ML20125B243
| ML20125B243 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/31/1973 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Oleary J NRC, US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9212090270 | |
| Download: ML20125B243 (23) | |
Text
, . e e
gulatorf D ocket File i g NORTHERH STATES POWER COMPANY MIN N E A PO Ll S. MIN N E SOT A 55401 W IJJJ
/3 // P 7 February 28, 1974 k,'gN:' ("QD
[
0L1J' i 7I]
i N
3
, ^ 3
. : APR16197%
AFP 161974> 9 g u .. .. , m,n cu 4u
, u. s II:,#s g g Mr. J F O ' Le a ry ,
C,x, ;.pd. S g NH Stens g Office of Regulation Directorate s-of Licensing'h'1 .br~g[I n NT#
United States Atomic Energp Coh61hs Washington, DC 20545
Dear Mr. O' Leary:
FDNTICELIO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Six-Month Operating Report #6 July - December, 1973 In accordance with Section 6.7.A.2 of Appendix A, Technical Specifica-tions, of Provisional Operating License DPR-22, enclosed are three copies of the subject operating report. The other required 37 copies vill be transmitted as a bulk shipment.
Yours very truly, h.
L 0 Mayer, PE Director of Nuclear Support Services IDM/1h cc: J G Keppler G Charnoff Minnesota Pollution Control Agency Attn. E A Pryzina enclosure c.n, v c c -1 9212090270 731231 PDR ADOCK 05000263 *.
R PDR s
3 TAM.E OF CONTENTS
- 1 Narrative Sumary of Operating Experience II Occitpational Exposure Report III Changes, Tests and Experitnents O
t 4 9
- 1. NARRATIVE StM48Rf 0F OPERATING EXPERII?CE 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.
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 stem threads, and damage to the packing indicated igroper packing installation.
Instructions were issued conceming packing adjustments andreplaced ing accumulator on 1 op/18/77 (Reportable Occurrence No.erability was M-RO~-77-01).
On 1/25/77 it was noted that the SBGTS flow rate had not been recorded monthly _in cog liance with Techni-
- cal Specifications which were issued 9/27/76. The test procedure waa revised to provide for a monthly flow record and the- administrative procedure for implementing 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. Le valve was tapped and it closed, ne 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).
2/23/77 A reactor scram occurred when a load rejection caused a turbine control valve fast closure. We load re-jection resulted when a ice stonn 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 temerature 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. The observed
' period was consistent with core analysis data (Report-able Occurrence No, M-RO-77-04). Following temera-ture reduction and xenon decay, plant restart was
,, initiated.
I-1
+
2/23/77 (Cont'd) Administrative and Operating Procedures were subsequent-
. ly revised and new core analysis procedures were in-stituted to identify hi gh reactivity worth notches, place restrictions on their withdrawal, and clarify administrative pmcodures pertaining to anomalous re-activity changes.
2/24/77 to Power was gladually 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 tonts IcVel indicators. Investiga-tion revealed that the actual toms water volume was slightly below the minimum Technical Specification limit.
Water volunc was returned to the normal and the failed transmitter was replaced (Reportabic Occurrence No, M-RO-77-05).
3/18/77 Conrnenced 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 /
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 valvos.
3/21/77 Returned to power operation and increased t ower to to 100% of rated.
3/25/77 On 3/21/77 the air-ejector sanple system was found isolated, such that both air-ejector radir. tion nonitors had been inoperable during startup. Operation of the nonitors was re-established. 'Ihe work coatrol process and startup procedures were revised to pievent a re-j currence. (Reportable Occurrence No. M-TO-77-06).
I-2
7
)
f I
3/26/77 Operated at 100% of rated power except fcr brief weekly to reductions for control md exercising ;ed valve testing.
4/11/77 .
On 4/4/77 during a routine surveillance test, the set- !
point of one of the HPCI steam line ates tenperature >
switches was found to have drifted above the allow-able Technical Specification limit. ne switch was replaced (Reportable occurrence No. M-PD-77-07).
4/12/77 Pwer was reduced to 66% of rated for load following.
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 4/19/77 Following routine maintenance on #12 Reactor Protection ,
}G Set, it was started and a transfer of load from the altemate source to the MG Set was atterpted, initiating an expected Channel B half scram. The FC 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 high flux on APRM #2. When the channel B power range neutron monitors were de-energized during the power source transfer, the shared LPMI inputs from APBf #6 were automatically removed from the APRM #2 averaging circuit. As a result of this action the LPRM average input increased, causing APRM #2 indication to increase above the scram setting. The MG Set transfer was com-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 minimize the affect on APRMs is in progress.
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 to reductions for control rod exercising and valve testing.
6/2/77 On 5/30/77, during the monthly RIR Motor Operated Valve Operability Test, "B" MIR Injection Valve bO-2013 failed to open. A line motor control center "open" contactor did not close as required when given the valve "open" signal causing the motor starter control power fuse to open. Re-fuse was replaced, the motor starter cleaned, and pro 77-09).per stroking and operation was demonstrated. (M-RO-1- 3_
4/23/77 On 5/31/77, during the monthly RCIC Fbtor Operated Yalve
! to Operability Test RCIC Outboard Steam Supply Isolation 6/2/77 (Cont'd) Valve}O-2076falledtoclose. ne main and limit switch gear train grease had deteriorated due to high ambient tenperatures. Worn gears were replaced, the gear trains
~
were cleaned and greased and valve operability was demonstrated. A ventilation modification has significantly lowered the ambient temperature. (WRO-77-09).
6/3/77 power was reduced to 921 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 perform operator licensing demon-
, to strations,'a CRD hydraulic return line isolation test 6/12/77 and the following maintenance:
- a. Plugged leaking tube in high pressure feedwater heater 14A.
- b. Replace outboard seals on reactor feedwater pum #12.
- c. Repacked miscellaneous valves.
6/13/77 Returned to power Operation. Increased power to 981 to of rated. Power was limited to 98% of rated due to low 6/19/77 feedwater pum 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 igroperly adjusted, such that the margin for normal deterioration was less than desired. The switches were subsequently properly adjusted and administrative procedures revised to clarify and inprove control over such work. (W RO- 77-10) .
On 6/17/77 a small Icak was discovered in a welded joint on the 1" drain line connected to the "C" moisture sep-arator drain line. He 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).
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 infomed by General to Electric Co. of an inappmpriate assunption used in the 6/23/77 (Cont'd) detemination of Cycle 5 !CPR limits. A conservative reanalysis increased the transient delta-CPR for all fuel by 0.08 At the tine of the occurrence, the reactor was operating within the new limits (M-RO-77-12).
On 6/23/77, the flow through Standby Gas Treatnent System was found to be below Technical Saccification requirenents.
An isolation damper was found to be operating inproperly due to a combination of nomal wear and inpmper installa-tion of the air supply to the danper control. Die valve controls were corrected and proper flow verified.
(M 77- 13) .
6/24/77 Power was reduced to 58% of rated due to high vibration on #11 Reactor Feedwater Pu p (RIT).
6/25/77 Scheduled outage to repair feed punp recirculation valves to J-3489 and CV-3490 and #11 RFP. Inspection of the ptmp 6/26/77 evealed damage to various conponents, including a severely rubbed shaft due to the suction flow guide being dis-11 aced, wiped inboard and outboard journal bearings, aroken discharge and first stage diffuser capscrews, first stage diffuser vane to sideplate weld cracking, a broken first stage diffuser vano and a damaged inpeller.
6/27/77 Returned to power operation. Power limited to 62% of to rated pending conpletion of repair of #11 RFP.
6/30/77 On 6/27/77, during plant startup, the indication of the air ejector off gas radiation nonitors was found to be abnormally low due to air leakage into the sanpling system caused by an inproperly open manual valve. 'lhe inproper 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.
I-5
_ _ _ . _ _ __ _ _ . . _ _ _ _ ~ _ _ _ _ _ __ _
l 1
, 7/5/77 On 7/7/77, during routine operator inspection, a steam l
, to leak was observed at the 1-inch Icak test connection to ;
7/15/77 (Cont'd) the RCIC steam supply line drain pot drain line. In- i spection revealed poor quality of a welded joint. 'Ihe l 1eak was tenporarily repaired with a clam? and pack-ing. During the 1977 refueling outage, tie faulty sec- i
. tion of pipe and weld was replaced. (M-RO-77-15). 1 On 7/9/77, the "A" recorbiner train off-gas flow control valve (ICV-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 acetrulation. (M-RO-77-16).
On 7/11/77, following maintenance of the acctrulator on CRD llCU 26-23, the nitrogen charging valve would not hold pressure. Inspection revealed that the valve was not properly seated in the instrumentation block, allow-ing nitrogen to leak past the seat and body 0-rings.
'Ihe 0-rings were replaced and the valve was seated properly.
(M-RO 17) .
7/16/77 Reduced power to 56% of rated for weekly control rod to exercising, valve testing and control rod pattern ad-7/19/77 j ustments. Gradually returned power to 100% of rated.
On 7/19/77, the nitrogen charging valve on CRD liCU 18-11 would not hold pressure. The nitrogen leaked past the valve stem as a result of defective packing, which had failed due to natural end-of-life. The valve was replaced.
(M-RO- 77-18) .
7/20/77 Operated at 100% of rated power except for brief re-to ductions for control rod exercising, valve testing, and 8/12/77 load following.
On 8/2/77, a small steam leak was discovered in a 45-degree elbow on the IIPCI steam supply drain line to the condenser.
Investigation revealed that a steam trap upstream of the cibow had failed causing crosion of the elbow. The leak was temporarily repaired with a clanp. During the 1977 refueling outage, the trap was repaired and the cibow was replaced. (M-R0- 77-19) .
On 8/5/77, during the daily HPCI auxiliary oil ptmp test, a resistor in the IIPCI governor control system failed resulting in a loss-of DC power to the system. 'Ihe resis-tor was replaced with an equivalent adjustable resistor.
(M-RO-77-20).
1-6
l
!. l 7/20/77 On 8/10/77, the nitrogen charging val've on CRD HCU 30-11 to would not hold pressure, ne nitrogen leaked past the 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 was replaced in all 121 accumulator charging valves (M-R0-77-21).
j i:
i 8/13/77 Pwer was reduced to 54% of rated for control rod exer-i to cising, valve testing and to adjust the control rod 8/18/77 pattern. Gradually increased power to 100% of rated, i
i l 8/19/77 Operated at 100% of- rated power except for brief reduc-j to tions for load following, control rod exercising and w
8/23/77 valve testing.
1 4
1 1 8/24/77 The recombiner system steam sup31y. valve failed closed to due to low air pressure caused by a Icaking solenoid j 8/25/77 valve on the condensate domineralizer system and a j partially plugged air filter. Reactor power was immedi-ately reduced to 45% of rated. The valve was reopened i 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 tenperature j were low. - Also, the SJAE off-gas radiation monitor L readings were gradually increasing. -A thorough investi---
gation led to the conclusion that recombination was occurr-t ing at the air ejector after condensers, Normal 03eration was restored -by shutting off the off gas flow to tie air
!- ejectors for a short time. It is believed that the trip i
of the reconbiner steam supply valve allowed flame propa-j 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.
4 i
8/27/77 Electrical noise caused by a lightning storm initiated a
trip of the recombiner trains. Power was redtcod to 46%
of rated until the recombiners were returned to operation
-and condenser vacuum was increased to normal. Power was
, then increased to 97% of- rated.
i 1
17-a - - -, ... -
i o
]
]
i
, 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- '
I 9/8/77 ing.
I On 8/28/77, during the monthly MR Hotor Operated Valve l l . Operability Test, "B" MR Injection Valve (W2013) failed l 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. Ris, in con- i j junction with a slightly worn contactor, prevented operation, i ne contactor was cleaned and repaired, and control relays
] were installed to reduce control wire voltage loss. (M-h77-22) 1 On 9/3/77 all control rods were fully withdrawn.
! 9/9/77 Comenced scheduled outage to refuel the reactor and per-form plant inspections, modifications and maintenance, i During operation of the MR System, PJR toms cooling valve, j- FO-2009, failed to operate properly. Inspection revealed that the stem clamp set screws had sheared allowing the stem i to rotate. A modified stem clamp utilizing a keyway was in-stalled. (M-RO 23) ,
In addition to refueling i
i outage (9/9/77 to 11/10/7Y) major included items the accomplished during following:
i i 1) Type "A", "B" and "C" containment leak rate testing, j 2) In-service inspection activities.
l 3) Surveillance tests and inspections.
I 4) CRD maintenance, including dye penetrant examination
- of collet housings.
j 5) Repair and repacking of miscellaneous valves.
i 6) Replacement of 10 LPRM strings.
p 7) Removal.of 4 neutron sources and source holders from the i Core.
- 8) - Capping of CRD hydraulic return line reactor vessel nozzle and drywell penetration and rerouting of return line to Reactor Water Cleanup System.
- 9) Installation of modified barrel assemblies' and balance drums in both reactor feedwater punps.-
- 10) Preventive maintenance on all 8 safety / relief valves.
,, 11) Replacement of 3 of 8 safety / relief valve discharge _line
- ramsheads with T-quenchers.-
- 12) Installation of 8-inch vacutzn breakers on safety / relief valve discharge lines.
'. 13) Replacement of feedwater low flow control-valve-CV.
6-13 with a drag-type valve. -
i
~
a
9/9/77 (Cont'd.) 14) P c ment of HPCI turbine steam isolation valve p
- 15) Leak and eddy current testing of main condenser and feedwater heater tubes. Plugging of leaking and suspect tubes.
Inspection of turbine front standard, #3 and #4 stop 16) and control valves, #2 and #4 CIV's and moisture separators.
- 17) Inspection of generator and exciter.
- 18) bbdification 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.
2, 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) bbdification of torus vent header supports to increase
, their load capability.
- 26) Installation of two additional condenser vacuum
, sensing lines to provide separation for condenser low vacuum scram sensors.
- 27) Machining of reactor vessel feedwater nozzles 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 Sbolt of the "F" safety / relief valve discharge line were fotad damaged due to inadequate restraint. The cibow and U-bolt were replaced and an additional support was installed to reduce displacement of the line (M-RO-77-26).
I-9
4 9/13/77 Results of calculations using recently approved model changes were found to require slightly more restrictive fuel thermal limits at some exposures. Of-RO 25) .
~ '
A license amendment request was submitted to the NRC to revise the Technical Specifications.
4 9/23/77 Upon disassembly of RlR toms cooling valve, hD-2008, for stem replacement the valve body seat ring threads were found to be stripped. Previous problems with shearing of the valve stem clanp set screws allowed the plug 1
to rotate during attenpts 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.
01-R0 27) .
9/25/77 Core reloading was com,leted and preparations for feedwater nozzle work began.
9/26/77 A weekly IRh1 Rod Block Test was found to have not been perfomed. 'Ihe surveillance file was revised to ensure identification of testing requirements associated with special plant conditions. Qi-RO-77-28),
10/2/77 'Ihree main steam line area tennerature switches were found to trip slightly above the Technical Specification allowable setting. The switches were recalibrated.
@t-R0-77-29).
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.
O!-RO-77-30).
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. 01-RO- 77-31) 10/30/77 During a routine surveillance test the setpoint of a main steam line low pressure isolation switch was found to have drifted lower than allowed by Technical Specifi-cations. 'Ihe switch was recalibrated. Of-RO- 77-32) .
T-10
l l
= I 1
I I
10/31/77 Cog leted 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 contml fuses for the transfer circuit. The wiring was corrected and the fuses replaced. (M-RO 33) .
11/1/77 Completed all feedwater nozzle and sparger work.
l 11/5/77 Completed reactor coolant leakage test.
11/7/77 Completed Type "A" primary containnent integrated leak rate test.
11/10/77 Returned te, power operaticn 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 freezing 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 l
1 l
I-11 l
l
12/19/77 Conducted safety /mlief valve discharge T-Quencher tests, to Upon conpletion of the testing, power was returned to e
12/22/77 100% of rated.
12/23/77 Operated at 100% of rated power exc.pt for brief weekly to reductions for control rod exetcising and valw testing.
12/31/77 5
e o
I-12
II NUMBER OF PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION NUMBER OF PERSONNEL ( E 100 mrem) TOTAL MAN-RDi CONTRACT CONTRACT STATION UTILITY WORKERS AND STATION UTILITY WORKERS AND EMPLOYEES EMPLOYEES OTHERS EMPLOYEES EMPLOYEES CrrHERS WORK & JOB FUNCTION REACTOR OPERATIONS & SURVEILLANCE 35 0 0 65.773 0.000 0.000 OPE 3ATING PERSONNEL 8 0 0 15.415 0.000 0.000 HEALTH PHYSICS PERSONNEL 28 11 19 19.803 6.662 6.973 SUPERVISORY & ENGINEERING PERSONNEL INSTRUMENT & CONTROLS PERSONNEL 7 0 0 8.056 0.000 0.000 ROUTINE MAIhTENANCE MAIhTENAhCE 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 d00 0.515 3.766 OPERATING PERSONNEL 0 0 16 0.000 0.000 18.567 CSPECIAL MAINTENANCE MAINTENANCE PERSONNEL 26 68 364 27.805 57.127 516.881 HEALTH PHYSICS PERSONNEL 4 0 26 1.760 0.000 31.365 INSTRUMENT & CONTROLS PERSONNEL 7 1 8 7.728 0.309 9.138
- [ WASTE PROCESSING MAINTENANCE PERSONNEL 11 0 0 5.157 0.000 0.000
& 3.977 0.000 6.364 OPERATING PERSONNEL 10 0 5 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 '
O.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
- TOTAL MAINTENANCE PERSONNEL 73 136 367 85.636 87.640 517.339 OPERATING PERSONNEL 65 0 24 73.910 0.000 25.321 HEALTH PHYSICS PERSONNEL 12 0 33 17.175 0.000 33.184 SUPERVISORY & ENGINEERING PERSONNEL 28 13 28 19.803 7.178 12.508 INSTRUMENT & CONIROL PERSONNEL 14 1 8 15.784 0.309 9.138 SECURITY 0 0 9 0.000 0.000 2.020 GRAND TOTAL: 192 150 469 212.308 95.127 i 599.510
- DESCRIPTION: 1. Maintenance Performed in Primary Containment During Shutdown
- 2. Feedwater Sparger Modification
- 3. Torus Modifications
- 4. Reactor Building Crane Modifications CCINDIVIDUALS MAY BE LISTED UNDER MORE THAN ONE WORK AND JOB FUNCTION.
I f- .
4
\
III. OWGES, TESTS AND EXPERIMDITS i
The following sections include a brief description and a sumary j of the safety evaluation for those changes, tests and experiments 4
which were carried out without prior NRC approval, pursuant to the l requirements of 10CFR50.59(b).
- 1. REACTOR HJIIDING MAIN STEAM CHASE E01AUST FRE (AIDENDlN #2
, to SRI 111)
Description of Change i
A manual transfer of operation for main steam chase exhaust fans V-EF-24A and V-EF-24B was installed.- The automatic transfer logic previously installed under SRI 111 was
- abandoned in place, i
The original design provided for autmatic 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
- > tiane available witha manual switching scheme is more than adequate for continued plant operation.
Sumary of Safety Evaluation
- Interlocks with TfS and reactor building supply fans i and Reactor building isolation capabilityare not affected by the modification.
- 2. INSTALLATIW OF PILOT INLET FILTERS 04 SAFETY / RELIEF VALVES
, (SRI 174)-
Description of Change- '
, Pilot inlet filters were installed on the 8 Target Rock safety / relief valves to maintain cleanliness of the pilot valves. --
- Sumary of Safety Evaluation The pilot filters had been removed previously because of concern that the filters could cause degradation of valve i response time. Ibwever, 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. I I
og III-1 i
I
, w , ~ ., .-. - n - - - - - - - - . - _- - . . -. -
i'
- 3. DELETE ANNJAL REPLACIMDR REQUIRIMINT FCR 'INE STACK FILTER l (SRI 175) i Description of Change 4
s l The requirenent for annual replacment of the IIEPA stack i
filter has been deleted. The Stack HEPA filters will be -
j retained in service as long as annual D0P testing verifies p . that they meet the filter efficiency testing requirements, i the filter pressure drop limit. is not exceeded and the -
- recebiner- system is not bypassed. In the event the recomb--
1 iner systs is bypassed the filter _ will be replaced after one year of service.
i Simnary of Safety Evaluation
[
! FSAR Section 9.3.3.3 states that the annual-filter unit l replacement is based on the activity-that would be released.
. from an explosion in.the off-gas filter after one year -
l operation at the stack release rate of 100,000 uci/sec (after i 30 minutes delay), which would result in 10% of the. filter's '
l activity being released to the environment.~- With the incorp-i oration of the modified off-gas system, the hydrogen and 5
oxygen (source of explosions) has_ been removed from the j 30 mirute delay pipe and the filter. Additionally, the off-gas is now filtered through 2 charcoal' and -2 HEPA units -
, prior to being cmpressed:and stored in tanks before release.
- 4. IDAD CINTER ADDITION (75M094)-
l ; Description.of Change-i The new 480V load center was installed to reduce-the loads carried by load ~ centers B12,-B14 and B23:and to provide-
- capability for future requirenents.
j-l- 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.
l 5. . TORUS MODIFICATIONS (781052) i
!' Description 'of Change L . 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'-
l . inch parallel reinforcing plates were added on each . side of the -
i ..
4 III-2--
I l -_-..
a._ , _ . . _ . . . . ., , . .. ~ . _ . - . _ . _ , _ . , , . - _ . . .
, web plate.
Sumary of Safety Evaluation
. The modification was in accordance with ASIE Cods,Section III (through the Winter of 1975 Addenda).
This modification increased the load carrying capabi-lity of the torus support structure.
- 6. RELOCATE HPCI GIAND CONDENSER RESTRICTING ORIFICE (77 M 012)
Description of Change The HPCI gland seal condenser restricting orifice, R0-2058, was moved from downstrem to upstrem 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. MJDIFY TORJS VENT HEADER SJPPORTS (77 M 017)
Description of Change The tonis 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. This modification increased the lod carrying capacity of the support connections.
Sunnary of Safety Evaluation This modification was perfonned in accordance with ASfE Code, Secticn III, Subsection NF. The FSAR references the AISC Code as the applicable construction code. The ASME Code was developed from the AISC Code, therefore, an updated version of the same code was used.
- 8. REACTOR BUILDING CRANE (772019)
Description of Change The originally installed, reactor building crane trolley was replaced with. a single-failure proof (redundant) trolley to III-3 l
J
- + .
s .
l , reduce 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. Be electr-
! ical controls were also modified to minimize single failure
!. , vulnerability. ,
l Sunimary of Safety Evaluation The structural and mechanical caponents were designed to meet the requirments of the governing codes and standards.
- All critical load bearing cmponents of the main hoist have
- a mininum safety factor of 5.
l-
- 9. INSTALLATI(N OF FEEDWATER CONTROL VALVE DIFFFRENTIAL PRESSURE ,
TAPS (77 M 021) i j Descripticn of Change
- + -
i Pressure taps were installed on existing drain lines-located i upstream and downstream of the "A" Feodwater Control Valve
, to permit measurement.of valve differential pressure.
Sumary of Safety Evaluation i
The modification was performed.in accordance with the original
[ design code, ANSI B.31.1, Power Piping.
}
- 10. LOAD MITIGATING SPARGERS (77 M 041)
Description of Change 1 The res heads on the A, E and'G safety / relief valve discharge
[ lines were replaced with load mitigating spargers.
j Summary of Safety Evaluation F Aspects of this modification which could conceivably affect the.
. probability or consequences of an accident or malfunction
- previously analyzed were evaluated.. The quencher.is designed 4
to result in an acceptable pressure dmp and thereby elimin-ate feedback to the' safety relief valve... Neither the safety relief valvenor nuclear stem supply systs are affected by e.
the modification.
Effects of the. quencher on the piping-has been specifically.
accounted'for in the quencher and sup: ort design. - he loads
. on contaiment with the existing ramstead were measured during i relief valve test at-Monticello during June,1976, and the t
structural adequacy oE the contalment was demonstrated.
L III-4' j.
.3 - - < .[-- -+ y - , , , ,--- ,.e ,-- ~h w...
,,,,,,--...-y . .---.my. , ,.,..-.[.w r----ve- - c w w ~ + , -w e i a-w. *.m-, - e w-- e it -*----w e (wwmev-- w r- v *'ev*
l I
- 11. GEAR RATIO OlATE W RCIC Clirl0ARD ISOLATIm VALVE, i
MO-2076 (77 M 043)
'Ihe_50 to 1 worm and wom wheel of the Limitorque
- -94B-000 motor oxrator on M0-2076 were replaced by a wom and worm wheel having a 68 to 1 gear ratic due to wear -
observed on the original worm and wom wheel.
, Stamary of Safety Evaluation i
The modification results in slower valve operating speed,
- but closing time remains within established limits. Valve cmponent stresses as a result of this modification are.
i acceptable.
) 12. RELIEF VALVE SOLENOID PLATES (77 M 046)
Descripticn of Change _-
The leaking bellows test-solenoid 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 located near the valve. The purpose of the modification is to simplify maintenance performed on the valves. The
! Prmpt Relief Trip solenoid valves were rmoved from the j
solenoid valve group at this time.
l Sumary of Safety Evaluation Relocation of the solenoid valves did not affect relief .
valve operation or bellows testability. At the present i time - there are no plans for using .the_ Prmpt Relief Trip Systm -
l l l- 13. CYCLE 6 RELOAD (77 M 050)- .
Description of Change The Monticello reactor core was changed for operation in CYCLE 6 by rmoving 132 fuel assablies and replacing them
! with a like number having 2.62 w/o enrichment. The CYCLE 6 i core configuration includes 100 MTB's (Reload 2), 48 GBIPs (Reload 3), 204 IJ's (Reload 4), and 132 IJ's (Reload 5.
All of the new fuel assablies have finger springs.
Sumary of Safety Evaluation ,
l The-Reload 5 Licensing Amendment prepared by. General Electric l Company NEDO-24032, contains sections which consider the !
. 1 1
III-5 l
I l
4 1
~
4 nochanical design, nuclear characteristics, thermal-Ir/draulic analysis and safety analysis pertinent to the Reload 5 fuel added to the Monticello core for CYCLE 6 operation.
14.
STR CYCLE 6 OMNGES (77 M 051)
Description of Change Thirteen (13) irradiated segmented fuel rods were removed from the STR 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 titirteen were 4
replaced with rods having two (2) unirradiated segments each, or a total of 12. Thus, 40 new, unirradiated segments were addo.1 to the SrR bundle.
Stanmary 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 SFR bundle, indi-cates that the changes node to MrB 001 should have no effect on the ability to operate the bundle and the core within all applicable thermal limits and safety considerations for-CYCLE 6 operations.
- 15. RIM 0/AL OF SCURCES (77 M 056)
Description of Change The four neutron sourceholders with sources were rmoved from the core at the end of CYCLE 5. (NOTE: The center core source had been removed in 1973, as previously reported.)
This was done in conjunction with GE recmmendations to preclude degradation of the sourceholders resulting from neutron embrittlenent. 'Ihe sourecholders 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 SkM countrate during shutdown. BOC-6 exposure was 8248 MND/SIU and is expected to be similar at the beginning of subsequent equilibrium cycles.
. III-6
'A i
t
- 16. RIMWAL W REACTOR VESSEL FEEDWATER N0ZZLE CIADDIE AND f ,
TNSTALIATION & SPARutES (77 M 063) 1 Description of Changes i
- Stainless steel cladding, along with fatigued base metal, 7 was removed from the reactor vessel feedwater nozzles by 3 machining. Spargers incorporating a piston ring seal thermal sleeve were installed.- These modifications were'
- made to reduce the probability for feedwater nozzle crack 4
initiation.
j Sumary of Safety Evaluation
' The modification was performed in accordance with require-ments of AS4E Code,Section III,1974 Edition with Addenda-through Summer, =1976 and Section XI Edition with addenda
- through Sumner,.1975
! 17. CRD REIURN RWTED TO RWCU (77 M 065)
Description of Change 1
!' To eliminate the possibility of cracking in the stainless j
steel CRD hydraulic return line, the line was isolated and re-l routed to the RNOJ return line downstream of the last motor operated isolation valve.. The new retum 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.
l Summary of Safety Evaluation N
- 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, sta11Lflow, settle time 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.
Descriptica of Change !
To eliminate cracking of the CRD return line RP7 nozzle, I the nozzle was capped during the refueling outage. The: !
drywell penetration was also capped and all return.line : i i
III-7 3
! i
! 0 1-i piping in the drywell was rmoved. The reactor nozzle i cap is. four-inch Sch.120 ASIN-A-182, Gr. 316 with a 0.02 percent carbon maxbntn, while the containnent vessel 4 cap is six-inch Sch. 80 ASIN-A-350, Gr. IF1.
Sunmary of Safety Evaluation l
l These modifications do not create the possibility of a new .
i accident, increase the probability or consequences of a i 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. REPLACBIENT OF DRYKELL CB1/ CAM WITH A PARTIOJIATE CAM j- (77 M 096) i-Description of Change The drywell CB1/CN4 was replaced by a particulate CAM.
. This was done'because of unreliable operation of the i
i drywell CB1/ CAM. The CM uses the same sample-inlet and discharge connections that the CB1/ CAM used.
Summary of- Safety Evaluation The replacunent of the drywell CBf/CA%1 with a particulate CM did not create any new potential accident considerations.
j Iodine monitorits of the Drywell atmosphere is not required '
i or deemed necessary, f 20. RELOAD RIEL ASSBIBLY (5S) -IJ 3736 FOR CYCLE 5 OPERATION l (75 M 087)
Smi-Annual Operating Report Number 10-reported that fuel
- bundle IJ 3736 contained 14 tabs on the water / spacer capture rod rather than the customary 7 tabs. Visual observation in l Septaber,1977, confirmed that; LI 3736 does in fact .contain-
- only 7 tabs.-
l l^
e 4
III .
.- - _ ~ . _ . . . .. ..._. .a __.x _ -.