ML20091A567
| ML20091A567 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/31/1976 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 9105160383 | |
| Download: ML20091A567 (34) | |
Text
._
NSE3 NORTHERH STATES POWER COMPANY MtN N E A POLI m, MIN N r sotA 99400
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Februa ry 28, 1977 y\\YT' 10 y
k(C VED 6 D
Hr J G Keppler, Director, Region III MAR 3 1977 g
Office of Inspection & Enforcement
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U S Nuclear Regulatory Cocnission awlsmas y
799 Roosevelt Road gp Glen Ellyn, IL 60137 g
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Dear Mr Keppler:
MONTICELLO NUCLEAR GENERATING PIMI Docket No. 50-263 License No. DPR-22 Operating Report No. 11 January 1 - December 31, 1976 In accordance with the Monticello Technical Specifications, Appendix A to Operating License DPR-22, we are submitting two copies of Operating Report No. 11 covering the period January 1 - December 31, 1976.
In accordance with our letter to Mr B C Rusche, dated December 27, 1976, the attached Annual Operating Report was formatted to meet the requirements of Technical Specification TS 6.7 A.2 and the guidance contained in Regulatory Guide 1.16, Revision 4, Paragraph C.1.b.
The section on Changes, Tests and Deperiments has been included in the 1976 Annual Operating Report as a con-venient means of meeting the annual reporting requirements of 10 CFR 50.59(b);
we may choose to report this information separately in future years.
Yours very truly, C\\
e-L 0 Mayer, PE Manager of Nuclear Support Services G
).
tcM/MHV/deh cc: Director, IE, USNRC (40) 5 Director, MIPC, USNRC (2)
G Charnoff MPCA
,N Attn:
J W Fe rman
' _ f,'
Attachment 2193
$$35*ISSS NbSh$3 R
NORT1IERN STATES poler COMPANY FDNTICEllD NUCIIAR GENERATING PLANT DOCKET NO. 50-263 LICENSE NO. DPR-22 I
t REPORT TO UNITED STATES NUCLEAR REGULATORY C0bt41SSION OPERATING REPORT NO.11 1976 r1
TABLE OF CONTINTS I.
Narrative Sumary of Operating Experience 1
i II.
Outages and Forced Power Reductions lh III.
Number of Personnel and hhn/Ren Exposure by Work and 24 1
Job Function.
IV.
Fuel Perfonnance 25 V.
Changes, Tests and Experiments 26
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1.
NARATIVE SIM MRY OF OPERATING EXPERIENCE 1/1/76 Operated at 98% of rated power.
Power ad-to ministrativelv limited to assure operation 1/2/76 within core thermal limits.
1/3/76 Operated at 100% of rated power except to for brief weekly reductions to perform 1/24/76 control rod exercising and valve testing.
1/25/76 Power reduced to 90% of rated due to high to strainer dP on the "E" condensate fi)ter de-1/26/76 mineralizer. After backwashing the strainer, rower was increased gradually, achieving 100% of rated on 1/26/76.
1/27/76 Operated at 100% of rated power except to for brief weekly reduction for control rod 2/13/76 exercising and valve testing.
2/14/76 Power reduced to 90% of rated due to high dP on the condensate domineralizer system.
1 Power was reduced further to 62% of rated to conduct control rod exercise test.
2/15/76 Following completion of testing, power was to increased to 90% of rated and held at this 2/16/76 level while the filter elements on "E" con-densate demineralizer were replaced and the strainer was backwashed.
Power was then gradually increased achieving 100% of rated on 2/16/76.
2/17/76 Operated at 100% of rated power.
to 2/18/76 e
2/19/76 Commenced power reduction in preparation for scheduled maintenance sbitdown.
2/20/76 Scheduled outage to perform the following to maintenance:
2/29/76 a.
Repair of seal on #12 reactor recircula-tion pump.
b.
Replacement of "A" Safety Relief Valve topworks.
2 c.
Repair of "C" condensate domineralizer effluent cont.01 valve.
4 d.
Repacking of miscellaneous valves inside drywell.
On 2/20/76 a leak was found on a 1/2" elbow downstream of a steam trap on the recombiner system steam line. The elbow was replaced during the outage (Reportable Occurrence No, M-RO-76-01).
Return to power operation was delayed one week due to overheating exnerienced with the #12 recirculation ptup seals. The overheating was caused by incorrect orien-tation of the stuffing box cover which allowed cooling water to bypass the seal faces.
The cover was oriented improperly due to a manufacturing error that resulted in the guide pin being improperly located.
3/1/76 Returned to power operation and increased to to 100% of rated power.
3/9/76 3/10/76 Operated at 100% of rated power except for 3
to brief weekly reductions for control rod ex-4/4/76 ercising and valve testing.
4/5/76 While perfoming the monthly RCIC surveil-lance test, the RCIC Turbine tripped due to high exhaust line pressure.
Investigation showed exhaust line set point of 25 psig was too low since RCIC exhaust pressure often spikes near the 25 psig trip setpoint during the startup transient. The set point was to increased to 40 psig (Reportable Occurrence No. M-RO-76-02).
i 4/6/76 Operated at 100% of rated power except for to brief weekly reductions for control rod 5/3/76 exercising and valve testing.
5/4/76 During a routine surveillance test, the primary containment oxygen concentration was found to be in excess of the Technical Spec-ifications limit of 5%. The oxygen concen-tration was approximately 6.5%.
The problem was caused by leakage from the drywell in-l strument air system into the containnent nitro-gen supply line via a solenoid valve. The
solenoid valve was isolated and the primary containment oxygen concentration was reduced i
j to below Technical Specification limits (Reportable Occurrence No. M-RO 76 03).
5/5/76 Operated at 100% of rated Power except for brief weekly reduction for Control Rod j
to 5/14/76 exercising and valve tests.
5/15/76 Power was reduced to 75% of rated to per-formcontrol rod exercise test.
A crack was discovered in the 3/4" vent con-noction to the 5 inch "B" moisture separator drain line. The cause was fatigue due to 3
cyclic relative motion between the 6" drain line and a 3/4" line restraint. The crack was renaired and the vent line removed on 5/23/76.
(this line can be vented at another location) (Reportable Occurrence M-RO-76-04).
5/16/76 Returned to 100% of rated power.
5/18/76 A reactor scram occurred due to an inadvertent trip of the condenser low vacuum scram switches during calibration of a pressure transmitter connected to a vacuum sensing line. Leakage past the transnitter isolation valve allowed the scram switches to trip.
Improved instrument isolation valves were installed on this transmitter and other similar transmitters sensing condenser vacuum.
The reactor was retumed to power operation.
5/19/76 Power was returned to 100% of rated.
5/20/76 The No. 3 TIP Ball Valve failed to close pro-perly durinn routine operation of the TIP System. After repeated attemnts to close the valve by nomal means, it was found that the valve could be closed by tapping on it.
The spring tension was increased on the valve and the valve was cicaned and relubri-cated. Repeated closures were demonstrated and the valve was returned to service.
I (Reportable Occurrence No. M-RO-76-05).
No.11 Cooling Tower Puno rubber ernansion joint failed, resulting in flooding of the discharge structure. The cause of failure was over-extension of the expansion joint at the time of its original installation.
.j.
The discharge stnicture was de-watered by using portable and submersible pumns. 'the electric motors were cicaned and dried. The failed rubber expansion joint was removed and replaced with a new joint. A steel spacer ring was also added to reduce the rubber ex-pansion joint extension.
5/21/76 Plant shutdown for scheduled outace to repair to pilot valve leakage on four safety / relief valves 5/23/76 in preparation for the Torus Response to Re-lief Valve Actuation Test.
5/24/76 Returned to power Operation.
Increased power to 72% of rated.
5/25/76 Power was reduced to 50% in preparation for it going to one circulating water numn operation due to indication of low river flow.
Sub-sequent investigation revealed that plant river flow indication was in error and power was returned to 68% of rated.
L 5/26/76 Operated at 68% of rated power as required dur-to ing the Torus Response to Relief Valve Actua-6/13/76 tion. Test.
During a surveillance test on 6/3/76, the HPCI steam line inboard isolation valve failed to close completely due to low torque switch set-tinn. The torque switch was time-delay by-passed until an outage on 6/15/76, at which tiine the torque switch was readjusted (Re-portabic Occurrence No, M-RO-76-06).
6/14/76 Plant Shutdown for repair of "A" Safety Relief to Valve which had failed to actuate when given 6/15/76 a manua) initiation signal. The air operator control solenoid valve plunger assembly was loose and the threads in the valve body were stripped. The faulty solenoid valve was re-placed and all similar valves were inspected.
(Reportable Occurrence No. M-RO-26-07).
6/16/76 Returned to power operation.
Increased power to 69% of rated.
5 6/17/76 Operated at 66-69% of rated power as required to during the Torus Response to Relief Valve Actua-6/18/76 tion Test (Test completed on 6/18/76).
6/19/76 Power reduced to 42% of rated and control rods were repositioned in preparation for operation at 100% of rated.
6/20/76 Reactor power was gradually increased to to 100% of rated.
6/25/76 6/26/76 Power to 1AR transfomer was lost due to lockout of substation transfomer caused by lightening strike.
Redundant power sources were available. Substation trans-fomer relay scheme was modified to prevent inadvertent trips of protective relaying due to lightening strikes (Reportable Occurrence No. M-RO-76-08).
6/27/76 Operated at 100% of rated power except for to brief weekly reductions to perfom control 7/4/76 rod exercising and valve testing. On June 28, 1976 power was reduced 5% to decrease the condensate demineralizer system differential pressure while back-washing one of the vessels.
(The flow throuch the condensate demineralizer system had increased over the past several months due to internal leaks in valves and low pressure heaters in the feedwater system.)
On June 29, 1976, the semi-annual emergency plan drill was conducted.
7/5/76 Power was reduced to 80% of rated for load following.
6-7/6/76 Operated at 100% of rated power except for to brief weekly reductions for control rod 7/17/76 exercising and valve testing.
On July lith, a high pressure feedwater heater developed an internal leak. Since the feed-water flow no::les are unstrean of the hich pressure heaters, the leakage flow is included in reactor heat balance calculations. This caused calculated reactor power to be about 1%
higher than actual power, 7/18/76 power was reduced to 51% of rated to conduct control rod exercising and special tests to locate and measure the high pressure feed-water heater leak. The leak was found to be in 14A heater. hkasured leakage flow was 68,000 lb/hr.
7/19/76 Operated at calculated 100% of rated power to (99% actual) with a brief reduction to 95%
7/25/76 of rated power to allow backwashing the i
condensate demineralizers on July 23 and 24, 1976.
While performing a routine surveillance test on July 22nd, it was discovered that the setpoint on a main steam line low pressure switch had drifted belev Technical Specification limits. The switch was reset and placed on an increased surveillance frequency (Reportable Occurrence No, M-R0-76-09).
On July 25, 1976, while performing a surveillance test, the drywell purge inlet valve, A0-2381, failed to close. Operability of valve was subsequently demonstrated. Cause for failure c,uld not be determined. The valve was placed on an increased surveillance frequency and there were no subsequent malfunctions !
(Reportable Occurrence No. M-RO-76-10).
7/26/76 power was reduced to 65% of rated to conduct control rod exercising and a special test to locate and measure leaks in the low pressure feedwater heaters.
Results of the test were l
inconclusive.
I m
. c
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7-7/27/76 Operated at Calculated 100% of rated power with a brief reduction to 95% of rated while back-to 7/31/76 washing a condensate demineralizer vessel on 7/30/76.
8/1/76 Operated at Calculated 60 - 100% of rated power to for load following, 8/3/76 On 8/2/76, during a surveillance test, the RCIC Turbine tripped on overspeed.
Investigation following this trip and a subsequent trip on 8/30/76 revealed two contributing causes:
Valve leakage from the feedwater system created and the con-a steam void in the discharge piping,ich was denser cooling water supply valve wh being maintained open pending valve repair allowed wam water circulation through the ptap. The cooling water supply valve was repaired and closed. An additional discharge valve was maintained closed to prevent formation of a steam void. The control oil system was modified to improve control valve response (Reportable Occurrence No. M-RO-76-11).
8/4/76 Scheduled plant shutdown to locate and repair to feedwater heater leaks. Tube leaks were found 8/9/76 and plugged in heaters 11B, 12A, 12B, 13A, 6 14A.
In addition the manhole cover in heater 11A (located inside the condenser) was found loose.
A diaphram was welded over the opening and the cover was reinstalled.
Other najor maintenance items accomplished during the outage included disassembiv and repair of the reactor feedptmp recirculation valves, replacement of the topworks for nain steam Safety / Relief valves A and G, and replacement of the air operator diaphragns for Safetv/ Relief valves, A, B, C and D.
On 8/5/76, a 5% reduction in M\\PUiGR limits below 85% core flow was impicmented. This reduction resulted from a generic ECCS model analysis for BWR-3's whichindicated the 2200 F peak clad temperature (regulatory limit) may be exceeded during a loss of coolant accident if operating at reduced core flow. Subsequent analysis indicated that the reduction in H\\PulGR limits should apply below 90% core flow.
(Reportable Occurrence No. M-RO-76-12).
s
.g.
8/10/76 Returned to power operation and increased power to 55% of rated.
8/11/76 Power held at 55% of rated for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> to to establish equilbrium xenon, reedwater heater 8/14/76 13A developed another tube leak. The leaking tube was plunced and then power was gradually increased to 100% of rated.
During a routine surveillance test on S/11/76, the #2 main stop valve closure scram linit switch failed to actuate. A recent adjustment had been made such that this switch failed to actuate under hot conditions. The actuating collar was modified and the limit switch was adjusted.
(Reportabic Occurrence No, M-R0-76-13).
On 8/11/76 the toms 03 concentration was not reduced below the 5% IImit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering the RUN mode as required by the Tech-nical specifications. A leaking n3 analv:er sample pump gave false indication of high 02 concentration. Therefore, all of the on-site nitrogen was used while inertine the drywell.
The sample pump was replaced, additional nitro-een was obtained and the torus was inerted approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after entering the RtN mode (Reportable Occurrence No. M-RO-76-14).
8/15/76 power was reduced to 48% of rated to conduct to control rod exercising and to repair leaking 8/17/76 eedwater pump seals.
8/18/76 Operated at 100% of rated power except for a to brief reduction for control rod exercisine and 8/27/76 valve tests. Also, on 8/24/76, a trip of No.12 reactor recirculation pump caused by a misniaced jumper during a surveillance test resulted in a brief reduction to 57% of rated.
On 8/35/76 loss of power to shutdown cooling out-board isolation valve was experienced due to failure of undervoltage relay coil. A new coil was installed. A modification to eliminate coil contacts from control circuit will be installed,~
(Reportable Occurrence No M-RC 76-15).
l e
'O 9
8/28/76 A reactor scram occurred from high neutron flux caused by a spurious reactor pressure transi-ent. The main steam pressure control system was thoroughly tested following the scram and found to operate properly.
A leak (caused by errosion) was discovered in a 1" elbow on the high pressure turbine inlet steam line U bend drain line. The elbow and a 2 foot secticn of drain piping were replaced (Reportable Occurrence No, M-RO-76-16).
8/29/76 During the restart following the 8/28/76 scram, power to the turbine auxiliary oil pump was lost due to a bus lockout relay trip. This resulted in turbine bypass valve closure, causing reactor pressure increase and a high neutron flux scram.
A plant restart was initiated.
)
8/30/76 to Power was gradually increased to 100% of rated.
9/1/76 9/2/76 The torus cooling injection valve (h0-2008) failed to open due to shearing of stem clama set screws. The stem clamp was welded to the valve stem.
(Reportable Occurrence No. ht-RO-76-17).
9/3/76 Operated at 100% of rated power except for a to brief reduction for control rod exercising and 9/5/76 valve tests.
9/6/76 A reactor scram occurred due to closure of the inboard hSIV's. The DC circuit for the valves was opened in an attempt to locate a grotmd.
The AC circuit was apparently in a tripped condition at the time. The plant was returned to power operation.
i 9/7/76 Reactor power was gradually increased to 100%
to of rated. On 9/9/76 power was reduced to 8M 9/9/76 of rated during repair of the #11 reactor feed-pump recirculation valve.
On 9/9/76 it was discovered that the volume of i;
water in the torus was below Technical Specifi-cation limits. The level had been lowered on 9/7/76 as a result of NRC desire that the plant operate close to the low limit. The volume vs level correlation did not take into account the effect of drywell-torus pressure differential.
Water volume was increased and a new vomme vs
10 l
level correlation was issued.
(Renortable Occur-rence No. 4 RO-76-18).
9/10/76 Onorated at 100% of rated noverexcept for a brief to reduction for control rod exercisinn and valve 9/13/76 tests.
The Main Steam 1.ine Radiation Monitor Punctional Test scheduled to be conpleted the week of 9/12/76 was not performed. To prevent a recurrence, a weekly surveillance test ".hecklist was pro-vided for the shift supervisors (Reportable Occurrence No. M-RO-76-23).
9/14/76 Power was reduced for a short neriod on each to dav for load followinn.
9/17/76 On 9/16/76 power was brief1v reduced to 62%
of rated to repair #12 reactor feedwater pump seal cooling line.
9/18/76 Operated at 100% of rated power excent for to brief reductions for weekly control rod 10/2/76 exercising, valve tests, and load following.
On 9/19/76 during a surveillance test, one of the PSIV closure scram relays failed to deenergize. The relay was placed in a tripped condition. The limit switch that deenergizes the relay was replaced during a later outage.
(Renortable Occurrence No. M-RO-76-19).
On 9/29/76 a crack was discovered at the connection of the No.12 reactor feedwater inboard seal cooler to the seal cooling return line. The crack was a result of vibration induced fatique. The seal cooler was isolated from the source of vibration by the installation of a flexible hose in the seal' cooling return line (Reportable Occurrence No M-RO-76-20).
10/3/76 Power was reduced to 74% of rated to conduct control rod exercising and valve testing.
The #2 Turbine Control Valve would not reopen af ter a test closure. The closing and opening time control relays were adjusted and the valve opened.
During a retest, the valve closed too rapidly, resulting in a reactor pressure increase and a high neutron flux scram.
The valve closing time was readjusted and the plant was restarted.
11 10/4/76 Reactor power was gradually increased (excent to for a brief power reduction or.10/5/76 due 10/8/76 to recombiner system problems) to 100% of rated.
10/9/76 Operated at 100% of rated nower except for a to brief reduction for control rod exercising and 10/14/76 valve tests.
On 10/10/76, the #11 Diesel Generator failed to start on the #2 starting system.
Rust particles had clogged the air relay orifice.
The starting system air control components were c1 caned and operability of the #2 start system was demonstrated (Reportable Occurrence No, M-RO-76-22),
10/15/76 Control lir.kages on two of the turbine con-trol valves broke.
Erratic operation of the control valves caused a reactor pressure in-crease and a high neutron flux scram. The linkages were found to have been weakened by bolt holes which had been irproperly located during original assembly. The control linkages were repaired and reinforced.
10/16/76 Plari remained shutdown to repack several to valves located inside primary containment.
10/18/76 10/19/76 Returned to power operation and gradually in-to creased nower to 100% of rated. The #1 TIP in-10/28/76 dexer malfunctioned. The rate of power in-crease was therefore limited due to inavail-ability of the #1 TIP machine for core monitoring.
On 10/27/76 the semi-Annual emergency plan drill was conducted.
10/29/76 Operated at 100% of rated ncxer except for r-to a brief reduction for control rod exercising 11/1/76 and valve tests.
11/2/76 Power reduced to 52% of rated to remove #11 renc-to tor feedpump from service due to unusual 11/3/76 noise and high axial vibration.
i Increased to 66% of rated power while per-forming maintenance. The coupling was replaced and the ptenp and motor were aligned.
1 12 -
11/4/76 Power was gradually increased to 100% of rated, to 11/5/76 11/6/76 Operated at 100% of rated power except for a to brief reduction for control rod exercisine 11/ 8 /76 and valve tests.
11/9/76 Power reduced to 54% of rated to remove #12 to Tr.ictor feedwater pumn from service due to high 11/10/76 v2bration.
Power was increased to 61% of rated while performing maintenance. The inboard bearing, and seals were replaced.
11/11/76 A reactor scram occurred due to a valving error while performing a surveillance test of the condenser vacuum switches.
During the shutdown, No.1 TIP machine was re-paired. The chain drive bracket had slipped out of position, allowing the chain to fall off and wedge against the indexer drive, pre-venting indexer rotation. The chain was.re-placed and the bracket repositioned and tightened.
The plant was returned to power operation.
11/12/76 Pover was gradually increased to 100% of to rated.
11/15/76
(
On 1/12/76 a steam leak caused by errosion was discovered in a 1" elbow on the H.P. turbine bypass drain line. The elbow was repair welded.
(Reportable Occurrence No, M-RO-76-24.)
11/16/76 Operated at 100% of rated power except for to brief reductions for. control rod exercising 12/13/76 and valve tests.
On 12/5/76 a temporary reduction of condenser vacuum occurred while attempting to operate with one circulating water pump. Power was reduced to 70% of rated, the second circulating water pump was started, and condenser vacutan returned to nonnal.
Power was returned to l
100% of rated on 12/6/76.
. On 11/24/76 a high pressure feedwater heater developt.d internal leakage. Allowance for the effects of the leakage on calculated power level was made and operation at 100%
of rated was continued.
12/14/76 Power was reduced to 53% of rated to remove
- 12 Reactor Feed Pump from service due to to 12/17/76 high vibration. Power was increased to 68%
of rated while perfoming maintenance on the pump.
Six of the eight 2nd stage diffuser bolts had broken, and there were cracks in the hub and vanes of the 2nd stage impeller. The bolts were replaced. The first and second stage impe11ers were interchanged.
12/18/76 Plant shutdown to repair tube leak in 14B to high pressure feedwater heater.
12/19/76 12/20/76 Returned to power operation and gradually in-creased to 100% of rated power.
to 12/24/76 12/25/76 Operated at 100% of rated power except for a brief reduction for control rod exercising to 12/31/76 and valve tests.
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4:48
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To: Date/ Time
'l3/1/76 18:20 5/18/76 23:48 253.53 8.6 i
Time Lest, Hours b,
e Cause/ System / Components
- l. Scheduled outage for repair of pilot Low Condenser Vacuum SCRV1 caused by valve for "A" Main Steam Safety Relief air leak through instrtnment valve during Valve and No.12 Reactor Recirculation surveillance test.
pump upper seal.
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Reportable tr.currences:
I None.
None LER Date/RO#.
ll 4,
More reliable instrument isolation valves Corrective tct2cn Not Applicable.
were installed on the condenser vacuum sensing lines. A design change to pro-vide condenser low vacuum scram switches 1
with independent sensing lines is planned.
b -
1.
Neutron monitoring - LPRM 04-29c:
None Major, Safety Related, Corrective 3'aintenance:
repaired under vessel connector.
System /Componcnt/Pescription 2.
Containment isolation - dPIS 2-116B: Re-pared leaking nahifold valve.
i l
4 j 3.
Safety relief valve "A", lightly Iapped main
!j poppet and installed spare top works.
4.
Reactor recirculation purp #12, replaced pump seals.
Critical Path Activity i
SCRAM Recovery and Start Up Activities.
Replace Recirc Pump Seals Single Releases or exposures.
> 10% Annual Allowable.
O.;.
- . n:r.% w u.e; ; ~.-
M?Le w
X
?rouctica
. rce cate,^ ire
-i 5/21L76 _. 23.:17 8-In: Date"T r.c
- h. 5/24/76-.-
5:5f>-
tie Dr.. $ars p__._,.54;65
!! Scheduled outage to refurbish four main steam safety / relief ratra/Syst w -am>on -nts ij valves in preparation for the " Torus Response to Relief Valve
[ Actuations" test.
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, l1 Reportahic Cccurrences:
l None.
LER Date/FOi.
q 0;
li,i
--..\\.m...
I
- - - ~ ~ *. ~ ~
i Corrective f.cticn Not applicabic.
l l
.,. ~.. -. - -. -
Major, Safer', Kelato!,
j1. Rod position infomation - RPIS Pmbe 30-23: Replace Probe.
Corrective Maintenance
, 2.
Reactor Protection "A" Yarway Column: Repaired insulation Svstr/ c:menent/!weription i
1 on column.
Containment Isolation - #3 TIP Ball Valve: Valve failed to close -
]! 3.
I Cleaned 6 adjusted spring tension.
4.
Mainsteam "II" Relief Valve bellows Leak PS:
P.S. Replaced.
Critical Path Activity
, Safety Relief Valve Maintenance 7
Single Rele.nes or e.rtx.'sures j
( None
> 10% Annual.ulowable.
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( 3 AN-1. 4 ^ED 'C OR G EC- ~ c:6 31? age X
Redi.ction, _
___t[ _ _ _ _ ___
From: Date rire t,-__.6/14f 76 23:42 T3: Date/ Time f 6/16/76 13:00 i-Time Lost. Ceurs j)i 37.30 o
C.luse/ System' Corponents
[
Scheduled outage to investigate failure of "A" main stea:,1 safety relief valve to respond to a nonnal "open" signal and to make re-pairs.
Reportable Occurrences:
July 14, 1976 / M-k0-76-07 l
LER Date/RO#.
,t V
m
. j Corrective Action Preventive ?hintenance Procedure was revised to require inspection of threaded connection between relief val re solenoid valve plunger assenbly and main solenoid valve body.
i Major, Safety Related, 1.
Safety relief valve "A", replaced air Corrective Maintenance:
operator solenoid valve.
Cvstem/compownt/Descric-ion l
2.
IIPCI WV-2034, adjusted closing torque switch.
~
f 3.
Safety relief valve "F" discharge line snubber #27 NI2A Peplaced with spare.
l l
l Critical Path Activity Repair Safety Relief Valve i
Single Releases or exposures None
> 10% Annual Allowable.
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Maga j _ _ _. _... L Reduction j.
X.
j s_cen; Pat _e._/.T_.L.me-
...,l!.8/3/76 22:42 8/24/76 12:00.
ib: "at e /Ti.re.
_ _ _.. _ __8/10/.76._
.. 15:38.
.8/25/7fL
- 3:00
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HnaLest,Hetrs
._ fn 160. 93.....
.... _ _ _ _ - - - - - 19 00..
Cause/Systc/ :'ogorera Scheduled outage to locate and repair Feed-No 12 reactor recirculation ptrnp tripped due water heater leaks and other miscellanems
{::maintenance. Leaking tubes were plugged in to improper application of jumpers during 4
a routine surveillance test resulting in a 14A,13A,12A,12B, and 11B feedwater heaters.
1:lAflangeleakwasrepairedon11Aheater.
power reduction to 57% of rated.
+
Reportable Occt:rrences:
l e
None LER Date/RC#.
None g
t a
e Corrective.'uticn A diaphram was seal welded over the Test switches will be installed to clininate i
11A heater nanhole. All leaking tubes were plugged.
the use of jumpers for routine surveillance.
mjor, Safety Related, lI 1.
Safety relief valves "A" 6 "G""
None Corrective m,ntenance:
i replaced top works with spares.
System /Componcnt/Descr2p lon 2.
Safety relief valves "A" "B" "C" "IT-replaced air operator diaphrapas.
t i
d Critical Path Activity Repair Leaking Feedwater IIcaters.
Recovery Activities.
I Single Releases or exposures;
> 101 Annual Allowable.
i, ' a 2 None
.i t 4
ff
s s
- 7
- n,v ";. : 22D %ER 1DCC1'irss 7_
yll_ __ _.
X y,,,,,,
Jutage c
Reduc.t i cn._.._.. _
Fro:n: Date/ Time iLE/2B/.76 _ _
12:37 9/6/76 _.._.
. 10:15..
I Time lost, Sours
.. 9/fl76.
21:53 To: Date/TI E i R/79/7ft __.
17*7?
r
__ J, L 75.
__._1.1 13_. _._ _.. _.
Cause/Systen! Components l.! lEgh neutron flux scram resulting from Reactor scram caused by closure of inboard FSIV's.
f a spurious pressure regu'ator mal-FEIV's closed when DC circuit was de-energized to function.
locate source of DC ground. The AC circuit apparent-r ly was in a tripped condition at the time.
None Reportable Occurrences:
g LER Date/R0r.
ec>
Corrective action Pressure regulator components FSIV reset switch was replaced. Procedure for functionally checked and found troubleshooting DC grounds is being revised to to operate properly.
assure FSIV AC circuit is energized prior to deenergizing DC circuit.
Major, Safety Related*
None 1.
Group I isolatien reset switch, panel C-05, Corrective 'hintenance
- 'ed~
Syste/Componcnt/Descripticn
- n..
1!
i l
Critical Path Activity SCRAM Recovery 6 Start Up SCRAM recovery and startrp activities.
f Activities.
Single Releases or exposures DC None
> 10% Annual A11onable.
l m
1 W,/ A ~;
17 '-c h ER Etn C W ~
Dutare ti
.X Reluicr.
.k g
'.10/03/76..
05:59 10/3/76 17:00 frcn: Nte/Ti~
Tc-Late / 7 c in/nT/26 _.
. 12:50
._.10/6L76
_ _ _ 11:3n Tinc im t. H.ars d_._ _ __.6.81 18.50 r.w e/Systc~/ Lnpenents il Reactor pcwer reduced to 54% of rated folicwing i High neutron flux scram caused by pressure recombiner system probicas that resulted in a i spike due to rapid closure of 82 turbine slight loss of condenser vacuun. Recarbiner lI Rapid closure caused by misadjustrent control valve during a valve exercise test.
syste. problens involved Icakage of non-condensi-I bles back to the condenser when one of the recort-
't nr hmm 2rlay.
Reportable L' currences:
f hiner_+rM m m returned. tn scrrice._____ ~. _
,,one LER Date/RGd.
Jl!
None u
+
Corrective.'.ctica The valve closa'e timing relav was -
.h e - leakage stopped w'nen cold train was placed
..preperiv adj.us t s,. E Co. is invest-in walrup e.
g.1gatingmodificationstodecrease sensitivity of tining relays.
l Major, Safety Related.
None Corrective ".aintenance:
Syste-/ Component /Descnction
'.i
+
' l, ti 1
t b-Critical Path /<tivity ll SCRA'4 Recovery and Startup_.
Recovery Activities.
ogActi?ities.
4.
Single Releases or exposures I..ene
> 101 Annual Allowable.
w il 1
l
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- j10/16/20 3
- 00 10/20/J6
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102. %
T m.e Lost, Hours o 60 Scheduled outage for repair of prir:ary sys-
~~
- I' Cause/Systen/ Corponents l High neutron flux scran caused by tem Icaks inside the drywell.
Pat. king was Pri-I pressure spike due to erratic opera-replaced or tichtened on nine valves.
tion of 81 ar.d 82 turbine control r.arv source of leak 3Re was through packing on valves. Erratic operation caused by
. reactor head vent valve XIV-4 llailure of control _ valvepition feedback linkagc
~-
l '
None Frportable Occurrences:
I None o
to LER Date/RCe.
[ --- - - -
t Corrective Action Replaced ar4 tightened packina Engineering Feedback linkage was repaired.
staff studvn.g possible replacenent of s:all si:ed valves with hermetically scaled valves.
~. -.
Major, Safety Related, 1.
Neutron renitoring - SDf Ch. 24: Repai r-Corrective raintenance:
1.
FSIV SOB, replaced DC solenoid ed detector drive.
Syste vCarpenent/Desc in. ion
- coil, i
n i
Critical Path Activity Repair Broken Turbine Control
! Repair Leaking Valves in Dnvell Linkages.
1 None Single Releases or exposures l he I
> 101 Annual Allowable.
f il f
._ 9_
m Mage
,1, _ _
Reducti?_, _
X x
i Fm
>tt.
h 11/1/76 22:00 11/9/76 10:M I;. Nte/T r..e..
!!_._11/5/76_. _
10.:45_.
11/11/76._.
7:37 Time Lest,i. cur.,
J, 34.75 45,62_ __ _.
Cause/Syst<n/ C n n.nte S,
h Pwer was reducaito approximately 681 Power was reduced to 61% of rated to wor,x on 812 II of rated to work on 811 reactor feed reactor feedrtr p.
Feed ptro was exhibiting hich purp. Feedptry was exhibiting hich vertical vibratien on the inbeard bearing.
1 axial vibration.
Reportabic Occurrences:
r LER Date/RC8
!l
\\.C""
k.UC o
i-
~
l t
i Correctise Action A new coupling war installed and the inboard bearing and seal was replaced and the pump the ptry and motor were ra-and notor were realigned.
aligned.
li 4
Major, Safety Relate!,
3.,'
l Corrective raintenance:
None Syste-/Corpen(nt/Pactrtionj a
i:l' b!
Critical Fath Activity ll Repair of Reactor Tm.
RepaEEcacter fE$.! water Pter.
~~~
' ' ~
l! Ptro.
1 Single Releases or exposures h None._ _
None
> 101 Annual Allowable.
I e
l
..j!ACT.~ L *GD3, rYoD dh.G* % -
~ ^
._ X 3tnage Reductim _
__ __' 8
.X Frce: M ic/rine
. 11411/I6__.__
__ 7;37... _
_12/5/76 _.____
__..._,_13:0Q_
_._7:00 To: Dote / Tine if 1i/1,/ 6 _ _
5:02 _.
..1226/26 ___
Tire Lest. !!mrs l'..
_ _ J,. 4.,
._18.0n CausejSystem! Cxmrents
' ?
Power reduced to app uximately 701 of rated due to loss Low condenser vacuum scram caused ll by valving error during sur-of condenser vacutri with one circulating water ptrp dveillancetest.
operating. Loss of vacutri apparently due to air build-h up on circulating water side of condenser.
_i i'eportabic Cccurrences:
M.C re l
LER Date/R08.
1 e
l I
Second circulating water pump placed in operation.
Corrective Action Reestablished proper valve Condenser scavanging syster. air vent valves repaired.
positions.
?todification of scavanging system is being imestigated by engineering staff.
Major, Safety Related, 1.
Neutron monitoring - IPM Ch.12:
!bne Corrective Matntenance:
Replaced Prearplifier.
SystpKonponcnt/Descriptien 2
Neutron !bnitoring - SR'! 01. 21 3
i Repaired period amplifier
.I Critical Path Activity SCRMt Recovery and Start UP Recovery Activities.
Activites Single Releases or exposures gone
> 10% Annual Allowable.
.i
J_'_~/ _:*^ ~ TG. C_T P_ w ? M. _10C_7.L._%__
r.
49 anage j [ _ _ ___._..
X Reluction il_. _.. _ _.
i Fro:.ute ime
'9 12f1Bf.76..
7M To: Date/ri.ne F 17/70/76 _.
._...'i-M.._._....
Time Lost, 5 urs L
50.Z5... _....- __
'l Scheduled outage to locate and repair Cause/Systeal Comants l:
high pressare feedwater heater tube leak and c+Jer i
i miscellaneous maintenance. One tube was found leaking on #14B heater.
I Reportabic Occurrences:
I LER Date/RO8.
None g
i i
i Corrective Action Leaking tube and other tubes in 814B
- heater that showed signs of *Jtinning were plugged.
3 i
Major, Safety Related, Corrective Maintenance:
None l
Systen/ Component /Desc*iptic.n it 1
It D
j Critical Path Activity k Repair Leaking Fcedwater Heater.
g Single Releases or exposuresI
> 10% Annual Allowable.
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25 -
j IV, ruel Performance No refueling, and therefore no fuel insrection, was conducted during 1976.
Ilowever, Analysis of off gas and reactor water chemistry shows no evidence of fuel cladding failures since the fall 1975 refueling outage.
Following that outare, analysis of the six principal noble gas isotopes indicated greater than a 95% re-coil pattem,confiming that the activity was from residual plate out of fuel material resulting from previous operation with substantial fuel claddine defects. This pattem did not chang,e durine,1976.
As expected, off gas release rates and reactor chem-ist n analysis for radiciodine and dissolved /undis-solved activity levels exhibited a 10% decrease during the year due to depletion of the plated out material.
f'
I V.
CIMNGES 1ESTS AND 12PERIM13'TS, The follos tag sections include a brief description and sta:rnary of the safety evaluation for those changes, tests and experitnents shich were carried out without prior NRC approval, pursuant to the require-ments of 10 Cit 50.59(b):
DRYWELL PLitPBACK SYST124 (SRI 102) 1.
N2 Desc ription of Cl.any.c A system was installed for pumping drywell atmosphere to the instrument gas supply for the valves and instruments located inside the dryvell.
This was done to eliminate the buildup of oxygen in the nomally inerted conta 'nment.
The system utilises a spare drywell penetration, X-48, for suction (see Semi-Annual Operating Report #5).
The system consists of two air operated isolation valves on the suction line, two 10 SC111 compressors, moisture and particulate removal equipment, two receiver tanks and associated piping. The system discharge line is connected to an existing nitrogen line which is connected to the dryvell instrument gas supply line, rurena ry o f Sa fe t y Evaluation The safety related portion of the design change includes the J
tie in to the spare dryvell penetration and installation of the isolation valves.
The inotallation meets the requirements of ANS1 B.31.1 Power Piping Code.
The process line is designed for Seismic Class 1 to the first restraint beyond the isolation valves. The valves are isolated via the existing primary con-tainment group 2 isolation logic. Control wiring for the valves meets the separation criteria of IEEE-279. The valves are designed to fail-closed on either loss of power or loss of air. The valves were subjected to a local leak rate test following in-sta11ation and vere found to be acceptable.
2.
FLOW TPR1SMITTER AND ANALOG TRIP UNIT FOR OPERATIONAL EVALUATION (SRI 172)
Description of Chant,e A Rosemont Inc. flow transmitter and analog trip unit were connected to the "C" main steam line flow element for an operational evaluation.
Stumry of Safety Evaluation The transmitter and associated tubing are equal in quality to the existing steam 11ow instrumentation.
The additional voitune of the transmitter does not have a significant ef fect on the response time of the instruments connected to the same flow element.
, 3.
tJPGRADE TIMPORARY RADWASTE St.lfDGE P1plNG _ Q175 19)
Description of Channe lhe temporary hose connections in the "B" hopper room required f or pumping radwaste sludge to a pot table processing unit were replaced with 1" pipe.
The flowmett, and backprosaure control valve were interchanged to exclude from the ftow measurements any flow returned to "B" hopper by means of the relief line.
A connection was also installed to allow the contents of "B" hopper to be discharged to Waste Sludge Tank T-35.
Suma ry of Sa fety Evaluation Replacing the temporary hone and hose clamp connections with pipe and threaded pipe connections increases the strength of the system, reduces the Itkelihood of spills, and reduces radi-ation exposure to plant personnel.
The piping installation meets the requirements for Bechtel Specification M 40, Class JB piping, which was used for other similar piping in the radwaste system.
4 RIMOTE FLUSil1NG CAPABILITY FOR RADWASTE SLUDGE Il0SE STATION (M75-25)
Description of Chang The temporary hose station used for filling a portable processing truck with radwaste sludge was replaced with pennanent piping and a flushing connection was made to the service condensate system.
Sumary of Safety Evaluation 9
Radwaste is prevented from entering the service condensate system by the prrsence of a gate valve which is closed except when flush-ing, a check valve, and a relief valve which prevents the pressure in the radwaste line from exceeding the pressure in the service condensate line. The relief valve also prevents overpressurizing the portable processing unit's pump, the valve operators for this r'
hose station are remote located to reduce radiation exposure to plant personnel.
The piping installation meets the requirements of licchtel Specifi-
{I cation M 40, Class JB piping.
5.
REPIACIMENT OF TURB1NE BUILDING ROOF f311AUST FAN TEMPERATURE CONTROLLERS (M75-53)
Description of Change The temperature controllers for the turbine buildins; roof exhaust fans were replaced with contro11ero capable of being set at a higher terareratut e.
The new controllers are set at 250 F to pre-clude operation for any purpose other than heat and smoke removal in the case of fire.
t-
, l Sunnary of Safety Evaluation Actuation of the roof exhaust fans at 250 F satisfies insurance company requirements.
Operation of these fans is not needed for normal plant operation.
6.
REP 1ACEMENT OT REACTOR WATER CLEANUP HEAT EXCHANGER TEMPERATURE SWITCH WI111 AIARM UNIT (M75 57)
Description of Change The reactor water clear.ap system downstream high temperature alarm switch (TS 12128) was replaced by an alarm unit connected in parallel with an existing thermocouple (TE 12-97) which pro-
)
vides temperature indication in the control room.
This modifica tion was made because calibration of the temperature switch was im.
practical due to the high radiation from the heat exchanger.
Sunnary of Safety Evaluation i
This replacement increases the accuracy and reliability of the ala rm function.
The temperature switch had no safety function.
7.
INSTALIATION Or DRWELL PRESSURIZATION SYSTEM (76M002)
Description of Change A system was installed to pressurize the drywell relative to the corus by pumping torus atmosphere to the drywell.
Fuintaining the drywell pressure positive relative to the torus reduces the dynamic loading on the torus in the event of a loss-of coolant accident.
The system consiste of a suction line from the torus (connected to the torus vent 2" bypass line downstream of CV 2384), an isolation valve (CV 7440), two Roots blowers, and a discharge line to the drywell (connected to the N2 pumpback system supply line and the containment nitrogen purge line).
Summary of Safety Evaluation 1he acw isolation valve installation meets the requirements of all i
applicable codes and specifications.
The quality assurance activities associated with installation were conducted in accordance with f'
Article NA 4000, ASFE B&PV Code,Section III.
The process line is designed for Seismic Class 1 to the first restraint beyond the isolation valve. The valve is Isolated via the existing primary containment group 2 isolation logic.
The control wiring for the valve meets the separation criteria of IEEE-279.
The valve will fail closed on loss of power or loss of air.
There is no high pressure piping or high speed unchinery in the bmnediate vicinity j
of the isolation valve. Check valves have been installed in the discharge line of each blower to prevent vacuum breaker bypass flow i,
i
~- ~m
,,_,...._,m.
~. - -, _... - _, _,.. _ _. - - - - _, -, - - - -. - - -.
m-
I c.!
.. =
-29 Surrna ry o f Sa f e t y Tvs lua t i on (Cen t i nue d),
i f rom the drywell to the torus.
1he bypass flow area as a result of check valve failure is less than 147, of the maxin. tun allowable.
The automatic isolation valves are required to close within 60 seconds of receiving an isolation signal.
The blowers are suspended f rom the ceiling over the torus.
If the blower supports failed, the blower would fall 2 feet to the top of the torus.
The energy available if this occurred would be less than 17. of the energy needed to penetrate the torus wall.
8.
INSTAL 1ATION OF TORUS SUPPORT C01,UMN REINFORCIMENTS (76M004 C) g[
Description of Chance Each torus support column was reinforced by weldinp. two 160 ten-inch schedule 160 carbon steel pipe sections over the existing eight inch support coltuans.
This modification was made to increase the snargin of safety in the coltunn support design by increasing the crossectional area. The pipe sections are made of SA 106 Grade B material.
The original support columns were designed in accordance with AISC Specifications. The modification was perfonned in accordance with ASME B6PV Codes Section 111. Subsection NF.
Strana ry o f Sa fe ty Eva lua tion Use of ASME Section III, subsection NF, (which did not exist at the time of original design 6 construction) is consistent with current practice.
Installation of the reinforcement increases the load carrying capability of the torus support coltunns.
9.
INSTAL 1ATION OF DRAIN TRAP ON AT-7731-B 0FF GAS 112 AMALYZER SAMPLE PUMP DISCHARGE (76M009)
Description of Change, A drain trap was installed at the discharge of the AT-7731-B offgas 112 Analyzer Sample Pump to facilitate drainage of condensate and prevent damage to the t, ample pump bellows.
Steina ry o f Sa fe ty Eva lua ti on The offcas H2 Analyzer Sample Ptunp discharge tubing is outside of the of fgas system ASME Code T,ounda ry.
The trap specifications are the same as existing traps.
l 10.
ROUTH CONTAINMElff OXYGEN A!;ALYZER AND CIM-CAM SAMPIE GAS TO REACTOR BUILDING EXHAUST PLENUM (76M027)
Description of Change l
l The sample returns f rom the primary containment oxygen analyzer and l
CIM-CAM were routed to the reactor building ventilation exhaust
.-_._-_7 Description of Change (Continued) l rather than returning them to the containment.
This modification prevents the introduction of air into the containment frma leaks in the CIM-CAM and provides a constant backpressw 0 fcf (Ac oxygen analyzer.
Suma ry _of Safety Evaluation luterial used in the modification is equal in quality to that used in the original portions of the system.
Administrative con-trols prevent discharge of the sample to the building atmosphere i
during abnormal ventilation system operating modes. The sample flow of approximately 2 SCFM is insignificant compared to the building e.xhaust flow and dius does not contribute significantly to the overall release rate.
- 11. RD10 VAL OF VENT LINE FROM SIX INCH MAIN TURBINE MOISTURE SEPARATOR DRAIN LINE (76M029)
Description of Change A 3/4" vent line was removed from the A and C Main Turbine Moisture Separator cormon drain line.
The vent line was replaced with a capped three inch schedule 40 pipo stub.
The purpose of this modi-fication was to remove a crack at the original 3/4" branch connection.
3 Summary of Safety Fvaluation The modification was performed in accordance with the original design code, ANSI B.31.1, Power Piping.
The original Bechtel Specification called for AISI Type 304 or 316 stainless steel in this application.
AISI Type 31CL was substituted becar 3 its mechanical properties are superior to requirements of type.,04 or type 316 and because chemical properties of type 31GL are such that chromium carbide pre-cipitation is idaibited during welding.
12.
INSTAL 1ATION OF TDiPORARY RHR SYSTDi TEST INSTRUME!UATION (76M064)
Description of Chan3e_
Test instruments were temporarily installed to monitor (1) pressure upstream of each of the inboard and outboard shutdown cooling suction isolation valve (2) flow in the A loop LPCI line, containment cooling line and in the reactor vessel head spray line (3)
"A" RllR pump start and (4) opening of "A" LPCI throttling valve M02012.
The l'
purpose of the test instrumentation was to detect the cause of the flow disturbances experienced during initiation of shutdown cooling.
Summary of Safety Evaluation i
Installation of the test instrumentation van shown to have no effect on the ability of the RHR System to perform its safety function. In-sta11ation of the instrumentation was completed in accordance with requirements for Quality Group B. NRC Regulatory Guide 1.26, and Appendix B to 10 CFR 50.
<r.,s, onu 195 NRCe U s. NucLa An atouLatomy co*
SsioN oocda v Nuwee no - 2. G 3 NRC DISTRIBUTION ron PART 00 DOCKET MATERIAL
"'[g"P0RSFILE IO' I 4HOW Daft or 00 W MENT Mr J C Keppler Northcon States pwr Co 2-28 77 Minneapolis, Mn oatt nectivt o l,
L 0 Mayer
> 3 -77 NLetten DNotonis t o enoe iNeurtonu Nuust n or cocits at etivto MonlosNAL
$ %ctAssitsgo Ocorv one signed pt sc nip fioN ENCLOSURE Ltr trane the following:
Annual Operating Report #11 for the period I
l-1-77 thru 12-31-77.......
1p (40 cys encl rec'd)
(-
m..
il
\\ inch DISTRIBUTION PER S. SHEPPARD 3-3-77 g
b v i s v.L
. - - - - + * ~ '
3 NOTEt LIC. ASST.(S)...LTR & MAIL CONTROL ONLY....'OCKET CLERK WILL SEND 1 CY EACH TO PIANT NAME:
Monticello ASST. SAFETY FOR DISTRIBUTION.... TOTAL (5) CY3i SAFETY FOR ACTION /INFORMAllON 3-4-77 chf PROJECT MANACER:
_"gg maan LIC._ ASST:
(5)
Q4y 1
.*SEE NOTE INTERNAL DISTRIBUTION 1 REG FILE 3 r
nw FDR I k.E_,G)
_ HIP _C_(2)
HANAUER JIELLO i
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_3MMES J._ COLLINS EX1ERNAL DISTH10UTION CONT HOL NUMeiE R LPDRj_Qgftftyp j_d { J A BJ N@b I
TIC:
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NRCFonM 196 (2-76)