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ENCLOSURE.

SER ON NUSCO 140-P, VIPRE-01

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!.1.0 INTRODUCTION

By letter dated July 30,1984 (Ref.1), Northeast Utilities Service Company(NUSCo) submitted a topical report NUSCo 140-2, "NUSCo Themal Hydraulic Model

Qualification Volume II (VIPRE)" (Ref. 2), for staff review. The purpose of thesubmittal is to demonstrate NUSCo's in-house reload capability in the use of theVIPRE-01 computer code (Ref. 3) for perfoming core themal hydraulicanalysis. The first planned reload application of VIPRE-01 will be for theupcoming cycle for the Haddam Neck plant. NUSCo also indicated that the

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application of VIPRE-01 may be extended to Millstone Unit 2 reloads.:

i- VIPRE-01 is a subchannel themal-hydraulic code developed by Battelle PacificNorthwest Laboratories under the sponsorship of the Electric Power ResearchInstitute (EPRI). In December 1984, the Utility Group for RegulatoryApplications (UGRA), which consists of more than 20 utilities, submitted theVIPRE-01 topical reports for staff review. VIPRE-01 was developed from theCOBRA series codes including COBRA-IIIC (Ref. 4), COBRA-IV (Ref. 5),-

I COBRA-IIIC/MIT (Ref. 6) and COBRA-WC (Ref. 7) by incorporating many featuresof these codes into one package. The staff review (Ref. 8) concluded thatthe VIPRE-01 code was acceptable for PWR application with the following

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i conditions:

(1) The application is limited to the heat transfer modes up to critical heatflux (CHF)..

(2) An analysis is made to ensure that the minimum departure from nucleate '

boiling ratic (DNBR) limit of a CHF correlation used in VIPRE-01 can

predict its data base of DNB occurrence with at least a 95 perpent) probability at a 95 percent confidence level. -

; (3) Documentation is submitted by each user to provide ,iustification for themodeling assumptions, choice of particular two-phase flow models,correlations and input values of plant specific data, etc.

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(4) If a profile fit subcooled boiling model which was developed based on-

steady state data is used in boiling transients, care should he taken in,

establishing the time step size used for transient analysis tn avoid aCourant number less than 1.

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(5) Each user should abide by the quality assurance program established byEPRI for the VIPRE-01 code.

Our evaluation of the report will concentrate on the application of theVIPRE-01 code by NUSCo for licensing calculations.

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2.0 STAFF EVALUATION,

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In using the VIPRE-01 computer code for the core thermal hydraulic analysis,' NUSCo will apply a single pass approach. In the single pass method, the

calculations of the hot channel flow conditions and minimum DNBR are done injust one computer run which accounts for the crossflow and turbulent mixingeffects of not only the neighboring subchannels but the entire core. This isaccomplished by a lumped subchannel core model where the subchannel layout of

the hot channel and its adjacent channels within the hot assembly is modeledin detail, and the remaining assemblies in the core whcse effects on the hot

I channel is relatively small are lumped into a few larger channels. The singlepass approach has been increasingly used in the nuclear industry due to thelarge capacity of the thennal hydraulic computer codes and has been accepted.in the past by NRC for using the thennal hydraulic codes such as COBRA-IIIC/MIT

(Ref. 9) and LYNXT (Ref.10).

Topical Report NUSCo 140-2 provides a detailed description of NortheastUtilities qualification efforts on the VIPRE-01 core thermal hydraulic modelwith the single pass approach. These efforts include (1) sensitivity studies'

to establish the VIPRE-01 input parameter values and options and (2) VIPREbenchmarks with comparisons to the COBRA-IIIC calculated results and test data.

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2.1 Sensitivity Studies:*

The sensitivity studies perfomed by NUSCo use the Haddam Neck Plant (HNP) asa typical base model. The objectives of the sensitivity studies are to examine

'the effects of the important parameters and options on the VIPRE-Ofcalculatedresults. The studies serve as bases for the input values, options of I

correlations and solution techniques available in VIPRE-01. A summary ofthe sensitivity study results and the selected values and options for the HNPVIPRE-01 model is provided in Section III.R of the NUSCo 140-2 report.,

Although many sensitivity study items, such as core model, water properties,and numerical solution schemes, etc., are generically applicable to the PWRfuel designs, some of the other sensitivity studies are dependent on specificfuel design. Since the studies were perfomed with the HNP fuel design, the-

|| results of the fuel-dependent items are applicable to HNP only. A summary ofthe staff review of the sensitivity studies follows.

(1) Using a i core model, the study varies the number of radial channelsranging from 8 to 43 channels with the number of channels in the hotassembly ranging from 4 to 22. The resulting minimum DNBRs generallydecrease with increasing radial node detail with a maximum difference ofabout 3% in DNBR. However, when a 3x3 subchannel array is modeled with

I the hot channel completely surrounded, the resulting minimum DNBRs arealmost the same regardless of how the remainder of the core is modeled.

NUSCo's choice of a 34 channel nodel with a 3x3 hot channel region layoutis therefore acceptable. In Addendum A to NUSCO 140-2, which was

submitted with an August 8,1986 letter (Ref.11), NUSCo perfomed asensitivity study using one eighth of the core, with the number ofchannels ranging from 8 to 28. It was found that with a 19 channel modelwhere the hot channel is located close to the center of the core and issurrounded by many rows of sub-channels, the resulting DNBR calculated isslightly more conservative than the 1/4 core model. Therefore, NUSCO's

proposal to use a 1/8 core 19 channel model for reload calcula;tions isacceptable. -

(2) An axial noding study was done with both uniform and non-unifom nodingsizes with the number of axial nodes ranging from 11 to 55. The results

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show little effect with the maximum difference in DNBR of about 1%.NUSCo's choice of an axial node size of 2.53 inches in the minimum DNBRregion and a node size of 5.06 inches elsewhere provides sufficient detail

i for reliable results._. _ _ _ _ _ _ _ . _ . _ _ _ . _ . _ _ . . _ _ _ _ _ _ _ _ -. _ . . . _. _ . _ _ .. .

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l(3) A study with respect to centroid distance was done with various values of.i

the gap width-to-centroid distance ratio (S/L) ranging from 0.25 to 1.0 |and various centroid distance values. The results show almost identical |minimum DNBRs in all seven cases studied. This confims the studies in I

the VIPRE-01 report (Ref. 3) showing the VIPRE-01 solution to M insen-.

sitive to the selection of centroid distance (L). NUSCo's use of the ,

actual physical 'istance between the centroids of adjacent channels asd

input for L is realistic and acceptable.!

(4) The effect of the single phase friction factor of, on DNBR was studied by'

using the Blasius relation and Waggener correlation and using various. values of coefficients for these two correlations. The results show

4 - kVIPRE-01 to be fairly insensitive to f under nominal high power or-

intennediate flow conditions, but slightly sensitive to changes in f' under low flow / low power conditions (a maximum difference of 3% in DNBR,

was observed). NUSCo will use the VIPRE-01 default options of the Blasiusrelation and Poiseville relation for the turbulent and laminar flows, i

respectively, which are commonly used in the industry.,

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The calculation of f is generally based on the bulk fluid temperature in anode. VIPRE-01 also has an option of using the Rohsensow-Clark viscosity

I model to modify f to account for the fluid viscosity variation near ai

heated surface. Even though a sensitivity study shows the use of theRohsensow-Clark model to have insignificant effect on the VIPRE-01

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results, this model will be used by NUSCo. i\.

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(5) The local loss coefficient effect is studied by doubling or halving the?

fonn loss coefficients for the mixing vane spacer grids and the fuel i

assembly inlet and outlet fittings. The results show that the minimum )DNBRs vary by no more than 1% for nominal, high power and intermediate j

flow, and by a maximum of 2.5% for the low flow / low power conditions. I

NUSCo will use the form loss coefficients which were obtained.for theConnecticut Yankee fuel assemblies. This is acceptable. For other plantssuch as Millstone Unit 2 appropriate fom loss coefficients pertinent tothe plant specific feul designs should be used in the licensingcalculation.

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(6) The crossflow resistance effect was studied using various values of thee

gap loss coefficient ,Kg, (ranging from 0.25 to 1000) and various Kg func-tions available in VIPRE-01 (such as Kg being a constant, a function ofReynolds' number, or proportional to the number of rows between centroids).The results show that VIPRE-01 is very insensitive to the va10 of Kg.NUSCo's choice of Kg equal to 0.5 which is recommended by the VIPRE-01

. report is acceptable..

(7) Various turbulent crossflow correlations, turbulent mixing coefficientsand turbulent momentum factors (FTM) were studied. The results show thata change in the turbulent crossflow correlation results in no more than 4percent change in the calculated minimum DNBR, that an increase of themixing coefficient from 0.01 to 0.04 results in an increase of minimum-

DNBR by about 3.5% for the high power and intermediate flow conditions,Y and that a change in the FTM between 0.0 and 1.0 produces an

insignificant effect on the VIPRE-01 results. NUSCo will use a VIPRE-01recommended value of 0.8 for the FTM, and a mixing coefficient of 0.019which is a conservative value for the HNP mixing vane grid fuel assemblies.

(8) Because of the smaller diameter at the 21-inch section at the bottom ofeach thimble, the channel flow area, wetted perimeter and gap width

! I will vary axially along the length of a channel. The effect of thenon-uniform axial geometry was studied. The results show very littlesensitivity of the VIPRE-01 results to various applications of the

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axial geometry variation modeling. Therefore, NUSCo's decision not touse the axial geometry variation is acceptable.

(9) VIPRE-01 has three options for obtaining the water properties: (a)awater property table created by VIPRE-01 with a user specified range andnumber of table entries, (b) the EPRI water property functions, and(c) user input. A sensitivity ::tudy using the first two methods shows'

that the VIPRE-01 solution is insensitive to the method of water propertiesgeneration. VIPRE-01 also has the option of using a unifonn pressure ora local pressure in detennining the fluid properties. A sensitivitystudy of the different pressure options shows VIPRE-01 calculations to beinsensitive to the options used. This is because the pressure drop acrossthe core is relatively small compared to the system pressure of 2200 psia.Therefore, NUSCo's use of a 24 entry table generated by VIPRE-01 and thelocal pressure option in the water properties calculation is acceptable.

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(10) There are many two-phase flow empirical corrections in VIPRE-01. Fore

subcooled boiling, the homogeneous equilibrium model, the Levy and the

EPRI correlation are available for calculating subcooled quali;ty. Thevoid / quality relationship includes the Zuber-Findlay void drif_t fluxcorrlation, the Zuber-Findlay-EPRI void model, the Armand slidcorref ationand the homogeneous model. For the two-phase friction multiplier, theoptions available include the homogeneous fomulation, the EPRI, Armandand Beattie correlations. However, certain combinations of correlationsare preferred to be used as a package. For the sensitivity s,tudy, a totalof 10 combinations of correlations are used. The results show that thechannel exit void fraction varies widely in some cases due to the use ofthe homogeneous void / quality relationship which results in a larger void

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fraction compared to other models considering vapor / liquid slip. However,-

. the minimum DNBRs calculated vary by less than 1%. Therefore, for the'

purpose of minimum DNBR analysis, VIPRE-01 results are relativelyinsensitive to the selection of correlations, and the EPRI package will beused by NUSCo for the subcooled boiling, void / quality relation and thetwo-phase multiplier calculations. This is acceptable.

(11) A sensitivity study was also performed for the heat transfer correlationsavailable in VIPRE-01 with respect to subcooled forced convection, nucleate>

I boiling and the critical heat flux. The study was performed with five setsof assumed operating conditions. The results show that VIPRE-01's calculations

of the minimum DNBR and flow conditions are insensitive to the heat transfercorrelations selected with the minimum DNBRs differing by less than 0.5 percentar.ong the correlations studied. The fuel centerline temperature and claddingtemperature calculated with the Chen correlations are higher than those cal-ulated with other correlations because the Chen correlations give considerably

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lower heat transfer coefficients. NUSCo will use the VIPRE-01 default heattransfer correlations, i.e., the EPRI single phase forced convection correlationand the Thom correlation for nucleate boiling.

The W-3 critical heat flux correlation with single grid factorvill beused for CHF and DNBR calculations. The use of the W-3 correlation is

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justified because the Haddam Neck fuel assemblies are Westinghouse fuel

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designs with simple mixing vane grids. In applying the spacer grid,

factor to the W-3 correlation, a thennal diffusion coefficient (TDC) isneeded to account for the turbulent mixing effect of the mixing vane. Asensitivity study was perfonned with the TDC ranging from 0.059 to 0.064.

I The results show that decreasing the TDC from 0.032 to 0.019 Osults inless than a 1.2% decrease in DNBR and increasing the TDC from 0.032 to 0.064results in a 2 percent increase in minimum DNBR. NUSCo will use the VIPRE-01

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recommended value of 0.032 for the TDC. We find that the use of 0.032 forthe TDC is acceptable.

However, though the W-3 correlation has been accepted with a DNBR limitof 1.3, the W-3 correlation and its spacer grid correction factor weredeveloped with the Westinghouse THINC thermal hydraulic code. NUSCo is-

therefore required to perform an analysis to demonstrate that the DNBR' ' limit of 1.3 for the W-3 correlation used in VIPRE-01 can predict the W-3

corre-lation data base of DNB occurrence with at least a 95 percentprobability at a 95 percent confidence level. For other plants, similaranalysis should be done for their CHF correlations (such as CE-1) todemonstrate the appro-priateness of the DNBR limits before theirapplications with the VIPRE-01 code.

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! (12) A study was done using the various numerical solution techniques ofVIPRE-01, i.e...the UPFLOW and RECIRC options with various convergence

criteria. The results show almost identical minimum DNBRs calculatedwith these solution techniques. NUSCo will use the RECIRC option and theVIPRE-01 default convergence limits. This is acceptable.

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(13) The time step size effect was studied using various time step sizes andtime step controls including variable step size and automatic step size.With the time step sizes ranging from 0.01 to 0.5 seconds, the

results show that the difference in the calculated minimum DNBR is4

minimal when the time step is less than 0.1 second. NUSCo wil.1 use the

time step nizes between 0.05 and 0.1 seconds depending on the nature ofthe transients analyzed. The use of these time step sizes w11.1 also1

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| result in the Courant number being greater than 1 since the axial nodei size for the HNP core model is 2.53 inches in the minimum DNBR region.

Therefore the time step sizes to be used are acceptable.

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(14) A sensitivity study was also performed for the fuel rod modeling.

including the axial power profile, fuel pellet power distribution,fuel-clad gap conductance, fuel rod nodalization and heat deposition inthe cladding. ."

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The axial power profiles studied are a chopped cosine, a UsinU top peak,'

and the HNP Cycle 12 BOL shape. The results show that the top peak shapeprofile produces the most limiting DNBR. NUSCO stated that an axialpower shape which ensures reasonably conservative results in the licensingcalculation will be used.

The radial power profile within the fuel rod was studied using a uniform*

density and a depressed (80% of nominal) center power density. The-

results show identical minimum DNBRs with the depressed center powerT density producing a much lower centerline temperature. The uniform power

density distribution will be used.

The fuel temperature, cladding surface heat flux and the minimum DNBR arestrongly affected by the gap conductance for the transient analyzed. Alower gap conductance results in a higher fuel centerline temperature andalso higher minimum DNBR. Therefore, the choice of a gap conductance

! depends on the transient considered as to whether the maximized orminimized gap conductance produces a conservative result.

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The number of radial fuel nodes has no effect on the minimum DNBRcalculation but does have an effect on the calculated fuel centerlinetemperature. The NUSCo will use a 10 radial node model, which is areasonable number.

The fractional power deposition in the cladding will lower the fuelcenterline temperature. Therefore no power deposition in the claddingwill be used.>

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(15) An inlet flow maldistribution with 80% flow through the hot assembly.

results in a minimum DNBR about 1.5% lower than the uniform inlet flowcase. This is due to flow redistribution within the first few feet fromthe inlet. However, for reload analysis, the inlet flow distrjbution andother operating parameters such as pressure, flow rate, powerSnd inlettemperature will be chosen to include all uncertainties to produceconservative results.

2.2 VIPRE Benchmarks:

Chapter IV of the topical report provides descriptions of the VIPRE-01benchmark studies performed by NUSCo for verification of the VIPRE-01 code and

the VIPRE-01 model to be used for licensing calculations. The benchmarks were-

done by comparing the VIPRE-01 results to comparable COBRA-IIIC analysis' results, the HNP final design safety analysis (FDSA) results, the HNP technical

specifi-cations core safety limit curves, and the available HNP core exittemperature data.;

The comparison to the COBRA-IIIC calculation was done using the HNP Cycle 7<

operating parameters and a VIPRE-01 core model essentially identical to that ofCOBRA-IIIC. The results show very good agreement between the two codes in the

I predictions of pressure drop, enthalpy, mass flow and crossflow. The VIPRE-01calculated minimum DNBR is slightly conservative with a difference of less,

than 1%. There is a slight difference in the heat flux due to the fact that.COBRA-IIIC calculates the heat flux at the mid-point of an axial node whereasVIPRE-01 calculates an integrated averace heat flux over the axial node.,

The comparison to the HNP FDSA was done with a full power rod ejection

transient at beginning of life and a four pump loss of flow transient. Exceptfor these parameters for which no infornation is available in the FDSAdocument, such as the axial power shape, the input to VIPRE-01 uses the same

| input used in the FDSA. For the rod ejection transient, the result,s show goodagreement in the fuel centerline temperature, but considerably lower fuelaverage temperature and cladding temperature for the VIPRE-01 prediction. Thisis attributed to

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the fact that DNB is predicated in the FDSA but is not predicted in VIPRE-01..

However, the ninimum DNBR calculated by VIPRE-01 for the rod ejection transient

is 1.1 which is lower than the DNBR limit of 1.3 for W-3. Therefore, if the'

DNBR limit of 1.3 rather than 1.0 is used in the determination of 9NB, thej VIPRE-01 prediction of.the cladding and fuel temperatures may actus11y be in better

agreement with that of the FSDA. This is confirmed by a reanalysis of the fullpower rod ejection case described in Addendum A to NUSCo 140-2 (ref. 11).

For the four pump loss of flow transient, the results show good agreement in<

the minimum DNBRs with the VIPRE-01 predictions slightly conservative.,

A VIPRE-01 analysis was done to generate core safety limit curves to comparewith the limit curves in the HNP Technical Specifications which were generated-

using COBRA-IIIC. The core safety limits are the loci of points of thermalT power, pressure, and temperature for which the minimum DNBR limit is not

; violated. The VIPRE-01 analysis used is almost the same model (including thesame design heat flux factor and enthalpy rise hot channel factor) as the

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COBRA-IIIC analysis except for some slight differences in the fuel geometrymodel to reflect the difference in the fuel geometries between Cycle 12 andCycle 6 of HNP fuel designs. The VIPRE-01 analysis also uses a slightlydifferent spacer grid loss coefficient and the EPRI correlations available in

! VIPRE-01. The results show that the VIPRE-01 predictions are in very goodagreement with the present safety limit curves.

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A comparison was also made for the VIPRE-01 predictions to the HNP exit

| temperature distribution obtained using 48 thermocouples positioned near theexit of 48 fuel assemblies. These data were recorded regularly throughout thefuel cycle. Except for a few assemblies in the core periphery, the results

| show most VIPRE-01 calculations of the exit temperature are within the

; measurement uncertainty of the data.

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For those peripheral assemblies where the calculated exit temperatures anddata differ by more than the measurement uncertainty of about 3 percent, the

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|VIPRE-01 calculation underpredicts the exit temperature for the lower powerassemblies while it overpredicts the exit temperature for the high power '

assemblies. To evaluate whether this discrepancy is due to underprediction of[

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crossflow and turbulent mixing by VIPRE-01, NUSCo reran VIPRE-01 by changing.

the crossflow resistance, turbulent mixing coefficient and friction factor,respectively, for each run. The results show little effect on the ,VIPRE-01calculated exit temperature. Therefore NUSCo attributed the discrepancy tothe measurement inaccuracy of the exit temperatures in the periphedalassemblies.

3.0 SUMMARY AND CONCLUSION

The staff has reviewed topical report NUSCo 140-2 and finds that the studies per-fomed for the VIPRE-01 code are acceptable for establishing the input valuesand selection if the available correlation options and solution techniques forlicensing calculations. Our review findings are summarized below.-

(1) Since the sensitivity studies were perfomed using Haddam Neck plantT

fuel design, the results are applicable to Haddam Neck only. Other plantssuch as Millstone Unit 2 having different fuel designs will require aseparatesensitivity study to establish the appropriate plant specificfuel-dependent inputs for licensing calculations.

(2) Since the W-3 correlation with the spacer grid correction factor will be! used in the licensing calculation, NUSCo should perform an analysis to

demonstrate that the minimum DNBR limit of 1.3 for W-3 used in VIPRE-01can predict its data base of DNB occurrence with at least a 95 percent .probability at a 95 percent confidence level. Similar analysis should bedone for other CHF correlations before their licensing applications withthe VIPRE-01 code.

(3) In order to maintain a consistent configuration in the EPRI developedVIPRE-01 code, NUSCo should abide by the quality assurance program

established by EPRI and connitted to by the UGRA regarding VIP.RE-01modifications. Otherwise, any new VIPRE-01 version with modificationsnot following the EPRI Q/A program should be assigned an NUSCo.desig-nation to disassociate it from the EPRI developed VIPRE-01 code. Inany case, any significant change to the VIPRE-01 code will reoufre staffreview and approval prior to licensing application.

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REFERENCES,

1. Letter from W. G. Counsil (NUSCo) to J. R. Miller (NRC), "Haddam Neck

Plant, Millstone Nuclear Power Station Unit 2. In-house Reload ,Capability". Docket Nos. 50-213,50-336,B11278, July 30,198C

;

2. NUSCo 140-2, "NUSCo Thermal Hydraulic Model Qualification Volume II

(VIPRE)" August 1, 1984, Northeast Utilities Service Company.

3. EPRI-NP-2511-CCM, "VIPRE-01, A Thermal-Hydraulic Analysis Code for>

Reactor Cores" Volumes 1 through 4. EPRI, April 1983, Revision 1,November 1983, Revision 2, July 1985.

: -

4. D.S. Rowe, " COBRA-IIIC: A Digital Computer Program for Steady-State and* Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements",

Richland, Washington: Pacific Northwest Laboratory, March 1973, BNWL-1695.1

:

] 5. C.L. Wheeler, et al., " COBRA-IV-I: An Interim Version of COBRA for; Thennal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores",

Richland, Washington: Pacific Northwest Laboratory, March 1973,;*

BNWL-1962.

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6. Bowring, R.W., and P. Moreno, " COBRA-IIIC/MIT Computer Code Manual,"

Prepared by Massachusetts Institute of Technology for EPRI, March 1976._

i7. T.L. George, et al . , " COBRA-WC: A Version of COBRA for Single-Phase

Multiassembly Thermal-Hydraulic Transient Analysis", Richland, Washingtonj Pacific Northwest Laboratory, July 1980, PNL-3259.!

8. Letter from C.E. Rossi (NRC) to J.A. Blaisdell (NUSCo), " Acceptance for,

: Referencing of Licensing Topical Report, EPRI NP-2511-CCM, VIPRE-01: A

) Thermal-Hydraulic Analysis Code for Reactor Cores, Volumes 1,.2, 3, and

; 4," May 1, 1986. -

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-- . - - ,,--r -.w .--.--..---...---.,----e--.,-y., - , , . , - - - -,,, ., ., en wem, -e,-w,..-e.,-ww-.--,--m.-. - ,e ,..--i.,_-m. . - , -. - , - - ,e.e.,. - -. , - - - - - - - . ._ , -

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9. Letter from R.A. Clark (NRC) to D.M. Musolf (Northern States Power,

Company), Docket Number 50-282 and 50-306, March 24, 1983.

10. Letter from H.N. Berkow (NRC) to J.H. Taylor (B&W), "Acceptanee forReferencing of Licensing Topical Report BAW-10156, LYNXT - Cori *

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Transient Thermal-Hydraulic Program", December 3, 1985.

11. Letter from J. F. Opeka (NUSCo) to C. I. Grimes (NRC) and A. C. Thadani(NRC), "Haddam Neck Plant, Millstone Nuclear Power Station, Unit 2 In-

house Reload Capability" Docket Nos. 50-213, 50-336, B12208, September 5,1986.

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ENCLOSURE 2.

SALP INPUT FOR NUSCO TOPICAL REPORT NUSCO 140-2

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1. Management Involvement and Control in Assuring Quality'

The quality of submittals is very good

Rating: Category 2

2. Approach to Resolution of Technical Issues from a Safety StandpointUnderstanding of issues is apparent.-

* Rating: Category 2

3. Responsiveness to NRC Initiatives

The licensee was in general very responsive to staff questions;

Rating: Category 2

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