IR 05000263/1974006
| ML20024G138 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/08/1974 |
| From: | Choules N, Johnson P, Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20024G137 | List: |
| References | |
| 50-263-74-06, 50-263-74-6, NUDOCS 9102070578 | |
| Download: ML20024G138 (2) | |
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O. S. ATOMIC ENERGY COMMISSION
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i DIRECTORATE OF RICULATORY OPERATIONS (
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RIGION III Report of Operations Inspection RO Inspection Report No. 050-263/74-66 Licensee Northern States Power Company 414 Nicollet Mall Mfuncapolis, Minnesota 55401 Honticello Nuclear Generating Plant License No. DPR-22
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Monticello, Minnesota CateEory: C Type of Licenseet BWR (GE), 545 Mwe Type of Inspection:
Routine, Unannounced
' Dates of Inspection:
July 16 - 18,1974 b
Dates of Previous Inspectiont June 18, 1974 (Operations)
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3fotn eon cleIn i
Principal Inspector:
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(Date)
Accompanying Inspector:p.
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Other Accoepanying Personnel None glj
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pt Reviewed By:
. C. Knop Senior Reactor Inspector
/(Dai.e)
g207057s749999 o
ADOOK 05000263 PDH
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SUKHARY OF FINDINGS
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Enforcement Action the following violations are considered to be of Category Il severity.
The licensee reported these violations to the AEC as abnormal occurrences, and was found during the inspection to have taken corrective actions te prevent their recurrencet A.
Paragraph 3.7.A.2 of the Technical Specifications states that " Primary Containment Integrity as defined in (the) Section 1, shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212*F.
." The definition of Primary Containment
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Integrity is given in Section 1.0.P. of the Technical Specifications.
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Centrary to these requirements, the licensee discovered on May 13, 1974,
that a 3/8" air supply line penetrating the primary containment had been
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open since the removal of a vacuum breaker and its air operator in September, 1973.
(Paragraph 10.b)
B.
Paragraph 3.3.B.3(b) of the Technical Specifications requires the Rod Worth Minimiter to be operable below 10% rated thermal power unless a second individual verifies control rod movemer.ts. Paragraph 4.3.B.3(a)
defince the test to be performed to verify ope' ability.
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Contrary to these requiremants, the RWH operability test was not performed prior to reducing reactor power below 10% rated on June 14, 1974.
(Paragraph 13)
Licensee Action on Previously Identified Enforcement Matters The licensee has taken actions to prevent further omissions of required chemistry analyses.
(Paragraph 8)
Unusual Occurrences A.
A 3/8" air supply line penetrating the primary containment was found in May,1974, to be uncapped.
(Paragraph 10.b)
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B.
Local leak rate tests performed during the 1974 refueling outage showed 8 primary containment isolation valves to be Icaking in excess of allowable limits.
(Paragraph 10.c)
C.
Stuffing box hold down cover studs on both recirculatien pumps were discovered in April, 1974, to have cracked because of stress corrosion.
(Paragraph 7)
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D.
Non-injurious cracks were discovered on a stainless steel fitting
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in the standby liquid control system in April, 1974 (Paragraph 15)
E.
A crackad'aocket veld was discovered in a feedvater pump varmup line on June 24, 1974.
(Paragraph 9)
F.
An intermediate range monitor heat balance calibration was not per-formed during a power reduction in June,1974.
(Paragraph 13)
G.
Surveillance required for the Rod Worth Minimizer was inadvartently omitted during a power reduction on June 14, 1974.
(Paragraph 13)
H.
Three hydrogen detonations occurred in the recombiner inlet piping between May 20 and July 8, 1974.
(Paragraph 16.b)
Other Significant Findings
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A.
Current Findinnn Mr. G. Jacobson, Plant Engineer, Technical, is being reassigne( as Superintendent, Nu;1 ear Projects, in the NSP Power Production Department. A succassor is expected to be named in the near future.
B.- Status of Previously Reported Unresolved Items: None reported.
Mananceent Interview
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The inspectors conducted a management interview with Messrs. Larson (Plant Hausger), Clarity (Superintendent - Plant Engineering and Radiation Protection),
and Anderson (Superintendent - Operation and Mainteuance) at the conclusion of the inspection.
The following matters vere discussed:
A.
The inspector stated that he had no questions regarding the investigation and intentions of the licensee with regard to the recirculation pump stuffing box cover studs.
The inspector noted that additional significant information regarding the occurrence had become available subsequent to the follovup report to the Directorate of Licensing, and asked that a concluding report be considered. The licensee stated that an additional report would be provided.
(Paragraph 7)
B.' The inspector stated that he had no comments concerning the veld repair on the feedvater pump varmup line orifice, but questioned what plans the licensee had for three similar velds. The licensee stated that the abnormal occurrence report which was before the Operations Committee for review recommended that the three additional velds be ground out and rcplaced. The inspector indicated that the Operations Committee's decision would be reviewed at a later date.
(Paragraph 9)
C.
The inspector stated that the back door of the turbine building ireake structure near the river, had been observed to be open during a plant tour 3-p__
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consistent with the use of armed guards at the plant boundary. The
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licenses acknowledged the inspector's comment.
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D.
The inspector stated that the licensee's corrective actions related to previously noted violations of reactcr coolant chemistry requirements had been reviewed and were considered to have been completed.
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(Paragraph 8)
E.
The inspector stated that the osission of the rod worth minimiter surveillance and f ailure to cap an air supply line penetrating the primary containment were noted to represent violations of Technical Specifications. He stated that in view of the licensee's rer'rt and the corrective actions completed by the licensee, no response to the enforcement letter would be required.
(Paragraph 10.b and 13)
F.
The inspector stated that the ongoing investigation related to hydrogen
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detonations in the off-gas system had been reviewed. He asked that R0:111 be inforned of the results of the proposed activation analysis.
(Paragraph 16.b)
G.
The inspector stated he had reviewed general plant operations including log booka, tour of the plant, and interviews with operators. He stated that in reviewing the log books he noted that there was an administrative instruction which identified the Reactor and Control Room Log and listed the general items to be in the log but there were no instructions which did the same for the other, plant logs. The inspector suggested that (,j '
an instruction identifying all plant logs and listing the general items to be entered would be desirable. The licensee stated that the inspector's comment would be considered.
(Paragraph 3)
H.
The inspector stated that the log sheets on which the operators record readings on a periodic basis do not require the operator to initial the set of readings that he has recorded.
The inspector stated that the operator's initials would be helpful if it were ever requir:d to go back and review log sheets for problems. The licensee stated that he would consider requiring initials on the periodic log sheets.
(Paragraph 3)
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REPORT DETAILS
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Persons C$ntacted Monticello Plant Staff
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C. Larson, Plant Manager M. Clarity Superintendent - Plant Engineering and Radiation Protection W. Anderson, Superintendent - Operation and Haintenance W. Sparrow, Operations Supervisor H. Nimmo, Maintenance Supervisor C. Jacobson, Plant Engineer, Technical D. Antony, Plant Engineer, Operations L. Eliason, Radiation Protection Engineer F. Schober, Shif t Supervisor
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R. Kmitch, Shif t Supervisor
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D. Nevinski, Engineer, Nuclear J. Pasch, Engineer L. Nolan, Engineer B. Jenness, Engineer B. Day, Engineer P. Pochop, Quality Control Engineer R. Jacobson, Plant Chemist D. Roisum, Lead Plant Equipment Operator and Reactor Operator
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W. Boehme, Plant Equipment Operator and Reactor Operator b
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SUNTAC Nucle _ar Corporation J. Sevier, Of f-Gas System Engineer L. Jordan, Engineer 2.
,Ceneral At the time of the inspection, the plant was operating at approximately 65% power af ter having experienced a scram the day prior to the inspectors' arrival. Operation during the previous week was stated to have been nominally at 85% power with a stack release rate of approximately 65,000 uC1/sec. During informal discussions on fuel
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performance, a licensee representative stated that the sipping performed during the recent outage appeared to have been effective, and that incipient failures present at the time of sipping were considered to be the cause of the high off-gas release rates presently being experienced.
The licensee representative also stated that the effect of the gadolinium in the new 8 X 8 assemblies was determined to be stronger than anticipated, and that this probably caused a slight relative power increase in the older 7 X 7 assemblies. At the time of the inspection, discussions were ongoing between tho licensee and General Electric Company cor.cerning the licensee's desire to couplete replacement of all. 7 X 7 fuel bundles during the next two refueling outages.
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January,1975, was being considered for the next refueling outage, with
the following outage possibly to be scheduled for the fall of 1975.
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Review of Plant Operations The inspector performed a review of general plant opsrations. The following items were reviewed..
Reactor and Control Room Logbook, Shif t Supervisor's Log, Radwaste a.
Log, Pump Log, and Jumper (Bypass) Log.
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Control Room Operations and Staffing c.
Interviews with Control Room Operators d.
A Detailed Tour of the Plant.
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There were no violations of AEC requirement noted.
The inspector noted in the review of the plant logs that the licensee's Administrative Control Directive (ACD) 4.7, plant Operator Activities, identifies the Reactor and Control Room Log.
The ACD specifies in general the items that are to be entered into the log.
For the Shift Supervisor log, the Radwaste Log and the Pump Log there was ne such ACD or ir.struction.
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The inspector reviewed several plant log sheets on which readings are (2/ 4 recorded on a pariedic basin, (e.g., hourly, 4 times a shift, etc.).
These logs are reviewed and signed at tha end of each shift by the lead plant equipe.ent operator but the periodic readings are not required to be initialed by the operator recording the readings.
Differences in handwriting on the log sheets indicated that readings were recorded by different operators durit.3 an eight hour shift. Since there were no initials for each set of periodic reading it was not readily apparent which operator recorded the readings.
4.
Main Steam Isolation Valve (MSIV) Maintenance Al A previous inspection report discussed actions taken by the licensee
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following f ailure of two MSIV's to close during surveillance testing in February,1974.
That report also stated that the licensee planned
further maintenance on the MSIV operator solenoid valves during the 1974 refueling outage.
A licensee representative stated during the current inspection that the AC and DC solenoid valves had been overhauled during the outage under the supervision of the vendor's representative.
The licensee stated that the blocks for all solenoid valves had been disassembled and cleaned, and that new internals had been installed. The new internals were of the same design as those installed during the February repairs; i.e., spring-compensated upper seats in the AC solenoids, and uncompensated seats in t c DC solenoids.
The representative stated that b
1/ RO:III Inspection Report Ha. 050-263/74-02.
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no noticeabla degradation of the removed parts was evident. He also
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stated that pisns call for disassembly and inspection of the solenoid
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valves associated with one inboard and one outboard MS1V during the next refueling outage.
5.
Safety / Relief Valve Testing During the recent refueling outage, the licensee removed the four Dresser safety valves which had been designed to relieve to the primary containment, and replaced them with four additional Target Rock safety / relief valvcs, with discharges piped to the suppression chamber.
Records examined during the inspection showed the setpoints of the eight safety / relief valves to have been established between 1065 and 1075 psig using the cold nitrogen method. The five relief valves previously on site were modified by the installation of a locking pin as requested by RO Bulletin 74-4 The licensee verified by examination of the second stage that this modification had also been completed on the three new safety / relief e
valves received from General Electric and verified by review of quality assurance documents which accompanied the new valves that all previous modifications had been completed. Records showed that on April 25, all eight air operators were satisfactorily stroked 20 times using less than 10 psig air pressure to verify proper air operator performance.
The control room log showed the relief valves to have been operationally tested following reactor startup as follows:
(1) all eight relief valves at 150 psig on May 18, (2) the "G" relief valve at 600 psig on May 19, (3) the "G" relief valve between 200 and 500 psig on May 21, kJ and the "E" and "G" relief valves at 1000,p,sig on,May 21., No abnormalities were noted during the tes't'ing.' The licens'ee' stated that
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the G relief valve had been subjected to additional testing since its air operator was the one which had malfunctioned during testing af ter the 1973 refueling outage.
6.
Missing Items in Reactor Vesse)
The inspector noted from review of Operatiot.. Committee minutes that two items were lost into the reactor vessel during the refueling outage.
Both were small pieces of plexiglass; one was approximately 1" square and the other was a crescent-shaped piece approxir tely 41/2" in length which came from a broken underwater light cover, dimilar pieces of plexiglass were analyzed by the NSP and General Electric testing
facilities, who concluded that the plexiglass would turn into a limp substance at reactor operating temperatures. Ca May 8,1974, the Operations Committee reviewed the question and concluded that the small pieces of plexiglass presented no safety problem. Although not known to have fallen into the reactor vessel, an operator's film badge was also discovered to be missing af ter he had been working on the refueling bridge for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Although minutes of the Operations Committee meeting were not yet availabic, a committee member stated that the matter had been discussed and concluded to represent no safety problem.
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The film badge holder is made of metal, and is approximately 1/2"/ y b
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f 1" in size. The inspector noted that a previous safety analysi>2 had Y
concluded that the loss of a larger componenet (flow switch paddle)
would have no deleterious effect on reactor operation.
7.
Cracked Studs on Recirculation Pumps _
An abnormal occurrence report 3/ discussed the discovery of cracked stuf fing box holddown cover stude on both recirculation pumps.
A subsequent letteti/ provided additional information resulting from the licensee's investigation of the f ailures, stating that the major factor in the failures was teproper heat treatment watch caused susceptibility to stress corrosion cracking.
The concentration cell effect of an elongation pin cavity and possibly floride-containing 0-rings were also considered to be contributing factors. The latter report stated that new studs had been procured for the Number 11 recirculation pump, and that spare studs had been used for Number 12
pump after seal welding the pin cavity. The inspector examined a final report of the evaluation prepared by the licensee, and noted no variances from the information provided in the referenced cor-respondence. However, the completed investigation report did not include the following information which had been subsequently determined:
a.
Abnormal sensitivity to stress corrosion can result if the stud material is improperly heat treated or tempered. Characteristic data from material suppliers indicated the proper hardness range k/
for the material (SA-193, B6) to,be Rockwell C20 to 28. Hardnesses s
for the failed studa were found to be in the range of Rockwell C31 that improper heat treatment could result in hardnesses as low as Rockwell C25, and suggested that the hardness be kept between Re 20 and 24.
b.
The licensee has ordered two new sets of stuffing box holddown cover studs which will beinstalled in the rceirculation pumps during the next outage.
The following criteria were specified for the new studst (1) maximum hardness Re 28 (NSP recommends Rc 20-24, (2) rolled threads, (3) minimum tensile strength 110 kpsi, (4) certification of each heat of material by metallographic examination, (5) no elongation hole (6) serial number to oc
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indicated on the end of each stud, and (7) hardness values for each heat of material to be verified by NSP upon receipt.
c.
The licensee plans to remove existing studs from both recirculation pumps during the 1975 refueling outage and. examine them for noticeabic deterioration.
2/ RO:III Inspection Report No. 050-263/72-06, Paragraph 16.a.
3/ Letter, NSP to Directorate of Licensing, dated 4/15/74.
4/ Letter, NSP to Directorate of Licensing, dated 5/31/74.
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The licensee stated during "the final interview that recently developed information related to the stud failures would be' included in a
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r supplemental letter to the Directorate of Licensing.
8.
Reactor Coolant Chemistry Requirements The reportb! of a previons inspection identified violations of two sections of the Technical Specifications which specified requirements for three reactor chemistry analyses.
The inspector verified that the revised surveillance procedures (revised to include Technical Specifications requirements) and memoranda to the chemistry group describing the to the enforcement letter. The inspector noted that a table had also been provided for guidance to chemistry personnel which identifies all Technical Specifications chemistry requirements and indicates particular conditions for which each analysis is required.
The licensee's corrective actions related to this violation are considered to have been completed.
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9.
Feedwater Pump Warmup Orifice An abnormal occurrence reporL1/ discussed the discovery of a cracked weld in a reactor feedvater pump warmup line. Discussion of the event with licensee representatives and examination of documents related to the investigation showed the licensee's actions to have been as described in the abnormal occurrence report. The final internal abnormal occurrence report which was being cubmitted to the Operations Committee for
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b review was noted to include a recommendation that the veld at the other end of the same orifice and both velds associated with the orifice in the other f eedwater pump we.nup line be ground out and rewelded during the next refueling outage.
Operations Committee review of the recommendation was pending.
10. Primary Containment a.
Integrated Leak Rate Test (ILRT)
The licensee conducted an ILRT on May 13-15, 1974. The inspector examined without cocuent the test procedure and data recorded during performance of the test.
Test records showed an ILRT result of
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0.252% per day, determined by *he error enalysis to be accurate to within 0.0033% per day.
The r3jort on the ILRT which is to be submitted to the Directorate of Licensing was stated to be in typing.
5/ RO Inspection Report No. 050-263/74-02.
6/ Letter, NSP to RO:111, dated 4/29/74 2/ Abnormal Occurrence Report No. 263/74-19, dated 6/24/74.
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8/ Letter, NSP to Directorate of Licensing, dated 5/24/74.
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b.
Primary Containment Leakage A licensee abnorm al occurrence reporeSI discussed the discovery during initial ccaduct of the ILRT of an open 3/8" air supply line penetrating the primary containment boundary.
Review of the discovery with a licensee representative and examination of related records showed tie event and corrective actions to have been as described in the licensee's abnormal occurrence report. A step was noted to have been added to a revised surveillance test procedure for the torus-drysell vacuum breakers to verify that the manual block
valves installed ftllowing the occurrence are closed upon completion of surveillance testing.
The inspector informed the licensee that the event representei a violation of the Limiting Condition for Operation expressed it. paragraph 3.7. A.2 of the Technical Specifications.
The licensee's correcti'e actions relating to this violttion were noted to have been completed.
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Local Leak Rate Tests c.
Local leak rate tests conducted by the licenree during the spring 1974 refueling outage showed 8 primary containment isolation valves to be leaking in excess of Technical Specifications limits. The valves found to be leaking, the reasons for excess leakage, and a description of corrective actions takeg/were provided in the licensee's abnormal occurrence report.-
Review of the licensce's summary data sheets shoved the as-Icit leakages to have been as
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stated in the referenced report.
In addition to the repairs described for main steam isolation valve 2-80A, similar work was performed on FGIV 2-86A, although its initial leakage was satisf actory.
Test data indicated the highest leak rate on the remaining six MSIV's to have been 5.08 SCFH (Technical Specifications allov 11.5 SCFH).
Total as-left leakage for testable penetrations (less MSIV's) and double gasketed seals were noted to be 95.15 and 6.76 SCTH, compared to Technical Specifications limits of 103.2 and 34.4 SCFH. respectively.
11.,High Energy Pipe Break Restraints The licensee has in previous correspondence described additional restraints to be added to feedwater piping and to the HPCI steam line
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in the torus to prevent pipe whip in the event of a line break. A licensee representative stated that these modifications had been
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Their completion ves coepicted during the 1974 refueling outage.
subsequently noted to have been discussed in the Cycle 3 Startup Report.IS/
8/ Letter, NSP to Directorate of Licensing, dated 5/24/74.
I/ Letter, NSP to Directorate of Licensing, dated S/20/74.
T0/ Letter, NSP to Directorate of Licensing, dated 7/19/74, Paragraph V.D.
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12. Residad Best Removal (RHR) Heat Exchantere_
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stated that the licensee was considering A previous inspection report sedifications to the RHR service water system to correct the relationship between RHR heat exhanger pressure rating and RHR service water pump During a phone conversation subsequent to the inspection, shutoff head.
a licensee representative stata.d that new bowl assemblics for all four RHRSW pumps were on order, with delivery scheduled for early 1975.
Installation of the new bowl assemblies will be scheduled upon receipt.
13. Omf sof on,_of Surveillance Tests Abnormal occurrence reportsE/ M / submitted by the licensee described the omission of intermediate range monitor (IRM) and rod worth minimiser (RWM) surveillance during a power reduction to hot standby on June 14-15, 1974.
The inspector discussed the omissions with licensee representatives during the inspection and verified corrective actions to have been taken
as described in the abnormal occurrence reports.
The inspector stated that the omission of the RWM surveillance represented a Technical Specifications violation, but that a response from the licensee describing further corrective actions would not be required.
14. Standby Liquid Control Tank Level Indication _
EI stated that the licensee was considering
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A previous inspection report the installation.of additional level indication on the standby liquid
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control tank. A representative stated during this inspection that the-new indicator to be installed would be a buoancy IcVel transmitter which will provide an output to the level alarm and the local narrow range indicator. The present bubbler level transmitter will continue to provide vide range level indication to the control room.
15. cracked Tee in Standby Liquid control system
! discussed the observation of indications An abnormal occurrence repori.
in a stainless steel tee associated with the flow switch in the standby liquid control system. The report stated that the cracks were not of
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suf ficient depth to be considered injurious. The inspector examined the tee during a plant tour, reviewed minutes documenting review of the
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resolution, and concluded the condition to have been adequately described in the licensee's report.
M/ RO:III Inspection Report No. 050-263/73-11.
M/ Abnormal Occurrence Report No. 263/74-17, dated 6/24/74.
M/ Abnormal Occurrence Report No. 263/74-18, dated 6/27/74 M/ RO:111 Inspection Report No. 050-263/73-5.
M / Letter, NSP to Directorate of Licensing, dated 4/22/74.
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16. Off-Cus Eyetem Modification'
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Evaluation of "A" recombiner n.
a.
Overheating of the "A" of f-gas recoa.biner in December,1973, and related investigative and corrective actigtg by the licensee are discussed in previous inspection reports.-
A copy of the licensee's investigation and evaluation report was obtained in May, 1974, for review by RO III.
Followup actions were discussed A
with a licensee representative during the current inspection.
fermal change per 10 CFR 50.59, reviewed by the Operations Committee on May 9,1974, docenented two deviations from the ASME Section III requirements f or the recombinert (1) the vessel shell does not comply precisely with SA-387 Crade C requirements, although materiad properties are defined in the evaluation report and were used in stress calculations which showed acceptable results; and (2) two heat cycles (one inadvertent and one for stress relieving) were imposed
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upon the recombiner shell subsequent to radiographic examination.
The inspector reviewed the safety evaluation performed for the design change as required by 10 CFR 50.59, and examined a letter dated April 29, 1974, from the Chief Inspector of the state Division of Boiler Inspection which approved the recombiner vessel for operational use.
It was noted that the licensee's investigation and evaluation report described a modification to the recou.biner vessel heater control circuitry which provides complete independence of the heater cutout feature from the normal heater control circuit.
k The inspector informed.the licensee representative that he had no further questions regarding usability of the "A" recombiner vessel.
e b.
Hydronen Detonations (1) During startup testing of the of f-gas recombiner system on May 20, 1974, the licensee observed a hydrogen detonation in the recombiner inlet piping.
'.no licensee's investigation concluded that a spark may have been initiated by a flow control valve in the inlet piping, and valve internals were replaced with non-spark.ing material. Upon completion of the licensee's investigation and inspection of the system, the recombiner subsystem was returned to service. Further details relatedtotheoccurrenceagdescribedinthelicensce's
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abnormal occurrence report.-
(2) A second hydrogen detonation was observed to occur on June 10, 1974, during operational testing of the recombiner systems at 25% power.
16] RO Inspection Reports Eo. 050-263 /7 3-12, No. 050-263/74-02, and 050-263/
74-03, 17/ 1.etter..NSP to Directorate of Licensing, dated 5/29/74.
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The recombiner system was again removed from service, and
the licensee assembled a.special task force to investigate i
the cecurrence.
Design changes were made to the system s
described fn the licensee's abnormal occurrence report, in an effoit to eliminate the cause of the detonations.
Special instrumentation and recording equipment were als) connected to the air ejector of f-gas lines and the recombiner inlet piping to assist in the analysis of any subsequent detonation.
(3) Prior to returning the recombiner system to service af ter the second detonation, as recommended by General Electric Company, the rupture discs at the air ejector discharge were replaced with blank flanges. The related design change and safety evaluation vere noted to have been reviewed by the Operations Committee. The licensee's safety evaluation concluded thats (a) the hazard to personnel and the resulting activity release following a potential hydrogen detonation outweigh any benefits to be derived f rom use of the rupture discs; (b) the recombiner
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system is designed to withstand pressures resulting from a hydrogen detonation, and (c) rupturing the rupture discs has a negligible effect on the peak pressure experienced following a detonation.
(4) Upon completion of the actions described above, the recocbiner system was returned to service. On July 8,1974, as testing of the recombiner system at 25% power was nearing completion, O
another detonation was observed. The recombiner system was again removed from service for investigation.
Since no physical damage was observed and no radioactivity was released as a result of the detonation, the licensee determined the event not to represent an abnormal occurrence. An investigation into the cause of the detonation was initiated, using the information recorded by the special instrumentation installed on the recombiner system. The licensee's investigation report stated thats (a) temperatures in the recombiner inlet piping were noted to have increased significantly prior to the hydrogen detonation; (b) data indicated the detonation to have originated in the train '
B inlet piping, traveled into train A, and back through the
buried piping into the air ejector room piping; (c) instantaneous pressures as high as 200-300 psig existed for one to two
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seconds, re-adjusting to within 3 psi of the initial pressure within 5 to 6 seconds.
The report concluded that catalytic recombination was occurring in the recombiner inlet piping.
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The catalyst for the recombination had not been conclusively identified at the time of the inspection. However, following overheating of the "A" recombiner vcsoel (See previous pt.ragraph),:
18/ Abnormal Occurrence Report No. 263/74-16, dated 6/20/74.
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the recombiner had been flushed through the inlet piping to
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remove foreign materials.' The licensee suspects that this may r,
have deposited palladium fines from the recombiner catalyst in
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the inlet piping. Attempts to determine whether palladium was present by chemical analysis were stated to have been un-successful, and the licensee was making arrangements for activation analysis of samples taken from the inlet piping.
(Phone conversations subsequent to the inspection reported that positive indications of palladium had been observed).
Upon identification of the catalyst, the licensee plans to evaluate possible means for removing or ne'J:ralizing it.
Licensee representatives stated that R0:111 would be informed when the catalyst had been identifild, and that a report summarizing the
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conclusions reached by the investigation would be submitted to the Directorate of Licensing.
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Of f-Can System Testing A licensee representative stated during the inspection that all preoperational testing for the of f-gas systoa had been completed.
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Operational testing at 25% power had also been mostly compleced, although it was expected to be repeated after corrective actions were taken to minimize the chances of further hydrogen detonations.
17. Water In off-Gas Line b
on July 15, while attempting to add water to a drain tank, water was introduced into the of f-gas line at the base of the stack.
The sequence of events which followed resulted in a decrease in the main condenser vacuum, loss of suction pressure on the reactor feedwater pumps, and a feedwater pump trip followed by a low water level scram.
Following the scram, the No. 12 off-gas HEPA filter, which had exhibited *
high differential pressure during the occurrence, was replaced and both HEPA filters were tested for ef ficiency- (using cold DOP tests).
The test showed No.11 filter to have failed, and No.12 was placed into se rvic,e. Upon examination the following day, water was found to be present in the annulus around the No. 11 filter. Remaining water was On July 17 an successfully removed from the off-gas line on July 16.
12 filter to have failed, possibly due to
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efficiency test showed the No.
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the introduction of moisture from the off-gas line the previous day.
At this time the No. 11 filter was replaced, efficiency tested, and returned to service.
The licensee stated during the inspection that he planned to report the filter failures to the AEC as an unusual event..
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UNITED STATES 1..',, -(
l ATOMIC ENERGY, COMMISSION
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DlV85loH OF COMPLIANCE paolowles
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GLan ELLYN, ILLINots 40137 m:m ses-ases
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050 4 34 b %
A.
R0 Inspection Report No.
Transenittal Date
August 8, 1974
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Distribution:
Distribution
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R0 Chie f, FS&EB RO Chief, TS6EB RO:HQ (5)
RO HQ (4)
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DR Central Files L D/D for Fuel & Materials Regulatory standards (3)
DR Central Files
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Licensing (13)
R0 Files RO Files I
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B.
RO Inquiry Report No.
Transmittal Date
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Distribution:
Distribution:
R0 Chief, FS&EB R0 Chtef FS6EB RO HQ (5)
R0 HQ DR Central Files DR Central Files Regulatory Standards (3)
RO Files Licensing (13)
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to Files
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Incident Notification From:
(Licensee 6 Docket No. (or License No.)
Transmittal Date
Distribution:
Distribution:
R0 Chief, FS&EB RO chie f, TS&E B RO NQ (4)
RO.HQ (4)
Licensing (4)
L D/D for Fuel & Materials DR Central Files DR Central Files R0 Files R0 Files t
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