ML20062F600
| ML20062F600 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/07/1978 |
| From: | Olshinski J Office of Nuclear Reactor Regulation |
| To: | Israel S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20062F594 | List: |
| References | |
| NUDOCS 7812200087 | |
| Download: ML20062F600 (8) | |
Text
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UNITED STATES 8"
NUCLEAR REGULATORY COMMisslON g
WASHING TON, D. C. 20555 s.,..o /
DEC 7 1978 MEMORANDUM FOR:
S. L. Israel, Section Leader, Reactor Systems Branch, DSS FROM:
J. A. Olshinski, Reactor Systems Branch, DSS t
SUBJECT:
EXPECTED FLOW RATES AND PRIMARY SYSTEM TEMPERATURES UNDER NATURAL CIRCULATION CONDITIONS As requested, I have briefly reviewed the frequency and extent of natural circulation events that have occurred to date.
Printouts of natural i
circulation / loss of offsite power events were obtained from the-Office of Management and Program Analysis (NRC) and from the Nuclear Saf'5ty' Information Center (ORNL).
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The more significant of these events include a Farley Unit 1 event in which forced reactor coolant system flow was lost for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the St. Lucie Unit 1 event in which the primary coolant system was cooled down under natural circulation conditions.
Additionally, some data were available from natural circulation tests con-ducted at Maine Yankee, 'Calvert Cliffs Unit 2, Fort Calhoun, and Connecticut Yankee nuclear units. The data-from these tests ranged from very, detailed data tb data that included only one or two temperature points.
Although the test conditions varied significantly from test to test, the test data indicated a consistency in expected temperature differentials and natural circulation flow rates.
The Fort Calhoun test was initiated by tripping all reactor coolant pumps at a reactor power level of 35%.
Thirty minutes after the reactor coolant
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pump trip, all feedwater addition was secured to the steam generators.
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Data was recorded for 105 minutes after the reactor coolant pump trip.
i The Calvert Cliffs Unit 2 test was initiated from a reactor power level of 40%. Data were recorded for 45 minutes after the reactor coolant pump trip.
The Maine Yankee natural circulation test was initiated from 35% reactor power. The test was continued for 63 minutes.
The Connecticut Yankee test was initiated from hot standby conditions.
The results from the tests indicated that the initial core AT's increased for 10 to 15. minutes, then decreased to a stabilized level for the remainder of the test.
Figure 1 shows a plot of the core AT from time 0 to time 100 minutes for the Fort Calhoun and Calvert Cliffs natural circulation tests.
Since the detail of the test data reported in the 7812200037
S. L. Israel DEC 7 1978 startup test reports varied from test to test, it was not possible to plot core AT versus time for all the tests.
Descriptions of the test results for all the tests do indicate, however, that the measured core AT followed similar patterns for all the tests.
It is also significant to note that for each test, the maximum TH0T recorded during natural circulation conditions was less than the TH0T recorded imediately before the start of the test.
Table I shows the initial TH H0Trecordedduringeachofthenaturalcirculationte![s.
and the maximum T Table 2 lists the power / flow ratios recorded at various times during each of the natural circulation tests.
The power levels were calculated either by using decay heat rates or by measuring boiloff from the ste'aigenerators.
Flow rates were calculated utilizing loop transport time calctilations or (e
core AT calculations.
Power is expressed as percentage of full power and flow rate is expressed as percentage of full forced circulation coolant flow rate.
Since the Connecticut Yankee core nower history nrior to the start of the natural circulation -test is available, it is possible to calculate the expected natural circulation temperatures and flow rates assuming a loss of forced circulation flow at 100% reactor power.
The heat flux and flow rates can pe calculated ' utilizing the relationship that-Q2 = K(aT)3 where Q = decay heat rate AT = core differential temperature Assuming that TCOLD egals TSECONDAR H0T is calculated to be 588.80F f
1 minute after loss of all forced re.Y, T f
actor coolant system flow from 100%
power. Assuming a hot channel factor of 2.0, TH0T hot channe ) at 1 i
minute after the loss of forced flow is calculated ;{o be 629.6gF. which is below the saturation temperature of the primary safety valve set pres-sure and is therefore acceptable. The peak heat flux at this point is 0.012 BTU /hr-ft2x10-6 The average natural circulation flow velocity is 0.1358 lbm/hr-ft x10-6 based on these assumptions-2 Reference 1 addresses the critical heat flux in a heated bundle cooled by pressurized water.
Figure'C-19 (attached) from reference 1 presents a graph of critical heat flux versus average mass velocity for pressures about 2200 psia.
As can be seen from examination of Figure C-19, the peak heat flux of 0.012 BTU /hr-ft x10-6 remains well below the critical 2
heat flux.
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1 S. L. Israel DEC 7 1978 In summary, then, a number of plant events have occurred during which all forced reactor coolant system flow has been lost for significant lengths of time without any apparent adverse consequences. Additionally, a number of natural circulation tests have been conducted which indicated that the
% power /% flow ratio during the tests was about 0.2.
Based on these tests, calculations indicate that neither burnout nor loss of reactor coolant inventory due to lifting of primary safety valves would be expected to occur during natural circulation conditions following a loss of all forced reactor coolant system flow at 100% power.
a - QM =f
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John A. 01shinski Reactor Systems Branch Division of Systems Safety Attachments:
1.
Table 1 2.
Table 2 3.
Reference 4.
Figure 1
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5.
Figure,C-19 cc:
T. Novak C
a t
TABLE 1 Temperatures Recorded During Natural Circulation Tests Initial T g Maximum TH0T Prior to test Recorded during test 0
Fort Calhoun 545 F 543 F 4^::
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0 Calvert Cliffs 556 F 546 F I
Maine Yankee 541 F 528 F 0
Connecticut Yankee
555 F-551 F
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TABLE 2
% Power /% Flow During Natural Circulation Tests
% Power
% Flow
..af Fort Calhoun 0.20
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Calvert Cliffs Unavailable i
Maine Yankee 0.22
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Connecticut Yankee 0.20 2
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Topical Report BAW-10000A, Correlation of Critical Heat Flux in a l
Bundle Cooled by Pressurized Water, May 1976.
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Figure C-19. Experimental Errors for the Bundle
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- 0. 9 0.8 Bands Around Data Repre s'ent 100 E.atimated Total Uncertainty o
(Fixed Error + 3 X Randorp 0.7 Error) Due to Experimental f
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Source: Topical Report BAW-10000A (May 1976)
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