ML19345D687
| ML19345D687 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 12/01/1980 |
| From: | Aliviankis N COMMONWEALTH EDISON CO. |
| To: | Case E Office of Nuclear Reactor Regulation |
| References | |
| NJK-80-479, NUDOCS 8012160479 | |
| Download: ML19345D687 (4) | |
Text
.
Commonwealth Edison l
/. Oi Quad Cities Nuclear Power Station (y g, 22710 206 Avenue North s
Ne.../ <
Corcova. INinois 61242 Telephone 309/654-2241 NJ K-80-479 i
i i
De ember 1, 1980 i
l L
i l
Mr. Edson G. Case, Deputy Director j
Office of Nuclear Reactor Regulation
~
U. S. Nuclear Regulatory Commission Washington, D. C.
20555 1
c
Dear Mr. Case:
Enclosed please find a listing of those changes, tests, and 1
experiments completed during the month of November, 1980, for Quad-Cities Nuclear Power Station Units I and 2, DPR-29 and DPR-30. A summary of the safety evaluation is being reported in compliance with 10 CFR 50 59 Thirty-nine copies are provided for your use.
Very truly yours, l
COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION W lA* N'I Y
,Y
<c N. J. Kalivianakis Station Superintendent NJK/bb Enclosure cc R. F. Janecek m
g
'M A
ty u
5; t.
- c? U
($ }t/df, r: r:
- * :.; w
~-
c
. e, L".
s i
a i.
E 80121 eo179
/
M-4-1-76-21 Containment Spray Permissive Switches Description of Modification This modification changes the alarm function of a Control Room annunciator.
This removed a useless continuous alarm to an alarm which warns the operator of low drywell pressure, less than I psig.
The operator then knows when Containment Spray Mode of RHR is inhibited and the loss of Drywell to Suppression Chamber dif ferential pressure.
The wiring, changes the alarm contacts of the pressure switches from normally open to normally closed.
Summary of Safety Evaluation The RHR Logic remains unchanged therefore, the probability as evaluated in the FSAR remains unaltered.
Improvement is made by eliminating a use-less alarm with one that warns the operator of changing conditions.
The setpoint of the pressure switches will remain the same, therefore, no safety margin is reduced.
M-4-1-74-35 RHRS Service Water Isolation Valves Description of Modification This modification involved installation of manually operated butterfly valves downstream of the 1-1001-5A and 5B Residual Heat Removal System (RHRS) Service Water valves.
The new valves will be used as isolation l
valves to permit maintenance on either the 1-1001-5A or eB valves without j
draining the complete RHRS Service Water discharge header. The new valves i
will be locked open during normal system operation.
Summary of Safety Evaluation The new valves were installed in accordance with the original design requirements and will be locked open during normal plant operation.
Since the valves are passive components, the possibility of a different type of accident or malfunction is not created. The function of all other components in the RHR Service Water system will be unchanged.
l l
M-4-1-76-57 HPCI Start Circuitry Description of Modification The change will prevent the HPCI discharge valve from cycling when both high drywell pressure and high reactor water level occur simultaneously.
Should a high water level exist the HPCI turbine will trip and reset will be inhibited until the condition has cleared.
Summary of Safety Evaluation The intent of HPCI Logic is unchanged and the possibility of an accident has not increased.
The loss of HPCI is evaluated in the FSAR and no new consequences result from this modification.
M-4-1-78-16 Diesel Generator Auto-Start Test Switches Description of Modification _
The modification installed a test switch in both the Unit One and tne One/
Two diesel generator auto-start circuits. The test switches were in-stalled between the main feed air circuit breaker contact and the automatic start relay of the respective diesel generator.
The test switches will be used during surveillance testing of the diesels to verify that the diesels auto-start from bus undervoltage relay operation rather than from the breaker actuation.
The test switch will be in the closed position for normal plant operation, and will be opened only during surveillance testing.
Summary of Safety Evaluation The installation of the test switches will not change the operation of the system as previously evaluated.
The emergency diesel generators will main-tain the same standby function and installation of the test switches does l
I not alter the overall safety of the system.
The Technical Specification requirements concerning the system will ramain unchanged.
1 1
I
SPECIAL TESTS Special Test 1-49:
Core Spray Flow Test This test was performed to test the flow rate capacity of a core spray pump with its minimum flow valve failed open.
The results of the test were such that the Tech Spec flow and pressure requirements could still be met with the valve failed open.
Summary of Safety Evaluation Only one core spray pump was tested. The other 100 percent core spray loop, LPCI, HPC I, Auto-Blowdown, and RC IC were operable.
The design, function, and intent of the core spray system were not altered.
The system was not operated in an unsafe manner, and was still fully capable of adding water to the reactor vessel, if needed.
Special Test 2-32:
LPCI Flow Test This test was performed to test the flow rate capacity of the LPCI Mode of RHRS with the pump minimum flow valves failed open.
The results of the test were such that the Tech Spec flow and pressure requirements for LPCI could still be met with the valves failed open.
Summary of Safety Evaluation Both of the core spray loops, HPCI, Auto-Blowdown, and RCIC were operable.
The design, function, and intent of tne LPCI Mode of RHRS were not altered.
The system was not operated in an unsafe manner, and was still fully capable of adding water to the reactor vessel, if needed.
9