ML19340A525
| ML19340A525 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/17/1960 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19340A524 | List: |
| References | |
| PROC-600317, NUDOCS 8008080718 | |
| Download: ML19340A525 (5) | |
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March 17, 1960 C
TESTING PROGRAM FOR POWER OPERATION 4
OF THE DRESDEN NUCLEAR POWER STATION I.
INTRODUCTION In conformance with the provisions of License No. DPR-2, issued to Commonwealth Edison Company on November 16, 1959, Docket No. 50-10, the following report is submitted describing the test-ing program planned to be conducted during power operation of the Dresden Nuclear Power Station, after installation of the reactor vessel head to completion of testing at 315 MW thermal (one-half rated thermal power).
II.
TESTING PROGRAM All standard operating procedures developea prior to and during the previous testing phases will be followed in the testing program during power operation, subject to modification from time to time as results obtained during this phase of the testing program may indicate to be desirable or necessary.
1.
Power 0: Pressure 0:
The procedure at this condition provides for:
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Jy ' : A.
Determining the primary loop operating characteristics for o
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- >; f'; open and closed loops.
The measurements made will provide
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operation.
During this phase control blades will remain
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1hlly inserted in the core.
B.
Start of control drive testing, continuing throughout the n
test program, to verify that for all reactor pressures the following specifications maintain:
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Maximum time from receipt of scram signal to 10% control rod travel 0.6 sec.
90% control. rod travel 2.5 sec.
2.
Power 0 to 10 MW (t): Pressure 0 to 975 psig at Primary Steam Drum (1000 psig Reactor Pressure): Steam Flow 0:
This phase is for the purpose of initial nuclear heating, calibrating the in-core monitors, and correlating in-core and out-of-core nuclear instrumentation.
The tests listed below will each be conducted at several temperature steps between the cold condition temperature and rated temperature.
A.
In-core wire irradiations B.
Temperature coefficient measurements C.
Primary loop flow vibration and expansion measurements D.
Radiation survey E.
Control rod scram F.
High pressure scram G.
Emergency conden ser H.
System heat loss Initial nuclear heating will be established by observation of in-core, out-of-core and loop temperature instrumentation.
The temperature of the syst c ' will be altered in eight st eps to 470 F (500 psig).
Combinations of tests listed above will be conducted at each step to assure safe operation at the next condition as. specified in the Power Operation Section of Appendix A to License No. DPR-2.
From 500 psig to 975 psig at the primary steam drum the heating rate will be governed..
-s by a limited pressure rise and limited by pressure regulator control set up in 100 psi increments; for each increment control rod scram performance will be recorded.
3 Power LO to 315 MW ( t) - Pressure 975 esiz: Primary Steam Flow-120.000 to 1,200,000 lb/hr. : Secondary Steam Flow at Minimum.
The reactor power will be increased to approximately 40 MW (t) with steam being utilized in part to roll the turbine and the remainder bypassed to the condenser.
Power level increases will be made in steps from the heating power of 10 MN (t) to insure satisfactory system operation.
l The test program during this phase involves operating the reactor at 1000' psig from a power level of approximately 40 MW (t)
(no voids in the core.1 to 315 MW (t) (half power ).
Operation will be with a minimum secondary steam flow, to provide rated voids at less than half rated power.
There will be five steps (minimum) in this increase to half power.
Data taken at each step will be extrapolated to higher power to verify that the plant will operate satisfactorily at higher powers as specified in App endix A to the License.
The turbine and its auxiliaries will be started during this period for the first time using nuclear steam.
Specific tests to be conducted and/or data to be taken at each of the power levels specified above include:
A.
Radiation Level Measurements B.
Reactor Recirculating Pump Tests i
C.
Pressure Regulator Tests j
D.
System Transient Tests
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E.
Radio-Chemical Tests F.
Stability Tests G.
System Expansion and Vibration Measurements H.
In-Core Monitor Calibration 4.
Power 130 - 315 MW (t): Pressur.e 975 usig: Primary Steam Flow 450,000 to 900.000 lb/hr.: Secondary Steam Flow Minimum to 590,000 lb/hr.
During this phase of the test program dual cycle control will be tested.
The selected steps provide exploration o f those portions of the primary-secondary flow which are permitted by turbine restrictions and the limitations o f the present 315 MW (t) license.
The same specific tests mentirned in (3) above will be conducted at 3 (minimum) conditions of primary to secondary steam flow rati o.
In addition, several power level changes will be made with secondary steam flow changing due to load change demand.
Stcady state data will be recorded to assure consistent
" tracking" on the dual cycle partial load characteristic curve 5
Power 315 MW (t): Pressure 975 osiz: Primary Steam Flow 1,200,000 lb/hr.: Secondary Steam Flow Minimum.
This phase is a sustained steady state power run at half rated power and simulated rated core void content.
The purposa of the test is to provide relatively long irradiation of the fuel at simulated rated power distribution.
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Following shut down a thorough power distribution study will be made by:
A.
A gam =a probing scan of the core to provide gross power distribution data.
B.
A gamma scan of a special fuel assembly that can be disassembled to provide information on local power peaking factors.
These data will be used to re-evaluate the in-core monitor technique as a means of assuring operation within thermal output limitations.
During operation, wire irradiations will be made to re-calibrate the in-core monitors.
This phase provides the first opportunity to check much of the radiochemical prccess equipment under equilibrium conditions.
NOTE:
All power levels, temperatures, steam flow rates, etc.
in the foregoing are to be considered as estimated approximate values.
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