattern to reduce the impact of accidental pile-up; and (7) Joint test of prototype unit - EURATOM-LANL.

[en] PEP-II is an e⁺e^- B-Factory Collider located at SLAC operating at the Upsilon 4S resonance. PEP-II has delivered, over the past five years, an integrated luminosity to the BaBar detector of over 131 fb^-1 and has reached a luminosity of 6.11 x 10³⁶/cm²/s. Steady progress is being made in reaching higher luminosity. The goal over the next several years is to reach a luminosity of at least 2 x 10³⁴/cm²/s. The accelerator physics issues being addressed in PEP-II to reach this goal include the electron cloud instability, beam-beam effects, parasitic beam-beam effects, high RF beam loading, shorter bunches, lower beta y*, interaction region operation, and coupling control. A view of the PEP-II tunnel is shown in Figure 1. The present parameters of the PEP-II B-Factory are shown in Table 1 compared to the design. The present peak luminosity is 204% of design and the best integrated luminosity per month is 7.4 fb^-1 that is 225% of design. The best luminosity per month is shown in Figure 2. The integrated luminosity over a month is shown in Figure 3 and the total integrated luminosity in shown in Figure 4. The progress in luminosity has come from correcting the orbits, adding specific orbit bumps to correct coupling and dispersion issues, lowering the beta y* in the LER, and moving the fractional horizontal tunes in both rings to just above the half integer (<0.52)

[en] In early 2009, preliminary excavation work has begun in preparation for the construction of the New Safe Confinement (NSC) at the Chernobyl Nuclear Power Plant (ChNPP) in Ukraine. The NSC is the structure that will replace the present containment structure and will confine the radioactive remains of the ChNPP Unit-4 reactor for the next 100 years. It is expected that special nuclear material (SNM) that was ejected from the Unit-4 reactor during the accident in 1986 could be uncovered and would therefore need to be safeguarded. ChNPP requested the assistance of the United States Department of Energy/National Nuclear Security Administration (NNSA) with developing a new non-destructive assay (NDA) system that is capable of assaying radioactive debris stored in 55-gallon drums. The design of the system has to be tailored to the unique circumstances and work processes at the NSC construction site and the ChNPP. This paper describes the Chernobyl Drum Assay System (CDAS), the solution devised by Los Alamos National Laboratory, Sonalysts Inc., and the ChNPP, under NNSA's International Safeguards and Engagement Program (INSEP). The neutron counter measures the spontaneous fission neutrons from the ²³⁸U, ²⁴⁰Pu, ²⁴⁴Cm in a waste drum and estimates the mass contents of the SNMs in the drum by using of isotopic compositions determined by fuel burnup. The preliminary evaluation on overall measurement uncertainty shows that the system meets design performance requirements imposed by the facility.

[en] Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd₂O₃) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available ²⁴¹AmLi (α,n) interrogation source strength of 5.7×10⁴ s⁻¹. Furthermore, the calibration range of the new collar has been extended to verify ²³⁵U content in variable PWR fuel designs in the presence of up to 32 gadolinia burnable poison rods with Gd concentrations of up to 12 wt%. Monte Carlo calculations predict that the EFC has a lower statistical uncertainty for measurements performed in the fast neutron mode than its predecessor neutron collar design. This paper describes the physics design and calculated performance characteristics of the EFC. The Gd response is presented over a realistic range for modern PWR fuel designs