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NUREG 0933
Displaying 1 - 25 of 113
The objective of this task was to demonstrate by testing and analysis that the relief and safety valves, block valves, and associated piping in the reactor coolant system (RCS) were qualified for the full range of operating and accident conditions; …
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The objective of this task was to improve the reliability and capability of nuclear power plant systems for removing decay heat and achieving safe shutdown conditions following transients and under post-accident conditions. ITEM II.E.3.1: RELIABILITY OF …
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The objective of this task was to improve the reliability and capability of nuclear power plant containment structures to reduce the radiological consequences and risk to the public from design basis events and degraded-core and core-melt accidents. ITEM …
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The objective of this task was to evaluate whether existing requirements for valve testing provided adequate assurance of performance under design conditions. ITEM II.E.6.1: TEST ADEQUACY STUDY DESCRIPTION Historical Background The purpose of this TMI …
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The objective of this task was to increase the reliability and diversification of the electrical power supplies for certain safety-related equipment. ITEM II.G.1: POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES, AND LEVEL INDICATORS DESCRIPTION …
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DESCRIPTION The issue was raised after the occurrence of various incidents of water hammer that involved steam generator feedrings and piping, emergency core cooling systems, RHR systems, containment spray, service water, feedwater, and steam lines. The …
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DESCRIPTION This item was originally identified in NUREG-0371 [1] and was later declared a USI in NUREG-0510. [2] (See Item A-3 for further details.) CONCLUSION This item was RESOLVED and requirements were established. (See Item A-3 for further details.) …
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DESCRIPTION During the conduct of a large scale testing program for an advanced design BWR pressure suppression containment system (MARK III), new suppression pool hydrodynamic loads associated with a postulated LOCA were identified which had not been …
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DESCRIPTION During testing for an advanced BWR containment system design (MARK III), suppression pool hydrodynamic loads were identified which had not been considered in the original design of the MARK I containment system. To address this issue, a MARK I …
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DESCRIPTION As a result of the GE testing program for the MARK III pressure-suppression containment program, new containment loads associated with a postulated LOCA were identified in 1975 which had not been explicitly included in the original design of …
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DESCRIPTION Historical Background Since the issuance of Appendix J to 10 CFR Part 50 in February 1973, certain requirements of the appendix have been found to be conflicting, impractical for implementation, or subject to a variety of interpretations by …
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DESCRIPTION Historical Background The Class 1E power sources provide the electric power for the plant systems that are essential to reactor shutdown, containment isolation, reactor core cooling, containment heat removal or are otherwise essential in …
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DESCRIPTION Historical Background This issue is documented in NUREG-0371 [1] and addresses the adequacy of safety-related DC power supplies which was questioned by a nuclear consultant in a letter to the ACRS in April 1977 (NUREG-0305, [2] Attachment B). …
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DESCRIPTION This issue was raised in NUREG-0371 [1] following a series of events at operating plants that involved the offsite power systems. These events indicated that the reliability of the preferred source of emergency power might have been less than …
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DESCRIPTION Historical Background This issue deals with a concern for the availability of adequate recirculation cooling water following a LOCA when long-term recirculation of cooling water from the PWR containment sump, or the BWR RHR system suction …
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DESCRIPTION The complete loss of AC electrical power to the essential and nonessential switchgear buses in a nuclear power plant is referred to as a "Station Blackout." Because many safety systems required for reactor core decay heat removal are dependent …
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DESCRIPTION The design criteria and methods for the seismic qualification of mechanical and electrical equipment in nuclear power plants underwent significant changes during the course of the licensing of commercial nuclear power plants. Consequently, it …
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DESCRIPTION Following a LOCA in an LWR, combustible gases, principally hydrogen, may accumulate inside the primary reactor containment as a result of: (1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of the …
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DESCRIPTION Historical Background This NUREG-0471 [1] item concerned staff positions [2] BTP EISCB 18 and BTP RSB 6-11 which required physical locking out of electrical sources to specific MOVs in the ECCS. The existing staff positions established the …
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DESCRIPTION Some prototype electrical penetration failures have occurred to date. In addition, failures of low voltage penetration modules have occurred at a licensed facility. It was originally postulated that failures of the low voltage penetration …
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DESCRIPTION Historical Background The description of this item given in NUREG-0471 [1] is as follows: "This is an ACRS generic concern. Evaluation and approval is required of various aspects of the MARK III containment design which differs from the …
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DESCRIPTION The calculations of differential pressure that occur in containment subcompartments from a loss-of-coolant event require a complex fluid dynamic analysis to assure that the subcompartment design pressures are not exceeded. To check the various …
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DESCRIPTION The rationale for normal and postaccident containment cooling will be reviewed to determine the adequacy of the design requirements imposed on the containment ventilation systems. By reviewing typical designs, the staff will develop a basic …
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DESCRIPTION Test data from the Marviken containment tests were obtained for the purpose of validating containment pressure codes used for performing independent calculations related to licensing reviews. The Marviken data are containment pressure …
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DESCRIPTION The CONTEMPT computer code is used by the NRC staff to perform independent containment analyses. This NUREG-0471 [1] task involves the maintenance and revision of the CONTEMPT code to accommodate new containment designs or new problem areas as …
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Page Last Reviewed/Updated 3/1/2026
Disclaimer: Some of the formatting in NUREG-0933 may not be correct. We are currently working on fixing the formatting.