ML17317B558

Notice of Deviation from Insp on 790312-16

ML17317B558
Person / Time
Issue date:	03/29/1979
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
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ML17317B559	List: ML17317B558 ML19270G311 ML19270G313
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REF-QA-99900002 NUDOCS 7906060108
Download: ML17317B558 (9)

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U.S.

NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III 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.

l.

Oyster Creek - On December 14, 1978, the reactor experienced a scram as control rods were being withdrawn for approach to critical, following a scram from full power which had occurred about 15 hours1.736111e-4 days 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Control rod 10-43 (first rod in Group 9) was being withdrawn in "notch override" to notch position 10, when the reactor became critical on an estimated 2.8 second period.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.

IE Bulletin No. 79-12 May 31, 1979 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 than 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 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 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 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 withdrqwn 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.

IE Bulletin No. 79-12 May 31, 1979 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 GAO B180225 (R0072); clearance expires 7/31/80.

Approval was given under a blanket clearance specifically for identified generic problems.

Enclosures:

l.

IE Circular No. 77-07 2.

List of IE Bulletins Issued in Last Twelve Months

'UCI EAR REGULATORY CO%MISSION OFFICE QF INSPECTION AND ENFORCBlEHT MASHINGTON, 0.

C.

20555 IE Cfrcular 77-07 Date.'prf L l4, 1977 Page I of 3 SHORT PERIOD OURIHG REACTOR STARTUP OESCRIPTION 0F CIRCUMSTANCES:

Recent events of concern to the NRC occurred at the Montfcello and Dresden IARs fnvol vfng fnadvertent high reactfvf ty fnsertf ons causf ng short per fods during t Iactor startup.

At Oresden Unft Ho.

2 on December 28, 1976 during a reactor star up followfng 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 wfthdrawal of one notch resul ted fn'a rapid power rise assocfatH with a reactor perfad of about one second and caused an Int rmedfa.e Range Nonitor (IRK) Hf-Hf flux scram.

The IRN was on f s most sensf-tfve scale.

The moderator was essentially without voids and the reactor water ta~perature was 338oF.

A similar event occurred at

, this Sacflity on August 17, 1972.

At Mantfce1lo on February 23, I977, following a reactor scram about IO bours earlier from unrelated

causes, a reactor period of abou one second was experienced during s artup before the re ctor tripp&

on XRM Hi-Hi flux.

The IRH was on its mos sensitive scale and ihe short period resu1ted from the wfthdr wal'of a control rod one notch.

The reac+wr Iederator had few voids and the water tempera"ure was 48OoF.

The two most re-ent events were sfmflar fn the fo11owing respects:

1.

Prior to the ear1fer, unra1ated

scram, both plants had been operating at or near fu11 power wfth axfa1 flux pe>kfng.fa the bottom per fon of the core.

R.

The time from the earlfer scrams to the subsequent startups

$4xhlfzed the xenon ioncentratfens 5n the core.

~ ~

r

~ y

~

~

IE Cfrcalar 77-07 Date:

April 14, 1977 P'age 2 of 3 Hfgh worth rod 'Ibcatfons were sfmf1ar and both plants were usfng the same generic control rod pattern (identified as Sl).

4.

Prfor to the I&8 scram at both facflftfes, dramatfc fndfca.fons ot'fgh notch worth had been seen wfth rod wfthdrawals resultfng fn periods rangfng from IO to 30 seconds, which were terafna Q

by refnser tfon o the rod.

Revfer of the events showed that all o the syst~

fncludfng the Peactor l'rotectfons Systen furc ioncd as required.

~lyses indicate that the coabfnatfon of essentially no voids in the moderator and high xenon concentration accounted for the conditions that resulted fn the co'ntrol rod notch acctufrfng an unusually hfgh differ ntial reactfvfty worth whfch approxfmatcd one-hal, pere nt delta K/K at Nontfcello.

TKs excessfva worth of rod notch was the result of essentfally no vofds fn the moderator and peak xenon conditions which neccssf ~ted the v'drawal oi sfgni

~ fcantly core control rods t>en fs normally requfred to re ch cr ftfcalfty.

The resultant flux dfstrfbu ion at criticality magnified the normal axfal peaking at the top of the core due to the hcavy xcncn conccntra.fons at the bottom.

Additionally, the radial contribution w flux pe king was enhanced due to the withdrawal of peripheral rods.

A revfew of XRC rccor"s showed that aft r the earlier event at Orcsden Qaft No. 2 on August 1T, 1972, correc ive measures were taken.or

~he subsequent startup consfstina of notcnwise wi hdra~al of the group of rods.

Thfs corrective action was taken only for that operatfng cycle.

. Evaluation of these events fndfcates tha. essentially trouble-f; e startups can be aecomplfshed by avofdfng the peak xenon HN no moderator voids Condftfon op possfbly by the use of a rod pattern developed for these partfcular condftfons.

These events fndfcatc a need for all Ifcensees ef operating SMRs to tevfew their startup procedures and practices to assure that their Operatfng staff has adequate fn.ormation to par'ore reac or swr ups

,avofdfng such shor t perfods in the event hat t4e above-descrfbed condi fons Of peak xenon with no Ioderator voids exist at the time of s~rtup.

Operators should be made aware that extremely high rod notch worths can 0

1.

ti Cfrcular 77-0'I Cate:

. Aprf1 14, lS77 Page 3 of 3 bc encountered under.these condf fons.

The procedures should fnclud requirements for a horovqh assessment fo11ow(nq the occurr rc o,

a short period be ore any further rod wfthcfravels are made.

The& can-sfdera fons should be included fn the operator trafnfnq and rej,ua'Jfff-catfon trafning programs.

Ko wrftten response to this Circular fs requir d.

Tf you need add',tfonal fnfonnation reaardfng hfs matter contact the Director of th cognf an:

'NC Regfonal Office.

Bulletin Sub) ec t No.

LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Date Issued IE Bulletin No. 79-12 May 31, 1979 Issued To 79-11 Faulty Overcurrent Trip Device in Circuit Breakers for Engineered Safety Systems 5/22/79 All Power Reactor Facilities with an OL or a CP 79-10 Requalification Training 5/11/79 Program Statistics All Power Reactor Facilities with an OL 79-09 79-08 1

Failures of GE Type AK-2 Circuit Breaker in Safety Related Systems Events Relevant to BWR Reactors Identified During Three Mile Island Incident 4/17/79 4/14/79 All Power Reactor Facilities with an OL or CP All BWR Power Reactor Facilities with an OL 79-07 79-06B 79&6A (Rev 1)

Seismic Stress Analysis of Safety-Related Piping Review of Operational Errors and System Mis-alignments Identified During the Three Mile Island Incident Review of Operational Errors and System Mis-alignments Identified During the Three Mile Island Incident 4/14/79 4/14/79 4/18/79 All Power Reactor Facilities with an OL or CP All Combustion Engineer-ing Designed Pressurized Water Power Reactor Facilities with an Operating Licensee All Pressurized Water Power Reactor Facilities of Westinghouse Design with an OL 79-06A Review of Operational Errors and System Mis-alignments Identified During the Three Mile Island Incident 4/14/79 All Pressurized Water Power Reactor Facilities of Westinghouse Design with an OL 79&6 Review of Operational Errors and System Mis-alignments Identified During the Three Mile Island Incident 4/ll/79 AllPressurized Water Power Reactors with an OL except BSW facilities Enclosure Page 1 of 3

IE Bulletin No.

79-'12 May 31, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject No.

Date Issued Issued To 79-05A 79-05 79-04 79-03 79W2 Nuclear Incident at Three Mile Island Nuclear Incident at Three Mile Island Incorrect Weights for Swing Check Valves Manufactured by Velan Engineering Corporation Longitudinal Welds Defects In ASME SA-312 Type 304 Stainless Steel Pipe Spools Manufactured by Youngstown Welding and Engineering Co.

Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts 4/5/79 4/2/79 3/30/79 3/12/79 3/2/70 All B6W Power Reactor Facilities with an OL All Power Reactor Facilities with an OL and CP All Power Reactor Facilities with an OL or CP All Power Reactor Facilities with an OL or CP hll Power Reactor Facilities with an OL or CP 79-01 Environmental Qualification 2/8/79 of Class IE Equipment All Power Reactor Facilities with an OL or CP 78-14 78-13 78-12B 78-12A Deterioration of Buna-N Component In ASCO Solenoids Failures in Source Heads of Kay-Ray, Inc., Gauges Models 7050, 7050B, 7051,

7051B, 7060,

7060B, 7061 and 7061B Atypical Weld Material in Reactor Pressure Vessel Welds Atypical Weld Material in Reactor Pressure Vessel Welds 12/19/78 10/27/78 3/19/79 11/24/78 All GE BWR facilities with an OL or CP All general and specific licensees with the subject Kay-Ray, Inc.

gauges hll Power Reactor Facilities with an OL or CP hll Power Reactor Facilities with an OL or CP Enclosure Page 2 of 3

IE Bulletin No. 79-12 May 31, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject No.

Date Issued Issued To 78-12 78-11 Atypical Weld Material in Reactor Pressure Vessel Welds Examination of Mark I Containment Torus Welds 9/29/78 7/21/78 All Power Reactor Facilities with an OL or CP BWR Power Reactor 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 Shock Suppressor Accumulator Spring Coils All BWR Power Reactor Facilities with an OL or CP 78-09 78-08 BWR Drywall Leakage Paths Associated with Inadequate Drywell Closures Radiation Levels from Fuel Element Transfer Tubes 6/14/79 6/12/78 All BWR Power Reactor Facilities with an OL or CP All Power and Research Reactor Facilities with a Fuel Element transfer tube and an OL 78-07 78-06 Protection afforded by Air-Line Respirators and Supplied-Air Hoods Defective Cutler-Hammer Type M Relays with DC Coils 6/12/78 5/31/78 All Power Reactor Facilities with an OL, all class E and F

Research Reactors with an OL, all Fuel Cycle Facilities with an OL, and all Priority 1

Material Licensees All Power Reactor Facilities with an OL or CP Enclosure Page 3 of 3