ML19225A793
| ML19225A793 | |
| Person / Time | |
|---|---|
| Site: | Atlantic Nuclear Power Plant  |
| Issue date: | 05/31/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Collier A OFFSHORE POWER SYSTEMS (SUBS. OF WESTINGHOUSE ELECTRI |
| References | |
| NUDOCS 7907200257 | |
| Download: ML19225A793 (1) | |
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j gy 3. m a In Reply Refer To:
RII:JP0 50-437 Offshore Power Systems Attn:
A. R. Collier, President P. O. Box 8000 Jacksonville, Florida 32211 Gentlemen:
No written The enclosed Bulletin 79-12 is forwarded to you for inforr.ation.
response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, k
5 4 ames P. O'Reilly J
Director
Enclosure:
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366 202 7 907 200 p5-)
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND EhTORCEMENT WASHINGTON, D.C.
20555 May 31, 1979 IE Bulletin No. 79-12 SHORT PERIOD f, CRAMS AT BVR FACILITIES Suma 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 meast respects to events which were previously described by IE Circular 77-37 (copy encloseo). The recent recurrenc s of this event indicate an apparent loss of effectiveness of the earlier Crcular.
Issuance of this Bulletin is considered appropriate to further reduce the number of challenges 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Dyster Creek - On December 14, 1978, the reactor experienced a scram as control rods were being withdrawn for approach to critical, following a scram from full power which had occurred about 15 hours1.736111e-4 days 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 xenon concentration the operators had not made an accurate estimate of the criticel rod -attern. The operator at the controls was u. sing the SRM count rate, wh.ch 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Control rod 10-43 (first rod in Group 9) was being withdrawn in " notch override" to notch posit on 10, when the reactor became critical on an estimated 2.8..cond perivo.
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 a scram during the initial approach to critical following refueling.
The 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 subtritical. A short reactor period, estimated at 5 seconds, was experienced. The operator.sas attempting to reinsert control rods when the scram occurred.
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366 203
IE Bulletin No. 79-12 May 31, 1979 Page 2 of 3 3.
Hatch Unit 1 - On Jcnuary 31, 1979, the reactor experienced a scram during an approach to critical.
Control rod 42-15 (fifth rod in Group 3) was being continuously 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 zenon.
As indicated above, these short period trips occurred under a wide variety of circumstances. They did have several things in comon, however.
In none of 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 tue ; actor was very subcritical, was pulling a rod on continuous withdrawal.
Action to be Taken by Licensees:
For all GE BWR power reactor faciliti.
vith 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 variables (e.g.,
core Xenon, saoderator tempera
- ure, etc.).
2.
Where inaccuracies in critical rod pattern estimates are.tnticipated due 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.
Review and evaluate your control rod withdrawal sequences to assure that they minimize the notch worth of individual control rods, especially those withdrawn imediately at the point of criticality.
Your review should ensure that the following related criteria are also satisfied:
Special rod sequences st.ould 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 "energency rod in" switch to perform its function 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 of the appropriate NRC Regional Of fice, 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 Comission, Of fice of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C.
20555.
For all BVR 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 expires 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
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366 205
Enclosure IE Bulletin No. 79-12 Page 1 of 3 May 31, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE M0hTHS Bulletin Sulject Date Issued Issued To No.
79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Facilities with an Device in Circuit Breakers OL or a CP for Engineered Safety Systems 79-10 Requalification Training 5/11/79 All Power Reactor Facilities with an OL Program Statistics 79-09 Failures of GE Type AK-2 4/17/79 l'1 Power Reactor Circuit Breaker in Safety I.ilities with an OL or CP Related Systems 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Facilities with an OL Reactors Identified During Three Mile Island Incident 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor Facilities with an of Safety-Related Piping OL or CP 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System His-ing Designed Pressurized Water Power Reactor aligaments Identified Facilities with an
. During the Three Mile Island Incident Operating Licensee 79-06A Review of Operational 4/18/79 All Pressurized Water Power Reactor t'acilities (Rev 1)
Errors and System Mis-alignments Identified of Westinghouse Design with an OL During the Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Pressurized Water Power Reactor Facilities Errors and System Mis-alignments Identified of Westinghouse Design with an OL During the Three Mile Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Power Reactors with an Errors and System Mis-alignments Identified OL except B6V facilities During the Three Mile Island lucident ki:nt;aggg
IE Bulletin No. 79-12 Enclosure May 31, 1979 Page 2 of 3 LISTING OF IE BULLETINS ISSUED IN LAST IVELVE MohTHS Bulletin Subject Date Issued Issued To No.79-05A Nuclear Incident at 4/5/79 All B&W Power Reactor Three Mile Island 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 Vaives 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 Welding and Engineering Co.
79-02 Pipe Support Base Plate 3/2/70 All Power Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts 0L 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 BWR 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 Veld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Velds 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 366 207 r1%. rj rw J
Enclosure IE Bulletin No. 79-12 Page 3 of 3 May 31, 1979 7.iSTING OF IE BULLETINS J5 SUED IN LAST 'lVELVE MohTriS Bulletin Subject Date Issued Issued To No.
78-12 Atypical Weld Material 9/2'/78 All Power Reactor in Reactor Pressure Facilities with an 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 BVR Power Reactor Shock Suppressor Accumulator Facilities with an OL or CP Spring Coils 78-09 BWR Drywell 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 Element 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 an 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 Defcetive Cutler-Hamer 5/31/78 All Power Reactor Type M Relays with DC Coils Facilities with an OL or CP 366 203
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