ML20245A166
| ML20245A166 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/03/1989 |
| From: | Robey R COMMONWEALTH EDISON CO. |
| To: | NRC |
| References | |
| RAR-89-19, NUDOCS 8904250084 | |
| Download: ML20245A166 (11) | |
Text
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fD CommonwyIth Edis n
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I; 22710 206 Avenue North -
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I Cordova, Illinois 61242 9740 -
Telephone 309/654-2241 RAR-89-19 April 3, 1989 Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Mail Station Pl-137 Hashington, D. C.
20555 Enclosed please find a listing of those changes, tests, and experiments completed during the month of March, 1989, for Quad-Cities Station Units 1 and 2, DPR-29 and DPR-30. A summary of the safety evaluations are being reported in compliance with 10 CFR 50.59.
Thirty-nine copies are provided for your use.
Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION
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.Q R. A.
bey Services Superintendent
.RAR/vmk/eb Enclosure cc:
R. Stols T. Hatts/J. Galligan
[h 8904250084 890403 PDR ADOCK 05000254 i
R PDC I \\
0027H/0061Z
SPECIAL TEST 1-126 Special Test No. 1-126 was completed on March 7, 1989. The purpose of this test involved observation of control rod movement at 260 psid and 320 psid drive differential pressure which was more extensive testing in response to Special Test No. 1-124, 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because rod manipulations were conducted in accordance with existing approved procedures.
2.
The probability for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the test was conducted under normal operating practice and procedure so any accident or malfunction was no different than previously analyzed.
3.
The margin of safety, as defined in the basis for any Technical Specifi-cation, is not reduced because no condition which could render the control rods inoperable was specified or required or expected for this test.
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SPECIAL TEST 2-90 Special Test No. 2-90 was completed on March 4, 1989.
The purpose of this test was to identify any missing or degraded balls in the ball check valve 305-115 (charging water check valve). At cold conditions with all rods fully inserted, the CRD pump was turned off and results identified immediately.
1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because in Section 10.5 the CRD pump is simply to supply pressure for charging the scram accumulators. At cold shutdown with the rods fully inserted the accumulators were not required.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the CRD pump is shutdown as part of normal operating practice during refuel outages when not required.
3.
The margin of safety, as defined in the basis for any Technical Speci-fication is not reduced because with the reactor in cold shutdown and control rods fully inserted this test (by tripping the CRD pump) will not reduce the margin of safety, i
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h Modification M-4-2-79-002 Description Modification M-4-2-79-002 was to install anticipated transient without scram-recirculation pump trip equipment (ATWS-RPT).
This system is designed to shutdown the reactor in-case of transients without a scram occurrence.
The ATWS will consist of two pressure transmitters, two water level transmitters and electronic trip units. The electronic trip units are initiated by the water and pressure transmitters and they in turn energ*.ze the trip relays.
The relay contacts are arranged in two out of two configuration to trip the field breakers of both recirculation pumps. The field breakers will be provided with two trip coils, one for each division.
Evaluation The addition of pressure and water level transmitters into the existing pressure lines does not create any new accident or malfunction since the RPS (Reactor Protection System) remains unchanged and RPT (Recirculation Pump Trip) is a redundant back-up to RPS system.
The RPT operation results in a power reduction aiding safe shutdown. This modification will increase the safety of the system.
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Modification M-4-0-83-4 1
Description i
This modification added an alarm in the Radwaste control room to notify the operator when the Laundry Sample Tank is 95% full. This minor change to the Quad Cities radioactive waste system is reported to the NRC even though no change is required to Station Technical Specifications or FSAR.
Evaluation The change reduces the probability of an accidental spill or localized plant flooding.
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Modification M-4-1/2-83-14 Description This modification was initiated to provide isolation capability for the HVAC fan and the cooling water pump for the 1/2 Diesel Generator. This modi-fication was initiated to bring the station into compliance with the require-ments of Section III.G cf 10CFR50 Appendix R,
' Fire Protection Program for Operations Nuclear Power Plants'.
The isolation of a fire detection relay contact and a control cable which trip the HVAC fan on the detection of a fire was accomplished by the installation of an isolation switch at the 1/2 Diesel Generator auxiliary control panel. The isolation of the redundant power feeds for the 1/2 Diesel Generator cooling water pump was accomplished
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by installing a transfer panel which automatically connects the energized
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power feed to the pump while isolating the alternate feed.
j Evaluation The affect of this modification was to reduce the possible consequences of a fire in various zones of the plant by increasing the availability of the auxiliary equipment associated with the 1/2 Diesel Generator. This increases the reliability of the on-site power system and, therefore, improves the margin of safety of the plant.
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1 Modification M-4-1(2)-83-049 Description As a result of Bulletin 7901B (environmental qualification or electrical equipment) several flow, level, differential pressure and pressure switches were replaced by a Rosemount transmitter / trip unit scheme. The transmitters were installed in the reactor protection, primary containment isolatio.4, residual heat removal, core spray, and high pressure coolant injectica systems.
The transmitters are located in areas considered to be harsh environments I
and the trip units are located in a mild environment.
All equipment. was installed seismically and divisionalized. No setpoints were changed and all systems function as per original design.
Evaluation The margin of safety during regular operations has not changed as a result of this modification.
The margin of safety as postulated during an j
accident, however, has increased significantly.
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Modification M-4-1(2)-84-13 Description This modification replaced the original Unit One and Unit Two refueling bridges with new refueling bridges. The replacement was performed to provide faster and more reliable refueling bridges so as to reduce the time required for refueling. The east rail was also replaced to provide a larger size and more level rail.
Evaluation The new refueling bridges have the same safety features as the original refueling bridges and they use the same existing refueling interlocks as addressed in the Technical Specifications.
The functions of the new bridges are the same as the functions of the old bridges, and therefore, the design basis refueling accident does not change.
Replacing the east rail with a larger size increased the seismic ability of the new refueling bridges.
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Modification M-4-1/2-85-007 Description 1
The existing General Electric CFD Diesel Generator (1/2) differential current protection relay was replaced with a new seismically qualified Westinghouse SA-1 type differential relay in order to satisfy OPEX 84-7551.
The new relay is in the same phyiscal location (at the 4KV switchgear) as before. The relay continues to provide trip signal to the lockout relay to disconnect an internally faulted Diesel Generator from its 4KV switchgear bus.
Evaluation The new SA-1 relays provide the same protection as the old CFD relays, but with seismic qualification to ensure operation capability before, during and after a seismic event.
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o Modification M-4-2-85-026 Description The existing General Electric CFD diesel generator.(Unit 2) differential current protection relay was replaced with a new seismically qualified Westing-house SA-1 type differential relay in order to satisfy OPEX 84-7551. The new relay is in the same physical location (at the 4KV switchgear) as before.
The relay continues to provide a trip signal to the lockout relay to disconnect an internally faulted Diesel Generator from its 4KV switchgear bus.
Evaluation I
The_new SA-1 relays provide the same protection as the old CFD relays but with seismic qualification to ensure operation capability before, during and after a seismic event.
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Modification M-4-2-86-005 Description Modification M-4-2-86-005 was to install a test tap and manual isolation valve on line 2-8806-2"L for the purpose of local leak rate testing primary containment isolation valves A0-2-8803 and A0-2-8804 from the drywell side of the valves. This will significantly improve the testing method which is now being used. This involves adding a two inch globe valve in line 2-8806-2"L downstream of the oxygen analyzer system isolation valve A0-2-8804 with the addition of a 3/4" branch connection with two more 3/4" globe valves. Valves A0-2-8803 and A0-2-8804 will be relocated upstream on line 2-8806-2"L to create a sufficient length of pipe for the additional valve and tee connection.
Evaluation No change in the function of the primary containment isolation valves A0-2-8803 and A0-2-8804.
The material and installation procedures meet or exceed the original code of construction. Therefore, the possibility of an accident or malfunction is of the same type that has previously been evaluated in the FSAR.
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4 Safety Evaluation #89-102 Modification to ODCS C-Model Program Changes are to be made in the ODCS C Model calculational model. These changes allow the calculations to incorporate time following shutdown and release duration into the dose calculations. This will allow more accurate dose projections.
1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated l
l in the Final Safety Analysis Report is not increased because the change does not involve plant equipment. The calculation model will provide more accurate dose projections for post accident assessment.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because this program is only used to assess conditions.
Improving the calculations will allow more accurate projections.
This will not create any unanalyzed transients or malfunctions since it does not effect plant equipment.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because post accident off-site dose projection is not addressed in the Tech. Specs.
Improving these calculations will not reduce the margin of safety.
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