Letter Sequence Other |
|---|
|
Results
Other: ML19247E427, ML19247E430, ML19247E431, ML19247E432, ML19247E433, ML19248C222, ML19350D685, ML19350D686, ML20004B075, ML20008F881, ML20033A354, ML20042C115, ML20042C117, ML20042C118, ML20042C120
|
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
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
esuuo SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 s street, Box 15830, Sacramento, California 95813; (916) 452-3211 March 17, 1982 A
Q
,s J,y
-, g. :
W DIRECTOR OF NUCLEAR REACTOR REGULATION h G m,29 fib '\\ g
'/
ATTN MR JOHN F STOLZ, CHIEF i'
n c?
G.
OPERATING REACTORS BRANCH 4 pK:f U S NUCLEAR REGULATORY COMMISSION tw (g"; /,
WASHINGTON DC 20555
&,g Dh,/\\,Y
(:y DOCKET NO. 50-312 REQUESTED INFORMATION ON NUREG-0737 ITEM II.K.2.13 In your letter of January 4,1982 you requested additional information so that review of NUREG-0737 item II.K.2.13, " Thermal-Mechanical Report-Effect of High Pressure Injection on Vessel Integrity From Small Break Loss of Coolant Accident With No Auxiliary Feedwater" could be completed.
The following requested information is provided to aid in this review.
1.
JperatingProcedureswhichinstructoperatorsfnthefeedandbleed operations and limit operations within the 100 subcooling limits following any small break loss of coolant accident and any significant cooldown transients.
The following procedures are appropriate to this question:
B.4 Plant Shutdown and Cooldown D.5 Loss of Reactor Coolant / Reactor Coolant Pressure D.14 Loss of Steam Generator Feed Copies of these procedures are attached as requested.
It should also be noted that the District is currently involved in a review of the Babcock and Wilcox Abnormal Transient Operating Guidelines (AT0G). When these guidelines are finalized, they will beccme the basis for new emergency procedures concerning the type of transients involved in NUREG-0737, item II.K.2.13.
Yoh
.s /
///
8203300223 820317 PDR ADOCK 05000312 P
PDR AN ELECIPIC SYSTEM SERVtNG YOEE THAN 600.000 IN THE HEART OF CAtlFORNIA
John F. Stolz March 17, 1982 2.
A basis for demonstrating that operations can control SB LOCA and overcooling transients.
This is demonstrated by the following facts. The District has instituted a training program in Mitigating Core Damage, which is
)
a 40-hour course directed toward controlling conditions such as the SB LOCA. Additionally, simulator training at the Babcock and Wilcox simulator includes training on small break accidents.
Approximately six hours of classroom and eight hours of simulator training specifically covers small break accidents.
- Finally, requalification training for licensed operators at Rancho Seco is an ongoing program designed to update and reenforce training and the use of procedures in small break accidents as well as other operational and accident conditions.
Ik t Acdr.a
{JohnJ.Mattimoe Assistant General Manager and Chief Engineer