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OEC 4 1964 Dr. Kamal Araj Department of Physics Harvard University Cambridge, MA 02138
Dear Dr. Araj:
During our telephone conversation on November 26, 1984 and November 27, 1984, you asked the following:
1.
Provide a diagram of a Mark II containment.
2.
Verify the diagram of the Surry containment, which I sent previously.
Focus on the reactor cavity.
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3.
Provide experimental data for decay heat curves, which I previously sent.
4.
Provide diagrams of a core / concrete interaction showing more detail than the diagrams which I previously sent.
5.
Provide specifications of the core used in Surry 1 and Peach Bottom 2 so that the data used for the Pl11-2104 analyses can be verified.
6.
Provice burnup figures for Surry 1 and Peach Bottom 2.
The figures shoulc be those used in the analyses and those actually found in a reactnr.
7.
Define the following tems; degraded core, deterministic approach, event tree analysis, external event, and rule making.
The responses to these items are given below:
Item 1.
A diagram of the Mark Il containtrent is enclosed.
Item 2.
'The diagram of the Surry containment, used in EMI-2104, is erroneous. The diagram is of a high pressure containment but it is not the Surry containment; the Surry containment has a flat mat, not a keyed mat as indicated; the crane is mounted on a crane wall, not the floor as indicated. These differ-1ces may influence a source tem.
E. Waman at Stone and Webster Engineering Corporation will send you the correct diagrams in several weeks.
Item 3.
Experimental data for a decay heat curve does not exist as such.
The curve is calculated using radionuclide decay series. The decay series are well known. The issue is not one of data but one of detemining how many radionuclides to include in the calculations. The curves are sufficient to describe the shut-down of a reactor.
Item 4.
I am compiling a set of diagrams of a core / concrete interaction with text.
Item 5.
Core specifications that would be useful to verify the BMI-2104 data are difficult to obtain because the core specifications may
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2 change each time a reactor is refueled.
The Peach Bottom 2 reactor has 764 assemblies with fuel rods in a S x 8 lattice. The Surry reactor has 157 assemblies with fuel rods in a 15 x 15 lattice. This is as of November 27, 1984.
Though the desions may change from one refueling to the next, the overall power is about the same. Surface areas and ccm-position may change slightly.
In the BMI-2104 analyses, the following fuel arrangements were
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used:
- of Assemblies Lattice Peach Bottom 764 7 x 7 ano 8 x 8 (mostly)
Surry 157 15 x 15 Zion 193 15 x 15 Grand Gulf 800 8x8 Limerick 193 17 x 17 The actual core configuration has little impact on the source term.
Given a core configuration, the source terms are sensitive to the power distribution.
For the BMI-2104 analyses, the power distri-bution is modelled as a chopped-cosine curve; actually, a power distribution is much flatter.
Though the core configuration is unimportant for a source term calculation, a consistent analysis is important. A single core configuration was used for each reactor to calculate source terms for the selected accident sequences.
Item 6. The table below shows the burnup figures that you requested.
Surry 1 Peach Bottom 2 Core Fraction Burnup (mwd /MT)
BMI-2104 1/3 33000 Average burnup is 1/3 22000 18434 mwd /MT 1/3 11000 Actual 31000 mwd /MT
- Predicted for next 18 months 19780 mwd /MT huREG/CR-3602, Fuel Performance Annaual Report for 1982 The BMI-2104 analysis for the Peach Bottom reactor used burnup figures for the seven types of fuel found in the Brown Ferry 1 core. This was done because the information was readily available from the Oak Ridge National Laboratory.
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3 The average burnup can be calculated using the data in table 6.6. p 6-53,
. Volume II, BMI-2104. For your convenience I have abstracted the data:
Number of Approximate Assemblies Burnup (mwd /MT) 87 30400 127 23800 140 22900 23 24000 87 16600 68 16900 232 8900 Weighted average = 18434 mwd /MT, A calculation using the ORIGEN code was done for each type of fuel.
The ORIGEN results were then added together Iten 7.
The commonly accepted definition of each term that you request are given as follows:
Degraded core - loss of fuel rod geometry, massive zircaloy oxidation, etc.
Deterministic approach - the standard approach to the specification and analysis of accidents in the licensing process; e.g., specifi-cation of a design basis accident followed by a prescribed, usually conservative analysis. The term is usually intended to contrast the term "probabilistic" approach used in probabalistic risk assessment (PRA).
Event tree analysis - an accident -initiator with an origin external to, and not caused by, faults within the plant; e.g., earthquakes, sabotage, meteorite strikes, etc.
Rulemaking - the process for changing a regulation in the Code of Federal Regulations (CFR). The process usually involves a puolic hearing.
Sincerely, th, %$
4 Christopher Ryder Accident Source Term Program Office i
Office of Nuclear Regulatory Research
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