ML20215D100
| ML20215D100 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Haddam Neck, 05000000 |
| Issue date: | 10/01/1986 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20215D088 | List: |
| References | |
| NUDOCS 8610100663 | |
| Download: ML20215D100 (14) | |
Text
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4-ENCLOSURE SER ON NUSCO 140-P, VIPRE-01
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 the submittal is to demonstrate NUSCo's in-house reload capability in the use of the VIPRE-01 computer code (Ref. 3) for perfoming core themal hydraulic analysis. The first planned reload application of VIPRE-01 will be for the upcoming cycle for the Haddam Neck plant. NUSCo also indicated that the application of VIPRE-01 may be extended to Millstone Unit 2 reloads.
i- VIPRE-01 is a subchannel themal-hydraulic code developed by Battelle Pacific Northwest Laboratories under the sponsorship of the Electric Power Research Institute (EPRI). In December 1984, the Utility Group for Regulatory Applications (UGRA), which consists of more than 20 utilities, submitted the VIPRE-01 topical reports for staff review. VIPRE-01 was developed from the COBRA 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 features of these codes into one package. The staff review (Ref. 8) concluded that the VIPRE-01 code was acceptable for PWR application with the following _
i conditions:
(1) The application is limited to the heat transfer modes up to critical heat flux (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 the modeling 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 a Courant number less than 1. ,
n-(5) Each user should abide by the quality assurance program established by EPRI for the VIPRE-01 code.
Our evaluation of the report will concentrate on the application of the VIPRE-01 code by NUSCo for licensing calculations.
f 2.0 STAFF EVALUATION 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 in just one computer run which accounts for the crossflow and turbulent mixing effects of not only the neighboring subchannels but the entire core. This is accomplished by a lumped subchannel core model where the subchannel layout of the hot channel and its adjacent channels within the hot assembly is modeled in 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 single pass approach has been increasingly used in the nuclear industry due to the large 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 Northeast Utilities qualification efforts on the VIPRE-01 core thermal hydraulic model with the single pass approach. These efforts include (1) sensitivity studies to establish the VIPRE-01 input parameter values and options and (2) VIPRE benchmarks 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) as a typical base model. The objectives of the sensitivity studies are to examine the effects of the important parameters and options on the VIPRE-Ofcalculated results. The studies serve as bases for the input values, options of I correlations and solution techniques available in VIPRE-01. A summary of the sensitivity study results and the selected values and options for the HNP
, VIPRE-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 PWR fuel designs, some of the other sensitivity studies are dependent on specific fuel design. Since the studies were perfomed with the HNP fuel design, the ll results of the fuel-dependent items are applicable to HNP only. A summary of the staff review of the sensitivity studies follows.
(1) Using a i core model, the study varies the number of radial channels ranging from 8 to 43 channels with the number of channels in the hot assembly ranging from 4 to 22. The resulting minimum DNBRs generally decrease with increasing radial node detail with a maximum difference of about 3% in DNBR. However, when a 3x3 subchannel array is modeled with I the hot channel completely surrounded, the resulting minimum DNBRs are almost 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 layout is therefore acceptable. In Addendum A to NUSCO 140-2, which was submitted with an August 8,1986 letter (Ref.11), NUSCo perfomed a sensitivity study using one eighth of the core, with the number of channels ranging from 8 to 28. It was found that with a 19 channel model where the hot channel is located close to the center of the core and is surrounded by many rows of sub-channels, the resulting DNBR calculated is slightly 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 is acceptable. -
(2) An axial noding study was done with both uniform and non-unifom noding sizes with the number of axial nodes ranging from 11 to 55. The results 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 DNBR region and a node size of 5.06 inches elsewhere provides sufficient detail i
for reliable results.
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. (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 l and various centroid distance values. The results show almost identical l 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 'istanced between the centroids of adjacent channels as 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 Blasius relation and Poiseville relation for the turbulent and laminar flows, i respectively, which are commonly used in the industry. ,
The calculation of f is generally based on the bulk fluid temperature in a node. VIPRE-01 also has an option of using the Rohsensow-Clark viscosity i
I model to modify f to account for the fluid viscosity variation near a heated surface. Even though a sensitivity study shows the use of the Rohsensow-Clark model to have insignificant effect on the VIPRE-01 .
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 the Connecticut Yankee fuel assemblies. This is acceptable. For other plants such as Millstone Unit 2 appropriate fom loss coefficients pertinent to the plant specific feul designs should be used in the licensing calculation.
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e (6) The crossflow resistance effect was studied using various values of the 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 of Reynolds' 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 coefficients and turbulent momentum factors (FTM) were studied. The results show that a change in the turbulent crossflow correlation results in no more than 4 percent change in the calculated minimum DNBR, that an increase of the mixing 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-01 recommended value of 0.8 for the FTM, and a mixing coefficient of 0.019 which 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 of each thimble, the channel flow area, wetted perimeter and gap width
! I will vary axially along the length of a channel. The effect of the non-uniform axial geometry was studied. The results show very little sensitivity of the VIPRE-01 results to various applications of the ,
axial geometry variation modeling. Therefore, NUSCo's decision not to use the axial geometry variation is acceptable.
(9) VIPRE-01 has three options for obtaining the water properties: (a)a water property table created by VIPRE-01 with a user specified range and number 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 properties generation. VIPRE-01 also has the option of using a unifonn pressure or a local pressure in detennining the fluid properties. A sensitivity study of the different pressure options shows VIPRE-01 calculations to be insensitive to the options used. This is because the pressure drop across the 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 the local pressure option in the water properties calculation is acceptable.
e (10) There are many two-phase flow empirical corrections in VIPRE-01. For subcooled boiling, the homogeneous equilibrium model, the Levy and the EPRI correlation are available for calculating subcooled quali;ty. The void / quality relationship includes the Zuber-Findlay void drif_t flux corrlation, the Zuber-Findlay-EPRI void model, the Armand slidcorref ation and the homogeneous model. For the two-phase friction multiplier, the options available include the homogeneous fomulation, the EPRI, Armand and Beattie correlations. However, certain combinations of correlations are preferred to be used as a package. For the sensitivity s,tudy, a total of 10 combinations of correlations are used. The results show that the channel exit void fraction varies widely in some cases due to the use of the homogeneous void / quality relationship which results in a larger void .
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 relatively insensitive to the selection of correlations, and the EPRI package will be used by NUSCo for the subcooled boiling, void / quality relation and the two-phase multiplier calculations. This is acceptable.
(11) A sensitivity study was also performed for the heat transfer correlations available in VIPRE-01 with respect to subcooled forced convection, nucleate I boiling and the critical heat flux. The study was performed with five sets of assumed operating conditions. The results show that VIPRE-01's calculations of the minimum DNBR and flow conditions are insensitive to the heat transfer correlations selected with the minimum DNBRs differing by less than 0.5 percent ar.ong the correlations studied. The fuel centerline temperature and cladding temperature calculated with the Chen correlations are higher than those cal-ulated with other correlations because the Chen correlations give considerably lower heat transfer coefficients. NUSCo will use the VIPRE-01 default heat transfer correlations, i.e., the EPRI single phase forced convection correlation and the Thom correlation for nucleate boiling.
The W-3 critical heat flux correlation with single grid factorvill be used for CHF and DNBR calculations. The use of the W-3 correlation is justified because the Haddam Neck fuel assemblies are Westinghouse fuel
. _ . . . _~ -
designs with simple mixing vane grids. In applying the spacer grid factor to the W-3 correlation, a thennal diffusion coefficient (TDC) is needed to account for the turbulent mixing effect of the mixing vane. A sensitivity 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 in less than a 1.2% decrease in DNBR and increasing the TDC from 0.032 to 0.064 results in a 2 percent increase in minimum DNBR. NUSCo will use the VIPRE-01 1
recommended value of 0.032 for the TDC. We find that the use of 0.032 for the TDC is acceptable.
However, though the W-3 correlation has been accepted with a DNBR limit of 1.3, the W-3 correlation and its spacer grid correction factor were developed 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 percent probability at a 95 percent confidence level. For other plants, similar analysis should be done for their CHF correlations (such as CE-1) to demonstrate the appro-priateness of the DNBR limits before their applications with the VIPRE-01 code.
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! (12) A study was done using the various numerical solution techniques of VIPRE-01, i.e...the UPFLOW and RECIRC options with various convergence criteria. The results show almost identical minimum DNBRs calculated with these solution techniques. NUSCo will use the RECIRC option and the VIPRE-01 default convergence limits. This is acceptable.
l (13) The time step size effect was studied using various time step sizes and time 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 is 4
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 of 1
the transients analyzed. The use of these time step sizes w11.1 also .
l result in the Courant number being greater than 1 since the axial node i 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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9 (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 in the cladding. ."
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n-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 shape profile produces the most limiting DNBR. NUSCO stated that an axial power shape which ensures reasonably conservative results in the licensing calculation 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 power T
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 are strongly affected by the gap conductance for the transient analyzed. A lower gap conductance results in a higher fuel centerline temperature and also higher minimum DNBR. Therefore, the choice of a gap conductance
! depends on the transient considered as to whether the maximized or minimized gap conductance produces a conservative result.
The number of radial fuel nodes has no effect on the minimum DNBR calculation but does have an effect on the calculated fuel centerline temperature. The NUSCo will use a 10 radial node model, which is a reasonable number.
The fractional power deposition in the cladding will lower the fuel centerline temperature. Therefore no power deposition in the cladding will 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 flow case. This is due to flow redistribution within the first few feet from the inlet. However, for reload analysis, the inlet flow distrjbution and other operating parameters such as pressure, flow rate, powerSnd inlet temperature will be chosen to include all uncertainties to produce conservative results.
2.2 VIPRE Benchmarks:
Chapter IV of the topical report provides descriptions of the VIPRE-01 benchmark 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 exit
- temperature 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 of COBRA-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-01 calculated 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 whereas
, VIPRE-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. Except for these parameters for which no infornation is available in the FDSA document, such as the axial power shape, the input to VIPRE-01 uses the same l input used in the FDSA. For the rod ejection transient, the result,s show good agreement in the fuel centerline temperature, but considerably lower fuel average temperature and cladding temperature for the VIPRE-01 prediction. This is attributed to n - . - - - - - - . - , - - - , - - - , - . . , . , - _ , , - - , - . , . - . . , - - ,
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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, the j 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 full power 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 compare with the limit curves in the HNP Technical Specifications which were generated using COBRA-IIIC. The core safety limits are the loci of points of thermal T
power, pressure, and temperature for which the minimum DNBR limit is not
- violated. The VIPRE-01 analysis used is almost the same model (including the same design heat flux factor and enthalpy rise hot channel factor) as the COBRA-IIIC analysis except for some slight differences in the fuel geometry model to reflect the difference in the fuel geometries between Cycle 12 and Cycle 6 of HNP fuel designs. The VIPRE-01 analysis also uses a slightly different spacer grid loss coefficient and the EPRI correlations available in
! VIPRE-01. The results show that the VIPRE-01 predictions are in very good agreement with the present safety limit curves.
A comparison was also made for the VIPRE-01 predictions to the HNP exit l temperature distribution obtained using 48 thermocouples positioned near the exit of 48 fuel assemblies. These data were recorded regularly throughout the fuel cycle. Except for a few assemblies in the core periphery, the results l show most VIPRE-01 calculations of the exit temperature are within the
- measurement uncertainty of the data.
l For those peripheral assemblies where the calculated exit temperatures and data differ by more than the measurement uncertainty of about 3 percent, the ,
VIPRE-01 calculation underpredicts the exit temperature for the lower power l assemblies 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-01 calculated exit temperature. Therefore NUSCo attributed the discrepancy to the measurement inaccuracy of the exit temperatures in the periphedal assemblies.
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 values and selection if the available correlation options and solution techniques for licensing calculations. Our review findings are summarized below.
T (1) Since the sensitivity studies were perfomed using Haddam Neck plant fuel design, the results are applicable to Haddam Neck only. Other plants such as Millstone Unit 2 having different fuel designs will require a separatesensitivity study to establish the appropriate plant specific fuel-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-01 can predict its data base of DNB occurrence with at least a 95 percent .
probability at a 95 percent confidence level. Similar analysis should be done for other CHF correlations before their licensing applications with the VIPRE-01 code.
(3) In order to maintain a consistent configuration in the EPRI developed VIPRE-01 code, NUSCo should abide by the quality assurance program established by EPRI and connitted to by the UGRA regarding VIP.RE-01 modifications. Otherwise, any new VIPRE-01 version with modifications not following the EPRI Q/A program should be assigned an NUSCo.desig-nation to disassociate it from the EPRI developed VIPRE-01 code. In any case, any significant change to the VIPRE-01 code will reoufre staff review and approval prior to licensing application.
, 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.
- 6. Bowring, R.W., and P. Moreno, " COBRA-IIIC/MIT Computer Code Manual,"
Prepared by Massachusetts Institute of Technology for EPRI, March 1976._
i
- 7. T.L. George, et al . , " COBRA-WC: A Version of COBRA for Single-Phase Multiassembly Thermal-Hydraulic Transient Analysis", Richland, Washington j 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. -
i I
-- . - - ,,--r .--.--..---...---.,----e--.,-y., - , , . , - - - -,,, ., ., en wem,
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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 for Referencing of Licensing Topical Report BAW-10156, LYNXT - Cori
- 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 Standpoint Understanding 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 m
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