ML20028C225
| ML20028C225 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 01/05/1983 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:83:173, NUDOCS 8301070224 | |
| Download: ML20028C225 (12) | |
Text
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- Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:173 m 0 51983 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Check:
TRANSMITTAL OF DEMO PIPE BREAK ANALYSIS
Reference:
Letter HQ:S:82:149, Longenecker to Check, " Meeting Summary for the Shutdown Heat Removal Meeting, December 16, 1982," dated December 20, 1982 As part of the agreements and commitments from the December 16, 1982, shutdown heat removal meeting, the Clinch River Breeder Reactor Plant (CRBRP) project agreed to provide the DEM0 code output assumptions used for the pipe break analysis. Enclosed are nine DEM0 generated curves that plot sodium temperatures as a function of time from the average fuel, inner blanket, and outer blanket for the three pipe break locations analyzed. Case I is a break at the reactor vessel inlet.
Case II is a break in the cold leg pipe at the top of the reactor vessel guard vessel.
Case III is a break in the hot leg at the top of the intermediate heat exchanger guard vessel. The nature of the pipe materials in CRBRP when used with sodium at low pressure and operating below its boiling
~ point precludesan abrupt rupture. Thus, the incredible event of a guillotine rupture is beyond the design basis. The event is, therefore, not analyzed as a design basis event.
The pipe break analyses were not conducted with the full conservatism applied to design basis safety analyses.
For the most part, best estimate values were used; but some quite conservative assumptions did enter the analysis that exceed those nonnally used for beyond the design basis events.
Specifically:
1.
All three breaks were sudden guillotine breaks. That is particularly impossible in Case I, the worst case, where the guard vessel / guard pipe prevents the 24-inch pipe from moving more than 5 inches away from the reactor vessel plus 5 inches tangetial to the reactor vessel.
8301070224 830105 D
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2.
No credit was taken for the check valve that should be assumed to operate in a best estimate case even though its operation is not a safety function.
3.
The maximum specification allowance for reactor trip time was assumed.
4.
The minimum insertion rate for the control rods was assumed.
5.
The minimum control rod worths were assumed for the analysis, including the assumption that only the primary rod bank operated and the highest worth rod in that bank failed to insert.
6.
No credit was taken for cell pressure slowing the leak.
Other assumptions include:
1.
Expected plant operating conditions at beginning of operations were used instead of thermal hydraulic design values (maximum pressure drop--minimumpumphead).
Degradation of heat transfer surfaces over the 30 year life of the plant could increase the temperature nominally by 140F.
2.
Pump coast down characteristics from the prototype pump test were used instead of minimum specification allowable pump coast down characteristics.
The FORE-II analyses provided at the December 16, 1982, meeting was a mix of conservative and best estimate values also, instead of the total conservatism used in design basis events. Specific conservatisms were:
1.
The hottest rod at the worst time in life (beginning of Cycle 1) for the fuel was used.
2.
The hottest rod at the worst time in life (end of Cycle 4) for the blanket was used.
3.
The peak temperatures at any time during the incident were compared with the lowest saturation temperature at any time during the incident regardless of the fact that the peak temperatures do not occur at the same time as the minimum saturation temperature.
4.
Maximum trip time was assumed.
5.
Minimum shutdown worth was assumed, including only eight of the nine primary rods assumed to scram.
?
On the other hand:
1.
The starting conditions were plant expected operating conditions rather than thermal hydraulic design values.
2.
The original hot spot was followed in time during the event instead of following the movement, if any, of the hot spot down the rod.
3.
Nominal values for decay heat, film coefficient, power distribution, and so forth, were used instead of the nonnal design practice of applying direct and statistically combined uncertainties to tne results to obtain what we refer to as a 3 sigma analysis.
In sumary, the DEM0 analysis performed with assumptions more conservative than appropriate for beyond design basis events clearly provids evidence that core coolable geometry is maintained following a double ended pipe break.
Sincerely, k
~ankij Jdbn R. Longenhter Acting DirectoN0ffice of Breeder Demonstration Projects Office of Nuclear Energy 9 Enclosures cc: Service List Standard Distribution Licensing Distribution 1
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