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REGION I DIVISIO.. Or COMPLIANCE Report of Inspection I
l C0_ Report No. 219/70-3 i
Licensee:
JERSEY CENTRAL POWER AND i,IGHT COMPANY i
Oyster Creek 1 t
License No. DPR-16 Category C Dates of Inspection:
March 18, 19 and 20, 1970 i
l Date of Previous Inspection:
February 5, 1970 Inspected by:
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D. L. Caphton, Reactor Inspecfor (In Charge)
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~L, j. McDermott, Rea tor Inspector b te 72.1 A
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L. B. Higginbotham,
., Radiati6n Specialist Date" Reviewed by) fs I8 D
R. T. Carlson, Senior Reactor Inspector (Dat5
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Proprietary Information:
None SCOPE Type of Facility:
Boiling Water Reactor Power Level:
1600 Mwt Location:
Forked River, New Jersey Type of Inspection:
Routine Accompanying Personnel: Messrs. R. J. McDermott (prepared Sections C in part, F, and N) and L. B. Higginbotham (prepared Sections E, K.4.,
P, Q, and R)
SUMMARY
Safety Items - No safety items were identified during the inspection.
Noncompliance Items - No items of noncompliance were identified during this inspection.
Status of Previously Reported Problems -
1.
The main steam isolation valves were tested and found to be leaking greater than technical specification limits.
It appears that the reactor 1
was operated with this condition.* The valves were subsequently
- C0 Report No. 219/70-2.
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repaired and tested within technical specification requirements.
Reference to Jersey Central Report to the DRL dated March 20, 1970, and paragraph K.3. of this report.
f Other Significant Items -
1.
The last of JC's three GE-trained nuclear engineers originally s
assigned to the OC-1 staff has resigned.
(Section B and Management i
Interview Item 4).
I 2.
GORB representation at PORC meetings is noticeably lacking.
(Management Interview Item 3).
3.
Plant physical security continues to be an item of concern to CO.
(Section C and Management Interview Item 5).
4 Reactor outages since the previous inspection are summarized.
(Section C).
s 5.
The control rod drive system scram accumulator precharge pressure has been reduced per GE recommendations.
(Section F.1.a.)
6.
GE's initial proposal for 0C-1 of safety circuit modifications recently incorporated at Nine Mile Point
- has been rejected by JC.
(Section F.2.
and Management Interview Item 2).
7.
JC-GE have inspected the Permali neutron shielding around the re-circulation system penetrations through the biological shield and found nothing abnormal. All five locations have been instrumented with thermo-couples.
(Section K.2.)
8.
An instance of the failure of one of the emergency diesel generators during routine testing while the reactor was shut down is discussed.
The licensee has discussed this incident in his semi-annual Report No.1.
(Section N).
9.
Sections E, K.4, P, Q and R discuss the results of a review of the health physics program.
Management Interview - The results of the inspection were discussed with Messrs.
McCluskey, Ross and Carroll on March 20, 1970.
1.
In response to inquiries by the inspector during the inspection regarding whether or not aluminum pipe was used for transporting concrete during construction of OC-1, Mr. McCluskey was very responsive in stating i
that an immediate evaluation had already been initiated by Burns and Roe.
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2 Mr. McCluskey stated that revisions to the scram reset circuitry would be done at the first opportunity. He stated that this matter will be given priority attention. He stated that PORC would review the change when GE provided a workable design which to date they have not.
- Scram circuit reset problem.
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3.
The: fact that 33*/. 'of ' the PORC assigned members had failed to attend,
- in person, any of the last 12 PORC meetings,-was discussed. The 1
absent members'are the'GORB representatives to PORC.
Mr. McCluskey 1
4 stated'that the PORC meetings have been held responsively, primarily 1-as problems arose, which was quite frequently, i.e.,
PORC has been problem oriented, therefore did not provide him the benefit of_ time-
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to notify the two GORB members. He stated that planned, scheduled meetings with'an agenda would be-held in the future. He stated that-j it was a matter of fact that both the GORB members that - are on PORC -
l-have spent quite a bit of time at OC-1, however not at times when_
PORC was meeting - He further stated that PORC had held many-informal meetings for which there were no documented minutes.
The content of subject material reviewed by PORC was discussed briefly.
The inspector stated that it was planned to discuss this matter in i
considerable detail during our next visit. An example of the type of
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material in mind for PORC to review and noted by the inspector was j
"Did PORC review the mode of the diesel failure which occurred during the October 1969 control rod drive filter outage." Mr. McCluskey appeared to view this as a contentious subject area, however, one that
.needed an understanding and resolution.
4 The observed staffing omissions in the JC technical organization was discussed. Mr. McCluskey stated that they were trying very hard to get qualified people. He emphasized that he was plugging holes in the JC organization with GE people and would continue to do so until qualified JC' people were in the jobs.
Mr. McCluskey responding to a question regarding when Mr.'Hetrick was j
leaving, stated "two months from March 16, 1970".* The inspector noted this to be approximately the middle of May.
The minutes of the March 17, 1970 GORB meeting, in which McCluskey had stated that a review was conducted of the staffing changes at OC-1, I
were not available for inspection at the time of the visit.
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Subsequent to the visit, on April 8, 1970, the inspector telephoned Mr. McCluskey and stated that Mr. Carroll, who had been designated to become the operations supervisor vice Mr. Retrick, appeared to not meet the technical specification requirement relative to having a senior-l operator's license for two years. The inspector stated that in order to maintain. compliance with the technical specification requirements, it appeared that a request for a waiver from DRL of this technical specifica-tion requirement should be effected by JC.
5.
The inspector pointed out the observed ease of access to vital parts of l
the building through unlocked exterior and interior doors. Locations specifically noted by the inspector were the 4160 switch gear room, the turbine bay, emergency diesels and the control room cable terminal room.
Mr. McCluskey stated that the exterior fence, with the 24-hour manned-vehicular gate, was considered the principal control of ingress and egress j
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- 1nquiry Memorandum No. 219/70-A.
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to the facility. The inspectors pointed out the fact that the fence did not control the influent and effluent canals.
Mr. McCluskey stated that they would review the matter of total plant security.
6.
The fact that the 10-day report to DRL (per paragraph 3C.1 of license DPR-16) was overdue and had still not been submitted, was discussed.
Mr. McCluskey stated he thought that it would be mailed this day.
The fact that the semi-annual report had not been submitted was discussed.
I Mr. McCluskey pleaded "no excuse" and stated it would be mailed in approximately two weeks. He stated that they had assembled the data, it justtooKmanpowertogetitout, that was the problem.
Subsequently on April 8, 1970, the inspector again inquired about the semi-annual report status during a followup telephone call to the site. The report was stated by Mr. McCluskey to still not be out.
The inspector stated that it was the inspector's position that JC was in noncompliance with technical specification requirements because of their failure to issue the report.
Mr. McCluskey stated that he hoped to mail it during the week.
7.
C'oncerning recommendations of GM thru GE for JC to initiate a sampling program of the emergency diesel generators lube oil for water, it had been determined that no samples had been taken.
Mr. McCluskey stated that they were waiting on the GPU nucicar group to evaluate and recommend the sampling frequency.
The fact that the GM diesel control cables appeared to be subject to being covered with water in the cable trench was discussed.
Mr. McCluskey stated that JC would evaluate the reported condition.
8.
The fact that the control rod drive scram time recorder had failed to start on several scrams was discussed.
Mr. McCluskey was asked if they intended to modify the recorder circuitry to provide for periodic testing of the recorder; he stated, they will do.
9.
During the telephone conversation on April 8, 1970, the inspector reviewed with Mr. McCluskey the JC commitment to sample the stack effluent at the 242 foot level and make a correlation with the permanent sample location at the bottom of the stack.
Mr. McCluskey stated chat he was not abreast of the current status, however, he would check. The inspector stated that CO:1 considered this matter to be very much belated at thic time and one in need of prompt resolution.
- 10. The inspector commented on the apparent fire hazard condition in the facility record storage room. Cigarette butts and paper had been observed on the floor during the inspector's tour of the facility.
Mr. McCluskey noted the inspector's comment.
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- 11. The inspector asked Mr. McCluskey if JC intended _ to evaluate piping-stresses, seismic restraints and hangers in light of problems at' the Nine Mile Point reactor.
Mr. McCluskey stated that they had already j
begun such a review. The inspector stated that an evaluation of their current program on drywell leakage surveillance and, monitoring also appeared to be in order.
Mr. McCluskey noted the inspectors comment.
12.
Mr. McCluskey stated that the inspectors had not made one comment in i
the exit interview regarding the improved housekeeping in the building.
l The inspector stated that he had mentioned the subject to Mr. Ross during an earlier contact; Mr. Ross acknowledged this. fact. The inspector stated that it was not this inspector's policy to generally zero in on complimentary items but to primarily focus on compliance
-l items.
Mr. McCluskey further stated that as he had previously stated that when GE was out, there would be a marked improvement in housekeeping.
Mr. McCluskey commented about the frequency of CO inspection visits being made to 00-1.
He stated that he uncerstood that the frequency was quarterly.
The inspector stated that Mr. McCluskey was referring to previously stated frequencies, however these were minimum inspection frequencies. The inspector stated that frequencies were primarily based upon inspection needs, therefore may and often do exceed the minimum i
frequencies.
DETAILS A.
Persons Contacted:
Jersey Central T. McCluskey, Station Superintendent D. Hetrick, Operations Supervisor J. Carroll, Operations Supervisor In Training D. Ross, Technical Supervisor E. Riggle, Maintenance Supervisor N. Cole, Shift Foreman J. Maloney, Shift Foreman D. Kaulback, Radiation Protection Supervisor J. Pelrine, Chemical Supervisor Mr. McKeon, A Operator j
Mr. Young, B Operator General Electric J. Nickols, Shift Foreman i
R. Elems, Shift Foreman B.
Administration and Organization l
Mr. McCluskey reviewed the status of the Oyster Creek organization with the inspector. He provided the inspector with a current marked-up organization chart (see Attachment No. 1).
The weaknesses in the JC organization continued to be
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a 6-shored ~up with qualified GE personnel. Bob Hurd, GE nuclear engineer, serves in a similar capacity to that of the continuing vacant technical engineer position.
There are currently no assistant technical engineers, since the only one remaining of the specified two, Charlie Agan,* resigned on March 20, 1970. Three
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assistant engineers plus two loan engineers from GPU are serving to provide-
-- t technical support.
Mr. Tom Robbins of Pickard and Lowe and Associates, also 1
continues to provide technical support.
One of the two GE shift foremen will be transferred from the site on March 20,
- i 1970. One will remain on shift. Discussions with both GE shift foremen and the JC foremen whom the GE shift foremen are supporting, indicated to the inspector that the release of the one GE foreman appeared appropriate.
Mr. Riggle was determined to be actively functioning as maintenance supervisor.**
His promotion was effective March 8, 1970.
Mr. Finfrock has been officially transferred to the GPU Nuclear Activities Group.
C.
Operations Mr. Carroll conducted inspectors Caphton and McDermott on a tour'of the facility.
The entire building was toured with the exception of the drywell as the reactor was operating.
The inspectors observed exterior and interior doors of the building to be open. One exterior door, adjacent to the cable terminal room, was observed to be unlocked and Mr. Carroll stated that this door was normally kept locked. He had no explanation as to why the door was unlocked. The door to the cable terminal room was also unlocked. Access to the turbine bay, a high radiation area, could also be easily accomplished from unlocked exterior doors. The doors to the emergency diesel generators were also observed to be unlocked. The inspectors observed that the influent and effluent canals have no fences to control access by water. There is a 24-hour guard stationed at the facility exterior gate. The visitor entrance to the building is kept locked and required a telephone contact or key to obtain access at this door.
The following facility shutdowns have occurred since the last inspection on December 5-6, 1969:
Date cause January 3, 1970 (#38)
Automatic scram from a power level of 89%.
Scram was initiated by " low vacuum" in main condenser i
following a load reduction.
Investigation of cause of " low vacuum" was determined to be a leak
'l in turbine to condenser expansion joint allowing air "in-leakage".
- The last of the JC OC-1 personnel trained at the GE fuel management course.
See Inquiry Memorandum No. 219/70-B.
- Inquiry Memorandum No. 219/70-A.
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, Date Cause January 9, 1970 (#39)
Automatic scram from a power level of 987..
Scram was initiated by " Reactor low water level" i
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following a loss of total steam flow input to the three element (steam, feedwater and level)
I fee.dwater controller. Both channels of main steam flow were observed to go to zero just i
prior to the scram.
Investigation disclosed a loose connection on the output of the square root function generator for the steam flow totalizer.
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February 15, 1970 (#40)
Automatic scram from a power level of 857..
Scram-
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initiated by "high flux" following turbine trip.
j Turbine trip was caused by moisture separator I
drain tank high level due to spurious signal during a load reduction. Level sensors for j
drain tank relocated back to the drain tank from
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the moisture separators.
February 18, 1970 (#41)
Automatic scram from u 77. power.
Scram was caused by a high flux in IRM range. Flux increased due to reactor water l' vel change during a e
period that the operator was holding a hot stand-by condition.
Operator did not range the IRM instruments to prevent the scram.
D.
Facility Procedures One instrument calibration check procedure was audited. This procedure was for calibration of the reactor water level instrumentation and was dated March 2, 1970. No discrepancies were noted by the inspector during this audit.
E.
Primary System Examination of records disclosed that samples of reactor coolant were analyzed for conductivity and chloride concentration at least each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as required by Section 4.3.E of the technical specifications. Records for the period December 16, 1969 through March 18, 1970 indicated the conductivity of the coolant ranged to a high of 0.5 micrombos and chloride concentrations were recorded with a high of 50 parts per billion (ppb). The technical specifications for the two parameters are (the most restrictive) 2.0 micromhos and 100 ppb.
F.
Reactivity Control and Core Physics 1.
Control Rods a.
Scrcm Accumulator Precharge Pressure Reduction By letter dated November 10, 1969,* GE recommended to Jersey Central that the accumulator precharge pressures at OC-1 be reduced. The
- Letter from K. W. Hess, GE. Company Site Manager to T. J. McCluskey, Station Superintendent.
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f inspector reviewed the letter and noted that it referred to Tsuruga, Nine Mile Point, Dresden 2 and Dyster Creek.
It had been determined that control rod drive scram insertion velocities were faster than desireable when the reactor was at atmospheric pressure. ' To correct
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this condition, the accumulator precharge pressures were reduced at 1
the other sites listed above and this change was recommended for Oyster Creek as soon as possible. The letter specified the change r'
could be made one drive at a time during plant operation or j
cellectively during the forthcaning outage. The change requirements l
were specified as follows:
(1) Reduce from the present value of accumulator nitrogen pre-charge of 850 to 870 psig down to a value of 565 to 585 psig at 700 F.
(2) The required precharge pressure will vary with the ambient temperature of the accumulator at the time of nitrogen charging.
The nominal precharge pressures of 500 F, 70 F and 90 F are 552, 575, and 598 psig as plotted.
It is important t h a t the pre-charge pressure be checked approximately 30 minutes after initial filling to account for temperature changes in the nitrogen during the charging operation.
(3) Reset pressure switch on the hydraulic control unit from 1350/1320 psig to 970/940 psig on decreasing pressure. Calibration data should be recorded.
(4) During the system operation the water side pressure must not exceed 1510 psig with the recommended operating range between 1400/1450 psig.
This pressure may be properly regulated by throttling the pump discharge valves (V15-7 and V15-10).
Using these valves will hold the correct pressure in all operating modes.
(5) When the pressure change is accomplished, perform the following scram tests:
(a)
Individual scram tests on all 137 control rods at 1000 psig, either in cold hydrotest or with the reactor operating or at zero reactor pressure. Tests at 1000 psig are preferred since this will provide data comparable to that observed during the startup test program. Whichever pressure is selected, all drives should be tested at the same pressure.
Record time from loss of voltage to 10, 50 and 90 percent insertion.
(b) From data in step (a) select the four slowest control rod 4
drives. At some convenient time during a startup or shut-j down, scram test each of the four selected CRD's at reactor
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J pressures of 300 and 800 psi. Record scram time to 10, 5v and 90 percent insertion.
(c) GE requested results of all tests be forwarded to APED t
Design Engineering via Mr. D. Bridenbaugh, APEP, San Jose, I
California. Also requested were the results of.any scram
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times obtained by the Brush recorder during future operations.
The accumulator precharge pressures were reset en masse and the rod
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scram times recorded for 10%, 50% and 90% on February 13, 1970, at i
rated reactor temperature and pressure conditions. The results of this test indicated an average increase in 90% insertion's times of 0.14 seconds from previous results. The data of individual rod scram times are attached as Attachment 3 of this report.
b.
Scram Timing The inspectors reviewed the data from the Brush recorder and no abnormalities in scram times were noted (see Attachment 3).
It was observed, however, that data was not available for scram Nos.'40 and 41 which occurred on February 15, 1970 and February 18, 1970, respectively.
Mr. Ross stated that the recorder had failed to start on these automatic scrams. The inspectors inquired if the recorder was presently functioning properly and Mr. Ross stated that it was and that testing to insure continued operation was planned.
It was made known during the review of this matter that the Brush recorder will not start on manual scrams but only on automatic scrams.
2.
Scram Circuit Reset Problem
- The inspectors inquired into the status at OC-1 of the safety circuit modifications that were recently completed at Nine Mile Point and were informed that the GE design proposal had been reviewed by PORC and returned to GE for further clarification. JC was currently awaiting GE's answer.
Mr. McCluskey stated during the exit interview that this modifica-tion would be made as soon as PORC could review and approve the recommended fix. This area will receive followup inspection.
J.
Electrical Systems The inspector audited cable tray temperatures being recorded per a special prior CO concern ** regarding cable tray lo'ading. All temperatures were recording within a band between 260 C and 280 C.
Mr. Carroll stated that tLe recorder charts were being sent to GE at San Jose for review and analysis.
K.
Containment j
1.
Nitrogen Inerting The nitrogen inerting supply system was inspected. The supply pressure to this system was observed to be regulated at a pressure of 5 psig.
- CO Report No.. 220/69-16, Section F.2.
- C0 Report No. 219/69-6,Section II.A.2.
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Mr. Carroll stated that the two supply control valves (in series) auto-i matica11y isolate (close) if the drywell pressure reaches 2 psig.
The percent of 02 in the drywell was obtained by the inspector from the 02 analyzer and was 4.85%, the torus percent of 02 was 4.1%.*
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The inspector also audited recorder charts taken from the 02 analyzer and
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filed in the record storage room for the period March'8, 1970 through March 15, 1970. The percent of 02 was within the technical specification
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i The N2 makeup to the drywell appeared to average approximately 800 scfd.
2.
Permali Material Because of a reported fire at the Tsuruga reactor which occurred in a wood type material called Permali, an inspection was conducted by GE t
of similar material at DC-1 which surrounds the recirculation piping at the biological shield. The inspector noted a letter to JC from GE dated January 21, 1970 regarding the incident. A preliminary report of the inspection dated February 11, 1970, was also audited.
The results of the GE inspection of the OC-1 Permali material revealed no problems. All five recirculation loops were inspected.
One Permali installation was completely dismantled for the inspection; the others were inspected without dismantling. No discoloration or other visible damage was determined. Thermocouples were installed in all five loops to permit future surveillance. The report stated that APED was c'mpiling i
o a more definitive report.
A 0-500 F recorder was installed at elevation 23.6 feet in the southeast side of the reactor building for the purposes of monitoring the Penmali.
The inspector observed one temperature to be 1340 F and all the others to average approximately 850 F.
The GE letter quoted the continuous operating temperature rating of Permali to be 2200 F.
The letter also stated that Permali discolors at 1650 C.
3.
Main Steam Isolation valves (MSIV)
An " Abnormal Occurrence MSIV Report" dated March 17, 1970 was reviewed.
After repair of three of the MSIV's, the final test results gave the following leak rates for the MSIV's, all of which were within the allowable 11.5 cfh limit.
Valve No.
Final Leak Test Rate (cfh)
NSO 3A 4 0.1 NSO 4A 4 0.3 NSO 3B
<; 0.1 NSO 4B
<; 0.3
- These percentages comply with technical specification requirements in paragraph 3.5A.6 of less than 5.0% 0 -
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JC submitted a report
- to the Commission concerning this matter. A special CO inspection also covered work done on the MSIV's.**
4.
Filter Testing DOP testing of particulate filters in the standby gas treatment system was performed January 19, 1970, with results reported at 99.98% for all i
four filter banks. Freon-112 testing of the charcoal filter banks was performed January 31, 1970 and the lowest results were recorded as 99.79%.
l These results are within the technical specification limits of 99% each.***
N.
Emergency Power The inspectors determined from a log review that one of the two emergency diesel generators (No. 1) had failed during a reactor shutdown period. The generator failed during routine surveillance testing.
The failure was due to a hydraulic lock from water in a cylinder. The water entered the cylinder through a cracked cylinder head.
Examination of the head by General Motors, the supplier, disclosed that casting defects (core sand, iron deposits and casting finish) had partially blocked cooling water flow and had resulted in overheating and cracking of the head. The diesel engine was replaced and the unit tested satisfactorily prior to resumption of operations.
This event was reported in the licensee's semi-annual report.****
The inspectors did not find evidence in the form of PORC meeting minutes that this failure had received the required review.
Mr. McCluskey stated, however, that PORC had reviewed this matter. The subject of quick reporting of failures of this type in accordance with license requirements was also discussed during the exit inter-view.
Mr. McCluskey stated that improvements in this area can be expected but at the time JC did not consider the item reportable as the reactor was shutdown at the time of failure.
P.
Radiation Protection 1.
Surveillance Testing Examination of records disclosed that satisfactory testing and calibration had been conducted for the following listed radiation monitors as required by Section 4.1 and Table 4.1.1 of the technical specifications.
a.
Main steamline monitors b.
Monitors on the reactor building operating floor c.
Monitors on the reactor building ventilation exhaust d.
Air ejector off-gas monitor
- Report to DRL dated March 20, 1970.
- C0 Report No. 219/70-2.
- Technical Specification 4.5.K.1 and L.l.
- Semi-annual report No. 1, May 3, 1969 to December 31, 1969.
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.s Test results included the verification of operation of the standby gas treatment system by monitors of b. and c. on the previous page as required by Table 4.1.2 of the technical specifications.
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2.
Internal Audits Facility records contained the results of an audit of health physics activities by the Nucicar Power Activities Group, General Public Utilities.
The scope of the audit, as outlined in the report, was a review of radio-j logical controls, radiation protection procedures and records, and radio-i active waste disposal records and procedures. The results of this audit are discussed below.
The GPU report indicated that the sealed sources held under AEC materials license 29-12773-1 had not been leak tested at intervals of six months as required by the license. However, Mr. Kaulback furnished the AEC inspector test results which indicated that the sources had been leak tested each six month period.
Mr. Kaulback explained that one set of leak test records could not be found during the GPU audit.
The GPU report indicated that records of calibration of portable radiation monitoring instruments were not being maintained properly. The AEC inspector found that instruments are normally tagged with the date of latest calibration.and the tag is initia11ed by the person performing the work. All instruments noticed by the CO inspector during a tour of the facility were examined and found to be marked indicating a calibration made within the previous three month period.
Mr. Kaulback said that the GPU audit disclosed that technicians were not maintaining the maintenance s
history card files up to date when instruments were calibrated. Further, instruments are calibrated or calibration checked each three months or removed from service until the calibration can be performed, i
The GPU report stated that the stack gas monitor and radwaste liquid effluent monitor had not been calibrated at the frequency required by j
technical specification, i.e.,
each three months. The C0 inspector l
determined from examination of records that both monitors had been calibrated initially (August 1969) with planchet-type sources and since l
that time the calibrations have been made and/or checked with grab samples taken of the effluent streams. The liquid effluent monitors are checked at least weekly in this manner with the results of the (required) grab samples analyzed for each liquid waste release. Grab samples of the off-gas system are taken 'each week and results of the analysis are used to calibrate and/or check the calibration of both the off-gas monitor and the stack gas monitor.
It is the CO inspector's judgement that this complies with the requirements of the technical specifications. This subject is discussed further in Section Q.1. of this report.
The GPU report stated that radiation monitoring systems having no calibration requirements in technical specifications, e.g., monitors 1
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. on the isolation condenser vents, monitors on the containment spray heat exchanger, area radiation monitors, etc., had not been calibrated recantly.
The CO inspector determined that such monitors had been calibrated twice in the latter part of 1969 and that action was taken to include t! <se monitors in a schedule for calibration at six-month intervals. The CO inspector examined the radiation monitoring panel in the control room and noted that all monitoring stations were indicating a response or reading comparable to the results of radiation surveys made by the hecith physics l
group. The alarm points for each monitor were noted to be set at a value l
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thought to be adequate for the area or system monitored.
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Surveys and Monitoring The CO inspector examined records of surveys and radiation work permits l
and found indication that adequate surveys and protective measures were I
required and/or taken for routine and special work throughout the facility.
I There were no indicatione in the records that airborne radioactivity had exceeded applicable MPC levels. Contamination found during routine and special surveys has not been at a level to present a hazard to personnel, and radiation levels through the plant (operating at near full power) are
. essentially the same as determined at 1600 Mwt in the GE start-up test No. 5, Radiation Measurements.
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Q.
Radioactive Waste Systems l
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Gaseous Calibration requirements of Section 4.6.A of the technical specifications for the atack gas monitor have been performed.
The monitor calibration is verified on a weekly basis from results of grab samples of the off-gas system. The alarm for the stack gas monitor is presently set at 1/10 of the continuous annual average release rate for noble and activation gases based on a calibration of 100 uCi/second per count per second (uCi/sec/ cps).
Results of the last three checks (grab samples of off-gas) prior to the inspection showed calculated values of 128 to 148 uCi/sec/ cps.
Mr. Ross said that if the results of the analysis continue to be greater than the 100 value presently used he will change the setpoint of the stack monitor accordingly. The calibration value of 100 was based on the results of 10 separate samples which varied in results of 72 to 101 uCi/sec/ cps.
The alarm point at 1/10 of the allowable release rate is sufficiently conservative to cover a factor of two calibration error. The inspector feels that Mr. Ross makes an adequate and continuing review of this area and will take the necessary corrective action.
The inspector reviewed the analysis and counting technique used for the particulate and halogen filters in the stack monitor and found the procedure and methods to be adequate and conservative.
In discussing this portion of stack monitoring with Messrs. Ross and Kaulback, they stated that the effort to correlate sampling at the top of the stack with the results of sampling at the base of the stack had not been made as yet (C0 Report No. 219/69-2,Section IV.2.)
They said they plan to do this sampling correlation in the near future but gave no definite date or time period.
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Examination of records showed the following releases recorded for January and February 1970.
i January February l
5 Average Average l'
Curies uCi/see Curies uCi/sec j
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Particulate 0.000562 2.1 x 10-4 0.000226 8.2'x 10-5 q.
'I Halogens 0.002543 9.5 x 10-4 0.000472 1.7 x 10-4 j
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Noble and activation 6,023.0 2,250.0' 3,502.0 1.300.0 2.
Liquid Examination of records disclosed that samples are analyzed to comply with
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the requirements of Section 4.6.B, C, and D of the technical specifications, i.e., each batch of liquid waste is sampled and analyzed prior'to release, reactor coolant is analyzed at least each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for total iodine concentration, and the total curie content of' waste ommple tanks, floor.
)
drain sample tanks and the waste surge tank is determined at least each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Examination of records showed the following releases recorded for January and February 1970.
January February i
Mixture, unidentified at MPC l
1 x 10-7 (Curies) 0.282887 0.424458 I
Tritium (Curies) 2.623 1.3796 l
Total Waste Volume Discharged 6
6 (gal) 1.31 x 10 1.03 x.10 9
~
Total Dilution Volume (gal) 6.81 x 10
' 9.67 x 109 l
l v
Average Concentration After Dilution Mixture (uCi/cc) 1.1 x 10-8 1.2 x'10-8 Tritium (uCi/cc) 1 x 10-7
- 3. 8 x 10- 8 j
Percent of Limit Mixture (%)
11 12 q
Tritium (%)
0.003 0.001
)
.1 See Attachment 2 for a summary of waste releases to date.
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Solid l
l Examination of ' records disclosed that one shipment of. solid wast e had' i
been.made. The shipment was made in January,1970 and consisted of 300 -
- )
55-gallon drums containing 0.58 curies. of material.
Mr. Kaulback said
}.
the majority of the radioactivity was.from filter floc.from backwash and i
regeneration of the condensate demineralizers.: Records showed a total of 15R '
j.
55-gallon drums on hand awaiting shipment and Mr. Kaulback informed the.
I j.
inspector that this shipment was due to leave before April 1, 1970.
t-f R.
Environment
.j
(..
I The environmental monitoring program continues to be conducted at'the.same level and scope: as-the preoperational monitoring program. Results'of sampling indicate that there has been no measurable contribution to radioactivity in'the.
environment from operation of-the facility.
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4 Monthly Summaries of Waste Releases for Oyster Creek 1 Liquid Releases August 69 September 69 October 69 November 69 December 69 January 70 February 70 sture, unidentified at MPC 1 x 10-7 (C1) 0.003 0.115 0.07 0.087 0.2 0.283 0.424
-itium (C1) 0.0 0.1 1.2 1.5 2.2 2.6 1.4 5
6 6
6 6
6
- tal waste volume discharged, gal 9.7 x 10 1.8 x 10 1.4 x 10 1.6 x 10 1.3 x 100 1.3 x 10 1.0 x 10 9
total dilution volume, gal 2.8 x 109 9.4 x 10 5.7 x 109 9
7.5 x 10 5.4 x 109 6.8 x 10 9.7 x 10' 9
?"erage concentration after dilution Mixture (uci/cc) 2.8 x 10-10 3.2 x 10-9 3.2 x 10-9 3.1 x 10-9 9.8 x 10'9 1.1 x 10-8 1.2 x 10-8 Tritium (uct/cc) 2.8 a 10-9 5.6 x 10-8 5.3 x 10-8 1,1 x 10-7 1.0 x 10-7 3.8 x 10-8 Percent of limit Mixture (%)
0.28 3.2 3.2 3.1 9.8 11 12 Tritium (%)
0.0001 0.0019 0.0018 0.0037 0.0033 0.0012 Caseous Releases Noble and activation gases (C1) 28.8 57.6 939.7 6000 6023 3502 Time average release rate (uci/sec) 11 22 360 2240 2250 1450 5
Percent of limit. 3 x 10 uC1/sec (%)
0.0037 0.0074 0.12 0.75 0.75 0.48 Halogens and particulates (C1) 0.000218 0.000019 0.000461 0.00228 0.003105 0.000698 Time average release rate (uci/sec) 8.4 x 10-5 y,1 x 10-6 1.8 x 10
8.5 x 10-5 1,1 x 10-3 2.9 x 10-4 Percent of ILeit, licensed 4 uC1/sec (%)
0.002 0.0002 0.0045 0.002 0.03 0.007 JERSEY CENTRAL POWER & LIGHT CO.
CO Report No. 219/70-3
JERSEY CE:. /,L POWZ!, UC CD Report No. 319/70-3 SCRAM INSERTION TIMES (SECONDS) fl
/
/.
Date of Scram Jantaary 9.1970 Reason for Scram Low Reactor Level Rod Secuence A
Rod Coordinate 1 Inserted 1 Inserted 1 inserted 10 50 90 10 50 90 10 50 90 06-43
.38 1.42 2.56 06-27
.37 1.45 2.64 06-19 a
.37 1.37 2.40 10-39 a
.38 1.45 2.55 14-19
.39 1.51 2.72 14-27 40 1.52 2.72 14-35
.38 1.43 2.51 14-43 a
.39 1.51 2.70 18-07 a
.38 1.46 2.57
/
18-23*a
.37 1.38 2.39 14 47 a 40 1.52 2.72 22-27
.40 1.46 2.47 22-35
.41 1.50 2.63 22-43
.39 1.43 2.50 30-19
.39 1.46 2.54 30-27 40 1.45 2.51 30-35
.41 1,44 2.48 30-43
.39 1.38 2.39 30-51_a
.38 1.49 2.78 38-11
.38 1.48 2.63 38-27
.39 1.55 2.75 l-38-35
.39 1.46 2.61 38-43
.37 1.40 2.51 46-11
.37 1.38 2.48 46-19
.38 1.46 2.66 46-35
.38 1.42 2.47 Average
.386 1.453 2.57 a = drives with filters
- = erratic scram times
y 00 Re;, art Nr. el'J/7D c Ittachmtnt 3 SCRAM INSERTION TIMES (SECONDS) 1 i
Date of Scram February 13, 1970 Testing following accumu Reason for Scram 1 star nrechare, creanne, radu c H.-n Rod Sequence N.A.
Rod Coordinate
% Inserted
% Inserted
% inserted 10 50 90 10 50 90 10 50 90 02 19
.43 1.41 2.31 02-23
.40 1.39 2.40 02-27 44 1.55 2.61 02-31
.42 1.56 2.62 02-35
.43 1.56 2.65 06-11 43 1.54 2.64 06-15 48 1.63 2.68 06-19
.45 1.55 2.60 06-23
.47 1.66 2.77 06-27
.43 1.57 2.71 06-31
.44 1.56 2.64 06-35
.43 1.47 2.52 06-39
.46 1.65 2.73 06-43 46 1.56 2.63 10-07 4o 1.74 2.93 10-11 43 1.51 2.54 10-15
.41 1.46 2.47 10-19
.43 1.59 2.67 10-23 42 1.50 2.56 10-27
.46 1.62 2.76 10-31 *
.47 1.63 2.74 i
10-35
.43 1.55 2.64 I
10-39 44 1.56 2.63 10-43
.47 1.61 2.72 i
10-47
.43 1.53 2.61 i
1 14-07 46 1.60 2.69 I
1 (continued) i* erratic scram times
V,
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- f Attcchment 3 c h Scram dated February 13, 1970 (Continued) 10 50 90 10 50 90
'10 50 90 y
14 11
.41 1.47 2.49 14-15
.48 1.58 2.62 14-19
.46 1.66 2.80 14 23
.46 1,67 2.79
.14 27
.44 1,56 2.65 14 31 45 1.65 2.73 14-35
.43 1.52 2.55 14 39
.43 1.58 2.66 14-43
.44 1.68 2.85 i
14 47
.45 1.56 2.60 18-03
.45 1.59 2.70 18-07
.43 1,47 2.50 18-11
.45-1.57 2.66 18-15
.43 1.52 2.62 18-19
.46 1.65 2.78 18 23 43 1.47 2.52 18-27
.45 1.60 2.68 18-31
.46 1.63 2.80 1
18-35
.49 1.74 2.85 18-39
.46 1,62 2.71 I
18-43 44 1.55 2.58 18-47
.43 1,59 2.69 18-51 42 1.44 2.43 22-03
.45 1.60 2.72 1
22-07
.46 1,58 2.66 22-11
.47 1.70 2,84 22-15
.46 1.59 2.68 22 19
.48 1.68 2.82 22-23
.44 1.54 2.63 22-27 46 1,54 2.57 22-31 49 1.66 2.78 (continued) e
P~
5
)
- Alt tchment 3 3
a l
y Scram dated February 13, 1970 (Continued) 10 50 90 10 50 90 10
. 50 90 22-35 47 1,62 2.79 22 39 46 1.65 2.81 22-43 45 1.56 2.63 22-47
.49 1.62 2.81 I
22-31
.41 1.51 2.61 26-03 *
.50 1,55 2.56 26-07
.47 1.59 2.67 i
26-11
.51 1.81 3.60 26-15
.48 1,69 2.86
)
26-19
.48 1.62 2.67 26-23 46 1.70 2.85 26-27
.51 1.74 2.93 26-31
.47 1.67 2.85 26-35
.55 1.97 3.39 26-39
.44 1.53 2.58 26-43
.46 1.69 2.84 26-47
.48 1.72 3.00 26-51
.50 1.69 2.82 30 03 valved out 30-07
.51 1.74 2.86 30-11
.51 1.74 2.95 30-15
.46 1.56 2.59 30-19
.47 1.60 2.72 30-23
.48 1.64 2.74 30-27
.47 1.60 2.69 30-31
.48 1.62 2.70 30-35
.47 1.54 2.61 30 39
.47 1.60 2.71 30-43
.46 1.49 2.52 30-47
.49 1.65 2.84 30-51 43 1.65 2.91 1
(Continued)
- erratic scram times
_m f.
a
[
j;
. d choent 3 i !
Scran dated February 13,1970 (Continued) 10 50 90 10 50 90 10' 50 90 l
i 34-03 46 1,55 2.62-34-07 48 1.61 2.68
~ {-
34-11 46 1.50' 2.49 i
\\
34-15
.50 1.72 2.87 34-19
.46 1.53 2.58 34-23
.48 1,67 2.80 34-27
.52 1.80 3,08 34-31
.50 1.69 2.94 34-35
'. 50 1.69 2.86 i
34-39 44 1.51 2.53 C
34-43
.48
'1.60 2.74-34-47
.52 1.88 3.22 34-51
.52 1.66 2.74
)
f 38 07
.48 1.64 2.79 38-11
.47 1.62 2.74 I
38-15
.52 1.72 2.80 i
38-19
.49 1,64 2.84 30 23
.49 1.65 2.75 38-27
.48 1,71 2.87 38-31
.48 1,57 2.64 l
38-35
.47 1.62 2.80
'i 38-39
.46 1.54 2.61 38-43
.47 1,61 2.75 38-47
.49 1.73 2.96
.f 42-07
.48 1,64 2.79 i
42-11
.50 1.65 2.75 42-15
.52 1.66 2.76 4
42-19
.50 1.65 2.74 42-23
.51 1.76 2.94 42-27
.52 1,69 2.83 42-31 47 1,62 2.73 (Continued)
. ')
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e ctaument 3 Scram dated February 13, 1970 (Continued) 10 50 90 10 50 90 10 50 90 42-35
.48 1.63 2.74 42-39
.54 1.95 3.32 42-43
.43 1.45 2.47 42-47
.43 1.48 2.54 46-11
.45 1.56 2.65 46-15
.49 1.56 2.62 46-19
.47 1.65 2.83 46 23
.49 1.66 2.75 46-27
.48 1.67 2.82 46-31
.47 1.63 2.76 46-35
.46 1.56 2.63 46-39
.46 1.52 2.57 46-43
.46 1,57 2.66 i
50-19
.46 1.53 2.59 i
50 23
.47 1.56 2.65 e
50-27
.51 1.67 2.79 50 31
.42 1.51 2.57 50-35
.46 1.54 2.57 i
Average 2.71 b
p,
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e July 2, 1970 J. P.-O'Reilly, Chief, Reactor Inspectior & Enforcement Br nch b
Division of Compliance, Headquarters
. INQUIRY MEtipRANDUM JERSEY CEtrfRAL POWER & LIGHT COMPANY, OYSTER CREEK 1, 219/70aH o
L ISOLATION CONDENSER INSTRLHENTATION SETTING - NONCOMPLIANCE Mr. T. McCluskey, Station 'uperintendent, was contacted by telephone on the morning of July 1,1970, and he informed the inspector of the following:
1.
The present trip settings of the steam line and condensate line instrumentation (used.to isolate the condensers in the event of a leak) are 20 peig d P'and 59" d P H O respectively. The condensate 2
line trip setting was changed from 27" to 59".1 P H O as per GE 2
instruction in the December, 1969 letter as a result of an inadvertent isolation of the isolation condenser when the isolation condenser was
' tested for the first time under load conditions.- The stated reason for the inadvertent isolation was that-the condensate flow instrument saw a false transient spike. A five-second time delay was addod to both the steam and condensate line instrumentation to preclude additional spurious closures.
i 2.
These settings were reHorted to represent 300% (steam line setting) and
.l 280% (condensate line setting) of rated flow which was assumed to be 5 lb/ hour. These numbers were provided to JC by GE representatives 3.2 x 10
.j in San-Jose on Joly 1,'1970.
-i 3.
The current technical specifications for Oyster Creek specify the settings (Technical Specification 3.1, page 3.1-10 H.) as < 20 psig A P (steam) cud $27" 4 P H 0.
2 4.
GE provided JC with calibration data points for the condensate instrumenta-tion as follows:
68" = 300% of the rated flow of 3.2 x 105 Lh/hr.i.e.,
(9.6 x 105 lb/hr); 59" - 280% of rated flow; and 27" = 190% of rated flow.
' Mr. McCluskey. was informed that the current trip setting of the condensate line instrumentation was in violation of the technical specifications for.a. limiting condition for operation and that the plant was operating in noncompliance with the Technical Specifications. Mr. McCluskey was subsequently contacted at.
2:30 p.m. on July 1, 1970 and requested to immediately contact Mr. R. Schessi of DRL to discuss their situation. Mr. McCluskey informed the inspector at 6:30 p.m. that discussions were held at approximately 5:00 p.m. on July 1, 1970 between Mr. I. Pinfrock, Manager, Nuclear Generating Station, JC and DRL.
S3050&t)05f
- as
, COMPLIANCE a ffIh
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2-At present we consider that the licensee is operating in noncompliance with Technical Specifications in that the trip setting of the condensate line instrumentation is in excess of the permissible setting.
I I
R. T. Carlson I
Senior Reactor Inspector cc E. G. Case, DRS (3)
P. A. Morris, DRL R. S. Boyd, DRL (2)
R. C. DeYoung, DRL (2)
D. J. Skovholt, DRL (3)
P. W. Howe, DRL L. Kornblith, Jr., CO Regional Directors, CO REG File i
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