ML20206M402
| ML20206M402 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/01/1986 |
| From: | Robey R COMMONWEALTH EDISON CO. |
| To: | Case E Office of Nuclear Reactor Regulation |
| References | |
| RAR-86-24, NUDOCS 8608210175 | |
| Download: ML20206M402 (8) | |
Text
e Commonwealth Edison 6 Telephone 309/654-2241 ouad Cities Nuclear Power Station 22710 206 Avenue North Corcova, Illinois 61242 p+Y
/
D, h S
RAR-86-24 August 1, 1986 Mr. Edson G. Case, Deputy Director Office of Nuclear Reactor Regulation U. 5. Nuclear Regulatory Commission Washingtcn, D. C.
20555
Dear Mr. Case:
Enclosed please find a listing of those changes, tests, and experiments completed during the month of July, 1986, for Quad-Cities Station Units 1 and 2, DPR-29 and DPR-30. A summary of the safety evaluation is being reported in compiiance with 10 CFR 50.59.
Thirty-nine copies are provided for your use.
Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION f
I R. A Services Superintendent bb Enclosure cc:
J. Wojnarowski J. Leider/E. Budzichowski
%8 hC R
0027H/0061Z j
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Modification M-4-1-83-7 Description This modification reorientated the 1-2001-3 and 1-2001-4 Drywell Floor Drain Sump Discharge Isolation Valves. Before, the valves were mounted upside down, and sediment collected in the valve internals, and inhibited valve stem movement.
In order to move the valves to the up-right position, the piping configuration had to be changed slightly.
However, the valves remain in the same approximate location.
Evaluation By orientating the 1-2001-3 and 1-2001-4 Drywell Floor Drain Sump Discharge Isolation Valves to the upright position, the valves are made more reliable. This will assure the valve's operability during Drywell Floor Drain Sump operation, and proper closure during an isolation condition.
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o Modification M-4-1(2)-85-14 Description This modification installed a 50 micro-farad capacitor across a feedback resistor in the amplifier circuit of the indicator trip unit of the Fuel Pool Process Radiation Monitors. The purpose of this modification was to add a time delay constant to the amplifier circuit to eliminate spiking and erratic operation of the radiation monitors.
Evaluation This modification will reduce spurious alarms and Standby Gas Treatment starts caused by voltage spikes and circuit noise. The ability of the monitors to detect a high radiation condition, to cause alarms, and to initiate Standby Gas Treatment remains the same.
A capacitor open circuit failure would have no affect on the amplifier.
A short circuit failure would cause the downscale alarm to sound, which is a conservative action.
m.
- o SPECIAL TEST 1-82
~0n July 21, 1986, Special Test 1-82 was completed to provide comparative pump data-based on differing flows for cavitation problems on RHR Service Water Pumps being investigated by Station Nuclear Engineering Department.
Safety Evaluation 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 three other RHR Service Water Pumps will be operational at the times of the test. This will be proven by running operability surveillances as required by limiting conditions for operation (Technical Specification 3.0/4.0) prior to test.
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 test is the same as the flow rate testing operations test (QOS 1000-4) with the exception of having the room cooler valved out and the opening of the 1001-3A(B) discharge valve.
The RHR System will not be placed in any unsafe configurations. -An Operator will be in attendance to valve in the room cooler should the need arise. Also, pressure gauges will be installed to the inlet &
outlet sides of the primary pump. This will not adversely affect the performance of the pump.
3.
The-margin of snfety, as defined in the basis for any Technical Specification, is not reduced because this test requires only the changing of valve positions and no physical changes will be made to the pump itself, so the margin of safety will remain unchanged.
i-SPECIAL TEST l-93 Special Test 1-93 was completed on July 11, 1986. The purpose of this test was to verify the functions of the K1 and K2 Relays for IRM Channels 11 through 18 power supply, while pulling the 24V DC fuse.
Safety Evaluation 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 this test will be performed while the mode switch is in the RUN mode. Therefore, detection and indication of neutron flux level is provided by the power range instrumentation.
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 test simulates a 24V DC fuse blowing and will verify that IRM logic functions as designed. All Reactor Protection Systems will be intact during this test.
3.
The margin of safety, as defined in the basis for any Technical Specification is not reduced because, while in the RUN mode, the IRM Reactor Protection logic is bypassed by the control switch and APRM logic.
1
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SPECIAL TEST 2-58 on July 3, 1986, Special Test 2-58 was completed. This test performed a temperature survey of the 2B RHR room. The test was performed while the 2C and 2D RHR Pumps were running, and while the ventilation and room cooler were off.
Safety Evaluation 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 this test is only a temperature survey. During the test, temperature will be constantly monitored to ensure that no temperature limits are exceeded. The test will not affect the system.
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 test is a temperature survey. -During the test, temperature will be constantly monitored to ensure that no temperature limits are exceeded. The tesc will not affect the system.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because this is a temperature survey.'
All limiting operating conditions for operation, as' stated in the Technical Specification were met prior to beginning the test.
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SPECIAL TEST 2-60 Special Test 2-60 was completed on July 2, 1986. The purpose of this test was to verify the functions of the K1 and K2 Relays for an IRM Channel 11 power supply, while pulling the 24V DC fuse.
Safety Evaluation 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 this test will be performed while the mode switch is in the RUN mode. Therefore, detection and indication of neutron flux level is provided by the power range instrumentation.
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 test simulates a 24V DC fuse blowing and will verify that the IRM Channel 11 logic functions as designed.
All Reactor Protection Systems will be intact during this test.
3.
The margin of safety, as defined in the basis for any Technical Specification is not reduced because, while in the RUN mode, the IRM Reactor Protection logic is bypassed by the control switch and APRM logic.
m s a SPECIAL TEST 2-61 On July 11, 1986, Special Test 2-61 was completed. This test consists of pulling the 24V DC fuse and verifying the functions of K1 and K2 Relays for all IRM's.
Safety Evaluation 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 this test will be performed while the mode switch is in the RUN mode. Therefore, detection and indication of neutron flux level is provided by the power range instrumentation.
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 test simulates a 24V DC fuse blowing and will verify that the IRM logic functions as designed. All Reactor Protection Systems will be intact during this test.
3.
The margin of safety, as defined in the basis for any Technical Specification, is not reduced because while in the RUN mode, the IRM Reactor Protection logic is bypassed by the control switch and APRM's.