Section
Resolution Product Category
Facilities Type
Related Technical Area
Image
NUREG 0933
Applied Filters
Displaying 1 - 25 of 61
TASK II.E.1: AUXILIARY FEEDWATER SYSTEM The objective of this task was to improve the reliability of the auxiliary feedwater (AFW) system. ITEM II.E.1.1: AUXILIARY FEEDWATER SYSTEM EVALUATION This item was clarified in NUREG-0737, [1] requirements were …
1
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 …
1
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 …
1
DISCUSSION On May 7, 1975, the NRC was informed by VEPCO that an asymmetric loading on the reactor vessel supports resulting from a postulated reactor coolant pipe rupture at a specific location (e.g., the vessel nozzle) had not been considered by W or …
1
DESCRIPTION Prior to 1978, operating experience with PWR steam generators was characterized by extensive corrosion and mechanically-induced degradation of the steam generator tubes, frequent plant shutdowns to repair primary-to- secondary leaks, and two …
1
DESCRIPTION During the course of the licensing action for North Anna Units 1 and 2, a number of questions were raised as to the potential for lamellar tearing and low fracture toughness of the steam generator and RCP support materials for these …
1
DESCRIPTION Safety-related equipment inside the containment of a nuclear power plant is qualified for the most severe accident conditions under which it is expected to function. In a PWR, this had been previously assumed to be the pressure and temperature …
1
DESCRIPTION Historical Background Several aspects of the MSLB analyses as currently provided by applicants and accepted by the NRC have been questioned. The concerns derive principally from Issues 1 and 15b of NUREG-0138. [1] Issue 1 in NUREG-0138 [2] …
1
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 …
1
DESCRIPTION Combinations of fabrication, stress, and environmental conditions have resulted in isolated instances of stress corrosion cracking of low pressure Schedule 10 Type 304 stainless steel piping systems. Although these systems are not part of a …
1
DESCRIPTION Historical Background Structural damage to the primary system, including the reactor pressure vessel and internals, associated piping and steam generator tubing in PWRs, can be caused by vibrations of sufficient magnitude. These vibrations can …
1
DESCRIPTION This NUREG-0471 [1] item is an ACRS generic concern which involves assessing the uncertainties in calculations of the control rod drop accident, including the choice of negative reactivity insertion rate due to a scram and the potential …
1
DESCRIPTION Historical Background This concern was identified by the ACRS as a result of their review [1] of LERs covering a period from 1976 through 1978 and, subsequently, a comprehensive review of the problem was completed which covered the LERs and …
1
DESCRIPTION Under the scope of the TMI Action Plan Item II.K.3 in NUREG-0660, [1] the NRC Bulletin and Orders Task Force conducted generic reviews of the loss-of-feedwater and small-break loss-of-coolant events on all operating reactors. These reviews …
1
DESCRIPTION Historical Background This is an ACRS concern raised by the Subcommittee on TMI-2 Implications in October 1979. This issue centers around the possibility of a breach in the reactor coolant system boundary caused by the failure of nonsafety …
1
DESCRIPTION Historical Background Cracking has occurred in PWR piping systems as a result of stress corrosion, vibratory and thermal fatigue, and dynamic loading. However, as of February 1981, no cracking had been experienced in the primary system piping …
1
DESCRIPTION Historical Background This issue effects all PWR-type reactors (Westinghouse, CE, B&W). The issue as described, [1] , [2] , [3] concerns postulated accidents resulting from a steam line break which consequentially results in a steam generator …
1
DESCRIPTION Historical Background Many PWRs have no positive means of detecting boron dilution during cold shutdown. [1] Some operations carried out during outages (e.g., steam generator decontamination) reduce the RCS volume thereby speeding up dilution. …
1
DESCRIPTION Background This issue addressed the high rate of reactor coolant pump (RCP) seal failures that challenge the makeup capacity of the ECCS in PWRs. At the time this issue was identified [1] in 1980, RCP seal failures in BWRs occurred at a …
1
DESCRIPTION Historical Background This issue was raised by the staff following a review [1] , [2] of operating events that indicated a significant number of ECCS spurious actuations, particularly, the four events that occurred at Davis-Besse during 1980. …
1
DESCRIPTION Pressurized water reactors are susceptible to certain types of hypothetical accidents that, under circumstances such as operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of …
1
DESCRIPTION Historical Background This issue corresponds to AEOD recommendation 4 highlighted in an AEOD memorandum [1] to NRR in December 1980. The AEOD recommendation resulted from a natural circulation cooldown event from full power at St. Lucie Unit …
1
DESCRIPTION Historical Background This issue was raised by AEOD [1] , [2] and involved isolation of the reactor coolant system charging and letdown system following a spurious safety injection transient at H. B. Robinson on January 29, 1981. Following the …
1
DESCRIPTION In AEOD/CO05, [1] AEOD identified potential safety problems concerning steam generator overfill due to control system failures and combined primary and secondary blowdown. As a result of discussions with the Commissioners and the EDO, NRR …
1
DESCRIPTION AEOD issued a memorandum [1] in which a potential safety issue involving combined primary and secondary system LOCAs was raised. The issue was discussed at Commission meetings on October 16, 1980 and on November 10, 1980. NRR informed AEOD of …
1
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.