IR 05000219/1972004
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U. S. ATOMIC ENERGY COMMISSION
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DIRECTORATE OF REGULATORY OPERATIONS
REGION I
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RO Inspection Report No. 50.219/72-04
Subject: Jersey Central Power & Light Company ovster Creek License No. DPR-16
Location: Forked River. New Jersev Priority Category C
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Type of Licensee: BWB. 1930 MWt l
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Type of Inspection: Routine
Dates of Inspection: Juiv 5 - 7. 10. 12-14. 1972
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Dates of Previous Inspection: April 21. 1972
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Principal Inspector:
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F. S. Cantrell, Readp6/ Inspector Date Accompanying Inspectors: None Date
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Date Other Accompanying Personnel: None Date
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Reviewed By:
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R. T. Carlson, Chief, Reactor Operations Branch
' bate Proprietary Information:
None
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9604170509 960213
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DEKOK95-258 PDR
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SECTION I Enforcement Action l
Noncompliance Items
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A.
Technical Specification 3.6.C states, in part, "The maximum amount of radioactivity,
.. contained in the radwaste storage tanks out-
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side the radwaste building shall not exceed 10.0 C1."
Contrary to the above requirement the amount of radioactivity in the outside radwaste storage tanks was 12.78 C1 when inventoried on June l
28, 1972.
(Paragraph 22)
-B.
Technical Specification 3.5.B.1 requires that secondary containment integrity be maintained at all times when the reactor is operating.
Technical Specification 3.5.B.2 states, in part, "The standby gas
treatment system shall be operable at all times when secondary con-I tainment integrity is required..."
Contrary to the above requirement secondary containment integrity was compromised gnd the standby gas system was rendered inoperable on April 10, 1972 with the reactor operating when the 1-13 supply fan motor breaker was racked-out to repair the fan motor. The isolation circuit which closes the reactor building ventilation system supply dampers when the standby gas treatment system is initiated was rendered inop-erable when the 1-13 breaker was racked-out. The supply dampers were open at the time. This condition existed until April 11, 1972.
(Para-graph 30)
Safety Items.
l A.
Gags which would prevent operation were installed on the safety valves for the operating pressure leak test following the refueling outage.
System pressure relief capabilities available at the time were less than code requirements.
(Paragraph 21)
j Licensee Action on Previously Identified Enforcement Matters l
l As a result of the June 23 - 25 and July 2, 1971 inspection, two items of noncompliance with Regulatory requirements were identified in a letter from J. P. O'Reilly, Director, Region I, to JCP&L on September 14, 1971. No
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reply was necessary for Item No. 1.
In a letter dated October 1, 1971,
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JCP&L replied to Item No. 2 as follows:
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"hecently the GORB Chairman has instituted procedures that require the General Public Utilities Safety and Licensing Group to review all GORB
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Audit Reports to determine if any license violations are involved. The I
results of this review are reported to the Chairman, and investigations l
and reports to the President of JCP&L are completed as necessary." The JCP&L letter further stated, 'Tich respect to Item No. 2 of your letter, an investigation will be conducted and the results reviewed at the next GORB meeting." Contrary to the above commitment, the minutes of the next GORB meeting, which was held on November 23, 1971, did not show that the investigation had been conducted or that the results had been reviewed.
Records did not indicate any other meeting had been held subsequent to October 1, 1971.
As a followup to the February 23, 24, 25, 29, and March 1, 1972 inspection, a letter from J. P. O'Reilly, Director Region I to JCP&L dated May 19, 1972 informed JCP&L that, ".
.our inspector was unable to determine that the
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l investigation referred to in your letter (dated October 1, 1971) had been l
conducted, or that the results of the investigation had been reviewed by the GORB."
The JCP&L reply dated June 8, 1972 stated "The failure of the GORB to carry out the investigation noted in your letter of September 14, 1971 was investigated and the res,ults were reported to the Company Presi-dent in a memorandum dated October 12, 1971. Subsequently, the original l
noncompliance item was investigated and the results were reported to the Company President in a memorandum from the General Office Review Board Chairman dated November 10, 1971. Copies of these memorandums are avail-l able for your future review." The subject memoranda were reviewed during recurrent inspections and the item is considered resolved.
As a result of the February 23, 24, 25, 29, and March 1, 1972 inspection two items of noncompliance with Regulatory requirements were identified in the May 19, 1972 letter to JCP&L.
In the June 8, 1972 letter, JCP&L re-plied to the enforcement action as follows:
1.
"In connection with Item No. 1 of the enciosore to your letter, we ex-pect to complete the investigation of our tagging system by August 15, l
I 1972 and implement appropriate corrective measures, if necessary, by October 1, 1972.
In the interim, all shift foremen have been formally notified to review the Technical Specifications as they relate to the availability of engineered safeguards and to ascertain that all licensed operators on their shift are also aware of these requirements. These reviews have been completed and documented." This matter was reviewed and is considered resolved.
2.
Statements were made reporting the adoption of new administrative pro-
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-3-cedures to assure more timely review and reporting of abnormal occur-rences. This matter was reviewed and is considered resolved.
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Unresolved Items A.
JCP&L has not established a schedule for checking the relief valves on the liquid poison system.
(Paragraph 41)
B.
Oyster Creek does not have a positive means to assure that water cannot siphon from the spent fuel pool via the fill line.
(Paragraph 39)
C.
JCP&L in evaluating the current use of paddle type flow switches in plant systems.
(Paragraph 37)
D.
JCP&L is considering modification to the controls for the dampers that supply air to the cooling radiators on the emergency diesel generator.
(Paragraph 36)
E.
Eleven persons received exposures in excess of 3 rems during the April -
June 1972 Quarter.
(Paragraph 33)
Status of Previously Reported Unresolved Items A.
Reactor vessel level instrumentation - The "A" GE/MAC level indicator does not agree with the "B" GE/MAC or the Yarway level indicator.
This l
problem is still unresolved.
(Paragraph 8)
B.
The protective devices for the No. 2 cmergency diesel generator were calibrated by the Jersey Central Relay Department during the May -
June 1972 outage.
This matter is considered resolved.
(Paragraph 9)
Unusual Occurrences A.
The scram dump volume level switch failed to actuate a high level alarm during a surveillance test on March 1, 1972 (Inquiry Report 219/72-04 and letter from JCP&L March 10, 1972).
B.
When one of the reactor building supply fans was racked out for mainten-
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ance on April 10, 1972, the isolation circuit associated with the reactor I
building ventilation supply dampers was rendered inoperable (IR 219/72-06
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and JCP&L letter April 20, 1972).
(Paragraph 30)
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C.
Following a scram on April 13, 1972, four control rod drives settled at notch "02" (IR 219/72-07 and JCP&L letter May 30, 1972).
(Paragraph 32)
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When the mechanical vacuum pumps were started during a plant atartup on April 14, 1972, the stack release rate reached 330,000 uct/sec.
(IR 219/72-08 and JCP&L letter May 30, 1972)
E.
An. inspection in the torus May 2 - 3, 1972 showed that f'ive baffles were loose (IR 219/72-11 and JCP&L lettar June 2, 1972).
(Paragraph 4)
F.
One fuel assembly was determined to be misoriented 90* while " sipping" fuel assemblies during the refueling outage.
(IR 219/72-12 and JCP&L letter May 24, 1972).
(Paragraph 7)
l G.
A design evaluation of the reaction forces on the discharge piping.from l
the electromatic relief valves indicated that the supporta might be un-der designed (IR 219/72-13 and JCP&L letter August 22, 1972).
(Para-graph 31)
H.
The expansion joint in the discharge line of the
"A" emergency service water pump failed during a surveillance test on June 15, 1972 (IR 219/
72-14 and JCP&L letter June 26, 1972).
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One of the main steam line iso'lation valves leaked excessively during
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a surveillance test on June 16, 1972 (IR 219/72-15 and JCP&L letter l
June 26, 1972).
(Paragraph 20)
J.
An odor of burning insulation was detected from a relay that operates the seal in contents for the outside MSIV's.
(IR 219/72-16 and JCP&L letter June 26, 1972).
(Paragraph 35)
K.
The cooling radiator shutters on the No.1 emergency diesel generator failed to open during a surveillance test (IR 219/72-17 and JCP&L
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1cteer June 26, 1972)
(Paragraph 36)
L.
The activity inventory in the outside radwaste storage tanks exceeded 10.0 ci on June 28, 1972 (IR 219/72-18 and JCP&L letter July 11, 1972).
(Paragraph 22)
M.
A leak test was performed on the primary system with the safety valves j
gagged on June 16, 1972 (IR 219/72-19)
(Paragraph 21)
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Eleven men received whole body doses in excess of 3 rems during the l
second quarter 1972 (IR 219/72-20 and JCP&L letter August 10, 1972).
(Paragraph 19)
O.
During an inspection of the turbines during the September - October 1971
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and.the May - June 1972 outage, 374 turbine blades retaining pins were replaced (17 due to cracks).
(Paragraph 24)
Flux shapes with a peak at the top and the bottom have been observed at
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Persons Contacted Mr. T. J. McCluskey, Station Superintendent
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Mr..D. A. Ross,, Technical Supervisor
- Mr.~ E. I.~Riggle, Maintenance Supervisor Mr.' Don Reeves,_ Engineer Mr. R. Staudnour,. Engineer Exit Interview - July 13. 1972 The following' subjects were discussed at the exit interview with Messrs.
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I-McCluskey, Ross, Riggle_and Reeves:
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A.
Program to check the relief valves on the liquid poison system - Mr.
McCluskey stated that this matter would be reviewed and that a sur-veillance testing program wculd be established as appropriate. Jer-say Central's resolution of'this matter will be reviewed during the
- next inspection.
(Paragraph 41)
h B.
Use of gags on the safety. valves while performing the operational
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hydro test - Mr. McCluskey stated that the pressure during the test was carefully controlled administrative 1y to not exceed Technical Spe-cification limits and that he did not consider this a violation of Technical Specifications, however, the safety valves will v t by gagged l
in the future unless a self actuating relief device is in service,
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C.. The temporary procedure change which placed the gags on the safety l
valves during the hydro test - The inspector stated that he did not consider that this was the type of temporary procedure change that was permitted by the Technical Specifications. Cagging one or two of.'the valves during the hydro would not change the intent of-the Technical Specifications.
In addition, there was no record that this change had subsequently been reviewed by the PORC. Mr.
'McCluskey stated that the recommendation to gag the safety valves was. reviewed with his. staff. The~ temporary procedure change was signed by two Senior, Reactor Operators as. required by Technical Spe-cifications. He stated that the minutes of.the next PORC meeting
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This matter is considered resolved pending a review of PORC meeting minutes during the next inspection.
(paragraph 21 & 25).
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D.
Exceeding 10 curies in outside tanks (12.78 C1) - The inspector sta-ted that the Technical Specification requirement to complete an in-ventory of activity in the outside tanks at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> was the minimum acceptable frequency.
If it requires more frequent inven-tories to assure compliance with the Technical Specifications, then more frequent analyses are required. A post mortum review of radwaste re-cords showed that the system was in trouble prior to the June 28 inventory and RO:I was not notified until June 30, 1972.
Mr. McCluskey acknow-ledged that more frequent inventories would be made if needed, but sta-ted that until this violation occurred, plant personnel did not believe that the 10 Ci limit could be exceeded. By the time the inventory on June 28 was completed, the tank containing the high activity water had been recycled back inside the radwaste building. The tank analy-sis and calculations were checked to determine how the inventory could be so high. An error was found in the analysis, however, the results were not verified until late June 29, 1972, and Region I was notified the next day.
The inspector stated that he accepted the explanccion given; however, a telephone report should be made in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the indicated viola-tion, not within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the confirmed violation. If a subsequent analysis shows no violation, the re-evaluation can be reported by a telephone call. The initial telephone report does not commit a licen-see to make a written report even if no reportable event occurred.
Mr. McCluskey stated that.a study would be initiated toward increasing their knowledge of the inventory of activity in the tank farm and to develop techniques'to avoid exceeding the 10 Ci limit.
(Paragraph 22)
E.
Information recorded in Forec3n's Log Book - The inspector stated that it appeared the Foremen's lug 'hould entain mo.
information about
abnormal plant operations. It is used to record surveillance testing, but it did not report that the waste neutralizer tanks overflowed or that the activity inventory in the tank farm exceeds' 10.0 C1.
After several minutes of discussion of what sh,uld be recorded, Mr. Mc Cluskey stated that emphasis would be placed on improving the informa-tion recorded.
(Paragraph 26)
F.
Note in Foremen's Log June 27, 1972
" Fire in white trailer North of Reactor Building.
Called Forked River Fire Department."
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In the event of adverse publicity it helps the AEC to know the status, and, public relations-wise, it could help JCP&L.
Mr. McCluskey stated he would try to relay this type of in-formation to R0 in a timely manner.
G.
Siphon breakers in spent fuel pool inlet line - Mr. McCluskey agreed to investigate a means of testing the check valves in the inlet line or to evaluate other methods to prevent a siphon from being established.
The resolution of this matter will be reviewed during the next inspec-tion.
(Paragraph 39)
H.
Sensitivity of containment leak detection systems - Mr. Ross stated that test work to supplement the present sump method of leak detection had been unsuccessful, and that the results of this testing would be reported in the January - June 1972 semiannual Report. A purchase order for a constant air monitor that will measure particulates, iodine, and gaseous activity is being considered; however, at present there are no spare pentrations into containment There has been no decision reached on the resolution of this matter.
(Paragraph 29)
I.
Paddle type flow switches - Mr. McCluskey agreed to have the PORC eval-uste the use of paddle type flow switches with respect to replacing these switches with a different type switch or to justify their contin-ued use.
(Paragraph 37)
J.
Modification to the standby gas treatment system isolation circuit -
Mr. Riggle stated the modification could be made while the plant was operating, and Mr. McCluskey stated that work would be complete by January 1, 1973.
(Paragraph 30)
K.
Relay failure in reactor protection system - Mr. McCluskey stated that subsequent testing had shown that the relay was still operable and capa-ble of it t forming its intended functions. He stated that he did not con-
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a reportable item per the Technical Specifications; however, a report has been submitted. The inspector stated that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> no-tification to R0 of an unusual event or failure does not automatically require a written report to be submitted for an unreportable event.
Mr.
McCluskey stated that he was embarraased that the relay was assumed to be bad when it was replaced and was not tested prior to submitting the written report of the failure.
(Paragraph 35)
L.
Failure of the dampers to open on one of the emergency diesel generators -
Mr. McCluskey stated that the failure is being reviewed with the vendor j
to determined if any modifications were recommended and how long the EDG
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-8-can we operated with the dampers closed. Any Ja:ommendation will be re-viewed by PORC and following this review the operators will be given spe-cific instruction for action if the dampers fail to open during an emer-gency statt, le., keep hands off, or shut down or other instruction.
Mr. McCluskey agreed to inform the inspector of the specific instruction given to the operators.
(Paragraph 36)
M.
Health Physics surveys while removing the tube bundle from the concen-trator-- The inspector stated that the air activity samples and surveys for contamination were not taken in a timely manner while removing the tube bundle.
It was pointed out that even though previous experience had shown the contamination was tightly adhering, good HP practices require that surveys and air samples be taken at the beginning as well as during the work. Air samples were not obtained until after the tube bundle was removed from the concentrator and the tube bundle was posi-tioned for cleaning.
Mr. McCluskey agreed samples should be taken at the beginning of the job and that this point would be re-emphasized to the HP Group.
In addition, he stated that another concentrator tube bundle was beit.g
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purchased and this will allow the tube bundle to be stored for a period while activity decays before attempting to unplug the bundle.
The inspector -stated that he would review the results of the whole body count during the next inspection, however, he would like to be notified if the results showed any abnormal uptakes.
Mr. McCluskey agreed to do so.
(Paragraph 23)
N.
Inspection of pipe hangers in the drywell - Mr. McCluskey confirmed that a program was being implemented to identify and inspect all pipe hangers in the drywell, and that this program would be imple-menced at least by the next scheduled refueling outage (spring 1973).
(Paragraph 28)
0.
Follow up report on safety valve cracking - Mr. McCluskey stated the followup report would be submitt'ed to Licensing by August 1, 1972.
(Para-graph 27)
P.
Report on additional restraints for the relief valve discharge piping -
Mr. McCluskey stated that the final report would be submitted to Licens-ing by August 25, 1972.
(Paragraph 31)
Q.
DC operated pressure switches in the reactor protection system or safeguards system - Mr. McCluskey cenfirmed that all of the DC operated pressurc switches in the subject system would be replaced with a GE type BZR-169 AC-DC rated switch during routine surveillance testing during the next three months.
(Paragraph 38)
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Flux wire samples removed from the reacter during the October - Novem -
ber 1971 outage - Mr, McCluskey stated that the results of the ana-lysis of these samples were being reviewed by Jersey Central and Gen-eral Electric and that the results would be reported in the July - De-cember 1971 semi-annual report if the analysis confirmed previous cal-j culation.
If not, the sample analysis would be the subject of an spe-
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cial report.
(Paragraph 40)
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S.
Personnel Overexposures - The inspector asked about Jersey Central's plans to investigate the eleven overexposures in the second quarter of
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1972.
Mr. McCluskey stated in a subsequent telephone call that the Chairman of the GORB had been directed by the President of JCP&L to investigate '.he overexposures and submit the required report to the j
Commiss ion.
(Mr. McCluskey stated he was aware of the 30 day report
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requirement.)
(Paragraph 33)
Meeting with the General Office Review Board July 27, 1972 The inspector met the members of the GORB at the beginning of the regularly
scheduled meeting and discussed the importance of GORB in assuring the safe
operation of Oyster Creek-1. Emphasis was placed on the need for audits of j
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plant operations by GORB and documentation of GORB's activities including
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the basis for decisions or recommendations. The inspector did not attend the the meeting, per se,
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SECTION II_
Additional Subject _s Inspec_ted', Not-Identified in Section"I,~Where-No D_oficiencies or Unresolved Items Were Found-
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G_eneral-L The reactor operated at gradually reduced power levels (1885 - 1830 MWt) from January 28, 1972 until April 13, 1972 when'the reactor Jacrammed on low water level as a result of a. valving error that caused a feedwat'er,, ump trip.
Subsequently the reactor' scrammed on' April 24,.
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1972 because of. low condenser vacuum as a result of a loss of'a trans-fer-pump sesi. The plant was returned'to power. and subsequently. shut down for the refueling' outage.on May 1, 1972. A turbine trip test was t
performed.from 1830 MW to initiate the shutdown. All systems responded as expected..
During this period,-stack release rates increased to approximately 107,000
'uCi/sec. Reactor power was generally adjusted as necessary to keep release rates less than'100,000 uC1/see..(TS-limit 0.21/E Ci/sec equi-'
valent to approximately 300,000 uci/sec)..During'the startup on April 15,-1972, the release rate reached 300,000 uci/sec, however, due to a lower.
-he limiting release rate was'approximately 800,000 uC1/sec.
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Relea
.ates' following the outage have been'in the range of 13,000 L
to 15,000 uCi/sec.
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The irradiated fuel was sipped to determine which assemblies con-L stained. failed fuel pins. Fifty-eight assemblies were identified as containing' failed fuel pins and were replaced. A total of 136 new fuel assemblies were charged to the core (132 GE Type II and 4 Jer-sey Nuclear' Type III*).
Twenty three control rod drives were re-l-
placed with refurbished drives (including the four drives that had F
settled at notch 02 following scrams **).
Twenty one local power range monitor strings were replaced.
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As a result of finding cracks in the' seat bushing of two spare safety l.
valves, all sixteen safety valves on the reactor were disassembled and inspected. Cracks'were found in seven additional safety valves.
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The defective seat bushings were replaced, the valves were retested
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at.the vendors shops and reinstalled.
Additional outage work performed included: replaced liquid poison
- check valves (CE Punchlist items).- inspected the' internal of all elec-tromatic relief valves, replaced the instrument taps in the throat of the main steam flow restrictors as recommended by GE FDI 339, made
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=* Described in JCP&L submittals April 6,.19,.1972
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panel connections for installation of the computer, modified second stage reheater to permit operation at full power, insta111ed addi-tional restraints on ERV discharge piping, inspected torus and re-moved 18 baffles (JCP&L letter 6/2/72),
and ran containment leak rate test.
The reactor was made critical June 19, 1972 and the plant was brought on line June 20, 1972. Power has been limited to ^'1900 MWt due to steam flow restrictions in the turbine control valves (Licensed level 1930 MWt).
2.
General Office Review Board (GORB)
Minutes for meeting during the period February 29, 1972 through June 7, 1972 were reviewed.
3.
Plans to install check valves in discharge lines of the air compres-sor and install an additional air compressor.
(Letter from JCP&L December 17, 1971)
4.
Removal of five displaced baffles and thirteen other baffles from the.corus-(Letter from JCP&L June 2,1972)
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5.
Trip of one feedwater pump and three recirculation pumps (Letter from JCP&L February 7. 1972)
6.
Ruptured expansion joint in an emergency service water discharge
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line (JCP&L letter June 26, 1972).
- 7.
Misoriented fuel bundle (90') in position 25-08 during cycle 1B (Letter from JCP&L May 24, 1972)
8.
Reactor Vessel Level Instrumentation (R0 Report 219/72-02 Paragraph 9)
The licensee's investigation is continuing as to why one GE/MAC level indicator reads 1.3 feet lover than the other GE/MAC level indicator and the Yarway level indicator.
9., Calibration of Protection Devices for No. 2 Emergency Diesel Generator (R0 Report 219/72-02 Paragraph 12)
-The protective devices were calibrated by the Jersey Central Relay Department during the May - June 1972 outage, however, the report
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had not been received at Oyster Creek.
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10. Electromatic Relief Va'.};es, j
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The' internals of all 5 ERV's were inspiscred by the licensee during the refueling outage.
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11. _Radwaste Facilities According'to Mr. Ross, a purchase. order has been issued to buy a j
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spare tube bundle for the radwaste concentrator, and is included l
.in cbs budget to install a second concentrator.
12. Containment Inerting Equipment The installation of the in-plant liquid nitrogen evaporator was com-pleted in March 1971'and first used for inerting containment April 21 - 22,1972 (when decision was-made to defer refueling outage).
13. Turbine Trip Test on May 1, 1972 14. Procedures for Purging Containment 415. Operating Voltage for DC Operated Relays q
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l 16. Load Test Station Batteries 17..JCP&L Letter of June 8, 1972, Replying to R0 letter of May 19, 1972
The corrective action reported in the JCP&L letter was verified.
l 18.-High Stack Release Rate on April 14, 1972 (Letter from JCP&L May 30, 1972
'19. Shutdown Margins Physics tests performed at the beginning of Core II demonstrated the required shutdown margin.
(1.65% AK)
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Details of Subjects Discussed. in Section I 20. Main Steam Isolation Valve (MSIV) Leakage Letter from JCP&L June 26, 1972 While checking the leakage of individual MSIV's as part of the contain-ment leak rate test, one of the inside. valves (NS03B) was found leak-ing in excess of 100 CJK. The maximum permitted leakage in 9.9 CPPr'~
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The valve was disassembled, and the seat was lapped. The pilot stem was found.out of alignment (38 mil) and was straightened. A new shaft without a cushion spud was installed as recommended by the vendor.
The valve was' reassembled and ratested with satisfactory results-(.0.1 CFM).
This valve failed to pass the-leak test in November 1971 as well as several times previously. At the direction of the General Office Review Board, a consultant has been employed to evaluate the long Lterm suitability of these valves. (Enclosure No. 4 gives a his-tory of.the MSIV leakage and the corrective action for each valve when required.).
21 1 Inoperable-Safety Valves Durina Hydro Test A note in the shift Foremen's log on June 16, 1972 indicates the gags were removed from the safety valves when reactor pressure was lowered to-approximately 870 psi following the operating hydro test. Discussions with Mr. McCluskey indicated that gags were placed on the Safety Valves upon.the recommendation of the S-Valve vendor's-representative and was done via a temporary procedure change. Records do not<show that use of the gags ware subsequently approved by the PORC.
Technical Specification 2.2 specifies the reactor coolant system pressure shall not exceed 1375 psig-whenever irraidated fuel is in the reactor vessel.
Technical Specification 2.3 Specific Limiting Safety System Settings:
Reactor High Pressure Safety 4 @ 1212 ).
-Valve Initiation 4 @ 1221 ) + 12 psi 4 @ 1230 )
4 @ 1239 )
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i-The control rod drive pump was used to provide reactor pressure, and it normally operates at approximately 1450 psi; (pump rated 1600 psi)
therefore, it was capable of exceeding the TS limit of 1375 psi with the'16 Safety Valves inoperable.
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Section III Boiler and Pressure Vessel Code, Article N-910.1 specifies that the vessel shall be protected while in service against the con-sequence of excursion of temperature and pressure, both transient and
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steady. state.
I Article'N-910.3 specifies that self-actuated safety relief devices shall not take advantage of relieving capacity of externally actua-ted. relief devices.unless they meet the requirements of Article N-911.4.
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N911.4A & B specifies that er.terna11y powered relief devices are not acceptable unless they are fail safe.
The electromatic relief valves are solenoid actuated; therefore, they are not acceptable per N911.4 A & B.
The relief valves on the cleanup system could be isolated from the reactor in the event of an isolation signal; therefore, they do not satisfy the code requirements for a self-actuating relief device.
Mr. McCluskey stated that as a result of the cracks detected in the seat bushings of nine safety valves (Paragraph 27), the vendors rep-resentative recommended that the safety valves be gagged during the hydro test to avoid the accumulation of water on the outside of the seat bushings because safety valves tend to weep when pres-surized cold.
Mr. McCluskey stated that gagging the safety valves was discussed with his staff and he made the decision to install the gags.
Instructions were issued to make a temporary procedure change to place the gags on the safety valves during the hydro test.
Mr. McCluskey stated that he felt the procedure change had been properly considered and that special administrative controls
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were in effect to avoid approaching the setpoint of safety valves.
He' stated that the administrative procedures were backed up by the electromatic relief valves and the self-actuating relief valves on the cleanup' system. Considering the above, he felt the reactor was adequately protected; however, in the future, prior to gagging the safety valves, a self-actuating relief valve will be installed that meets all interpretation of the code requirements.
22. Exceeding Activity Inventory Limit in Outside Radwaste Tanks
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(JCP&L letter July 11, 1977.)
The activity inventory that was made on June 28, 1972 showed that the outside tank farm contained 12.78 C1.
The inventory was calculated as of 8:30 a.m.
Technical Specification 3.6C limits the activity to 10 Ci in the tank farm. Pluggage of the waste con-centrator following the refueling outage caused a higher than nor-mal inventory of liquid waste. While regenerating one of the con-densate demineralizers, the A waste neutralizer tank overflowed on
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June 26 - 27, 1972. This overflow water, which had not been neu-tralized, was pumped to the floor drain collector (FDC) Tank and processed through the floor drain filter to the A Floor Drain Sample Tank (outside). As a rasult of the regeneration water not being neutralized prior to being.tranrferred to the Floor Drain Collector' Tank, additional activity may have been extracted from
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stray resin _that is believed to be in the FDC Tank or'an the floor Drain Filter.
At approximately 2:00'p.m., June 28, 1972 the plant chemist noti-fied the shift foreman that it appeared the activity in the tank farm was greater than 5 Ci and suggested recycling the A floor drains sample tank. Based on this recycling, the inventory was less-than 5 Ci by 3:30 p.m.
The complete inventory results_were not available until later in the day when all of the tank analyses were complete.
23. Radwaste Concentrator As a result of reduced flow through the concentrator due to plug-gage of the tube bundle, it was necessary to remove the-tube bundle from the concentrator on July 7,1972. This is accomplished by un-
bolting the head cf the concentrator and lifting the tube bundle out of the concentrator and placing it in a horizontal position on the roof of the cell areas. Steam is supplied to the jacket to j
saften the plug while each individual tube is reamed out.
This op-
eration hac been performed several times at Oyster Creek. A shed has been built fqr the mechanics to protect the mechanics from the
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weather. The end of the tube bundle is placed in the open end of the j
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shed, i
A survey by the rad protection personnel showed the following maximum readings:
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Radiation 1500 mR/hr @ 1" from concentrator (in cell)
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L 1000 mR/hr @ 1" from tube bundle 100 mR/hr @ 3' from tube bundle Contamination i-6780 d/m B Top of Tube Bundle 2300 d/m B Roof Area, first day
26,000 d/m B Roof Area, second day
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260 d/m B Crane j
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Air Sample
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3.9 X 10-10 uCi/cc gram B @ l' from the open concentrator flange.
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C*tma Spectrum on Air Sampla 1.62 X 10-10 uCi/cc Co-58'
1.62 X 10-10 uci/cc Co-60
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2.35 X 10-10 uC1/cc Cs-134 1.55 X 10-10 uci/cc Cs-137
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The removal of the tube bundle from the concentrator was observed by the inspector. At the time, personnel were not using respiratory protection, air samples had not been taken, and air activity was not being monitored. When questioned Mr. Reeves and subsequently Mr.
Sullivan stated that previous experience had shown that air activity
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.was not a problem, however, air monitoring was initiated after the tube bundle was positioned for work.
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Arrangements were made to have a whole body count made on all en-ployees that worked in regulated areas during the outage. The count-ing had not been performed on men involved in removing the tube bun-die. Preliminary field evaluation of the whole body counting of the men involved in the concentrator work did not show any abnormal up-takes according to Mr. McCluskey (by telephone August 4, 1972). The
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men involved in cleaning the concentrator and their estimated exno-sure for the job is shown in Enclosure No. 2.
24. Failure of Turbine Retoining Pins i
During the inspection of the A B, and C low preasure turbines dur-ing the September - October 1971 and the.May - June 1972 outages, a total of 374 turbine blade retaining pins were replaced as part of turbine blade repairs and due to cracks found in the pins (17 due to cracks).
Records supplied by Mr. Riggle showed the following distri-bution:
I-A Turbine - 95 B Turbine - 153 C Turbine - 126
The repairs were performed under GE supervision.
25. Plant Operation Review Committee (PORC) Meeting Minutes PORC meetings were reviewed for the period February 10, 1972 through
June 19, 1972. The PORC minutes did not show that the temporary pro-
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cedure change to tne hydro procedure was subsequently approved by the PORC.
Mr. McCluskey stated that this_ was on the agenda for the next PORC meeting.
26. Foreman's__ Log Reviewed the log for the period June 16,1972. through July 12, 1972.
The general comment is that the log is used to record routine infor-mation, however, there was ne note to show that the waste neatralizer tanks overflowed on June 26 - 27, 1972 or that the activity in the tank farm exceeded 10.0 Ci on June 28, 1972. A question exists as to whether all other unusual events are recorded.
Mr. McCluskey agreed that additional emphasis would be placed on expanding the information recorded and assuring that the log contained information about unusual events.
27. Cracks in Safety Valve Seat Bushings Letter from JCP&L May 1, 1972
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After a liquid penetrant inspection of the seat bushing of the five spare safety valvas (Removed from service during the October - Novem-ber 1971 outage) showed cracks in two of the. seat bushings, all of the safety valves were removed and inspected during the May - June 1972 outage.
Cracks were found in the seat bushing of seven of the sixteen installed valves. The cause of the cracking was evaluated to be chloride stress corrosion by General Electric and JCP&L's con-sultant MFR Associates Mr. J. Collins, Regulatory Operation re-viewed the analysis performed by GE and concurred in the above eval-uation.
New seat bushings were purchased to replace the cracked seat bush-
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ings.
All of the valves were decontaminated to the levels required for ship-rent to the vendors shop (Dresser) and inspected at the Todd Shipyard in Galveston, Texas. The valves were reinspected at the vendors shop ant rebuilt using new or sound parts. The inspection included a diman-sicial check as well as a liquid penetrant check. The inspection at bot h Dresser and Todd were witnessed by JCP&L representatives. The re-Ifef set point was set to the specified value on steam (1212 to 1239 psig) and the equivalent cold nitorgen relief correlation was deter-mined before the valves were returned to Oyster Creek. Mr. Ross sta-ted that as expected the radiation level of the valves with cracked seat bushings was higher than the radiation level of the valves with-out cracks.
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According to Mr. McCluskey, the vendors repressntative stated that
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28. Vibration Failure of Main Steam Line Vent'(Letter from JCp&L January 12,-1972)
The pipe supports for this line were reinspected during the May -
June 1972 outage. No problems were detected. According to Mr.
Ross, OC does not have a program to systematically inspect all l
pipe hangers, however, a program is being established to inspect all pipe supports in the drywell. According to Mr. McCluskey, this program will be implemented during the Spring 1972 refueling out-age.
29. Sensitivity of the Containment Leak Detection System The only method that Oyster Creek has for determining leakage in the containment is based on the filling and pump out times of the dryweil equipment drain tank and the floor drain sump.
The equipment drain tank receives identified leakage from pumps and valves.
(T.S. limit 25 gpm). Other leakage in the drywell is routed to the floor drain
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sump (T.S. limit 5 gpm). The leakage is calculated at least once per shift. The amount of water pumped out of the sump is measured and the total is shown on a flow integrator. The total flow over a four hour period is used to calculate the leakage rate. The floor drains sump has high and low level alarms connected to a timer that measures the filling time. The timer is set to alarm if the filling rate reaches 4 gpm (if sue tank fills in less than 20 minutes).
Based on the above, a 5 gpm unidentified leakage rate could be determined in approximately 16 minutes if the timer works - if not in four hours.
At present the operation of the sump timer is not verified routinely, however, Mr. Ross stated that a surveillance program would be estab-lished to assure proper operation.
Mr. Ross stated that with the pre-sent system, a 1 gpm leak can be determined with an accuracy of i 0.1 gpm within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. k'ith a 5 gpm leak, the accuracy is 10.02 gpm in 24 hears.
Jersey Central has previously committed itself to a research program to develop additional methods for determining leaks using grab samples of air and water.
Mr. Ross stated that this program had not been successful and that the results would be reported in the January -
June 1972 semi-annual report.
Consideration is being given to instal-ling a constant air monitor to determine particulate, iodine and gas-eous activity. The lack of sufficient spare penetrations is the main drawback to this approach.
30. Loss of Secondary containment capability (Letter from JCP&L 4/20/72)
The 1-13 breaker for the reactor building ventilation supply fan was racked out to repair the fan motor on April 10, 1972. During a sur-vei11ance test of the radiation detectors in the reactor building on
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April 11, 1972, the standby gas treatment system started as required; however, the dampers associated with the supply and did not close as required to complete isolation of the reactor building. The supply fan motor leads were removed and the breaker was racked in restoring the isolation circuit to normal.
Caution tags were placed on the controls and the ' fans to' identify the problem and to prevent recur-rence.
General Electric has recommended modification to the isolation cir-cuit (GE - FDI - 324, reactor building vent modification) to permit racking out the supply f an' breakers without deactivating the reactor building isolation circuit. Caution tags were placed on the fan con-trols and standing crder No. 14 was issued April 19, 1972 to adminis-tratively prevent deactivating the isolation circuit until the modi-fications are complete. According to Mr. Riggle the modifications will be complete by December 31, 1972.
31. Insufficient Restraint - Relief Valve Discharge Piping According to Mr. Ross a design evaluation by MPR Associates of the
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reaction forces in the relief valve discharge piping indicated mar-ginally insufficient restraint of the piping when the relief valves
initially relieve (however, valves have been tested without any re-sulting damage to the piping). The corrective action was to add three hydraulic snubbers to one line and two hydraulic snubbers to the other line...Mr. McCluskey stated that JCP&L would submit a written report to the Commission by August 25, 1972 that included the basis for the re-evaluation and the justification of the adequacy of the corrective
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action.
32. Control Rods Settling at "02" Position following a Scram Letter from JCP&L 1/25/72
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Twenty-one control rod drives were changed out during the refuelin_g
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outage including the four drives that settled at 02 position fol-
lowing a scarm. These four drives were dirty.
Two of the four had i
one or more stop piston carbon saals broken and one had a carbon bushing broken. The cooling water orifice was plugged on one drive.
j All of the drives were rebuilt, and reinstalled and tested satisfac-j torily.
j 33. Exposures (Letter from.__JCP&L_8/10/72)
Film badge results received from Landauer on July 13, 1972 for the
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second quarter of 1972 showed that 11 employees received exposures in excess of 3 rems / quarter (3010 to 3360 mrem). The exposures of
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two employees were reported as 3000 mrom and 33 employees were repor-ted with 1500 to 3000 mram. Ten of the eleven employees with expo-sures in excess of 3 rem were maintenance mechanics and one was an HP technician. At the end of the inspection, film badge results had not been received for all employees. The inspector was subsequently in-formed by telephone August 4,1972 by Mr.,McCluskey that there were no o?.her exposures in excess of 3 rem. Mr. Ross stated that indivi-dual records were kept for each employee and that estimated exposures
'(dosimeter readings) were limited to 2500 mram until film badge re-suits were received. After film. badge results are received, additional radiation work may be permitted; however, no one is allowed to (knowingly)
receive more than 3 ram. Previous comparisons have shown that the es-timated exposures were higher than the film badge results.
Mr. Ross sta-ted that the high exposures were not tracable to a particular job during the outage, that each of the men had worked on several jobs and had had
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different. assignments, however, the investigation of the high exposures j
was not complete..
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Mr. McCluskey stated that the high exposures had been reported to Mr.
Bovier, President JCP&L as required by the T. S.
During a phone con-versation on July 18, 1972,'the inspector was informed that the GORB had been directed to perform a special investigation of the high expo-
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sures that were accumulated during the outage. Enclosure No. 1 is j
a list of the persons that received 3 rem exposure or greater dur-
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ing the April - June 1972 quarter.
t-34. Axial Flux Shape
Axial flux shapes are determined for the purpose of calculating the
. minimum cr itical heat flux ratio (MCHFR) using the traveling incore
probe sys. tem (TIP). Normally the flux only has one peak region de-pending on the rod configuration.
Enclosure No. 3 is a copy of a l
TIP trace with 2 peaks. Three of the adjacent rods are at 4e step
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(fully withdrawn) and one rod is at step 08. With this flux shape, i
the McHFR was calculated to be 3.0 (Operating limit = 1.9).
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Mr. Ross indicated the saddle flux shape is a bit unusual but it has been observed before.
- 35. Relay Failure in the Reactor Protection System
JCP&L lettet June 26, 1972
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An odor of warm electrical insulation was traced to the 6 K 28 relay i
L on June 15, 1972.. The relay was replaced and the new relay was checked for proper operation. The failure was. reported to Licensing
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by letter dated June 26, 1972.
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A subsequent investigation failed to show the cause of the odor
and indicated that the relay was operable and would have performed its. intended function.
.The relay was a GE relay, type CR120A 022202AA.
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In response to a question, Mr. McCluskey stated that the apparent
failure was reported before the investigation was complete in order
to comply with TS requirements for a 10 day written report. The relay was replaced while the plant was shutdown, however, the in-vestigation of the cause of the relay failure was not completed until the plant was back on line.
i 36. Emergency Diesel Generator - Failure of Shutters to Open during l
Surveillance Test (JCP&L letter June 30, 1972)
The failure of the radiator shutter to open during a surveillance
test on June 26, 1972, caused a high temperature alarm that tripped the EDG off line. The high temperature trip is only operable in the
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test mode - not in the emergency start mode.)
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According to Mr.'McCluskey and Mr. Riggle, JCP&L has asked the vendor
'3 (General Motors) for recommendation to prevent recurrence of this par-
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ticular failure (failure of temperature sensor) and an auxiliary means j
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for opening the radiator shutters.
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According to Mr. McCluskey on August 1, 1972, a standing order No. 15 i
was issued to the operating personnel to provide interim operating in-
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struction in the event of a trouble alacm on one of the EDG's when op-
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erating in the emergency start mode. The instructions require the i
control room operator to determine that all c: the emertency equipment
associated with the other EDG is operating normally, then from the
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control rc:m shutdown the EDG with the trouble alarm, and dispatch an
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operator to determine the cause of the alarm.
If all of the equipment j
on the other EDG is not operating normally or if only one EDG is op-erating, the investigation of the cause of the trouble alarm will
be conducted with the EDG running.
37. Paddle Type Flow Switches
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Mr. Riggle stated that paddle type flow switches were installed in the inlet Jtnes to the cleanup filters, seal leakage lines from the recir-
culation pumps, the injection line from the liquid poison system, and the discharge of the dilution pumps (outside the plant).
Prints show that only the injection line from the liquid poison system-has
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Mr. McCluskey stated that the possibility of a broken paddle switch getting into.the reactor from the liquid poison system would be pre-sented to the Plant Operation Review Committee for review and recom-mendation.
38. Use of AC Rated Micro-Switches for DC Service In response to a previous inquiry about micro-switches, Mr. Riggle stated that an investigation had shown that AC rated micro-switches were being used with the isolation condenser pressure switches. These micro-swithces on the isolation condenser were replaced with a DC ra-ted micro-switch, Barksdale Model BZK-169 and was set to operate at
the recommended 60% pull in voltage. Mr. Riggle stated that the sur-vey is continuing to determina the namber of other AC rated switches in DC service.
Mr. McCluskey confirmed Mr. Riggles's statement that these switches will be replaced with the BZR-169 Model switch as sur-veillance tests are performed during the next three months.
39. Siphon Breakers for Spent Fuel Pool Fill Line An inspection showed that a check valve is installed in each fill line (2) to the spent fuel pool, however, there are no provisions for check-ing the operability of these valves. After discussing the possibility of lowering the level of the spent fuel pool by siphon action, Mr.
McCluskey agreed to investigate a method of checking the operability of the check valves or some other method of breaking a siphon.
40. Irradiation Test Specimen Holder CO Report 219/71-04, Paragraph 6 Efforts to reinstall the specimen holder that was removed from the reactor during the September - October 1971 outage were unsuccessful at the time and again during the May - June 1972 outage. Another at-tempt will be made with special tools fabricated for this purpose during the planned April 1973 outage.
- The results of the testing performed on the flux wire samples in the specimen holder have been received and are being reviewed by Jersey Central and General Electric. Following this review the results will be transmitted to the Commission by the Semi-Annual Report if the re-suits are as predicted. Otherwise the irradiation cample results will be the subject of a special report according to Mr. Ross.
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41 lelief Valve - Standby Liquid Control System According to Mr. Riggle, OC does not have a schedule for checking the relief valves, on the liquid poison system.ad as such they have not been checked since plant startup (1969).
It was pointed out to Mr.
McCluskey that if the relief valves relieved at too low a pressure, there would not be enough force to inject the poison solution into the. reactor under accident conditions.
If the relief valves failed to relieve (high pressure) with the positive displacement pumps in the system, the pump or piping could rupture and allow the poison solution to be lost via the rupture.
Mr. McCluskey agreed to review this metter and establish a test schedule as is appropriate.
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. ENCLOSURE NO. 1.
Exposure (Film Badae Raoulcs)
April --June 1972 F. Anderson 3110'
R. Keating 3050 R. Litson 3090 E. Wacha'
3090 J. Keating 3150
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F. Kossatz 3360 J. Buckalow 3050 R. Hoatson 3010
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T. Johnson 3060.
J. Groomm 3040
. T. Rayment 3030 H.'Wilkins 3000 A. Wacha 3000
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ENCLOSURE NO. 2
. Concentrator Work Exposure (Est)
Name
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DATA SHEET
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e J. G. Keppler, Chief, Reactor Testing & Operatf.ons Branch Directorate of Regulatory Operations, BQ
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RO INQUIRY REPORT No. 50-219 /72-31Q JERSEY CENTRAL POWER-AND LIGHT COMPANY OYSTER ~ CREEK - BWR EXCEED TECHNICAL SPECIPICATIONS - BOTH LIQUID POISON PUMPS INOPERABLE The subject inquiry report is forwarded for your action in that this-
,
problem may be generic. As previously reported in Inquiry Report No.
-
50-219/72-06, the effectiveness of the standby gas treatment system was
-
compromised when the integrity of secondary containment was violated by the failure of the reactor building ventilation supply fan despers to close when the standby gas treatment system was initiated during a
-
surveillance test. This failure of the dampers to close was caused by
'
racking out one of the supply fan power breakers, which defeated the logic for closing the supply dampers. This may be generic in the design of GE supplied equipment or in the equipment designed by Burns & Roe.
b addition, the Technical Specifications require specific surveillance tee.s be conducted when specific pieces of safeguards equipment are I
inopeable if reactor operation is to continue; however, the Technical Li.+a Specifmtions do not res. ire that this surretilance sa mada immediate1y.
It is recommended that future Technical Specifications be written to clearly identify the requirenv o immediately confirm the operability l
of redundant equipment when i serability is required for continued i
plant operation.
The licensee plans to conduct an investigation to determine if cther safeguards equipment has protective functions wired through power breakers s
that could be defeated if a power breaker is racked out. The results of
'
this investigation will be included in the written report to the Director-ate of Licensing. We plan to review this matter of prompt sarve111ance testing of safeguards equipment (when required) during the next routine j
inspection.
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'ISIC, DTIE and' State of New Jersey, af ter the licensee has reviewed it for proprietary information.
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'I R. T. Carlson, Chief Reactor Operations Branch Enclosure:
Subject. Inquiry Report (21 cys)
cc: RO:1:Q (5)
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f NEW ARK. NEW JERSEY 07102 RO Inquiry Report No. 50-219/72-31Q Licensee:
Jersey Central Power & Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 License No.:
DPR-16 Facility:
Oyster Creek - BWR Forked River, New Jersey Descriptive Title:
Exceed Technical Specifications - Both Liquid Poison Pumps o erable Prepared by:
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J 2-7. S. Cantrell, Reacp6
/spector Dat/e A.
Date and Manner AEC was Informed :
September 26, 1972, by telephone call from the licensee.
"Vf*!
B.
Description of Par'*.cular Event or Ciret.Mance:
The "A" liquid poison pump was removed irov. sarvice and its breaker racked out at 10:45 a.m., September 25, 1972, to tepack the pump seals.
Technical Specification 3.2.C requires a daily check of the operable liquid poison system pump when the reactor is operating if one pump-ing circuit becomes inoperable. The firut surveillance check on the
"B" pump was conducted at 4:20 a.m., September 26, At that time, the
"B" pump would not start. A controlled shutdown was initiated immediate-ly.
At 4:32 a.m. the breaker fc,r the "A" pump was racked in and the
"B" pump was started and de: onstrated to be operable. Power had been decreased from 655 MWe to 645 MWe, and at that time, the controlled ehutdown was terminated.
C.
Action by Licensee:
1.
Only one liquid poison pump is needed to inject the contents of the e
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liquid poison system into the reactor in the required time. An interlock'is provided in order to prevent operating both pumps.
at one time. This interlock operates through a contact on the power breaker of each pump, and when the breaker is racked out,
l neither pump will start. In order to complete repairs to the
'
"A" liquid poison pump, the starting permissive for the "B" pump
'['
was restored by jumpering the interlock centact on the "A" pump
power breaker during the repair period.
.
2.
Operating personnel have been instructed that when' daily surveillance checks of safeguards equipment are required because of other in-operable equipment, the surveillance check must be performed immed-intely when the inoperable condition is determined and daily-thereafter when the reactor is in operation.
<
3.
The licensee plans to check ether safeguard equipment to determine i
if any safety functions are wired through the power breakers such that the safety function of operable equipment would be defeated j
when the power breaker for the inoperable equipment is racked out.
-
4.
The licensee. plans to modify the interlocks for the "A" and "B" liquid poison pumps to prevent defeating the starting logic i
'
'
for the operable pump"when the power breaker for either pump is
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racked out.
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5.
The licensee will submit a written report to the Directorate of Licensing within 10 days as required by the Technical Specifications.
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SEP 2 61972 J. C. Kappler, Chiaf, Reactor Testing & Operations Branch Directorate of Regulatory Operations, HQ RO INQUIRY REPORT NO. 50-219/72-30 JERSEY CENTRAL POWER & LIGHT CCMPANY OYSTER CREEK - BWR EXCIMD TECHNICAL SPECIFICATION LIMITS - RAD WASTE STORAGE TANK INVEhTORY The subject inquiry report is forwarded for your information.
This makes the third time (IR No. 50-219/72-18 and 50-219/72-27)
s since the Technical Specification limit for tank farm inventory was ine:reased from 0.7 Ci to 10.0 Ci that the new limit has been exceed-The licensee was agaic.1;Jormed of our concerns regarding the
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apparent inability of Jersey Central to implement effective controls Mr. McCluskey stated with regard to rad waste storage tank inventory.
that Ja.rsey Central was putt!ng forth efforts to correct the general problem and that this included the help of their consultants, MPR As-He reiterated sociates, for both short range and long range measures.
to the the views expressed by Mr. Ross at the time of the last incident, effect that although this ocilasion was more in the line of a "one of a kind" occurrence, it still reflected negatively overall and that both he and his management shared the concern expressed by RO:1.
Further, that tilis would be factored into the review of the incident by the Plant q '
P Op. rating Review Committee. The writer requested that the General Office Review Board be informed of RO:I views with respect to these occurrences.
Mr. McCluskey stated that the GORB was scheduled to meet on September 22 and that they would be so informed.
'
We intend to continue following the licensee's actions with respect to this general problem and will keep your office informed as is appropriate.
As is noted in the report, the licensee will submit a written report within ten days as required by the Technical Specifications.
.
R. T. Carlson, Acting Senior Reactor Inspector Enclosure:
Subject Inquiry Report (original and 1 cy)
h}
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P. Morris, RO i
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RO INQUIRY REPGtT NO. 50-219/72-30
Subject:
Jersey Cectral Power & Light Compauy
.
Facility:
Oyster Creek - BWR
'
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License
.:
!
Title:
Exceed Technical Specification Limits - Rad Waste Storage
l Tank Inventory
4 Prepared by:
j R. T. Carlson, Acting Senior, Reactor Inspector Date
!
A.
Date and Manner AEC was Informed:
'!
September 21, 1972, by telephone call frees Mr. T. J. McCluskey, Plant Superintendent.
.
E.
Description of Particular Eve _at or Circumstance:
l The inventory of the _outside rad wasta storage tank farm was found
to be 27.74 Ci when a sample of the contents taken at 8:00 a.m.
on September 20 was analyzed. Technical Specification paragraph 3.6.C limits inventory to 10.0 C1,and directs that the conta sta be
,g recycled if the inventory exceeds 5.0 Ct. During operations in the
,.
.
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rad waste area on the weekend of September 16 and 17, a transfer cart moving a drum of liquid waste sludge tipped, dropping the drum
and spilling the contents. Significant decontamination was necessary
,
in order to affect repairs to equipment damsged in the incident.
Drtmaning operations were halted, pending completion of this work.
Processing of the liquid easte generated as a result of the close
,
j up operation resulted in the violation.
i i
C.
Action by Licensee:
.
Recycling of the tank farm contsats had reduced the inventory to less j
l than 10 CL by 11: 00 p.m. on September 20 and to 5 Ci by 3:00 a.m.
i
'i on September 21. As in previous cases, some of the excess inventory wau also being trucked off site by the licensee's rad waste con-l tractor, Nuclear Engineering Corporation. The Plant Operating Re-l view Consnittee and the General Office Review Board will review this violation. The licensee will submit a ten day written report of this occurrence to Licensing.
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SEP 211972
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J. C. Keppler, Chief, Reactor Testing & Operations Branch i
,
Directorate of Regulatory Operations, HQ
RO INQUIRY REPORT No. 50-219/72-29 i
JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK - BWR EQUIPMENT FAILURE - RANDOM SCRAM OF INDIVIDUAL CONTROL RODS
"
The subject inquiry report is forwarded for your information.
Mr. McCluskey was on vacation during the period of the scram and upon return
"
from vacation, was called to jury duty for over a week. Mr. McCluskey stated
that.the scram was reviewed by the PORC while he was away and was not con-sidered reportable; however, he agreed that the sequence of events was unusual.
Upon further reflection, he agreed to submit a written report to Licensing, describing the events that led up to the scram, within 10 days. While we
<
are of the view that the event was reportable under the TS, we agreed that a citation would not be made if the report was submitted as agreed. We plan to f
reveiw this matter during the next inspection.
'
s R. T. Carlson, Acting Senior Reactor Inspector Reactor Operations Branch Enclosure:
Subject Inquiry Report cc:
P. Morris,'RO H. Thornburg, RO R. Engelkan, RO
.R. Minogue,'RS (3)
R. Boyd, L (2)
R. DeYoung, L (2)
D. Skovholt, L (3)
'
H. Denton, L'(2)
h RO Files i
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RO Inquiry Report No. 50-219/72-29 l
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1, Subj ect : Jersey Central Power & LiRht Company Facility: Oyster Creek - EWR
License !;o.: DPR-16 i
Titles _E_quipment Failure - Random Scram of Individual Control Rods Prepared by: __ F. S. Cantrell, Reactor Inspector Date
,
a A.
Data and Ifanner AEC was Informed:
September 14, 1972 by telephone call frow Mr. T. J. McCluskey,, Station Superintendent.
B.
Description of Particular Event or Circumstance:
With the reactor operating at 1900 !!We, a low water level in the reactor fun initiated a reactor scram on August 25, 1972. An investigation diaelosed a loose wire on a 3way solanoid valve (NC 16 B) that supplies air to the scram valve pilot air header. The loose wire caused the solenoid i
to deenergize and close the air supply. Leakage from this header caused
)
the air pressure in the header to drop to the point ttuit individual * control i
rod scram solenoids were operating to drive the control rods into the re-Sufficient roda drove in to cause the reactor icvel to drop to
'
actor.
the initiation point of a low water level scram (minimuci level reached was i
9 feet 2 inches above the fuel).
C.
petion by_ Licensee:
The loose wire was believed to be caused by vibraticn. Other connections
,
'
were checked for tightness. After returninF the ";eactor to hot standby conditions, scran times were measured for the six of the eight conitored rods that scramned individually prior to the low reactor level scram.
)
l Tiw average screw time was 2.68 seconds.
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AUG 2 31972
J. G. Kappler, Chief, Reactor Testing & Operatio s Branch Directorate of Regulatory Operations. HQ RO INQUIRY REPORT NO. 219/72-28 JERSEY CENTRAL POWER & LIGHT CGiPANY O'lSTER CREEK 1 - BWR OTHER - DEATH BY D't0WNING IN INIAKE CANAL The subject inquiry report is forwarded for your information.
Copies of the two newspaper reports of the drowning are attached to your copy of this inquiry report. The newspaper did not connect the event with Oyster Creek Nuclear Generating Station.
We agree with the licensee that a written report is not required. We do not plan any further action.
R. T. Carlson, Acting Senior Reactor Inspector
W Enclosures:
1.
Subject Inquiry Report 2.
Asbury Park Su,wiay Press Article dated August 20, 1972 3.
Sunday Star ledger Article dated August 20, 1972 cc:
P. Morris, RO H. Thornburg, R0 R. Engelken, RO R. Minogue, RS (3)
R. Boyd, L (2)
R. DeYoung, L (2)
D. Skovholt, L (3)
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H. Denton, L (2)
RO Files DR Central Files s,*,
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RO INQUIRY REPORT NO. 219/72-28 Subject: Jersey Central Power &_ Light Company
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Facility:
Oyster Creek 1 l
License No.:
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Descriptive Title:
Other - Deat'a By Drowning In Intake Canal Prepared by:
Date Floyd S. Cantrell, Reactor Inspector i
A.
Date and Manner AEC was Informed:
August 19, 1972, by telephone call-from Mr. T. J. McClusky, Station t.dditional to Mr. J. P. O'Reilly, Director, Region I.
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Superintendent, information was provided in telephone conversation between the assigned I
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inspector and Mr. McClusky on August 21, 1972.
Description of Particular Event or Circumstance:
B.
At approximately 10 a.m. on August 21, while crabbing from a pipeline across the south branch of Forked River, which is the intake canal for Oyster Creek Nuclear Generator Station, an eleven year old boy fell in A man in the same party attempted to help the boy to shore, the water.
but collapsed from exhaustion and drowned. Efforts by other members of the party to aid the victim were unsuccessful. The body was found by
% 11 a state marine police scuba diving team, down stream about equal dis-tance between the pipeline and the plant intake screens (400 yards apart). The Sund'ay Star-Ledger identified the victim as Ismael Lugos, age 23.
Oyster Creek was at full power at the time of the incident, with one
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dilution pump operating.
C.
Action by Licensee:
1.
The licensee assisted local authorities.
the incident for in-2.
The licensee telephoned Region I to report formation only.
The licensee does not consider the matter to require a written re-3.
port.
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ASBURY PARK, N.J., SUNDAY, AUGUST 20, 1972 -
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_ _,,,, _,, u m Vktor Gereno,11, vi ws photos of his formly and friend, Ismael Lugos (left), who collapsed and drowned yester.
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doy offer pulling Victor to safety in Forked River, Lacey Township. Mr. Lugos and the Gereno f amily, w (AP)
in the some Bronx, N.Y., oportment house had been crebb:ag in the river.
Mr. Lugos, pulling the boy The Iscy Township l'irst
} 1{}}} ~. j p^f.j (,n his back swam toward Aid Squad had sc.ta divers p, g ia the water for about one
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thore where another friend shortly after the hour'wning happened at 10
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ar,d a bystanc'er pu!!cd %.c" Silt 7Ill,. r.Lm = n safety. nut Mr. t.ums,
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But they wcce tinable to
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lalced and drown (d find Mr. I,ugos' body. They
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"lle gave his hic for the were, joined, la the afternoon.
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by state. Marine Pohce and
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- aid Sgt' I Wilham hs a state pouce scuba div.
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.'N isc team. Philhp Powell and 23 ) cur.old IIront m:m n, lh 5
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I drowned in the math branch :.nd Army ' Sgt. Ih.ver
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the body, police said. at
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of l'orked Rher yem rd.sv n1K Peakrboa Tomhip, about 1 p m.
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nitor -saving an ll year.ui,1 Irnf tu ra.e Mr. Lu.;us bt.t Authoritics raid Victor. his uere unab!c to eu to parents. Mr. Lugos and Mr.
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Velez live in the same apart.
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.Whoritics w ere unsure
,,,,,. h e the l'ov Lipml to la 1. ment complex in the lironx.,
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l'oy Victor th rena, had !.41 Jaa. r.<. cr.<',arra n ere fre.
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cn from a gas pipe spann:n;: r:m ntly drop the it knos from
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Section One: Page 28 iO SUNDAY ' STAR. LEDGER, August 20, 1972
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A'MO U00 Prtu W rg)"OlO Victor Cerena. I1 sits in his Bronx home af ter being saved from drowning
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kOSCUer drowns
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A N. year ol(! Dronx man after the boy fell frorn a gas l
drowned in the south Franch p:pe spar.ning the river. He
of the Forked liner n Lacev mana;'ed to toa, the boy to Township. Ocean Cou:m, yes. safety on his back. but E.f ter i
I terday after rescum; an 11 Yelcz and a bvstander pul!cd j
year-old liny, tho rf:re >::.d.
the boy ashori, the exhauslet!
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The victim. lomaci 1 u00<.
I;ugos enl1; psed m the 13 foot-
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I had been (rabh:r';; ;uith the deep rner. the puhcc said.
bay, Victor Gere:u. the W. 's Vele.t N. nr.d.\\rmy Sgt
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parents and i.:.v.hi r forrL R9;:er Girt:5 of Pemberton Luis Yeler. oli of v h"m were Med to sr e Loy, but failed.
neighbors in the flitans. palire Ilis ho@ rit rrrnvered m e.
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cral noot s I r.cr r.y a scuba l
.Lugos dard ha the e ver dwir.:: k m.
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%G 2 21972 j
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J.' G. Kappler, Chief, Reactor Testing & Operations Branch j
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RO INQUIRY REPORT No. 50-219/72-27 JERSEY. CENTRAL POWER & LIGHT COMPANY
,
OYSTER CREEK - BWR EXCEED TECHNICAL SPECIFICATION LDf1TS - RADWASTE STORAGE TAE INVENTERY
,
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The subject inquiry report is forwarded for your information.
This is the second time (IR No. 50-219/72-18)~since the Technical Specification limit for tank farm inventory was increased from 0.7 Ci to 10.0 Ci that the new limit has been exceeded. The 0.7 Ci limit had been exceeded several times previously. Mr. Ross was informed of oar concern,regarding the apparer.c inability of Jersey Central to imple-ment effective controle in this area, and that we needed to be provided assurance that adequate steps would be taken to prevent further such occurrences. Mr. Ross stated that both he and his management shared these concerns and that this would be factored into PORC's review of j
the subject occurrence, to be conducted' on August 21 or 22 Es
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stated that Region I would be informed of the results of this review.
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We intend to follow closely the licensee's investigation and evaluation of this latest occurrence and will keep your office informed as is appropriate. As is noted in the report, the licensee will submit a written report within 10 days, as required by the Technical Specifications.
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l R. T. Carlson, Chief, Reactor Operations Branch i
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Enclosure:
Subject Inquiry Report cc:
R. Minogue, RS (3)
R. S. Boyd, L (2)
R. C. DeYoung, L (2)
D. J. Skovholt, L (3)
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H. R. Denton,' L '(2)
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P.'A. Morris, RO l
l H. D. Thornburn. RO R. H. Engelken, RO
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RO Inquiry Report No. 50-219/72 4 8,
Subject: Jersey Central Power 6 Light Ceny i
Facility: Oyster Creek - BWR License No.:
^
Exceed Technical Specification Limits - Radwaste Storage Tank Title:
' Inventory I
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l Prepared by:
Date R. T. Carlson A.
Date and Hanner AEC was Informed:
Au8ust 18, 1972, by telephone call from Mr. D. Ross, Assistant Plant Superintendent.
Description of Particular Event or Circumstance:
B.
The inventory of the outside rad waste tank farm was 13.0 Ci when Technical the contents were analyzed at 8:30 a.m. on August 18.
Specification paragraph 3.6.C limits inventory to 10.0 Ci and directs The that the contents be recycled if the inventory exceeds 5.0 C1.
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cause for the excessive inventory was attributed to an operator error that permitted the overflowing of the waste concentrator tank and thance via the rad waste f1Loor sump to the subject storage tanks.
C.
Action by Licensee:
Recycling of the tank farm contents had reduced the inventory to 10.5 Ci at the time of the call. The licensee estimated that the inventory would be reduced below 10.0 Ci within an additional four hours and
below 5.0 Ci within 24 te 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The operation was being per-formed around the clock. Fome of the excess inventory was to be trucked off site by the licensee's rud waste contractor, Nuclear Engineering Corporation. The Plant Operating Review Cormaittee will review this violation. The licensee will submit a 10-day written report of this occurrence to Licensing.
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J. G. Kappler, Chief, Reactor Testing & Operations Branch i
Directorate of Regulatory Operations, HQ.
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RO INQUIRY REPORT No. 50-219/72-26 l
JERSET CENTRAL POWER & LIGHT CCHPANY j
UYSTER CREEK - BWR EQUIPMENI FAILURE - CONTROLS FOR CONTROL ROD DRIVES
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The subject inquiry report is forwarded for your information.
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'The licenses has researched this problem and considers it unusual in nature, but not a particularly serious one in that the controi
rods could still be scrammed. His proposed action in case of a
future switch failure with the reactor at power, i.e., to replace
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the defective switch without a abutdoen, is acceptable to us on
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i the basis that in any case it would be necessary to jum'per the
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failed switch in order to effect a shutdown by normal means.
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Reactivity changes can still be effected by recirculation pianp
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flow.
1" c4 We are of the view that this failure should be reported in writing and are encouraging the licensea to do so.
As a
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minimum, this matter will be reviewed during the next inspection.
j R. T. Carlson, Chief
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Reactor Operations Branch
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Enclosure:
Subject Inquiry Report cc:
R. Minogue, RS (3)
,
R. S. Boyd, L (2)
R. C. DeYoung, L (2)
i D. J. Skovholt, L (3)
j H. R. Denton, L (2)
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P. A. Morris, RO H. D. Thornburg, R0
R. H. Engelken, RO (
,
RO Files-DR Central Files LJ
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RO Inquiry Report No. 50-219/72-26 i
Ettject: Jersey Central ?ower & Limht Company i
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Facility: Oyster Creek - BWR
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License No.: DPR-16
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Prepared by:
F. S. Centre 11, Reactor Inspector Date
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Date and Manner AEC was Informed:
August 15, 1972, by telephone call from Mr. T. J. McCluskay,
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Station Superintendent.
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l B.
Description of Particular Event or Ciretsastance:
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With the reactor at low power and while increasing temperature, the selector switch for control rod drive 14-11 failed in such a manner i
that the drive control was locked in that control rod. Under this condition (open circuit), it was not possible to select any other
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control rod for, insertion or removal. All control rods could still be
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screened; however, because the peactor pressure was low at this particular thne, a decision was made to jtsaper the switch such A
that other rods could be selected for operation and to shut down in a normal fashion.
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C.
Action by Licensee:
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After the plant was in a cold shutdown condition, the selector switch was replaced.
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2.
The circumstances relating to the problem were subsequently reviewdd by the Plant Operations Review Consnittee which concluded that the proper course of action had been taken.
3.
The failure was discussed with the Geveral Electric Company, which
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stated that this particular problea had been reviewed with Licensing
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in the original application. JCP&L concluded that if a failurc i
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occurred with the reactor operating at power, the correct action j
would be to jtzaper and replace the selector switch while at power
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rather than to shut down for switch replacement.
4.
A procedure will be generated to cover any future replaceraents
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of this switch.
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The licensee is considering submitting an informational report 5-
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to the Cocraission concerning this failure.
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