Letter Sequence Other |
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Results
Other: ML19247E427, ML19247E430, ML19247E431, ML19247E432, ML19247E433, ML19248C222, ML19350D685, ML19350D686, ML20004B075, ML20008F881, ML20033A354, ML20042C115, ML20042C117, ML20042C118, ML20042C120
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MONTHYEARML19248C2221980-10-29029 October 1980
Provides Insights on Overcooling Transients in Plants W/B&W Nsss.Frequency of Overcooling Transients Diminishes Prior to Commercial Operations.Summary of Overcooling Events Encl
Project stage: Other
ML19247E4321981-03-0303 March 1981
Provides fracture-mechanics Parametric Analysis of Overcooling Accident at Facility,In Response to Request
Project stage: Other
ML19247E4331981-04-10010 April 1981
Updates Re Safety Implications of Control Sys & Dynamic Characteristics of Nuclear Facilities.Pressurized Thermal Shock Phenomena & Control Sys Implications to Safety Have Not Been Analyzed
Project stage: Other
ML20042C1171981-04-11011 April 1981
Plant Shutdown & Cooldown, Revision 19
Project stage: Other
ML19247E4301981-04-21021 April 1981
Forwards Preliminary Assessment of Frequency of Excessive Cooldown Events Challenging Vessel Integrity,In Response to 810417 Request
Project stage: Other
ML19247E4311981-04-21021 April 1981
Forwards Modified Table 1 to 810421 Memo.Further Consideration of B&W Two Bus Design W/Failure of Single Bus Has Led to Revision of HEP
Project stage: Other
ML19247E4271981-04-28028 April 1981
Forwards Preliminary Assessment of Thermal Shock to PWR Reactor Pressure Vessels,In Response to 810416 Memo
Project stage: Other
ML19350D6861981-04-30030 April 1981
EPRI Research on Properties of Irradiated Matl Pertinent to Overcooling Transients
Project stage: Other
ML20008F8811981-05-0505 May 1981
Informs That Data Presented in BAW-1511 P, Radiation- Induced Reduction & Charpy Upper-Shelf Energy of Reactor Vessel Welds, Is Applicable to Facility.Completion of Work in Other Phases of Program Will Justify Plant Operation
Project stage: Other
ML19350D6831981-05-12012 May 1981
Forwards Ltr Rept in Response to NRC 810429 Request for Addl Info Re Thermal Shock Issue.Also Forwards Summary of EPRI Programs
Project stage: Request
ML20004A4921981-05-12012 May 1981
Discusses Results of Preliminary Review of Thermal Shock to PWR Pressure Vessels.Reasonable Assurance Found That Plant Can Continue to Operate During Generic Evaluation
Project stage: Approval
ML19350D6851981-05-15015 May 1981
Ltr Rept on Reactor Vessel Brittle Fracture Concerns in B&W Operating Plants
Project stage: Other
ML20004B0751981-05-22022 May 1981
Notifies That Info Provided in Re Reactor Vessel Brittle Fracture Concerns in B&W Plants Is Applicable to Facility.Util Will Perform Further Fracture Analyses & Participate in Owners Group Activities Re Thermal Shock
Project stage: Other
ML20030A7701981-07-20020 July 1981
Notification of 810728,29 & 30 Meetings W/B&W,Westinghouse & C-E Owners Groups in Bethesda,Md to Discuss Thermal Shock to Reactor Pressure Vessel Issue
Project stage: Meeting
ML20030D6551981-08-21021 August 1981
Requests Addl Info Re Pressurized Thermal Shock to Reactor Pressure Vessels,Per Review of PWR Owners Group 810515 & Licensees 810522 Responses to NRC
Project stage: Approval
ML20033A3541981-10-30030 October 1981
Forwards NRC Evaluation of ORNL Rept on Pressurized Thermal Shock.Rept Does Not Change Previous NRC Conclusion That Probability of Occurrence of Severe Pressurized Overcooling Transients Too Low to Require Immediate Action
Project stage: Other
ML20042C1201982-03-0808 March 1982
Loss of Steam Generator Feed, Revision 16
Project stage: Other
ML20042C1181982-03-0808 March 1982
Loss of Reactor Coolant/Rcs Pressure, Revision 17
Project stage: Other
ML20042C1151982-03-17017 March 1982
Responds to NRC & Submits Addl Info Re NUREG-0737 Item II.K.2.13, Thermal Mechanical Rept. Procedures Re Plant Shutdown & Cooldown,Loss of Reactor Coolant/Reactor Coolant Pressure & Loss of Steam Generator Feed Encl
Project stage: Other
ML20054H6621982-06-21021 June 1982
Comments on Proposed Regulatory Position Re Pressurized Thermal Shock,As Discussed at 820609 Meeting W/Nrc.Programs Undertaken to Prevent Overcooling,Following 780320 Event, Also Discussed
Project stage: Meeting
1981-05-12
[Table View] |
Text
k e smur SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, Box 15830, Sacramento, California 95813; (916) 452-3211 N.
j May 22,1981
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n DIRECTOR OF NUCLEAR REACTOR REGULATION p
ATTENTION DARRELL G EISENHUT DIRECTOR i os DIVISION OF LICENSING U S NUCLEAR REGULATORY COMMISSION WASHINGTON DC 20555 DOCKET 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT NO 1 THERf1AL SH0CK TO REACTOR PRESSURE VESSEL The Sacramento Municipal Utility District has received your letter of April 20, 1981 requesting information on the effects of thermal shock to reactor pressure vessels. As Chairman of the Babcock & Wilcox Owners Regulatory Response Group, I responded on May 12, 1981 and provided you with a letter report on reactor vessel brittle fracture concerns in Babcock & Wilcox operating plants and a summary of the Electric Power Research Inrtitute programs pertaining to brittle frac-ture. This response provided the information requested in your letter and during meetings with NRC staff on March 31, 1981 and April 29, 1981.
We have determined that the information provided in my letter of May 12, 1981 is applicable to Rancho Seco Unit No.1 and support the conclusions presented therein. One of the conclusions was that the analysis presented in BAW-1648, " Thermal Mechanical Report - Effect of HPI on Vessel Integrity for Small Break LCCA Eveat with Extended Loss of Feedwater", represents a bounding cooldown event. We commit to perform further plant specific analyses to demonstrate that considerable time exists before there are any concerns over brittle fracture during such an event at Rancho Seco Unit No. 1.
These analyses will consider the actual borated water storage tank temperature of 800 and limiting weld locations. Also, we have previ-cuiy committed to install an upgraded auxiliary feedwater control system which will provide safety grade control of auxiliary feedwater, including ODl s
D 8105270\\90 P
aN E L E c;? t C SYSTEV iER7tNi MCRE THAN 600 000 IN THE HEART OF C All F O R N I A
2.
DARRELL G EISENHUT May 22, 1981 a means to prevent primary system overcooling. These controls will be installed during the first extended outage following equipment delivery in 1982. This system was described in our letter of November 17, 1980.
We further commit to participate in generic Owners Group activities which develop in the area of reactor vessel thermal shock. We feel confident that these actions will provide adequate justification for operation of Rancho Seco Unit No.1 for its design lifetime.
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I John. Mattimoe Asisstant General Manager and Chief Engineer i
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