Safety analysis criteria

Safety analysis criteria Was the protocol for selecting projects, items of hardware, or operations for detailed analysis clearly defined and adequate Were appropriate hazard analysis techniques recommended or specified ... 

The CPT is responsible for integration and implementation of the core design and safety analyses, which includes responsibility for design specifications, verification that fuel received is within specifications, verification that core conditions lie within core design and safety analysis criteria, operating limits, physics ihd thermal-hydraulic data for operational assessment and operational data for core design and safety analyses. The CPT is also responsible for core performance analysis which includes estimated critical position predictions, shutdown reactivity predictions, control rod worth curves, operational anomaly analysis and observed performance summaries. 

Optimum design solutions in the creation of multi-barrier systems for longterm storage and final disposal of radioactive materials should be chosen based on probabilistic safety analysis for each viable option, supplementing the latter with the cost estimates for the implementation of the option, i.e., using the complex criterion risk-benefit. ... 

In order to facilitate the implementation of this nuclear safety design criterion DOE 5480.28, NATURAL PHENOMENA HAZARDS MITIGATION, specifies the requirements for each new and existing DOE facility. The evaluation criteria for this DOE Order are built around a graded approach. DOE 5480.28 specifies that an NPH analysis will be performed for all new and existing dOE faciI ities. Specifically ... 

The acceptance criterion for GSI 022 is that new plants shall minimize the consequences of inadvertent boron dilution events by meeting the intent of SRP Section 15.4.6. Specifically, when performing a safety analysis to evaluate the consequences of an inadvertent boron dilution, plant designers should consider (1) design limits for maximum RCS pressure and minimum DNBR, (2) moderate frequency events in conjunction with a single failure or operator error and their possible effects on fuel integrity and radiological dose calculations, (3) and time limits specified for each mode of plant operation, if operator action is required to terminate an inadvertent boron dilution. 

In 1967, E R. Farmer of the United Kingdom proposed that the probabilities as well as consequences of potential accidents need to be estimated to assess the associated risk. Farmer used 1-131 as a surrogate for consequences. By plotting the probability and consequence of each postulated accident, one could distinguish those with high risk from those with low risk. He proposed a boundary line as a criterion for acceptable risk. Farmer s work was a conceptual breakthrough in nuclear reactor safety analysis. Farmer takes full credit as the originator and pioneer of PRA. 

Non-compliance with the single failure criterion shall be exceptional, and shall be clearly justified in the safety analysis. 

Step 2 The safety classification of structures, systems and components reflects the internal postulated events and external events as set forth in the safety analysis of the plant (box (3) of Fig. 1). The definition of the defence in depth levels and barriers [2], the application of the single failure criterion and the assessment of the potential for common cause failures are identified in box (2) of Fig. 1 [19], bearing in mind the categorization of the facility. Next is the evaluation of the need for emergency procedures, both on and off the site. This is followed by identification of the internal events to be considered as a consequence of an external event or as contemporaneous to an external event, and therefore of the safety functions to be maintained in case of an external event (e.g. cooling of radioactive material, reactivity control, confinement). 

Criterion 42. Safety analysis of the plant design Anticipated operational occurrences (AOOs) and DBAs are evaluated in the safety analysis and include design of a physical protection system to protect against malevolent acts. 

Section 2.1 provided a discussion of design-basis accidents, as included in Chapter 15 of the Safety Analysis Report (SAR). For containments, the design must preclude exceedance of the 10 CFR 100 dose guidelines, given the most limiting accident evaluated in Chapter 15. Specifically, the requirements of 10 CFR 50, Appendix A, General Design Criterion 50 state ... 

The fourth criterion assesses the ability of the full complement of plant systems to perform their safety functions. By application of this criterion, the plant is shown to be Within the bounds of the existing safety envelope for SRS reactors. Further, this process aids in identifying specific areas for improving the documentation of the SRS safety analysis. 

As a part of the Unreviewed Safety Question (USQ) review, the proposed activity must be evaluated to the criteria set forth in the Technical Specifications and the Safety Analysis Report (SAR). Seismic concerns and designs are both covered by the Technical Specifications and the SAR and are addressed by the USQ process. The procedure further requires that the USQ evaluation be approved by DOE prior to implementation of the activity when a USQ is determined to exist. The requirement to use the USQ process applies to all systems thereby enveloping the requirements of restart criterion 3. 

The responsibility criterion also requires that Reactor Operations establish the necessary interface agreements with the SRL for needed support. The staff found that these interface agreements have been defined in SCS-RGP-89-0020, "Centralization of Reactor Safety Responsibility" (Reference 3). These agreements establish the responsibilities for the flow and exchange of various information that is needed by the various organizations to accomplish the core design and safety analysis. 

This restart criterion requires that WSRC demonstrate the structural safety margins stated in the Safety Analysis Report (SAR) are consistent with current plant configuration and will be maintained. These margins should be consistent with the national codes and standards pertaining to seismic (discussed in the seismic section of the SER) and other postulated design basis accidents. 

The staff s evaluation of the program elements addressing Restart Criterion 1 included a review of (1) Chapter 13 of the Safety Analysis Report (SAR), "Conduct of Operations" (2) WSRC Technical Specifications (TS), Category 5 (5.0, "Administrative Controls") and the Reactor Engineering Department (RED) draft of procedure RP 5.1101, "Conduct of Reactor Engineering Activities."... 

The staff noted that the SRS emergency notification list (i.e., DPSOP 287, Appendix G) and its draft replacement, 6Q12 Manual (Reference 31), which is located in the TSC Communications Center, identify affected program personnel who are called out for reactor area emergencies. This list indicates that the identified personnel have reactor safety analysis expertise and report to the TSC, which adequately addresses this restart criterion. 

For reactivity insertion accidents (RIAs), the pellet enthalpy criterion of 230 cal/g UO2 is taken. It is the same as for LWRs. For abnormal transients with reactivity insert over 1, the criterion is set as 170 cal/g, again taken from that of LWRs. However, this criterion has not been applied to the safety analysis of the Super LWR and Super FR because no transient is followed by reactivity insertion over 1. It should be considered in the future study whether the pellet enthalpy criterion for transients is necessary, as in LWRs, or not, as in sodium cooled reactors. 

This is a typical flow increasing transient. The demand of the main coolant flow rate is assumed to rise stepwise up to 138% of the rated flow as is assumed in the feedwater control system failure of Japanese ABWRs. Since increase in the core coolant flow rate is mild in ABWRs due to the large recirculation flow, the feed-water flow rate is assumed to increase stepwise. This assumption is too conservative for the Super LWR. The main coolant flow rate is gradually increased by the control system in the safety analysis. The calculation results are shown in Fig. 6.31. The reactor power increases with the flow rate due to water density feedback. A scram signal is released when the reactor power reaches 120% of the rated power. The maximum power is 124% while the criterion is 182%. The increase in the pressure is small. The sensitivity analysis is summarized in Table 6.15. 

Consequently, separate experiments for the determination of extraction efficiency are often not required. An expert statement based on the results of metabolism studies is sufficient in most cases. These statements should also refer to the extraction solvent used for the analysis of samples of supervised trials. Residue levels found in these trials are the criterion for GAP and the basis for the setting of MRLs. Even if a solvent with insufficient extraction efficiency is used for samples from supervised trials, the later choice of better solvents would not result in lower safety for the consumer. 

Structural analysis of the solid rocket case-grain system using experimentally determined propellant response properties may permit a complete description of the combined stresses and resultant deformations, but a statement expressing the ability of the propellant to withstand these stresses is also required. Such a statement, which relates the physical state at which failure occurs to some material parameters, is called a failure criterion. The criterion for failure permits a prediction of safety margins expected under motor operation and handling and defines the loading regimes where abnormal operations will occur with intolerable frequency. 

The analysis and evaluation problem involves modeling the alternatives in an appropriate fashion and then developing an evaluation criterion so that one can compare them to each other. Developing a suitable criterion is often a nearly impossible task. How does one compare the safety of two processes For the problem of reaction path synthesis, how does one compare alternative chemical routes to the same molecule when one cannot predict kinetics If it proves possible to develop a suitable evaluation function, then one must still be able to do the analysis and evaluation quickly. One is usually faced with an enormous number of alternatives in synthesis, and a trade off is necessary between analysis and evaluation speed and accuracy. 

Department of Energy systems analysis should specifically include safety, and it should be understood to be an overriding criterion. 

The existing analysis for the uncontrolled depressurisation of all steam generators is based on a scenario where neider main steam header separation nor steam generator isolation occurs, enabling all steam generators to blow down through the ruptured main steam line (i.e. the isolation valve on the ruptured line fails to close), however the faulted SG is automatically isolated on feed side by the logic. For this scenario, the analysis was finalised to include safety injection termination criterion for boron concentration. 

A decision regarding hfe extension is a multidisciplinary analysis. Expertise of design engineers and manufacturer, safety engineers, system analysts, material degradation and structural integrity experts, financial analysts and hinnan/organizational factors researchers is combined to arrive at usefiil life criterion. 

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