ML20085H637

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Amend 9 to Change Request 32 to License DPR-4,clarifying Info Re Engineering Hot Channel Factors
ML20085H637
Person / Time
Site: Saxton File:GPU Nuclear icon.png
Issue date: 11/24/1969
From: Neidig R
SAXTON NUCLEAR EXPERIMENTAL CORP.
To:
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ML20083L048 List: ... further results
References
FOIA-91-17 NUDOCS 9110280329
Download: ML20085H637 (5)
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SAXTON NUCLEAR EXPERIMENTAL CORPOIMTION DOCKET UO. 50-lh6 LICENSE DPR-b Amendment No. 9 to Change Request No. 32

1. By application dated September n,1969, Applicant submitted a report entitled "EAFEGUARDS REPORI FOR SAXTON CORE III - Revision 1." paragraph 2 3 3 entitled, "Thernal and !!ydraulic Analysis," pre lented and discusced hot channe) factors and the tiethod used for calculating the enthalpy rice in the hot channel. That part of Paragraph 2 3 3 pertaining to hot channel factors has been revised to clarify the method used in calculating the enthnipy rise in the hot channel.
2. Applicant hereby submits Amend nent No. 9 to Change Request No. 32 revicing Pages 2-16, 2-17, 2-18, and TABLE 2 3-1 of the above report.

SAXTON FUCLEAR EXPERIMENTAL CORPOPATION py /s/ R. E. Neidig President Dated-11/24/69 9110280329 910424 PDR FOIA DEKOK91-17 PDR

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Evaluations were conducted at both nominal and overpower conditions using the THINC code.U )

2.3.3 THEPJML AND HYDRAULIC hhALYSIS Best estimat? X-Y power distributions are shown in Figure 2.3-1. These values are increa d by an 8% nuclear uncertainty before being used in the thermal and hydraulic analysis. Figure 2.3-2 shows the detailed. rod by rod power dis-tribution for the limiting loose-lattice assembly (D-2) and the limiting load-follow assembly (E-3). These power distributions include an 8% nuclear uncer-tainty factor.

Hot Channel Factors The total hot channel factors for heat flux and enthalpy r.ise are defined as the maximum-to-core average ratio of these quantities. The heat flux factors consider the local maximum at a point (the " hot spot"), and the enthalpy rise factors involve the maximum integrated value along a channel (the " hot channel").

Each of the total hot channel factors is the product of a nuclear hot channel factor describing the neutron flux distribution and an engineering hot channel factor to allow for variations from design conditions.

(a) Heat Flux (Fq)

The total hot channel factor is the product of a nuclear hot channel factor describing the nuclear power distribution effects and an engineering hot channel factor to allow for variations from nominal design conditions. The ,

engineering heat flux factor includes the variations in fuel rod diameter and fuel pellet diameter, density, enrichment and eccentricity and is identical to that used in the Core II design. Tables 2.3-1 and 2.3-3 summarize,the heat flux hot channel factors for Saxton Core 111.

(b) Enthalpy Rise (FAH)

The enthalpy rise in the hot channel is calculated osing THINC. For convenience, the engineering enthalpy rise factor is presented as the simple product of several subfactors. .In actuality, the various itons such as power distribution, s

(1) Chelemer, H. Weisman, J. , Tong, L. S. "Subchannel Thermal Analysis or Rod Bundle Core", WCAP-7015, January, 1967.

2-16

f mixing and flow redistribution are net independent of one another. These factors which are obtained from the THINC analysis will vary with the opera-ting conditions. For this reactor at nominal operating conditions a value for the engineering enthalpy rise factor is 1.00. The subfactors used in obtaining this value are described in the following paragraphs and presented in Table 2.3-1 for the DNB limiting channel. .

Pellet Diameter, Density and Enrichment Variation and fuel Rod Diameter, Pitch and Bowing:

Based on the applicable tolerances and consistent with the probability limit of three standard deviations for the measured Yankee, Selni, Sena, SCE, Connecticut Yankee and Indian Point data, the hot channel enthalpy rise was increased by 8%

when no mixing was permitted in the hot channel.

Inlet flew Ma1 distribution:

Possible inlet flow maldistribution effects are taken into account by conserva-tively reducing the flow by 7% to the fuel, assembly containing the hot channel.

This 7% reduction in inlet flow is the same as used in the Core I and Core II design analyses. The effect of the 7% inlet flow reduction on the hot channel enthalpy rise is determined by the THINC analysis when comparing the case of uniform flow to all assemblies to one where the inlet flow to the hottest essem-bly (including the hot channel) is reouced 7%. The net effect of this inlet flow reduction as evaluated in the THINC analysis results in an increase of 1% in the hot channel enthalpy rise.

Flow Redistribution:

The flow redistribution accounts for the reduction in flow in the hot channel resulting from the high flow resistance in the channel due to the local or bulk boiling. A nominal value for this hot channel subfactor when evaluated by the THINC code is 1.00. Because the Saxton core is greatly subcooled at the outlet, this subfector has no adverse effect on enthalpy rise.

Flow Mixing:

The effect of mixing on the hot channel enthalpy rise is evaluated using the THINC code. The hot channel enthalpy rise evaluated by THINC assuming no mixing 1

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  • l 2-17 l

is compared to the hot channel enthalpy rise evaluated from the use of a thermal diffusion coefficient (see Ref. (1) ) based on experimental tests con- i ducted with spacer grids (no mixing vanes) in a full size assembly. The bene-fit obtained from mixing is to decrease the hot channel enthalpy rise 8%.

Because of unheated rods, the total enthalpy rise cannot bc considered as a simple product of nuclear factors and engineering factors. Enthalpies at the exits of the hot channel are calculated by THINC and are reported with the nu-clear hot channel factors in Tabic 2.3-3; Departure From Nucleate Boiling The evaluation of the Sexton Core III, DNB conditions have been made using the W-3 correlations.(2) The W-3 DNB design minimum value of 1.30 has been chosen statistically to insure a 95% probability that DNB will not occur with a con-fidence level of 95%. For fuel channels adjacent to the assembly enclosure, the presence of the unheated wall affects the amount and degree of coolant mix-ing inside the channel. The effect of the unheated wall on the DNB ratio has been considered (modified W-3 correlation used)(3) in the des'ign analysis using experimental data.

The minimum DNBR during overpower conditions is 1.30, as insured by the reactor trip setpoints (See Section 3.1).

Thermal and Hydraulic Design Parameters Detailed THINC analyses were conducted for the various limiting thermal and hy-draulic conditions in the core. The following overall limits apply at nominal steady state operation:

1. 28 MWt
2. 19.9 Kw/ft in load-fellow assemblies

' 3, 24.0 Kw/f t in loose-lattice assemblies l

l ~(2) L. S. Tong. " Prediction of Departure from Nucleate Boiling for an Axially

! Non-Uniform Heat Flux Distribution", J. of Nuc. Energy Vol. 21, pp. 241-248,(1967).

l (3) L. S. Tong, et. al., " Critical Heat Flux on a Heater Rod on the Center of i

Smooth and Rough Square Sleeves, and in Line Contact with an Unheated Wall",

ASTM -WA/HT-29 (1967).

2-18

s.
  • n Table 2.3-1 Engineering Hot Giannel Factors Pellet Diameter, Density F

E Enrichment, and Eccentricity 1.045 Rod Diameter, (Pitch and Bowing) b

.i Pellet Diameter, Density, Encichment E Rod Diameter, Pitch and Bowing 1.08 ,

7 AII Inlet Flow Maldistribution 1.01 -

Flow Redistribution 1.00

  • I Flow Mixing C.92 Resulting F E g 1.00 Y

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