ML19225A751
| ML19225A751 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/31/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Clayton F ALABAMA POWER CO. |
| References | |
| NUDOCS 7907200162 | |
| Download: ML19225A751 (2) | |
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4 UNITED STATES I
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NUCLEAR REGULATORY COMMISSION E
REGION 11 3
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ATLANTs, GEORGIA 30303 MY 3 1 1979 In Reply Refer To:
RII:JPO 50-348 50-364 Alabama Power Company Attn:
F. L. Clayton, Jr.
Executive Vice President Post Office Box 2641 Birmingham, Alabama 35291 Gentlemen:
The enclosed Bulletin 79-12 is forwarded to you for information. No written response is required.
If you desire additional ir. formation regarding this matter, please contact this office.
Sincerely, h
4 ames P. O'Reilly J
Director
Enclosure:
" WW
Ib4E Alabama Power Company 3, pp77 cc w/ encl:
A. R. Barton Executive Vice President Post Office Box 2641 Birmingham, Alabama 35291 C. Biddinger, Jr.
Manager-Corporate Quality Assurance Post Office Box 2641 Birmingham, Alabama 35291 H. O. Thrash Manager-Nuclear Generation Post Office Box 2641 Birmingham, Alabama 35291 H. G. Hairston, III Plant Manager Drawer 470 Ashford, Alabama 36312 R. E. Hollands, Jr.
QA Supervisor Post Office Box U Ashford, Alabama 36312 P
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UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 May 31, 1979 IE Bulletin No. 79-12 SHORT PERIOD SCRAMS AT BWR FACILITIES Summa ry:
Reactor scrams, resulting from periods of less than 5 seconds, have occurred recently at three BWR facilities.
In each case the scram was caused by high flux detected by the IRM neutron monitors during an approach to critical.
These events are similar in most respects to events which were previously described by IE Circular 77-07 (copy enclosed). The recent recurrences of this event inoicate nn apparent loss of effectiveness of the earlier Circular.
Issuance of this Bulletin is considered appropriate to further reduce the number of challenge; to the reactor protective system high IRM flux scram.
Description of Circumstances:
The following is a brief account of each event.
1.
Dyster Creek - On December 14, 1978, the reactor experienced a scram as contrcl 7.;ds were being withdrawn for approach to critical, following a scram from full power which had occurred about 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> earlier. The moderator temperature was 380 degrees F and the reactor pressure was 190 psig.
Because of the high zenon concentration the operators had not made ar. accurate estimate of the critical rod pattern. The cperator at the co'ntrols was using the SRM count rate, which had changed only slightly, (425 to 450 eps) to guide the approach.
Control rod 10-43 (first rod in Group 9) was being withdrawn in " notch override" to notch position 10, when the reactor became critical on an estimated 2.8 second period. The operator was attempting to reinsert the rod when the scram occurred. Failure of the " emergency rod in" switch to maintain contact, due to a bent switch stop, apparently contributed to the problem.
2.
Browns Ferry Unit 1 - On January 18, 1979, the reactor experienced aThe scram during the initial approach to critical following refueling.
operator was continuously withdrawing in " notch override" the first control rod in Group 3 (a high worth rod) because the SRM count rate had led him to believe that the reactor was very suberitical. A short reactor period, estimated at 5 seconds, was experienced.
The operator was attempting to reinsert control rods when the scram occurred.
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033 7006060168
IE Bulletin No. 79-12 May 31, 1979 Page 2 of 3 3.
Hatch Unit 1 - On January 31, 1979, the reactor erperienced a scram during an appror.ch to critical. Control rod 42-15 (fifth rod in Group 3) was being contin:ously withdrawn in " notch override" when the scram occurred, with a period of less then 5 seconds. The temperature was about 200 degrees F with effectively zero xenon.
As indicated above, these short period trips occurrei under a wide variety of In none of circumstances. They did have several things in common, however.
these cases was an accurate estimate of the critical position made prior to the approach to critical.
In each case a rod was being pulled in a high worth region. Finally, in each case the operator, believing that the reactor was very subcritical, was pulling a rod on continuous withdrawal.
Action to be Taken by Licensees:
For all GE BWR power reactor facilities with an operating license:
1.
Review and revise, as necessary, your operating procedures to ensure that an estimate of the critical rod pattern be made prior to each approach to critical. The method of estimating critical rod patterns should take into account all important reactivity vari. >1es (e.g.,
core Xenon, moderator
- 6 erature, etc.).
Where inaccuracies in critical rod pattern estimates are anticipated due 2.
to unusual conditions, such as high xenon, procedures should require that notch-step withdrawal be used well before the estimated critical position is reached and all SRM channel indicators are monitored so as to permit selection of the most significant data.
3.
Reviev and evaluate your control rod withdrawal sequences to assure that they minimize the notch worth of individual control rods, especially those Your review should withdrawn immediately at the point of criticality.
ensure that the following related criteria are also satisfied:
Special rod sequences should be considered for peak xenon a.
conditions.
b.
Provide cautions to the operators on situations which can result in high notch worth (e.g. first rod in a new group will usually exhibit high rod worth).
4.
Review and evaluate the operability of your " emergency rod in" switch to perform its f
- tion under prolonged severe use.
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IE Bulletin No. 79-12 May 31, 1979 Page 3 of 3 5.
Provide a description of how your reactor operator training program covers the considerations above (i.e., items 1 thru 3).
6.
Within 60 calendar days of the date of issue of this Bulletin, report in writing to the Director ef the appropriate NRC Regional Office, describing your action (s) taken, or to be taken, in response to each of the above items. A copy of your report should be sent to the United States Nuclear Regulatory Commission, Of fice of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C.
20555.
For all BWR facilities with a construction permit and all other power reactor facilities with an operating license or construction permit, this Bulletin is for information only and no written response is required.
Approved by GA0 B180225 (R0072); clearance erpires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.
Enclosures:
1.
IE Circular No. 77-07 2.
List of IE Bulletins Issued in Last Twelve Months l
IE Bulletin No. 79-12 Enclosure May 31, 1979 Page 1 of 3 LISTING OF IE BULLETINS ISSUED IN LAST IVELVE M0hTHS Bulletin Subject Date Issued Issued To No.
79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety OL or a CP Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-09 Failures of GE Type AK-2 4/17/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems OL or CP 79-08 Events Relevant to BWR 4/14/79 All BVR Power Reactor Reactors Identified During Facilities with an OL Three Mile Island Incident 79-07 Seisnic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 79-06B Review of Operational 4/14/79 All Combustion Ent neer-i Errors and System His-ing Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with r
Island Incident Operating Liceu.ee 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev 1)
Errors and System Mis-Power Reactor Facilities ali,inments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06A Review of Operational 4/14/10 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactors with an alignments Identified OL except B&W f acilities During the Three Mile Island Incident 409 006
Enclosure IE Bulletin No. 79-12 May 31, 1979 Page 2 of 3 LISTING OF IE BULLETidS ISSUED IN LAST WELVE M0hTHS Bulletin Subject Date Issued Issued To No.79-05A Nuclear Incident at 4/5/79 All B&W Power Peactor Three Mile Islan Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three Mile Island Facilities with an OL and CP 79-04 Incorrect Weights for 3/30/79 All Power Reactor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP Manufactured by Youngstown helding and Engineering Co.
79-02 Pipe Support Base Plate 3/2/70 All Power Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or CP 79-01 Environmental Qualification 2/8/79 All Power Reactor of Class IE Equipment Facilities with an OL or CP 78-14 Deterioration of Buna-N 12/19/78 All GE BVR facilities Component In ASCO with an OL or CP Solenoids 78-13 Failures in Source Heads 10/27/78 All general and of Kay-Ray, Inc., Gauges specific licensees Models 7050, 7050B, 7051, with the subject 7051B, 7060, 7060B, 7061 Kay-Ray, Inc.
and 7061B gauges78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 409 007
Enclosure IE Bulletin No. 79-12 Page 3 of 3 May 31, 1979 LISTING OF IE BULLETINS ISSUED IN LAST IVELVE M0hTHS Bulletin Subject Date Issued Issued To No.
78-12 Atypical Weld Material 9/29/78 All Power Reactor Facilities with an in Reactor Pressure Vessel Welds OL or CP 78-11 Examination of Mark I 7/21/78 BWR Power Reactor Containment Torus Welds Facilities for action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monticello and Vermont Yankee 78-10 Bergen-Paterson hydraulic 6/27/78 All BWR Power Reactor Shock Suppressor Accurulator Facilities with an OL or CP Spring Coils 78-09 BWR Dryvell Leakage Paths 6/14/79 All BWR Power Reactor Associated with Inadequate Facilities with an OL or CP Drywell Closures 78-08 Radiation Levels from fuel 6/12/78 All Power and Research Elcment Transfer Tubes Reactor Facilities with a Fuel Element transfer tube and an OL 78-07 Protection afforded by 6/12/78 All Power Reactor Air-Line Respirators and Facilities with aa OL, all class E and F Supplied-Air Hoods Research Reactors with an OL, all Fuel Cycle Facilities with an OL, and all Priority 1 Material Licensees 78-06 Defective Cutler-Hammer 5/31/78 All Power Reactor Type M Relays with DC Coils Facilities with an OL or CP nc rna lt u -)
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