IR 05000289/1976009
| ML19276H185 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/10/1976 |
| From: | Hurd R, Kellogg P, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19276H183 | List: |
| References | |
| 50-289-76-09, 50-289-76-9, NUDOCS 7910150404 | |
| Download: ML19276H185 (22) | |
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IE:I Form 12
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n 75) (Rev)
NOTICE g,MgAy 14 1976 REGION 1 HAS ff0T CBTAIMED PRCPdlGARY
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CLEARANCE IN ACNtCE bahn W us W U. S. NUCLEAR RECUL\\ TORY CC:CliSSIO:i 0FFICE OF INSPECTIO!! AND EliFORCD!E!T
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REGION I
IE Inspection Report No:
50-289/76-09 Docket No:
50-289 Licensee:
Metr Politan Edison Company DPR-50 e m No*
P.O. Box 542 Priority:
Reading, Pennsylvania 19603 C
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Safeguards P-Three Mile Island 1, Middletown, Pa.
Location:
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Tvoc of Licensee:
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PWR, 2535 MWe, B&W Type of Inspeqtion:
Routine, Unannounced Dates of Inspection:
4/6-9, 12-14, 21/76 Dates of Previous Inspection:
3/23-25/76 Reporting Inspector:
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". O. Hurd, Reactor Inspector
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Accompanying Inspectors:
DATE DATE DATE Other Acccmpanyin;; Personnel:
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P.J.Kello/g,ReactorInspector DATE
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_ icved By:
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E. C. McCabe, Section Chief, Nuclear Support Section
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No. 1, RO&NS Branch
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SUMMARY OF FINDINGS Enforcement Action Items of Noncompliance
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Duriciency Contrary to Technical Specification 6.8.1 (formerly 6.2.3) which re-quires written procedures to be inplemented, the data for Quadrant Tilt and Core Imbalance were not recorded between 1600 and 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> on February 14, 1976 in accordance with SP 1301-1.
(Details, 5.b)
Licensee Action on Previously Identified Enforcement Items A.
Correct!.ve actions taken in response to Region I Inspection Report 50-289/75-25 were completed.
(Details, 24.a)
B.
Corrective actions taken in response to Regiin I Inspection Report 50-289/75-14 were completed.
(Details, 24.b)
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Other Significant Findings A.
Current Findings 1.
Acceptable Areas I
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These are. areas which were inspected on a sampling basis and findings did not involve an Item of Noncompliance, Deviation, or an Unresolved Item.
a.
Reactor Protection System Surveillance.
(Details, 2)
b.
High and Low Pressure Injection Systen Analog Channels.
(Details, 3)
Reactor Building Spray Actuation Setpoints.
(Details, 4)
c.
d.
Shift and Daily Checks, Rod Misaligned.
(Details, 5.a)
e.
Precritical Checks.
(Details, 6)
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Raactor Building Pressure Setpoints.
(Details, 8)
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Pressurizer Heatup and Cooldown Rates.
(Details, 9)
h.
Reactor Coolant Oxygen Limit.
(Details, 11)
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Reactar Coc'. ant Leakage Evaluation.
(Details, 12)
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Maintenr:.ce on ECCS and E=ergency Power Systems.
(De-tails, 13)
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Mav.imum Rod Worth Test.
(Details, 14)
1.
Ra,airements for Containment Integrity.
(Details, 15)
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Containment Isolation Valves.
(Details, 16)
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Maintenance Activities.
(Details, 18)
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Fuel Inspection.
(Details, 19)
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Fire Drill.
(Details, 20)
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CO Discharge System Test.
(Details, 21)
2.
Unresolved Items I
These are items for which more information is required in order to determine whether the items are acceptable or Items of Noncompliance.
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OP 1102-2, Plant Startup.
(Details, 7)
b.
Emergency Diesel Generator Monthly Test.
(Details, 10)
c.
Corporate Technical Support Staff Review.
(Details, 17)
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d.
SP 1302-5.4, RC Flux Flow Calibration.
(Details, 23)
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TS 3.3.11a, BWST Manual Discharge Valve.
(Details, 25)
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Status of Previous Unresolved Items The following items are resolved.
1.
Nuclear Ovarpower Trip Setting.
(Details, 22.a)
2.
Operability of Safety Instrumentation.
(Decails, 22.b)
3.
Review of 50.59 Evaluation by GOR 3.
(Details, 22.c)
4.
FSAR Amendment, Post Accident.
(Details, 22.d)
5.
Corporate Technical Support Staff (CTSS).
(Details, 22.e)
6.
NDE Test Personnel Qualificiation Records.
(Details, 22.f)
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7.
Test, Vents and Drains.
(Datails, 22.g)
8.
Electrical Penetrations.
(Details, 22.h)
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9.
Flow Meter Orientation.
(Details, 22.1)
10.
Flow Meter Bypass.
(Details, 22.j)
11.
Pipes.
(Details, 22.k)
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12.
Surveillance Procedure 1303-5.1 Reactor Building Cooling and Isolation System Logic Channel and Component Test.
(Details,
22.1)
13.
Flux-Reactor Coolant Flow Comparator Channels.
(Details,
22.n)
14.
Support Bolts on Seismic Category I Components.
(Details, 22.m)
Management Interview An exit interview was held onsite on April 14, 1976 at the conclusion of that portion of the inspection.
Further discussions were held with the Unit ?operintendent cohcerning the inspection on April 21, 1976.
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Personnel Attending Mr. J. Colitz, Unit Superintendent Mr. C. Hartman, Electrical Engineer Mr. W. Potts, Supervisor, Quality Control The following items were discussed:
A.
OP 1102-2, Pl' ant-Startup.
(Details, 7)
B.
Emergency Diesel Generator Monthly Test.
(Details, 10)
C.
Corporate Technical Support Staff Review.
(Details, 17)
D.
SP 1302-5.4, RC Flux Flow Calibration.
(Details, 23)
E.
Shift and Daily Checks.
(Details, 5)
F.
The inspector briefly reviewed the other items covered in this inspection which are "ound acceptable, included in this review were the previously unresolved items which were resolved.
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DETAILS 1.
Persons Contacted
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Discussions were held with the following persons during the conduct of the inspection activities documented in this report.
Mr. M. Benson, Engineer, Nuclear Mr. K. P. Bryan, Shift Foreman Mr. J. J. Chwastyk, Shift Supervisor
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Mr. J. J. Colitz, Unit No. 1 Superintendent Mr. W. E. Cotter, Lead Mechanical Engineer Mr. F. H. Grice, Safety Director Mr. C. Hartman, Lead Electrical Engineer Mr. J. Hilbish, Nuclear Engineer Mr. G. A. Kunder, Supervisor Station Operations Mr. J. P. O'Hanlon, Engineer, Senior I
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Mr. V. Orlandi, Lead I&C Engineet Mr. R. H. Porter, Shift Supervisor Mr. W. Potts, QC Engineer it'S Mr. M. J. Ross, Shift Supervisor
!'l Mr. M. A. Shatto, Engineer, Associate I Mr. E. M. Sheets, Electrical Engineer Mr. D. M. Shovlin, Maintenance Supervisor Mr. M. G. Snyder, Instrument Foreman
Mr. G. E. Stambaugh, Maintenance Foreman 2.
Reactor Protection System Surveillance The inspector reviewed SP 1303-4.1, Rev. 17, Reactor Pressure System Surveillance Procedure, which is performed every four weeks on each RPS channel and prior to each startup.. The inspector also reviewed the data sheets from this procedure from the RPS surveil-lance procedures performed during January, February and March of 1976.
The purpose of this review was to determine that the follow-ing Technical Specification requirements were satisfied.
a.
T.S. Table 2.3-1 Item 3, Reactor Trip at less than 91% of rated power with only one Reactor Coolant Pump in each loop operating, Reactor Trip on loss of both RCP's in one loop and loss of one.or two RCP's during two pump operations.
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T.S. Table 2.3-1, Item 7, Reactor Trip equal to or less than a reactor coolant maximum temperature'of 5190F.
c.
T.S. 3.5.1.6, verified that, during all tests, the trip de-vices did not fail in the untripped state.
The inspector found no inadequacies in the areas itemized above.
3.
High and Low Pressure Iniection System Analog Channels The inspector reviewed SP 1303-4.11, Rev. 7, High and Low Pressure Injection System Analog Channels which is performed monthly to
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satisfy the requirements of T.S. Table 4.1-1 Item 15 to test monthly the HPI analog channels.
The inspector reviewed the data sheets for this procedure performed January 8, 1976 and January 30, 1976.
No testing was required in February or March of 1976 due to the refueling outage in progress during that period of time.
The setpoints for the HPI actuation are set in a range of 1540 to 1550 psig and 554 to 550 psig.
The data recorded indicated that the
setpoints are required to be greater than 1500 psig and 500 psig by fy T.S. 3.5.3.1.
The inspector found no inadequacies in this area.
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4, Reactor Building Spray Actuation Setpoint The inspector reviewed SP 1302-5.11, Reactor Building 30 psig Pressure Channel, and the data for this test performed 12/18/75.
The inspector noted that the procedure verifies or adjusts the six pressure switches in the spray actuation channels to trip at be-tween 28.0 and 28.9 psig and that, for the test data, all switches were. set to be within that range.
The setpoints are required to be less than or equal to 30 psig by T.S. 3.5.3.1.
The inspector found no inadequacies in his review of this area.
5.
Shift and Daily Checks The inspector reviewed the data sheets for SP 1301-1, Shift and Daily Checks performed in January, February and March of 1976 to verify the following Technical Specification requirements were satisfied.
T.S. 4.7.1.2, Control rod misaligned with its group average by a.
more than sh indicated nine inches.
b.
T.S. Table 3.5-1, Item 7, at leasttwooffourfgfigbapaydj/
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flow Channels are operable when reactor critical
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T.S. Table 3.5-1, ESF Item 1.a. at least two HPI Low Reactor Coolant Pressure Channels operable when reactor critical, d.
T.S. Table 3.5-1, ESF Item 4.a, at least two Reactor Building 30 psig Instrument Channels are operable when the reactor is critical.
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T.S. 3.5.2.4.d, Quadrant Tilt monitored at least once every e.
two hours durins power operation above 15 percent.
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T.S. 3.5.2.5.a. Operating rod group overlap shall not exceed 25 vercent + 5 percent between the two sequential groups except for physics test.
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T.S. 3.6.4, Reactor is not critical when reactor building in-ternal pressure exceeds 2.0 psig or 1.0 psi vacuum.
.i Based on review of the above items, the inspector had the following findings.
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(1) Rod Misaligned On January 15,1976 at 2026 Rod 3 in group 1 dropped from the fully withdrawn position (the group average was 100 percent) to the fully inserted position.
The inspector reviewed the Shift Foreman's log to determine if the
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actions taken were in accordance with the requirements of Technical Specifications 3.5.2.2 and 4.7.1.2.
The core i
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shutdown margin was determined to be 1.45% ak/k thus
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satisfying the requirement.of T.S. 3.5.2.2.c that the shutdown margin be greater than one percent.
The reactor power was ic=ediately decreased to 45% (of rated) thus satisfying the requirement of T.S. 3.5.2.2.d to reduce power below 60% of the thermal power allowable.
Subscquently, at 0555 on January 16, 1976 the plant was shutdown to replace the failed CRD stator on the dropped rod.
On return to power all rods were operable.
The inspector found no inadequacies in this area.
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(2) Monitoring Requirements for Ouadrant Tilt Technical Specification 3.5.2.4.d requires that the quandrant tilt be monitored at least every two hours during operation above 15 percent power.
The inspector's review of SP 1301-1, Shift and Daily Checks noted that the procedures require recording of the values of the four power range channels and the calculation of the
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Quadrant Power Tilt per the definition given in T.S.
1.6.1 and recording of that value.
The data sheet for SP 1301-1 completed on February 14, 1976 indicated that, with the reactor at 100 percent
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power, the readings of the four power range channels, the core Quadrant Tilt, and the Core Lebalance were not re-
corded a,e 1800, 2000, and 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br /> as required by SP 1301-1.
The id:ensee provided the inspector with the hourly computre loge for that period which indicated that the Quadrant lower Tilt and Core Imbalance were within
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the Technical Specification required values.
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discussio.'s with the licensee and review of other hourly computer logs to determine the adequacies of these logs the inspector determined that the plant computer logs satisfied the technical specification require =ents.
The bases for Technical Specification 3.5.2 states, "the plant computer will scan for tilt and imbalance and will
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part, " Written procedures and administrative policies
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shall be established, implemented..." the data for Quadrant Tilt and Core imbalance on February 14, 1976 between 1600 and 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> was not recorded in accord-ance with SP 1301-1.
This is a deficiency.
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The inspector found no inadequacies in the other areas itemized above.
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Precritical Checks
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The inspector reviewed the " Precritical Checklist" which is enclos-ure No. 1 of OP-1102-2, Plant Startup.
The " Precritical Checklist"
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completed prior to startup of 1/18/76 was properly signed off to indicate that the following Technical Specification requirements were satisfied prior to the reactor being made critical.
a.
T.S. 3.3.1.1.a, BWST contains a minimum of 350,000 gallons of
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water with a minimum concentration of 2,270 ppm boron and a l
temperature not less than 45 F.
b.
T.S. 3.3.1.1.b, Two makeup pumps are operable in the safe-t guards mode and powered from independent safeguard buses.
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c.
T.S. 3.3.1.2.c, The motor operated discharge valve for each
core flood tank is open and their breakers opened and tagged.
d.
T.S. 3.3.1.3.a. The Reactor Building spray pumps and their associated spray nozzle headers are operable.
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T.S. 3.3.1.3.b, The thiosulfate tank contained between 34,000 e.
g and 36,000 pounds of sodium thiosulfate and the sodium hydroxide tank contained between 16,000 and 17,000 pounds of sodium hydroxide.
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T.S. 3.3.1.4.a, b and d, Two nuclear service closed cycle
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cooling water pumps, two nuclear service river water pumps and
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two decay heat river water pumps are operable.
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T.S. 3.3.1.5, Engineered Safeguard valves and interlocks associated with the systems in specifications 3.3.1.1-3.3.1.4 are operable.
h.
T.S. 3.3.1.1.c, Two decay heat removal pumps are operable.
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T.S. Tabic 3.5-1, item 7, At least two of the four flux /_mbal-ance/ flow channels are operable.
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T.S. Table 3.5-1, ESF item 1.a, At least two of the reactor coolant pressure channels are operable.
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T.S. Table 3.5-1, ESF item 4.a, At least two of the reactor building 30 psig Instrument Channels are operable.
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T.S. 3.6.4, Reactor building pressure does not exceed 2.0 psig or 1.0 psi vacuum.
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T.S. 3.7.1.a. All engineered safeguards buses, engineered safeguards switchgear, and engineered safeguards load shedd-ing systems are operable.
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T.S. 3.7.1.e, Engineered safeguards diesel generators are operable and at least 25,000 gallons of fuel oil are available in the storage tank.
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The inspector found no inadequacies in the iters listed above.
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OP-1102-2, Plant Startuo
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During review of OP 1102-2, Plant Startup, the inspector noted that, on page 39, Enclosure No.1, procedure SP 1303-4.12 was listed.
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j The correct listing is SP 1303-4.11, High and Low Pressure Injection Analog Channels.
The licensee stated that SP 1303-4.12 has been combined with SP 1303-4.11 and that OP 1102e2 would be corrected.
This item is unresolved pending issuance of a corrected OP 1102-2.
8.
Reacter Building Pressure Setpoints
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Technical Specification 3.5.3.1 requires that the ~ Reactor Building Pressure setpoint actuation of reactor building cooling and isola-tion and actuation of high pressure injection system is less than or equal to 4 psig.
The inspector reviewed SP 1303-4.13, Sections 6.1.7, 6.2.7 and 6.3.7 which require the checking and/or adjusting of the Reactor Building Pressure bistables to 2.5 to 2.8 psig.
The inspector reviewed the completed SP 1303-4.13 performed February 3,1976 and verified that the results indicated the setpoints were within the range specified in the procedure.
The inspector found no inadequacies in this area.
9.
Pressurizer Heating and Cooldown Rates Technical Specification 3.1.2.3 lLsits the pressurizer heatup and cooldown rates to 1000F/hr.
The inspector examined the com-pleted trend recorder traces of RC Loop temperatures and pressures recorded during plant cooldowns on 1/16/76 and 2/21/76 and the plant heatup on 1/17/76.
The inspector determined that maximum temperature change rate for the pressurizer during these events i
was 76 F/hr.
The inspector found no inadequacies in this area.
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Emergency Diesel Generators Monthly Test Technical Specification 4.6.1.a requires a monthly test on each Diesel Generator which includes manually initiated start, manual synchronization with the other power sources and loading to 3000 Kw.
Surveillance Procedure 1303-4.16, Emergency Power Systems is designed to satisfy the Technical Specification requirement for the monthly test.
The inspector reviewed the SP 1303-4.16 data sheets for the tests performed during January, February and March.
During this period, a total of 21 tests were performed on the "A" Diesel-Generator and 19 tests were performed on the "B" Diesel-Generator.
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Since T.S. 4.6.1.a required only 3 to-'s during this time, the
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majority of the testing was performe6 demonstrate operability
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prior to and following maintenance on.he Diesel-Generators.
The inspector noted that the data for this test required recording of the individual cylinder exhaust temperature and the maxi =um differ-ential temperature between cylinders.
The data sheet listed a normal range for exhaust temperature on each cylinder and a tem-perature limit.
The data sheet also listed a maximum differential temperature limit of 2500F. The inspector expressed concern that
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on most of the data sheets the cylinder exhaust temperatures were k.)
outside the nor=al range listed on the data sheet and on 22 of the 40 data sheets, che enterec. value for differential temperature
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exceeded the 2500F limit.
The maximum differential temperature recorded van 3000F for D-G "B" on March 1, 1976.
The licensee stated that the diesel vendor had been consulted on the accept-ability of this data and the vendor stated that it was acceptable.
The licensee is presently requesting that the vendor determine more realistic temperature limits.
This item is unresolved until the data sheet for SP 1303-4.16 is modified to reflect the actual temperature limits.
11.
Reactor Coolant System Oxygen Limits Technical Specification 3.1.5.1 requires an oxygen concentration limit in the pri=ary coolant of 0.1 ppm during power operation.
The inspector reviewed the data of SP 1301-3, Reactor Coolant Sample Procedure which samples and determines the oxygen concen-tration in the RCS five times weekly.
The data for SP 1301-3 indicated that for the period of January through Feburary 21, 1976 (date when refueling outage coc=enced) the maximum Oxygen Concer.cracion was 0.075 ppo during power operation.
The inspector found no inadequacies in this area.
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Reactor Coolant Leakage Evaluations The inspector revieued the Shift Foreman's log from January 1 to February 21,1976 (date refueling outage commenced) and the data sheets for SP 1301-1, RC System Leak Rate, for the same period.
The review indicated that the Reactor Coolant did not exceed any of the limitations of Technical Specification 3.1.6 during this period of time. The inspector found no inadequacies in this area.
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13. Maintenance on ECCS and Emergency Power Systems i
The inspector reviewed the Work Request documents for the seven maintenance items performed during February, 1976 on emergency
electrical system components.
The Work Request forms were properly signed and approved by the Supervisor of Maintenance and the
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l Quality Control supervisor. The inspector also reviewed two Work Requests for maintenance performed on ECCS components during power operation in January and February. The inspector found no inade-quacies in this area.
14.
Maximum Rod Worth Test
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The inspector reviewed Enclosure 5 of SOP 199 dated 5/17/75,
" Ejected Control Rod Worth Measurements" this test is performed to verify compliance with T.S. 3.5.2.3 which states in part "the worth
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of a single inserted control rod shall not exceed 0.63 percent ak/k at rated power or 1.0 percent ak/k at hot zero power." The data
l from SOP 199 performed 6/22/75 indicated the worth of control rod H-14 was 0.58 percent ak/k at zero power hot condition. The licen-
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see stated that H-14 had been determined to be the strongest rod
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from computer code analysis and that the rod would be worth less at rated power.
The licensee stated the evaluation and measurement was valid for the period of January through March of 1976 since the only core alteration was defueling the core which commenced on March 3, 1976 and was completed on March 10, 1976.
The inspector had no further questions in this area.
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Requirements for Containment Integrity Technical Specification 3.6.2 states, " Containment integrity shall be maintained when both the reactor coolant system is open to the containment atmosphere and a shutdown margin exists that is less than that for a refueling shutdown."
The inspector determined by discussions with the licensee and re-view of the Shif t Foreman's log that the reactor coolant system was opened to the atmosphere on 2/26/76 and remained open through the inspection period.
The Shif t Foreman's log of daily boron con-centration determinations indicate that the concentration was main-l tained greater than 1900 ppm during this period.
A boran concentra-
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tion of 1800 ppm provides the refueling shutdown margin (Refer to T.S. bases for Section 3.8), therefore containment integrity was not required.
The inspector had no further questions in this area.
16.
Containment Isolation Valves
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Through discussions with the licensee and review of the Shift Fore-man's Log and the data sheets for SP 1303-5.1, " Reactor Building (3 Cooling and System Logic Channel and Component Test," the inspector
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determined that, during the period of January through March, no reactor building icolation valves were inoperable when the reactor was critical.
Therefore, the actions described in Technical
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Specification 3.6.6 were not required to be performed.
The inspector found no inadequacies in this area.
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17.
Corporate Technical Support Staff Review Technical Specification 6.5.2.A.2.a requires that the Corporate
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Technical Support Staff (TSS) review evaluations of proposed changes to procedures, equipment or systems to verify that an unreviewed safety question was not involved.
The licensee identified 133 procedure changes and 7 change / modifications made to equipment described in the FSAR which were Unplemented during January and February of 1976.
The inspector determined by review of memorandums from the CTSS which were approved by the Manager, Generation Engineering, to the Maintenance Supervisor that all of the change /
modifications were reviewed by the CTSS.
The inspector determined by review of memorandums from Manager, Generation Engineering, to the Unit No. 1 Superintendent that 98 of the 133 procedure changes have been reviewed by the CTSS.
This item is unresolved pending documentation of CTSS review of the remainlag 36 PCR's.
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Maintenance Activities The inspector reviewed the procedures used to control the mainten-ance activitics listed below which were performed during the re-fueling outage:
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a.
Core Support Assembly Repair During remova'l of a surveillance capsule from its holder tube, damage to the holder tube was observed.
Subsequently, pro-cedures were written and approved to remove major portions of the surveillance holder tube and modify the remaing portions of the holder tube.
The inspector reviewed the following procedures used to control these activities:
(1)
SOP No. 348, 3/27/76, Survefilance Specimen Holder Tube Controlling Procedure.
(2)
SOP No. 351, 3/28/76, Surveillance Tube Removal Tools.
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(3)
SOP No. 365, 3/29/76, Radiation Control Procedures for Surveillance Holder Tube Ra=cval.
(4)
SOP No. 349, 3/27/76, Initial Inspection Procedure.
(5)
SOP No. 350, 3/27/76, Installation of Work Platforms on Core Support Assembly.
(6)
SOP No. 358, 3/28/76, Reactor Vessel and Fuel Transfer Canal Loose Part Debris Recovery.
(7)
SOP No. 361, 3/28/76, Shroud Tube Ram Operation.
(8)
SOP No. 362, 3/28/76, Internal Tube Puller.
(9)
SOP No. 359, 3/27/76, Speciment Holder Tube Rotating Tool Procedures.
(10)
SOP No. 360, 3/20/76, Use of Boroscope for Inspection Operations.
(11)
SOP No. 379, 4/2/76, Diver Controlled Oxy-Arc Bearing Operations for Removal of Pintle Pin Tool.
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(12)
SOP No. 381, 4/3/76, Core Support Assembly Inspection i
and Cleaning.
The inspector found no inadequacies in this area.
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b.
Makeup and Purification Pump (HU-P-1A) Repair The inspector reviewed Work Request No.14131.
This work request provided the instructions and controlled the repair of MU-P-1A.
The procedure included cleaning, flushing and hydro-static testing of the pump after re-installation.
The in-spector found n6 inadequacies in this area.
c.
Control Rod Drive Stator Replacement The inspector reviewed Work Request No. 13475.
This approved document provided the instructions and controlled the repl'ce-ment of control rod drive stators and the change of the cab connectors.
The work was divided into the following ti.ree parts:
(1)
CRDM stator removal and replacement per ?rocedure MP 1401-5.1, Rev. 3.
(2)
Connector Replacement, all connectors were replaced by temperature resistant silicone connectors.
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(3)
Retent of CRDM's to verify latching capability, hold capability and measurement of dropout current.
The inspector found no inadequacies in this area.
19.
Fuel Inspection The inspector reviewed SOP No. 328 dated 3/19/76.
This procedure defines the fuel inspection steps for visual inspection of fuel which is returned to the reactor core for fuel cycle II.
The inspection included Television scanning and recording of each fuel assembly and completion of Data Sheet "IMI-1, Fuel Visual Inspec-tion Record" for each fuel assembly.
The new fuel to be placed in the reactor core for fuel cycle II was inspected per the require-ments of Procedure 1503-1, " Receipt, Inspection, Fit-up and Storage of New Fuel Control Ccmponents." The inspector verified on a sampling bases that the appropriate inspection data sheets were available for each fuel assembly to be placed in the reactor core for fuel cycle II.
The inspector noted that 19 fuel assemblies had indicatiens of damage to the fuel spacer grids.
This damage consisted of slight chipping of * the outer spacer grid material.
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Based on review of the inspection data sheets, licensee memorandums and through discussions with the licensee the inspector had.the i
following findings:
a.
All damaged assemblies are acceptable for further operation based on the fuel vendors inspection and safety evaluations.
b.
All damage was sustained during fuel removal from the core based on the lack of formation of oxides in damaged areas.
c.
Additional precautions and limitations have been added to the fuel movement procedures to preclude this type of damage in the future.
l The inspector had no further questions in this area.
i 20.
Fire Drill The inspector witnessed a licensee conducted fire drill.
The inspector observed the response to a simulated oil fire at Reactor Coolant Pump "A" in the reactor building.
The response was in ( ])
accordance with EP-1202-31, Fire Emergency Procedure.
The licensee stated that such drills are held monthly and that a report concerning each drill is made by the Three Mile Island Safety
,
Director to the Metropolitan Edison Manager of Generation.
The
'
inspector found no inadequacies in this area.
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21.
CO2 Discharge System Test During the period of this inspection, the licensee conducted the initial test of a CO2 Discharge System in the cable spreading room.
The system provides additional fire protection in this area.
The inspector reviewed the data concerning the CO2 con-centration during the test.
The licensee stated that the test satisfied the design criteria and the system would be armed for operation when safety training of personnel was completed.
The inspector had no further questions in this area.
22.
Previously Unresolved Items a.
Nuclear Overpower Trio Settine Reference:
Report 50-289/75-25, Detail,12.b(1) and 12.b(3)
,
The inspector reviewed the required changes to data sheets 1 and 2 of'SP 1302-1.1.
The changes are documented by TCN 76-13 and the approved PCR 76-030.
These items are resolved.
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Operability of Safety Instrumentation Reference: Report 50-289/75-25, Detail 14 The inspector reviewed OP 1102-2, Enclosure No.1, " Precritical Checklist" page 39, Rev. 12, dated 2/5/76.
Item 22A of this checklist contained RPS surveillance procedure.
This item is resolved.
, c.
Review of 50.59 Evaluation by GORB i References: (1) Report 50-289/75-14, Detail 7.c (2) Report 50-289/76-01, Detail 7.1 The inspector reviewed AP 1016, Rev. 9, and verified the re-quired changes were incorporated.
This item is resolved.
, I d.
FSAR Amendment, Post Accident Purge System 'Referecces: (1) Report 50-289/74-18, Detail 3.b(1) '5) (}) (2) Report 50-289/75-04, Detail 6.c - (3) Report 50-289/76-01, Detail 4.c The inspector reviewed FSAR page 6-2a Rev. 1 (2-18-76), Supplement 1, Part II concerning the Post Accident Purge System.
This item is resolved.
, j e.
Corporate Technical Suncort Staff (CTSS) ] Reference: Report 50-289/76-01, Detail 14.b(2) The inspector reviewed the Routing Sheet for PORC Meeting #312 which indicated the CTSS had the reviewed the minutes of this PORC meeting which contained the review of two Items of Non-compliance.
This item is resolved.
f.
NDE Test Personnel Qualification Records Reference: Report 50-289/76-05, retail 6 The inspector reviewed the visual examination reports for specific temporary employees and verified that the modified reports indicated requirements of SNT-TC-1A had been meet.
This item is resolved.
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Test, Vents and Draining Reference: Report 50-289/76-06, Detail 8.a The inspector reviewed the approved TCN 76-125 which adds Enclosure 14 to SP 1303-11.18.
This enclosure provides verification that valves and caps are closed and tagged.
This verification is performed at the frequency of local leak rate testing.
This item is resolved.
h.
Electrical Penetrations ' Reference: Report 50-289/76-06, Detail 8.b FSAR Section 5.7.4.2 states in part, "All electrical pene-trations have at least one epoxy insulator structurally bonded between wire and steel, and full penetration, steel- - to-steel welds as the barrier between the containment atmos-phere and environment." The licensee stated that based on the above, leak tests are not required.
This item is resolved.
( ) 1.
Flow Meter Orientation , Reference: Report 50-289/76-06, Detail 8.c
The inspector reviewed approved TCN 76-125 which has added . to Enclosure 3 of SP 1303-11.18 a requirement that the rota-meter is vertical and checked with a plumb bob.
This item is resolved.
j.
Flow Meter Bypass Reference: Report 50-289/76-06, Detail 8.d The inspector reviewed approved TCN 76-125 which has added to data sheet 1303-11.18D a requirement that the bypass valve be tested weekly to verify that no leakage exists.
This item is resolved.
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Pipes Reference: Report 50-289/76-06, Detail 8.e The licensee stated that the four pipes observed by the inspector were installed for insertion of temperature probes to measure the penetration shell temperature during the initial startup program.
These pipes do not penetrate the containment wall.
This item is resolved.
- 1.
Surveillance Procedure 1303-5.1, Reactor Building Cooling and Isolation System Loeic Channel nnd Component Test Reference: Report 50-289/76-06, Detail 6 , . The inspector determined by review of the appropriate drawings and discussions with the licensee that the Fluid Block Control Valves and Purge Inner-Space Supply Valves do not open during ESAS and therefore verification of position is not required.
The inspector also reviewed page 49 of SP 1303-5.1 and verified the position of the Penetration Isolation Valves (PPV-151 and ( 152) and Electrical Penetration Post-Accident Pressurization - Valves (PPV-147 and 165) are verified after ESAS.
This item is resolved.
{ m.
Support Bolts on Seismic Category I Comoonents References: (1) Report 50-289/74-33, Detail II.3 (2) Report 50-289/75-24, Detail 11
i The inspector reviewed memorandum GEM 0155, dated January 21, 1976, subject, AI 75-754 and 75-766, Engineering Evaluation of Anchor Bolt Surveillance Program, SP 1300-2.
The memorandum was approved by the Manager, Generation Engineering and indicated that SO 1300-2 was capable of providing assurance of continuing integrity of anchor bolts.
This item is resolved.
n.
Flux-Reactor Coolant Flow Coenarator Channels Reference: Report 50-289/76-03, Detail 4.a The licensee stated that SP 1303-4.1, Jection 6.5.9 meets all Technical Specification requirements concerning channel tests and that further expanding the test points would be redundant with SP 1302-5.4, "RC Flux Flow Calibration." This item is resolved.
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SP 1302-5.4 RC Flux Flow Calibration Based on the above item, the inspector reviewed the data sheets for SP 1302-5.4 performed 5/27/75 and TCN 75-74 which was in affect at that time.
The procadure perferms and documents through use of data sheets the calibration verification of the square root conver- ' ters in the RC flow channels.
The square root converters are checked by inputting a signal from the Flow Converter Test Module ! and measuring the square root converter output and comparing against an acceptance criteria.
The inspector noted that the pro-cedure did not document calibration of the Flow Converter Test ! Module.
The licensee stated that SP 1302-5.4 will be revised to provide this documentation.
This item is unresolved pending review of the revised procedure.
' 24.
Corrective Action for Previously Identified Items of Noncompliance a.
Region I Inspection Report 50-289/75-25, Region I letter dated - December 24, 1975 and Licensee's response dated January 19, 1976.
}' ) The inspector reviewed changes to SP 130314.1 and SP 1302-1.1 i which were detailed in the licensee's response and found no inadequacies.
- This satisfactorily completes the licensee's corrective action
for this Item of Noncompliance.
b.
Region I Inspection Report 50-289/75-14, Region I letter dated August 19, 1975 and Licensee's response dated September 15, ' 1975.
(Apparent Deficiency E) Reference: Report 50-289/76-01, Detail 5.a(3) ' The inspector reviewed GP 0013, Rev. I dated 2/23/76 and GP 4010, Approved Vendors List dated 2/17/76.
The Vendors List had been modified to delete sales offices and establish a separate information sheet for each approved vendor.
This satisfactorily completes the licensee corrective action for this Item of Noncompliance.
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T.S. 3.3.1.1.a BWST Manual Discharge Valve During review of the requirements for startup (Detail 6), the inspector noted that T.S. 3.3.1.1.a contained a requiremene that a manual discharge valve from the BWST be locked open.
During inspection 50-289/75-01 conducted January 6-9 and 17, 1975 it was determined that no manual valve existed on the discharge of i the BWST, (Refer to Report 50-289/75-01, Detail 5.h).
At that time the licensee stated that the Technical Specification 3.3.1.1.a - would be corrected by licensee amendment.
This item is unresolved pending correction to T.S. 3.3.1.1.a.
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