ML19225A748
| ML19225A748 | |
| Person / Time | |
|---|---|
| Site: | Sterling |
| Issue date: | 05/31/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Arthur J ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| NUDOCS 7907200150 | |
| Download: ML19225A748 (1) | |
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Docket No. 50-485 Rochester Gas & Electric Corporation ATTN: Mr. J. E. Arthur Chief Engineer 89 East Avenue Rochester, NY 14649 Gentlemen:
The enclosed Bulletin 79-12 is forwarded to you for information.
No written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, C
oyce Grier Director
Enclosures:
1.
List of IE Bulletins Issued in Last Twelve Months cc w/encls:
C. R. Anderson, Manager, QA Lex K. Larson, Esquire N. A. Petrick, Executive Director, SNUPPS Gerald Charnoff, Esquire 7907200150 c
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UNITED STATES
"'lCLEAR REGULATORY COMMISSION LV t ? 0F INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 IE Bulletin No. 79-12 Date:
May 31, 1979 Page 1 of 3 SHORT PERIOD SCRAMS AT BWR FACILITIES Summary:
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 prev!ously described by IE Circular 77-07 (copy enclosed).
The recent recurrences of this event indicate an apparent loss of effectiveness of the earlier Circular. Issuance of this Bulletin is csnsidered appropriaie 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.
Oyster Creek - On December 14, 1978, the reactor experienced a scram as control rods were being withdrawn for approach to critical, following a scram from full power which had occurred about 15 hours1.736111e-4 days 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 critical rod pattern.
The operator at the controls was using the SRM count rate, which had changed only slightly, (425 to 450 cps) 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 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 subcritical.
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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IE Bulletin No. 79-12 Date:
May 31, 1979 Page 2 of 3 3.
Hatch Unit 1 - On January 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 xenon.
As indicated above, these short period trips occurred under a wide variety of circumstancus.
They did have several things in common, 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 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 cri tical.
The method of estimating critical rod patterns should take into account all important reactivity variables (e.g., core xenon, moderator temperature,etc.).
2.
Where inaccuracies in critical rod pattern estimates are anticipated 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 chanc:1 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:
a.
Specia'i rod sequences should be considered for peak xenon 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 function under prolonged severe use.
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IE Bulletin No. 79-12 Date: 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 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, Office 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 GAU 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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Enclosure to IE Bulletin No. 79-12
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NUCt. EAR REGULATORY COMNISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D. C. 20535 s
IE Circular 77-07 Date:
April 14,1977 Page 1 of 3 SHORT PERIOD DURING REACTOR STARTUP DESCRIPTION OF CIRCUMSTANCES:
Recent events of concern to the NRC occurred at the Monticello and Dresden SWRs involving inadvercent high reactivity insertions causing sbort periods during reactor startup.
At Dresden Unit No. 2 cn December 28, 1975 during a reactor startup following a scram frem unrelated causas about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> earlier, a rain withdrawal of one nctch resulted in a rapid pcwer rise associated with a react:r period of about one second and caused an Intermediata Range Monitor (IRM) Hi-Hi flux scram.
The IRM was on its most sensi-tive scale.
The moderator was essentially withcut voids and the reactor water tamperature was 3380F.
A similar event cccurred at this facility on August 17, 1972.
At Monticello en February 23,1977, following a reactar scram abcut 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> earlier fr m unrelated causas, a reacter period of abcut one second was experienced during startup before the reactor tripped cn IRM Hi-Hi flux.
The IRM was on its most sensitive scale and the short period resulted from the withdrawal'of a contrcl rod one not:h.
The reacter moderater had few voids and the water tenparature was C
48 0 ?.
The two cost recent events were similar in the following respe'* :
1.
Prior to the earTier, u.1related scram, both plants had been operating at or near full power with axial flux peaking.in the bott:m portion of the core.
2.
The time frem the earlier scrams to the subsequent startups.
maximized the xenon concentrations in the core.
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Enclosure to IE Bulletin No. 79-12 IE Circular 77-07 Date: April 14.1977 Page, 2 of 3 High worth rod locations were similar and both plants wete,using the sama generic control red pattern (identified as 81).
Prior to the IF.M scram at both facilities, dramatic indications of high notch worth had been seen with red withdrawals resulting in periods ranging frca 10 to '30 seconds, which were terminated by reinsertion of the rod.
Lview of the events showed that all of the'syste=s including the P.eactor rotections System functioned as required.
Analyses indicate that the
- cmbination of essentially no voids in the =oderator and high xenon encantration accounted for the conditions that resulted in the co~ntrol red notch acze f ring an unusually high differential reactivity worth hich approximated one-half percent delta X/K at Monticello.
This excessive worth of rod notch was the result of essentially no voids in the moderator nd peak xenen conditions which necessitated the withdrawal of significantly core control rods than is normally required to reach criticality.
The esultant flux distribution at criticality magnified the nor=al axial eaking at the top of the core due to the heavy xenen concentrations at
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.he bott:m.
Additionally, the radial centribution to flux peaking was.
enhanced due to the withdrawal of peripheral rods.
' review of MRC records shewed that after the earlier event at Dresden nit No. 2 an August 17. 1972, corrective measures were taken for the subsequent startup consisting of notchwise withdrawal of the grcup of rods.
This corrective action was taken only for that operating cycle.
Evaluation of these events indicates that essentially trouble-free startups can be acccmplished by avoiding the peak xenon with no moderator voids condition or possibly by the use of a rod pattarn developed for these particular conditions.
These events indicate a need for all licensees of operating EWRs to review their startup procedures and practices to assure that their operating staff has adequate infomation to perform reactor startups avoiding such short periods in the event that the above-described conditions of peak xenon with no moderator voids exist at the' time of startup.
Operators should be made aware that extremely high red notch worths can 410 0.2
Enclosure to IE Sulletin No. 79-12 IE Circular 77-07 Data:
April 14,1977 Page 3 of 3
^
be encountered under.these :enditions.
The procedures should include require =ents for a thoroug~n assessment following the cccurrence cf a short period befo-any further red withdrawals are made.
Thest c:n-siderattens should be included in the operator training and requalifi-caticn training programs.
No writtan response to this Circular is required.
If you need additional infor=ation regarding this mattar contact the Director of the c:gnizant NRC Regional Office.
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IE Bulletin No. 79-12 Date: May 31, 1979 Page 1 cf 5 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
78-06 Defective Cutler-5/31/78 All Power Reactor Hammer, Type M Re. lays Facilities with an Operating License (OL) or Construction Permit (CP) 78-07 Protection afforded 6/12/78 All Power Reactor by Air-Line Respirators Facilities with an and Supplied-Air Hoods OL, all class E and F Research Reactors with an OL, all Fuel C,ucle Facilities with an OL, and all Priority I Material Licensees 78-08 Radiation Levels from 6/12/78 All Power, Test and Fuel Element Transfer Research Reactor Tubes Facilities with an OL having Fuel Element Transfer Tubes 78-09 BWR Drywell Leakage 6/14/78 All BWR Power Paths Associated with Reactor Facilities Inadequate Drywell with an OL (for action)
Closures or CP (for information) 78-10 Bergen-Paterson 6/27/78 All BWR Power Reactor Hydraulic Shock Facilities with Suppressor Accumulator an OL or CP Spring Coils 78-11 Examination of Mark I 7/24/78 BWR Power Reactor Containment Torus Facilities with an OL Welds for action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monticello and Vermont Yankee.
All other BWR Power Reactor Facilities with an OL for information 410 0.4
IE Bulletin No. 79 12 Date: May 31, 1979 Page 2 of 5 LISTING 0F IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued To No.
78-12 Atypical Weld Material 9/29/78 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 78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-13 Failures In Source Heads 10/27/78 All General and of Kay-Ray, Inc, Gauges Specific Licensees Models 7050, 70501, 7051, with the subject 7051B, 7060, 7060B, 7061 Kay-Ray, Inc.
and 7061B Gaugos 78-14 Deterioration of Buna-N 12/19/78 All GE BWR Faci-Components In ASCO lities with an OL Solenoids (for action), and all other Power Reactor Facilities with an OL or CP (for infor,..ation) 79-01 Environmental Qualif-2/8/79 All Power 'teacto' ication of Class IE Facilities with aa OL, Equipment except the il Systematic Evaluation Program Plants (for action), and all other Power Reactor Facilities with an OL or CP (For Information) 79-02 Pipe Support Base Plate 3/8/79 All Power Reactor Design Using Concrete Facilities with an OL Expansion Anchor Bolts or CP
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IE Bulletin No. 79-12 Date: May 31, 1979 Page 3 of 5 LISTING 0F IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued to Mo.
79-03 Longitudinal Weld Defects 3/12/79 All Power Reactor in ASME SA-312 Type Facilities with 304 Stainless Steel Pipe an OL or CP Spools Manufactured by Youngstown Welding and Engineering Company 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 79-05 Nuclear Incident at 4/1/79 All Babcock and Three Mile Island Wilcox Power Reactor Facilities with an OL, Except Three Mile Island 1 and 2 (For Action),
and All Other Power Reactor Facilities With an OL or CP (For Information)79-05A Nuclear Incident at 4/5/79 Same as 79-05 Three Mile Island -
Supplement 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactor Facil-alignments Identified ities with an OL Except During the Three Mile B&W Facilities (For Incident Action), All Other Power Reactor Facil-ities with an OL or CP (For Information) 410 046
IE Bulletin No. 79-12 Date: May 31, 1979 Page 4 of 5 LISTING 0F IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued to No.79-06A Same Title as 79-06 4/14/79 All Westinghouse Designed Pressurized Power Reactor Facil-ities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information)79-06A Same Title as 79-06 4/18/79 All Westinghouse (Revision 1)
Designed Pressurized Power Reactor Facil-ities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information)79-06B Same Title as 79-06 4/14/79 All Combustion Engineering Designed Pressurized Power Rcactor Facilities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information) 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 410 047
IE Bulletin No. 79-12 Date: May 31, 1979 Page 5 of 5 LISTING OF IE BULLETINS ISSUED IfPLAST TWELVE M0flTHS (CONTINUED)
Bulletin Subject Date Issued Issued to No.
79-0C Events Relevant to 4/14/79 All BWR Power Boiling Water Power Reactor Facilities Reactors Identified with an OL (For During Three Mile Action), All Other Island Incident Power Reactor Facil-ities with a.i OL or CP (For Information) 79-09 Failures of GE Type 4/17/79 All Power Reactor AK-2 Type Circuit Facilities with an Breaker in Safety OL or CP Related Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety OL or CP Systems fb h /[ 8