IR 05000219/1971001
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U. S. ATOMIC ENERGY COMMISSION
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REGION I
DIVISION OF COMPLIANCE l
Report of. Inspection C0 Report No. 219/71-1 Licensee:
JERSEY CENTRAL POWER & LIGHT COMPANY Oyster Creek - 1 License No. DPR-16 i
Category C l
Dates of Inspection:
April 6-8, 12, 13, 1971 l
DatesofPreviousInspectign:
October 13-16, 28, 29 and November 5, 6, 1970 Inspected by: /*.NUh 6/8/!7[
R.
. McDermot, Reac' tor Inspector (Principal)
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Reviewed by :
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R. T. Carlson, Senior Reactor Inspector
%)atd Proprietary Information:
None
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i SCOPE Type of Facility:
Boiling Water Reactor Power Level:
1690 MWt I
Location:
Forked River, New Jersey Accompanving Personnel:
Mr. D. Pomeroy, Technical Support Branch, CO:HQ on April 6-8,fl971toreviewliquideffluentmonitoringand the performance of the liquid rad waste system. His report is i'ncluded as Attachment No. 3.
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Dr. C.Pelletier, Technical Support Branch, CO:HQ accompanied for orientation.
Mr. G. Fredrickson, GAO, was present on all dates.
Scope of Inspection:
An announced routine inspection was made to review plant operations for the inspection period, review the status of previously identified problems, review liquid effluent monitoring practices and review the performance of the liquid radwaste system.
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9604170006 960213 PDR FOIA DEKOK95-250 PDR
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SUMMARY
- Safety I'tems I
1.'LLicuid Waste System Performance'- A review of the liquid radwaste-system design objectives was not made because the' licensee.has been
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unsuccessful in determining what the objectives are. The process
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idesign drawings'usually available for review at the site apparently Ldo not exist ~ for the Oyster Creek station. A review of the expected performance of the radwaste system as outlined in the FD and SAR was
made and compared with the system performance and substantial variances j
were observed. '(Attachment No. 3 and Noncompliance. Item No. 5)
Liauid Effluents It appears that the licens'ee exceeded the annual 2.. average concentration limit specified in Technical Specification 3.6.b.1.a.
for calendar year 1970. JC has reported * the release of 4.5 curies for calendar year 1970 (approximately 177..of annual limit) which was escablished
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by a gross beta analysis of liquid waste.
Plant records and independent
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sample analyses ** indicate a nonconservative error of approximately 3007.
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in the gross beta analysis which was being used by JC.
Additionally,.the
. ratio of total activity in OC-1 liquid waste to the activity determined by an accurate gross _' beta analysis by JC has been measured as' approx-imately 3007. for a total nonconservative error of 9007..
Based on these errors, it would appear that the actual plant releases may have been as high as 40 curies (icotopic ' analysis) during calendar year 1970 and that theaver9geannualconcentrationlimitforunidentifiedradionuclides***
(1 x 10" uCi/ml) was exceeded by a factor of approximately 1.5.
From
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these findings, it would appear that a gross. beta analysis alone is in-
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sufficient to determine the total activity in liquid wastes.
(Attachment No'. 3)
. j Noncompliance Items -
Title 110 CFR Part 20.201 (Surveys) states "each licensee shall make or
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q cause to be made,.such surveys' as may be necessary for him to comply with
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the regulations in this part".
Contrary to the above,~it appears that adequate surveys (analyses) were
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not. conducted to evaluate the radiation hazards incident to the presence
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and release of radioactive liquid waste. A significant. nonconservative
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error'was noted in the method used for gross beta analysis of the activity
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in the radioactive liquid waste. This was determined to have resulted from
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an incorrect calibration of the counting equipment.
It was further evident
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that the ratio of total activity to the activity determined by gross beta analysis in liquid waste was not.used as a factor in determining and-reporting. actual plant releases or in controlling the activity contained
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in outside-storage tanks.
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- Semi-annual Report No. 3.
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" SamplesLanalyzed by OC-1, NYOO.and Idaho' Operations Office.
. ?9fechnical.' Specification'3.6.B.1.a.
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2. ' Technical Specification 3.6.B.1.a. states "The annual average. concentration
of unidentified radionuclides shall not exceed'1'x 10-7 uCi/cc."
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. Contrary to the above, it appears that the 1970 annual average-concentra-wastesexceeded-1x10~gionuclidesreleasedfromtheplantinliquid l
tion of unidentified ra
'uCi/cc. This finding was based on an appropriate
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upscaling of releases that have been reported to the'AEC in semi-annual
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report No. 3.
The appropriate correction was determined by a review of-analysis records at 0C-1 and by comparisons of independent samples analysed by AEC. laboratories with similar samples analyzed at 00-1.
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3.. Technical Specification 3.6.B.2. states " Radioactive liquid effluents being released into the discharge canal shall be continuously monitored".
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Contrary to the above, the continuous in-line monitor that is used to satisfy this. Technical Specification requirement was considered to be incapable of performing its intended function as a result of high back-ground radiation. Based on our review, it is estimated that the facility
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could have released at the rate of 400 curies / year without tripping the in-line monitor alarm.
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4.
Technical Specification 3.6.C. states in part that "the maximum amount of
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radioactive liquids in storage in the waste sample tanks, the floor drain
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sample tanks, and the waste surge tank shall not exceed 0.7 curie."
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i Contrary to the above, the gross beta analysis disclosed a total of 0.316 curies in these tanks on March 10, 1971.
By appropriately upscaling the
gross beta analysis to correct for identified errors in this counting
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technique, it appears that the total curie content of these tanks may have
been at least as high as 1.8 curies.
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5.
Section 3.C.(2) of the license states that " Jersey Central shall report to the Director, DRL, within 30 days of its observed occurrence, any substantial variance disclosed by operation of the facility from performance specifica-
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i tions contained in the facility description and safety analysis report
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(safety analysis report) or the Technical Specifications",
Contrary to the above, no written report was submitted when the liquid i
radwaste system performance varied substantially from information prosented in the FD and SAR.
Specific examples of the above include:
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a.
Section IX-4, paragraph 4.2 of the FD and SAR states the waste collection drains are " essentially always reused as condensate."
During 1970 less than 10% of waste collection drains were reused as condensate, the remaining was released to the circulating water i
system.
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b. ' Table IX-4-2 of the FD and SAR lists the expected concentrations of radioactivity'in the floor drain system as a function of fuel performance and interpolation of this table for an off-gas release
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rate of 10,000 uCi/second (which is typical oi station performance for December, 1970 to March, 197 indicates that the expected floor drainconcentrationis0.gx10'j)Ci/ml.
u Actual concentrations range from 10 to 50 x 10" uCi/ml or 20 to 100 times the expected value.
c.
Section IX-3, paragraph 3.1.3.3. of the FD and SAR scates that the chemical waste will be sampled and concentrated as required.
Our review indicated that these tanks had never been sampled and that during 1970 essentially no concentrating of these wastes was performed.
d.
Table IX-4-3 of the FD and SAR lists the concentrations of various isotopes expected in the liquid effluents for operation with and without fuel failures. The expected I-131 for operation with failed fuel (100,000 uCi/second stack gas) is listed as 0.037. of the total
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activity.
I-131 is not expected in the absence of fuel failures.
l The Oyster Creek semi-annual report for the period ending December 31, 1970, lists the isotopic composition of a floor drain sample and i
irdicates that iodine-131 comprises approximately 16% of the activity
I in this batch. Other isotopes were also determined to be at substantial j
l variance from the expected values.
Unusual Occurrences -
1.
Torus Vacuum Breaker Valves - Failure to Open* - Both torus vacuum i
breaker block valves failed to open during surveillar.c e testing on i
December 18, 1970. Adjustments were made to the valves and the valves
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were successfully retested on that date but testing on the following dat e resulted in both valves again failing to open. Ths problem was determined l
to be an incorrect valve operator linkage adjustme.nt which was corrected.
l Subsequent tests have been performed without additional problems.
l (Section K.1.)
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Leaking Containment Isolation Valve ** - During the containment integrated leakage test conducted in October, 1970, excessive leakage was noted through
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the torus 02 sample line isolation valve which was closed.
Both the torus and drywell 0 sample line isolation valves were replaced with a different design valve.2 JC plans to install double isolation valves and to reroute the 02 sampler extaust back to the drywell.
(Section K.3.)
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Status of Previously Reported Problems - Following the last routine inspection, a form AEC-592***, identifying six (6) items of noncompliance, was sent to the l
licensee on December 7, 1970. A satisfactory reply was received from JCP&L on l
- Letters from JC to DRL dated Decemter 23, 1970 and February 8, 1971.
Inquiry Memoranda Nos. 219/70-r. and 219/70-L.
- Letter from JC to DRL dated November 2, 1970 and Inquiry Memorandum No. 219/70-I,
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dated October 29, 1970 j
- Form AEC-592 sent to Mr. R. H. Sims, Vice President, December 7,1970.
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December 22, 1970.* The proposed corrective measures discussed in the reply were reviewed during this inspection and found to have been implemented.
Other Significant Items -
1, Administration and Organization - A review of meeting minutes for both the on-site and off-site safety committees reflected improved performance.
Audit schedules have been developed and implemented by the off-site safety committee to review plant operations. The technical staff at OC-1 has been increased by seven new additions since the last routine inspection and attempts are being made to add five additional engineers to the site staff.
Three additional persons in the operating organization have received SOL's and additional maintenance personnel have been added.
The surveillance testing program is now being coordinated by one responsible individual.
(Section B.)
Control Rod Performance - Scram time performance appears satisfactory with only a slight upward trend in the time required to fully insert.
Totalized withdrawal stall flows for all rods appears to have stabilized.
No unusual operating problems were noted.
(Section F.3.)
3.
Core Spray System - OC-1 has been experiencing water hammer in both core spray loops during surveillance testing (pump starts).
Pipe movements of up to 6" are being experienced for piping outside of the drywell. Cause is attributed to: (a) the pump discharge piping not being full of water and completely vented and, (b) starting of the core spray booster pumps which closes the bypass check valve around these pumps. The off-site safety committee has requested Burns & Roe to perform a stress analysis of the piping system and this is currently in progress. Based on the results of this analysis, NDT inspection of selected areas of the piping system will be performed,if warranted. GE is providing a field change to consist of a small filling pump which takes its suction from the torus to keep the piping systems full. JC has requested GE to determine if these pumps could develop sufficient pressure to prevent the core spray pump discharge check valves from opening.
It was also determined that a number of pipe anchors had been pulled free during preoperational testing of this system and that the pipe anchors were subsequently modified (strengthened).
A weld defect in the core spray system was identified in October, 1970 during JC's review of the original construction NDT records.
T, bis review wa required by Section 6.6.A. of the Technical Specifications. This, weld.was radiographed because the original RT record could not be located.** Repair welding was also performed in October, 1970. Repair records were reviewed and no deficiencies were noted.
(Section L.)
4.
Drywell Leak Detection Procedures - A review was made of the leak detection procedures and equipment.
Revised procedures went into effect during the inspection.
Evaluation of an independent means to detect unidentified leakage by weekly sampling of the drywell atmosphere is continuing by JC but no meaningful results have been achieved.
(Section K.4.)
- Reply to Form AEC-592 dated December 22, 1970.
- Letter to DRL from JC dated September 30, 197.
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5.. Mini-Stretch Power Increase - Licensed power level was increased from
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J 1600 MWt to 1690 MWt on December 2,1970 The facility was shut down.on
. December 5, 1970 to make plant modifications required for the power increase.
The facility achieved 1690 MWt on December 10, 1970. Testing was completed
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at this power level and all test results were reported to have met the acceptance criteria.. (Sections C and S)
6.
Turbine Initial Pressure Regulator Performance
' Operation of this control
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i system has remained steady since the lasr inspection. GE replaced four (4)
.of the aluminum horizontal'" push-pull" linkages for the bypass valves with
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carbon steel material. GE investigation of the failed linkage * disclosed
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the material did not meet design specifications.
(Sections H.1. and H.2.)
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7.
Torus Vacuum Breaker Block Valves ** - No additional problems have been experienced with the proper opening of these valves. Valve internals were
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inspected and rubber liners were replaced in January, 1971.
(Section K.1.)
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Containment Integrated Leak Rate Test - This test was successfully i
completed in October,1970. A summary report of this test is being
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prepared by JC.
Repairs were required to the torus 02 sampling isolation i
valve due to excessive leak-through and therefore the containment will be
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retested in one year as required by Technical Specifications. The contain-ment isolation valves for both the drywell and torus 02 sampling lines were j
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replaced with a different design and'GE is to provide a design modification to install double isolation valves in these lines and additionally return the exhaust of the 02 analyzer back into contairment.
(Section K.3.)
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Local Power Range Monitors - OC-1 has experienced 13 failures of LPRM l'
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channels.
Cause is attributed to moisture entering cable connectors within the drywell.
(Section F.2.)
i 10. Gaseous Effluents - A review disclosed that reported results of stack off-gas may berv 10% low. The error is attributed to JC's failure to correlate the six radionuclides measured in off-gas releases with the
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sum of the 22 radionuclides present. JC corrected this condition following the inspection.
(Section Q.2.)
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11.
Emergency Power - A review disclosed that load testing of the 175 V station
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batteries resulted in a terminal voltage at the end of the 8-hour test that was below the acceptance 1 Lait. JC corrected this condition following the inspection by cleaning the cell interconnectors.
(Section N.2.)
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'12 Compressed Air Systems - No problems were identified in a review of air systems out the licensee.has stated that surveillance practices to determine proper, system operation will be reviewed.
(Section I.)
- CO Report No. 219//0-6.
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- Letters:from JC to DRL dated December 23, 1970 and February 8,1971; Inquiry Memoranda Nos. 219/70-K and 219/70-L entitled " Failure of Torus Vacuum Breakers"
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-7-Management Interview - Exit interviews were ccaducted with Mr. McCluskey on April 8 and 13, 1971 and by followup telecon on May 11, 1971.
Mr. G. Fredrickson o.f the GAO was present during both site exit interviews.
Exit Interview on' April 8,1971 - Messrs. Pomeroy and McDermott discussed deficiencies and apparent items of noncompliance that were identified during-Mr. - Pomeroy's review of liquid effluent monitoring and radwaste system performance.
(See Attachment No. 3)
Exit Interview on April 13, 1971 and Followup Telecon on May 11, 1971 - Mr. McDermott held discussions with Messrs. McCluskey, Ross, Carroll, Riggle and Reeves.
Issues
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discussed were as follows.
1.
Control of Liquid Radioactive Wastes - The inspector stated that his review on April 13, 1971 disclosed that JC had not corrected (in entirety)
the discrepancies identified during the April 8, 1971 exit interview concerning the release and storage of liquid effluents. Mr. McCluskey was informed that the facility was still controlling on a gross beta
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analysis and even though the errors had been corrected in this method of analysis, JC was still not relating this analysis to the total activity in the wastes (isotopic analysis).
Mr. Ross proposed then to measure l
gross beta activity and add to this result the concentration of radio-nuclides known to be only gamma emitters (Mn-54 and Cr-51). After further discussion, Mr. McCluskey agreed to do an isotopic analysis for each batch of liquid waste before discharging it.
The mixture MPC would be used to control effluent releases from the plant on all future dumps until a complete review could be made of the relation between the total activity and a gross beta analysis.
He further stated that isotopic analysis would be used to control the curie content of outside storage tanks in the future.
(Section Q and Attachment No. 3)
2.
Core Spray Syste - Water Hammer - The inspector stated that he would
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closely follow the planned review of pipe stresses and NDT inspection.
JC was requested to provide a written report to DRL when the stress evaluation is completed for the piping system.
(Section L)
3.
Generator Load Rejection and Turbine Stop Valve Closure Scrams - The inspector stated that his raview had disclosed that the trip point for
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the pressure switches that bypass these two scram features below 45%
power were not being verified during quarterly surveillance testing.
The inspe'ctor pointed out that the Technical Specifications did not specifically require such testing but that testing was bnplied and expected to be. performed.
Mr. McCluskey stated that the surveillance test procedures would be changed to require this testing during future quarterly tests.
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Mr. McCluskey was informed that no documentation was available to establish the basis for the set point for the 45% scram bypass pressure switches and that the present setting was established by verbal communication with' a GE representative.
It was pointed out by the inspector that the pressure switch setting was in effect a scram set point as two scram features are added to the safety circuit by actuation of these pressure switches and it would be expected that JC would have required more than a verbal communication to establish the set point.
It was further pointed out that no testing was performed to verify the accuracy of this set point when the turbine was restarted after the modification was made. The inspector stated that a review of scram 15 indicated that the set points were adequace or conservative but that testing should_have been performed following the modifications. Mr. McCluskey stated'that GE would be requested to supply the necessary documentation to provide the basis for the set points.
Diesel Generator - Shutdown Devices and Past Failures - The inspector stated that a review of the records for the annual diesel generator inspection (required by Technical Specifications) indicated that no testing had been
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performed on the devices which shut the diescis down.
It was pointed out that a premature actuation of these shutdown devices would reduce the reliability of the system.
Mr. McCluskey stated that these checks will be performed, if possible, during future annual inspections.
The reportability aspects of the past failures * of the diesel generators to start on the first attempt were discussed and Mr. McCluskey stated a report would be submitted in the near future.
5.
Emergency Power - The inspector stated that the load testing records for the "B" 125 V station battery during March 1971 had indicated that the
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terminal voltage at the end of the 8-hour load test had dropped below the acceptance value and that no apparent action was taken to investigate the Mr. McCluskey subsequently informed the inspector by telephone that cause.
the battery cell connectors had been removed and cleaned and another load
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test was completed satisfactorily.
6.
1rftrumentation - The inspector stated that an audit sampling of the c-libration frequency for instrumentation associated with Technical Specification limits and with safeguard equipment indicated that no periodic schedule was in effect.
Mr. McCluskey stated that a review of such instrumentation would be made and a program developed. The inspector stated that this area would receive followup review..
7.
Stack Releases - The inspector pointed out that JC's method for analyzing stack releases could be as much as 10-15% low.
Mr. McCluskey, in a followup telecon, stated that Mr. Ross was now using the necessary correction to eliminate this potential error.
- C0 Report No. 219/70-7, Unusual Occurrence item 2.
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' DETAILS
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A'.
Persons Contacted:
O LMr. T. McCluskey, Station Superintendent, OC-1
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Mr.:D. Ross, Technical Supervisor,'OC-1
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Mr. J. Carroll, Operations Supervisor,.0C-1
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Mr.yW.'Riggle, Maintenance Supervisor, OC-1 l
Mr.!J.-Sullivan, Technical Engineer, OC-1
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i Mr. F.'Kossatz, Mechanical. Foreman, OC-l'
l-Mr.'D.1Kaulback, Radiation Protection Supervisor, OC-1 Mr.cN. Goodenough, QA Engineer (Radiography), GPU Mr. T. Pelrine Chemistry-Supervisor, 00-1
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Mr. D. Reeves,. Technical Engineer, OC-1
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Administration and Organization L
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1.
Site Organization a.
Additions'to Technical Staff The present technical staff organization is outlined in Attachment No. 1.
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P Since the last routine inspection, the following personnel have been j
added:
(1) Mr. J. Budd joined as an Associate Engineer on November 30, 1970.
Mr. Budd graduated from the New York Maritime Academy with a
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Bachelor of Science in 1968 and majored in Nuclear Science.- He holds a second engineer's license.
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Mr. R. Stoudnour joined as a chemical Engineer in December, 1970.
Mr. Stoudnour has a Bachelor of Science in Chemistry and has worked
in various water laboratories at conventional stations since June, 1964.
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(3)
Mr. E. Rosenfeld joined as an Associate Engineer in December, 1970.
Mr. Rosenfeld holds a Bachelor of Science in Chemical Engineering
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and has no previous industrial experience.
i (4) ' Mr. A. Rone -joined as an Associate Engineer in December,1970 Mr. Rone has a Bachelor of Science in Electrical Engineering and
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d no previous industrial experience.
l(5).Mr. Di T. Allred joined as a staff engineer in December, 1970.
LMr.-Allred isJon loan from the Jersey Nuclear Corporation.
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Mr. C. Konta joined as a Chemical Technician in January,1971.
(7)' ' Mr. M. ' Oberstadt' joined in January,1971 as a Radiation Technician.
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b.
Recruitment Efforts
Mr. McCluskey informed the inspector that attempts are being made to recruit two additional staff engineers, two additional assistant jtechnical engineers, and one additional associate engineer.
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c.
Operations Organization
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Mr. McCluskey informed the inspector that there are currently five complete operating shif ts with the exception of the shift foremen who
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l position are as listed below:
(1)
Shift Foremen - 4 (2)
Operating Foremen - 3
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(3) A Control Room Operator - 5; B Control Room Operator - 5 (4) A Equipment Operators - 10 ;
B Equipment Operators - 5
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d.
Maintet;ance Organization The present maintenance organization is as outlined in Attachment No. 2 There are plans to add one additional welder and one additional A mechanic.
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2.
Plant operating Review Committee (PORC) Performance
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u Between the period between October 2, 1970 and March 11, 1971, there were 13 meetings of the Plant Operating Review Committee.
Four of these meetings had no attendance by GORB members. The remaining nine had one or the other of the two permanently assigned GORB members present. The quality of the PORC meeting minutes reflected an improvement in the documentation of the reviews from past performance.
3.
General Office Review Board (GORB) Ferformance During the time period between October 26, 1970 and January 15, 1971, j.
there were four meetings of the General Office Review Board. The present committee membership is made up of the following personnel:
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Mr. W. H. Hurst, Chairman i
Mr.-D. E. Hetrick, GPU, Vice Chairman
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Mr. W. W. Lowe, Pickert and Lowe Associates Mr.- D. D. Reese, GPU
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Mr. T. M. Schneider, GE
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Mr.- W. Southerland, GE Mr. 'J. R. Thorpe, GPU Mr. I. R. Finfrock, Jr., JCP&L The only changes in the committee membership were the addition of f
. Mr. Hetrick and the removal of Mr.
G.. H. Ritter. Meeting minutes were
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reviewed by the inspector and considered to be of improved quality.from
'past: performance..
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GORB Audits of Site Activities
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Audits of site activities were conducted by Mr. W. W. Lowe and Mr. N.
Goodenough on October 14,'1970 and by Messrs. Hetrick, Crimmins, Reppert
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At the time of.the inspection and Goodenough on January 12 and 13, 1971..
The - minutes ' of GORB
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an audit was being' conducted by two GORB members.
meetings reflected that the General Office Review Board has reviewed the results of the-audits, as well.ap the action taken to correct identified i
deficiencies.
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5.
Surveillance Testing Program Administration-t Discussions-were held witN Mr. D. Reeves during the inspection and the-following information was pbtained.
Mr. Reeves was assigned overall
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coordinating responsibilit'y for the surveillance testing program in
October, 1970. The system currently in effect to schedule, test, and review
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,results was: reviewed with Mr. Reeves and the only identified deficiencies
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were as,follows:
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a.
Not all testing procedures have been developed to date. Test.
procedures that are still lacking involve testing that is required during refueling outages.
b.
Surveillance test procedures that are used during operational checks have not all been reviewed by Mr. neeves to determine their adequacy.
c.
Th'e testing records do not lend themselves in entirety to evaluating l
the performance of the systems being tested.
Examples of this would
be no operating data is recorded during the testing of core spray
'
pumps, and in many instances the as-found instrument calibrations or trip points are not recorded during calibration and trip checks.
These discrepancies were discussed with Mr. D. Reeves and improvement in these areas is expected.
6..
Jersey Nuclear Company Mr. McCluskey informed the inspector that Mr. D. Allred was provided 't[o
.
Jersey Central on loan from Jersey Nuclear Corporation in December, 19,70.
' Jersey Nuclear Corporation has made available to the JC organization up to four nuclear engineers. Jersey Central has contracted' Jersey Nuclear
.
to supply reload fuel. The stated purpose of Jersey Nuclear providing
. nuclear engineers is to familarize these personnel with the operation of-the Oyster Creek-facility. At present, Mr. Allred is working for Mr. D.
Ross on specia1' assignments. Jersey Nuclear Corporation has not stipulated any restriction in the-work assignments for the personnel they would provide to Jersey Central,
s.
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C. - Operations The; facility was shut down on October 17, 1970 to perform the containment i
integrated leakage test. Miscellaneous maintenance was:also accomplished during
.the shutdown and the reactor resumed operation on October 27, 1970.
Operations continued,until December 5, 1970 when the. reactor was shut down.for one day to make the final tie-in modifications required for the mini-stretch power increase (1600 -
,
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i 1690 MWt). Reactor operations resumed on December 6, 1970 and continued until l
Decembe'r:12, 1970 when the reactor was shut down because of an air leak within the
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drywell. The plant reached 1690 MWt on December 10, 1970. Operations resumed on December 13, 1970 and continued until December 25,1970 when scrans 14 & 15 (described below}
' occurred. Operations resumed on December 27 and continued until January 25,.1971 when the reactor was shut down because.the identified leakage increased to approx-
' imately 14 gpm in the drywell. The source of the leakage was determined.to be a
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leak.in the closed cooling water system' coil located inside the drywell equipment
< drain tank.. Operations resumed on January 27, 1971 and continued until February 13,
.'1971 when the reactor was shut down to inspect the turbine control valve seats.. This investigation was prompted by the recent failure of the valve seat captive screws
.
which occurred at the Nine Mile Point reactor. - Operations resumed on February 19,
-1971 and continued until March'3,.1971 when the unidentified leakage into the dry-well. increased to approximately 4.5 gpm. The source of this leakage was determined to be a pipe nipple leaking-in the closed cooling water system supply to the
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cooling coil in the drywell equipment drain tank. Operations resumed on March 4, 1971 and no additional shutdowns have been experienced by the facility during the
inspection period. 00-1 plans a fall 1971 outage for poison curtain removal.
. Attachment No. 4 lists all scrams since the beginning of csmmercial operation.
Listed below is a description of the unscheduled shutdowns during the inspection period:
Date Cause j
December 12, 1970 An orderly reactor shutdown was prompted by in-creasing 02 levels (up to.
4.5%)'within the drywell.
The source of oxygen was determined to be an instru-ment air leak that resulted from a broken air line oiler in the drywell. Operations resumed on October 13, 1970 December 25, 1970 An automatic scram from 100% reactor power resulting I
(Scram No. 50)
from turbine stop valve closure. Just prior to the
- (No.14. since start of scram,-turbine stop valves were being individually-commercial operation)
. tested. When the No. 2 turbine stop valve was tested, all four turbine stop valves closed._ Strip charts for key parameters were reviewed by the inspector and the following information was obtained. The pressure
.
I
. peaked at 1120 psig. All 8 APRM's dropped to 0
'immediately with.no increase from the pre-scram equilibrium value.. Turbine bypass valves operated normally and went full open following the closure of
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Date cause the turbine stop valves. The events recorder did not function properly following the scram
and no useful information could be obtained from this source. Electromatic relief valves C or D (or both) lifted for a short period during this transient. Reactor vessel level dropped to approximately 8 feet 6 inches above the top of the active fuel (1 foot on the level recorder) and then increased to the full scale v,alue (8 feet on the recorder) momentarily.
See Section H.1 for additional details of this event. Following the scram, the turbine stop valves were investigated by the licensee and no cause could be found to explain the closure.
The reactor was restarted on the same date and power level increased to approximately~250 MWe where additiona turbine stop valve: testing was performed.
See scram 51 below.
December 25, 1970 Automatic scram from approximately 45% reactor (Scram No. 51)
power due to the closure of the turbine stop (No. 15 since start of valves..Just prior to the scram, the No. 2 commercial' operation)
turbine stop valve was being tested. This resulted in the closure of all four turbine stop valves, which in turr. caused the scram.
The reactor remained shut down until December 27, 1970 while a detailed inspection was made of the No. 2 turbine stop valve.
It was determined that this valve was closing in approximately 4 - 5 seconds.
It was postulated by JC and GE that this closure time resulted in a hydraulic shock wave being established in the turbine control oil system which resulted in the closure of the other three turbine stop valves.
The No. 2 turbine stop valve was adjusted to close in 12-17 seconds. The change in the closure time which resulted in the scrams was attributed to a loose lock nut on one of the pilot valves which controls the valve closure speed.
January 25, 1971 Unscheduled reactor shutdown when the identified leakage rate increased to approximately 14 gpm.
Investigation by the licensee disclosed that a
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cooling coil within the drywell equipment drain tank had become eroded and was leaking fresh water coolant into the drain tank. The problem was corrected by reorienting the coil within the tank and weld repairing the cooling coil.
Operations resumed on January 27, 1971.
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Date Cause
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March 3,'1971 An orderly reactor shutdown resulted when the unidentified leakage rate into.the containment l
drywell reached approximately 4.5 spm..
Investigation by the licensee disclosed a pipe i
nippleleakintheclosedcoolingwatersystem
supply (outside the tank) to the cooling coil l
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in the drywell equipment drain tank. The nipple leak was repaired and operations resumed on
.
March 4, 1971.
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E.
Primary System
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1.
Electromatic' Relief' Valves
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-Mr. Riggle informed the inspector that the pilot valves for the electro-matic relief' valves were inspected in December, 1970 and adjustments made to the stroke of the pilots to preclude the binding experienced at
j This was considered-to be a temporary fix another boiling water reactor.
and General Electric Company has provided JC with a field change to.
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j permanently correct the problem by changing the pilot internals. Mr. Riggle informed the inspector that this field change will be installed during the i
planned October, 1971 poison curtain removal outage.
2.
Isolation Condenser Isolation Relav Failure
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Mr. Riggle informed the inspector that during surveillance testing in
March, 1971 one of the four, g51ve-second time delay. relays used to
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isolate the isolation condenser had failed to operate properly. The nature
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of the failure was the time delay feature was lost due to a bellows failure
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in the relay which resulted in the relay operating instantaneously. He
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(Agastat) and that the other two relays are scheduled to be replaced when
replacement parts become available.
It was determined that the loss of-the single relay would not have affected the normal isolation feature.
.
Mr. McCluskey stated (telecon) that a report of this event would be submitted to DRL within 30 days in accordance with Paragraph 3.C.*(2) of the license.
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F.
Reactivity Control and Core Physics
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1.
. Turbine Scram Trips
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The inspector reviewed the modifications made to the plant safety system in early December,1970 ~ that were required 'for the mini-stretch power increase.
Two'new additional scram circuits were added; turbine stop
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valve closure scram at greater than 45% of power level and generator load
,
,
rejection (acceleration relay) scram trip at greater than 45% of power
level. - The; installation and testing records did not reflect the presstre setting
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of the pressure-switches mounted on the turbine third stage that affect
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the 45% bypass provision for these two scram trips. Discussions with
Mr. Riggle indicated that the trip settings had been provided verbally by a GE representative at the site and that the. instruments had been set Calibration at.200 psig (representing approximately 45% of rated power).
records for the switches were reviewed.and the last noted trip check of the instruments was performed on December 29,l1970. As found, readings during this check varied from 150 to 205 psig for the four switches. The inspector informed Mr. Riggle that the installation of these pressure switches on the third stage of high pressure turbine did not agree with the information provided in the application, i.e., that the pressure switches would be installed on the first stage of,the high pressure turbine. Mr. Riggle stated that sufficient instrument tap-in points were not available to satisfy the separation criteria (for the pressure switches) on the first stage, and therefore, the third stage was selected for installation.
He stated that the pressure switch setting (200 psig) was based on the expected third stage
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pressure that'would be experienced at 45% of rated power. The inspector
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inquired-into whether or not the 45% pressure setting had been verified (during turbine startup) following resumption of operation on December 6, 1970.
Mr. Riggle stated that it had not.
A review was made of the quarterly surveillance test procedures that are The required by the Technical Specifications for these trip circuits.
test procedures did not require that the 45% bypass function pressure switch setting be verified during each check. It was pointed out to Mr. Riggle that as most of the surveillance testing is done at 100% of rated power,
'
that the 45% trip setting would not be verified with the proposed method of testing. This item was also discussed during the exit interview and Mr. McCluskey agreed at that time to perform set point verification for Also these pressure switches during future quarterly surveillance tests.
discussed during the exit interview was the concern that JC had set safety system trip points solely on the basis of verbal consunications between GE and JC.
Mr. PcCluskey agreed to provide documentation to the inspector during the next routine inspection verifying the correctness of the 45%
pressure switch settings.
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2.
Local Power Range Monitors (LPRM)
,
During discussions with Messrs. Ross and Sullivan, the inspector was informed that there are currently 13 failed LPRM's. The outputs from j
these LPRM's are bypassed from their respective APRM inputs. A review j
]
of the locations of-the LPRM's disclosed that the facility is operating
,
within the requirements of the Technical Specifications (3.1.B).
.l i
Jersey Central submitted change request No. 6 to DRL on March 3,1971
~
requesting a change to Technical Specification 3.1.B.2.
This change would
permit the continuous use of a traveling in-core probe (TIP) chamber as an
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input to an APRM if all LPRM monitors in.a particular radial core position
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Mr. Ross informed the inspector that he conoidered the cause for the failure of the LPRM's to be due to moisture entering connectors external to the reactor vessel, but within the drywell.. Replacement of LPRM's' requires the reactor vessel head removal.
3.
Control Rod Performance
.
Control rod scram times were reviewed by the inspector for scrams 49, 50 and 51.
Summarized below are the results of the scram times for the 26
selected rods.
.
Scram Average For Maximum Minimum No.
90% Insertion Individual Rod Individual Rod
2.72 seconds 2.92 seconds 2.46 seconds
2.79 seconds 3.11 seconds 2.59 seconds
2.85 seconds 3.18 seconds 2.60 seconds The totalized withdrawal stall flow results on March 29, 1971 indicated a total of 255 gpm. The previous monthly totalized stall flow (February, 1970)
was 273 gpm.
It was reported to the inspector that the reduction resulted from the venting of some drives. Currently, one control rod (42-19) has a stall flow of greater than 5 gpm. Withdrawal stall flow for this rod is i
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5.8 spm and it is reported that the rod operates satisfactorily at near normal pressure.
It was reported that there are no operating problems'with 'the I
drives.
4.
Reactor Vessel Level Instrumentation Mr. Riggle was questioned by the inspector as to whether or not 0C-1 had j
experienced any problems with the collection of noncondensable gases in the steam condensing pots used for the reference legs for the GE/MAC reactor
.
vessel level instruments (used in the feedwater control system) and for the low-low-low level instruments (used in the reactor protection system).
He informed the inspector that prior to plant turnover from GE to JC, some
' problems had been experienced and at that time the steam condensing pots were reoriented 90 from a vertical to a horizontal position.
He also informed
the inspector.that no measurements had been taken by JC to determine if the steam condensing pots were free venting, i.e.,
if the slope of the line from
.
the' condensing pot to the point where the instrument line taps into the reactor vessel is positive during reactor operation.
Mr. Riggle did inform the. inspector that one of the GE/MAC level transmitters was reading low by -
approximately 1 foot'below the indicated Yarway indication.
This issue was discussed with Mr. McCluskey subsequent to the inspection and he stated that a determination will be made as to whether or not the subject steam condensing pots are free venting during reactor operation.
This issue will rec
/e followup during the next routine inspection.
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H.
Power Conversion System 1.
Turbine Initial Pressure Regulator Performance Since the last routine inspection on October 13-16, 1970 no additional pressure disturbances have been experienced as a result of malfunctions or oscillations of the turbine initial pressure regulator system.
Scram No. 50 (See Section C) highlighted the possibility of a new transient not previously expericarcd by the facility.
Mr. Carroll informed the inspector that when.the turbine stop valves are closed by an electrical signal or on demand from the control room operator, the turbine control valves also close. This results in prompt opening of the turbine bypass valves.
Scram No. 50, which resulted from the closure of the turbine stop valves due to hydraulic oil pressure disturbances would rygt have resulted in the prompt opening of the turbine
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bypass valves. The effect of a delay in turbine bypass valve opening could not be obtained from recorder charts as the control room operator j
had promptly initiated (control switch action) tubine stop valve closure when he heard the stop valves shut.
.
2.
Turbine Bypass Valve Linkage Replacement The linkage failure in the turbine-bypass valve control system which was described in Section C.3. of CO Report No. 219/70-6 resulted in the licensee
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replacing all the aluminum horizontal push-pull bars (four) with carbon steel material. The GE LSIG Division performed a metallurgical analysis of the failure.
Mr. McCluskey stated that GE had informed him that the aluminum material used in the bar was not in accordance with design specifications.
He also stated that GE was testing an aluminum linkage made with the specified material but no results were available. Mr. McCluskey stated that efforts had been made by JC to obtain a metallurgical report of the failure, but that GE has been reluctant to provide any information on the subject.
I.
Auxiliary Systems 1.
Air Systems Discussions were held with Mr. J. Carroll and the inspector informed him of i
the problems with contaminants in air systems experienced at two other facilities. Mr. Carroll informed the inspector that the instrument air system at 0C-1 which serves the majority of the vital equipment is constructed with both carbon steel and copper pipe.
He stated that no routine dew point measurements are obtained of the air supply to determine drying tower performance, but the original preoperational test verified that the drying tower was capable of achieving a -25 F dew point. He stated that current surveillance practices include a blowdown of air receivers once per shift and a check by the operators to insure that the drying tower is operating normally and transferring properly.
He stated that there was no preventive
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i maintenance schedule *for testing the dessicant.in the drying towers or
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inspecting of filters or equipment within the system.. Mr. Carroll stated that he will review the surveillance practices in effect with the ' operations-group and. additionally' review the need to' perform preventive maintenance-on air system equipment..This item will receive followup inspection.
K.
Containment 1.. Torus Vacuum Breaker-Valves
'
~ Discussions were held with Mr. W.,Riggle to determine the nature'of the problem.with the torus vacuum breaker valves and the repairs that were made.
to-the system. Mr. Riggle informed the inspector that during surveillance
. testing on December 18, 1970, both. vacuum breaker valves failed to open.
'The butterfly valves in question are designed to open on approximately 0.5
,
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negative pressure dif ferential between the. torus /and the reactor building.
i The problem was-reported to be that the butterfly discs had locked in the seats in the closed position and that the air operators were unable to open the valves.
One of the valves was found to have an improperly adjusted linkage which resulted in'over travel in the closed direction and the other valve was reported to have a misadjustment of the butterfly stem which caused the binding condition. Both conditions were corrected and the valves successfully retested on that same date.
During surveillance testing on the following day, December 19, 1970, the valves again failed to open and manual assistance was required. Once open, they continued to operate satisfactorily.
It was noted that when the valves were left in the closed position for approximately 5 minutes, that the linkage. assembly on the operator continued to exert force on the butterfly
,
disc.and continued to drive it into the seat. An adjustment was made to-the air regulator which positions the valve operator so that when traveling in a closed direction the linkage would not tend to drive beyond the closed position.
Subsequent testing on December 20, 21, 22. 23, 24 and 31,1970 and January 7 and 13, 1971 and in monthly tests thereafter have not dis-closed any additional failures. The valves were disassembled in January, 1971 and the rubber liners replaced in both valves as the liners appeared worn.
Following the final linkage adjestment, local leak rate measure-ments.at 35 psig were conducted to determine linkage adjustments which
'
resulted 'in the minimum valve leakage while still permitting satisfactory block valve operation. -These tests determined <a total of approximately
,
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18 SCFH valve 21eakage for both valves. Technical Specifications * limits i
the total leakage rate through these valves-and the containment air purge valves (measured' leakage rate 7.98 SCFH) to 111.0 SCFH.- The butterfly
valves in question are 20 inch-Rockwell Model'LCW 2434 W, designed for.
50 psig.
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2.
Drywell Nitrogen System
'JC stated in their reply to the form AEC-592 on December 22, 1970, that the target date for the installation of a nitrogen system to supply
,
instrument air needs in the drywell was February, 1971. During the inspection, Mr. Riggle informed the inspector that the drywell N2 system is 85-90% complete. The final design will incorporate.two compressors and an air receiver. The nitrogen supply will be from the liquid nitrogen l
,
storage tank and evaporator. The system will supply (through a ball transfer valve) N2 to the drywell instrument air system. The function of the ball transfer valve is to transfer back to the instrument air supply should the drywell N system fail. Mr. Riggle stated that the necessary parts
to complete the system should be at the site in the near future and would be installed'at:the first outage of sufficient duration.
,
3.
Torus' 0xygen Sample Line. Isolation Valves During the conduct of the containment integrated leakage test in October, 1970, a considerable amount of air leakage was observed through the oxygen analyzer located downstream of the torus oxygen sample line isolation valve.
The valve was. checked and found to be closed, but the flow as measured
)
through the oxygen analyzer with an air flow meter disclosed a leakage of
'
greater than 180 CFH. The subject valve was replaced with a new solenoid
.alve of different design. The local leakage rate through the new valve
,
was measured and added to the final recorded containmsnt leakage rate.
.
Mr. Riggle informed the inspector during this inspection that both the torus
'
and the drywell 02 sample line isolation valves had been replaced.
In j
i addition, the General Electric Company is to provice a field change to provide double isolation valves in each line and additionally reroute the
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l-exhaust from the 02 sampler back to the drywell. The current status of the 2 analyzer is that i t exhausts into the ventilation exhaust from the
"
h reactor building.
4.
Drywell Leak Detection Procedures and Equipmeni *
.
The "small leak detection procedare" and equipment in use at the facility
.
were reviewed by the inspector. Also reviewed were the two unscheduled shutdowns that resulted on January 25, 1971 and March 3, 1971 from high identified and unidentified' leakage rates.
Discussions were held with Mr. J. Carroll, Operations Supervisor to determine.the leak detection procedure.s in effect at the station. The
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control room copy of the small leak detection procedure was also reviewed with Mr. Carroll. 'Pending changes to this procedure, to reflect the
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.Dresden 2. experience,'were reported by Mr. Carroll to have been circulated E
for. comment to all concerned parties, but that the final revised procedure had not yet been placed into effect. Mr. Carroll stated that the final revision to the procedure would be put into effect on that date. A copy of the revised procedure was provided to the inspector by Mr. Carroll and
- Memo J. 4 P. O'Reilly to Regions " Leak Detection Equipment and Procedures" dated March 2, 1971.1
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a review of same indicated that the procedure appeared adequate. The i
procedure included specified action points for increasing leakage rates at 5 gpm for unidentified loakage and the 25 gpm for total leakage (both identified and unidentified). If the action points were reached, it specified that the reactor should be shut down to the cold standby condition.
Leak detection equipment for unidentified Isakage consists of level probes installed in the drywell sump. These probes, and associated timer, function during sump pump out.
If the level between the two measuring probes can not be reduced by pump' operation within the specified timer setting, an alarm sounds and the pump continues to empty the sump. The alarm condition ccn only be reset in the rad waste system control room where the timer can also be read. There is a correlation between the timer setting and sump in-leakage that permits an evaluation to be made of the unidentified leakage rate. After the initial alarm is received it requires approximately 29 minutes (at the 5 gpm feak rate) to refill the sump to re-check the alarm point. The operators obtain pump flow integrator readings at four hour intervals as'un additional check on the leakage rate. Twenty-four hour integrator readings are also obtained byube
te engineering group and the operations group to plot the daily ayert
.<-leakage for this
'
source of Icakage.
As a result of the ACRS letter to JC on December 12, 1968, JC committed to provide a supplemental and redundant method of leak detection.
Station records were reviewed with Mr. D. Ross and disclosed that weekly drywell air samples have been taken for analysis during the past year.
Sample collection is achieved through both particulate and charcoal filters and
,
both filters are subjected to a gamma scan. JC has been investigating selected radionuclides identified in this gamma scan in attempts to correlate measured activities wi-th known leakage rates. To date, no meaningful results have been obtained with this method of analysis and Mr. Ross stated that the sample analysis will be continued in attempts to evaluate the method of leak detection.
.
Identified leakage into the drywell is monitored in a similar manner as described above for unidentified leakage.
A review of station records for the periods preceeding and including the two unscheduled shutdowns on January 25 and March 3, 1971 disclosed that the identified leakage rate increased to approximately 14 gpm on January 25, 1971 before the plant was shut down. The source of this leakage was determined to be a leaking cooling coil inside the drywell equipment drain tank. The coil was repaired by welding and reoriented within the tank.
Station records also disclosed that the unidentified leakage rate increased to approximately 4.5 gpm on March 3, 1971. The sourge of leakage was determined to be a pipe nipple leak in the closed cooling water system supply to the cooling coil within the drywell equipment drain tank. This nipple was repaired and operations resumed on the following date.
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Emergency Core Cooling Systems
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1.
Core Sprav Nozzle Wall Thickness Determinations
'Due to.the inspector's previously stated concerns, reinspection of t'he north core ~ spray nozzle wall thickness (by radiography) was performed -
.by Mr. N. Goodenough of the GPU Quality Assurance Staff during October, 1970. Wall thicknces was evaluated at 8 points with the minimum thickness being 0.465 inches. The required (by code) wall thickness was stated to be 0.401 inches. The GORB reviewed the, reinspection results during meeting.
No. 22 on January 15, 1971.
.
Core Spray System Water Hammer During GORB meeting No. 22 a review of plant pip'ing system vibrations.
was undertaken. The chairman of the GORB had previously requested information from the site regarding any systems that were known or suspected to have excessive vibration levels.
Mr. J. Carroll, Operations _ Supervisor, generated a list of systems and provided this list to the GORB. Based on a. review of this list, the GORB requested the GPU Quality Assurance Group to determine the feasibility of testing all'of'the welds in the core
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spray piping which had been subjected to vibration from wa t er hamaer.
GORB also requested qualitative vibration measurements for all piping systems identified to enable a determination if they are within design i
l limits.
Two GORB members were at the site on February 3,1971 to observe and measure the vibration levels on the. core spray systems. A review of a_
letter generated by these two members that was sent to both the PORC and
,
GORB members on February 8, 1971 disclosed the following information:
a.
There is no high point vent on the core spray loop piping associated with pumps A and C.
This piping is filled by watching loop pressure in the centrol room after opening the condensate fill line valve. The other core spray system does have a high point vent.
.
b.
A review of the pipe motion measurements discussed in this letter indicated pipe movements at selected points of up to 6 inches-i following pump starts.
,
c.
The letter also discussed that during the preoperational program, when severe hammers were noted, a system was added to supply
.
condensate to fill and pressurize the system to 15 psig.
It was later determined that this~ system could not be maintained in constant, use due to small leakage past the core. spray pump check valves,
[
which allowed the condensate to leak into the torus. This in turn created a water handling problem as the torus at 0C-1 is chromated
'
and the State of New Jersey has a O release limit for chromates.
'
The system is now used to fill and vent the core spray loops prior to'each monthly test.
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The chainnan of the GORB- (two other GORB members also present) was contacted byLtelephone by the assigned inspector on March 12, 1971.
Mr. 'W. Hurst, GORB chairman, informed the inspector that-GORB had requested Burns and Roe to measure and determine the stress levels in i
the core spray piping. In addition, the GORB has also recommended
- that the results of this stress analysis be used to determine the negd to perform an NDT inspection'of_ piping system materials.
I
'JC has been working with'GE on a design change to the system w'hich it is hoped will eliminate the water hammer problems. The design change
- will.inc'orporate two small pumps (one for each core spray loop) which.
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will take a suction from the torus and keep the core spray piping filled and pressurized at all times. Thus, any leakage back to the torus will not increase the level in the torus. This design change was reviewed by-It is the PORC.and has been returned to GE with additional questions.
anticipated that all questions will be satisfactorily resolved and that-
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the design change will be installed within three months.
i During the exit _ interview conducted on April 13, 1971, the assigned inspector stated that Compliance.would closely follow the progress and results of the GORB recommended actions.
3.
Core Spray System Weld Defect
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Jersey Central provided DRL in letters dated July 31, 1970 and September 30, 1970 information on the nondestructive testing report for selected balance
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E of plant systems outside the primary containment.
JC reported *that they
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were unable to locate a radiograph or acceptance report for one veld
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j located in a 10 inch line at the discharge of a core spray pump. JC
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planned to perform a radiographic examination of the weld at the earliest convenient time permitted by plant operations. During this inspection, it was determined that on November 11, 1970 the subject wgid was re-
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radiographed and two defective areas were identified. Discussions were held with Messrs. Goodenough, GPU and Kossatz, Maintenance Foreman, and
they informed the inspector that the defects consisted of both porosity and
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excessive " suck-back" in the root weld.
Defective areas were repaired by i
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grinding and rewelding and acceptable radiographs of the repairs were obtained on November 11, 1970.
A review of the repair records disclosed that the radiographic inspection had been performed by Conam. The NDT personnel were qualified 'in accordance with SNT-TC-1A requirements. 'The weld repair procedure, dated
. November 10, 1970 was also reported to be used. The welding procedure The used was qualified (in 1969) by the welder performing the repair.
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finished repair weld'was LP examined using an approved GPU LP procedure.
Mr. Kossatz stated that the film was interpreted using USAS B31.7 - 1969
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requirements..
-* Letter to DRL from JC dated September 30, 1970.
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' Emergency Power 1.
Diesel Generator Annual Inspection Station records disclosed that both diesel generators were given a thorough inspection as required by Technical Specifications (4.7.A.3)
on November 12-13, 1970 Present during the inspection was the General
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Motors representative. The inspection report from the GM representative to JC was reviewed and the only problem areas noted were that new starting
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solenoids were installed on both the No. 1 and No. 2 diesel generator's
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starters. This' item is considered resolved.
(See Management Interview
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item 4 for discussion on diesel generator shutdown devices).
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125 Volt Station Batteries The review was made by the' inspector of surveillance testing records for the A and B 125 volt station batteries. Review disclosed that monthly checks are now being performed to determine specific gravities and the temperature of every fifth cell, the voltage o:' every cell, the water level of every cell and the amount of water added. Iaese tests are now considered to satisfy the Technical Specification (4.7.B) requirements.
Discharge load tests were noted to be conducted. for both the A and B batteries in March, 1971 with acceptable measured ampere-hour capacity.
One battery cell in the B battery was replaced on January 8,1971 as a result of a lower than recommended cell voltage which was noted at the.
termination of the discharge load test conducted in September,1970.
The inspector noted during the review of the load discharge testing records for March, 1971 that the B battery terminal voltage had been reduced to less than 105 volts. The 105 volts were previously stated by Mr. Riggle to be an acceptance criterion for the test.
Mr. Riggle stated that he doubted the accuracy of the recorded terminal voltage as the summation of individual cell voltages would indicate a higher terminal voltage (at the
'
completion of the load discharge test).
He stated that this aspect of the test would be reviewed and resolved. Subsequently, Mr. McCluskey informed the inspector (telecon on May 11, 1971) that this issue had been resolved by cleaning cell interconnections and successfully load testing the B battery.
3.
Diesel Generator Starting Batteries A review of the surveillance records required by the Technical Specification (4.7. A.5) for these batteries indicate that all the required tests are now
,being performed. A review of the load discharge test records conducted on-December 4, 1970 indicated that the batteries had a measured capacity in
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excess of the 240 empere-hour requirement that was stated by Mr. Riggle to i
be the acceptance criterion for the batteries, l
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4. 'Reportability of Past Diesel Generator Problems
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Mr. McCluskey was' informed that the past instances of fdilures of the diesel generators to start on the first attempt were considered reportable
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by Compliance (3.C. (2) of license).
He stated that a written report of I
these events will be submitted to DRL in the near future.
- 5.
Diesel Generator Shutdown Devices The inspector noted in a review of the' diesel generator annual inspection records (See N.I. above) that no checks had been made to determine the'
accuracy of the set points for. devices which can shut down the diesel generator. This issue was discussed with Messrs. Riggle and McCluskey and they stated that these devices will be tested during future annua'.
inspections.
LQ.; Radioactive Waste Systems
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1. -Liquid Effluents.
i See Attachment No. 3.
2..
Independent Analysis of Liquid Radioactive Sample As a result of the discrepancies noted in Attachment No. 3, a sample of
liquid waste was split with 0C-1 for comparison analysis.
Independeht-analysis of the C0 sample was performed by the New York Operations Office (NYOO) and the Idaho' Operations Office (100).
The results of the analyses for this sample are discussed below:
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Gross Beta Analysis (uCi/ml)
OC-1 NYOO 100 1.3 x 10-5 7.46 x 10-6 (1)
3.3 x 10-5 (4)
3.6 x 10-5 ((5)
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6)
l 2.12 x 10-5 (2)
6.86 x 10-6 (3)
3.5 x 10-5 (7)
(1) Method in use prior to inspection. Counted with GM tube using
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ST-90/YT-90 as calibration source. Counted 1505 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.726525e-4 months <br /> on April 9, 1971.
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(2)
Counted with an internal proportional counter using thallium-204 as'a calibration source.' Counted 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> on April 9, 1971.
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(See Q.4. below)
~(3) Counted with an internal proportional counter using thallium-204 as'a calibration source. Counted 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> on April 12, 1971.
(See Q.4. below)-
(4) Counted by liquid scintillation technique using ST-90/YT-90 as a calibration source. Counted on April 9, 1971.
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(5). Counted with an internal proportional counter using Cs-137 as a calibration source. Counted on April 14, 1971.
(6) Counted by liquid scintillation technique using Cs-137 as a calibration source. Counted on April 14, 1971.
(7)- Counted by Cerenkov radiation detector using Cs-137 as a
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calibration source. Counted on April 14,-1971.
The gross beta analysis performed by 0C-1 resulted in a difference (between old and new method) of N 3 which indicated that release concentrations which were determined by the "old" analysis were low by a factor of 3.
In addition, the "new" method of gross beta analysis is still a factor of 3 lower than the concentration determined by isotopic analysis.
(See Q.3.b. below). These results support Mr. Pomeroy's findings as discussed in Attachment No. 3.
b.
Isotopic Aralysis (uCi/ml)
Results were corrected to the time of sampling for decay.
Isotope OC-1 (2)
NYOO (1)
100 (1) (4)
~7 Ba-140 1.7 x 10-6 4.7 x 10-7 46.9 x 10 La-140 1.7 x 10-6 1.9 x 10-6 45.5 x 10-6
.Co-60 1.5 x 10-6 1.3 x 10-6
x 10-6 Co-58 1.3 x 10-6 1,1 x 10-6
x 10-6 Mn-54 1.2 x 10-7 2.2 x 10-7
x 10-7 Fe-59 4.9 x 10-7 8.6 x 10-8
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I-131 1.6 x 10-6 1,7 x 10-6 1.7 x 10-6 I-133 el.4 x 10-6 1.8 x 10-7
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Np-239 1.1 x 10-5 7.7 x 10-7
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Cr-51 1.2 x 10-5 1.2 x 10-5 1.2 x 10-5 Zn-65 2.5 x 10-7
x 10-7
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Tc-99m 2.3 x 10-6 2.8 x 10-6
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Mo-99 e 2.3 x 10-6 2.5 x 10-6
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Ce-141 8.6 x 10-7 4.4 x 10-6
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Xe-133 1.4 x 10-5 (3)
9,9 x 10-6
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Xe-135 1.8 x 10-5 (3)
4.0 x 10-6
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TOTAL 6.9 x 10-3 4.0 x 10-5 2.9 x 10-5 (1) As the exact time of analysis was not available, a 24-hour correction for decay was applied to the NYOO results.
Similarly, a six-day correction for decay (sample counted April 14, 1971) was applied to the reported 100 results.
Both NYOO and I00 counted a liquid sample using a GE (Li)
detector.
(2)
Sample counted after evaporating to dryness using a NaI crystal.
(3)
Sample counted in liquid form.
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(4)
100 reported Hg-203 at a concentration of 0.5 x 10-6 uCi/ml and 'Ce-143 at a concentration of 4.0 x 10-5 uCi/ml. NYOO and 100 were recontacted to request that they recheck their analyses
.
in an attempt to resolve this discrepancy. No resolution was
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obtained.
3.
Gascous Radioactive Effluents
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DiscussionswereheldwithMessrs.RossandPelrinetodeterminethe licensee's method for analyzing and reporting the stack release rate.
The following information was obtained:
a.
The concentration of six radionuclides are analyzed for in grab samples which are obtained on a weekly frequency from upstream of the steam jet air ejector off-gas delay line. The six radionuclides include Xe-138, Kr-85, Kr-88, Kr-85m, Xc-135 and Xe-133.
b.
The off-gas delay time is computed by weighing a sampic of off-gas before and after spark testing. As the test is done, submerged in water, the spark test results in the recombination of the major constituents (02 and H2) of radiolytic gas. A major fraction of s
the remaining sample is assumed to be due to air in-leakage into the turbine main condenser.
Typical delay time measurements as determined by this method were approximately 62 minutes.
c.
Flow through the off-gas system is recorded on three ALP instruments in the control room. This source of measurement is not used however, in determining the off-gas flow rate. A comparison of this method with themethoddescribedinb.aboveindicated71cfmthroughthethree24P
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instruments vs 105 cfm as determined from the spark test. The use of the 105 cfm value would result in the shortest holdup time determination.
d.
The stack gas monitor is calibrated each week. The calibration is performed using the results of the grab sample analysis which is corrected for delay ta determine the stack activity.
OC-1 uses a, i
stan'dard 1.9 x 105 cfm dilution flow in the stack. A counts per second/
microcurie per second correction factor is determined for the stack gas monitor on a weekly basis.
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Results reported in the semi-annual report to the AEC are derived from e.
weekly "eynhall" integrations of the stack gas recorders.
i f.
Trip functions for the stack gas monitor were currently set at 300 cps, which corresponds to approximately 0.03 Ci/sec. This was noted to be one decade below the annual average release rate specified in Technical Specification 3.6.A.1.a.
The'second high-high alarm is set at 1,000 cps which corresponds to approximately 0.1 Ci/sec, which is one. decade below the 48-hour release rate specified in Technical Specification 3.6.A.1.b.
g.
Off-gas monitor alans set points are set to trip at approximately 0.3 Ci/sec for a high-high alarm and 0.03 Ci/sec on the high alarm.
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Mr. Ross was questioned as to why OC-1 did not utilize the GE recommended practice of measuring the activity for the sum of the six gases analyzed and converting this (by the GE-supplied computer code' printout)1to the total activity contained in 22 gases which are present. - Mr. Ross did not
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indicate an awareness of this GE code, but Mr. Pelrine' stated that he was aware of the code and additionally provided a copy of the code. In the i
inspector's review of information, it appeared that a 10-15% (low)
deviation would result from measuring only the'six gases for the current values of holdup time. The inspector was informed (telecon May 11, 1971).
following the inspection that the code conversion to determine the activity of 22 gases would be used in the future.
4.
OC-1 Calibration Error - Gross Beta Counting The calibration error discussed in. Attachment No. 3, item A, was reported to the inspector by Mr. D. Ross, to be corrected on April 8, 1971. The new calfkration. source (standard) for gross beta analysis was reported to involve the use of thallium-204 deposited on a 2-inch diameter planchet.
I Gross beta analysis will be performed in the future at,OC-1 by use of an internal proportional counter. -The counting efficiency of this counter l
has been measured at 56%.
5. -Carbon-14*
~Mr. Ross informed the inspector that a sample of liquid radioactive waste
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was sent to Isotopes, Inc. in February, 1971 for C-14 analysis. No results have been reported back on this sample.
Mr. Ross also stated that a sample will be sent to Radiation Management Corporation for analysis and OC-1 will attempt an analysis with their liquid scintillator. The results of the analysis will be reviewed during the next routine inspection.
6.
Control of Curie Content in Outside Storage Tanks (Item of Noncompliance)
Section C of Attachment No. 3 discusses an apparent item of noncompliance with Technical Specification 3.6.C. which states that "the maximum amount of radioactive liquids in storage waste sample tanks, the floor drain sample tanks, and the waste surge tank shall not exceed 0.7 curies". A gross beta analysis (which has been established as being in error by at least'a factor of 6 from the total activity) performed on March 10, 1971
_
indicated 0.316 C1.
Scaling up this analysis result by the factor of 6
.would yield 9e 1.8 Ci vs the allowabl,e 0.7 Ci limit.
Discussions were held at exit interviews on April 8 and 13, 1971 and
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Messrs. McCluskey and Ross stated that the C1 content of these tanks will be controlled in the future on a gamma spectrum analysis.
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- CO Report No 219/70-7, Other Significant Items No. 10.
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Experiments and Tests Mini-Stretch Power Testing - Meeting minutes indicated that the PORC reviewed and approved the test procedures that were used for the power level increase
~from 1600 to 1690 MWt* during their meeting on November 20, 1970.
It was reported by Mr. D. Ross that the test procedures had been developed by Mr. D. Hetrick's group at GPU.. This group was reported to have also reviewed the test results and provided test result. summaries.
Mr. Ross stated that all test results had met the acceptance criteria.
U.
Miscell'aneous Aircraft Overflights - Mr. McCluskey informed the inspector that he had no knowledge of any scheduled military or commercial aircraft overflights at the 00-1 facility.
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- Described in Amendment No. 55 to the license.
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JERSEY CENTRAL POWER & LIGHT CO.
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CO Report No. 219/71-1
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Attachment No. l'
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' TECHNICAL GROUP D. A. Ross Technical Superintendent
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Staff Engineers Technical Engineers Chemical Supervisor Rad. Prot. Supervisor E. O' Conner E. Growney J. R. Pelrine D. E. Kaulback
-D. Reeves J. L. Sullivan D. Allred (on Loan JNC)
i Assistant Technical Engineers
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2 Vacant Positions Chemical Engineer R. Stoudnour
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Associate Engineers
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J. Budd
A. Rone E. Rosenfeld Chem. Technicians Rad. Technicians
1 Vacant Position R. Neuman J. Cook I
R. Peterson J. Scull i
C. Konta M. Oberstadt
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JERSEY CENTRAL POWER & LIGHT CO.
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C0 Report No. 219/71-1 Attachment No. 2
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PRESENT MAINTENANCE ORGANIZATION Maintenance Supervisor
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W. Riggle i
Instrument &
Mechanical Foreman Electrical Foreman F. Kossatz T. Johnson Machinist (1)
A Instrument Technician (4)
Welders (2)*
B Instrument Technician (2)
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A Mechanics (2)*
A Electrician (2)
B Mechanics (2)
B Electrician (2)
Station Helpers (8)
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- One additional person planned to be added in each category.
Attachment No. 2
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UNITED ST A1ES
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ATOMIC ENERGY COMMISSION
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W ASHINGToN. O C 20 W
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AFR 2 31971 R. T. Carlson, Senior Reactor -Inspector, 00 I -
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THRU:
H. R. Denton, Chief, Technical Support Branch, CO:HQ 4,
<
ASSIST INSPECTION - OYSTEk CREEK LIQUID WASTE SYSTEM, DOCKET No. 50-219
j Int roduction This memo contains the results of an assist inspection conducted by Mr. D. Pomeroy, CO HQ, on April 6-8, 1971, of the liquid waste system Dr. C. Ps11etier, CO:HQ, accompanied the inspector for at Oyster Creek.
The assigned inspector, Mr. R. McDermott, participated in the trainin g..
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discussions of the noncompliance items with the licensee's management.
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Scope The inspection followed the procedures in PI-3300/1 for liquid waste j
systees with particular emphasis on Section 3315.04.C which required a
f acilities are being
)
determination of the extent to which waste treatment e
utilized. Operations for 1970 and for the first three months o' 1971
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Within these areas the inspection consisted of selected were reviewed.
examination of representative records, interviews with plant personnel, I
and observations by the inspector.
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Principal Results A.
Monitoring of Radioactivity in Liquida
The licensee discharges liquids in batches and each batch is sampled and analyzed prior to release to the discharge canal. The analysis
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consists of the evaporation to dryness of a small quantity of the j
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waste semple on a 2-inch diameter planchet and the subsequent count-ing of the planchet with a 1-1/4-inch diameter thin window G-M tube
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(end window or gross beta counter). The counter has been calibrated
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and found to have a 10 percent efficiency with a 1-1/4-inch diameter strontissa 90 disc source. This counting efficiency is used for the
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2-inch diameter planchets even though more than half the area of the l
planchet is not under the detector.
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In order to determine the adequacy of gross-beta counting an examina-tion of the records of a complete gamma analysis was made for the batch whose analysis had been summarized in the licensee's third
i JERSEY CENTRAL POWER & LIGHT CO.
CO Report No. 219/71-1 Attachment No. 3
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t Page 1.of 12 pages.
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-2-semiannual repert. The activity in the batch saalyzed in the semi-annual report totalled 9.6 x 10-4 uci/ml. The release records showed that the concentration for this batch determined by the routine method was only 1.41 x 10-4 pCi/m1, indicating a y/g of 6.8.
Mr. D. Ross, the technical supervisor, stated that the reported gasuna analysis was the only one performed during 1970 and that they were now i
performing isotopic analyses about twice a month. An examination of the records of gamma analyses conducted during 1971 was made. For l
those ar.alyses where a gross-beta count was also available a comparison is made in table I which indicates an average v/S of over 6.
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Mr. Ross also stated that a check of the and window counter with thallium 204 on a 2-in h diameter planchet yie'ided an efficiency of about 4 to 5 percent.1-This result indicates that Oyster Creek gross-beta monlysis is low by a factor of 2 to 2-1/2. Results of the licensee's analysis of waste water on their internal proportional i
counter indicate that their end window gross-beta analysis produces
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values that are low by about a f actor of 3.
The results from a sauple the licensee split with Isotopes Incorporated indicates that the licensee's routine gross-beta analyses are low by a f actor of 5 to 7.
At the time of the inspection the licensee was still controlling and formally recording the release and storage of liquid wastes based on the end window gross-beta analysis at a 10 percent efficiency. It appears that the licensce's surveys conducted to evaluate the radia-tion hazards incident to the presence and release of radioactive materials are inadequate and, therefore, in non-compliance with Title 10 CFR Part 20.201 (Surveys).
B.
Control of Radioactivity in Liquids Discharged An estimate of the actual concentrations and amounts of radioactivity at Oyster Creek can be made considering the monitoring errors reported.
above. These es Nine Mile Point. pmates may then be compared with recent data from The Oyster Creek results indicate that their gross beta, when compared to total setivity, is low by about a factor of
~ 7. The results also indicate that beta counting errors above account for about a factor of 2.5.
The factor of 3 remaining is then presumed to be the gamma activity not counted by the gross beta detector or true v/S ratio. The results from Nine Mile Point indicate a factor f1/ Nine Mile Point usec a 4% efficiency for their end window counter by calibration with a 2-inch diameter thallium 204 source.
2/ Memo Pomeroy to Carlson March 27, 1971, Assist Inspection - Nine Mile Point.
I i
JERSEY CENTRAL POWER & LIGHT CO.
CO Report No. 219/71-1 l
Attachment No. 3
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Page 2 of 12' pages
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of 2.5 more beta than Oyster Creek (4% Vs 10% counting efficiency) and l
the average y/8 ratio for the past 14 months at Nine Mile Point has l
been about 3.
nis confirmation of the limited Oyster Creek data l
indicates that the overall f actor 7 is reasonable.
Using this factor the 4.5 gross beta curies reported by the licensee I
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l (and shown in table I) for 1970 would be 31.5 total curies. The
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Technical Specifications (3.6.B.1.a) state that the annual average shallnotexceed1x10~pfiedradionuclidesinthedischargecanal l
concentration of umident l
pCi/cc. The basis for the specification considers conditions of minimum flushing in the bay and reconcentra-tion in biota. De basis defines the specification to mean a limit of "0.067 curies per day or a yearly limit of about 25 curies." On
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l the gross beta basis the lic9nsee's report indicates an annual average concentration of 0.166 x 10~
pCi/m1, the factor of seven t fortotalcurieswouldindicateconcentrationof1.16x10"9. account We licensee, therefore, appears to be in noncompliance with the Techni Specification requiring that the discharge canal not exceed 1 x 10"yal pCi/cc on an annual average basis.
I It should be noted that if the Technical Specifications required only control on gross beta (e.g., Nine Mile Point and Dresden) the licen-see's report would be low by only the gross beta error factor of 2.5, l
yielding 11 curies and 42% of the annual limit. It should also be
noted that if isotopic identification and mixture MPC's were allowed by the Technical Specifications, the one sample for 1970 indicates that 1131 is controlling and that the mixture NPC is about 18 times larger than the 1 x 10~I pc/mi required. On this basis the curies
released are still 31.5 but the percent of annual limit is only 4 6%.
C.
Control of Radioactivity in Stored Liquids l
Technical Specification 3.6.c limits the radioactivity stored in the l
five outside tanks to 0.7 curies. An examination of the surveillance records indicated that the required sampling and analyses were being performed (required by Tech Spec 4.6D - every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). W e limit of 0.7 curies is considered by the licensee to include tritium. We basis for this specification indicates that only the unidentified
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Semiannual Report No. 3 JERSEY CENTRAL POWER & LIGHT CO, C0 Report No.- 219/71-1 Attachment No. 3 Page 3 of,12 pages
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radionuclides'were considered, and it is, therefore, not necessary to include tritium in this calculation. The review of recent sampling indicated that most results were in the 0.1 to 0.5 curia range, including tritium'
The higher numbers usually indicate that the.
.
100,000 gallon surge tank is being used. The highest recent result is listed below:
.Date Gross Beta Tritium Total 3/10/71 0.316 Ci 0.347 0.663
'
The gross beta is determined by using the same methods and is subject to the errors discurmed above.
Applying a correction f actor of 7 to
'
the gross beta analysis would result in greater than 0.7 C1 of activ-ity stored in the outside tanks, even if tritium was excluded. The licensee, therefore, appears to be in noncompliance with the 0.7 curie limit for storage of radioactive liquids.
D.
Continuove Liquid Monitor An examination was made of the calibration records for the continuous liquid monitor. Also examined were plant operating procedures for liquid effluent releases and records of actual releases. From these it appaared that a normal release of '1 x 10-3 pCi/mi water would be mad-
..c 10 gpm and that the continuous liquid monitor would be expected to read about 2,000 counts per second above background. Mr. Ross stated that considerable trouble had been experienced with the back-ground of this instrument and that the background was currently 14,000 counts per second and that the alarm was set at 100,000 eps.
Mr. Ross stated that the high background was caused by a buildup of contamination in the spool of pipe that the detector is mounted on and that a replacement spool of stainless steel is being fabricated.
The instrument has only been calibrated to 5,000 cps. Using this calibration, an estimate of the concentration required to trip the a
l alarm at 100,000 cps of about 1 x 10-1 pCi/cc appears reasonable.
l This concentration is higher than the current primary coolant activ-
,
ity and with this setting and the current volumes of waste, the
!.
facility could release at the rate of 400 curies per year without
~
tripping the alarm. ' Mr. Ross agreed with the inspector's statement that the instrument appeared to be out of service. Technical Speci-
. fication 3.6.B.2 states that " radioactive liquid effluents being
.
released into the discharge canal shall be continuously monitored."
Therefore it appears that the licenseo has been in noncompliance
with this Technical Specification in that the high background, set-
'
point problem has rendered the continuous in-line monitor incapable of performing its intended service.
'
JERSEY CENTRAL POWER & LIGHT CO.
'
CD Report No. 219/71-1 Attachment No. 3
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'Page. 4 of 12 pages
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E.
System Performance A review of. the liquid radwaste system design objectives was not made because the licensee has been unsuccessful in detormining what the objectlyes are. The process design drawings usually available for i
review 2/ at the site apparently do not exist ' for the Oyster Creek station. A review of the expected performance of the radwaste system as outlined in the FD and SAR was made and compared with the system performance. The following substantial variances were observed:
1.
Waste Coller. tion System Section IX - 4 Paragraph 4.2 of the FD and SAR states that waste collection drains are " essentially always reused as condensate."
During 1970 less than 10% of waste collection drains were reused as condensate. The majority of this water was released to the circulating water system. Mr. Ross stated that a chemical engineer (Mr. B. Stoudnour) was added to the plant 's staff in December 1970 and that his first priority was to make condensate from all waste collection waters. Table II indicates that approximately 50% of this water is now being recovered.
2.
Floor Drain Activity Table IX - 4-2 of the FD and SAR lists the expected concentrations of radioactivity in the floor drain system f *r an offgas release
.
"
performance.
Interpelation in this table 5 fo
'
rate of 10,000 microcuries per second (which is typical of station performance for December 1970 to March 1971) indicates that the expected floor drain concentration is 0.5 x 10-4 pCi/ml. Actual concentrations range from 10 to 50 x 10-4 pCi/ml or 20 to 100 times the expected value.
3.
Chemical Waste System Section IX 3.2, paragraph 3.1.3.3 of the FD and SAR states that the chemical waste will be sampled and concentrated as required.
Mr. Ross stated that these tanks had never been sampled and that during 1970 essentially no concentrating of these wastes were perforised. The wastes were rsleased to the circulating water
.
4_/ Mean Pomeroy to Thornburg, Dec. 4,1970. Assist Inspection Dresden 2.
. 5/ : As shown on Figure 1.
JERSEY CENTRAL POWER & LIGHT CO, CO Report No. 219/71-1
,
'
Attaciunent No. 3 Page 5 of 12 pages
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system through the floor ~ drain system. ' Mr. Ross stated that the evaporator is now operational and the chemical wastes are being -
concentrated and that it is planned to use the evaporator to the i
j fullest extent practical by concentrating floor drains during l
periods when it is not being used for chemical wastes.
l 4.
Isotopic Distribution
!
Table IX - 4-3 of the FD and SAR lists the concentrations of
>
various isotopes expected in the liquid effluents for operation j:
with and without fuel failures. The expected iodine 131 for
!
operation with failed fuels (100,000 pC1/see stack gas) is listed
'
I as 0.03%.
Iodine 131 is not expected in the absence of fuel f ailures. The Oyster Creek semiannual report for the period ending December 31, 1970 lists the isotopic composition of a floor drained sample and indicates that iodine 131 comprises approximately 16% of the activity in this batch. Other isotopes are also a* substantial variance from the expected values.
'
The licensee appears to be in noncompliance with section 3.C(2) of
,
the facility license in that he has not reported the above four sub-
'
stantial variances disclosed by operation of the facility from the L
performance specification contained in the FD and SAR.
l Exit Interview
'
L l
During the exit interview at the conclusion of the inspection Mr. McCluskey was infomed that the abnormal performance of the radwaste system was con-sidered reportable under the license. He stated that they were not in a
,
j position to challenge the alleged variances and that they will consider j
reporting them.
l Mr. McCluskey was informed that we consider the continuous liquid effluent
monitor to be out of service and that liquid releases without the monitor
to be in noncompliance with the Technical Specifications. Mr. McCluskey i
stated that a fix was smderway which would reduce the background and restore i
the instrument to an operable condition. Mr. McCluskey phoned the instru-
{i '
,
ment shop during the interview and reported that the installation of the
'
stainless steel spod had been completed and that the system was being i
.
hydroed and that the instrument would be operational today.
l l
Mr. McCluskey was~ informed that we consider the " surveys" (that is, the
'
counting methods) currently being used to control the release and pres-ence of radioactive material in liquids to be inadequate and that a
,
JERSEY CENTRAL POWER & LIGHT CO.
CO Report No. 219/71-1 J
Attachment No. 3
'
Page 6 of 12 pages i
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-7-review of the station records indicate that the current procedures are
]
underestimating the activity released and stored by about a factor of 7.
He was also informed that if the factor of 7 is the appropriate correc-tion the annual average limit for liquid released during 1970 was exceeded; the current release rate indicates that the 1971 level will also be exceeded and that the curie limit for the outside tanks has been exceeded on a number of occasions. Mr. McCluskey responded by saying that they are aware of certain inadequacies in their counting techniques and that they are not surprf sed that we found this problem. He stated that he believed that they have made an honest mistake; that it was not recognized at the time of the December 16 sample and not fully realized until February and that they are taking action to correct the problem but they are not ready to change the basic system at this time. He outlined the progress on the problem by stating that changes had been
,
made in the gross-beta system to reduce the loss of iodine by adding
'
silver nitrate to the solution before drying and by drying of a smaller sample volume to reduce coincident counting errors. He also stated that their bi-monthly isotopic analysis, their internal proportional counting and the sample sent to Isotopes Incorporated were all demonstrations of their effort to solve the problem and obtain a conservative method for controlling radioactivity in liquids. He stated that they were not pre-pared to change the method yet and offered no time schedule. No commit-ment was made concerning the correction of the previously reported data.
A
- ff&n I
D. L. Pomeroy l
Nuclear Engineer Technical Support Branch, CO
i
!
cc:
A. Giambusso, CO
>
R. H. Engelken, CO
'
L. Kornblith, CO i
J. P. O ',Reilly, CO j
N. C. Mose le y, CO :I i
W. Seidle, C0:11
!
G. Fiore111, C0:111 i
H. D. Thornburg, CO:III J. W. Flora, CO:1V'
G. S. S pencer, CO :V
,
!
f
JERSEY CENTRAL POWER & LIGHT CO.
CO Report No. 219/71-1 Attachment No. 3
,Page 7 of 12 pages
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TABLE I
'
l Floor Drain Tank Samples
'
Gross Beta Vs Total Activity
'
.
t Gross-Beta (I Total Date uC1/mi Activity uC1/mi Difference t-4 (2)
12/16/70 1.41 x 10 9.6 x 10 6.8
~
-4
~4 1/26/71 4.04 x 10 17.4 x 10 4 +3
~4 2/6/71 1.96 x 10F 22.5 x 10 11.3
~4 16.5 x 10 '
52.7 x 10 3.2
~
3/18/71
.
'
i
!
1 (1) Based on 10% counting efficiency
'
i (2) Reported in semiannual report No. 3 i
i i
,
JERSEY CENTRAL POWER & LIGHT CO.
CO Report No 219/71-1 Attachment No. 3 Page 8 of 12 pages
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TABLE I Floor Drain Tank Samples Gross Beta Vs Total Activity
.
I Total Gross-Beta Date UCi/mi Activity pCi/mi Difference (2)
12/16/70 1,41 x 10 9.6 x 10 6.8
.
1/26/71 4.04 x 10 17.4 x 10 4.3 2/6/71 1.96 x 10 22.5 x 10 11.3 3/18/71 16.5 x 10 52.7 x 10 3.2
.
l
-
.
(1) Based on 10% counting efficiency (2) Reported in semiannual report No. 3
,
.
JERSEY CENTRAL POWER & LIGHT CO.
CO Report No 219/71-1 Attachment No. 3 Page 8 of 12 pages
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OfSTER CREEK LIQUID WASTE SYSTEH TABLE II Waste Collection System Floor Drain System Callons x 10J Gallons x 103 Gross Beta 1970 Curies input Dumped Dumped Jan 0.283 560 All 763
"
Feb 0.424 460 574 Mar 0.196 690 520 729 Apr 0.185 600 All 448
"
May 0.060 480 492
"
June 0.613 700 418 July 0.721 680 509
"
Aug 0.461 860 646
"
"
Sept 0.353 780 523
Oct 0.392 800 438
"
Nov 0.495 827 740 406
Dec 0.320 980 340 333 j
Total 4.5 6,300
1971 Jan 0.354 900 120 472
,
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Feb 0.518 940 518 245 Mar 0.952 e 900 680 285 JERSEY CENTRAL POWER & LIGHT CO.
CO Report No. 219/71-1 Attachment No. 3 Page 9 of 12 pages
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TABLE III I
Frimary Coolant Activity - WCf /ml
)
Total Iodine Tritium Grossgeta 10-3 10-4 10-j Date 1970
<.
I Jm.
5
Feb.
5
<
f Mar.
8
Apr.
7
May 03-12 0.06-17 2-4
June
15
July
20
Aug.
25
'
Sept.
25
Oct.
35
.
Nov.
'40
'
Dec.
40
,
1971 Jan.
50
!
Feb.
50*
9
- Method change - reduced voltsse of sample to reduce coincidence losses.
JERSEY CENTRAL POWER & LIGHT CO, CO Report f:o. 219/71-1 Attachment No. 3
Page 10 of 12 pages
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TABLE IV
'
Sample Comparison
Gross Beta DCi/ml x 10 '3
,
.
Sample O. C. No_rm_al 0. C. IPC Isotopes Inc.
A 1.27 4.40 8.88 B
1.26 4'.72 6.91 I
.
Notes:
i
Oyster Creek Normal - Sr
- 10% eff.
>
Oyster Creek IPC
- Internal proportional counter
,
i
i l
TABLE V j
l Estimate of Liquid Releases
,
i Reported Correction Est.
Date Curies factor Actual Ci 1970 4.5 6.8
'
,
Jan '71 0.35 4.3 1.5
,
Feb '71 0.52 11.3 5.8 I
Mar '71 0.95 3.2 3.0
JERSEY CENTRAL POWER & LIGHT CO.
C0 Report No. 219/71-1 Attachment No. 3 Page 11 of 12 pages
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OfSTER CREEK LIQUID WASTE SYSTEM i
TABLE 11
'
Waste Collection System Floor Drain System Callons x 103 Gallons x 103
,
Gross Beta 1970 Curies laput Dumped Dumped
-
Jan 0.283 560 All 763
"
Feb 0.424 460 574
Mar 0.196 690 520 729 Apr 0.185 600 All 448 i
e.I May 0.060 480 492
"
J j
"
June 0.613 700 418
"
July 0.721 680 509
.
Aug 0.461 860 646
"
'
.
h-Sept 0.353 780 523
"
Oct 0.392 800 438
"
Nov 0.495 827 740 406 Dec 0.320 980 340 333 Total 4.5 6,300 1971
Jan 0.354 900 120 472
Feb 0.518 940 518 245
'
Mar 0.952 f 900 680 285 JERSEY CENTRAL POWER & LIGHT CO, CO Report No. 219/71-1
.__ Attachment No._3.
._
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Jersey Central Power & Light Company-(Oyster Creek 1)
(Comunercial Operation - Dec.1969)
Scram fl> ate of flo. -
Heram
_lwer Level Primary Cause Conunents i
12/il/69 1600 MWt Equipment low water level trip resulting i.
Failure from malfunction of three element controllers on feedwater flow
'
caused by stuck excess flow check
valve on steam line sensor.
1/ 3 /70 1420 MWt Equipment Condenser low' vacuum trip caused Failure by in-leakage through the expansion
joint between "C" condenser and the
'
,
exhhust section of the
"C" low pressure turbine.
1/ 9 /70 1600 MWt Equipment Low water level trip resulting from
' Failure a loss of the steam flow signal to the feedwater control system.
-
2/15/70 1350 MWt Design Following a load reduction,.high
. Inadequacy level in the moisture separator drain tank cause a turbine trip and a resulting reactor high neutron flux scram.
'
5 2/18/70 160 MWt Operator fligh neutron flux in the IRM's while s 'or holding the reactor in hot standby.
i
4/ 7 /70 1600 MWt Operator Main steam isolation valves closed
,
Error due to a steam-line low pressure
'
trip signal. While making a trip test of the spare exciter breaker, a short tripping three (3) re-circulating pumps occurred lowering the D.C. control voltage. The slow closure of the valves caused the pressure to momentarily drop below 850 psig, thereby causing the low steam pressure signal.
4 /14/70 1600 MWt Equipment Low water level trip resulting from Failure EPR pressure oscillations.
'
4/19/70 160 MWt Operator Neutron monitoring scram in the Error IRM-APRM overlap range while shutting down plant.
'6/17/7b 1600'.MWt Equipment Main steam isolation valves closed Failure-after receiving a main steam line break signal due to high ambient tunnel temperatures.
JERSEY CENTRAL POWER & LIGHT CO, CO REPORT NO. 219/71-1
~
Attachment No. 4 Page 1 of 2 pgs
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Oyster Creek I.(C<nt'd).
.
Scram Date of-Corrunents No.
Scram Power Level Primary cause 1'O
'7/11/70 960 MWt Natural ~
Turbine load was reduced manually to 200 MWe and a turbine trip manually Phenamonom initiated. Reactor scram caused by high pressure following turbine trip.
Ileavy grass at intake structure required load reduction and turbine trip.
>
Power oscillations occurred at 1600-MWt,
9/17/70 1200 MWt Equipment reactor power was reduced to 1200 MWt Failure by reducing recirculation flow..A turbine trip occurred caused by high
level in moisture separator drain tank.
,
The reactor scranmed on high flux.
MPR malfunction caused low main steam
9/22/70 1520 MWt Equipment Reactor scrammed on Failure line pressure.
MSIV closure.
Electrical load was reduced to 290 MWe
10/ 2 /70 1000 MWt Design Inadequacy to repair condenser leak. Turbine tripped on high moisture separator drais tank level. Reactor scrammed on high pressure.
Rapid closure of #2 stop valve during
12/25/70 1690 MWt Equipment valve test caused actuation of turbine'
r Failure trip anticipatory scram.
After complete check of all systens
12/25/70 810 MWt Equipment pertinent to stop valve closure and Failure anticipatory trip, #2 stop valve was retested at light load with a repeat
of the event described in scram #90.
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JERSEY CENTRAL POWER & LIGHT CO.
.i CO REPORT NO. 219/71-1 Attachment No. 4 '
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Page 2 of 2 pages.
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3944
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(t*,4 April 26, 1971
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J. P. O'Reilly, Chief, ~ Reactor Testing & Operations Br.
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Division of Compliance, Headquarters
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INQUIRE MEMORANDUM NO. 219/71-D
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JERSEY CENTIAL POWER & LIGHT CCB(PANY (OTSTER CREEK 1).
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-LOSS OF HEATING BOILER - COLLAPSE OF HEATING BOILER STACK
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The following enformation was received by the assigned inspector, Mr. R.
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McDermott,'in a telephone call from the plant superintendent, Mr. T.
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]p McCluskey on April 23, 1971. Supplementary information concerning the occurrence was obtained in a telephone call to Mr. D. Ross, Technical
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Supervisor, by Mr. Higginbotham on April 26, 1971.
3 1.
During the period April 20-22, 1971 it was noticed that the fined l
particulate and charcoal filters on the stack monitor required j
changing at a shorter frequency than was normal. The air flow
through the filters is self-campensating to maintain a camin=1
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2 cfm; an alarm is received in the control room when this flow drops j-and stays below normal.
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2.
Investigating possible causes for this occurrence, a maintenance man inspect.ed the inside of the main ventilation stack and found that the ventilation stack for the heating boiler had collapsed.
(The heating
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boiler stack was fabricated from " light weight asbestos-like" material.)
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l This was found and the heating boiler was secured on Thursday, April 22, 1971.
i 3.
This caused soot and flue gas to be picked up in the sample probe and
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lines where it was deposited on the filters of the stack sampler
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restricting the air flow.
l 4.
Inspection of the interior of the stack revealed visual damage to the
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support structure for the isokinetic sampling probes for the main stack I>,,_
monitor. The licensee intends to repair this damage as soon as it is Possible to do so.
However, the repair will kequire a shutdown of the
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plant.
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5.
Inspection of the chart trace of the stack noble gas monitor showed that j
there has been no discernible change in indications of gas releases. The
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normal, for several weeks, indication has been about 100 to 150 counts /
L second (equivalent to about.10,000 to 15,000 uCi/sec); no particular -,
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deviation from these nominal indications were found.
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N Flant Operations Review Cossaittee reviewed the oscurrence and,
"$.06 based on information in itesa 5 above, desided that plantJoperation M
eeuld continue.
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l 7.
At the time the heating boiler was secured on Thursday, April 22, 1971, dresh particulate and charcoal filters were installed in the stack
1 monitor. Gese filters vers +='- xM on Honday, April 26, 1971, and the results of counting and analysis of these filters w n 1 be
compared to results of filter counting during and prior to the
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occurrence. An attesept will be made to ascertain if the "cred" deposited on the filters affected the evaluation of releases of
particulates and halogens during the time the heating boiler stack d
. jg, was damnaged.
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8.
A tasaporary stack was erected for the boiler, outside the main ventila-
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tion flua, and the boiler was returned to service Sunday evening, i
j April 25, 1971. The use of the radioactive liquid wasta caneaarrator was lost during the period of time Thursday to Sunday evening.
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j 9.
Inspection by plant personnel revealed no visual structural damage to i
the main ventilation stack.
The licenses plans to subadt a written report of the occurrence to DRL within r...+ 30'dayoita-asserdamse with.the requirements ef the,das111ay 11ammas. 1The.-
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matter will be reviewed further during the next routine inspection of the facility and C0:10 will be informed of additional details as appropriate.
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R. T. Carlson l
Senior Reactor Inspector
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cc:
E. G. Case, DRS (3)
P. A. Morris, DRS
R. S. Boyd, DRL (2)
i R. C. DeYoung, DRL (2)
, D. J. Skovholt. DRL (3)
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l l P. W. Ikme, DRL (2)
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A. Giambusso, CO L. Kornblith, Jr., CO R.11. Engelken, C0
Regional Directors, CO REG Files
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3944.
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l 31 March 24, 1971
l J. F. O'Reilly, Chiaf, Reactor Testing and
Operations Branch, Division of Compliance, HQ INQUIRY MEMORANDtB( 219/71-C
JERSEY CENTRAL PCWER AND LIGIT CCRIFANY (0YSTER CREEK 1)
i ISOLATI(Et C(alDENBER ISCLATIGt RELAY FAILURE j
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Mr. T. McCluskey, Station Superintendent, informed the assigned
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I inspector, Mr. McDemott, during a telecon on March 23, 1971, that
during routine surveillance testing of the isolation condenser
isolation feature on that date, one of the five second time delay
relays in the isonation logic failed to operate. The ability to
isolate the isolation condensers was not lost but the relay failure resulted in a loss of redundancy. The subject relay.,vas manufactured
j by Nimmatic Corporation and was exchanged for an Agastat type rday.
JC is currently planning to change thrde additional Numatic relays in the initiating logic. Mr. McCluskey stated that JC would submit
M a 30 day report of this event in accordance with license require-
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meats. We plan a routina site inspection during early April 1971, at which time this event will be reviewed further.
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R. T Carlson Senior Reactor Inspector
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cct E. G Case, DRS (3)
F A. M rris, DRL
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R. S. Boyd, DRL (2)
R. C. DeYoung, DEL (2)
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D. J. Skovholt, DRL (3)
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A. Giambusso, CO L. Kornblith, Jr.. CO
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R. H'. Engelken, CO l
Regional Directors REG File
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,e t/jd Carlson SURNAME >
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Forum AEC-31s (Rev,9-53) ABCM 0240 -
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3944
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March 2, 1971
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J. P. O'Railly, Chief, Reactor Testing de Operations Br.
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l Division of compliance, Headquarters INQlmtf MEMORANDON No. 219/71-5 JERSEY CENERAL POWER & LIGHT CCHPANY (0YSTER CREEK 1)
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MAIN STEAM ISOLATION VALVE PERFORMANCE - FAILURE OF VALVE TO MEET LIMITING LEAK RATE
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i The assigned inspector, Mr. R. McDermott, was contacted by Mr. T.
2,.. *
McCluskey, station superiataa4= on February 19,19717 and informed that, during a recent outage (to inspect turbine control valves), all
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sain steam isolation valves had been tested for leak tightness in
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i accordance with prior commitments made in their letter to DEL dated
March 20, 1970 Mr. McCluskey stated that the valves were tested using
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20 pai air and that the two inside valves shound no detectable leakage.
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The measured leakage for the north outside isolation valve was 3.6 SCFN.
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He reported that the final test on the south outside isolation valve measured 4.3 SCFM. The technisal specification leakage limit for an individual valve is 11.5 SCFH.
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Mr. McCluskey stated that the initial closure of all main steen isolation
valves was accomplished with less than 100 peig in the main steam system and with no stama flow 2 He stated that the first two tests conducted on the south outside main steam isolation valve were unsuccessful. The first and second test resulted in a measured leakage oft MS80 SCFH. This isola-
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tion valve was cycled between these two tests at atmospheric pressure
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conditions in the main stema system. The final leakage test (results -
l 4.3 ::CFH) for this valve was conducted following a hydrostatic test on the F
primary system without further valve cycling. Mr. McCluskey stated that his basis for confidanos that the valve would meet the leak tightness
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l requirements of technical specifications was that it was tested in October, 1970 and found to be leak tight and that the station's prior experience
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with this valve indicated that it would not be leak tight if closed when
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there was less than 100 pai steam pressure in the, system. He further
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stated that normal steam flow through the main steam isolation valves is in such a direction as to assist the closure of the valve and improve the leak tightness measurements, If the valve were to close with stoma flow to
the turbine. He stated that the timeng of the closure of the main steam
isolation valve did not indicate any abnormalities with past performance and
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that there were no obvious mechanical difficulties with the valve. Operation
of the reactor was resumed on February 19, 1971.
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- laquirymemorandumsubmittedatrequestofJ.Keppler,COHdonMarch1,1971.
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McDERMOTT CARLSON sumuur >
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Mr. Hocluskey was resentacted on March 1, 1971, and agreed to submit,
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30 days required by the lisense. We plan to review these leek tests in a written report on the failure of the south outboard to DRL within the
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more detail during the name routine inspection scheduled for the week of l
March 15, 1971.
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j R. T. Carlson
Senior Reactor Inspector
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cc:
E. G. Case, DRS (3)
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P. A. Morris, DEL l
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R. S. Boyd, DEL (2)
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R. C. DeYoung, DEL (2)
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D. J. Skovbolt, DRL (3)
P. W. Howe, DEL (2)
A. Gianbusso, CO L. Kornblith, Jr., CO l
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R. H. Engelken, CO Regional Directors, CO REG Files
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