ML19256B420
| ML19256B420 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/31/1979 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Crane P PACIFIC GAS & ELECTRIC CO. |
| References | |
| NUDOCS 7906180374 | |
| Download: ML19256B420 (1) | |
Text
6 q >R klGQ
,J UNITED s TATES
];q t
NUCLEAR REGULATORY COMMISSION 5 *:,"Y'hl, ;.h,.
i HEGION V I]QPfis o, df.dj'h
,f 1990 ft CAL IFORfil A BOULEVARD
- [
SUIT E 202, WAlf;UT CHE E K PL AZA e,,,,"
WALNUT CGI F K, calif 0Rf4f A 94596 May 31,1979 Docket rios. 50-275 50-323 Pacific Gas and Electric Company 77 Beale Street San Francisco, California 94105 Attention:
Mr. Philip A. Crane, Jr.
Assistant General Counsel Gentlemen:
The enclosed Bulletin No. 79-12 is fontarded to you for information.
No written response is required.
If you desire additional information regard-ing this matter, please contact this office.
Sincerely,
/
/ M' f
Lff f',i[( lD'2(
.~
Ri H.(Engelken, f n
['irector
'[~' #
Enclosure:
IE Bulletin No. 79-12 with Enclosures cc w/ enclosures:
W. A. Raymond, PG&E J. D. Worthington, PG&E R. Ramsay, Diablo Canyon 7906180374 k
UNITED STATES fiUCLEAR 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 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 previously 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 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.
Oyster Creek - On December 14, 1978, the reactor experienced a scram as control rods were being withdrawn for approach to critical, following a scram from full power which had occurred about 15 hours1.736111e-4 days 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 estirate 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 gritical 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.
IE Bulletin tio. 79-12 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 (fif th 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 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 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 should take into account all important reactivity variables (patterns 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 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 immediately at the point of criticality.
Your review should ensure that the following related criteria are also satisfied:
Special rod sequences should be considered for peak xenon a.
conaitions.
b.
Provide cautions to the operatork on situations which can result exhibit high rod worth)g. first rod in a new group will usually in high notch worth (e.
4.
Review and evaluate the operability of your " emergency rod in" switch to perform its function under prolonged severe use.
~
IE Bulletin flo. 79-12 fiay 31, 1979 Page 3 of 3 5.
Provide a description of how your reactor operator training program covers the considerations above (i.e., itens 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 !!RC 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 GA0 B180225 (R0072); clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.
Enclosures:
1.
IE Circular flo. 77-07 2.
List of IE Bulletins Issued in Last Twelve !<onths L
Circular 77-07 Date: April 12, 1977 Page 1 of 2 SHORT PERIOD DURING REACTOR STARTUP DESCRIPTION OF CIRCUMSTANCES:
Recent events of concern to the NRC occurred at the Monticello and Dresden BURS involving inadvertent high reactivity insertions causing short periods during reactor startup.
At Dresden Unit No. 2 on December 28, 1976 during a reactor startup following a scram from unrelated causes about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> earlier, a rod withdrawal of one notch resulted in a rapid power rise associated with a reactor period of about one second and caused an Intermediate Range Monitor (IRH) Hi-Hi flux scram.
The IRM was on its most sensitive scale.
The moderator was essentially without voids and the reactor water temperature was 338F. A similar event occurred at this facility on August 17, 1972.
At Monticello on February 23, 1977, following a reactor scram about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> eariler from unrelated causes, a reactor period of about one second was experienced during startup before the reactor tripped on IRM Hi-Hi flux.
The IRM was on its most sensitive scale and the short period resulted from the withdrawal of a control rod one notch. The reactor moderator had few voids and the water temperature was 480F.
The two most recent events were similar in the following respects:
1.
Prior to the earlier, unrelated scram, both plants had been operating at or near full power with axial flux peaking in the bottom portion of the core.
2.
The time from the earlier scrams to the subsequent startups maximized the xenon concentrations in the core.
3.
Iligh worth rod locations were similar ahd both plants were using the same generic control rod pattern (i'dentified as B1).
4.
Prior to the IRM scram at both facilities,. dramatic indications of high notch worth had been seen with rod withdrawals resulting in periods ranging from 10 to 30 seconds, which were terminated by reinsertion of the rod.
~-
(heF
t Circular 77-07 Page 2 of 2 REVILW OF THE LVEilTS Sil0WED THAT ALL OF THE SYSTLMS II?CLUDIi!G THE REACTOR PROTECTION SYSTEM FUNCTlodEU AS REQUIRED.
Analyses indicate that the combination of essentially no voids in the moderator and high xenon concentration accounted for the conditions that resulted in the control rod notch acquiring an unusually high differential reactivity t; orth which approximated one-half percent delta K/K at lionticello.
This excessive worth of rod notch was the result of essentially no voids in the moderator and peak xenon conditions which necessitated the withdrawal of significantly more control rods than is normally required to reach criticality.
The resultant flux distribution at criticality magnified the normal axial peaking at the top of the core due to the heavy xenon concentrations at the bottom.
Additionally, the radial contribution to l
flux peaking was enhanced due to the withdrawal of peripheral rods.
l A review of HRC records she ad that af ter the earlier event at Dresden Unit flo. 2 on August 17, 1972, corrective measures were taken for the sebsequent startup consisting of notchwise withdrawal of the group of rods.
This corrective action was taken only for that operating cycle.
Evaluation of these events indicates that essentially trouble-free startups can be accomplished 'y avoiding the peak xenon with no inoderator o
voids condition or possibly by the use of a rod pattern developed for these particular conditions.
These events indicate a need for all licensees of operating BWRs to review their startup procedures and practices to assure that their operating staff has adequate information to perform reactor startups avoiding such short periods in the event that the above-described con-ditions of peak xenon with no moderator voids exist at the time of startup.
Operators should be made aware that extremely high rod notch worths can be encountered under these conditions.
The procedures should include requirements for a thorough assessment following the occurrence of a short period before any further rod withdrawals are made.
These considerations should be included in the operator training and requalification training programs.
flo written response to this Circular is required.
If you need additional infonaation regarding this matter contact tip Director of the cognizant URC Regional Office.
IE Bulletin No. 79-12 Enclosure f4ay 31, 1979 Page 1 of 3 LISTING OF IE BULLETINS ISSUED Ill LAST TWELVE tiONTHS Bulletin Subject Date Issued Issued To No.
79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities wi 11 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 CWR 4/14/79 All BWR Power Reactor Reactors Identified During Facilities with an OL Three flile Island Incident 79-07 Seismic 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 En9ineer-Errors and System flis-ing Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating Licensee 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev 1)
Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three liile with an OL Island Incident 79-06A Review of Operational k/14/79 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 Systen Mis-Power Reactors with an alignments Identified OL except B&W facilities During the Three fiile Island Incident
IE Bulletin No. 79-12 Enclosure May 31, 1979 Page 2 of 3 LISTIllG OF IE BULLETINS ISSUED Ill LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.79-05A Nuclear Incident at 4/5/79 All B&W Power Reactor Three tiile Island Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three flile Island Facilities with an OL and CP 79-04 Incorrect Ueights 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 ASitE SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP flanufactured 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 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 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 L
flodels 7050, 7050B, 7051, with the subject 70518, 7060, 70608, 7061 Kay-Ray, Inc.
and 7061B gauges78-12B Atypical Weld !1aterial 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
IE Bulletin No. 79-12 Enclosure May 31, 1979 Page 3 of 3 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE ft0NTHS Bulletin Subject Date Issued Issued To No.
78-12 Atypical Weld liaterial 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-11 Exanination of flark 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 Accumulator Facilities with an Spring Coils OL or CP 78-09 BWR Drywell Leakage Paths 6/14/79 All BWR Power Reactnr Associated with Inadequate Facilities with an Drywell Closures OL or CP 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 G/12/78 All Power Reactor Air-Line Respirators and Facilities with an OL, Supplied-Air Hoods all class E and F Research Reactors with an OL, all Fuel Cycle Facilities with an OL, and all Priority 1 Material Licensees b/31/78 78-06 Defective Cutler-Harmer All Power Reactor Type !! Relays with DC Coils Facilities with an OL or CP