ML20107C122
See also: IR 05000219/1970007
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U. S. AIOMIC ENERGY COMMISSION
REGION I
DIVISION OF COMPLIANCE
Report of Inspection
CO Report No. 219/70-7
Licensee:
JERSEY CENTRAL POWER AND LIGHT COMPANY
Oyster Creek 1
License No. DPR-16
Category C
Dates of Inspection:
October 13-16, 1970
Dates of Previous Inspec on:
September 24-25, 1970
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k f0(V M At h
/2,///70
Inspected by:
R.
. McDe
t, Reactor Inspector
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Reviewed by :
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R. T. Carlson, Senior Reactor Inspector
IDate
Proprietary Information:
Nor.e
ECOPE
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Type of Facility:
Boiling Water Reactor
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Power Level:
1600.MWt
Location:
Forked River, New Jersey
Accompanying Personnel:
Mr. R. T. Carlson, Senior Reactor Inspector, CO:I on
October 13-16, 1970
Mr. F. J. Nolan, Senior Reactor Inspection Specialist,
C0:HQ on October 13-15, 1970
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Mr. J. G. Keppler, Senior Reactor Inspection Specialist,
CO:HQ on October 13-16, 1970
Scope of Inspection:
A routina announced inspection was made to review;
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(1) the status of outstanding items identified in
previous reports, (2) operations for the inspection
period and (3) complete the requirements of PI 3000/1.
Messrs. Carlson and Keppler reviewed and tested the manage-
ment systems and controls that are in effect at the site.
Mr. Nolan reviewed and tested the administrative system
for insuring the adequacy of operating, maintenance and
emergency procedures and the adequacy of the surveillance
testing program. The findings of Messrs. Carlson,
Keppler, and Nolan will be documented in CO Report No.
i
219/70-8.
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SUMMARY
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Safety Items a None
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Noncompliance Items -
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'1.. Contrary to the requirements of Technical Specification 4.7.A.3,
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- each diesel generator has not been given a thorough annual inspection.
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No inspection has been' performed since the issuance of the provisional
operating license on April 9, 1969. -(Section N.6.)
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2 '. Contrary to the requirements of Technical Specification 4~.7.A.5, the
diesel generator starting battery surveillance checks are not being
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performed in entirety.
Specific required checks that are not being
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performed include the quarterly temperature'and electrolite measure-
ments.
(Section N.3.)
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3.
Contrary to the requirements of Technical Specification 4.7.B.3, the
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125 volt station battery surveillance checks were not being performed
in entirety.
Specific required checks that.are not being performed
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include the quarterly check on electrolite level and temperature readi,ng
of every fifth cell.
(Sewtion N.5.)
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Contrary to the requirements of Technical Specification 3.1, Table
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3.1.1,. item H.2, the trip setting for the "high flow" instrument in
the condensate'line.(input into the isolation condenser isolation
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circuitry) was set in excess of the6 27 inches AP H O required.
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(Section F.3.)
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5.
Contrary to the requirements of Technical Specification 3.5.A.6, the
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02 level within containment exceeded the specified 57. limit during
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operation on June 4,1970, for a period of at lea'st 20. hours.
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.(Section K.1.)
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Contrary to the requirements of Technical Specifications 4.5.K and 4.5.L,
the standby gas treatment charcoal and particulate filters were not
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tested for removal efficiency within the required six-month interval.
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(Section K.6.)
Unusual occurrences -
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Unidentified Leakage Into Containment - The measured unidentified
leakage rate into containment increased from 1 to 4.5 spm over a
,
period of approximately two weeks. The reactor was shutdown on
September 16, 1970,.to investigate the source. A recirculation pump.
discharge valve packing-was found to be leaking and was repaired.
Measured unidentified leakage returned to approximately 1 gym following
the repairs.
(Section K.2.)
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2.
Diesel Generator Failure to Start
Surveillance testing during the
period of February through October, 1970, disclosed that the No. 1
diesel generator failed to start automatically on the first attempt
on fear occasions. The problem was reported by the licansee to be
related to poor alignment between the diesel cranking motor pinion
gear and its associated ring gear. The mounting brackets for the two
starting motors (that act together to crank the diesel) have been re-
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located on both the No.1 and the No. 2 diesel generators. Electrical
problems have been experienced which have prevented proper operation
on three occasions during this period.
(Section N.2.)
3.
Loss of Main Circulating Water - On July 11, 1970, and August 2, 1970,
sea grass plugged the main circulating water intake screens to the
'
extent that the main circulating water pumps had to be shut down. On
both occasions the reactor was operating and following the first
occurrence, the' reactor scrammed for high reactor pressure. Following
the second occurrence, the reactor was manually scrauned.
(Sections C
and H.)
Statue of Previously Reported Problems - A formal enforcement letter * was sent to
Mr. R. F. Bovier, President, Jersey Central Power & Light Company listing seven
items of noncompliance identified in previous CO reports ** and other concerns
ragarding administrative systems and staffing for the operation of the 0C-1
fa:ility. A reply to the letter was received September 29, 1970.***
Other Significant Items -
1.
Control Rod Performance - Following the April - May, 1970 rod work outage,
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all control rod drives were scrammed in the hot pressurized condition.
The maximum time observed for any drive for 90% insertion was 3.4 seconds.
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During the inspection period there have been three scrams for which the
scram times for 26 monitor control rods had been recorded and the maximum
time observed for any drive for 90% insertion was 3.1 seconds. Totalized
stall flows have been taken monthly and have increased from 167 gallons
No
per minute to 218 gallons per minute during the inspection period.
operating difficulties had been experienced with the control rod drives.
Followup inspection items previously identified by Mr. D. Pomeroy, TSB,
,
CO:HQ, during Ms May,1970 assist inspection of the control rod drives
were reviewed.
(Section F.1.)
2.
Control Rod Inadvertent Drop - During operation on September 26, 1970,
a control rod inadvertently dropped into the reactor from position 32.
The cause for the inadvertent drop was determined to be an improper
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valving ~ arrangement which occurred during mainten'ance on the control
rod drive on the previous day. The improper valving resulted in
removing the air supply to the scram valves which then allowed the
air to slowly bleed from the scram valve diaphragms and to eventually
open the scram valves. Rod recovery was made within approximately two
hours.
(Section F.2.)
- 1-tter to Mr. R. F. Bovier, President, JC from Mr..L.D. Low, Director, CO:HQ,
,
dated September 9, 1970.
- C0 Repert Nos. 219/70-1, 219/70-2 and 219/70-5 Noncompliance Items.
- Letter to Mr. L. D. Low, Director,'CO:HQ from Mr. R. F. Bovier, President, JC,
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dated September 24, 1970.
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.I'salation Condenser Initiatina Lonic Circuitry - Circuitry' changes
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, vare made during August, 1970 to prevent the closure of a' single excess
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flow check valve from removing the automatic actuationLof the isolation
condsnsers on a high reactor pressure. As previously reported in CO
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' Report lNo. 219/70-5, the closure of an excess flow check valve resulted'
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in.a circuit review by JC and the review' disclosed the isolation
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condenser automatic initiating logic-circuitry had been defeated. The
cause for the-loss of function was stated by the licensee to' be. due to
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a design' error in the initiating logic circuitry. ~(Section E.4.)
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4.
Temporary Strainers in' Condensate and Feedwater Systems -:JC has
' stated that all temporary . strainers have been removed from all of the
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nuclear systems at the OC-l' facility. The inspector's inquiry was
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prompted by a recent-failure of a temporary strainer in the feedwater
! $ystem at.the Nine Mile Point reactor.
(Section E.5.)
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.S . Turbine Initial Preseure Regulator - As discussed in C0 Report No,
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219/70-6, during the period of September 17-28, 1970, five steam
pressure (flow)disturtaccas resulted from malfunctions with the
initial pressure regulator.
Performance of the system since September 28,
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1970, has been satiefactory and no additional disturbances from this
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source have resulted to date.
(Section H.2.).
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6.
Secondary Containmer Testing
A review was made of the method of
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testing the secondary containment and the records of the results.
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It was disclosed that tha test is conducted with both airlock doors
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at any one. penetration closed. The results of the tests indicate the
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spa f fied leak tightness for secondary containment. is within the
retaire.ments of the Technical Specifications.
(Section K.5.)
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7.
Facf_lity Staffing - There has been a noticeable improvement since the
. May, 1970,. inspection in the numbers of people at the site who are
currently preparing for either the senior reactor operator license or
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a reactor operator license examination.
(Section B.2.)
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8.
PI 3000/I " Survey of Security Measures for Emergency Power Systems" -
The requirements of PI 3000/1 were. completed.
Station security aspects
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previde. reasonable assurance that an. authorized person cannot gain
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a'icese to the emergency power controls without detection. The review
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did disclose that some key components within the DC emergency power
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system could be defeated without control room annunciation.
(Section N.I.)
9. ,Gasecus Release Rate - Tne current gaseous release rate from the
facility was reported to be approximately 7000 uCi/second.
(Section Q.3.)
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10.
Carbon-14' - Per the. memo from CO:HQ on this subject,* the substance of
the Public, Health Service's findings at the Yankee reactor were
discussed with the licensee. JC has stated that' samples will be obtained
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to assess the magnitude of Carbon-14 in their effluents.
(Section Q.1.)
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Eiencrandam, O'R6111y to Senior Reactor Inspectors, dated October 17, 1970.
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11. Chanistry - A review of the chemistry records for the period from May
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throudh September, 1970, indicated that no Technical Specification limits
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had been exceeded. -Typical values and the ranges of the values for
measured variables were obtained.
(Section'E.2.)
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12. Facility Plans - An outage is planned for the week of October 18-25,
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- 1970. Major maintenance items for the, outage include:-
a. .Containsent integrated leak rat _e testing.
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b. . Main steam isolation valve testing.
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c.
Turbina initial pressure regulator control system modifications
including the replacement of control valve cams, and the replacement
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of control linkages.
Exit Interview - Messrs. Carlson, Keppler and McDermott conducted the exit
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interview with Messrs. McCluskey, Ross, Carroll, and Riggle on October 16,
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1970. Messr.s. Carlson and Keppler discussed their inspection findings and
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their discussions will be included in a separate report.
Mr. McDermott discussed
items of apparant noncompliance and other areas of concern as follows:
1.
1.
Diesel Generator Annual Inspection (Item of Noncompliance)
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The inspector statad that the annus1 inspection test of the diesel
gen-trators had not been performed since the issuance of the provisional
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operating license on April 9, 1969.
Mr. Ross stated that it was JC's
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intantion to perform this test and that Mr. Riggle had prepared a
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maintenance procedure that was awaiting the review and approval of PORC.
He further stated that the inspections required by the Technical Specifica-
t. ions would be completed in November, 1970. This was identified as
an itam of noncompliance.
2.
Diesel Generator Starting Battery Surveillance Checks (Item of Noncompliance)
.
The inspa: tor stated that the surveillance checks required by Technical
Spe:f fication 4.7. A.5 were not being done in entirety.
Specific
rapirad checks that ware not being performed were stated to include
the quarterly temperature measurements and the electrolyte measurements.
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Mr. Ross stated that it was JC's intention to perform the temperature
and electrolyte measurement tests.
3.
Statin Battery Surveillance Checks (Item of Noncompliance)
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The inspa:ter stated that the surveillance te-sts required by
Technical Specification 4.7.B.3. on the 125 volt station batteries
wara n0t being parformed in accordance with the Technical Specific'ations.
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Spaific re gired chacks that were not being performed were' identified to
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in:Lude ':tca parterly check on electrolyte level and the temperature
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readi_g of tha fifth cell., Mr. Ross stated that it was JC's. intention
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to parform these tests. This was identified as an item of noncompliance. 2
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4.
Instrumer.t Trip setting Isolation Condenser Isolation circuitry.
(Item of Noncompliance)
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The inspector stated that the trip setting' for the high flow instrument
-in the condensate line had exceeded the Technical Specification limit
between the~ period of December 9,1969 and July 1,197.0.
This was
identified as an item of noncompliance.
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5. -Containment Inerting (Item of Noncompliance)
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The inspector stated that a review of the records had disclosed that
the 02 content within containment had exceeded the Techniczt Specification
(3.5.A.6.) limit of 5% during operation on June 4,1970. The inspector
further stated that there was a strong indication that the 5% limit may
have been exceeded for the six previous days as proper operation of the
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0 8ampling instrument was suspect during this period.
Mr. Ross was
2requested to provide the basis for the PORC committee's review of this
occurrence and their findings that this event did not violate the
Technical Specification limit.
Mr. Ross' justification for this finding
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was that the Technical Specifications allow a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
following a startup and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before scheduled shutdowns during which
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is invalidated and that based on
periods the specified 5% limit for 02
this, it was JC's understanding or interpretation of the Technical
Specifications that the 02 limit could exceed the specified 5% for up
to a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during operation. The inspector stated that he
did not concur with this interpretation and identified this as an item
of noncompliance.
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6.
Diesel Generator's Performance
The inspector stated that his review of the records had disclosed that
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on five occasions during the period of February through October 1970,
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the No. 1 diesel generator had either failed to start on the first
attempt or had tripped out from an electrical fault. The inspector
also stated that his review of the work requests had disclosed that
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on only two of the five occasions had any followup action been
initiated. The inspector stated that it was apparent to him that
an adequate review was not being made of surveillance testing records
as all of these faults had been indicated in the surveillance test
-records. Mr. Ross informed the inspector that following Mr. Don
Reeves' successful completion of the October,1970 senior operating
licensing examination, he will be assigned overall responsibility for
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the surveillance testing progran,and that this should result in an
improvement in this area.
7.
123 Volt Station Battery Test
The.inspertor stated that the review of the semi-annual discharge
testing of the A and B batteries had indicated that on five occasions,
one or tha other of the battery banks had failed to' meet the minimum
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acceptance criteria for the test.' Mr. Ross was also informed that
the maintenance supervisor was only aware of the most recent failure
which occurred on October 3, 1970
These test results were stated to
indicate that an inadequate' review of the surveillance testing records
was being made. Mr. Ross again stated that Mr. Don Reeves will be
assigned overall responsibility for the surveillance testing program
and it is expected that improvement in the coordination and review of
tests will result.
DETAILS
A.
Persons Contacted:
Mr. T. M:Cluskey, Station Superintendent, OC-1
Mr. D. Ross, Technical Supervisor, 0C-1
Mr. J. Carroll, Operations Supervisor, OC-1
Mr. W. Riggle, Maintenance Supervisor,- 00-1
Mr. J. ' Sullivan, Technical Engineer, OC-1
Mr. R. Toole, Technical Engineer, 00-1
Mr. T. Johnson, Electrical Foreman, 0C-1
Mr. F. Kossatz, Mechanical Foreman, OC-1
Mr. D. Kaalback, Radiation Protection Supervisor, 0C-1
Mr. N. Goodenough, QA Engineer (Radiography), GPU
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B.
Administration aod Organization
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1.
Managa. ment - Administrative Controls
Mr. R. T. Carlson, Senior Reactor Inspector, 00:1 and Mr. J. Keppler,
Senior Reactor Inspection Specialist, CO:HQ were at the site during the
period of October 13-16, 1970, to review and test the management systems
and controls at the site. Mr. F. Nolan, Senior Reactor Inspection
Spe:ialist, CO:HQ was at the site during the period of October 13-15,
'1970, to review and test the system for insuring the adequacy of opera-
ting, maintenance, and emergency procedures and to review the controls
used in implementing and reviewing the results of surveillance testing.
Their inspection findings will be discussed in a separate report.
2.
Operations Organization
Mr. J. Carroll informed the inspector that the present operating
organization consists of the following:
a.
Shift Foremen - These positions are cur'rently being filled with
four senior licensed operators.
In addition, there are two un-
licensed operating foremen, one of which is scheduled for an
October, 1970, senior operating license exam.
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b.
Shift Operators - These positions are currently being filled by
three licensed operators and~one senior licensed operator.. Of
the four control room "B" operators, three have operating licenses.
Five. individuals are in training for operating licenses and
scheduled to take the exams in October of 1970.
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Mr. Don Reeves is also in training for the senior operating license
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exams scheduled for October, 1970. It was reported that following
successful completion of the exams, Mr. Reeves will assume the over-
all responsibility for the surveillance testing program.
3.
Site Technical Support Staffina
Mr. D. Ross informed the inspector that the present staffing in the
Radiation Protection Group and the Chemical Group is as shown in Figure 1
attached.
Mr. Ross reported that Mr. Peirein, Chemistry Supervisor
has had 10 years of radiochemistry experience at KAPL and 8-10 years
experience in radiochemistry and chemistry at Industrial Research
Laboratories reactor.
Mr. Kaulback, Radiation Protection Supervisor, has
had prior health physics experience at the Saxton reactor and was present
at the OC-1 facility prior to reactor startup.
Mr. Ross was asked to provide the inspector with the summary sheets or
other reports that he requires from the Radiation Protection and Chemistry
Groups. Mr. Ross provided the inspector with copies of summary sheets
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that ha maintains in his office that are originated within the health
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physics and chemistry groups. .The records consist of the following:
a.
Monthly reports which include a summary of the solid wastes dis-
charged from the facility, personnel exposures, building surveys,
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airborne radioactivity surveys, batch radioactive liquid dumps,
identification and amount of activity of liquid waste discharged,
total volume of liquid waste discharged.
b.
Weekly summary sheets of gaseous releases from the facility including
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an isotopic breakdown of the gas.
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c.
Weekly summaries of primary system chemistry.
Records appeared to be current and in sufficient detail to permit
auditing of the health physics and chemistry groups' activities.
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C.
Operations
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Following tha April - May,1970 rod work outage, reactor operation was resumed
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on May 22, 1970, and continued until June 17, 1970, when a scram was initiated
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from 100% power from closure of the main steam isolation valves (MSIV's). The
- valves closed on.a signal from two faulty bi-metal switches that are located
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. above the turbine control valves. These switches are used to monitor for a
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- limited steam break. All post-scram functions following this scram were
- reported by the station superintendent to have functioned normally. Reactor
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operations resumed on June 17, 1970, and continued until. July 11, 1970, when a
scram was. initiated by a partial loss of main circulating water for the condensers
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which resulted from sea grass accumulation on the intake screens.. The reactor
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was restarted on July 12, 1970, and continued until August 1, 1970, when sea grass.
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accumulation on the. intake screens again resulted in a reactor scram (manual).
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Operations resumed on August 2, 1970 and continued until September 16, 1970, when
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a manual shutdown was initiated as required by Technical Specifications for a
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high identified leakage rate in the containment which was caused by a leaking
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valve packing. Operations resumed on September 17, 1970. During the period of
September 17 through September 28, 1970, three additional scrans resulted from
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malfunctions of the turbine initial steen pressure regulator (IPR). On
September 28, 1970, an additional plant disturbance was experienced as a result
of a malfunction of the feedwater controller system but the plant did not scram.
The plant was shutdown on October 17, 1970, for a planned eight-day outage. Majoz
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activities that were scheduled for the outage included:
(1) a containment integrated
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leakage test; (2) a main steam isolation valve leakage test; (3) modifications to
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the turbine control system including the replacement of linkages and the control
valve cams. Listed below is a description of the unscheduled shutdowns during
the inspection period:
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Date
Cause
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June 17, 1970
Automatic scram from 100% power resulting from the
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(Scram No. 45)
closure of the main steam isolation valves. Just
prior to the scram there had been two instances of
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spurious half-trips on the main steam line break
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circuit, which appeared to be caused by high readings
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from the temperature sensors. Checking of the RTD's
located in the area above the turbine control valves
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indicated an ambient temperature of approximately 1300
F which did not indicate that there should have been
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any spurious trips. Before all circuits could be
checked out, two temperature detectors, one in each
channel, picked up and caused the main steam isola-
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tion valves to close which resulted in a reactor scram.
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Reactor pressure reached 1040 psig during the ensuing
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transient and the isolation condensers were manually
initiated to cool'down and depressurize the reactor.
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All monitored control rods reached 90% insertion
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within 3.11 seconds. Following this scram and
turbine trip, the 4160 volt power supply to MCC 1-A
was automatically switched to' the startup trans-
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former.
However, MCC 1-B did not transfer auto-
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matica11y and resulted in the starting of diesel
generator No. 2.
Before the diesel generator picked
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up the loads on MCC 1-D, the operator had synchronized
and manually closed breaker SID restoring power to
MCC 1-B, (which feeds HCC 1-D) and the diesel
generator was shut down. Thus, all safety equipment,. _
which might have been required, would have operated
if needed. A check of breaker SID after the scram
showed it.to be operating properly and all interlocks _ -
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' functioning normal. Further investigation disclosed
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Cause
'
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that the malfunctioning may have beaa attributed to
dirty contacts in the breaker.
.
The check of the bi-metal temperature gaitches dis-
.
closed that two of the detectors had set points of
.
'
sv 500 F less than 1800 which would be the allowable
set point for the bi-metal switches in the vicinity
of the turbine control valves (Technical Specifications
allows a 500 span for the trip point above the ambient
i
measured temperature). The two temperature detectors
were' reset to 1800 F. .
July 11, 1970
Automatic scram'from approxdmately 43% power which
.
(Scram No. 46)
was initiated by high reactor pr : cure. Just prior
to the scram the load was being reduced due to a
buildup of grass' at the intake structure which caused
the shearing of pins in the travelling screens and
compounded the situation. Two (of four) circulating
water pumps were removed from service due to the
cavitation which was caused by loss of suction
pressure. The water level had decreased at the intake
structures and started effecting the service cooling
4
water systems and when the reactor recirculat(ng pump
i
and turbine oil temperatures started to incresse, the
load was dropped to approximately 200 MWe with re-
J
'
circulation flow and the turbine generator was tripped
with the emergency trip buttons. About one minute
i
after the turbine tripped, with nine bypass valves
opened, the pressure increased and the reactor scrammed
on high pressure which was apparently caused by
insufficient condensing ability of the main condenser.
The vessel water level dropped to approximately 9 feet-
l
4 inches above the active fuel on this transient. No
i
control rod scram times were obtained on this scram
as the recorder apparently stuck. The rod buffer
times were excmined and appeared normal. Conditions
were returned to normal at the intake structure by
reducing the flow, installing new sheer pins in the
travelling screens, and running the screens continuously
until the grass conditions tapered off.
August 1, 1970
During operation at 100% power, sea grass again began
(manual shutdown)
to plug the intake structure travelling screens.
The reactor was manually shut down and the main
circulating water pumps turned off. The reactor was
returned to service August 2, 1970.
,
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Date
Cause
September 17, 1970
Automatic scram from 100% power which was initiated
(Scram No. 47)
by a high flux signal.
Immediately pre:eeding ,this
scram, the reactor was operating at 1600 MWt, when
a power oscillation began. The operator dropped load
with recirculation flow from 530 MWe to 500 MWe where
a power oscillation again occurred. The load was
dropped further to 400 MWe at which time-everything
'
leveled out. The turbine then tripped from a high
level in the moisture separator drain tank. The
turbine trip caused the pressure to increase to 1010
psig which in turn caused the high flux scram from
void collapse. Reactor water level dropped to
,
approximately 9 feet-4 inches above the active fuel.
Main steam line low pressure (850 psi) occurred
approximately 17 seconds after the scram, followed
Lumediately by a main steam isniation valve closure.
All control rods reached 90% insertion within 3.06
seconds for the 26 monitored rods. Buffer actions
appeared normal on all monitored rods.
j
l
Just prior to the scram, the condensers were being
i
backwashed, which caused the load to decrease, then
'
increase as each condenser half's flow was reversed.
It was thought by JC that these power swings might
'
have contributed to the start of the oscillations.
Prior to resuming operation, the operators were
instructed to' decrease generator load to a more stable
cam position before backwashing condensers.
j
i
September 22e 1970
Automatic scram from approximately 95% power, which
(Scram No. 48)
was initiated by closure of the main steam isolation
valves. The valves were closed when the main steam
line pressure reached 850 psig. Reactor water level
dropped to 8 feet 7 inches above the active fuel and
theru was no data available from the 26 monitor
control rods as the recorder switch was in the off
position at the time of the scram.-
Prior to the scram, the reactor was operating at
approximately 1520 MWt. The first indication of
'the problem was a very small swing up in electrical
load, accompanied by a decrease in reactor pressure.
The operator immediately started to reduce load, but
pressure continued to drop. The pressure drop
resulted from a malfunction of the electric pressure
i
regulator (EPR) portion of the turbine steam inital
pressure regulator (IPR). The specific component
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Date:
Cause
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that malfunctioned was the MOOG valve (hydraulic
4'
pilot valve), which was found to be plugged with
.
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foreign matetLal.
l
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This is discussed more fully.in CO Report No.
l
219/70-6.
$
October 2, 1970
Automatic scram from approximately 50% power w' hich
(Scram No. 49)
was initiated by a turbine. trip.
Immediately
proceeding the scram the load was being reduced to
check the north side of the "C" condenser for salt
i
water (tube)LIgaks. Recirculation flow was at the
rated 1.6 x 19 gpm. The load had been reduced to
'
290 MWe by control rod insertions when the' turbine
trip resulted from highLlevel in the moisture
separator drain tanks. This initiated a reactor
scram from high pressure. The cause for the un-
,
planned reactor trip was attributed.to a false high
4
level from the moisture. separator drain tanks.
Control rod scram times of the 26 monitored rods
,
ranged from 2.46 to 2.92 for 90% insertion.
It was
reported that the isolation condensers were not
i
-
automatically initiated on this scram as the time-
pressure conditions were not satisfied to initiate
,,
automatic initiation,
i.e.,
reactor pressure did not
'
remain above 1040 psig for 15-seconds.
<
l
D.
Facility Procedures
i
'
Mr. F. Nolan, CO:HQ, was at the site during the period of October 13-15,
. 1970 to review:
a
4
'
1.
Assigned responsibilities for initiating required procedures or
.
j
test documents.
1
2
Review and approval methods for procedures.
1
3.- Periodic updating controls..
1
4
Controls-to assure procedure or test modification following equipment
modification.
,
5.
Controls to insure effective communications of procedure or test-
changes to operating personnel, including related retraining.
6.
Responsibilities and assignments of individuals reviewing and approving
""
' surveillance ~ test results..
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The adequacy of the surveillance test records to permit evaluating
of the results of the tests.
Mr. Nolan's findings will be included in a separats report.
.
E.
Primary System
1.,
Unidentified Leakage in the Containment
The unidentified leakage in the centainment increased from approx-
imately I spa to 4.5 spm 'during a period of about twa weaks operation.
The reactor-was shut down on. September 16, 1970 to investigate the
source of the leakage. Investigation disclosed that the source of the
leakage was a packing on the "E" recirculation pump discharge valve.
.
Following repairs to the packing, the measured indicated unidentified
leakage rate in the containment returned to approximatsiy 0.8 gpm.
This packing leak had been suspected as the unidentified leakage into
'
containment had started to increase after eyeling the discharge valve.
The "E" recirculation pump had been removed from eervice and isolated
(suction and discharge valves closed) to work on the re:irculating pump
MG set brushes on September 1, 1970.
The method used to calculate the unidentified leakage rate in the
,
containment was reviewed with Mr. J. Carroll and he informed the
inspector that the integrated flow readings of the containment sump
.
pump are obtained each hour and that the change in integrated flow
for a 24-hour period is converted into a gallon par minute leak rate.
The leak rate is plotted daily by both the operations supervisor and
'
technical engineers. Discussions were held with Mr. J. Sullivan and
Mr. J. Carroll to determine if other sources are available to
independently identify or verify leakage in the cantainment. Both
indicated that at present there is no other reliable instrument to
measure unidentified leakage in the containment although containment
temperature and humidity are plotted daily. The inspector reviewed
the temperature and humidity data and could not corralate variations in
these measured parameters with the recent increase in unidentified
leakage in the containment.
Subsequent to this September 16, 1970 shutdown, JC implemented'a program
l
to sample the containment atmosphere for radioactivity on a weekly
frequency. They: intend to use the results of the sampling program to
determine. if this method could provide a sensitive leak detection method.
Mr. Carroll also informed the inspector that the instrument used for
i
determining.the relative humidity in containnsat will be relocated to a
new position (adjacent to the temperature sensor) during the planned
,
.
October, 1970 outage. The stated purpose for this relocation was that
it'is expected to provide more meaningful data for determining. unidentified
l
1eakage in the containment.
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2.
Chemistry
.)
,
1
Primary coolant chemistry records'were reviewed for the period from
1'
-
May 28, 1970 to October 12, 1970
Typical values, as'well as the
,
-
. recorded range of values, are tabulated below:
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R_ age
Typical Value
,
.
'pH
5.8 - 7.3
6.1
< 20 - 140 ppb
.-< 20 ppb
SiO2
40 - 900 ppb
70 ppb
4 x 10-4 to 7 x 10-4 mci /mi-
5 x 10-4 mci /mi
H3
Grossd-[
5.8 x 10-4 - 1.5 x 10-2 mC1/mi
'
activity
-
1 x 10-2 - 1.6 x 10-1 uCi/ml
-
4
Conductivity
0.28 - 9.6 umho/cm
s.
A review of the records reflected that the main coolant chemistry has
remained within the Technical Specification limits for the inspection
,
period. OC-1 has experienced some salt leaks in the main condensers
,
which have been corrected by plugging of tubes.
I
3.
North Core Spray Nozzle Wall Thickness Determinations
As was.previously reported,* linear defects were observed during a
LP check of the 00-1 north core spray nozzle safe end overlay cladding.
l
i
This LP check had been made during the April - May,1970 outage as' a
l
result of the Nine Mile Point nozzle cracking problem. The investigation
of'the defects (boat sample) disclosed that the material was inconel and
not 308-L as stated in'the application. The defects (microfissuring)
were determined to be the result of weld solidification during the
application of the overlay cladding. The licensee removed all the
defects by grinding during the April - May outage and measured the
remaining safe end wall thickness by radiographic techniques. CO:I
(Tillou and McDermott) reviewed the techniques used for measuring the
wall thickness and advised JC (following the May 18-22, 1970 inspection)
-
that it would be prudent to perform additional wall thickness measure-
,
t
ments in the light of the indicated small margin over minimum co'de
requirements.
Mr. McCluskey stated that additional checks would be
'
considered.
During this inspection, Mr. McCluskey was asked if additional wall
thickness. determinations were planned. He informed the inspector that
- C0 Report No. 219/70-5, Other Significant Items No. 8 and Section E.2.
,
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Mr. N. Goodenough, QA Engineer (Radiography), GPU was scheduled to
perforia additional wall thickness determinations.during the planned
'
October 1970 outage. The results of these measurements will be
reviewed during the next inspection.
'
4.
Isolation Condenser Initiatina Lonic Circuitry
.
A licensee review of schematics was prompted by a separation of a fitting
'
l
on a primary system instrument sensing line. This separation resulted-
in a closure of an excess flow check valve, which in turn resulted in a
loss of sensed pressure.* The JC review disclosed that the high steam
pressure (1060 psig for 20 seconds) initiating logic for the. isolation
i
condensers was defeated by the closure of the excess flow check valve
4
in the sensing line. As this was not the design intent, GE was requested
-
to provide the required design change and testing procedure. Both the
'
design change and the testing procedure were supplied by GE and reviewed
7
!
and approved for installation by PORC. The design change was completed
.
in August of 1970.
The corrective action consisted of interchanging the sensors (high.pttessure)
j
that operate relays 6K10 and 6K11.
(See attached Figures 2 and 3).
The
present automatic initiating logic for the isolation condensers on sensed
high reactor pressure will not be: defeated by the closure of a single
s'
excess flow check valve (which would remove the sensed pressure from two
,
high pressure sensors). Additional fuses were added to the power supply
to relays 6K9, 6K10, 6K11 and 6K12 to preclude the failure of a single
s
fuse from negating the initiating logic circu!.try. The change details are
4
shown on Figures 2 and 3 attached.
5.
Temporary Strainers
,
' Based on the Nine Mile Point experience ** of a failed strainer -in the
feedwater system, the inspector asked Messrs. Carroll and Riggle if there
were any temporary strainers in any of the systems at 0C-1.
They
-
a
informed the inspector that temporary strainers had been installed
during the ccustruction phase, but that all temporary strainers had been
removed from the systems by February, 1970.
'
F.
Reactivity Control and Core Physics
.
1.
,
- a.
Scram Times
Station records were reviewed for the period which began on May'22,
s.
i-
1970, at the completion of the rod inspection outage. All control
rod drives were removed, repaired and reinstalled during a five-week
-
-
- C0 Report-No. 219/70-5, Section F.2.g.
I'
- Inquiry Memorandum No. 220/70-C, Niagara Mohawk Power C.orporation " Failure of
'
Temporary' Strainer in Feedwater System"
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outage in April - May, 1970. Following reassembly, cold depressurized
scram times were obtained, as were hot pressurized scram timas which
were taken on May 22, 1970.
Since that time, five scrams have resulted
j-
and meaningful scram time data has been obtained fer three of the
scrams. Listed below are the results of the measurements:
J
Date"
Range for 907. Insertion
,
Hot pressurized scram
May 22, 1970
2.42 - 3.4 seconds
(average time 2.77)
test (all rods).
.
,
Scram No. 45
June'17, 1970
2.42.- 3.11 seconds
(26 rods)
(average time 2.82)
Scram No. 46
July 11, 1970
No scram time data
obtained as the recorder
apparently stuck
'
Scram No. 47
(26 rods)
Sept. 17, 1970
2.53 - 3.00 seconds
Scram No. 48
Sept.'22, 1970
No scram tima d'ata
collected as the recorder
switch was.found in the
'
off position
Scram No. 49
Oct. 2, 1970
2.46.- 2.92 seconds
'
(26 rods)
(average time 2.72)
Buffer times appeared normal for all scrams that were recorded on
the scram time nonitor (brush recorder).
b.
Stall Flows
,
-Station records were reviewed to monitor the performance of the seals
as indicated by stall flow measurements which are taken monthly.
Tabulated below are the results of the review:
No. of individual rods with stall
Totalized stall flow
-
flows in~ indicated range:
measurements
'22: 3'gpm
3:4 gpm*
E!=5 spm*
May
4
1
1
167 gpm
June
8
4
1
191 spm
July
8
4
0
199 gpm
August
9
0
0
189.gpm
September
7
2
0
212 gpm
October
14
.0
0
218 gpm
- High individual stall flows were corrected by reworking the directional
,
solenoid; operated control valves that control normal movement of the rod.
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c.
Followup Inspection Items *
.
Mr. D. Pomeroy, TSB, CO:HQ identified four items for followup
inspection during his assist inspection at the facility on May 20
.
l
and 21, 1970. These included:
(1) The reassembly reports that
.
had been contaminated and were unavailable for review during this
5
l
!
inspection, .(2) The results of repeat friction test for four drives
l-
!
whose original test indicated marginal conformance,l(3) The results
i
of pressurized scram and stall flow tests, and (4) The results of
continuing surveillance,
i.e.,
scram times, buffer action, monthly
stall flow tests.
,
!
Items (3) and (4) are discussed in paragraphs a. and b. above.
The inspector was informed during this inspection that no reassembly
i
reports were available for. review. Mr. Pomeroy was previously
j
. informed by station personnel that these forms were contaminated
and were in the process of being copied. During this inspection,
discussions were held with Messrs. Carroll and Goodenough and they
informed the inspector that these forms had not been used, but that
the GPU QC engineers had observed the reassembly of all drives.
'
Mr. Goodenough informed the inspector that the GPU QC engineers had
rejected 15-20 drives (primarily for bulged index tubes) during the
4
inspection of the drives that otherwise would have been re-
'
installed into the reactor by GE.
4
l
During Mr. Pomeroy's May 1970 assist inspection he ide'ntified four
,
rods that did not appear.to meet the specified acceptance criteria
'
..
of a 15 psi deviation for continuous rod withdrawal. JC stated at
'
that time that individual." notch out and settle" tests would be
performed for these rods. The stated acceptance criteria for the
" notch out and settle" tests was a minimum of 30 psi settling pressure.
During the most recent inspection the records for these tests were
reviewed and in all cases the rods met the acceptance criteria.
,
i
2
Inadvertent Rod Drop
i
During power operation on September 26, 1970 a control rod inadvertently
dropped into the reactor from notch 32
On September 25, 1970 (the day
i
before) this rod had been valved out of service to remove and repair
the scram accumulator. The accumulator had been reinstalled and it was
thought.that all valving was returned to normal for the drive.
Subsequent ~
)
-
investigation of the cause of the rod drop disclosed that the scram
'
- .
' inlet and discharge valves had opened and that this was caused by.a manual
l-
air supply valve being left closed which allowed the' air pressure to
'
slowly bleed off the diaphragms of the scram valves. The air pressure
on.the diaphragms eventually reduced to'the point where the spring
- .
loading on the scram valves opened the valves and scransned the rod
,
'into the reactor. Following the finding of the valving error,-the
rod was withdrawn from the reactor to its normal position within two hours..
.
- CO Report No. 219/70-5, Addendum 4
4
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The procedure for isolating the charging scram accumulator calls for
' closing the air supply valve *in question and also the check list calls
for the valve to be reopened after the work is completed. The check
list used was reported to have been signed off by the operator but it
is believed by station management that an error was made on the operator's
,
s
part and that the valve was not reopened.
t
3.
Instrumentation Setting for Isolation Condenser Isolation (Item of
Noncompliance)*
Hr. T. McCluskey informed the' inspector during a telecon on the morning
of July 1, 1970, that the present trip settings (at that time) for the
steam line and condensate line break instrumentation (used to isolate
the condensers in the event of a line break) were 20 psig d P and 59 inches
,
4 P H O respectively. The Technical Specifications Paragraph 3.1.
2
item H.2. requires that the condensate line instrumentation be set at
2E 27 inches A P H20. At that time Mr. McCluskey was informed that the
current trip setting of.the condensate line instrumentation was in
violation of the Technical Specifications for a~1imiting 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 DRL to
discuss this situation. Mr. McCluskey info'rmed the inspector at 6:30 p.m.
that discussions were held with' DRL and that JC had decided to reduce
the trip point to the specified value and to report in writing to pRL
when this had been accomplished. JC made the required change and did
report on July 2, 1970.**
'
4
During this inspection it was ascertained that the original change in
,
instrumentation setting from 27 inches to 59 inches A P H O was made on
'
2
December 9, 1969, as a result of GE instructions to the JC maintenance
'
group. GE had requested that the instrument be set at 68 inches 6 P H O
2
but during an attempt to set the instrument at 68 inches, it was found
that the total range of the instrument was limited to 60 inches.
It
was therefore decided to establish the trip point at 59 inches. The
,
inspector was informed that the change was prompted by an inadvertent
I
i
isolation of the isolation condensers when they had been tested under
full load conditions for the first time. At the time of the inadvertent
'
isolation, the~ condensate line' break instrumentation was set at'27 inches
A P.
At that' time it was decided to incorporate the four-second time
delay (initiates isolation of the condensers four seconds after the flow
3
trip) and also to increase the trip set point from 27 inches to 68 inches
d P H 0.
.
2
During the July 1,'1970, telecon between Messrs. McCluskey and McDermott,
,
Mr. McCluskey informed the inspector that he considered the Technical
4
Specifications to be in error as GE had verbally provided calibration
'
data for the condensate line flow instrument. The calibration indicated
4
~
- Inquiry Memorandum Ne. 219/70-H.
'
,
- 1VX to Dr. P.$ . Morris, Direc. tor, DRL from Mr. I. Finfrock, Manager,
A
'
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Nuclear Gene ating Stations, JC, dated July 2, 1970.
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that a setting of 68 inches would be permissible and that this setting
would satisfy the conditions specified in the basis of the Technical
'
- _
'
Specifications of isolating the isolation condensers at less than or
1
equal to 3007 of rated flow through the condensate line. Mr. McCluskey
i
was informed at-that time that information available to compliance.
Region I.(Mr. D. Pomeroy's calculations) were not in agr:..aaent with
the utsabers provided by GE and that based on the past history of GE
.
supplying various numbers for-the trip set points, JC was encouraged
!
'
by the inspector to make an independent review.
I
.
During the most recent inspection, followup correspondence between JC
and GE on July 31,~1970, on this matter, was reviewed by the inspector.
GE stated in this correrpondence that the error in the specified set
i
7
!
- point provided by GE to JC on July 2,1970 (letter), was due to errors
1
- made in the assumptions on the elbow radius and a calculational error
involving a misplaced density term.
GE, at this time, provided a new
4
i
value of 3007. of design flow, which was 25 inches d P H O and 19.4 psig
2
for the steam.. JC requested the GPU technical support group to recheck
,
!
the figures. GPU has reviewed the latest settings supplied by GE and
their calculations agree with GE supplied numbers.
~ '
Since the instrument setting change on July 2, 1970 (back to 27 inches
j
d P. H2O), the isolation condensers have been manually placed in service,
without any additional spurious isolations of the condensers. No auto-
matic initiation of the isolation condensers (from high pressure)'have
4
l
resulted but the manual initiation.should simulate to a large degree
[
. the system response.
It appears that, based on the experience with
'
several manual initiations of the isolation condensers, the reduced
setting,
i.e., 27 inches d P H O on the condensate break instrumentation .
2
-
,
j
will not result in additional spurious isolations of the system. This
l
issue was discussed during the exit interview as e.n item of noncompliance.
J
H.
Power Conversion System
,
3-
1
Loss of Mair Circulatina Water
.
On two occasions (July 11 and August 1,1970), sea grass plugged the
traveling intake screens for the main circulating water suction wells,
'
and resulted in two reactor scrams.* The heavy accumulation of sea
t
{-
grass was reported by the licensee to be a seasonal condition which
i
results from the grass breaking off the bottom of Barnegat Bay in heavy
quantities during the summer months. During these periods, the screens
j
_are run in a'bontinuous advance" mode and station personnel man the heavy
.
.
los screens (upstream of the traveling screens) to assist in the removal
of the sea grass. On one occasion heavy sea grass accumulation on the
traveling screens resulted in the shearing of pins which then did not
.
- .
allow the screens to advance and eventually resulted in a partial loss
of the main circulating pumps suction supply. The water which is
'
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passed through the screens also supplies, in addition to the main
-
7
.
- Section C of this report.
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circulating pumps, the two service water pumps and the fotr emergency
service water pumps. These pumps would eventually losa suction if'the
main circulating pumps were allowed to continue to operate when the
.
-
. screens are plugged. Mr. McCluskey informed the inspector that there
are two separate suction baya (with separate intake. screens) and the
'
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two bays are physically isolated from each other so that a loss of a
l
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single bay does not effect the other bay. One bay supplies two main
-
circulating pumps, one service water pump and two emergency service
-
water pumps.
2
Initial Pressure Regulator' Performance
l
i
Five recent disturbances to the steam pressure-(flow) at the OC-1
facility resulted durings the period of S e p t . 17-28, 1970. These
)
disturbances have resulted in part from poor design of the cams which
'
-
operate the main turbine inlet control valves and in part, from dirt
)
,
'
within the control oil system. The disturbances had been manifested
j
by. power oscillations ranging up to i 25 MWe and in power spiking which
'
has ranged up to 55 MWe. These events are discussed fully in CO Report
No. 219/70-6.
,
K.
Containment
,
1.
Containment Inerting (Item of Noncompliance)
Station records were reviewed during the inspection.
It was observed
-
,
l
on June 4, 1970, that the 02 level within containment exceeded the
specified 57. limit.* The time period for which the 02 concentration
,
was logged to be in excess of the 57. specified was'# 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Records
disclosed a step change in 02 from. 407. to 7.57. occurred at M1300 on
'
l
June 4, 1970 and the 02 level remained 7 57. for the stated 20-hr period. 0
'
2
levels of 4 07. were recorded for a period of six- days from May 28
,
- .
through June 4, 1970.
Mr. Carroll stated that the instrument had not
!
been performing properly and that during this period, 02 samples had
!
been taken with a portable instrument and found to be within limits.
The inspector requested Mr. Carroll to provide him with records of the
i
sample results which were taken with a portable sampler. Mr. Carroll
informed the inspector after reviewing the shift supervisor and operations
,
!
log and finding no entries of 02 sampling, that apparently no 02 samples
had been taken until after June 4, 1970
PORC comunittee minutes were reviewed and indicated that the committee
-
'
did review this item on June 9, 1970. The minutes did not reflect
_
'
that any' recommendations for followup or that a Technical Specification
limit had been exceeded.
This subject was discussed during the exit interview.
Mr. D. Ross
was questioned to provide the basis for the statement in the PORC
'
-minutes that no Technical Specification limit had been exceeded.
He
~
- Technical Specification paragraph 3.5.A.6
.
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informed the inspector that JC had interpreted the Technical Specifica-
!
tions to allow for 02 to be in excess of the specified 5% limit for
l
periods of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In that the logged values of 02 in
containment did not indicate that the containsent atmosphere was in
'
excess of 5%.02 for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, they therefore concluded that
,
the specifications had not been exceeded. The inspector requested
-
.
,
l
Mr. Ross to provide the basis of how the measured 02 in containment
l
took a step change (between hourly readings) from 0% to 7.5% on June 4,
l
1970
(This is the period when this is the start of the 20-hour period
j
that the 02 was in excess of the 5% limit). Mr. Ross could provide no.
!
such basis to the inspector. The inspector informed Mr. Ross that CO:I
would review this matter further but that the tentative finding was
that the plant had operated in noncompliance with the Technical Specifica-
'
.
tions.
1
t
.
2
Unidentified Leakane in Containment
!
on September 16, 1970, the plant was shutdown when the unidentified
,
leakage rate into the containment reached 4.5 gym. The leakage had
.
i
increased during the period of September 1-15, 1970, during which time
leakege increased from 1 spa to,4.5 gym. Following the shutdown it was
i
found that the packing was leaking on the E recirculation pump discharge
j
. valve. Repairs were made to the packing and the reactor was returned to
1
service on September 17, 1970
The unidentified leakage returned to
0.8 spm following the repairs. Discussions were held with Mr.' Carroll
i
and he informed the inspector that the unidentified leakage is determined
,
!
by flow integrator readings that are taken hourly on the containment sump
!
pump. The unidentified leakage rate into containment is determined over
'
a 24-hour period by using the flow integrator readings taken at midnight
each night and calculating the average in-leakage. Mr. Carroll informed
the inspector that the other indicators that are also measured in
L
containment (relative humidity, temperature, and pressure) are not
presently able to provide a more sensitive or equally sensitive means
for determining in-leakage. Humidity, temperature, and pressure records
were reviewed by the inspector and it was noted that a poor correlation
of these parameters could be made with the increase in in-leakage that
j
occurred from September 1 to September 15, 1970.
Mr. Carroll also
informed the inspector that following the shutdown on September 16, 1970,
,
a weekly grab sempling program was implemented to obtain base line data
,
'
for future investigation into the sensitivity of using containment
g
activity as a leak detection method.
3
Containment Integrated Leak Rate Testing
].
Mr. Mc Cluskey informed the inspector that Chicago Bridge & Iron Company
(CB&I) has been contracted to perform a containment integrated leak rate
test. The test is currently scheduled for October, 1970
The results
4
.of this test will be documented in the next report.
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4
Main Steam Isolation Valves (MSIV's)
Mr. McCluskey informed the inspector that the main steam isolation
valves will be tested for leak tightness during the planned October,
1970 outage. The results of this inspection will be documented in the
,
}
next report. 'It was also made known that GE has supplied 00-1 with
i
eight pneumatic valves to replace those presently installed at 0C-1.
These valves are the pilot valves that control the closure of the MSIV's.
Dresden 2 has had poor experience with their pneumatic valves.
Mr. McCluskey_was unsure as to whether or not the pneumatic' valves
'
would be changed out.and stated that JC was currently questioning GE
to obtain the basis for the changeout'as 00-1 has not experienced
difficulties with these valves.
'
5.
Secondary Containment Testina
A review was made by the inspector of the method of secondary containment
testing.
It was disclosed that the testing is performed with all double
air-locked doors closed and sealed. The results of the recent testing
!
indicated that the leakage limit specified in the Technical Specifications
was being met. Mr. D. Ross was questioned by the inspector .to determine
if 0C-1 had plans to test secondary containment with the airlock doors
'
in various positions (one door opened and one door closed). Mr. Ross
stated that JC had no such intentions of testing in this manner due to
I
practical considerations. The inspector asked Mr. Ross if JC had
considered routine monitoring of some variable within the ventilation
'
control scheme to continuously monitor the status of secondary containment.
Mr. Ross informed the inspector that JC had not' considered a continuous
monitoring scheme. He was responsive to this question and stated ba
'
!
would review the matter to ascertain if this was feasible and practical.
i~
l
6.
Standby Gas Treatment ' Filters (Item of Noncompliarce)
!
Testing records were reviewed and it was disclosed that both the charcoal
l
and particulate filters-had not been tested within the required six-month
interval (Technical Specifications 4.5.K and 4.5.L).
The charcoal
,
filters were tested on January 31, 1970 and again on August 20, 1970
exceeding the specified six-month interval. The particulate filters
were tested on January 19, 1970 and again on August 18, 1970 also
exceeding the specified six-month interval.
Mr. McCluskey stated during
'
the exit interview that the retest of both sets of filters has now been
- -
scheduled for intervals of less than six months to prevent a recurrence.
N.
Emeraency Power
1.
Provisional Instruction 3000/1 " Survey of Security Measures for
bergency Power System'
a.
Access Control
_
The physical. arrangements and barriers to preclude or control
access were examined and the following is a description of these
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barriers and controls. The entire facility is surrounded by a
security fence, the gates to which are manned 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day by
.
1
a security guard. Access to the reactor and turbine building is
controlled by locked outside doors. .The administrative security
,
'
aspects of the facility include authorization for entry, escorts
for personnel outside the area of the administrative offices,
t
sign-in and sign-out requirements. The diesel generators and
j
their fuel oil supply are housed in separate outside buildings
(within'the security fence) that do not have controlled access.
'
1
Physical access to the diesels or.the generator controls is
restricted by locked cubicles within the structures housing the
diesel generators. There is unlocked access (within the confines
of the security fence) to the fuel oil storage tank and the fuel
j
oil supply valve from the storage, tank to the day tanks. The rooms
j
- .
within the reactor and turbine building that house the vital
rotating and electrical switch gear equipment are controlled by-
i
locked entry. Entry into these areas is authorized'for station.
staff personnel, the operations group personnel, and the maintenance
personnel. A roving operator patrols the reactor, turbine and diesel
generator buildings each shift.
,
I
b.
Controls and Control Indications
A review of schematic drawings was conducted at the site and
'
discussions were held with Mr. Riggle to determine if control room
indications would be obtained to alert the operator for abnormal
conditions that would result in a loss of availability of emergency
'
power systems or equipment. Figures 3 and 4 attached are elementary
one-line diagrams of the AC and DC normal and emergency power systems
at the facility. These figures should be used to assist the reader
1
in the following description.
,
Aa shown on Figure 3, the normal station power is fed to McC's 1A
'
and 1B from the output of the main generator during normal operation.
1
j
Breakers S1A and S1B are normally open when the generator is on the
'
line.
In the event of a main gener'ator trip or when the generator
is shut down for extended periods, S1A and S1B close automatically
to provide power to MCC 1A and MCC IB from the startup transformers
4160 volt MCC's 1C and ID are norma 11y' fed from MCC 1A
and MCC IB respectively.
In the event of an undervoltage condition
on either MCC IC or MCC 1D, its associated diesel generator starts
automatically and assumes the load of that bus. The cross-tie
breaker between MCC's 10 and ID, although normally opened, can be.
. closed to parallel at the 4160. volt level.
'
The diesel generator breakers DG-1 and.DG-2 are controlled by an' auto-
t
manual selector switch located inside the locked cubicle for each
,
diesel generator. Control room annunciation is provided when the
auto-manual' selector switch is placed in 'a manual position which
would result in a loss of auto-start capability. There are no other*~"
. front panel controls within the locked control cub'icles for the diesel
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generators that could defeat the diesels automatically being
i
placed in service and assuming the load on McC's 1C and 1D.
l
The six 460 volt motor control centers 1A1, lA2, 1A3, IB1, 1B2
and 1B3 power the majority of the operating equipment within the
{
turbine and reactor buildings. There are provisions to cross-tie
1A1 to IB1, lA2 to IB2, and 1A3 to.1B3 but interlocks are provided
i
_
to not permit cross-tieing in such a manner as to parallel the
"
4160 volt level system through the 460 volt level i.e.,
one of the
a
two associated feeder breakers for either 1A1 or 1B1 must be opened
2
l
before the cross-tie breaker is permitted to close.
1A2 and 1B2
feed the vital MCC's 1A2 and 1B2. The vital MCC's in turn power
critical power panels. Loss of either vital MCC-1A2 or MCC-1B2 would
be accompanied by numerous control room alarms. The majority of the'
auto transfer switches that feed power to vital loads and that are
!
powered from either vital MCC-1A2 or vital MCC-1B2, have "off-normal"
-
alarms which would alert the control room operator to a loss from
'
i
either of these buses. During a loss of off-site power, vital buses
MCC-1A2 and MCC-1B2 are powered from the diesel generators as the
.
feeder breakers from MCC's 1C and ID to 1A2 and 1B2 respectively
'
r'emain closed. The loss of either MG set No. 1-1 or MG set No. 1-2
i
(feed protection system panels No. 1 and No. 2) would be annunciated
in the control room as this would result in a 1/2 scram. All
'
460 volt McC's and vital switch gear are located in controlled entry
areas (locked doors) within the reactor and turbine buildings with
,
the exception of the isolation valve MCC'1AB2 which is located
on the 23 foot level in the reactor building.
'
The DC system including the emergency supplies is shown in Figure 4
i
attached. The battery chargers MGA and MGB which are powered from
vital motor control systems 1A2 and 1B2 respectively, normally assume
the station DC load.
In the event of a loss of AC motor power for
.
the battery chargers, the 125 volt station batteries (A and B) assume
the DC load and are designed to carry the load for an eight-hour
'
period. A static charger which is also powered from either vital
MCC-1A2 or MCC-1B2 is provided to accommodate' planned maintenance
'
on either MG set A or B.
All rotating and electrical switch cear
for the DC system is located in a locked room'within'the reactu.
building. Mr. Riggle informed the inspector that the opening of
the battery breakers, either Battery A or Battery B, would not be-
annunciated nor indicated within the control room but would be
detected duringethe' normal shift inspection by a roving operator of
these spaces. This condition would be detected by a loss of trickle
charging. current.
!
c.
Surveillance of Emergency Power Equipment
Discussions were held with Mr. J. Carroll and he informed the inspector
that a roving operator patrols both the reactor building and the
'
turbine building,and the diesel generator building on a shift basis.
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Check sheets that require initialing, are provided the operator to
ensure that all spaces are checked. A review of the che:k sheets
'
disclosed that there are no specific entries required for logging
.
information such as breaker positions, battery charging currents,
or any other locally monitored variables in the emergency power
,
-
system.
Mr.. Carroll informed the inspe:ter that the operator would
be expected to record all off-normal conditior.s on (Le remarks section
i
of the log sheet.
,
i
j
Surveillance checks to determine the availability and functional
operability of the emergency power systems include a 1.1-monthly
startup and partial loading (20%) for each diesel generator and the
functional test of the diesel generators (100% loading in 15 seconds)
during each refueling outage. .These tests are required ty the
1
Technical Specifications. In addition, the 125 volt station batteries
.are discharge load-tested each six months.
.
'
'
2
Diesel Generator Performance
Review of the station surveillance testing records for the diesel
4
4
generator weekly starts disclosed the following:
a.
During the period from February 21, 1970 to October 3, 1970 there
-were five instances in which No. I diesel generator did not either
start on the first attempt (three crank,ing cycles without starting)
a
or tripped from some other problem. On four occasions (February 21,
1970, July 26, 1970, September 20, 1970, and October 3,1970) diesel
'
,
i .
generator No. I would not start on the first attempt. After several
trys at starting the diesel, the diesel did start and came up to
,
speed.
On June 28, 1970, the No. I diesel generator tripped as a
,
result of a 55 relay actuation for which no cause could be found.
s
The diesel generator was restarted shortly thereafter with no trouble.
On September 20, 1970, after starting the diesel, the diesel generator
.
tripped out while trying to synchronize. This condition was reset and
,
the diesel started again satisfactorily but while shutting down the
diesel the governor was very unsteady. Following the October 3,1970,
<
event when'the diesel generator did not start automatically on the
-
first attempt, a work order was issued and repairs were currently in
progress during the inspection to relocate the holding brackets for
the dual starting motors on each diesel generator.
b. 'During a surveillance test of the No. 2 diesel generator on June 28,
1970, the output breaker opened at a load of approximately 400 kilo-
watts. No explanation could be provided to the inspector to either
~
the. corrective . action taken or the problem with the breaker.
The subject of.the diesel generators performance was reviewed during the
' exit interview. Mr. Ross was informed by the inspector that it was
apparent that'the surveillance testing records were not'being reviewed
~"
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to determine potential or real problems with safeguards equipment.
Mr. Ross was also informed that.out of five instances.of problems with
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the No. 1 diesel generator there were only two' work requests that had
j
been issue.d to correct the troubles and one of these work requests was.
incomplete in that it did not include the failure of the diesel generator-
,
to start on the first attempt. Mr. Ross informed the inspector that as
i
soon as Mr. Don Reeves completes the senior operating exam (scheduled for
'
October, 1970) he will be assigned the overall responsibility for the
'
surveillance testing program and he would be expected to review
L
surveillance testing records to detect such problems.
,
3.
DieselGeneratorStartinaBatteryTestina(ItemofNoncomplianceh
l
No records were available to determine if the quarterly temperature and
electrolyte levels for the starting batteries was being accomplished in
i
accordance with Technical Specifications paragraph 4.7.A.S.
Mr. Riggle
!
- also informed the inspector that'these tests were not being accomplished,
j
This item was identified as an item of noncompliance during the exit
interview and Mr. Ross stated that the tests would be accomplished as
required.
4.
125 Volt DC Station Load Testing
i
'
Surveillance testing records for the semi-annual station battery discharge
load test were reviewed with Mr. Riggle. Records for the A battery
indicated that on two of the last five discharge load tests, 100% of
ampere-hour capacity was not obtained. In addition, on four of the last
,
five tests, the manufacturer's recommended minimum cell voltage was
i
,
exceeded during the discharge load test.
Mr. Riggle informed the
inspector that the acceptance standards for the load testing were:
(a) 1200 ampere-hours capacity over an eight-hour discharge rate,
(b) a minimum cell voltage of 1.75 volts, and (c) a minimum battery
terminal voltage of 105 volts. The discharge test is started after
completing a 24-hour equalizing charge on the battery. A tabulated
summary of the A battery discharge testing is provided below:
,
i
A Battery
Duration of
% of
i
Beginning
Rate of
Discharge
Pilot Cell
Terminal
Ampere Hour
!
'Date
Voltane
Discharge
Test
Voltage
Voltage
Capacity
.
.3/16/69'
125
150
8 hrs
1.70
107
100%
,
9/18/69'
121
150.
8 hrs
1.79
108
100%
4
4/ 1 /70
125
150
7l hrs 40 min.
1.67
105
95.7%
9'/29/70
124
150
7 hrs 40 min
1.57
103
89.6%
10/13/70
124
150
8 hrs
1.62
105
100%
,
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Surveillance test records for the B battery indicated that on two of
!
the last four tests, the measured ampere-hour capacity.has been less
than rated. Tabulated below are the results of the tests.
4
B Battery
>
l .
$
'
'
% of
'
Beginning . Rate of
Time of
Pilot Cell. Terminal. Ampere Hour
'
~Date
Voltage
Discharge
Discharge
Voltage-
Voltage
Capacity
'
3/12/69
129
150
7 hrs 15 min
1.75
106.5
90%
9/19/69..
126
150
8 hrs
1,74
106
100%
i
3/24/70
122
150
7 hra 30 min
1.72
105
92%
i
9/23/70
125
150
8 hrs
1.76
107
100%
l
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This subject was discussed during the exit interview and used as an
,
example to indicate that management was not adequately reviewing the
-
results of surveillance tests. Management was only aware of one
(September 29, 1970)' failure of the load testing on these batteries to
meet the acceptance criteria. Mr. Ross.again indicated that Mr. Don
Reeves will assume the overall responsibility for surveillance testing
pending his successful completion of the senior operator test which was
l
scheduled for October, 1970
The inspector did not identify this issue
as an item of noncompliance but did state his concern for the failure of
i
the station management to recognize that the test had not meet the
-
specified acceptance criteria. The inspector was informed during the
exit interview that JC was currently discussing the battery performance
l
with the battery supplier to determine if replacement cells were warranted.
j
5.
Surveillance Testing of 125 voit Station Batteries - Quarterly Tests
l
(Item of Noncompliance)
.
,
'
Surveillance testing records were reviewed and they disclosed that the
quarterly tests on that 125 volt station battery had not been performed
in entirety. Specifically lacking were records to reflect that_ electro-
l
lyte' level and the temperature of every fifth cell had been measured, as
required by Technical Specification 4.7.b.3.
This subject was discussed
during the exit interview and Mr. Ross stated that it was JC's intention
to perform these tests. The inspector identified this as an item of
noncompliance.
,
6.
Diesel Generator Annual Overhaul Surveillance Testing (Item of Noncompliance)
Station records disclose that the annual diesel generator inspection
'-
required by Technical Specification 4.7.A.3. had not been completed
since the issuance of the provisional operating license on August 9,
,
1969.'_ Discussions lwith Mr. W. Riggle disclosed that a maintenance
procedure for the inspection'had been written and was awaiting PORC
~
review and approval _before its implementation. This issue was discussed
'
~ during the exit interview with Mr. Ross. He sta'ted that it was JC's
'
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intention to meet this Technical Specification 2nT fliat the' annual
'
inspections would be done during the month of November, 1970
The
'
inspector identified this issue as an item of noncompliance.
).
Q.
Radioactive Waste Systems
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1.- Carbon-14 Issue *
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,
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Discussions were held with Mr. D. Ross and'the inspector informed him
'
of the Public Health Service survey of the Yankee effluents. He was
informed that carbon-14 had been detected and he was appraised of the
4
';
relative-concentrations in both the' radioactive liquids and gaseous
wastes _ at the Yankree plant.
Mr. Ross was encouraged,'in view of the
fact that no sampling for carbon-14 has been performed at the Oyster
Creek facility, to analyze-for this radioisotope. He informed the
'
inspector that. a sanple of gaseous and liquid wastes would be obtained
and the concentrations of carbon-14 in the samples would be determined
by_an off-site laboratory.
2.
Stack-Sampling
As a' result of prior comunitments and understandings between the licensee
and 00:I, JC installed a stack sampler to sample at the 240 foot eleva-
,
.
tion. Sampling was performed during the period of July 13-17, 1970.
'
Plant conditions at the start of the,sanpling were that the plant had
been operating at
530 Hwe for three weeks with the chemistry in the
,
primary system in equilibrium. Two sample holders were installed
approximately 18 feet down from the origin of the sample point at the
'
240 foot elevation. Tabulated below are the results of the sampling
)
'
program:
.
Iodine 131
'
i
Sample Location
.
Date and Time
Base of Stack
240 Elevation
0830 6/30/70 to 1317 7/2/70
1.66 x 10-10 ccifcc
4
tI
1320 7/ 2 /70 to 1005 7/7/70
_1.73 x 10-10 et,i/cc
2
21
1010 7/ 7 /70 to 0825 7/10/70
1.70 x 1p-10 uCi/cc
0830 7/14/70 to 1600 7/10/70
P.06 x 10-10 uci/cc
-;
1555'7/14/70 to 0831 7/15/70
l'.77 x 10-10 uci/cc
,
0831 7/15/70 to 2000 7/15/70
1.48 x 10-10 uci/cc
2000 7/15/70 to 0840 7/16/70
2.06 x 10-10 uci/cc
l0840 7/16/70 to 2010 7/16/70
1.64 x 10-10 uci/cc
~2020 7/16/70 to 0855 7/17/70
1.25 x 10-10 uci/cc
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- Memorandum from J. P. O'Reilly to Senior Reactor Inspectors, " Detection of
Carbon-14 in Power Reactor Effluents", dated June 17, 1970.
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Iodine 133
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Sample Location
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Date and Time
Base of Sta;k
240' Elevation
'
[
0830 7/14'/70 to 1600 7/14/70
1.23 x 10-10 uct/c
1555.7/14/70 to 0831 7/15/70
1,41 x 10-10 uCi/cc
I
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l
(
0831 7/15/70 to'2000 7/15/70
1.19 x 10-10 uCi/cc
2000 7/1$/70 to 0840 7/16/70
2.25 x'10-10 uCi/cc
0840 7/16/70 to 2010 7/16/70
1.53 x 10-10 uCi/cc
2020 7/16/70 to 0835 7/17/70
1.76 x 10-10 uCi/cc
'
The samples were collected on a standard CESCO char;oal cartridge and
counted on a Nuclear Data model 2200 multi-channel analyzer utilizing
i
'a 3 x 3 inch NaI (T1) crystal. Calculations were based on the 0.36 and
i
0.53 Mev peak of I-131 and I-133 respectively.
,
i
.
During the sampling periods of July 14-15, 1970 and July 15-17, 1970,
,
^
charcoal cartridges were in both the 242 foot elevation and,the normal
[
stack sampler at the base of the stack,
i.e.,
filters in series offering
a check on efficiency of the individual cartridge for iodine retention.
,
'
In both instances the first cartridge (240 foot) retained greater than
+
80% of-the iodine collected.
l
'
Gelman fiber type E filter (99.57. removal for DOP) preceded the
cht.rcoal cartridge. After allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for decay of the short-
,
lived daughters of the fi'ssion gases, the filters were counted on a
multi-channel analyzer. No g emitting nuclides were noted.
Repeat tests for particulate and iodine are planned when either (a) the
fission gas levels increase by a factor of approximately 5, or (b)
>
'
evidence of long-lived (greater than eight day 1/2 life) particulate
'
i
activity is found on the routine weekly s.amples of the gaseous effluents.
3.
Current Gaseous Release Rate
,
Records were reviewed during the inspection and disclosed that the off-gas
i
3 uCi/second.
grab sample records indicated a range of 3.27 to 7.0 x 10
.
Grab samples are taken each week from the discharge of the air ejector
l
condenser after the gas has been delayed for approximately 30 minutes.
Normal flow through the off-gas line was reportted to be approximately
100 cfm.
'
!
'4
Liquid Radioactive Waste
i
.
Liquid sampling records for activity released from the facility for the
month of August, 1970, were reviewed and disclosed the following informa-
.
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. tion. Tritium - 2.56 curies identified isotopes 0.46 curies.
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T.
Facility Modifications
.
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.1.
Modifications to the Isolation Condenser Initiating Logic Circuitry
}
A change in the relay matrix logic was made in Augurt, 1970, to the
initiating circuit.to prevent the closure of a single excess flow check
s
valve from preventing automatic initiation of the condensers (1060 psi
-
1
for 20 seconds).
(See Section E.4.)
,
.
,
2.
Control Rod Drive Inner Filters
.
All operating control rod inner filters were changed to 10 mil filters
during the April-May, 1970, rod work outage.
4 .
V. +Re, liability Information
1.
Diesel Generator Performance
.,
From February to October, 1970, there have been five occasions when the
No. 1 diesel generator either failed to start on the first attempt or
l
tripped off the line due to electrical problems. During this period the
,
No. 2 diesel generator also tripped off the line'due to an electrical
,
problem during surveillance testing.
(See Section N.2.)
.
i
2.
125 Volt Station Battery Load Testing
'
Since March of 1969, the semi-annual load testing of battery cells A and
'
B failed on five occasions to meet the minimum acceptance criteria
recommended by the battery manufacturer.
(See Section N.4.)
,
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Technical Supervisor
,
D. Ross
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Radiation Protection
Chemistry Supervisor
1
Supervisor
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D. Kaulback
D. Pelrine
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2 - Technicians
2 - Technicians.
2
Ass't Technicians
2 - Ass't Technicians
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JERSEY CENTRAL POWER & LIGiff CO.-
.CO Report No. 219/70-7
,
Figure 1
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1Ao and 2Ao are relays that put the isolation condenser in service when de-
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energized,
i.e., open the condensate return line valve.
NOTE: The original circuit would have, under conditions of a loss of pressure
sensing capability for one instrument rack (RK01 or RK02), been unable to
actuate the isolation condensers on a high reactor pressure. With the
-
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modified design shown above, a loss of RK01 would only prevent 6K9 and
6K11 from proper operation as the actuating pressure switches for these
relays (RE15A and RE15B) would be the only pressure switches affected. Under
-
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conditions of a loss of one instrument rack, the power can still be
interrupted to 1Ao and 2Ao by operation of 6K10 and 6K12 which are actuated
from RE15C and RE15D and which are located in the other instrument rack
'
RK02
The circuit modifications from the old to the new involve the' inter-
change of. actuating pressure switches RE15B and RE15C which previously
actuated 6K10 and 6K11 respectively. REISA and RE15B remain in RK01 and RE150
and RE15D remain in RK02 as in the original design. Two additional.. sets of
fuses were added to provide individual fusing for all of the actuating relays
,
4
6K9,10,11 and 12 to preclude one blown fuse,from preventing proper opera-
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tion.
In the original design relays 6K9 and 6K10 were powered through a
,
-
consnon fuse as were relays 6K11 and 6K12. Thus, if a fuse had blown..the
automatic-actuating capability would have been lost.
,
.
JERSEY CENTRAL POWER & LIG}rr CO.
.CO Report No. 219/70-7
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