ML18347B306

Steady-State Thermal Hydraulic and Neutronics Analysis of for Operation at 2530 Mwt

ML18347B306
Person / Time
Site:	Palisades
Issue date:	07/15/1977
From:	Correll G, Kelley R Exxon Nuclear Co
To:	Office of Nuclear Reactor Regulation
References
XN-NF-77-22
Download: ML18347B306 (27)

Text

I XN*NF*77 *22 I

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STEADY-STATE THERMAL HYDRAULIC AND NEUTRONICS I ANALYSIS OF THE PALISADES .REACTOR

'I FOR OPERATION AT 2530 MWT I

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• 1 JULY 15, 1977 I

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~1 RICHLAND, WA 99352 I

I . EJ${0N NUCLEAR COMPANY, Inc.

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I Issue Date: 07/15/77 XN-NF-77-22 I STEADY-STATE THERMAL HYDRAULIC AND NEUTRONICS ANALYSIS OF PALISADES REACTOR FOR OPERATION AT 2530 MWt I

I by R. H. Kelley I and G. R. Correll I

,, Approved:

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f?5/~ C~&t*~i {;v11 ~~---7_,_/t~L~/_7_7_

C1';""' E. Leach, Mana er /' 'Da-te I Thermal Hydrauli Engineering I

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Approved ~-.r\~r---

w. S. Nechodom, Manager I Licensing & Compliance 1:

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IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY I

This technical report was derived through research and develop-ment programs sponsored by Exxon Nuclear Company, Inc. It is being I

submitted by Exxon Nuclear to the USNRC as part of a technical con-tribution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear-fabricated reload fuel or other techni-cal services provided by Exxon Nuclear for light water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, I

information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulations.

I l~i thout derogating from the foregoing, neither Exxon Nu cl ear nor any person acting on its behalf:

A. Makes any warranty, express or implied, with I

respect to the accuracy, completeness, or useful-ness of the information contained in this docu-ment, or that the use of any information, apparatus, method, or process disclosed in this document I

will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages re*sulting from the use of, any information, apparatus, method, or process I

disclosed in this document.

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I I XN-NF-77-22 I TABLE OF CONTENTS I

I Page l .0 INTRODUCTION. l I 2. 0

SUMMARY

. . 3 I 3.0 4.0 THERMAL HYDRAULIC DESIGN.

THERMAL DESIGN. .

7 14 I 5.0 NEUTRONICS CONSIDERATIONS 17

6.0 REFERENCES

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1.0 INTRODUCTION

I This report presents the evaluation of the thermal hydraulic and neutronics performance of the Palisades core and Cycle 2 Reload fuel types I at the core power level of 2530 MWt. The nominal primary coolant operating

'I conditions selected for this analysis are 2060 psia and 537°F core inlet temperature.

I The Exxon Nuclear Company (ENC) fuel .assemblies are designed to be com-patible with the Palisades reactor core and with Batch D fuel. This was dem-I onstrated in the Thermal Hydraulic Analysis Report submitted in support of I the Cycle 2 license application (XN-76-3 [P]) and Supplement 1.

fuel assemblies are designed to operate at a core power of 2650 MWt The reload I steady-state and anticipated transient conditions. The mechanical integrity of the fuel at pressure up to 2100 psia, at nominal core inlet I temperatures up to 543°F, and at peak fuel rod powers very nearly the same as will be experienced at 2530 MWt was reported in XN-76-52, submitted in support I of operation at 2100 psia. Operation at 2100 psia has been reviewed and I approved by the NRC. (l)

The limiting fuel types for the analysis are Batches D and E fuels.

I Batch D fuel was supplied by Combustion Engineering. Batches E and F fuels were supplied by Exxon Nuclear Company. Batch F fuel is 1.5% enriched and I has .low power generation and is therefore not limiting. Batch E fuel is I representative of future anticipated ENC Palisades reload deisgn. It is anticipated that future reload fuel performance will be enveloped by this I analysis.

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I The Palisades reactor core is currently licensed at 2200 MWt. Operation I

at 2530 MWt with adequate thermal margins is principally achieved by reduction of ,allowed nuclear peaking factors; the peak linear heat generation rate at I stretch power is nearly the same as for the currently licensed 2200 MWt.

Analysis of the neutronics characteristics of the Cycle 2 core at

'I 2530 MWt and an exposure of 15 GWD/T are presented in Section 5.0 of this I report.

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SUMMARY

'I The required reactor thermal margins and thermal margin trip limits to maintain MDNBR ~ 1.3 (W-3 correlation) during steady-state and transients I for the limiting D and E fuels is presented in this document. The steady-

,I state thermal margin analysis was performed at 115 percent of 2530 MWt (2910 MWt) to provide margins for the transients. On the basis of core I conditions including power peaking given in Table 2. l, the MDNBR at over-power was calculated to be l.30. This indicates that 15 percent overpower I results in an MDNBR lower than calculated for the most restrictive anticipated plant transient, i.e., l. 35. ( 2 )

I The method of analysis employed are consistent with prior ENC submittals I for Palisades, and the application of the W-3 DNB correlation is consistent with the methods described in Reference (3). A summary of the steady-state I DNB analysis results are presented in Table 2. l. The table shows the nuclear I peaking and engineering factors assumed in the analysis for fuel types D and E. This table also compares the limits to those derived in support of Cycle 2 I operation at 2200 MWt.

Steady-sta~e plant thennal margins were determined parametrically as a I function of power, core inlet temperature, and primary system pressure. The I results of this analysis are given in Section 3. These limits were used to derive set point equations as prescribed in the plant technical specifications.

I These set point equations are required in performing the plant transient analysis as report in XN-NF-77-18. The results of the transient analysis I indicate that adequate thermal margins are achieved to protect an MDNBR I > 1.3 during anticipated transients (Class 2 transients).

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I XN-NF-77-22 I

The neutronics characteristics of the Palisades Cycle 2 core have been I evaluated. These cha~acteristics are ~uch that routine procedures for main-tenance of thermal margins enable safe operation of the core at power levels I up to 2530 MWt and Cycle 2 average core exposures up to 15 GWD/MTU.

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I I TABLE 2.1 XN-NF-77-22 I

SUMMARY

OF DNB ANALYSIS I Design Reactor Core Conditions 2530 (MWt) 2200 (MWt)

Design Overpower, (MWt) 2910 2684 I Nominal Total Core Flow Rate, (10 6 lbm/hr) 121 . 7 124 Nominal Active Core Flow Rate, (10 6 lmm/hr)

I Primary Pressure (psia) 114.4 116.8 2060 2100 I Core Inlet Temperature (°F) 537 543.

Core Pressure Drop (psia) 13.5 + 0.5 13.5 + 0.5 I Fuel Bundles in Core 204 204 I Average Linear Heating Rate, (kw/ft)

Maximum Linear Heating Rate,(kw/ft) nominal 5.3 13.8 4.6 13.9 I *Maximum Linear Heating Rate, (kw/ft) at overpower 15.9 17.0 Fraction of Heat Generated in Fuel 0.975 0.975 I MDNBR (at overpower) l .3 1.3 I Nuclear Peaking Factors 2530 2200 Radial 1 .45 1.6 I Axi a 1 1.4 l. 5 Local 1.22 1. 21 I Engineering 1.03 (E) l . 03 ( E)

I 1.05 (D) 1 .05 (D)

Total 2.55 (E) 3.05 I 2.60 (D)

I -*----* -----------------

• The lower maximum linear heating rate at the higher core power is accounted for by the allowance of lower nuclear peaking factors at stretch power.

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I TABLE 2.1 (continued)

I Fuel Bundle Description Batch

- --- D Batch E I Rad Diameter 0.4175 in. 0.415 in.

Rod Pitch 0.55 iri 0.55 in I

Active fuel length 131 . 4 in 131 . 8 in. I Number of active rods 216 208 Number of poison rods 0 8 I Instrument tubes l Number of guide bars 8 8 I

Total positions 225 225 1*

Number* of Spacers 10 10 Average heating rate at 2530 MWt 5.244 kw/ft 5.429 kw/ft I II II II 2200 MW 4.560 kw/ft 4. 721 kw/ft I

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I I XN-NF-77-22 I 3.0 THERMAL HYDRAULIC PERFORMANCE I The evaluation of the thermal hydraulic design of the Palisades Reactor core and the ENC reload fuel was performed with the XCOBRA-IIIC( 4 ) thermal I hydraulic computer code employing the W-3 critical h~at flux correlation.

A hydraulic model of one-eighth of the Palisades Reactor core (see Figure 3.1)

I for a typical core loading pattern was used to calculate the flow to each I fuel assembly. The effect of a 5% lower plenum flow maldistribution was included in this analysis.

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A separate subchannel hydraulic model (see I Figure 3.2) consisting of an octant of the individual fuel assembly was used to determine the MDNBR of limiting fuel bundles. The design basis axial I heat flux profile, the bundle radial power distribution, and the bundle local

.I power distribution used in the'hydraulic models were based upon neutronic calculations. The core flow rate was determined as described in paragraph 3.3.

I The results of the MDNBR analysis with the design basis axial power I distribution and at reactor design overpower (115 percent) show that the MDNBR of the limiting ENC and CE fuel bundles are never less than l .3 at the design I operating conditions. Reactor conditions and assembly characteristics con-I sidered in this analysis are presented in Table 2.1.

I 3.1 HYDRAULIC CHARACTERISTICS The hydraulic characteristics of the ENC reload fuel assemblies I used in the hydraulic models are based on experimental results from hydraulic tests on fuel assembly components that are similar to those specified for I the Palisades Reactor fuel assemblies. Hycraulic characterization includes I

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I XN-N F-77-22 I

single-phase loss coefficients for:

• Lower tie plate and core support (combined)

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• Grid spacers I

• Upper tie plate and ho1ddown plate (combined)

• Bare rod friction. I The loss coefficients as measured for both ENC fuel assemblies and Combustion Engineering assemblies were used in the analysis. of each fuel I

assembly type in the XCOBRA-IIIC hydraulic models. I 3.2 HYDRAULIC COMPATIBILITY I

The hydraulic compatibility of the ENC Reload fuel with the Palisades Reactor is measured in part by the impact upon the core flow distribution of a I mixed core loading of existing and reload fuels. The split in flow among fuel assemblies is due t6 the different hydraulic resistances of the various assem-I blies as reflected in their individual hydraulic characteristics. The flow I

rate to the hot channel is strongly dependent on the assembly radial power (higher power generally results in lower flow rates) while the variation in I axial power profile among bundle has negligible effects. Thus, the same axial power profile was used for all assemblies while the hot bundle was assigned I

the design radial power factor.

be nonuniform, and th~

The lower plenum inlet flow was assumed to flow to the core region with the highest power assembly I

was 5% less than core average inlet flow. I The flow distribution for a typical loading pattern of D and E fuels showed less*than one percent variation in bundle mass flows from core I

average for a nominal power distribution and uniform inlet flow. The combi- I nation of the maximum design radial bundle power and five percent less than core average inlet plenum mass flow on highest power (FR= 1 .45) E and D I I

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assemblies was used in predicting hot bundle flow factors.

were then applied when modeling the limiting fuel assembly .

3.3 These flow rates MINIMUM DEPARTURE FROM NUCLEATE BOILING RATIO (MDNBR)

I The MDNBR is a function of limiting bundle power, primary system pressure, coolant inlet temperature, and coolant flow rate. The primary I coolant flow rate was determined *by a conservative relationship that accounted for the changing coolant density with coolant temperature (including core power)

I conditions based on measured flow at H~P. The effective core flow rate was I reduced by 3% to account for core bypass fl ow and a furthe*r 3% to account for uncertainty in flow measurements. The bundle flow rate is finally reduced I in accordance with the applicable hot bundle flow factor as described in Section 3.2. The XCOBRA-IIIC limiting fuel assembly model was evaluated at 115 percent I of 2530 MWt power and design radial to find the MDNBR as a function of coolant I inlet temperature and primary system pressure. A conservative value of the core inlet temperature, 5°F above nominal, as well as a conservative value of I primary pressure, 50 psia below nominal, were used in the analysis. These conservatisms account for unfavorable impact on the thennal margin of measure-I ment uncertainties and *normal operating fluctations of temperature and pressure.

I 3.4 THERMAL MARGIN For nominal operations, the limiting values of primary coolant pres-I sure, reactor inlet temperature, and reactor power level are defined for a I broad range of each parameter for which the thermal criteria are not exceeded.

The limits of operation are designed to assure:

I .. MDNBR > l .3 I

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• Quality of primary coolant

• Fl ow stab i l ity .

< 15 percent at MDNBR location I

The derivation of the thermal margin curves was accomplished by the I utilization of the XCOBRA-IIIC subchannel hydraulic model. The core operating conditions (primary system pressure, core inlet temperature, and core power)

I were parametrically analyzed so that the combinations of the above primary I system parameters that resulted in MDNBR = 1 .3 were determined. The assembly mass flow rate was determined based on its dependence on core inlet temer- I ature and power with appropriate adjustments for core bypass, measurement uncertainty, and the hot bundle flow factor. The design nuclear peaking factors I

were assumed throughout the analysis. The reduction of slope of. the curves at low I

power and low pressure (see Figure ~.3) is due to the occurrence of increased coolant quality limiting the allowed core inlet temperature rather than MDNBR I (i.e., W-3 correlation is valid only up to 15% local coolant quality).

Current low primary coolant flow and low primary system pressure I

trip points are adequate to prevent the possibility of DNB resulting from local I

flow oscillations. A parallel channel flow stability analysis was performed.

The results of this analysis showed that within the allowed technical specifi- I cation operating limits no flow instability will occur.

The limit lines shown on Figure-3.3 are provided for the formulation I.

of operating thermal set points in an algorithm compatible with the reactor I

safety system parameters.

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I XN-NF-.77-22 I

I D D D I .604 .768 .794 I D D D F E I Shrou

.. 544 .939 1 . 219 .889 1 . 141 D E F D F I 1. 001 1. 252 .978 1.418 * . 901 I .983 F

1 . 271 E

. 901 F

1 .138 E

I E F F 1 . 216 . 833 . 805 I

II .xxx Fuel Type 1 .129 E

1 .137 xxx Radial Peaking I Type D - Combustion Engineering 1 . 133 Type E & F - Exxon Nuclear I Reload I*'I

• Raised to design radial peaking (1.45)

I FIGURE 3. 1 CORE POWER FOR DISTRIBUTION FOR DETERMINATION OF CORE FLOW DISTRIBUTION I

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XN-NF-77-22 I

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Assembly Instrument Tube I

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@ I Fuel Rud Peaking Factor I

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L /___ _

___ r_____ _ GUIDE BAR I

• Increased to 1.22 for this analysis.

FIGURE 3. 2 PALISADES TYPE E ASSEMBLY SUBCHANNEL MODEL I

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Pressure I

590 I

570 LL 0

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rO

~

560 QJ Q_

E QJ I-

+>

QJ r-e 550 ~

QJ

~

0 u

540 530

% of Maximum Bundle Power at 2530 MWt FIGURE 3. 3 PALISADES THERMAL MARGIN LIMITING OPERATING CONDITIONS AT DESIGN CORE PEAKING FACTORS (Table 2.1)

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4.0 THERMAL DESIGN I

4.1 FUEL TEMPERATURE 5

The GAPEX( ) computer code was used to calculate the pellet-to-I clad gap conductance and to calculate the maximum steady-state fuel tempera- I ture at design overpower conditions. The fuel temperature calculation con-sidered the effects of the worst fuel pellet and cladding tolerances, fuel I densification, pellet cracking, and temperature and density on therm.al conductivity. Lyo~s( 6 ) uo 2 thermal conductivity data was used to define the I

temperature dependence of the fuel thermal conductivity. Lyon data which is I for 95 percent theoretical density fuel was corrected to the reload fuel design density. I A total power peaking factor of 2.55 was used for maximum heat generating rate of 115 percent of 2530 MWt. Individual peaking factors are I

listed in Table 4.1. I Steady-state fuel and cladding temperature for the beginning-of-life, low-enrichment fuel were most limiting. All temperatures as shown in I

Table 4.1 are below design limits.

I The GAPEX computer code was used to determine the maximum steady-state poison pellet and poison rod cladding temperatures at overpower. I Modifications were necessary to account for:

• The conductivity of B4C alumina at manufactured density as I

a function of temperature. I I

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• The decrease in conductivity of B c alumina with irradiation .

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• The thermal expansion of s c alumina.

I 4 The linear heat generation rate of the poison rod included factors I for core radial peaking, overpower (115 percent), axial peaking, and varia-tions in boron concentration. No densification or cracking effects were I considered. Beginning-of-life geometry (largest pellet-to-clad gap) and end-I of-life conductivity (maximum conductance) were used to envelope all condi-tions. Results are listed in Table 4.1.

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I TABLE 4. l I

FUEL, CLADDING AND POISON TEMPERATURES FOR ENC RELOAD E FUEL RODS WITH A LOCAL PEAKING TO 1.22 I I

Average heat generation rate at

. 115 percent power 6.24 kw/ft I

Peaking Factors:

I Radial l. 45 I Axial l. 4 I

Local 1.22 Engineering 1.03 I Total 2.55 I

.Percent Power Deposited in Fuel 0.975 I Temperatures:

I Maximum Fuel Centerline 4330°F I Maximum Clad O.D. 662°F I

Maximum Clad I.D. 799°F Maximum Poison Pellet 1100°F I

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XN-NF- 77-22 I

5.0 NEUTRONICS CONSIDERATIONS I The effects of operation at 2530 MWt and of a potential increase in I the length of Cycle 2 upon the neutronics characteristics of .the Cycle 2 core have been evaluated. Although the Cycle 2 core is somewhat atypical of cores I planned for use in the future (i.e., radial power peaking factors for Cycle 2 are somewhat higher than those projected for future cycles), the character-.

I istics of the Cycle 2 core are generally expected to envelope those of future I cores.* The characteristics of future cores will, of course, be individually analyzed in detail .and compared to safety analysis limits at the time of I final design of each of the cores.

The analyses reported here relied primarily upon a quarter-core, I pin-by~pin PDQ7/HARMONY model for calculation of detailed radial power dis-1* tributions and differential effects (rod worths, moderator temperature coefficients, etc.) and upon a three-dimensional, quarter-core XTG model I for calculation of exposure-dependent core reactivity, axial effects, etc.

(See Reference 7 for a basic description of the methodology utilized).

I It was assumed that at a Cycle 2 exposure of ~10,500 MWD/MTU the core power will be increased to 2530 MWt and that' core p9wer will be I maintained at that level until EOC2. Due to favorable operating experi-I- ence with the Cycle 2 core and due to a slightly lower rate of reactivity depletion versus exposure than originally expected, Cycle 2 is expected to I operate beyond the 12 GWD/MTU EOC2 exposure heretofore addressed. (The core is'calculated to be capable of operation at 2530 MW t to a \ core I exposure of 13, 140 .MWD/MTU.) To envelope this .potential for increased cycle I

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length and to evelope a potential for power coastdown operation beyond this exposure, th~ neutronics characteristics of the Cycle 2 core were evaluated I

at 2,530 MWt and an EOC2 exposure of 15 GWD/MTU.

I The neutronics characteristics of the Cycle 2 core at 2,530 MWt I and 15 GWD/MTU ar,e compared to those for ~,200 MWt at 12 GWD/MJU and to the revised safety analysis 1 imits in Table 5.1. The safety-significant I neutronics param,eters lie*well within the safety analysis limits with the

~xception of radial power peaking which at 15 GWD/MTU ~s projected to I

exceed the 2,530 MWt limit by rv2%. (The increase from the 1.35 maximum I radial peaking factor calculated at 2,200 MWt and 12.GWD/MTU is primarily the result of increased fuel exposure and increased depletion of the Batch E I burnable poison rods). This radial peaking will not, however, be deleterious to core safety since power peaking factors are coritinuously monitored by I

in-core detectors and the core power level is adjusted as necessary to I maintain margin to thermal limits. (The excess of rv2% in the calculated radial peaking at 15 GWD/MTU is not expected to a<:tually constrain core I operation since core reactivity is expected to be .insufficient to maintain 2530 MW beyond 13,140 MWD/MTU).

I Control rod worths, r~acti vity a11 owances, and ca 1cul ated shutdown I margins at 2,530 MWt-15 GWD/MTU are compared to values at 2,200 MWt-12 GWD/MTU in Table .5.2. The gross rod worth is calculated to ,be slightly higher at I 15 GWD/MTU,consistent with the increase calculated between.BOC 2 and 12 GWD/MTU, but the maximum stuck rod worth is higher by essentiall.y the same I

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I I XN-NF-77-22 I amount and the net rod worth at 15 GWD/MTU is essentially unchanged from that at 12 GWD/MTU. The basic reactivity defect (sum of moderator temperature I and Doppler defects) is calculated to be 1.84% ~P at 15 GWD/MTU (versus 1.29% Ap at 12 GWD/MTU) and a conservative reactivity allowance of 2.0% Ap I is provided for these effects.

I An excess shutdown margin of 0.33% Ar is conservatively projected from HFP at 15 GWD/MTU. Considering the revised HZP shutdown requirement of I 2.0% Ap and assuming no change in the HZP PDIL rod insertion from that quoted in Reference 8, an excess shutdown margin of 1.41% Ap is conserva-I tively projected for HZP at 15 GWD/MTU.

I It is concluded that the neutronics characteristics of the Palisades Cycle 2 core in combination with routine procedures for maintenance of thermal I margins enable safe operation of the core at power levels up to 2,530 MWt and exposures up to 15 GWD/MTU.

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TABLE 5 .1 EFFECTS OF INCREASED POWER AND CYCLE LENGTH ON PALISADES I

CYCLE 2 CORE NEUTRONICS PARAMETERS I 2200 MWt, 2530 MWt, Safety Analysis I

12 GWD/MTU 15 GWD/MTU Limit Moderator Temperature Coefficient, I

10-4 b.p/oF -1.42 -2.14 +0.5 to -3.5 I

Doppler Coefficient, 10- 5 ~p/°F -1.49 -1. 55 . - . 87 to -1. 64 I

Power Peaking Factors:

Radial Axial Total (Including Local)

1. 35

1. 10

1. 74

1. 48* '

1.12

1. 92 1.45

l. 40 2.55 I

Max. Ejected Rod Worth,  %~p I

HFP <0.2 <0.2 <0.6 HZP <0.90 <0.84

<l. 24 I

Delayed Neutron Fraction 0.0052 0.0049 >0.0045 I

Reciprocal Boron Worth, ppm/%~p 84 85 <100 I

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• Operation to 15 GWD/MTU would probably require a slight derate due to radial peaking being greater than l .45. Core would be in power coastdown due to reactivity depletion at this exposure. I I

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I I XN-NF-77-22 I TABLE 5.2 PALISADES EOC2 SHUTDOWN MARGIN I (All Entries in-Units of %6p)

I 2200 MWt, 12 GWD/MTU 2530 MWt, 15 GWD/MTU HZP HFP HZP HFP I Total Full Length Rod Worth* 9.15 9.15 9.33 9.33 I Stuck Rod Worth Total Minus Stuck Rod 3.52 5.63 3.52 3.69 3.69 5.63 5.64 5.64 I Uncertainty .56 .56 . 56 .56 Net Shutdown Rod Worth 5.07 5.07 5.08 5.08 I Doppler Defect 0 1. 00 0 1.10 I Moderator Temperature Defect 0 . 80 0 .90 Moderator Void Defect I Axial Flux Redistribution 0

0

.10

.50 0

0

.10

.50 I Required Shutdown Margin 3.40 2.00 2.00 2.00 Total Reactivity Allowances 3.40 4.40 2.00 4.60 I Available for Maneuvering l. 67 0.67 3.08 0.48 -

I PDIL Rod Insertion 1. 67 0.15 1. 67 0.15 Excess Margin 0.00 0.52 1.41 0. 33 -

I *EOC Rod Worth = EOC (Calculated) x BOC Measured/BOC Calculated I

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6.0 REFERENCES

I l. Letter, USNRC (H. Schwencer) to Consumers Power Company, May 11, 1977, Docket No. 50-255.

I 2. G. E. Koester, J. D. Kahn, D. J. VandeWalle, "Plant Transient Analysis of the Palisades PWR for 2530 MWt", XN-NF-77-18.

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3. K. P. Galbraith, et al.,"Definition and Justification cff Exxon.

I Nuclear Company DNB Correlation for PWR's", XN-75-48 (NP), October 6, 1976.

I 4. K. P. Galbraith and T. W. Patten, "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient I Core Operations 11 , XN-75-21, April l, 1975.

5. K. P. Galbraith, GAPEX: A Computer Program for Predicting Pellet-I to-Cladding Heat Transfer Coefficients 11 , XN-73-25, August 13, 1973.

I 6. M. F. *Lyons, et al., "U0 Pellet Thermal Conductivity from Irradiation

- with Central Melting 11 , 2GEAP-4624, May 1964.

I 7. F. B. Skogen, 11 Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors, 11 XN-75-27, June 1975 and XN-75-27 Suppl. 1, September, 1976.

I 8. Letter from D. P. Hoffman, CPC, to A. Schwencer, NRC, dated April. 29, 1977.

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I I XN-NF-77-22 I DISTRIBUTION I CA RH GE Brown Kelley Koester TL Krysinski I CE GA Leach Sofer RJ Ehlers I KP Galbraith NRC (40) I GF Owsley I Consumers Power ( 95) / RJ Ehlers *.

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