ML19225A746
| ML19225A746 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 05/31/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Martin T Public Service Enterprise Group |
| References | |
| NUDOCS 7907200142 | |
| Download: ML19225A746 (1) | |
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'u KING OF PRUSSIA, PENNSYLVANI A 19406 May 3i, 1979 Docket Nos. 50-354 50-355 50-311 Public Service Electric & Gas Company ATTN:
Mr. T. J. Martin Vice President Engineering '.nd Constructic.1 80 Park Place Newark, New Jersey 07101 Gentlemen:
The enclosed Bulletin 79-12 is forwarded to you for infarmation.
No written response is required.
If you desire additional ir. formation regarding this matter, please contact this office.
Sincerely, (O
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Boyce H. Grier Director
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Enclosures:
1.
List of IE Bulletins Issueu in Last Twelve Months cc w/encls:
E. N. Schwalje, Manager - Quality Assurance, Engineering and Construction Department 79072001R t l. O C ', 3 i
UNITED STATES NUCLEAR REGULATORY COME 5SION OFFICE OF 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 Sun Reactor scrams, resulting from periods of less than 5 seconds, have occurred recently at three BWR facilities.
In cach 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 #urther 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Oyster Creek - On December 14, 1978, the reactor experienced a scram as control rods were being withdrawn for approa< h to critical, following a scram from full power which had occurred abou+ 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 s< ram 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 ic 1 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 sr. ram 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 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 Tak'en by Licensees:
For all GE BWR power reactor facilities with an operating license:
1.
Peview 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, moderator temperature,etc.).
2.
Where inaccuracies in critical rod pattern estimates are anticipated due to unusual conditions, such as high xenon, procedures should require tnat notch-step withdrawal be used well before the estimated critical p.ition is reached and all SRM channel indicators are monitored so as to ;>e ait selection of the most significant data.
3.
Review and evaluate your control rod withdrawal sequences to assure that tney mirimize the notch worth of individual control rods, especially those withdra n immediately at the point of criticality.
Your review should ensure that the following related criteria are also satisfied:
a.
Special 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 shculd 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 No. 77-07 2.
List of IE Bulletins Issued in Last Twelve Months 410 0;6
'. 1 Enclosure to IE Bulletin No. 79-12 NUCt. EAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D. C. 20555 1
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Date:
Aprq ]4,1977 Page 1 of 3 SHORT PERIOD DURING REACTOR STARTUP DESCRIPTIO" 0F CIRCUMSTANCES:
Recent events of concern to the NRC oc:urred at the M nticello and Dresden SWRs involving inadvertent high reactivity insertiens causing short periods during reactor startup.
At Oresden Unit No. 2 on December 28, 1975 during a reactor startup following a scram from 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 re&=
withdrawal of one notch resulted in a rapid power rise associated with a reactor period of about one second and caused an Intermediata Range Monitor (IRM) Hi-Hi flux s: ram.
The IRM was on its most sensi-tive scale.
The moderator was essentially without voids and the reactor water tamperature was 3380F.
A similar event occurred at this facility on August 17, 1972.
At Monticello en February 23, 1977, following a reactor scram abcut 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> earlier frca unrelated causes, a reactor period of about one second was experienced during startnp before the reactor tripped cn IRM Hi-Hi flux.
The IRM was on its most ensitive scale and the short period resulted frca the withdrawal *cf a control rod one notch.
The reactor moderator had few voids and the water :=nparature was 480 F.
The two cost 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.
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Enclosure to IE Bulletin No. 79-12 I~, Circular 77-07 Date:
April 14, 1977 Page, 2 of 3 3.
High worth red locations were similar and both plants wege using the same generic centrol rod pattern (identified as 31).
4.
Prior to the IEM 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 tarminated by reinsertion cf the rod.
Review of the events shewed that all of the syste=s including the P.eactor Protections System functioned as required.
Analyses indicate that the ccabination of essentially no voids in the =cderator and high xenon concentration acccunted for the conditions that resulted in the co~ntrol red notch aczeiring an unusually high differential reactivity worth which approximated one-half percent delta X/X at Monticello.
This excessive worth of red notch was the result of essentially no voids in the moderator
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and peak xenon conditions which necessitatsd the withdrawal of significantly
=cre control rods than ir normally required to reach criticali ty.
The resultant flux distribution at criticality magnified the nor=al axial peaking at the top of the core due to the heavy xenon concentrations at
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the bott:m.
Adoitionally, the radial contribution to flux peaking was.
enhanced due :
the withdrawal of peripheral rods.
A review of NRC rec:rds showed that after the earlier event at Cresden Unit No. 2 on August 17, 1972, corrective measures were taken for the subsequent startup c:nsisting of notchwise withdrawal of the gr up of rods.
Tnis corrective action was taken only for that operating cycle.
Evaluation of these events indicates that essantially trouble-free startups can be accenplished by avoiding the peak xenon with no moderator voids condition or possibly by the use of a red pattern developed for these particular conditions.
These events indicate a need for all lfcansees of operating BWRs to review their startup procedures and practices to assure that their operating staff has adequate infor sation to perform reactor startups aYoiding 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 rod notch worths can 4 t0 0:3
Enclosure to IE Sulletin No. 79-12 IE Circular 77-07 Date:
April 14,1977 Page 3 of 3
' e encountered under.these conditions.
The procedures should include c
requirements for a thorough assessment folicwing the oc:urrence of a short period before any further r:d withdrawals are =ade.
Thest con-siderations should be included in the operator training and requalifi-catien training programs.
No written response to this Circular is required.
If you need additional infor=ation regarding this mattar contact the Di ector of the c0gni:3nt NRC Regional Offica.
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IE Bulletin No. 79-12 Date: May 31, 1979 Page 1 of 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 Relays Facilities with an Operating License (0L) 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 Cycle 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 (fo~ 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 C,3
IE Bulletin No. 79-12 Date:
May 31, 1979 Paga 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 Materiul 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-128 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, 70508, 7051, with the subject 7051B, 7060, 7060B, 7061 Kay-Ray, Inc.
and 7061B Gauges 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 information) 79-01 Environmental Qualif-2/8/79 All Power Reactor ication of Class IE Facilities with an OL, Equipment except the 11 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 No.
79-03 Longitudinal Weld Defects 3/12/79 All Power Reactor in ASME SA-317. Type Faci'.ities 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, Fxcept Three Mile Island 1 and 2 (For.,ction),
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 0:2
IE Bulletin No. 79-12 Date:
May 31, 1979 Page 4 of 5 LISTING OF 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 Actica), 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 Reactor 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 l
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IE Bulletin No. 79-12 Date: May 31, 1979 Page 5 of 5 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued to No.
79-08 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 an OL or CP (For Information) 79-09 Failures of GE Ts de 4/17/79 All '>ower 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 wi'.h an OL 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers cacilities with an for Engineered Safety JL or CP Systems i;10 0J4