ML20247E251
| ML20247E251 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 12/31/1988 |
| From: | Eenigenburg E COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3286A, 89-236, NUDOCS 8904030011 | |
| Download: ML20247E251 (35) | |
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A
-1988 ANNUAL NRC REPORT OF
SUMMARY
OF OPERATING EXPERIENCE, CHANGES, TESTS, AND EXPERIMENTS PER 10 CFR 50.59 FOR DRESDEN NUCLEAR POWER STATION-COMMONWEALTH, EDISON COMPANY 1 Mil DOCKET LICENSE 1
050-010 DPR-2 2
050-237 DPR 3 050-249 DPR-25 li N
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TABLE OF CONTENTS
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'1.0 In't roduc tion -
.2.0 ' Annual. summary of plant or procedure changes,-tests and experiments.
2.1 Amendments'to Facility License or~ Technical Specifications
.2.1.1 Unit 2 and Unit 3 Shared
. 2.1. 2 '
Unit 3' 2.2 Changes to Procedures Which are. Described in the Final Safety 1
Analysis' Report (FSAR) (Units.2 and.3) 2.3 Significant Tests and Experiments Not Described in the FSAR (Units 2 and 3) 2.4 Completed Safety Related Modifications 2.4.1~
Unit 2
'2.4;2 Unit 3 2.5_~ Temporary. System Alterations (Units 2 and-3) 3286a r
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1.0 Introduction The Dresden Nuclear Power Station (DNPS) is a three-unit facility owned i
by Commonwealth Edison Company and located near Morris, Illinois. Unit 1, a General Electric Boiling Water Reactor with a design net electrical output rating of 200 megawatts electrical (MWe), is retired with all fuel removed from the reactor vessel. Therefore, no Unit 1 operating data is provided within this report. Units 2 and 3 utilize General Electric Boiling Water Reactors and have an initial design net electrical output l
rating of 794 MWe. Waste heat is rejected to a man-made cooling lake using the Kankakee River for make-up and the Illinois river for blowdown. The architect-engineer for the Dresden Units was Sargent and Lundy.
I The 1988 Annual Report of Channes. Tests. and Experiments oer LQ CFR 10 59 for_Drgsden Nuclear Power Station. (DNPS). Commonwealth Edison Cnmgany (1988 DNPS 10 CFR 50.59 Report) is submitted to the Nuclear Regulatory Commission (NRC) pursuant to the requirements of Title 10, Code of Federal Regulations (10 CFR), Section 50.59(b)(2).
For each of the changes, tests, and experiments described herein, a safety evaluation was performed which determined that the change, test, or experiment did not constitute an unreviewed safety question.
This report was compiled by Gerrine Paramore of the Technical Staff, telephone number (815) 942-2920 extension 2364.
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2.1 Amendments to Facility'Liccnsa or Tschnic 1 Specifiestions P04!
2. 1. 11 Unit 2 and Unit 3 (Shared)
Amtadment No. 101 and 97 The NRC approved amendments 101 and 97 to.the Unit 2 and 3 Technical!
Specification Sections 3.5/4.5 to change the operability requirements i
of the Emergency Core Cooling Systems (ECCS) during cold shutdown and I
refueling operational modes.
Amendment No._103 and 99 The NRC approved amendments 103 and 99 to the Unit 2 and 3 Technical Specification Sections 3.7/4.7; these. amendments clarified primary containment oxygen concentration and drywell to torus differential pressure requirements.
Amendment No. 98 and 93 The NRC approved amendments 98 and 93 to Unit 2 and 3 Technical Specification Sections 3.5 and 4.5 to change the diesel generator testing requirements associated with inoperable ECCS components.
6mendment No. 97 and 92 The NRC approved amendments 97 and 92 to Unit 2 and 3 Technical Specification Section 6.0 for miscellaneous changes to Administrative Controls.
' Amendment No. 102 and 9B The NRC. approved amendments 102 and 98 to the Unit 2 and 3. Technical Specification 3.2/4.2 to modify.the minimum calibration frequency and location requirements for various post-accident instrumentation in accordance with Regulatory Guide 1.97.
Amendment Nos. 99 and 100: Amendment Nos. 95 and 96 The NRC approved the above amendments to the Unit 2 and 3 Technical Specification Sections 3.1/4.1 for the elimination of the Average Power Range Monitor (APRM) downscale scram requirement associated with the Reactor Protection System (RPS).
The ebove amendment states the following*
This Technical Specification amendment was submitted in response to the corrective actions given in Licensee Event Report (LER) 87-22 on Docket 050-237. The elimination of the APRM downscale scram requirement was supported by a General Electric (GE) analysis. To summarize, the removal of the APRM downscale scram essentially elminates the Intermediate Range Monitor (IRM) scram function in the run mode. This is acceptable because the IRM scram is not credited in the Rod Drop Accident analysis while the Rod Withdrawal Error is prevented by the APRM downscale rod block. It should be noted that GE's discussion of plant transient and safety analysis is consistent wit.h Advanced Nuclear Fuels (ANF) methodology. As such, no additional analysis was required.
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2.1 Amendments to Fccility Licznsa or Tschnical Spacifiestions 2.1.2 Unit 3 Anlendm_en.t No. 94 The NRC approved Amendment No. 94 to Unit 3 Technical Specification Sections 1.0, 1.1/2.1, 3.1/4.1, 3.2/4.2 and 3.5/4.5 to implement Unit 3 reload licensing requirements prior to startup from a refuel outage.
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2.2 Chingssito procsduras which era not dsscribsd in ths Final Safety Analysis. Report (FSAR)
Table 2.2 lists those procedures. referenced in the DNPS Units 2 & 3 FSAR which were revised in calendar year 1988.
in.the majority of cases, the changes to these procedures fall into one or more of.the following four categories:
(1) An administrative change, such as revising the responsible individual or procedure' reference, which does not alter the intent of the procedure; (2) A change to clarify the actions to be performed under the procedure, such as _ the adding of cautions or. additional procedure detail,_which does not alter the intent of the!
procedure; and
-(3) A change to incorporate details-of a newly utilized component,_
either as a subject of-the procedure or used in performance of 4
the procedure, which does not alter the intent of the procedure.
(4) - A change to implement testing of calibration methods for equipment and/or instrumentation, which does not alter the-intent of the procedure.
- Procedure revisions of these types are noted in Table 2.2.
If the revisions to a given procedure are exclusively'of one or more of these three types, then no further discussion of the change is provided.
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r 2.3 Significant Tasts and Exp2riments Not Describ:d in ths FSAR (UNITS 2 and 3)
-Significant.Special Procedures involving tests not described in the Finali Safety Analysis Report during the calendar year 1988 are listed
'below.
Procedure No.
Procedure Title / Description
. SP 1-2 Test procedure for Control Blade Guide Rack assemblies.in the Unit 1. fuel building
- SP-88-2-5 Test' procedure for monitoring Limitorque Valve cperator maintenance and surveillance SP-88-2-6 Test procedure f>
verifying the adequacy of the Reactor piping:svotem thermal expansion SP-88-3 Test c ocedure for determination of quarterly valve timing reference valves testing for In-Service Testing (IST). program in ac'cordance with DOS 1600-1. Valves were stroked 3 times.
-SP-88-3-12 Test procedure for monitoring the turbine feedwater cycle maximum capability.
SP-88-3-13 Diegnostic test procedure for Unit 2 and and 3 CRD stall flow testing.
'This procedure investigated the. December 29, 1987 SP-88-3-15
,
- event when the Unit 2 203-1B MSIV failed to fully close when its instrument air line pulled out of
+
the manifold on the valve operator.
SP-88-4-26&27 Test procedure to ensure proper concentration of
~ Carbon Dioxide (C0 ) in the Unit 2 and 3 3
diesel generator rooms upon initation of the Co, fire suppression systems.
l SP-88-4-31 Test procedure for collecting, analyzing and s'
periodically trending the Unit 2/3 ce'ntrifugal pump operating data to identify pump degradation.
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2.3 Significant Tests and Experiments Not Described in the FSAR (UNITS 2 and 3) fIDredute_Han.
Procedure Title /Descriplinn SP-88-5-39 Modification test procedure for isolating the Unit 3 250V DC Battery and paralleling its associated loads to the Unit 2 250V DC battery.
SP-88-5-55 Diagnostic procedure for air testing the Unit 2 MSIV pilot manifold assemblies and accumulators.
SP-88-5-68 This procedure provided guidelines for testing.
Unit 3 primary containment air-operated vent and purge valve air-operator accumulator systems.
SP-88-6-76 Test procedure to verify the function of a new Unit 3 Feedwater Low Flow Regulator Valve and associated drag valve control function.
S P-88 7 7 Unit 3 start-up test procedure for walkdown of essential switchgear prior to startup and after a refuel outage.
SP-88 78 Unit 3 startup test procedure for characterizing the capacity and operation of the new Feedwater Regulating Valves, SP-88-8-97 Procedure provided test for Unit 2 and 3 Core Spray pump minimum flow operation to verify adequate flow.
SP-88-8-99 Procedure determined the appropriate High Pressure Coolant Injection (HPCI) steam line flow used for calculating a 300% high steam flow isolation setpoint.
SP-88-10-135 This procedure outlined the requirements for a HPCI turbine overspeed trip test.
3286a i
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l 2.3 Significant ?ests and Experiments Not Described in the FSAR 3
(UNITS 2 and 3)
^
l Erac_edutc_E h Procedure Title / Description j
I SP-88-10-140 Test procedure for operating the Unit 2 2A Recirculation pumps above 28% pump speed with i
e less than 20% feedwater flow to determine the i
cause of abnormal LPCI injection line vibration.
j j
SP-88-11-149 Modification procedure for testing the Unit 2 2B
)
Feedwater Regulating Valve (FWRV) ability to transfer from the low flow valve and its ability I
to handle 3 inch and 10 inch reactor water level l
changes at various power levels.
SP-88-11-144 Test procedure outlined the steps for performing j
a chemical decontamination of the Unit 2 reactor q
recirculation system piping SP-88-11-150 Test procedure verified the proper electrical wiring of the Unit 2 MSIV inboard and outboard Primary Containment Isolation System (PCIS) position indication and determined if improper grounding of the circuits or improper fuse size caused fuses to open at abrormally high rates.
SP-88-12-197 Modification test for visual inspection of field boundary welds and pipe connections on the CRD drive water supply header.
SP-88-11-157 During the months of November and December of SP-88-11-158 1988, a Unit 2 and 3 Service Water System outage SP-88-11-159 was conducted to facilitate replacement of the SP-88-11-160 service water strainer isolation valves. An SP-88-11-161 alternate service water system was established SP-88-11-162 in place of the normal Service Water System for each unit. Since the removal f rom service of the entire Service Water System was not a normal evolution governed by existing station procedure (s), seven Special Procedures (SP) were written and implemented. The first SP l
implemented for this evolution was a pre-operational testing procedure to functionally test and verify the operation and leak tight integrity of the alternate cooling water systems. Five operating Special Procedures were implemented to repair and operate the alternate cooling water systems, and then restore the service water system to normal. Additionally, a Special Procedure was written as a contingency procedure to provide operational guidance in the event either alternate cooling water system failed.
3286a
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2.4 c -nlated Safety Lelated Modifications This section provides a brief description of the changes to'the DNPS
- facility as' described in the FSAR which'were completed in' calendar year.-
1988.. Only modifications which have been completely closed are listed.
within this' report. Modifications authorized for use but not completely closed are not-listed; tlwy will be reported based.upon the date_of their-final closure._ For: ease of reference,. the changes.have been identified by their design change control modification number.
9 2.4.1 Unit 1" No Unit 1 modifications were completed in 1988.
3286a
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7 i?
2.4 Ca=pleted Safety Related Modifications
'2.4.1-Unit 2.
a;
! Modification _Ns.
. Description
.M12-2/3-88-16 This modification involved removing diesel 1 generator air' start piping / filter supporti mounte,d to non-safety personnel. handrails and installing new seismic supports to hold air filters / piping that meet FSAR seismic load requirements. The safety evaluation. concluded that the seismic supports reduce the probability of a' seismically induced failure.
~ M12-2-85 75 This modification replaced the existing' General Electric (GE) type CFD differential relays for
~
the Unit 2,.3 and 2/3 diesel generator output' breakers with seismically. qualified' Westinghouse type SA-1 differential' relays. The safety
- evaluation concluded that the new relays will; provide improved reliability during a postulated '
seismic event.
~
M12-2/3-82-24
.This modification adapted the Dresden Unit 1 fuel'
- storage rack to' provide a means'for long term storage of. eighteen thoria rods previously located in the Unit 2/3 high density fuel storage racks..The safety evaluation concluded that fuel racks loaded with the 18 thoria rods meet the Quadrex qualifications.. In addition,.the-
.K-effective-value for the 18 thoria rods is lowered and will remain below 0.95.
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ta I.
~2.4 Comp 1stcd S$fsty R31ctid Modifientions
' 2.4.1 Uriit' 2 >
]
1 Modification _ h -
DREIIPtiDR '
M12-2-86-9 This modification involved the replacement of a
, control switch and relay for HPCI valve 2301-4 with a FSAR qualifled and. seismically mounted control switch and relay. This modification alters the method by which the operator controls the steam inlet valve to the HPCI system by providing throttle open' capability and a valve pull-to-lock moveme.nt funetion'90'.from center, and modifies the method by which the HPCI
' initiation signal is sealed in.
The safety evaluation concluded that modification does not affect the primary sensors for automatic' initiation or the availability of the HPCI system.
M12-2-80-90 This modification replaced existing spent fuel racks in the Unit 2 spent fuel pool with new absorber high density spent fuel racks. The-safety evaluation concluded that the twic! ion of fuel racks remains the same and a more restrictive design criteria'has been implemented.
3286a
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y 2.4 Comp 12t;d S:fety R31ct;d Modifications 2.4.2 Unit 3 MoslificatiotLN m D ucIlplisn M12-3-86-34 This modification separated the control logic of a LPCI minimum flow bypasa valve sensor in one loop from the second loop LFCI valve sensor to resolve a single failure concern in accordance with I.E.Bulletin 86-01.
The safety evaluation concluded that the separation of the control logic for the LPCI minimum flow bypass valves assures the FSAR redundancy requirements.
M12-3-88-27 This modification involved installation of en improved support for a 1B MSIV pneumatic operator supply line. The safety evaluation concluded that the support complied with the FSAR design criteria.
M12-3-86-10 Installation of a remote control systen (Unit 2 & 3 shared)
(control switches, reset buttons, indicating lights and cable / conduit between turbine floor and centrol room) for the Reactor Building ventilation isolation dampers in the control room. The control room reset capability allows the operator to reopen isolation valves without requiring a second operator to reset and position the damper valves locally. The safety evaluation concluded that the modification has no effect on the isolation signal or initiation of the standby gas treatment system.
M12-3-85-90 This modification involved the removal of the redundant, temporary Scram Air Dump System (SADS). The SADS is now superseded by a permanent Scram Discharge Volume (SDV) level scram protection system which directly detects the accumulation of water in the SDV, unlike SADS which indirectly monitors water level. The safety evaluation concluded that the margin of safety is not reduced.
M12-3-86-31 Modification of the Emergency Core Ccoling System (ECCS) pump minimum flow valve logic to provide the control room operator with ability to close valves even with an indicating accidert signal, and to perform the close loop isolation function General Design Criteria (GDC) 51 requirements.
The safety evaluation determined that the ECCS logic changes will allow one function of the ECCS to be executed without jeopardy to the second function.
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h" 2.4 Complat;d S fsty R31st d M:dific tions 2.4.2 Unit 3 r
Modification Nai Description M12-3-87 This change involved mounting vibration isolator assemblies to the Main Steam Line (MSL) Lou Pressure switches to prevent spurious actuation of the switches. The safety evaluation determined that the Limiting Safety System Settings are not changed and MSL Low Pressure isolations will still protect against rapid reactor depressurization.
M12-3-85-83 This modification replaced the existing Unit 3 250 VDC,120 cell batteries with GNB NCX-1500, 116 cell lead calcium batteries and installed seismically qualified racks to support the batteries. The safety evaluation concluded that the new batteries are sized to provide sufficient DC power for the worst case accident load profile in accordance with Technical Specification Section 3.9.B.
In addition, the new seismically qualified rack will ensure battery operability.
M12-3-84-108 Installation of reactor pressure and level instrument based on th, requirements of Appendix R Reverification Analysis and Regulatory Guide 1.97.
The safety evaluation concluded that the installation of new instrumental'on and revision of existing instruments range will improve safe shutdown capability by providing two redundant and physically separated channels.
M12-3-84-119 This modification allows the simultaneous operation of two Standby Liquid Control System (SBLC) injection pumps and increases the minimum sodium pentaborate solution concentration to n 14 weight percent. The change is intended to accomplish an additional performance objective by providing a deliverance capability of 80 GPM of sodium pentaborate solution that will control, reduce and terminate certain anticipated j
transients in conjunction with a reactor scram
{
failure (ATWS events).
In addition, a second j
pump suction line was added to eliminate l
injection pump Net Positive Suction Head (NPSH) j conce rns. The safety evaluation concluded that the total boron content remains unchanged and the single pump time requirement for insertion is maintained.
3286a
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./2.45. Comp 10ted'S:fsty R212ttd Modifications E
2.4.2 Unit' 3-tic.difica.Li.on_.No -
Description
[
g
-M12-3-85-85L This modi ition involved the removal of f
deactivated Unit 3 Scram Discharge Volume (SDV) air header scienoid valve S02-302-18. Thic valve,.which was s' redundant manual test valve.
- associated with the SDV vent and drain system, had been superseded by the installation of:
individual manual test valves for each SDV vent / drain under. modification M12-2-81-11..The safety evaluation concluded that this-modification had no effect on the function of the SDV system.
M12-3-85-96' The objective of this modification was to increase-the reliability of the Unit 3 diesel
- generator. By replacing the engine idler gear stubshaft assembly and lower idler gear turbocharger components with more reliable ones, gear wear was reduced. This modification is designed for partially loaded accelerated starts
-or fu11' speed applications..The safety evaluation concluded that the replacement of the
'above components does not affect the design or-operability characteristics of the diesel generator.
'M12-3-88-26.
The intent of'this modification was to replace-the existing Unit 3 125 volt battery, which consisted of 58 Gould type NCX 1344 cells, with a seismically qualified,-type 58 GNP 21 cell-battery. 'The existing battery was replace 1 due to degrading battery discharge test performance.
The safety evaluation concluded that the new battery is adequate for four (4) hours, with an end of discharge.tenninal. voltage of 105 volts.
M12-2/3-86-3.6 This modification involved replacement-of
.and existing HACR-I synchronizing relay used for the l
M12-3-86-42 Unit 2/3 and Unit 3 diesel generator synchronization system with an improved HACR-IV relay. The safety evaluation determined that the replacement of the relay will provide improved protection during synchronization.
l 3286a 1
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L.' 2 Comp 1sted S foty R21stad Modification s i
'2.4.2: Unit 3 tiodification No.
Description M12-3-86-29' Modification of the. Unit 3 ECCS pump minimum and flow valve logic to provide.the control room M12-3-86-27 operator with the ability to'close the minimum flow valves'even with'an accident signal-p re s en t'.. The safety evaluation concluded that the logic changes provide a minimum flow path for ECCS. pumps in order to prevent dead-heading without jeopardizing the containment isolation function.
M12-3-88-03 This modification involved prov'iding new safety-related HPCI system'and feedwater small bore tap line supports to ensure proper support for vibration and flexibility required for long term plant operation. The safety evaluation.
concluded that the additional pipe supports.
increase the reliability of the associated piping during a postulated seismic event.
1 3286a q
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(2 5. Tcaporsrv Systens Alterations (Unit 2'end 3) l f-1
, Tempora ry.-.
System
- Installation,
~ Removal-411 CIA 1. inn _Ng2
. Description Dale Date f-?L <
l 11-1-88 Remove erroneous numeral two 1-12-88 11-15-88:.
~
-inithe Units column of Control-Rod. Drive (CRD) E-8 position!
indic'ation; II-2-88:
Supply auxiliary. power for-
- 1-22-88 1-22-88 Radiation Waste Panel 2223-4 Lto repair ground-on normal feed cable.
L II-3-88 Lift cable to. clear ground on 1-22-88 3-4-88'
^
waste neutralizer (P.H.)
Transmitter.
II-4-88
. Clear ground.on1 alarm relay at 1-27-88
' 3 - 3 7,
. Panel 2203-56B.
II-5-88 Removed off gas condenser outlet 2-8-88 2-10-88 temperature indicator for repair.
II-6-88 Replaced Reactor Water Cleanup 2-10-88 2-10 _(RWCU) non regen heat exchanger temperature-switch.
-II-7 - Bypass thermostat to repair' heat 2-13-88 8-3-88 trace for-the' Chimney Separate Particulate Iodine Noble. Gas (SPING) effluent monitor.
Replace relay to prevent a rod 3-15-88 3-17-88
'II-8-88
. block when the accumulator for CRD A-9.is out of service (00S). 9-88 Replace relay to prevent a 3-15-88 3-17-88 rod block when the accumulator s
for CRD K-8 is 00S.
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2.5' Temporcry Systema Alterations. (Unit 2 and 3)
L,..
Temporary System
. Installation
-Removal Alteration No.
Description Dalg.
Date.
1 II-12 T0perata crane 3-24-88 3-25-88
.II-13-88 Update camera security system 4-11-88 4-16-88 by Unit 2/3 discharge canal II-14-88 Install external feed to move - 13-88 7-6 grapple over Unit 3 reactor cavity 11-17. Reroute a portion of service 4-19-88 water radiation monitor piping in an area with low background
-radiation II-18-88
' Provide emergency feed for 5-12-88 1-20-89 control room lighting and Heating Ventilation and Air Conditioning (HVAC) systems II-19-88 Calibration of HPCI' Room 5-13-88 5-14-88 temperature switches.
II-20-88 Provide. emergency feed for 5-14-88 1-20-89
' control room lighting and HVAC systems 11-21-88 Disconnect and reconnect 5-18-88 5-22-88 MSIV solenoids in Unit 2 drywell to facilitate testing.
II-22-88 Bypass failed thermocouple for 5-19-88 5-24-88 continued operation of the shutdown cooling system II-23-88 Simulation of " Full In" signal 5-22-88 5-27-88 to clear rod block on CRD E-8 II-24-88 Simulation of " Full In" signal 5-22-88 5-27-88 to clear rod block on CRD C-9 JI-25-88 Simulation of " Full In" signal 5-22-88 5-27-88 to clear rod block on CRD G-8 3286a
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4 2.5 Tempor:ry System 3 Alterations (Unit 2 cnd 3)
)
Tempora.y System Installation Removal Alteralinn_HL Description Data pate t
II-26-88 Provide temporary feed to 5-25-88 2-11-89 i
Unit 2 and 3 Control Room l
Emergency Lighting II-27-88 Removal of failed low level 5-31-88 11-11-88 cut out switch from the drywell equipment drain sump pump's trip logic.
II-28-88 Alteration to divert more 7-29-88 1-23-89 cooling air flow to the Recirculation System MG Set motor, motor windings and override the low flow trip of the MG Set vent fans II-29-88 Alteration to divert more 7-29-88 1-23-89 cooling air flow to the Recirculation System MG Set motor windings and override the flow trip of the MG Set vent fans II-31-88 Tape edge connector to remove 8-4-88 11-17-88 the false indication of digit
- 2 of the CRD portion indicator II-32-88 Lifting the seal-in contacts to 9-21-88 9-24-88 give the Low Pressure Coolant Injection (LPCI) valve a throttle capability to perform maintenance.
II-33-88 HPCI drain line temporary 10-4-89 1-10-89 repair.
II-34-88 Operating the torus cooling 10-4-88 10-6-88 valve in the throttle mode 11-35-88 Provide for live-bus transfers 10-31-88 2-10-89 to support maintenance and relay calibration for Unit 2 Busses 25 thru 29 feed breakers during refuel outage.
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"2 2.5 Tempor2ry Systems Alterations (Unit 2 and 3) o Temporary System Installation Removal Alteration Not Description Dalg '
Date II-36-88 Installation of temporcry exhaust 10-26-88 2-11-89 ducting to perform torus recoat project on Unit 2 II-37-88 Installation of temporary 11-20-88 2-11-89 torus drain fittings on Unit 2 II-38-88 Installation of special flange 10-29-88 2-11-89 to support torus recoat II-39-88 Failure of thermocouple in 10-30-88 1-3-89 recirculation loop causing isolation of shutdown cooling system 11-40-88 Installation of lighting for 10-31-88 2-8-89 torus recoat II-41-88 Removal of main steam area 10-31-88 2-5-89 temperature switches for calibration 11-42-88 Connector checks on Average 11-1-88 1-1-89 Power Range Monitor (APRM) and Low Power Range Monitor (LPRM) channels during outage II-43-88 Establish throttle capability 11-1-88 1-9-89 for the recirculation of cross tie valve during piping decontamination 11-44-88 Temporary disabled the generator 11-3-88 2-10-89 alteration cardox fire protection system II-45-88 Blind Flange cardox supply 11-3-88 2-10-89 line to facilitate removal of alterrex system II-46-88 Temporary removal of instruments 11-7-88 12-12-88 to facilitate core boring for repairs to the flued head 3286a 9
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'2.5 Temporary Systema Altdrctions.(Unit 2 and 3)
.b Temporary System-'
Ins tall'a tion Removal Alteration _Ho, description Dat0 Date
- II-47-88 Temporary insta11at[on,of.;
11-7-88.
2-3-89 of-blank flange oil s'upply.
line to'Alterrex system-
'II-48-88 Alteration of feed cables =for 11-8-88 1-19-89 torus recoat project 11-49-88
' Utilization of a spare breaker 11-11-88' 1-7-89
'at Bus 23-1 to supply power to recirculation piping deconination equipment II-50 Prevention of rod block 11-11-88 1-7-89 occurrence from control Rod Position Indication System (RPIS) w' nile fuel'is out of core II-51-88 Temporary system alteration 15-88 12-7-88 for alternate cooling during service water outage II-52-88 Temporary connection of pipe 11-17-88 12-9-88 between 2-1501-4A Reactor
. Building Closed Cooling Water System (RBCCW) heat exchanger for service water outage.
k II-54-88 Installation of jumper to bypass 11-17-88 12-6-88 fuse to feed. power to LPCI/CCSW Heat' Exchanger AP controller 3A during dual unit outage-11-55-88 Temporary modification of 11-18-88 12-3-88 temperature indicating controller to allow for manual throttling of throttle control valve, during service water outage l
.II-56-88 Temporary modification of 11-18-88 12-3-88 temperature indicating controller to allow for manual throttling of throttle control valve, during service water outage l
3286a l'
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0 2.5 Temporary Systems Altarctions (Unit 2'cnd 3)
Temporary Eystem Installation Removal A.lteration Not Description Dalf Qale 11-57-88 Jumpers'placed to prevent 11-21-88 2-11-89 group II isolations from Drywell Radiation Monitors II-58-88 Provide power to paint storage 11-25-88 2-11-89 trailer for torus recoat project II-59-88 Placement of jumper cable for 11-29-88 1-8-89 Units 2 and 3 refueling. platform hoists to support CRD guide tube vacuuming 11-60-88 Removal of control power to the 12-13-88 1-19-89 primary containment air sampling valves to facilitate rearrangement of the control switches as per modification on panel 902-3 II-61-86 Insertion of a mechanical stop 12-14-88 12-20-88 around valve stem of semp discharge valve to prevent closing II-62-88 Insertion of a mechanical stop 12-14-88 12-20-88 around valve stem of sump discharge valve to prevent closing 1
II-63-88 Installation of an air source 12-15-88 12-20-88 for the air operator of valve 2-8941-727 to perform maintenance on CRD scram air 4
pressure regulator II-64-88 Installation of an air source 12-15-88 12-20-88 for the air operator of valve
[
2-8941-728 to perform
}
maintenance on CRD scram air pressure regulator, II-65-88 Installation of an air source 12-15-88 12-20-88 for the air operator of valve 2-8941-729 to perform maintenance on CRD scram air presrure regulator 11-66-88 Block / gag open the valve to 12-28-88 1-8-89 enable drywell exhaust ventilation path 3286a l
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2.5 Temporary Systcca Alt 3 rations (Unit 2 cud 3) 9
' Temporary.
' Installation Removal.
' System..
Description Egig Date Alteration No.
II-67-88_
. Block / gag open the: valve to
~12-25-88 1-8-89 enable drywell exhaust..,
ventilation-path 11 -89:
II-68-88; LRemoval of normal electrical
~'12-19-88; 2
, feed and provide a temporary Lsupply to area radiation power supply 1705-7A;while Reactor Protection System (RPS) bus 1 is 00S II-69-88
' Removal of normal electrical-12-30-88.
1-2-89 feed and provide a temporary.
supply.to area radiation-power _ supply 1705-7B while RPS bus Breaker 1 is 00S 3286a 8
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V,.6 2.5 Tempor:ry Systems Alterations (Unit 2 cnd 3)
Temporary System Installation Removal-1 Alteration _Hai
. Description Dalm-Dale
'III-1-88 Prevent RWCU isolation while 1-29-88' 1-29-88.
calibrating temperature switch-on_RWCU non. regenerative heat exchanger III-2-88 Replace defective heat 1-20-88 2-10-88 temperature switch on RWCU III-3 Replace indicator on Panel 903-3 3-23-88 3-23-88 III-4-88
' Disconnect.LPRM cables from 3-29-88 6-10-88 APRM at panel 903-37 for work to be done on CRD platform III-5-88' Disconnect LPRM cables from 3-29-88 6-10-88 APRM #2 at panel 903-37 for work to be done on CRD platform III-6-88 Disconnect LPRM cables from 3-29-88 6-10-88 APRM #3 at pane 1L903-37 for work to be done on CRD platform
'III-7-88 Disconnect LPRM cables from 3-29-88 6-10-88 APRM #4 at panel 903-37 for work to be done on CRD platform-
'III-8-88 Disconnect LPRM cables from 3-29-88 6-10-88 APRM #5.at panel 903-37 for work a
to be done on CRD platform
)
III-9-88 Disconnect LPRM cables from
'3-29-88 6-10-88
(
APRM #6 at panel 903-37 for work to be done on CRD platform III-10-88 Disconnect LPRM cables from 3-29-88 6-10-88 Group I Rod Block Monitor at panel 903-37 for work to be done~on CRD platform IIT-11-88 Disconnect LPRM cables from 3-29-88 6-10-88 Group II Rod Block Monitor at panel 903-37 for work to be done on CRD platform III-12-88 3C Reactor Building supply fan 4-4-8d 5-28-88 breaker, sent to GE for overhaul III-13-88 3A Reactor Building supply fan 4-4-88 S-28-88 breaker, sent to GE for overhaul 3286a
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- l y
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' 2.5 Temporary Systems Altar 2tions (Unit.2 cnd.3)
. ;a
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- Temporary 3-
' System-Installation Removal Altiration No.
Description-Dalm Date' LIII-14-88 Installation of jumpers on 4-7-88 5-24-88 panel 903-28'to prevent restriction of control-rod movement.
III-15 Installation of jumpers at 4-7-88 5-15-88
. panel 903-17;for work to be performed on scram contactors (590-109B and 590-109D) while fuel is out of Reactor vessel.
III-16-88 Installation of jumpers at 4-7-88 5-15-88 pane 11903-17. for work to be performed on scram contactors
'(590-108B and 590-108D) while fuel is out of Reactor vessel
-III-17-88 Insta11ation'of jumpers at 4-7-88 15-88' panel'903-15.for work to be performed on scram contactors
-(590-108E and 590-108G) while-fuel is out of Reactor vessel-III-18-88 Installation of jumpers at 4-7-88 5-15-88 panel 903-15 for work-to be performed on scram contactors
.(590-109A'and 590-1090) while fuel is out of Reactor vessel.
III-19-88 Installation of jumpers at 4-7-88
. 5-15-88 panel 903-15 for. work to be performed.on scram contactors (590-108A and 590-1080) while fuel is out of Reactor vessel.
1 III-20-88 Installation of jumpers at 4-7-88 5-15-88
~
panel 903-17-for work to be performed on scram contactors (590-108F and 590-108H) while fuel is out of Reactor vessel III-21-88 Installation of jumper on control 4-13-88 4-14-88 room panel 903-5 to keep relay 595-147 energized to replace Standby Liquid Control (SBLC) control switch III-22-88 Installation of jumper on control 4-13-88 4-14-88 room panel 903-5 to keep relay 595-147 energized to replace Standby Liquid Control (SBLC) control switch 3286a j
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- 2'.5 Tempor:ry Sy tems Altsrctions (Unit. 2 cnd 3) j N#
.' Tempora ry -
s System
-Installation Removal Alteration No.
Description Date Date III-23-88 Take CAM heat trace for' torus 4-14-88 9-88 systems 00S to reroute CAM piping during refuel outage 111-24 Repair wire SLA in penetration 4-15-88 4-18-88 X204-m in Unit 3 tip room III-25-88 Paralleling busses 35, 36 and 4-28-88 6-20-88 77-(when busses 33 and 34 are taken 003) for breaker and relay maintenance III-26-88 Paralleling busses 38 and 4-30-88 6-20-88 39 with both busses energized (when busses 33-1 and 34-1 are taken OOS) for breaker and relay maintenance III-27-88 Provjde alternate feed to 5-5-88 5-6-88 radiation monitors while testing RPS MG Set output breaker III-28-88 Provide alternate feed to 5-6-88 5-10-88 radiatioa monitors while testing.RPS MG Set output breaker III-29-88.
Installation of jumpers on 5-10-88 5-l'2-88 panels 903-14 and 903-17 to prevent group II and.III isolations, caused by deenergized panels 2203-73A and 73B III-30-88 Installation of jumpers at Panel 5-16-88 5-18-88 (Card #1) 903-17 from HFA relay 590-110B (term. 1) to HFA relay 590-116B (term. 3) for work performed on scram contactors, while fuel is out of Reactor vessel III-30-88 Installation of jumpers at Pansi 5-16-88 5-18-88 (Card #2 903-17 from HFA relay 590-124B (term. 2) to HFA relay 590-107D (term. 4) for work performed on scram contactors, while fuel is out of Reactor vessel l
3286a u
g-Lv? ' ?#h W j f. [ 2i5.' Temporary' Systems Altsrcti:na (Unit 2 and 3)'
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. Temporary..
l _,
System Installation Removal l'
Alleration Noi Description Date' Dale
^'
E
'III-30-88
. Installation of jumpers at. Panel 5-16-88 5-18-88
.(Card'#3
'903-17Efrom HFA relay 590-124F
.(terminal 2).to HFA relay 590-107H (terminal' 4) for work performed on -
scram contactors, while fuel is o
out of Reactor vessel
~
.III-31-88 Installation of jumpers at panel 5-18 5-24-88 903-15 for work to be performed on scram contactors (590-109A and--590-1090) while fuel is out of Reactor vessel
.III-32-88
-Installation of jumpers at panel 5-18-88 5-24-88 903-15 for, work to be performed on scram contactors (590-108E and 590-109G) while fuel is out
.of Reactor vessel III-33-88 Installation of jumpers at panel 5-18-88 5-24-88 903-15 for work to be: performed on scram'contactors (590-108A and 590-109C) while fuel is out of Reactor vessel III-34-88 Clearance of half scram from "C" 6-10-88 6-14-88 Main Steam Line (MSL) radiation monitor 3286a
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_.,[_y[2.59Tempor:rySystemsAltarctions(Unit 2cnd3)-
d#f TemporaryJ System:.
Installation Removal
. Alteration No.
. Description Dalg Date III-40-88 Clear CRD high temperature' 6-22-88
~
alarm, caused by. failed-thermocouple and/or failed cable-connectors.
~
III-42-88 Installation of jumper on 6-28-88
- 7-1-88 failed MSL temperature switch
- 3-261--15D) to prevent half.
Group I isolatio1 during repairs.
III-43-88
. Installation of jumper on 6-30-88 7-1-88 failed MSL temperature switch
- 3-261-18D) to prevent half-Group'I isolation during repairs.
111-44-88 Insta11atiion of jumper on 7-4-88 7-15-88 failed MSL temperature switch
- 3-261-17D) to prevent half Group I' isolation during repairs.'
III-45-88 Installation of jumper on 7-5-88 7-15-88
~ failed MSL temperature switch
- 3-261-16B) to prevent half
' Group'I isolation during repairs.
III-46-88 Installation of jumper on 7-7-88 7-15-88 failed MSL temperature switch
- 3-261-18D) to prevent half Group I isolation during repairs.
III-47-88 Check calibration of HPCI area 7-23 7-27-88 temperature switches III-48-88 Disconnection of lead from A 7-30-88 11-29-88 hood alarm pressure switch, to prevent masking of alarms at panel 903-34 3286a
.e $;
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Temporary System Insta11ation' Removal-Alteration Noi:
Description.
Dalg ;
Date III-49-88 Installation of' jumper:on 8-17-88:
8-17 Panel 903-18 to bypass'high temperature isolation to stroke valves III-50-88 Removal of HPCI temperature 8-29-88 8-31-88 switches.for calibration by jumpering the temperature switch contacts to prevent any possibility of a spurious isolation of the HPCI system III-51-88 Installation of jumpers on 10-1-88 10-4-88 HPCI area temperature switches to allow-the switches to be
. removed for calibration.
III-53-88 Temporary feed to Unit 2's 10-27-88 10-30-88 (bus 33-1) torus recoat air compressors III-54-88 Temporary installation of a 10" 11-17-88 12-7-88 pipe between the 3-1501-4B valve and the RBCCW heat exchanger for dual unit service water outage-III-55-88 Temporary connection of a 4" 11-17-88 12-7-88 hose.between the 3A/3B TBCCW heat exchanger and the fire header for the service water outage III-56-88 Temporary modification of the 11-18-88 12-3-88 temperature controller to allow for the manual-throttle of TCV valve 'during service water outage
-111-57-88 Disconnection of power supply to 11-28-88 12-7-88 the main turbine diff erential expansion detector for maintenance on the turbine bearings 3286a
v-guem 10 2
- - -o A].
Commonwealth Edison g
.d Dresden No le:r Power Stati:n
/
$-o R.R. C1 G
Morris, tilinois 60450
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Telephone 815/942-2920 March 15, 1989 EDE LTR: #89-236 U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Document Control Desk
Subject:
1988 Annual Report Dresden Nuclear Power Station Commonwealth Edison Company Docket Nos. 050-010, QjiO-237 and 050-249 Gentlemen:
Enclosed please find the 1988 Annual Report of Changes._ Tests. and Experiments per 10 CFR 50.59 for Dresden Nuclear Power Station.
ConunQnwealth Edison Company.
This report is submitted pursuant to Title 10, Code of Federal Regulations, Part 50, Section 50.59(b)(2).
Please direct any communications concerning this report to G. Paramore of the Dresden Technical Staff at extension 2364.
Sincerely, g?
E. D. Eeni rg j
Station Manager Dresden Nuclear Power Station EDE:GP:jt cc:
A. Bert Davis Regional Administrator Directorate of Regulatory Operations Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137 S. Dupont USNRC Region III Dresden Site Senior Resident Inspector G. Paramore R. Whalen File / Tech. Staff File /NRC File / Numerical 3286a
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