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O U. S. ATOMIC ENERGY COMMISSION
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REGION I I
DIVISION OF COMPLIANCE s
q Report of Inspection l
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CO Report No. 219/69-11 t-I Licensee:
JERSEY CENTRAL POWER & LIGHT COMPANY Oyster Creek i License No. DPR-16 Category B Dates of Inspection:
November 4-6, 8 and 9, 1969 l
Date of Previous Inspection:
October 7-8, 1969 l
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Inspected by:
R. T. Dodds, Reactor Inspectdr date j y w
Reviewed by:
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R. T. Carlson, Senior Reactor Inspector hatl Proprietary Information:
None SCOPE Type of Facility:
Boiling Water Reactor Power Level:
1600 Mwt
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' Location:
Lacey Township, Ocean County, N. J.
Type of Inspection:
Routine, Announced Accompanying Personnel:
D. L. Caphton, Reactor Inspector -
j November 4-6 (Prepared sections E.
and S.3.)
L. Higg'inbotham, Radiation Specialist -
November 5-6 (Mr. Higginbotham pre-i pared Sections P, Q and R of the report)
Scope of Inspection:
Review of health physics records and-records of power ascension tests; obsefve transient tests of turbine trip, generator runback, and isolation valve trip and a reactor startup; and tours of the facility.
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SUMMARY
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f Safety Items - No items of safety significance were identified during the i
visit.
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Noncompliance Items - The following three items of noncompliance were iden-tified for which a Form AEC-592 was issued.
1.
The General Office Review Board had not audited plant operations at least quarterly as required by the technical specifications.
(Section B.3.)
2.
The reactor building containment was treached on November 6 by a main-tenance mechanic who ran an electr'.a1 cord through one of the person-net airlock doors while working ra the interlock for the other door.
(Section K.1.)
3.
The decision to continue the power ascension test program following the failure of the hot core vibration test equipment was not reviewed and approved in accordance with the provisions of 10 CFR 50.59 or the technical specifications.
(Section S.2.)
Status of Previously Reported Problems - The licensee had not yet had time to respond to the previous Form AEC-592 since it was issued just prior to this inspection. However, Mr. McCluskey did indicate that corrective action had been,taken on the referenced surveillance items.
i Other Significant Items -
l 1.
Power testing of the reactor was resumed at 800 Mwt on November 6, following the repair and subsequent testing of the control rod drives.*
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Most of the 800 Mwt power ascension tests have been completed.
Still remaining to be completed were (1) rerun of the recirculation system five pump trip test, (2) loss of auxiliary power and (3) steam carry-over tests. Operation at 1200 Hwt was expected the week of November 17.
Provided there was no more significant delays in the test program, the 100-hour warranty run should be completed by December 12, 1969.
(Section C.)
- Details of the control rod drive problem are contained in CO Report No. 219/69-10.
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2.
Five scrams have occurred during transient tests since the September y
visit. Only one was considered to involve safety considerations.
The five pump trip test disclosed a possibility for a technical conflict with the " safety limit" conta,ined in the technical specifications.
It had been expected that the natural recirculation flow would exceed 20%;
.1 however, the flow was about 16-18%.
Should a five pump trip occur at 800 Mwt or above and the reactor not scram, the safety limit of 20%
power at less than 20% flow as then specified in the technical speci-fications, could be exceeded. The licensee subsequently requested and was granted a change to the technical specifications to accurately reflect the true safety limit.
(Sections C. and S.3.)
3.
One of the Oyster Creek Nuclear Engineers has terminated his employment.
Jersey Central has now lost the services of two of three Nuclear Engine-ers who had received boiling water reactor training by General Electric.
(Section B.1.c.)
4.
It did not appear that all of the JC Shif t Foremen were exerting them-selves to the extent necessary for complete shift takeover once their GE counterparts leave.
(Section B.1.a.)
5.
The unidentified reactor coolant system leakage into primary containment was about 3.3 gph.
(Section E.1.)
6.
There appears to be a rod block interlock problem because of high and low count rates (instrument overlap) on the SRMs, IRMs, LPRMs and APRMs.
(Section F.1.)
7.
The health physics program was reviewed. The overall program appeared satisfactory; however, the stack gas monitor was set to alarm at about 5 Ci/sec rather than 0.03 Ci/sec. This was an oversight that occurred because the station order had never been corrected after calibrating 3
the monitor.
(Sections P., Q. and R.)
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8.
Section S. contains the results of the power ascension tests that have been completed since the September 29-3.0 visit.
1 Management Interviews - Two management interviews were held at the conclusion of the visit, one at Oyster Creek with Messrs. McCluskey, He1 rick, Einfrock and Hess on November 9 and the other at Parsippany, New Jersey with Messrs. Ritter and Kelcec, Jersey Central upper management, on November 10.
The results of the discussion with Messrs. Ritter and Kelcec are contained in CO Report No. 219/69-12.
Results of the interviews conducted at the site will not be included'with this report since they are basically the same as those in the above listed CO report.
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DETAILS b'
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A.
Personnel Contacted:
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.o Personnel contacted during the visit included the following:
Jersey Central Power and Light Company (JC)
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T. McCluskey, Station Superintendent D. Hetrick, Operations Supervisor i
D. Ross, Technical Supervisor i
I. Finfrock, Maintenance Supervisor C. Agan, Assistant Technical Engineer E. Riggle, Instrument Foreman S. DaAlgard, Chemical Supervisor (Consultant)
D. Kaulback, Radiation Protection Supervisor J. Malone, Shift Foreman J. Sullivan, Associate Engineer General Electric Company (GE)
W. Hess, Site. Operations Manager W. Bibb, Operations Supervisor D. Diefenderfer, Principal Test Design and Analysis Engineer R. Siemer, Transient Analysis Group i
L. Oxen Shift Supervisor G. Baston, Shift Supervisor B.
Administration and Ornanization 1.
Operating Organization s.
Shift Supervision During one reactor startup on November 4 that was observed, all operator direction was provided by the GE Shift Supervisor, Mr. G.
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Baston. The JC Shift Foreman, Mr. N. Cole, just observed and l
" puffed on his pipe".
According to the facility administrative procedures, the GE Shif t Supervisor should be instructing the JC
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Shift Foreman who, would in turn direct the JC operators.
j From observations of the other three shifts, they appear to be better balanced. The Shift Supervisors and Shift Foreman worked together with most operator direction coming from the JC Shift Foreman.
.b.
Chemical Supervisor JC still has not been able to hire a station Chemical Supervisor to replace Mr. Doyle who left in April 1969. This position is currently filled by a consultant, Mr. DaAlgard.
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Technical Engineers W
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Mr. R. Sullivan (Nuclear Engineer) terminated his employment with JC on October 31, 1969 to accept a position with Bette11e Laboratories, Columbus, Ohio. JC has now lost the services of two of the three Nuclear Engineers who had received boiling water reactor training by i
General. Electric at APED, San Jose, California. According to Mr. Ross, several nuclear engineers were being interviewed this week.
In the interim, any gaps created will be filled with consultants.
It was noted that Mr. T. Robbins, MPR Associates, has almost been a full time consultant to Oyster Creek since the startup.
Mr. Ross also pointed out that there were four other young engineers in training at the site.
They have been assigned to shift work and follow the test program on shift for JC, 2.
Plant Operations Review Committee (PORC)
The minutes of PORC meeting were reviewed with Mr. McCluskey.
The following constitutes the membership of the committee:
T. McCluskey, Chairman, Station Superintendent D. Hetrick, Operations Supervisor D. Ross, Technical Supervisor
- 1. Finfrock, Maintenance Supervisor D. Rees, Consulting Specialist, GPU (member of GORB)
R. Heward, Manager Safety Division, GPU.(member of 00RB)
The two Committee members from GORB have attended very few PORC meetings.
Of the 19 PORC meetings held since the first of the year, Mr. Rees has i
only attended 3 and Mr. Heward has only attended 1.
Later, during dis-l cuseions with Mr. Ritter, he stated that this matter had been considered by GORB and it was recognized that the GORB members were not required l
to be at every meeting.
During the startup phase, attendance would be j
difficult since meetings were not always scheduled well in advance of the meeting date. Later on, the two GORB members will attend the meet-ings regularly. However, in view of the inspector's interest, this matter will be discussed at the next GORB meeting to encourage member attendance.
He did point out that the two' GORB-PORC members were aware of the problems at Oyster Creek since they were participating in the solution of some of the day-to-day operating problems.
The minutes of the meetings disclosed that PORC has been functioning in accordance with the requirements of section 6.1.C.1. of the technical specifications.
For example, the following types of items were being reviewed:
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Scram Histories and Abnormal Occurrences.
1y Description of event Ni Cause' or causes
~h Remedial action taken V-Safety significance, if any Comparison to transient analysis I
b.
Startup Test Procedures c.
Proposed Channes to Procedures d.
Proposed Channes to Technical Specifications e.
Proposed Plant Modifications 3.
General Office Review. Board (GORB)
The minutes of GORB meetings and activities were reviewed.with Mr.-
McCluskey. The following constitutes the membership of the committee:
W. H. Hirst, Chairman, General Public Utilities (GPU)
R.. W. Heward, Vice Chairman, JC-GPU G. Kelcec, JC W. W. Lowe, Pickert-Lowe Associates T. M. Snyder, General Electric Co.
W. A. Sutherland, General Electric Co.
The committee has met on seven occasions since its inception, four since issuance of the operating license in April 1969.
For the most part,'the I
minutes disclosed that GORB has functioned in accordance with the' require-I ments of section 6.1.C.2. of the technical specifications..For example, 00RB reviewed in detail the abnormal occurience which concerned 'the re-duction of scram times of control rods and the~ proposed solution to the
.f problem. Proposed facilityLchanges and proposed amendments to the tech-i nical specifications have also been reviewed and approved. The minutes did-not reflect that GORB had been reviewing all of the PORC minutes..
However,. during a subsequent conversation with Mr. Ritter, he ' assured the inspector that all minutes of PORC meetings have been thoroughly reviewed by the' members of GORB and, in the future, all items reviewed will be.
properly documented in the minutes.
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- h The reactor license was issued on April 9,1969. The only audit of l
plant operations by GORB per se was conducted on July 2,1969.
t This involved a two-man audit of health physics practices (Mr. R. W.
gy;j Heward, Jr., and Mr. D. H. Reppert) and a one-man audit of reactor 33 operations (Mr. W. H. Hirst). The reports of the audit indicate that-detailed reviews were made in the subject areas. The audit of re-i actor operations disclosed sot.a of the same observations about the 7
plant startup check-off lists that were made by the Compliance in-spectors during the August visit.*
The technical specifications require that GORB " conduct periodic audits of plant operations at least quarterly." Therefore, the failure to have conducted no more than one audit since issuance of the license was in noncompliance with section 6.1.C.2.d.(4) of the technical specifications. When confronted with this possible item of noncompliance by telephone on November 7,1969, Mr. Heward stated that a second audit had been scheduled for October but had been postponed until November. He did not know the exact date. The item of noncompliance was included in the Form AEC-592 that was sent to the licensee.
C.
Reactor. Operations i
The reactor was shut down on October 3 because of long control rod scram times.** The long scram times have been attributed to an inner strainer in the driv'e that was clogged with a fine rouge-like iron oxide. Apparently, the iron oxide came.from the condensate system as the various feedwater heaters were
. placed in service during the ascent to higher power operation. The clogged con-
' trol rod drive strainers were removed but not replaced. Power testing was re-aused at 800 Mwt on November 6 following the repair and subsequent testing of j
'the control rod drives.
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As of November 9,1969, most of the 800 Mwt power ascension tests had been completed. Still remaining to be completed were (1) rerun of the recirculation system five pump trip test, (2) loss of auxilliary power and (3) steam carryover tests. Operation at 1200 Mwt was expected the week of November 17.
Provided there' are' no more significant delays in the test program, the 100-hour warranty run should be completed by December 12, 1969.
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- CO Report No. 219/69-8, Section D.I.
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- Details of the control rod drive problem are contained in CO Report No. 219/69-10.
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I The following reactor scrams have occurred since the time of the last visit
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(September 30) to 0200 on November 9, 1969.
1.
Five Pump Trip Test q
1 At 9:08 a.m. on October 2,1969, a flux-flow scram occurred following a recirculation flow five pump trip test.- The reactor power' decreased to about 187 Mwt initially following the pump trip and then increased to 287 Mwt where the reactor scrammed from high flux flow indication (time from pump trip was about 80 seconds). It appears that the core flow may have decreased to about 12% at the time of the scram, however, the indicated flow was about 6%. The indicated flow following the scram was a minus 6%.
This test was not expected to scram the reactor; however, the flow in-strumentation (false low flow indication) biased the flux-flow indication to the trip point.* The test will be rerun following recalibration of the flow instrumentation and modifying the flux-flow computer to more nearby bias the scram setpoint in accordance with the limit allowed in the technical specifications (i.e.:
6(8.66x10-7) W + 67.2) percent of rated flux).
Also, the technical " safety limit" could have been exceeded (20% flux with less than 20% flow). Therefore, the licensee subsequently requested and was granted a change to the technical specifications to accurately reflect the true safety limit.**
(See Section S. of this report for additional details of five pump trip test.)
2.
Turbine Trip Test As expected, the reactor scrammed because of high pressure (1060 psig -
max pressure noted) at 2131 on November 6, 1969 during the turbine trip test at 800 Mwt, 250 Mwe. The maximum APRM reading was 112%,or 1020 Mwt.
The turbine stop valve closed and the bypass valves opened in about 0.3 seconds. The bypass valves almost had sufficient capacity to override a 50% trip. They are rated at 41% capacity but appear to hrve about 49%
full flow capacity. No significant problems were experienced during the turbine trip.
(This test was witnessed by the inspectors.)
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- CO Inquiry Memorandum 219/69-G
- Change No. 1, dated
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Generator Trip Test Prior to the generator trip test, the reactor power level was lowered i
from 800 Mvt to 780 Mwt (i.e., the bypass valves have sufficient capacity to handle a 49% load rejection ~but not a 50% load rejection). The re-actor scrammed shortly after the generator trip at 8:30 a.m. on November 8, 1969. The scram occurred from high level in the high pressure turnine moisture separator drain tank and not from load rejection. The drain tanks for the steam separator and steam reheater for the turbine are normally drained by condenser vacuum. However, it appears that, at low turbine loads, the vacuum fluctuates and the tanks do not drain properly.
This can be corrected by turning on the large drain pump for the aux-illiary flash tank. However, this was not learned until after the scram that occurred during the generator run back test. Other than the drain tank problem, the generator trip test went as expected.
4.
Generator Run-back Test The generator runback test which was started at 800 Hwt, 240 Hwe at 1230 on November 8, 1969 also resulted in a scram because of high level in the steam separator drain tank. This test was observed by the inspector.
No significant problems were noted other than the drain tank problem, 5.
Main Steam Line Isolation Valve Test As expected, the reactor scrammed at 2:00 a.m. on November 9, 1969 from 10% closure of the main steam line isolation valves. This test was observed by the inspector. The scram occurred before there was any sig-nificant increase in flux or reactor pressure.
No problems were noted during the test.
D.
Administrative Procedures 1.
Experimental Data Review t
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The subject of experimental data review was discussed with Mr. Ross.
He stated that all data is reviewed by his group, operations and event-ually by PORC. The initial review is by the JC shift engineer.
During the inspector's review of the Brush recorder charts that showed the control rod drive scram insertion times for the hot scram tests, it was seen that there was no position indication from 0 to about 65% in-sertion on control rod No.50-27.
The engineer had estimated the 10% and 50% insertion times.
Further, the 90% insertion time was not conclusive.
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The fact that these were estimated times was not entered on the recorded l
data sheet. However, the shift engineer had written Mr. Ross a note t
5 that these were estimated scram times but this notation had been over-looked. Once this situation was brought to the attention of Mr. Ross by the inspector, the rod position probe was repaired following the scram that resulted from the turbine trip test.
4 Mr. Ross acknowledged that JC should be scrutinizing all experimental data.
E.
Primary System 1.
Reactor Coolant System Leakage An audit was made to determine the quantity of reactor coolant leakage into the primary containment from unidentified and identified sources.
Mr. Hetrick conducted a test over a 3-hour period to obtain an actual sample check. The check determined that there were 10 gallons of un-identified leakage in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (ne3.3 gph) and 300 gallons of identified leakage in~one hour (5 gpm)*.
These leakage rates were well within re-quirements of the technical specifications.
Mr. Hetrick stated that normal seat leakage accounted for approximately 2 gpm of the identified leakage. Valve stem leakage and condensate from coolers provide a good part of the remainder of the identified leak-age. He further stated that no absormal problems have been experienced to date with leakage into the primary -containment.
F.
Reactivity Control and Core Physics 1.
Nuclear Instrumentation During the reactor startup observed by the inspectors on November 4, 1969, it was observed that there was a rod block interlock problem with instru-ment overlap for the SRMs, IRMs, LPRMs, and APRMs. The SRMs may go above 5 x 105 eps before the LPRM-APRMs get far enough on scale to clear their 2% low level interlock. Each of the eight LPRMs that feed each APRM system has an individual low level interlock. This is not required by the technical specifications but was felt to be a desirable feature as an annunciator since the LPRM should be taken out of service if it fails.
- Subsequently Submitted. Lemer, Ritter to Morris, dated 11/25/69.
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Each APRM channel has the required low level interlock. GE has q
proposed'that bypass switches be installed on each LPRM to be used q
during startup. Further,,JC will be requesting change to the tech-nical specifications to allow the SRMs to be bypassed when the IRMs l
are in ranges 8 and 9.
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j To get around the rod block interio,ck problem, the reactor is taken to as high a power level as possible in the startup mode with min'imum 4
core flow. The reactor is then placed in the run mode of operation and reactor power level is increased by raising core flow. This arrange-ment will no longer be necessary once the bypass switches have been in-stalled on the LPRMs and the SRMs can be bypassed in ranges 8 and 9 of the IRMs. The change request has been reviewed and approved by GORB for submittal to DRL.*
K.
Containment 1.
Personnel Air Locks - Reactor Building During a tour of the facility in the company of Mr. Agan, it was observed that the solenoid operated interlocks on the personnel double doors for the reactor building had not yet been repaired.* Further, a maintenance mechanic had run an electrical cord (used for power equipment) through the ground level SE inner door of the reactor building, thereby causing the door to be ajar.
Unknowingly, Mr. Agan and the inspector opened the door and found the mechanic working on the other door which was wide ope n.
Therefore, reactor building containment had been broken in two ways (1) the electrical cord in the doorway prevented the door from l
sealing and (2) containment was thoroughly breached when the inspector and his escort opened the door. No warning signs against opening the door had been posted.
During subsequent discussions with Mr. Hecrick, he stated that replace-j ment parts had been ordered to repair the door interlocks and that peep
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hole windows would be installed as an additional precautionary measure.
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The failure to maintain reactor building containment during reactor
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operations was in noncompliance with section 3.5.B.1. of the technical specifications.
- C0 Report No. 219/69-9, section K.2.
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Radiation Protection S
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Health Physics Administration l
Mr. Kaulback stated that new personnel are required to study the radiation 4
protection procedure manual and take quizzes on each section. A final exam of about 110 questions is given.
Mr. Kaulback personally supervises I
all training and writes the examinations which appear to be fairly com-prehensive.
Health physics coverage is provided now on a schedule of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day, seven days a week; by a minimum of one radiation protection technician and one chemistry technician, Mr. Kaulback said that this arrangement of schedule would continue and eventually plans to include one of the assistant technicians in the 7-day / week schedule.
During times of plant startup and special testing, chemistry technicians have been providing coverage about 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per day.
2.
Personnel Monitoring Section 903 of the station procedures manual describes the licensee's
" Personnel Protection and Control" services.
A spot check'of files and records disclosed no discrepancies.
Baseline bioassay and whole body a)unting have been performed.* A firm schedule for repeat checks' of this type has not been established as yet.
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3.
Startup and Shield Surveys
- Surveys were completed during the power ascension program at zero power, 80 MWt, 400 MWt, and 800 MWt. The data sheets and results were reviewed
.and it appears that the surveys were adequate to identify any problem areas with regard to general radiation levels.
Mr. Kaulback said that, with one exception, no radiation levels were found that were unexpected or would present any problems. The exception was radiation levels ex-perienced in the TIP room when the probes were retracted into their j
shields. This area is located on the other side of the back wall (about j
i 18 inches concrete) of the counting room and, with the TIPS retracted, the resultant increase in background on health physics counting in-strumentation was undesirable. The procedures and immediate corrective action has been to relocate one counter and to caution the HP technicians to ascertain the status of equipment before beginning an instrument count.
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- C0 Report 69-1, section V.D.
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j Final corrective action, according to Mr. Kaulback, will-be the' addition
.of shielding in the TIP room as a first attempt. 'This item will be l
l followed by the inspector to determine the final solution.
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4 Radiation Monitorina Systems 7
A review of records and discussions with Mr. Kaulback indicated the
'l following with regards to monitoring systems. Those indicating.need for l
followup will.be checked during subsequent inspections.
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Initial calibrations of the monitors were performed using adequate
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procedures and techniques.
Calibration records are adequate.
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b.
The inspector found that the setpoint for the stack monitor was higher than any allowable release limit stated in technical-specifications section 3.6.
The alarm was set at an equivalent rele'ase rate of about 5 Ci/see where the allowable'15 min, release rate is 3 Ci/sec.
(Alarm not required by technical specifications.) During the dis-cussion with management this was found to be an oversight and/or=
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misunderstanding of instructions with regard to making the alarm set-l ting.
Mr. McCluskey said that this matter would be corrected immedi-
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ately. 'The alarm will be reset to an equivalent release rate of 0.03 Ci/sec, which is 0.1 of the continuous annual average. release rate.
1 c.
The point at which the.off-gas monitor performs its shutdown and/or 1
isolation function was determined to be about 0.3 Ci/sec, which is
.l the technical specification limit for continuous annual average re-lease rate for gaseous radioactivity.
"I" d.
No action has been taken as yet to demonstrate the adequacy of rep-resentative. sampling in the stack monitoring system.*' Mr. Kaulback said that this was planned during-operation at full power.
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l No action has been taken as yet to calibrate the off-gas line mon'itor.* '
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Mr. DaAlgard said that GE personnel have been taking grab samples of
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the off-gas periodically but that'results of.such sampling were un-known to JC personnel although requests have been made of GE to supply this data.
Mr. DaAlgard said he had been informed that'GE plans to train JC personnel in-system sampling and analysis beginning November 8,1969 af ter which JC personnel will assume the responsibility for this work.
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- CO Report 69-2, section II.A.7.
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- f.. The inspector observed the indication and operation of the off-gas
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and stack monitoring systems.as evidenced by the strip chart recorders I
in the control room. With the' plant operating at about.800 MWtlthe.
l monitors exhibited a satisfactory response to off-gas concentrations:
~ Jj of-about'50 uCi/sec.(calculated from licensee's calibration data).
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The authotized continuous annual' average release rate. for: noble gases i
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is 3 x 105 uCi/sec.
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Q.,Radioactivo Waste' Systems Discussions with Mr. DaAlgard, consultant chemist, and a review of chemistry records indicate that the licensee is conducting an adequate chemical. surveillance program. The program includes sampling and analysis required by sections 4.6.C.
.and D. of the technical specifications.
R.
Environmental Monitorina 1
' Discussion with Mr. Kaulback and a review of records indicates that the pro-gram is now conducted at the same. level and scope of the preoperational program.* '
S.
Experiments and Tests The results of the power ascension program to November 9 (operation up to 800 Mwt) were reviewed with Mr. Diefenderfer. The review included an examination of appropriate charts and data. The following is a brief summary of significant i
startup test results not previously reported by the inspector. The' opinions ex-
'l pressed-are those of the GE test engineer and/or Mr. Diefenderfer.
1.
System Expansion System expansion measurements made after the installation of re:irculation pump "D" did not disclose any anomalies with predicted values.--
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2.
Hot Core Vibration Measurements When questioned about the failure of the hot vibration test equipment for reactor vessel intervals, Mr. McCluskey stated that the failure of the incore test equipment had not been reviewed in accordance with the pro-visions of 10 CFR 50.59.
Section 3.0 of Amendment 48 explicitly states that sufficient redundant equipment would be installed to verify the hot
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vibration characteristics of the reactor vessel internals.
Mr. McCluskey added that, had it been realized Amendment 48 definitely committed JC i
- CO Report 69-2,Section II. A.7.
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J to obtaining hot vibration test results, the failure of the equipment on August 28, 1969 and the decision to proceed with the power ascension program would have been reviewed and CO would have been promptly in-formed. JC and GE were under the assumption, as a result of meetings 1
with ACRS and DRL, "we will get the information if it is at all possible."
Failure to have reviewed the loss of core vibration instrumentation and the decision to continue the power ascension test program without this confirming information was in noncompliance with the provisions of 10 CFR 50.59.
Further, sections 6.1.C.1.d.(4) and 6.1.C.2.e.(2) of the Technical Specifications require PORC and GORB to review and approve for implementation changes or modifications to plant systems or equipment provided such changes or modifications do not involve unreviewed safety questions. This item was included in the Form AEC-592 that was sent to the licensee.
3.
Recirculation System Performance The recirculation flow-reactor power transient resultin8 from recircula-tion pump trips (one and five pump trips) will be measured and evaluated at power levels of 800, 1200 and 1600 Mwt. The results of a single pump trip and then a five pump trip at 800 Mwt on October 2, 1969 were as follows:
Condition Power Level Change MCHFR One pump trip 755 Mwt to 592 Hwt 5.65 to 5.60 Isolated one loop 755 Mwt to 626 Mwt No change Five pump trip 800 to 187 to 287 Mwt (scram)*
5.65 to 5.4 l
{
The following information concerning the five pump trip test was obtained during an inspection on October 7, 1969 by Mr. Caphton, Reactor Inspector, CO:1.
j Following the pump trip at 800 Mwt, the reactor power decreased to approx-
)
imately 187 Mwt during pump coast down and development of convection cir-culation of the primary coolant; then, power gradually increased as en-thalpy changes occurred within the coolant until APRM flow biased scram tripped at 287 Mwt average reactor power.**
1
- Technical specification 2.1.B. "When the reactor pressure is less than 600 psia l
or reactor flow is less than 20% of design, the reactor thermal power shall not exceed 320 Mwt.
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The eight individual APRM recorders ranged from 27% to 35% or 252.5 Mwt to 327.6 Mwt, respectively. The reactor pressure was 988 psi
- at the time of the scram. The total core flow obtained from a GE test recorder was an approximate indicated 6%* at the time of the scram. The scram 7
occured approximately 80 seconds after the initiation of the five pump trip.
f.
~
The inspector determined from a review of the actual recorder charts monitoring the essential parameters that the technical specification limits had not been exceeded.
The total core flow recorder printed below zero af ter the scram.
Mr.
Hetrick stated that the five primary system individual venturi-transmitter instrumentation had been calibrated at cold conditions, therefore the reason for the flow error (negative indication) at hot conditions.
It appeared that the total core flow may have been greater than 6% at the time of the scram considering the amount of negative flow indicated on the recorder chart.
Mr. Hetrick stated that hot condition calibrations would be made before resuming power. operations.
The inspector discussed the fact with Mr. Hetrick that the margin between the actual average reactor power and the 320 Hwt safety limit for the 50%
power level recirculation pump trip was only 33 Mwt and the possibility existed for the 320 Mwt safety limit to be exceeded if a five pump trip were to occur at power levels above 501.
Mr. Hetrick stated that it looked as if the Technical Specifications limit was too restrictive.
Discussions with Messrs. McCluskey, Hetrv:k and Ross determined that the flow biased APRM scram set point had actually been more conservatively set than would be expected at higher power levels.
It therefore appears that resolution of this matter would be necessary before proceeding above 50% power level to avoid the potential for exceeding a technical safety limit. The inspector urged Mr. McCluskey to seek prompt resolution of this matter.
I
?
4 Recirculation Flow Control r
Plant response (load following) to recirculation flow control will be determined at 800 Mwt, 1200 Mwt and-1600 Mwt. The results of test at 800 Mwt performed on October 2 were as follows:
- See recorder charts, Attachment 1.
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s Initial Final-Relative l
Time Power Level Core Flow Channe Power Level Dampenina Ratio l
0210 800 Mwt 160 to 144,000 gpm 746 Mwt'
.0.12 ft 0230 800 Mwt 160 to 126,000 gpm 682 Mwt (0.10 t
y 0240 695 Mwt
.126 to 113 to 100,000 gpm
.556 Mwt (0.10' 0250' 556 Hwt
'100 to 75,000 gpm 457 Mwt
<0.10
]
0300 457 Mwt 75 to 651,000 gpm,
396 Mwt
<0.10 l
0330 396 Hwt 65 to 120,000 gpm 613 Mwt-
<0.10 i
0350 613 Mwt 120 to 160,000 gpm 771 Mwt
<0.10 I
i 5.
Steam Separator-Dryer -
~
Steam separator-dryer tests will be performed at 800, 1200 and 1600 Mwt f
for the purpose of measuring the carryunder performance of the steam separator assembly and the carryover performance'of the. steam dryer assembly.
Results of the tests performed at ;800 Mwt on September 27 I
were as follows:
- i 845 Mwt Power level Recirculation flow 160,000 gp Steam flow 2.71 x 100 lb/hr 250.001%*
Water carryover Too low to measure Steam carryunder
- Na-24 content measured to determine moisture content.
i 6.
.4 o
The inspectors were present during the turbine trip test that was con-l ducted at 800 Mwt on November 11 at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />. The results of the test
[
were as follows:
i S00 Mwt (250 Mwe)
Initial power level Peak power at scram 895 Mwt y
Peak pressure 1060 psig*
l Reactor level change -
dropped 0.5 ft.
l 150 meec l
Time for bypass valve operation i
I (5,000 gpm Core flow' change
- Scram occurred when pressure reached 1060 psig.
It appeared that the high pressure trip point would not have been reached had the reactor power level been 1% less at the start of the test.
(i.e.:
The bypass valves appear to have load capacity of about 49% of the rated reactor power level.)'
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1 7.
Generator Trio q
The generator trip test at 800 Mwt (780 Mwt) was performed lst 0839 i
hours on November 8, 1969. The generator runback test of 240 Mwe to j W,f 20 Mwe, which was performed at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br /> on November 8, was observed <
by the inspector. Results of the tests were as follows:
n Generator Trip *
+ 90 Mwt
( <1 see)
Reactor power level
+ 6 psig ( 41 see)
Reactor pressure
+ 0.3 ft. (ee2 sec)
Reactor level Generator Run Back*
No significant effect on reactor operating parameters.
- Note: A reactor scram occurred about five minutes after each of these tests. However, the. scram was caused by high level g
in the turbine moisture separator drain tanks, not;from the generator trip or run back. Apparently the level controls for the moisture separator drain tanks (2), reheater drain "a
tanks (4) and the flash tanks (2) were not properly " tuned"
~
and/or the drain system needs to be modified.
' '.8.
Main Steam Line Isolation Valve The main steam line isolation valves will be tested for proper opera-4 tion at 800, 1200, and 1600 Mwt.. On November 9, the inspector observed the test of isolation valve closure performed at a power level of 800 Mwt. The followup test which simulates-the. " steam line' break 'out-side the drywell accident" was performed afterwards.
' The four isolation valves closed in 6.7-9.4 seconds. The reactor scrammed within 0.8 seconds, from 107, closure of.the isolation v'alves.
The reactor pressure only increased 6 psig before'the scram. There was no increase in reactor power level. All systems functioned normally.
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