Safety requirements/criteria analysis

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 ... 

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 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. 

Recently, however, limited use of best estimate plus uncertainty analysis methods has been undertaken. This is consistent with the international trend toward use of such methods. In this approach, more physically realistic models, assumptions, and plant data are used to yield analysis predictions that are more representative of expected behavior. This requires a corresponding detailed analysis of the uncertainties in the analysis and their effect on the calculated consequences. Typically, the probability of meeting a specific numerical safety criterion, such as a fuel centerline temperature limit, is evaluated together with the confidence limit that results from the uncertainty distributions associated with governing analysis parameters. The best estimate plus uncertainties approach addresses many of the problematic issues associated with conservative bounding analysis by... 

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 ... 

Shutdown systems have a separate acceptance criterion. Modem PHWRs have two independent, redundant and diverse shutdown systems with separate logic and reactivity devices from the control system and from each other [5]. Each system, on its own, must be capable of shutting the reactor down after any accident, independently of the mitigation provided by the normal reactivity control devices. In general, two diverse trip parameters are required on each shutdown system for each accident over the range of operating conditions (unless it is impracticable or detrimental to safety to provide dual parameter coverage). As a result, it is not required to perform analysis of either transients or accidents without shutdown [6],... 

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. 

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