ML20056B843
| ML20056B843 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/17/1975 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Boyd R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9102130362 | |
| Download: ML20056B843 (3) | |
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Y M r. R. S. Boyd, Acting Director
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Washington, DC 20555 D. 4, 1 3~r
Dear Mr. Boyd:
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MONTICELLO NUCLEAR GENERATING P1 ANT Docket No. 50-263 License No. DPR-22 d./
t Main Steam Line Flow Trip Setting During the Monticello startup program the pressure differential across the main steam line flow restrictors did not respond as expected.
During the interim period prior to identifying and correcting the problem, the trip settings were reduced to tested levels.
Since identifying the cause and making appropriate modifications to the flow nozzles, we have observed over three years of opera-tion with pressure differentials responding as expected.
During discussions on this matter the staff questioned the use of the 140% flow trip setting used in the safety analysis and in the Technical Specifications.
In a January 10, 1972 letter to Dr. P. A. Morris, we agreed to restrict the setting to 120% until the overall effect on plant operation could be determined.
Table 3.2.1 of the current Technical Specifications allow a setting of less than or equal to 140%
of rated flow.
Since making the commitment not to exceed a setting of 120%
while evaluating operation we have experienced a scram when one MSIV isolated during testing, an event the plant is designed to withstand without scram based on a 140% flow trip setting.
We also experienced a substantial loss of capacity by reducing power to prevent inadvertent scrams on MSIV exercise tests until in-stalling improved MSIV operators.
We have also been forced to reduce power to allow full stroke tests of MSIV's and turbine stop valves.
The safety evaluation presented below has been reviewed and found to fully support the 140% flow set-ting. We therefore intend to reset the main steam line flow trip setting from 120% to 140% during the refueling outage scheduled to terminate October 15, 1975.
The 120% high steam flow isolation signal was originally established on the two steam line plants, Oyster Creek and Nine Mile Point.
The basis for the setting was not due specifically to a safety limit but was set high enough to avoid spurious trips during nomal operation and low enough to minimize the con-sequences of breaks of any size in the main steam line for a plant of that steam line configuration.
The radiological consequences were calculated based on the technical specification MSIV closure time of 10 seconds.
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i 9102130362 750917 CF ADDCK 05000263 CF yi
NORTH N OTATED POWER COMP Y
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i Mr. R. S. Boyd 2
September 17, 1975 j
Monticello is a four steam line plant with a technical specification MSIV l
closure time of 5 seconds.
Ihus for Monticello, the basis for the selection i
of the 140% of rated steam flow as a requirement for the automatic isolation is that this setting permits the plant to continue to operate at full power with i
with one of the four main steam lines isolated.
This results in an average j
133% of rated steam flow in each of the steam lines.
a t
Large steam leaks (greater than 140% of rated flow) outside containment are i
j detected by the main steam line flow restrictor differential pressure sen-i sors.
If flow exceeds the trip setting in any line an isolation of all main i
steam lines will be initiated.
The MSIV's will close in 3 to 5 seconds af ter j
receiving the isolation signal. One of the design basis accidents analyzed in the Monticello FSAR is a guillotine break of one of the main steam lines. Choke flow of the two phase blowdown is assumed to exist until MSIV closure, which is conservatively assumed to be 10.5 seconds following the break, resulting in the loss of 85,000 pounds of coolant.
n e calculated thyroid dose at the site boundary is 80 rem relative to a 10 CFR Part 100 limit of 300 rem.
The whole body gamma dose is a much smaller fraction of the respective limit.
d Small steam leaks (less than 140% of rated flow) are detected by one of three 3
means. A leak in the main steam line tunnel would be sensed by temperature e
sensors; the response time for the sensor to a nearby steam line break would depend upon the size and location of the break in relationship to the nearest j
sensor. A leak in the turbine building would be detected by area radiation l
monitors.
These monitors would trigger alarms in the control room upon exposure 4
to high radiation. A leak of any significant size would be apparent to the operator because the control room indication of flow in one steam line would exceed that in the other lines and the electrical output would drop for a given i
j reactor power level.
The operator would respond to each indication and isolate i
the reactor if required.
W e maximum size small leak is limited by the main t
steam flow restrictor trip setting.
The small break analysis is based on the conservative assumption of a maximum size leak corresponding to a 140% trip setting.
This leak size is sufficiently small such that the blowdown consists of on'.y steam.
The iodine partitioning in steaming results in a smaller dose l
j for a given mass of coolant released.
If the maximum size leak exists for 5 minutes without being terminated by the operator 226,000 pounds of steam are released, resulting in a thyroid dose at the site boundary of 4.3 rem.
The whole body gam.a dose is again a much smaller fraction of its respective limit, 4
i j
The above analyses assumed reactor water iodine concentrations based on FSAR 1
Section 14.6.5.2 values ratioed to correspond to an offgas release rate of 480,000 microcuries per second.
The carry-over from reactor water steam was taken as 2%.
Regulatory Guide 1.5 meteorology was used in the inhalation dose evaluation.
The analysis assumed tha t there was no coincident loss of AC power and therefore the feedwater system would continue to make up water to the reactor vessel, i
Should a loss of AC power occur simultaneously with the break, a low water level trip would occur resulting in isolation and initiation of HPC1.
The change in main steam line flow restrictor trip setting from 120% to 140%
does not change the design base accident analysis reported in the FSAR.
The calculated dose for a small break based on the 140% trip setting is much less than the design base accident considered in FSAR section 14.6.5 and is less l
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N ORTH F.
4 CTATE] POWER COMP. tY Mr. R. S. Boyd 3
September 17, 1975 than 2% of the 10 CFR Part 100 limit.
We therefore believe the initial plant design based on a 140% trip setting is thoroughly justified in that it assures that a minimum dose to the general public during the highly unlikely hypothesized We believe the design trip setting of 140% of rated main steam line event.
flow is desirable in that it will reduce the likelihood of inadvertent scrams and allow an improved plant capacity factor.
For these reasons we intend to the main steam line flow restrictor trip setting to 140% during the reset current refueling outage which is presently scheduled to terminate October 15, 1975.
Yours very truly, g;qr,iC).
L. O. Mayer, PE Manager, Nuclear Support Services LOM/MHV/ deb cc:
J. G. Keppler G. Charnoff MPCA Attn:
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W.
Ferman
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