IR 05000289/1975025
| ML20125B072 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/18/1975 |
| From: | Hurd R, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20125B061 | List: |
| References | |
| 50-289-75-25, NUDOCS 7910190526 | |
| Download: ML20125B072 (20) | |
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o U. S. NUCLEAR REGULATORY COMMISSION
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OFFICE OF INSPECTION AND ENFORCE!ENT
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REGION I
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Inspection Report No:
50-289/75-25 Docket No:
50-299
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censse:
Metropolitan Edison Company
' License No:
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P.O. Box 542 Priority:
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Reading, Pennsylvania 19603 Ca tegory:
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Safeguards Group:
Middletown, Penn'sylvania(Three Mile Island)
cation:
pc of Licensee:
.. iof Inspection:
Routine, Unannounced November 10-19, 1975
$::
t Inspec tion:
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tOofrreviousIusPection:
neve ser 3. 1975
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porting Inspector:
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DATE R. O. Hurd, Reactor Inspector companying Inspectors:
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DATE
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DATE DATE har Accompanying Personnel:
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E. C.,McCabe, Section Leader
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vieved By:
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M, IE IO TAT ~r E. C. McCabe, Nuclear Support Section Leader g
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React,or Operations and Nuclear Support Branch
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V SUMMARY OF FINDINGS
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Enforcement Action
A.
Items of Noncompliance Deficiency E-Contrary to Technical Specification 6.2.3 requirements for ad-
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herence to procedures:
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1.
SP 1303-4.1 requirements for data recording were not met on
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July 22, 1975 and Se.otember 24, 1975.
(Detail 12.a); and
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2.
SP 1202-1.1 admin 1A ative requirements for power range
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amplifier resetting..are not met on August 7, 1975.
(Detail 12.b. (2.))
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(Limiting safety system settings were not exceeded by the above.)
B.-
Deviations None
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Licensee Action on Previously Identified Enforcement Items
Report 50-289/75-14 The licensee's corrective actions with respect to Items of Noncompliance designated as Details 3.a, 3.b, 3.c, 4.c and 5.b of the above referenced report were reviewed with respect to the licensee's response letter to Region I dated September 15, 1975.
The inspector had no further ques-tions on these items.
(Detail 17)
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Other Significant Findings
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Current Findings l
1.
Acceptable Areas The following items were inspected on a sampling basis and
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findings did not involve an Item of Noncompliance, Deviation or an Unresolved Ite=.
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Limits on Reactor Building Pressure While Critical.
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(Detail 2)
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High and Low Pressure Injection Analog Channel Test.
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(Detail 3)
Em iij:;;r Reactor Building Spray Actuation Setpoint.
(Detail 4)
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Reactor Protection System Channel Bypass Key.
(Detail 5)
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Restriction on Reduction of Boron Concentration.
'(Detail
6)
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Reactor Coolant Leakage Limitation.
(Detail 7)
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.Shif t and Daily Checks - Core Flood Tank Pressure.
(De-
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tail 8)
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Shift and Daily Checks - RC Pressure Temperature Compara--
tor.
(Detail 9)
1.
Reactor Protection System Surveillance Procedure.
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tail 10)
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Variable Low Coolant Pressure Trip Serpoint.
(Detail 11)
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Reactor Building Pressure Trip Setpoint.
(Detail 13)
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Design Change Control Procedure.
(Detail 15)
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2.
Unresolved Items
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(These are items for which more information is required in g
H order to determine if the item is Acceptable, a Deviation or
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an Item of Noncompliance.)
a.
Power Range Amplifier Calibration Procedure SP 1302-1.1 clarification of instructions on Data Sheet 1.
(De tail'
12.b.(1))
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b.
Power Range Amplifier Calibration Procedure SP 1302-1.1
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format clarification on Data Sheet 3.
(Detail 12.b. (3))
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Plant Startup Procedure OP 1102-2 Precritical Check' List c c#
changes to procedure numbers in ite:a 22A.
(De:sil 14)
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3.
Infractions and Deficiencies Identified by Licensee None y
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B.
Status of Previous Unresolved Items
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Resolved Items 3:.;;..
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- i;;;;;e The following items identified in Report 50-289/75-14 were ihr examined and resolved.
REE En a.
Reactor Containment System Leak Race, reference report
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detail 8.a.
(De' tail 16.a.(1))
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b.
Incore Neutron Detectors. reference report detail 8.c.
(Detail 16.a.(2))
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c.
Power Range Amplifier, reference report detail 8.f.
naa (Detail 16.a. (3))
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d.
Main Steam Isolation Valves, reference report detail
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(Detail 16.a.(4))
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Station Batteries, reference report detail 8.k.
(Detail (
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16.a.(5))
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Reactor Buildf.ng Emergency Cooling and Isolation System ii
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Analog Channels, reference report detail 8.n.
(Detail E.f.
16.a.(6))
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Tests and Experiments, reference report detail 7.d.
(Detail 16.a. (7))
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A0 75-19, reference report detail 10.d.(3).
(Detail 16.a.(8))
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A0 75-18, reference report detail 10.d.(8).
(Detail 16.a.(9))
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PORC, reference report 13.b.
(Detail 16.a.(10))
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Borated Water Storage Tank, reference report detail 8.p.
(Detail 16.a.(11))
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2.
Items Still Unresolved
- l The following items identified in Report 50-289/75-14 were examined and remain unresolved.
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a.
Turbine Overspeed Testing, reference report detail 8.g.
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(Detail 16.b.(1))
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b.
Transmittal of Safety Evaluations to General Office Review Board, reference report detail 7.c.
(Detail 16.b.(2))
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Management Interview h
A.
A =anagement interview was conducted at the site on. November 14,
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1975 to. discuss the findings at that point in the inspection with
the following licensee attendees.
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Mr. J. J. Co'litz, Unit No. 1 Superintendent Mr. J. C. Herbein, Manager - Generation, Operations - Nuclear Mr. D. L. Good, Technical Analyst III
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Mr. G. A. Kunder, Supervisor, Station Operations Mr. J. P. O'Hanlon, Senior Engineer, I
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The following summarizes the items discussed.
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1.
Power Range Amplifier Calibration.
(Detail 12.b)
2.
Plant Startup Procedure.
(Detail 14)
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3.
Nuclear Overpower Trip Setting.
(Detail 2.a)
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4.
High and Low Pressure Injection Analog Channel Test.
(Detail 3)
5.
Reactor Building Spray Actuation Setpoint.
(Detail 4)
6.
Restriction on Reduction of Boron Concentration.
(Detail 6)
7.
Reactor Coolant Leakage Limitation.
(Detail 7)
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B.
An exit interview was conducted at the site on November 19, 1975 to Q
discuss inspection findings. not covered in the previous meeting,
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with the following attendees.
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Mr. J. J. Colitz, Unit 1 Superintendent Mr. D. L. Good, Technical Analyst III
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The following summarizes the items discussed.
1.
Operability of Nuclear Safety Instruments.
(Detail 14)
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2.
Reactor Protection System Channel Bypass Key.
(Detail 5)
3.
Design Change Control Procedure.
(Data 11 15)
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DETAILS
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Persons Contacted
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- (I Discussions were held with the following persons either onsite or at the Corporate' Headquarters during the conduct of the inspection activities documented in this report.
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Mr. R. O. Barely, Engineer II Ms. R. S. Brown, Technical Analyst III Mr. J. J. Colitz, Unit No.1 Superintendent Mr. P. Chalecki, Control Room Operator (in-training)
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Mr. D. L. Good, Technical Analy.,st III
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Mr. R. A. Klingaman, Manager Generation Engineering
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Mr. G. A. Kunder, Superviscr, Station Operations Mr. S. Lehigh, Instrument Man, Nuclear Mr. J. P. O' Hanlon, Engineer, Senior I y.)F
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Mr. V. Orlandi, Lead I and C Engineer
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Mr. R. H. Porter, Shif t Supervisor (SRO)
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Mr. W. Potter, QC Engineer Mr. D. Reich, Instrument Man, Nuclear Mr. M. J. Ross, Shift Supervisor (SRO)
Mr. M. A. Shatto, Engineer, Associate I Mr. W. J. Sawyer, Assistant Supervisor, Maintenance
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Mr. D. M. Shove 11ng, Maintenance Supervisor Mr. J. Smith, Shif t Foreman (SRO)
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Mr. P. Tinnes, Instrument Man, Nuclear
2.
Limits on Reactor Building Pressure While Critical j
The inspector reviewed SP 1301-1 Surveillance Records f.or Shift and
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Daily Checks for the months of July, August and Sap: ember to verify
!!1 that the Reactor Building Pressure limit of 2.0 psig or 1.0 psi vacuum with the reactor critical (T.S. 3.6.4) was maintained.
?2.
The inspector also reviewed Operating Procedure 1102-2 " Plant
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Startup" to determine if that procedure provided for assuring that reactor building pressure is within the above limits prior to
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criticality.
The inspector found no inadequacies in this area.
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High and Low Pressure Injection Analog Channel Test (3
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The inspector reviewed Surveillance Procedure 1303-4.11, Rev. 8 dated 8/12/75.
This procedure is used to implement the monthly test required by Technical Specification 4.1-1, item 17.
The three
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Reactor Coolant Channels are tested by applying an analog voltage F
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signal to the pressure trip bistables and verifying that the vol-tage required to trip them corresponds to pressure trip points of
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1540 to 1550 psig and 540 to 550 psig.
Technical Specification 3.5.3.1 requires these setpoints to be, respectiver 1500 psig and 500 psig, or greater.
The inspector revieued the surveillanc'e records for SP 1303-4.11 which were perforced on 7/6/75, S/6/75 and 9/5/75.
The data shcets indicated that the procedure was completed at the required frequency (monthly), and that the pressure setpoints were in accordance with T.S. 3.5.3.1.
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The inspecto'r found no inadequacies in this area.
4.
Reactor Building Spray Actuation Setpoint The inspector reviewed Surveillance Procedure 1302-5.11 "Rea.ctor ('~'}
Building 30 psig Pressure Channel" which is used to adjust or check the Reactor Building Spray actuation setpoint bd be less than or
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equal to 28.5 psig.
Technical Specification 3.5.3.1 requires this (_c j
value to be less than or equal to 30 psig.
The inspector reviewed
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the surveillance records for SP 1302-5.11 which was performed on 1/15/74.
These records are still applicable since the channel
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calibration is required each refueling period (T.S. 4.1.-1 T21).
This.vas the initial calibration and the pressure sett.oint was set at 28.5 psig.
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The inspector found no inadequacies in this area.
5.
Reactor Protection System Channel Bypass Key The inspector determined, by discussion with the licensee, that only one channel bypass key was in the control room in accordance with Technical Specification 3.5.1.2.
Plant keys are kept in a key locker adjacent to the Shift Foreman's desk.
This key locker is normally locked.
When any key is removed, the person charged with z.;.
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the key is entered on the key log.
The inspector reviewed this log
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to verify that only one RPS Channel Bypass Key was checked out at'
fll any given time.
The inspector also observed that only one such key E!:T!.?
was available in the key locker.
By inspection and discussion, the
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inspector determined that the licensee scheduled RPS surveillance
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testing on a one channel per week basis and therefore the surveil-
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lance schedule aids in preventing any attempt to bypass two channels J
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simultaneously.
'jj The inspector found no inadequacies in this aren.
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Restriction on Reduction 6f Boron Concentration
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Technical Specificatien 3.1.1.1b ste tes, "The boren concentration N
in the reactor coolant system shall nec be reduced unless at least
- d one reactor coolant pump or one decay heat removal pump is circu-
lating reactor coolant." The inspector reviewed SP 1301-1 Shift i
and Daily Ch.ecks and determined that all of the Reactor Coolant Ej!
Pumps were shutdown from 9/27/75 to 9/30/75.
By discussion with sj the licensee and review of the Shift Foreman's log the inspector
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determined that Decay Heat Removal System Loop B was placed in ser-
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vice at the time the last Reactor Coolant Pucp was shutdown an'd that
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loop B continued to operate until the RCPs were placed in operation.
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The inspector found no inadequacies in this area.
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Reactor Coolant Leakage Limitation SII $
$
The inspector reviewed Surveillance Procedure 1303-3.1 Rev. 5, H
10/14/75 " Reactor Coolant System Leak Rcta" which is pe-rfor.ed daily j
to icplement Technics 1 Specification 4.1-2T11.
The acceptance
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criteria for total Reactor Coolant Leakage is less chan er equal F ::
to 10 gpm.
The inspector reviewed the SP 1303-1.1 data sheets for m.)
July, August and September of 1975 and verified that the calculated-ll1
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total reactor coolant leakage did not exceed the 10 gpm valua of
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T.S. 3.1.6.1 at any time.
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In addition, the inspector observed that, during September 1975,
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11 c here e:1-the calculated total losses plus leakage from the primary system
=F were less than 2 gpm.
Linear extrapolation of the least squares G.py, fit of the computed data values correlated to an unidentified in-leakage of-1.5 gpm to the reactor coolant system.
However, the
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mean value of unidentified leakage rate times the total time period-
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of measurement (45 hrs) indicated a measured total of unidentified
in-leakcge of about 12 gallons.
The licensee stated that this was
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vithin the expected accuracy limitations of the measurements involved.
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The inspector hcd no further questions in this area.
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Fh " f t end Daily Chtchs-Core. Flood Tank Pressures
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The inspector reviewed SP 1301-1, " Shift and Daily Checks", sur-ve111ance records for September and October '1975.
These data sheets are used to document implementation of Technical Specifica-
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tion 4.1-1 T25.a which requires a. check of Core Flood Tank Pressure 7. 7 Channels every shif t.
Recorded values are co= pared against an
- gl acceptance criteria in SP 1301-1 which states in part "...within 7,'.....
normal expected range." The inspector observed that for all shifts
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the values recorded were acceptable.
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The inspector had no further questions on this item.
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Shift and Daily Checks-Reactor Coolant Pressure /Tempertture
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Comparator si;s
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[5 The inspector reviewed SP 1301-1, Shift m d hily Checks, surveil-lance records for July and August 1975.
Thase data sheets are used
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to document implementation of Technical Specification 4.1-1T11 which requires a check of the RC Pressure / Temperature Comparator
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Channel every shift.
The recordec values are compared against an
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acceptance criteria in SP 1301-1 which states in part, "...within normal expected range...".
The inspector observed that for all
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shifts the recorded values were acceptable.
The inspector found no inadequacies in this area.
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The inspector reviewed Surveillance Procedure 1303-4.1 Rev. 15, mm 9/24/75 which is.used to implement the Technical Specification
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requirements for monthly channel tests on the following channels.
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Reactor Coolant Temperature.
(TS 4.1-1T7)
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b.
High Reactor Coolant Pressure.
(TS 4.1-1T8)
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Reactor Coolant Press ~ure Temperature Comparator.
(TS 4.1-g);
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Tha inspector uitncsced the performance of this test en the instru-
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cent associated with RPS channel D.
The inspector noted that the g..
test was performed per the procedure.
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The inspectoi reviewed the Data Sheets for SP 1303-4.1 for the tests performed in July, August and September of 1975.
The inspec-tor noted that each RPS Channel was tested each month and that therefore the instrument channels listed above were tested each
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month per the requirements of T.S. 4.1-1.
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The inspector expressed concern to the licensee that there were several examples where the "as left" value of the recorded Digital
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O Voltmeter (DVi!) readings were not entered on the Data Sheet.
Since
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the "as found" values were within Technical Specification and Administrative Ld=its, the "as left" values should have been the same as the "as found" values in these cases, and it appeared that the blanks indicated a fcilure to rc:ord an unchanged value.
Tha E
licensee stated that more detuiled completion of Data Sheets had been emphasized to the technicians.
The inspector reviewed data
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sheets for 1303-4.1 completed in October 1975 and noted that as lef t" values had been completed, therefore, verifying ef fectiveness
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of licensee corrective action.
The inspector had no further questions in this matter.
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Variable Low Coolant Pressure Trip Setpoint The inspector review'ed SP 1303-4.1, " Reactor P: icction System Surveillance," which is used to check and/or ad3ust the Variable Low Coolant Pressure Trip Setpoint at a value greater than or equal to the requirements of T.S. 2.3-1T6.
(16.25T out-7756).
Procedure
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1303-4.1 Section 6.8.5 verifies that at a Reactor Coolant Outlet temperature of 5900F the pressure trip setpoint is greater than or equal to 1831.5 psig and at a temperature of 604cr t.he setpoint is
.g greater than or equal to 2059 psig.
The procedure steps perform
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E this verification by inputing to the trip bistable the voltage
value corresponding to the temperature listed above and then de-
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creasing the voltage input which simulates a decreasing reactor E
pressure until the bisttbic trips.
Tae trip voltat;c is then re-f corded and compared to licensca determined voltages.which correc-pond to the pressure values listed above.
gjj
- +a The inspector reviewed those Data Sheet 6, items 4 and,7 of SP 261 1303-4.1 which were completed in July, August and September of 1975 for each pressure and verified that the "as found" Variable Low Coolant Pressure Setpoint based on an analog input voltage was greater than required by T.S. 2.3-1T6.
The inspector noted thet for Data Sheet 6 step 6, completed on 7/22/75, the analog voltage (,
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used to simulate reactor temperature was recorded as -0.084, how-
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ever, the required accuracy is +0.0002.
The li.densee stated that y;
the value should have been recorded as -0.0840 and that Data Sheet
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V 6 of SP 1303-4.1 had been revised en 8/22/75 to state this more p
clearly.
H The inspector had no further questions in this area.
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12.
Nuclear Overpower Trip Setting
The inspector reviewed the following procedures which are used by
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the licensee to implement Technical Specification 2.3-1T1, Nu-lear Power Trip Setting of 105.5%.
SP 1303-4.1, " Reactor Protection System Surveillance," is performed monthly on each RPS Channel.
Section 6.5.8 of that procedure verifies that the High Flux trip bistable is set at less than or equal to 105.5% and adjust the trip setting to be 104.75%.
SP 1302-1.1, Power Range Amplifier Calibra-tion, is performed on a minimum of twice weekly and is used to i
adjust the output of the power range amplifier to a value equal to j
core power as determined by a heat balance calculatien.
The power
'l range amplifier providas the analog input to the dirh Flux trip bistabic and therefore, the calibration of this sapitfier is necessary to i;he accuracy of the Nuclear Overpower Irip Setting.
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The inspector reviewed the survaillance records for SP 1303-4.1 and
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SP 1302-1.1 for the period of July, August and September of 1975.
This review indicated that the nuclear overpower trip setting was 74 4 maintained at less than 105.5%.
The inspector had the following comment on these surveillance records,
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a.
RPS surveillance test SP 1303-4.1 was performed 12 times during July, August and September of 1975.
The inspector reviewed Data Sheet 3 for each of these surveillance tests to
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determine that the High Flux Eistable was set at 1 css then or equal to 105.5%.
The inspector noted that, on two occasions on the 7/22/75 Surveillance of P,PS channel C, and en the pq 9/24/75 Surve11L a.:t v T.:S Channel D, t ne :.m:tru...nt tc :h-b~l nical negiccted to enter the High Flux bistf.de inpu: voltage a
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at the trip points as required by the procedure.
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A power. range channel consists of two neutron dettetors (an
",l upper and a lower), each having a linear amplifier and a q
sumndng amplifier which provides an output proportiocal to core power.
This output provides indication of core power
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and input to the high flux bistable.
The trip point of this bistable is established by first setting the test input. voltage
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to one of the neutron detector linear ampli,fiers at a givea p }
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value and then increasing the test input voltage to the other linear amplifier until the bistable trips.
This test 16put
e-(_)3 value is then compared against a voltage criterion which corr-
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R esponds to the 105.5% overpower trip setting.
The procedure i
requires that the voltage to the bistable also be recorded and comparad againct a voltage criterien *:hich c'.on correspor.ds
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-to the 105.5% overpcwcr trip s2tting.
1he licensee stated thct
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recording of the bistable input was a " backup" to the previous-ly recorded value of the input to the linear amplifier and therefore, since that value had been recorded, the bistable trip was in compliance with T.S. 2.1-1T1 (105.5% overpower trip).
Technical Specification 6.2.3 states in part:
"rit t en
.
procedures shall be strictly adhered to in all matters re-lating to nuclear safety." Contrary to this Technical Spec-ification, the requirements of SP 1303-4.1 Data Sheet 3, item 3, were not adhered on two occasions 7/22/75 and 9/24 75.
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This item is an example of a Deficiency.
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SP 1302-1.1, " Power Range Amplifier Calibration" was performed y
264 times in July, August and September of 1975.
The inspec-di-tor reviewed the data sheets associated with the performance
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of each of these tests and had the following com ents.
"
(1)
The icstructions included on Data Sheet 1 of SP 1302-1.1 differs from procedural step 6.1.2.
The procedure re-
, p quires recalibration of the pcuer range chcnnel if the
,
computer read out of neutron power differs from the heat
[F" balance calculation of core power by more than 1.*:, or if
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the console indication of neutron power differs by nore Ed than 2% from the same calculction of core power.
Mcw-E~i cver, the Data Sheet specifies "anc/or" for those two
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conditions.
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'This is an unresolved item.
N Eb (2)
The inspector noted that, for SP 1302-1.1 Data Sheet No.
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1 dated 8/7/75, power range channel NI-7 computer readout value was 1.277% above the core power determined by heat i
balance.
Contrary to step 6.1.2 of SP 1302-1.1 the power J
range channel was not recalibrated.
Since the power l
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range channel indicated greater than core power this l
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conservative deviation aff ectively it<!ered the nuclear p
overpower trip setting.
However, failure to recalibrate V
the channel is contrary to Technical Specification 6.2.3
-.;
which requires strict adherance to written procedures, 4j and such failure with non-conservative deviations could
'
lead to failure to meet Limitir.g Safety Syste Settingc.
This item is' an exampic of a Deficiency.
(3)
Data sheet 3 of SP 1302-1.1 is used to document Power e.y
Range Channel values following recalibration.
The inspec-tor noted that the Core Thermal Power entry on Data Sheet 3 is not followed by a blank.
This apparently contributed Jj to the operators failure to enter the core thermal power su value following all of the six amplifier recalibrations f;j performed in July, August and September of 1975.
The
'
licensee stated that such a blank wculd be added to the sheet, g
This item is unresolved.
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Reactor Building Pressure Trip Setpoint k-The inspector reviewed Surveillance Procedure 1302-4.13
" Reactor
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Building En.4 gency Cooling and Isolation System Analog Channels,"
which is us2d to check and/or adjust the setpoints for the Reactor Building high pressure trip points.
The test verifies or adjusts the setpoints to a value of 2.4 psig to 2.6 psig by using an analog
-
test input voltage to the bistable.
This voltage corresponds to the voltage signal received from the pressure transmitter.
Tech-nical Specification 2.3-lT8 requires a Reactor Building High Press-
[g ure Setpoint of less than or equal to 4 psig.
The inspector reviewed
.
the^ surveillance records for SP 1302-4.13 for tests performed on (i 7/1/75, 8/Q/75 and 9/4/73 and verified that the "as found" setpoints r?.E vere within the r:.cuircuer:s of FP U?2 a.13 and therefore len than the 4 p.sig require. ants et I. S. 1. 3-lTS.
E The inspector found no inadequacies in this area.
g 14.
Operability of Safety Instrumentation
The inspector reviewed Operating Procedure 1102-2 to v.erify it implemented the operational requirements imposed by Technical Specification 3.5.1.1 on,the following instrument channels.
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a.
Reactor Building 30 Psig (Reactor Euilding,, Spray) Instrucent O-Channel.
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b.
Reactor Building 4 Psig (Low Pressure Injection) Instrument Channel.
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. Reactor Coolant Pressure (Lou Pressure Injection) ir. strut?r.t (
c.
Channel.
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Manual Pushbutton, High Pressure Injection Logic Channels.
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Reactor Building 4 Psig (High Pressure Injection)' Instrument
$: H Channels.
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Power / number of pumps Instrument Channels.
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- Flux / imbalance / flow Instrument Channels.
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Power Range Instrument Channels.
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Intermediate Range Instrument Channel.
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Source Range Instrument Channel.
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OP-1102-2 contained a "Procritical Check List," which included a
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specific signoff for each of the above items.
This signoff in-dicated that the appropridte Surveillance Test had been performed successfully in the required nuabar of enar.nels prior to sterrup, de j,
This implenents Technicc1 Specification ~.3 vaica doi'.r.es "Oparahls" and states in part "... tested p+-todically in acccrdance with 5pec-
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ification.4 and has met its performance require =ents."
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E During this. review the inspector noted that item 22A of the Pre-critical Check List references by number surveillance procedures
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which are not consistent with the presently used procedure nu=ber, i==
For example, item 22A lists Pressure Temperature surveillance procedure as SP 1303-4.8.
The correct reference is Section 6.8.5 ti
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of SP 1303-4.1.
The licensee stated that procedure nu=bers would C
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be reviewed and corrected for this checklist.
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15.
pesign Change Control Procedure I
The inspector revieued CP 1003. Rev. 2. 7/22/75. " Control of Desis.n h
Change / Modification" which tnc licenace us=s to inple ent the
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rr quirement of Technical Specification 6.1.1.2.b.2 and 6.1.1.2.b.3.
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ically requires a " safety evaluation" to determine if the proposed change would constitute an unreviewed safety question or change to the Technical Specification.
The inspector toured the Technical Support Staff of fices and verified procedure GP-1003 was available for the case of the Technical Support staff.
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'fechnical Specification 6.1.1.2.b.3 requires that Corporate Tech-lii.
nical Support Staff specify any tests that must be performed f
following plant change or modification.
To determine if T.S, E
6.1.1.2.b.3 was implemented, the inspector reviewco all Change /
=1. l Modification (C/M) packages of all 27 C/M's approved by the Man-T
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ager, Generation Engineering during July, August, September and
- ,[
October of 1975.
Each C/M package contains a C/M Design Checklist
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GPF 1003.002, R2v. 2 dated 7/22/75, which the Cc;nizant Engineer assigned to review the C/M cc: pletos.
Item 21 of that checklist is i;;:i.
a s.ignoff that testing recuircocnts were reviewed and includa.i in r
C/M approval uhen recuired.
Ihc inspector expressed concern that W.
the C/M checklist was not with the C/M package at the site.
Also with the e.<ception of 5 ef :1.,2 3 C/N's revic:ee, : c : y:cific menti 3n was made in the C/M SPprev:1 memo that testing rar,uircr:m s taa
...
Seen reviewed.
The licensee stated that a memo was written en
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11/10/75 to Generation Engineering Personnel from the Manager,
- .g Generation Engineering, requiring that all approved memos for i
- 1 Change / Modifications would specifically mention that review of Test / Retest requirements had been made and detail the results of that review.
- .
E The inspector had no further questions on this item.
It will, how-3: L
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ever, be subsequently reviewed as a part of the routine inspection program.
.
16.
Previously Unresolved Items
..
a.
Resolved Itams The following ite=c id2ntified in Report 50-2G9/7[-14 were examined and resolved.
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(1)
Reactor Containment System 1.eak Rate (Ref:
Detail 8.a)
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SP 1303-1.1, Revision 5, 10/14/75, has been issued.
DVM traceability is provided by recording of model and serial number, providing a reference to calibration recoras.
The inspector had no further questions on this item.
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Incore Neutron Detectors (Ref:
Detail 8.e)
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s SP 1301-5.3, Revision 3, 10/22/75, has been issued.
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traceability is provided by recording model and serial 7;;;;;
number. Th'e inspector had no further questions on this
,
item.
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(3)
Power Range Amplifier (Ref:
Detail 8.f)
SP.1302-1.1, Revisio.1 6, 10/20/75, has been issued.
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traceability is provided by recording model and serial
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number. The inspector had no further questicas on this
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iten.
.
(4). Main Stear Isolation Valves (Ref:
Detail 8.h)
SP 1303-4.17, Rev. 2, 10/8/75, has been issu.ed.
Step 6 calls for stationing an Auxiliary Operator at the valve to verify that it is 100% orrn, to listen for abnor al-ities, and to verify that tha v 1ve moves the required
,
10% and goes back open.
The Inspector had no further
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questions on this item.
- .
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(5)
Station Batteries (Ref:
Detail 8.k).'
SP 1301-4.6, Revision 2, 9/10/75, has been issued.
DVM model, serial and calibration data are required to be
. recorded.
The inspector had no further questions on this item.
.
(6)
Recctor Building Exercenev Cooline and Isolation Svstem Analog Channels (Ref:
Detail 8.n)
,.
FE q SP 1301-4.13, Revision 4, 10/20/75 has been issued.
t Ll Licensee review was accomplished.
Step 6.2.7 now
_
specifies resetting of the bistable, reference to a QA
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procedure in step 6.3.2.e was considered unnecessary and
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deleted, and step 6.3.9 was considered satisfactory.
DVM data providing traceability is required to be recorded.
The inspector had no f urther questions on this item.
.
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(7)
Tests and Experiments (Ref:
Detail 7.d)
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The licensee stated that the present procedure had been
,
evaluated with regard to T.S. 6.1.C.3 which requires safety evaluations in accordance with 10 CFR 50.59 for tests and experiments and that any test or experiment performed would require Special Operating Procedure, Temporary Change Notice or a Permanent Change Notice prior to their perfor=ance.
All of the above must be reviewed and approved by the PORC and 'Jnit Superintend-ent.
The inspector had no further quiscions on this item.
.
.
(8)
A0 75-19 (Eif:
DetM 1 10.d._()h T6 inspector rcviewed K?S surveillance for Variable Reactor Coolant Pressure Trip Setpoints for July, August and September of 1975.
There was no further indication i;;;
of setpoint drif t and all tests indicated that setpoints were within procedural and Technical Specification limits.
The inspector had no further questions on this item.
(9)
A0 75-18 (Ref:
Detail 10.d.(8))
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The inspector reviewed TMI Unit No. 1 Sto,res Procedure l
No. 1. " Procurement. " The purpose section of this pro-Fl
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cedure states in part:
"... establish administrative con-O, trols and requirements related to the requisitioning of material, equipment or outside technical services...".
- q The inspector had no further questions on this item.
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(10) PONC (Ecf:
Catail 13.b)
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The inspector reviewed Technical Specification Change '
Request No. 5, Amendment No. 3, transmitted to the NRC by licensee letter GQL 1579, dated October 3, 1975.
The inspector.had no further questions on this item.
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(11) Borated Water Storage Tank (Ref:
Detail 8.p)
SP 1301-4.4, Rev. 2, 9/12/75, has been issued.
Step No.
5 now references CP 1912, Determination of Boron.
The purpose of the procedure now includes a phrase, " Boron Concentration requirements of T.S. 3.3.1.la only." The
,
licensee stated that the water volume and temperature specification of T.S. 3.3.1.la are addressed in other appropriate procedures.
The inspector verified that this requirement was included in the Plcnt Startup Precedure 1102-2.
The inspector had no further questions on this
item.
.
b.
Iter s Sd!' U-r ::olw d The fo31owir.g items, identified in Report 50289/75-14 were examined and remain unresolved.
(1) Tur'bine Overspeed Testing (Ref:
Detail 8.g)
,.
DVM traceability is not yet provided in SP 1303-11.19.
,
IA (2)
Transmittal of Safety Evaluations to General Office Review
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Board (Ref:
Detail 7.c)_
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The inspector reviewed AP 1001, Rev. 3 of March 7,*1975.
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Impact Evaluations of procedure revisions and changes and also requires (Line 5) that such evaluations be sent to
- .
the GORD.
The insoecter reviewed a Procedure.Chr.nr.e b
Request (PCR) to nP 1016, Sectica 7.. 2 'thich has beau
cntered for approval.
This ch.snge vill requira that
,
safety evaluations transmitted with C/M's will be sent
'p to the GORB.
Issuance of the change to AP 1016 will
- /t resolve this item.
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p" 17.
Noncompliance Resolution The inspector reviewed the status of the following noncompliances from inspection 50-289/75-14.
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Recording Information on AO's (Ref:
Detail 3.a)
The licensee's response to this citation identified logging of a
the item in question and emphasizing to personnel of the responsibilities in logging abnormal occurrences.
Inspector
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review of the Control Room Log Book during this inspectiun showed written emphasis upon cntering AO's in the Control
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Room Log on the part of the operatiens supervisor.
The in-spector had no further questions on this item.
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b.
Log Book Revieu (Ref:
Detail 3.b)
Instances of documancation of log beok revieu by the Operau tions. Supervis ?r was no:cd in tho' Cent.:.;l
'c :. I.ct Ee: : hy tl.'
inspector.
L: sit Revisden 4 cf A? 2012 r;:cificcl:y requir.-s
..."
Operations Supervisor docum:ntation of review of the Cantal Room Log Book.
This item is unresolved pending revision.of AP
,
1012.
c.
Jumper Log Review (Ref:
Detail 3.c)
-
Review of Jumper /Lif ted Loads-1!echnical !!odification Log Book
37 identified no Shift Foreman failures to initial entries
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..
since the citation for this noncompliance.
However, two
-
(
)
instances,of failure to identify the " installer" in the coluna
.
recorded for entries on October 16, 1975 for Jumper Tags 11
'
O and 14 were identified.
Generic corrective action adequacy
,
for ensuring that this log provides the necessary data is, consequently, unresolved.
d.
Record Recoine Ti'.e Custodian Cd:
O c t o i'. 4.c)
GP 4407, Revision 0, 10/17/75, Regulatory Retention and Stor *
age of Quality Control Department Records, designates the Administrative Assistant-Quality Control as records custodian.
This item is resolved.
_
c.
Engineering Drawing Control (Ref:
Detail 5.b)
.
The licensee's memo GDI 3321 of 9/29/75, GQL 1507 TMI-1, con-cerning THI-1 NRC Inspection 75-14 documented the committed
.
licensee review of 20 drawings with 1 deficiency identified and reaudit scheduled in June 1976.
The inspector had no
_
'further questions on this item.
.y 90012542
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