ML20127H255
| ML20127H255 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/28/1978 |
| From: | NORTHERN STATES POWER CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 9211180461 | |
| Download: ML20127H255 (23) | |
Text
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MSP & .
NORTHERN STATES POWEH COMPANY MINN E A PO Li S. M8NN E SOTA S5401 1
g)~; -
Februa ry 28, 1978 '
Mr J G Keppler, Director, Region III "
Office of Inspection L Enforcement t-U S Nuclear Regulatory Cocznicsion 5 799 Roosevelt Road _
Glen Ellyn, IL 60137
Dear Mr Keppler:
MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Annuct Report of Occupational Exposure and Changes, Tests & Experiments Janua ry 1 - December 31 1977 In accordance with Appendix A Technical Specification 6.7. A,2 for the sub-ject license, enclosed are two copiec of the Annual Report of Occupational Exposure. The section on Changes, Tests and Experivants has been included in this report as a convenient means of meeting the annual reporting re-quirements of 10CFR50.59(b).
s In our license amendment request dated October 31, 1977, which proposed to delete the requirement for an Annual Operating Report, we comitted to submittal of a Narrative Summary of Operating Experienen for the year 1977.
The sucznary it centained in the attached report.
Yours very truly, L 0 Mayer, PE Manager of Nuclear Support Services LQi/deh cc: Director, IE c/o Distribution Services Branch, DDC, ADM (40)
G Charnoff MPCA Attn: J W Ferman O Attachment i g
9211180461 780228 PDR ADOCK 05000263 l
' i R PDR ;,, ,,)-
______-__-___-_-____-____-__:-_ -- t--
s TAB 1.E OF CONTENTS ,
i I Narrative Sunnary of Operating Experience II Occupational Exposure Report III Changes, Tests and Experiments 4
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I. NARRATIVE SGM\RY OF OPERATING EXPERIENCE 1/1/77 Operated at 100% of rated power except for brief week-to ly reductions for control rod exercising and valve 2/22/77 testing.
4 On 1/18/77 the accumulator on CRD hydraulic control unit 14-19 would not hold pressure due to stem packing leakage in the nitrogen charging valve. Evidence of packing material in the sten threads, and damage to the packing indicated improper packing installation.
Instructions were issued concerning packing adjustments and accumulator operability was restored with the pack-ing replaced on 1/18/77 (Reportable Occurrence No.
M-RO- 77-01) .
On 1/25/77 it was noted that the SBGTS flow rate had not been recorded monthly in compliance with Techni-cal Specifications which were issued 9/27/76 The test procedure was revised to provide for a monthly flow record and the administrative procedure for irplementing new or revised Technical Specifications was also revised (Reportable Occurrence No. M-RO-77-02).
On 1/27/77 No. 3 TIP Ball Valve failed to close during routine operation of the TIp System. The valve was tapped and it closed. The cause of the failure could not be determined. The valve was replaced with the latest model, which was an improved operator (Reportable Occurrence No. M-RO-77-03).
2 2/23/77 A reactor scram occurred when a load rejection caused a turbine control valve fast closure. The load re-jection resulted when a ice storm caused line problems whichtripped the plant output breakers.
During plant startup a scram occurred due to high neutron flux when a reactor period of less than 5 seconds was obtained while withdrawing an in-sequence control rod. Analyses indicated that the combination of high temperature and high xenon concentration at the time of the occurrence established conditions such that criticality occurred on an unusually high reactivity worth control rod notch. 'Ihe observed period was consistent with core analysis data (Report-able Occurrence No. M-RO-77-04). Following tempera-ture reduction and xenon decay, plant restart was initiated.
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2/23/77 (Cont'd) Administrative and Operating Procedures were subsequent-ly revised and new core analysis procedures were in-stituted to identify high reactivity worth notches, place restrictions on their withdrawal, and clarify administrative procedures pertaining to anomalous re-activity changes.
, 2/24/77 to Power was gradually increased to 100% of rated.
2/26/77 2/27/77 Operated at 100% of rated power except for brief-to weekly reductions for control rod exercising and 3/17/77 valve testing.
On 3/1/77 following the installation of a redundant torus level transmitter, a discrepancy was noted between the two torus level indicators. Investiga-tion revealed that the actual torus water volume was slightly below the minimum Technical Specification limit.
Water volum was returned to the normal and the failed transmitter was replaced (Reportable Occurrence No.
M-RO- 77-05) .
3/18/77 Commenced power reduction in preparation for scheduled maintenance shutdown.
3/19/77 Scheduled outage to perform the following maintenance:
to 3/20/77 a. Repaired pilot valve leakage on 4 reactor safety /
i relief valves and installed filters in the pres-sure sensing lines for all 8 valves.
- b. Plugged leaking tube in low pressure feedwater heater 13B.
\
- c. Replaced inboard seals on reactor feed pump #12.
- d. Replaced 2 main condenser air vent valves.
3/21/77 Returned to power operation and increased power to to 100% of rated.
3/25/77 On 3/21/77 the air-ejector sample system was found isolated, such that both air-ejector radiation monitors had been inoperable during startup. Operation cf the monitors was re-established. The work control process
- and startup procedures were revised to prevent a re-I currence. (Reportable Occurrence No. M-RO-77-06).
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3/26/77 Operated at 100% of rated power except for brief weekly _
, to reductions for control rod exercising and valve testing.
4/11/77 .
i on 4/4/77 during a routine surveillance test, the set-
, point of one of the HPCI steam line area-temperature "
! switches was found to have drifted above the a13cw-able Technical Specification limit. 'Ihe switch was replaced (Reportabic Occurrence No, h!-RO-77-07).
L l 4/12/77 Power was reduced to 66% of rated for load following.
+
I 4/13/77 Operated at 100% of rated power except for a brief. '
to reduction for control rod exercising and valve testing, "
4/18/77 i
4/19/77 Following routine maintenance:on f12 Reactor Protection L l MG Set, it was started and.a transfer of load from the altetr. ate source to the MG Set was attempted,' initiating-
! an expected Channel B half scram. The MG Set output - ,
- circuit breaker had~not been reset which caused a delay in the transfer. After approximately 1.5 seconds a .
reactor scram occurred due to a false indication of j high flux on APRM #2. hhen the channel B power range -
i neutron monitors were de-energized during the power source transfer,the shared 1.PRM inputs-from-APRM #6-were automatically removed-from the APRM #2 averaging l circuit. As a result of this action the LPRM average 4
- input increased, causing _APRM #2 indication to increase _ '
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- above the scram setting. The MG Set transfer was com-r pleted and the plant was restarted.. Procedures have now been prepared _to calculate the.effect on APRM's
' prior to making such transfers. An investigation of possible changes in LPRM assignment to minimite the affect
- on APRMs is in progress.-
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i- 4/20/77 to Power was gradually increased _to 100% of rated.-
- 4/22/77 4/23/77 Operated at:100%. of rated ' power except for brief weekly 1
, to: reductions for control rod exercising and valve testing, i 6/2/77 4
On 5/30/77,-during the monthly RHR Motor Operated Valve {
' Operability Test, "B" RHR Injection Valve hD-2013 failed-l to open. _ A line motor control center "open" contactor J
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did not close as required when given the valve "open" signal causing the motor starter control power fuse to open. The fuse was replaced, the notor starter cleaned,- ,
l and 77-09p) roper stroking and operation was demonstrated.:(M-RO-- j
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- , . . .. .L,.. - -- - - - , , , - - - ~~~ ~ ~ ~-~ - - " ~ ~~ ' ' ~ ~ ~ ~ ~ ~ ~ ^
- 4/23/77 On 5/31/77, during the monthly RCIC Fbtor Operated Valve to Operability Test, RCIC Outboard Steam Supply Isolation 6/2/77 (Cont'd) Valve hD-2076 failed to close. The nain and limit switch gear train grease had deteriorated due to high ambient temperatures. Wom gears were replaced, the gear trains were cleaned and greased and valve operability was demonstrated. A ventilation modification ha,s significantly lowered the ambient temperature. (M-R0- 77-09) .
6/3/77 Power was reduced to 92% of rated for load following.
6/4/77 Operated at 100% of rated power except for a brief weekly to reduction for control rod exercising and valve testing.
6/9/77 6/10/77 Scheduled outage to perfom operator licensing demon-to strations,'a CRD hydraulic return line isolation test i 6/12/77 and the follodng maintenance:
- a. Plugged Icaking tube in high pressure feedwater heater 14A.
- b. Replace outboard seals on reactor feedwater pump #12.
- c. Repacked miscellaneous valves.
6/13/77 Returned to power Operation. Increased power to 98%
to of rated. Power was limited to 98% of rated due to low 6/19/77 feedwater pump suction pressure caused by Icakage through the feed pump recirculation valves.
On 6/14/77 the torque switches for the RCIC steam line outboard isolation valve were improperly adjusted, such that the margin for nomal deterioration was less than desired. The switches were subsequently properly adjusted and administrative procedures revised to clarify and '
improve control over such work. (M-RO- 77-10) .
On 6/17/77 a small leak was discovered in a welded joint on the 1" drain line connected to the "C" moisture sep-arator drain line. The leak was temporarily patched.
The original weld was found to be of poor quality. On 6/26/77, the original weld was ground off and.the joint rewelded (M-RO-77-11).
4 6/20/77 Operated at 98% of rated power, to 6/23/77 I-4
6/20/77 On 6/20/77, plant personnel were informed by General to Electric Co. of an inappropriate assumption used in the 6/23/77 (Cont'd) determination of Cycle 5 FCPR limits. A conservative reanalysis increased the transient delta-CPR for all fuel by 0.08. At the time of the occurrence, the reactor was operating within the new limits (M-RO-77-12).
On 6/23/77, the flow through Standby Gas Treatment System was found to be below Technical Specification requirements.
An isolation damper was found to be operating improperly due to a conbination of normal wear and inproper installa-tion of the air supply to the danper control. The valve controls were corrected and proper flow verified.
(M-RO- 7 7-13) .
6/24/77 Power was reduced to 58% of rated due to high vibration on #11 Ret.ctor Feedwater Pump (RFP).
6/25/77 Scheduled outage to repair feed pump recirculation valves to CV-3489 and CV-3490 and #11 RFP. Inspection of the pump 6/26/77 revealed damage to various components, including a severely rubbed shaft due to the suction flow guide being dis-placed, wiped inboard and outboard journal bearings, broken discharge and first stage diffuser capscrews, first stage diffuser vane to sideplate weld cracking, a broken first stage diffuser vane and a damaged impeller.
6/27/77 Retumed to power operation. Power limited to 62% of to rated pending completion of repair of #11 RFP.
6/30/77 On 6/27/77, during plant startup, the indication of the air ejector off gas radiation monitors was found to be
' abnormally low due to air leakage into the sampling system caused by an improperly open manual valve. The improper valve position was due to an incorrect valve identifica-tion tag. The valve was closed and valve tags were corrected, (M-RO-77-14).
7/1/77 Following repair of #11 RFP, power was gradually in-to creased to 100% of rated.
7/4/77 7/5/77 Operated at 100% of rated power except for a brief re-to duction for control rod exercising, valve testing, and 7/15/77 load following.
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7/5/77 On 7/7/77, during routine operator- inspection, a steam' l
!- to leak was observed at the 1-inch leak test connection to-l
^ 7/15/77 (Cont'd) the RCIC steam supply line drain pot drain line. In-spection revealed poor quality of a welded joint. The j
' leak was temporarily repaired with a clamp and pack-ing. During the '1977 refueling outage, the faulty sec-tion of pipe and weld was replaced. (M-RO- 77-15) .
On 7/9/77, the "A" recombiner train off-gas _ flow control-i valve (KV-7489A) failed to stay closed after receiving
- a trip signal due to an accumulation of dirt in the asso-ciated solenoid valve. The valves operated properly after
, replacement of the solenoid valve internals. Previously installed supply line filters should prevent future dirt accumulation. (M-RO- 77-16) .
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On 7/11/77, following_ maintenance of the accumulator on i CRD HCU 26-23, the nitrogen cJiarging valve would not j- hold pressure. Inspection revealed that the valve was
[ not properly seated in the instncentation block, allow-t ing nitrogen to leak past the seat and body 0-rings.
The 0-ripgs were replaced and the valve was seated properly.
4 l (M-RO- 77-17) .
i 7/16/"7 Reduced power to 56% of rated for weekly control rod
- to exercising, valve testing and control rod pattern ad-J 7/19/77 j us tments. Gradually returned power to 100% of rated.
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, On 7/19/77, the nitrogen charging valve.on CRD HCU 18-11 L would not hold pressure. The nitrogen leaked past the-i valve stem as 'a result of defective packing, which had
! failed 'due to natural end-of-life. The valve was replaced.
j (M-RO-77-18) .
I I 7/20/77 Operated at 100% of. rated power except for- brief re--
to ductions for control rod exercising, valve testing, and j 8/12/77 load following.
p On 8/2/77, a small steam leak was discovered in aL45-degree elbow on the HPCI steam supply drain line to.the condenser.
' Investigation revealed that a steam trap upstream of the-
' elbow had failed causing erosion ~of the elbow. The leak was temporarily. repaired with a; clamp. During the 1977 refueling outage, the trap was repaired and the elbow was
' replaced. (M-RO-77-19) .
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1 On 8/5/77, 'during the daily HPCI auxiliary oil pump test,--
a resistor in the HPCI governor control system failed
}- resulting'in a _ loss of DC power to the system. The resis-
[ -tor was replaced with an equivalent adjustable resistor.
L (M-RO-77-20).
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7/20/77 On 8/10/77, the nitrogen charging vafve on CRD HCU 30-11 would not hold pressure. The nitrogen leaked past the to 8/12/77 (Cont'd) valve stem as a result of defective packing which had failed due to natural end-of-life. The packing was re-placed. During the 1977 refueling outage, the packing 2
was replaced in all 121 accumulator charging valves (M-R0- 77-21) .
8/13/77 Power was reduced to 54% of rated for control rod exer-3 to cising, valve testing and to adjust the control rod 8/18/77 pattern. Gradually increased power to 100% of rated.
8/19/77 Operated at 100% of rcted powcr except for brief reduc-to tions for load following, control rod exercising and 8/23/77 valve testing.
8/24/77 The recombiner system steam supply valve failed closed to due to low air pressure caused by a Icaking solenoid 8/25/77 valve on the condensate demineralizer system and a partially plugged air filter. Reactor power was immedi-ately reduced to 45% of rated. The valve was reopened 1
by installing a temporary air supply line. The recombin-er system and condenser vacuum were restored to normal, The solenoid valve was repaired and the filter was cleaned. As power was being increased, it was observed that the off-gas flow rate and recombiner bed temperature were icw. Also, the SJAE off-gas radiation monitor readings were gradually increasing. - A thorough investi-gation led to the conclusion that recombination was occurr-ing at the air ejector after condensers. Normal operation
' was restored by shutting off the off gas flow to the air ejectors for a short time. It is believed that the trip of the recombiner steam supply valve allowed flame propa-gation from the recombiners back to the air ejectors. Power was returned to 100% of rated and then reduced to 93%
, for load following.
8/26/77 Operated at approximately 97% of rated power. Power coast-down due to end-of-cycle reactivity depletion in progress, 8/27/77 Electrical noise caused by a lightning storm initiated a trip of the recombiner trains. Power was reduced to 46%
of rated until the recombiners were returned to operation and condenser vacuum was increased to nomal. Power was then increased to 97% of rated.
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!. 8/28/77 Operated at 92_ to 96% of rated power except for brief re-to ductions for valve tests, rod exercising and-load follow-3 9/8/77 ing.
3 On 8/28/77, during the monthly RfR hbtor Operated Valve
- Operability Test, "B" RFR Injection Valve (h0-2013) failed i
to open. The control leads for the "open" contactor for this valve are longer (about twice) than for any other con-tactor, resulting in a larger voltage loss. This, in con- .'
junction with a slightly worn contactor, prevented operation.
i The'contactor was cleaned and repaired, and control relays - <
were installed to reduce control wire voltage loss. (M-RO-77-22)
On 9/3/77 all control rods were fully withdrawn.
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! 9/9/77 Comnenced scheduled outage to refuel the reactor and-per-
{ fom plant inspections, modifications and maintenance. '
During operation of the RHR System, RHR torus ' cooling valve, FD-2009, failed to operate-properly. Inspection revealed- ,
that the stem clamp set screws had sheared allowing the stem j-to rotate. A modified stem clamp utilizing a keyway _was in-stalled.- (M-RO 23) . .
' In addition to refueling outage (9/9/77 to 11/10/7h) included majorthe items accomplished during the following:
- 1) Type "A", "B" and "C" containment leak rate testing. -
- 2) _In-service inspection activities.
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- 3) Survenlance tests and inspections.
- 4) CRU maintenance, including _ dye penetrant examination l of collet housings.
l 5) Repair and repacking of miscellaneous valves.-
l 6) Replacement of 101.PRM strings.
- 7) Removal of 4 neutron sources and source holders fromLthe .
Core.
- 8) Capping of .CRD hydraulic retum line reactor vessel nozzle and drywell: penetration and rerouting of return -l l-line to Reactor. Water' Cleanup System. l
' 9) Installation of modified barrel assemblies;and balance
_)
drums in both reactor feedwater pumps, i
- 10) Preventive maintenance on all 8 safety / relief valves'. I
- 11) Replacement' of- 3 of 8 safety / relief valve discharge line
[ ramsheads with T-quenchers. ;
- 12) Installation'of 8-inch vacuum breakers on safety / relief. ]
J. valve discharge lines.
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- 13) Replacement-of feedwater low flow control valve.CV
! 6-13 with-a drag-type valve._- i l
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- 14) Replacement of HPCI turbine steam isolation valve 9/9/77 (Cont'd.) FO-2036,
- 15) Leak and eddy current testing of main condenser and feedwater heater tubes. Plugging of leaking and suspect tubes.
- 16) Inspection of turbine front standard, #3 and #4 stop and control valves, #2 and #4 CIV's and moisture separators.
- 17) Inspection of generator and exciter.
- 18) hbdification of generator phase connection blocking.
- 19) Eddy current testing of hydrogen, exciter, stator and lube oil coolers.
- 20) Miscellaneous electrical inspections and maintenance.
- 21) Installation of new 480V load center.
- 22) Installation and maintenance of instrumentation in torus and drywell for relief valve discharge T-Quencher Tests.
- 23) Repair painting of drywell and torus interior skin.
- 24) Dredging of intake structure.
- 25) hbdification of torus vent header supports to increase their load capability.
- 26) Installation of two additional condenrer vacuum sensing lines to provide separation for condenser low vacuum scram sensors.
- 27) Machining of reactor vessel feedwater noz:les to re-move stainless steel cladding and provide machined surfaces for themal sleeves.
- 29) Inspection of 8 x 8 surveillance fuel bundle, re-constitution of the segmented test bundle, and in-spection of selected fuel channels.
- 30) Installation of modified controls for turbine stop valve and control valve testing.
- 31) Completed installation of CW Pump flood protection trip circuitry.
- 32) Overhaul of both condensate pumps.
- 33) Miscellaneous maintenance and minor modifications.
9/11/77 During inspection of the drywell, an elbow and U-bolt of the "F" safety / relief valve discharge line were found damaged due to inadequate restraint. The elbow and U-bolt were replaced and an additional support was installed to reduce displacement of the line (M-RO-77-26) .
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9/13/77 Results of calculations using recently approved model changes were found to require slightly more restrictive fuel thermal limits at some exposures. O!-RO-77-25).-
A license amendment request was submitted to the NRC to revise the Technical Specifications.
9/23/77 Upon disassembly of RIR torus cooling valve, FD-2008, for stem replacement, the valve body seat ring threads were found to be stripped. Previous problems with shearing of the valve stem clamp set screws allowed the plug to rotate during attempts to operate the valve. Rota-tion of the plug unscrewed the seat which caused the threads to strip. The valve body was modified using a holding ring to position and secure the seat ring in place.
(M-RO 27) .
9/25/77 Core reloading was completed and preparations for
- feedwater no
- z'e work began.
9/26/77 A weekly IRM Rod Block Test was found to have not been performed. The surveillance file was revised to ensure 4
identification of testing requirements associated with special plant conditions. (M-RO 28) ,
10/2/77 Three main steam line area temperature switches were found to trip slightly above the Technical Specification allowable setting. The switches were recalibrated.
1 @!-RO 29) .
10/10/77 During shutdown cooling operation, a fatigue crack was discovered on a "B" PJR loop relief valve 2-inch boss connection. The crack was ground out and the boss replaced with a weldolet. A restraint was installed.
@t-R0 30) ~.
4 10/13/77 Inspection of the torus internal catwalk revealed in-correct welded and bolted support attachment. All attachments were corrected to meet original construc-tion requirements.- @f-RO-77-31) 10/30/77 During a routine surveillance test the setpoint of a i main steam line low pressure isolation switch was found to have drifted lower than allowed by Technical Specifi-cations. The switch was recalibrated. Of-RO-77-32),
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10/31/77 Completed all CRD nozzle work.
During a surveillance test, the automatic transfer ,
circuitry for Emergency Bus #15 did not function. Im-proper wiring of a new 480V load center, had resulted in a short circuit through the control fuses for the transfer circuit. The wiring was corrected and the fuses replaced. 01-RO 33) ,
11/1/77 Completed all feedwater nozzle and sparger work.
11/5/77 Completed reactor coolant leakage test.
11/7/77 Completed Type-"N' primary containment integrated leak rate test.
11/10/77 Returned to power operation and increased power to to 100% of rated.
11/15/77 11/16/77 Operated at 100% of rated power except for brief to . weekly reductions for control rod exercising and valve 12/14/77 testing.
On 12/7/77, the secondary containment isolation dampers associated with reactor building vent supply _ unit, V-AH-4A, were found blocked by ice in the open position.
Corrosion of the preheat coils inner steam-distribution tube resulted in the stagnation and free:ing of the condensate within the coil causing the tube to rupture. The ice was thawed, the coil was repaired and the isolation dampers returned to service (M-RO-77-34),
12/15/77 Plant shutdown for scheduled outage to install and re-to pair relief valve discharge T-Quencher test instrumenta-12/16/77 tion and replace the _topworks- on "A" safety / relief valve.
12/17/77 Returned to power operation and increased power to_85%?
to of rated.
12/18/77 I-11
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! 12/19/77 Conducted safety / relief valve discharge T-Quencher tests.
to Upon conpletion of the testing, power was returned to 12/22/77 100% of rated, d
4.
12/23/77 Operated at 100% of rated power except for brief weekly
. to reductions for control rod exercising and valve testing.
12/31/77 1
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II NUMBER OF PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION NUMBER OF PERSONNEL ( E 100 mrem) TOTAL MAN-REM _
CONTRACT CONTRACT STATION UTILITY WORKERS AND STATION UTILITY WORKERS AND-WORK & JOB FUNCTION EMPLOYEES EMPLOYEES OTHERS EMPLOYEES EMPLOYEES 'OTHERS
_ REACTOR OPERATIONS & SURVEILLANCE OPERAf1NG PERSONNEL 35 0 0 65.773 0.000 0.000 HEALTil PHYSICS PERSONNEL 8 0 0 15.415 0.000 0.000 SUPERVISORY & ENGINEERING PERSONNEL 28 11 19 19.803 6.662 6.973 INSTRUMENT & CONTROLS PERSONNEL 7 0 0 8.056 0.000 0.000 ROUTINE MAINTENANCE MAINTENANCE PERSONNEL 31 57 1 51.963 28.871 0.220
- INSERVICE INSPECTION HEALTH PHYSICS PERSONNEL 0 0 1 0.000 0.000 0.447 SUPERVISORY & ENGINEERING PERSONNEL 0 2 3 0.000 0.515 3.766 OPERATING PERSONNEL 0 0 16 0.000 0.000 18.567 0SPECIAL MAINTENANCE MAINTENANCE PERSONNEL 26 68 364 27.805 57.127 516.881 11EALTH PHYSICS PERSONNE7 4 0 26 1.760 0.000 31.365 INSTRUMENT & CONTROLS PERSONNEL 7 1 8 7.728 0.309 9.138
[ WASTE PROCESSING A MAINTENANCE PERSONNEL 11 0 0 5.157 0.000 0.000 OPERATING PERSONNEL 10 0 5 3.977 0.000 6.364 SUPERVISORY & ENGINEERING PERSONNEL 0 0 0 0.000 0.000 0.000 REFUELING MAINTENANCE PERSONNEL 5 11 2 0.711 1.642 0.238 OPERATING PERSONNEL 20 0 3 4.161 0.000 0.390 HEALTH PHYSICS PERSONNEL 0 0 6 0.000 0.000 1.371 SUPERVISORY & ENGINEERING PERSONNEL 0 0 6. 0.000 0.000 1.769-SECURITY 0 0 9 0.000 0.000 2.020 00 TOTAL MAINTENANCE PERSONNEL 73 136 367 85.636 87.640 517.339
{ OPERATING PERSONNEL 65 0 24 73.910 0.000 25.321 i
HEALTH PHYSICS PERSONNEL 12. 0 33 -17.175 0.000 33.184 I SUPERVISORY & ENGINEERING PERSONNEL' -28 13 28 19.803 7.178 12.508 I 14 15.784 i
INSTRUMENT & CONTROL PERSONNEL ' 1. 8 0.309 9.138 SECURITY O O 9 0.000 0.000 2.020 GRAND TOTAL: 192 150 , 469 212.308 95.127 i 599.510 j ODESCRIPTION: _1. Maintenance Performed in Primary Containment During Shutdown s
- 2. Feedwater Sparger Modification 3.-' Torus Modifications "4. . Reactor Building Crane Modifications OOINDIVIDUALS MAY BE LISTED UNDER MORE THAN ONE WORK AND JOB FUNCTION.
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III. Cd M ES, TESTS AND EXPERIMENTS The following sections include a brief description and a sumary of the safety evaluation for those changes, tests and experiments which were carried out without prior NRC approval, pursuant to the requirements of 10CFR50.59(b).
1 REACTOR IUILDING MAIN STEA\f CHASE EX11AUST FANS (ADDENDLBf #2
.c SRI 111)
Description of Change A manual transfer of operation for main steam chase exhaust fans V-EF-24A and V-EF-24B was installed. The automatic e
transfer logic previously installed under SRI 111 was abandoned in place.
The original design provided for autoraatic switching to the alternate fan if failure occurred in the operating fan.
Since that concept was presented,however, it was established
' that imediate fan switching is not necessary, and that the time availabic witha manual switching schene is more than adequate for continued plant operation.
Summary of Safety Evaluation Interlocks with GGTS and reactor building supply fans and Reactor building isolation capabilityare not affected
- by the modification.
- 2. INSTALLATION OF PILOT INLET FILTERS ON SAFETY / RELIEF VALVES T$ill 174)
Description of Change Pilot inlet filters were installed on the 8 Target Rock safety / relief valves to maintain cleanliness of the pilot
, valves.
Starnary of Safety Evaluation The pilot filters had been removed previously because of concern that the. filters could cause degradation of valve response time. 1-bwever, drain and-vent groove modifications -
to the valve eliminate the potential delay time problem.
Testing has shown that a substantial amount of buildup in the filter does not affect valve performance.
III-1.
a a _
_= _
- 3. DELETE ANNJAL REPIACHENT REQUIRBENT FOR THE STACK FILTER (SRI 175)
Description of Change The requirment for annual replacment.of the IEPA stack filter has been deleted. The Stack IIEPA filters will be
, retained in service as long as annual DOP testing verifies that they meet the filter efficiency testing requirements, the filter pressure drop limit is not exceeded and the recmbiner system is not bypassed. In the event the recomb-iner system is bypassed the filter will be replaced after one year of service.
Sumury of Safety Evaluation FSAR Section 9.3.3.3 states that the annual filter unit replacement is based on the activity that would be released from an explosioa in the off-gas filter after one year operation at the stack release rate of 100,000 uci/sec (after 30 minutes delay), which would result in 10% of the filter's activity being released to the environment. With the incorp-oration of the modified off-gas system, the hydrogen and oxygen (source of explosions) has been removed from the 30 minute delay pipe and the filter. Additionally, the off-gas is now filtered through 2 charcoal and 2 IEPA units prior to being cmpressed and stored in tanks before release.
- 4. LOAD CINfER ADDITION (75M094)
Description of Change The new 430V load center was installed to reduce the loads carried by load centers B12, B14 and B23 and to provide capability for future requirments.
Sumnary of Safety-Evaluation Only the non-essential loads of B12, B14 and B23 were trans-ferred to the new load center. The new load center has an alternate supply from load center B2 which is supplied from the 4160 volt bus #14.
- 5. TORUS MODIFICATIONS (7G1052)
Description of Change The torus-to-support column connections were reinforced by adding weld metal to the existing web plate-to-shell weld, the lower wing plate-to-shell weld, the upper wing plate-to-shell weld, and the vertical stiffner-to-lower wing plate weld. Also, one
, inch parallel reinforcing plates were added on each side of the III-2 i
web plate.
Sumary of Safety Evaluation The modification was in accordance with ASNE Code,Section III (through the Winter of 1975 Addenda).
This modification increased the load carrying capabi-lity of the torus support structure.
- 6. RELOCATE IIPCI GIAND CONDENSER RESTRICTING ORIFICE (77 M 012)
Description of C}mnge The HPCI gland seal condenser restricting orifice, R0-2058, was moved from downstream to upstream of the gland seal condenser. The purpose of the change is to reduce the pressure on the condenser during HPCI system startup and thereby minimize the potential for gasket extrusion from the shell-bonnet interface.
Sumary of Safety Evaluation System operating characteristics were not changed by this modification.
7, MODIFY TORJS VENT HEADER RJPPORTS (77 M 017)
Descriptien of Change The torus vent header support connections were reinforced by replacing the existing pins with high strength pins.
The upper connection reinforcement clevis plates, with spacer plates, were bolted to the vent header collar and pinned to the support column. Thi: modification increased the load carrying capacity of the s.ipport connections.
Swmiary of Safety Evaluation This modification was perfomed in accordance with ASE Code,Section III, Subsection NF. The FSAR references the AISC Code as the applicable construction code. The
' ASbE Code was developed from the AISC Code, therefore, an updated version of the same code was used.
- 8. REACTOR IUILDING CRANE (77Z019)
DescIiption of Change The originally installed. reactor building crane trolley was rep 3 aced with a single-failure proof (redundant) trolley to III-3 4
rMuce the probability of dropping a fuel shipping cask or other heavy load. The new trolley incorporates an 85 ton capacity dual redundant configuration main hoist and a 5 ton capacity conventional auxiliary hoist. The electr-ical controls were also modified to minimize single failure vulnerability.
Sumary of Safety Evaluation The structural and mechanical caponents were designed to meet the requirments of the governing codes aM standards.
All critical load bearing cmponents of the main hoist have a minint:m safety factor of 5.
- 9. INSTALIATION OF FEEDWATER CONTROL VALVE DIFFERENTIAL PRESSURE TAPS (77 M 021)
Description of Change Pressure taps were installed on existing drain lines located upstream aM downstream of the "A" Feedwater Control Valve to permit measurement of valve differential pressure.
Summary of Safety-Evaluation
! The modification was performed in accordance with the original design code, ANSI B.31.1, Power Piping.
10.
LOAD MITIGATING SPARGERS (77 M 0411 Description of Chance The rams heads on the A, E and G safet;/ relief valve discharge lines were replaced with load mitigating spargers.
Summry of Safety Evaluation Aspects of this mMification which could conceivably affect the probability or consequences of an accident or malfunction previously analyzed were evaluated. The quencher is designed to result in an acceptable pressure drop and thereby climin-ate feedback to the safety relief valve. Neither the safety relief valvenor nuclear steam supply system are affected by the modification.
Effects of the quencher on the piping has been specifically accounted for in the quencher and support design The loads on containment with the existing ramshead _were measured during relief valve test at Monticello during June, 1976, and the structural adequacy of the containment was demonstrated.
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- 11. GEAR RATIO CliAM3E m RCIC OtRBOARD ISOLATION VALVE, Ffo-2076 (77 M 043) 5 i
The 50 to I wom and wom wheel of the Limitorque 3!B-000 motor o?crator on MO-2076 were replaced ~ by a wom and worm wheel having a 68 to 1 gear ratio due to wear observed on the original worm an:1 wom wheel.
Sumrnry of Safety Evaluation The modification results in slower-valve operating speed,
] but closing time remains within established limits. Valve >
c mponent stresses as a result of this modification are 4
acceptable.
i i 12. RELIEF VALVE SOLENGID PLATES (77 M 046) i Descriptian of Change i
2 The leaking bellows test solencid valves and air actuator
-solenoid valve for each safety / relief valve were relocated fra a direct mounting on the valve to a plate which was
- j located near the valve. The purpose of the modification is i to simplify maintenanco perfomed on the valves.- The.
] Pronpt Relief Trip solenoid valves were renoved from the j
solenoid valve group at this time.
Summary of_ Safety Evaluation I
!- Relocation of the solenoid valves did not affect relief -
i valve operation or bellows testability..' At'the present time there are no. plans for using the Pranpt Relief Trip '
System.
- -13.
CYCLE 6 RELOAD (77 M 050)
, Description of Change
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The Monticello reactor core was changed .for operation in 4
CYCLE 6 by removing 132 fuel assenblies and replacing them with a_ like number having 2.62 w/o enrichment. The-CYCLE 6 I
core configuration includes 100 MTE's (Reload 2), 48 GBil's (Reload 3), 204 IJ's (Reload 4), and 132 IJ's (Reload 5.
All of the new fuel assemblies have: finger springs.
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. Sumary of Safety Evaluation -
The Reload 5 Licensing Amendment prepared by General Electric Company NEDO-24032, contains sections which consider the III-5 l
mechanical design, nuclear characteristics, thermal-hydraulic analysis and safety analysis pertinent to the Reload 5 fuel added to the Monticello core for O'CLE 6 operation.
- 14. STR CYCLE 6 CHANGES (77 M 051)
Description of Change Thirteen (13) irradiated segmented fuel rods were removed-from the SIR II fuel bundle, MTB 001. Seven (7) of these rods were replaced by unirradiated segmented rods contain-ing a total of 28 segments; six (6) of the thirteen were replaced with rods having two (2) unirradiated segments eadt, or a total of 12. Thus, 40 new, unirradiated segments were added to the STR bundle.
Sumnary of Safety Evaluation The safety evaluation contained in GE document NEDE-20179 with sections concerning Results from Design Evaluations, Fuel Operating and Developnent Experience, Nuclear Characteristics, and Safety Analysis for the reconstituted STR bundle, indi-cates that the changes made to MTB 001 should have no effect on the ability to operate the bundle and the core within all applicabic thermal limits and safety considerations for CYCLE 6 operations.
- 15. RD40/AL OF SCURCES (77 M 056)
Description of Change The four neutron sourteholders with sources were removed i
from the core at the end of CYCLE 5. (NOTE: The center core source had been removed in 1973, as previously reported.)
l This was done in conjunction with GE recommendations to preclude degradation of the sourceholders resulting from neutron embrittlement. The sourceholders and sources were not replaced for CYCLE 6 operation.
Summary of Safety Evaluation For core average exposure above 8000 MWD /SIU, sufficient neutrons are produced by fission product poisons to provide the required SRM countrate during shutdown. .BOC-6 exposure was 8248 MWD /SIU and is expected to be similar at the beginning of subsequent equilibrium cycles.
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- 16. _RU40/AL T REACf0R VESSEL FEEDWATER NO22LE CLADDING AND INSTMJATION OF SPARGERS (77 M 063) i Description of Changes Stainless steel cladding, along with fatigued base metal, wus removo:1 fium the reactor vessel feedwater no :les by machining. Spargers incorporating a piston ring seal thermal sleeve were installed. These modifications were made to reduce the probability for feedwater no::le crack initiation.
- Summary of Safety Evaluation The modification was perfonned in accordance with require-ments of AS4E Code,Section III,1974 Edition with Addenda through Summer,1976 and Section XI Edition with addenda through Saner,1975
- 17. _CRD REIURN RWTED TO Rhw (77 !! 065)
Description of Change To eliminate the possibility of cracking in the stainless steel CRD hydraulic return line, the line was isolated and re-routed to the Rhw return line downstream of the last motor operated isolation valve. The new return line is constructed completely of carbon steel which is resistant to the type of cracking stainless steel is susceptable to. The isolation is maintained by two manually-operated stop valves.
Summary of Safety Evaluation Operation of the CRD hydraulic system is in no way degraded by isolation of the return line. A special test proved that isolation did not affect any significant operating parameters,
" including normal rod movement, stall flow, settle tiJne and pressure, exhaust water pressure and charging water pressure.
Scram insertion time was also unaffected. The new return line was installed and tested in accordance with the applicable codes.
If the CRD pumps are required to provide coolant to the vessel, the two manual stop valves can be opened.
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CAP CRD REIURN LINE N022LE (77 M 069)
Description of Change To eliminate cracking of the CRD return line RPV no::le, the no :le was capped during the refueling outage. The drywell penetration was also capped and all return line III-7
piping in the drywell was removed. . The reactor nozzle cap ir four-inch Sch.120 ASDI-A-182, Gr. 316 with a 0.02 percent carbon maximum, while the containment vessel cap is six-inch Sch. 80 AS111-A-350, Gr. LFL
&mrnry of Safety Evaluation These modifications do not create the possibility of a new accident, increase the probability or consequences of a previously analyzed accident or decrease the margin of safety for any Technical Specification. The caps were installed and tested in accordance with the applicable codes.
- 19. REPLACRM OF DRYh' ELL CBl/CANI WITH A PARTIOJLATE CA\1 (77 M 096)
Description of Change The drywell CD!/C#1 was replaced by a particulate CA\l.
This was done because of unreliable operation of the drywell CDl/CAh!, The CAsl uses the same sample inlet and discharge connections that the CBf/CA\1 used.
&nmary of Safety Evaluation The replacmient of the drywell CBi/CA\1 with a particulate CANI did not create any new potential accident considerations.
Iodine monitoring of the Drywell atmosphere is not required or deemed necessary.
- 20. RELOAD FUEL ASSBELY (SS) IJ 3736 FOR CYCLE 5 OPERATION (75 M 087)
Semi-Annual Operating Report Number 10 reported that fuel bundic IJ 3736 contained 14 tabs on the water / spacer capture i rod rather than the customary 7 tabs. Visual observation _in l
Septonber,1977, confirmed. that LI 3736 does in fact contain only 7 tabs, r
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