Nuclear criticality safety evaluation

Independent Audits and Appraisals Operations should be reviewed frequently to ascertain that procedures are being followed and that process conditions have not been altered so as to affect the nuclear criticality safety evaluations. 

A cardinal principle in nuclear criticality safety evaluations is the validatioii of any computational method by correlation with pertinent criticality experiments to establi its bias. Commonly, the correlation is expressed in terms of values of keff calculated for the experiments, in which case the bias is the deviation of keff from unity. Both the bias and its uncertainty are generally functions of the material com--position and geometric arrangement. The uncertainty in the bias is an estirhatedf both the precision ofthe computational method and the accuracy of the experimental data. The uncertainty in the bias must allow for extension of the bias beyond the range of experimental data. Such an allowance is always required. Otherwise, there would be no need for calculations, because directly applicable experimental data would exist for the conditions being evaluated. 

In conclusion, benchmark experiments are of great importance in nuclear criticality safety evaluations. A huge debt is owed to those who have been so diligent in supplying data. However, many uncharted areas still remain. 

A Study bn a Nuclear Criticality Safety Evaluation Technique, T. Naito, /. Katakufa, M.Yokota(JAERI-Japan)... 

The three steps of the criticality safety evaluation—contingency analysis, limit determination, and control specification—are presented in a document generally referred to as a Nuclear Criticality Safety Evaluation (NCSE) (although some sites separate out the first step into a separate document referred to as a Nuclear Criticality Safety Assessment (NCSA)). Within a given organization or processing site, the structure and format of NCSA/NCSEs are usually strictly proscribed for consistency of development and ease of use. 

DOE 0 420.1 Facility Safety Requires fire hazard analysis and natural phenomena analysis for all facilities. For Hazard Category 2 or 3 nuclear facilities only, requires a criticality safety evaluation. Criticality Safety Analysis Fire Hazard Analysis Effects of natural phenomena hazards on facility systems, structures, or components (SSCs) included as part of safety analysis documented in the Safety Analysis Report (SAR), Basis for Interim Operation (BIO), or Auditable Safety Analysis (ASA). 

M. C. Brady and T. L. Sanders, A Validated Methodology for Evaluating Bumup Credit in Spent Fuel Casks, Proceedings of International Conference on Nuclear Criticality Safety, Christ Church, Oxford, United Kingdom, September 9-13, 1991. 

RESEARCH ON NUCLEAR CRITICALITY SAFETY AND ACCIDENT RISK EVALUATION FOR NUCLEAR FUEL CYCLE FACILITIES... 

Nuclear Criticality Safety and Accident Risk Evaluation for Nuclear Fuel Facilities 45... 

A CSA is required for operations that involve total amounts of fissile material in excess of the threshold amount for non-aqueous forms (700 g U) where formal criticality safety evaluations and controls are required [Philbin, 1998, Criticality Safety Supplement to the SNL ES H Manual, Applicability Section]. The CSA requires that process designs incorporate a double-contingency principle so that at least two unlikely, independent, and concurrent changes in process conditions must occur before an accidental nuclear criticality is possible. Table 6.4-3 is a qualitative summary of the double contingency analysis from the CSA for isotope processing [Mitchell and Romero, 1999 (draft). Section 5]. 

Any drop test should be conducted with the contents of the package simulated to its maximum weight. More than one drop may be necessary to evaluate all possible drop attitudes. It may also be necessary to test specific features of the package such as hinges or locks to ensure that containment, shielding and nuclear criticality safety are maintained. 

Hie effect of composite reflectors of steel and water on the reactivity of single cylinders of aqueous solution of low-enriched uranium has been studied. Ihe results are applicable in the evaluation of the nuclear criticality safety shipping containers and provide bases for the verification Of calculatlonal models. 

During operation of the plant, deviations from procedures and changes in process conditions that affect nuclear criticality safety must be investigated and acted upon promptly. Maximum use should be made of instrumentation for early detection of process changes so that adjustments can be made before the process control limits are approached. Before process or equipment changes are made they must be evaluated as to their effect oh the criticality safety of the plant. The depth of this review must be the same as that required for the initial evaluation. 

Critical Experiments and Analysis. The nuclear safety parameters of hydrogenous systems containing mixtures of uranium and plutonium will be based largely on calculations that are keyed to relatively few critical experiments. We must continue to emphasize the devel-ment of reliable computational techniques and evaluated differential cross-section data, and to select critical experiments that will provide for maximal benefits. Credence in computed data must be established by thorough reviews of possible errors and independent assessments by responsible authorities in nuclear criticality safety. 

Three headend processed that use soluble poison as a primary control to avoid a criticality have been developed at the Idaho Chemical Processing Plant. These processes are sirconium-allbyed fuel dissolution, stainless-steel-clad fuel dissolution, and Rover fiiel dissolution. The equipment Was evaluated for nuclear criticality safety at flow-sheet concentrations of nuclear poison using the KENO-II code and 16-group Hansen and Roach cross sections. Validation of code and cross sections by comparison with critical experiments involving soluble. 

The third aid was a film entitled "Safe NaiuUihg of Enriched Uranium." The film was made using scenes Of enriched uranium processing in the Oak Ridge Y-12 Plant, and animated descriptions of the parameters affecting criticality. Again, it was an oi rtunity to inform personnel of the conditions the nuclear criticality safety engineer must evaluate to establish safe controls. Also, it showed these controls in practice. 

This paper presents a sununaiy of evaluations made by the Babcock Wilcox Company (B W) in support of its customers. General Public Utilities. (GPU), to assess the potential for reciiticality of the damaged Three Mile Island Unit 2 Reactor (TMI-2) core and to provide a basis for recommendations to prevent die occurrence, of such an event. The TMI-2 incident occurred on March 28, 1979 on March 31, B W formed a Criticality Analysis Task Force composed of penoniiel from the Company s reactor physics unit and their nuclear criticality safety unit. The Task Force at maximum strength utiliaed ten individuals and completed all pertinent evaluations in two weeks advisory activity continued for several months. 

Two concurrent evaluations were made for all probable (and a few improbable) fliel configurations that could be identified. One was made by reactor physics personnel using mactor design codes, while the other was made by nuclear criticality safety specialists using criticality safety codes. In addition to evaluating specific fuel configurations, basic data Such as K-infinite vs fuel-to-water ratios were tabulated for different enrichments, fuel particle sizes, water temperatures, boron levels, etc. 

The objective of the event-independent part of the Emergency Operating Procedures (EOPs) is to provide means to evaluate and restore the plant nuclear safety. The concept is based on the premise that radiation release to the environment can be minimised if the barriers to activity release are protected (barriers of defence in depth). In order to accomplish this goal, a set of functions has been defined which are critical from the plant nuclear safety point of view. These are the Critical Safety Functions. To be able to evaluate the status of these functions. Status Trees have been designed, one per CSF. Once the state of the CSF is evaluated, based on their state and the rules of priority one can designate a Function Restoration Guideline to be implemented for restoring CSF (see Appendix 3). 

Calculational methods used to determine the neutron multiplication should be validated, preferably against applicable measured data (see Appendix VII). For irradiated nuclear fuel this vahdation should include comparison with measured radionuclide data. The results of this validation should be included in determining the uncertainties and biases normally associated with the calculated neutron multiplication. Fission product cross-sections can be important in criticality safety analyses for irradiated nuclear fuel. Fission product cross-section measurements and evaluations over broad energy ranges have not been emphasized to the extent that actinide cross-sections have. Therefore, the adequacy of fission product cross-sections used in the assessment should be considered and justified by the safety analyst. 

The mix tank (72 X 24 x 2 -in.) in the mixing room was the subject of the first experiment. As a check on the previous nuclear safety evaluation for the tank, an approach-to-critical procedure was employed when the 300 g/i plutonium solution was transferred to the tank. The critical mass of the tank was determined to be >20 kg of plutonium The tank is reflected with a 0.030-ln. Cd layer be-. tween the polyethylene reflector and the 0.25-in- stainless steel tank wall. 

As the nuclear ftiel recycle industry grows, the probability of a criticality accident will tend to grow proportionately unless current criticality safety control is improved. (Accident Probability < Probability per Plant times Number of Plants.) Optimum control of criticality safety requires an overview of the entire system. The relationships between the criticality safety philosophy, the criticality data, the use of the data, and various human Victors must be considered. Application of the foult tree analysis (FTA) methodology will permit identification of the important control foctors and their relationships, indication of what is required to reduce the accident probability per plant, and systematic evaluation of our current criticality safety position. FTA is not new but the application to criticality safety presented in this... 

The above practices have been used for many years as the basis of accepting moderation control for criticality safety in processing plutonium. These practices perhaps can provide a foundation for evaluating the use of moderation control in other parts of the nuclear fuel cycle. 

This imper describes a systematic approach for evaluating the criticality safety of an array of equipment items in a large nuclear facility. The approach has the greatest utility where complicating factors are present, such as the existence of a large number of units of varying sizes, sh es, orientations, contents, spacings, and variable reflectors. 

This paper describes a modular system of computer programs called SCALE. SCALE is an acronym for standardized case analysis for licensing evaluation. The system is being developed to perform criOcality safety, radiation shielding, and heat transfer analyses of shipping casks and other equipment and facilities associated with the nuclear fuel cycle. In this paper, emi aSls is placed on the criticality safety features of I ALE mid on the use of pin-lattice critical experiments in validating the SCALE system analytical sequences and data libraries. 

A systein whose criticality should be diedced is often too complex to End directly applicable experimental measurements. In such a case, computer codes are used to evaluate the criticality safety. The codes rmd data used must be validated by many benchmark calculations. A computer code system named JACS (Ref. 1) has been developed at the Japan Atomic Energy Research Institute (JAE for evaluating nuclear oiticality safety, and more than 700 benclunark calculatiotis, as riiown in Table 1, wm carried out to validate it. Tte computed keff s of the calculations were widely S ead (0.90 to 1.03), and in some cases absolute values of keff woe far from 1.0. Hence, problems are submitted, one on how to improve tim calculation method and another on how to waluate tire criticality safety of a system from the result computed by the cpde. A study has been carried on to answer the latter problem. ... 

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