ML20215H625

Forwards Info Requested by NRC to Support 870303 Submittal, Changes to 1981 LOCA Model for Operation of Unit 1 Cycle 12, Including Loading Pattern,Relative Power Distribution Graph & Addl Info on Proposed 1981 LOCA Model Changes

ML20215H625
Person / Time
Site:	Prairie Island
Issue date:	04/08/1987
From:	Musolf D NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8704200465
Download: ML20215H625 (16)

Text

,

Northern States Power Company 414 Nicollet Mall Minneapoks. Minnesota 55401 Telephone (612) 330-5500 April 8, 1987 10 CFR Part 50.46 Director (

office of Nuclear React <>r Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAI!11E ISLAND NUCLEAR GENERATING PLINT Dociet Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Surplemental Information to Support the Changes to the 1981 LOCA Model Attached is informition requested by the NRC Staff to support the March 3, 1987 submittal entitled: Changes to the 1981 LOCA Model for Operation of Unit 1 Cycle 12.

Attachment 1 contains the Unit 1 Cycle 12 loading pattern (Figure 1), a graph of Relative Power distribution for the Westinghouse and Exxon assemblies as a function of cycle exposure (Figure 2). Tables 1, 2 and 3 provide the relative assembly power for assemblies at 0.1 CWD/MTU, 8 GWD/MTU and 14.0 GWD/MTU.

Attachment 2 coitains additional information on the proposed changes to the 1981 LOCA Model. The proprietary version of this attachment will be submitted unde: separate cover.

D & N-e David Musolf Manager - Nuclear Support Services DMM/TMP/tp c: Regional Administrator-III, NRC NRR Project Manager, NRC Resident Inspector, NRC MPCA Attn: F W Ferman G Charroff Attachment 2: 1) Unit 1 Cycle 12 Relative Power Information

2) Responses to Questions on the Changes Made to the Large Break LOCA Analysis for Prairie Island 8704200465 87040s PDR l P

ADOCK 05000282 pgg

m Attachment 1 I

-* PRAIRIE ISLAND UNIT 1, CYCLE 12 LOADING PATTERN BY RELOAD P11 CYC8 RELOAD,TWICE BURNED 2.55 W/O U235 - Exxon 'IOPROD P11 CYC10 RELOAD.TWICE BURNED 3.62 W/O U235 - Exxon 'IOPROD

$ Pl1 CYC11 RELOAD ONCE BURNED 3.80 W/O U235 - Westinghouse OFA Pl1 CYC12 RELOAD, FRESH 3.80 W/O U235 - Westinghouse OFA l

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NSP-NUCLEAR ANALYSIS DEPT.

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0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 CYCLE EXPOSURE (GWD/MTU)

RELATIVE ASSEMBLY AVERAGE POWER FOR EXXON AND WESTINGHOUSE FUELS FOR PRAIRIE ISLAND UNIT 1 CYCLE 12 FIGURE 2

Attachment 1 TABLE 1 PRAIRIE ISLAND UNIT 1 CYCLE 12 POWER DIST. @ 100 MWD /MTU,HFP 7 8 9 10 11 12 13 G 0.85 1.32 1.03 1.23 1.26 1.16 0.78 H .1.32 1.07 1.20 1.03 1.34 1.10 0.40 1 1.03 1.20 1.00 1.12 1.15 1.05 J 1.23 1.03 1.12 0.88 0.97 0.59 K 1.26 1.34 1.15 0.97 0.39 L 1.16 1.10 1.05 0.59 M 0.77 0.40 NSP-NUCLEAR ANALYSIS DEPT.

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PRAIRIE ISLAND UNIT 1 CYCLE 12 POWER DIST. @ 8000 MWD /MTU,HFP 7 8 9 10 11 12 13 G 0.82 1.26 0.97 1.35 1.14 1.06 0.73 H 1.26 0.97 1.10 1.02 1.22 1.13 0.41 I 0.97 1.10 1.02 1.35 1.13 0.98 J 1.35 1.02 1.35 1.00 1.09 0.61 l

K 1.14 1.22 1.13 1.09 0.46 L 1.06 1.13 0.98 0.61 l

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Attachment 1 TABLE 3 PRAIRIE ISLAND UNIT 1 CYCLE 12 POWER DIST. @ EHFP (14,864 MWD /MTU) 7 8 9 10 11 12 13 G 0.85 1.24 0.96 1.29 1.10 1.06 0.79 H 1.24 0.97 1.08 1.00 1.17 1.15 0.46 1 0.96 1.08 1.01 1.31 1.10 1.00 J 1.29 1.00 1.31 1.00 1.12 0.66 l

I U K 1.10 1.17 1.10 1.12 0.51 L 1.06 1.15 1.00 0.66 M 0.79 0.46 NSP-NUCLEAR ANALYSIS DEPT.

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Attachm:nt 2

" Westinghouse Class 3" RESPONSE TO QUESTIONS ON THE CHANGES MADE TO THE WREFLOOD INPUT IN THE LARGE BREAK LOCA ANALYSIS' FOR PRAIRIE ISLAND Nuclear Safety Department March 1987 4

WESTINGHOUSE ELECTRIC CORPORATION

! Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 I

l

Attachment 2 4

A Large Break Loss-Of-Coolant-Accident analysis was recently performed by Westinghouse using the 1981 Evaluation Model for the Prairie Island Nuclear Plant operated by Northern States Power. The results of this analysis were transmitted by Westinghouse to Northern States Power in a report entitled " DEMONSTRATION OF THE CONFORMANCE OF PRAIRIE ISLAND UNITS TO APPENDIX K AND 10CFR50.46 FOR LARGE BREAK LOCAS". The report was subsequently submitted to the NRC in February / March 1987 for review. The rcport contains the following passage:

"The release of metal heat during the reflood transient is limited by conduction heat transfer, so that a solutien of the conduction equation provides a realistic representation of reflood metal heat release. Such an approach has been found to be an acceptable modeling of the release of metal heat during reflood (Reference 8). HREFLOOD input values were adjusted to simulate the conduction limited metal heat releases during reflood.

In a telephone conference on March 23, 1987 between representatives of wastinghouse, Northern States Power, and the NRC, Mr. W. Jensen, NRC rcviewer for the report, forwarded several questions regarding the rGport. Mr. Jensen requested additional information and clarification of -

the above-referenced passage with respect to changes in metal heat release input values in WREFLOOD. Specifically, he requested information on how the changes were determined and the justification for the changes.

The Large Break LOCA analysis for Prairie Island was performed with the 1981 Evaluation Model as described in WCAP-9220-P-A, Revision I (Raference 1). The 1981 Evaluation Mi> del (81 EM) consists of the following codes: SATAN-VI, WREFLOOD," COCO, and IDCTA-IV. The SATAN-VI l

codo (Reference 2) calculates the thermal-hydraulics of the reactor coolant system during the blowdown portion of the transient and actchlishes the initial conditions for the WREFLOOD code (Reference 3).

WREFLOOD calculates the refill and reflood system hydraulics. The COCO code (Reference 4) operates interactlyely with the WREFLOOD code to svaluate the containment response. The LOCTA-IV code (Reference 5) calculates the thermal response of the fuel rods in the core. The SATAN VI code provides the core average and hot channel fluid conditions during blowdown to the LOCTA-IV code which, in turn, calculates the fuel cladding tGmperature response. The WREFLOOD code provides the core flooding rate, inlet subcooling, and system pressure to the LOCTA-IV code which then uses the FLECHT heat transfer correlation to predict fuel rod cladding tsmperature response during reflood.

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Attachment 2 No changes were made to the WREFLOOD code or any of the other codes comprising the 1981 Evaluation Model. The codes described above were the codss used to determine the Peak Clad Temperature in the Large Break LOCA cnnlysis of Prairie Island. The code versions used in this analysis were the versions which were, at that time, under computer configuration control in accordance with the Westinghouse Quality Assurances program, and were not modified in any way. All code versions were reviewed and determined to be the appropriate ones for use as part of the 1981 l Evaluation Model as described in Reference 1.

The section of the report quoted above refers only to the refinement of four WREFLOOD input values and should not imply a change to the WREFLOOD computer c.nde. All inputs to these codes were determined through sngineering calculations based on plant geometry, plant Technical Spncifications, and other substantive data in accordance with the anthodology developed by Westinghouse for use with the 1981 Evaluation Modal. Westinghouse has internally compiled a set of standard methods for the determination of many of the inputs to the approved ECCS codes.

l Howsvar, in some instances, these methods are refined to provide a more i

occurate calculation of one or more input values for a specific application. The only modification to the standard input calculation anthodology for the Prairie Island Large Break LOCA analysis was the i

calculation of the inputs used for the determination of Lower Plenum and

! Lowar Downcomer metal heat releases during reflood. These input values are used only in the WREFLOOD code.

Tha WREFLOOD code (Reference 3) has long been the primary tool used in the cniculation of the system refill /reflood hydraulics in Westinghouse Large Break LOCA analyses. WREFLOOD takes a simplistic approach to the calculation of metal heat releases during the reflood transient. The total heat release from Lower Plenum and Lower Downcomer metal components in modeled as an exponentially increasing function asymptotically approaching the total initial heat available in the metal components at thn initiation of the accident. No credit is taken for the blowdown cooling of the metal structures which would quickly reduce the metal surface temperature resulting in conduction-limited heat fluxes through ths lower conductivity stainless steel structures. The total available heat is calculated as the product of the metal mass, the constant pressure spncific heat of the metal, and the temperature difference between the I

steady-state vessel inlet temperature and the containment temperature.

Tho " decay constants" used to describe the rate of heat release are l typically taken to be standard values based on generic conduction l calculations which were conservative and bounding for all plant Lower Plenum and Lower Downcomer configurations. Such a calculation is conservative since it does not reflect the blowdown cooling transient which would create a temperature gradient in the outer layers (exposed to coolant) of the metal components. The heat flux actually delivered to the enolant reflects this conduction-limiting temperature gradient, decaying with time as the thermal gradient penetrates into the solid structure and tha effective conduction path becomes larger.

Attaciment 2

. In order to appropriately refine the WREFLOOD input values to more accurately represent the conduction limited metal heat release, it was n cassary to establish a curve of initial heat release as a function of tico from which the WREFIh0D input values could be calculated.

Establishing a conduction limited heat release curve applicable to Prairie Icland required a numerical solution to the conduction equation for the Prairie Island Lower Plenum and Lower Downcomer metal components. The cost readily available tool for such a calculation was that used in the INTERIM REFLOOD code i.e., subroutines for the numerical solution of the conduction equation.

Tha INTERIM REFLOOD code is a WREFLOOD code version developed for use with th3 BART code (Reference 6). The conduction solution contained in the INTERIM REFLOOD code subroutines takes a more mechanistic approach to the calculation of metal heat releases during reflood through the modeling of Lowar Plenum and Lower Downcomer metal components as metal slabs, cylinders cr hollow spheres. Metal heat flux into the fluid is then calculated by numerical solution of the transient conduction equation.

Tha heat flux is calculated to be conduction limited due to the large heat trcnsfer coefficient existing between the metal components and the coolant. The pertinent discussion of this calculation along with the SER cover letter for this topical appear as Exhibit 1 of this report.

Sections 1.6 and 1.7 of WCAP-9561-P-A, Addendum 3, Revision 1 (Raference 7) contain a brief discussion of the conservatism of the 81 EM WREFLOOD version using the original input values. Pertinent sections of this topical report along with the SER cover letter for this topical appsar as Exhibit 2 of this report. WREFLOOD input values were modified, based on the conduction solution results such that the metal heat calculated by the 81 EM in WREFLOOD would better agree with the solution to the conduction equation as shown below.

i

. - Attaciment 2 WREFLOOD performs separate calculations of reflood metal heat releases for thick and thin metal components in the Lower Plenum and Lower Downcomer.

consequently, the standard input values used for the thick and thin metal daccy constants are different, and the metal heat energy remaining at Cottom-Of-Core Recovery time is calculated individually for thick and thin actal. The calculation of metal heat release is based on four WREFLOOD input values which are defined as follows: _, a, c 6 m S

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. Attactment 2 argnitude of overconservatism contained in the WREFLOOD calculation of cstal heat releases before input refinement. The curve upon which the r0 vised WREFIDOD inputs are based is labelled " SIMULATED INTERIM REFIDOD" and approximates the conduction limited " INTERIM REFLOOD" curve. The curves are coincident at the time of Peak Clad Temperature (170 s. after Bottom-Of-Core Recovery) and near the point of maximum curvature. The fourth curve co'ntained in Figure 1 is labelled "WREFIDOD W/ NEW ALPHA"[

1

]f# The thin metal curves demonstrate similar behavior, but differ in magnitude from the thick metal curves of Figure 1.

.ab i Tha values of[ Jestablished by the above l prccess and based on the " SIMULATED INTERIM REFLOOD" curve were then used

! Os input to the WREFLOOD code as part of the normal 1981 Evaluation Model l ccquence.

Tha use of WREFLOOD input values determined by the above procedure allows for an accurate modeling of the Lower Plenum and Lower Downcomer metal l hant releases during the reflood transient. The metal heat releases as a l

function of time were based on a transient conduction calculation which is l technically reasonable and appropriate for the calculation of metal heat l rolcases during reflood. The calculational technique used to establish tha base curve was performed through computer calculation employing cquations consistent with newer, approved safety analysis methodology.

l The Large Break LOCA analysis performed for the Prairie Island Nuclear Plant employed the 1981 Evaluation Model as described by WCAP-9220-P-A, Ravision 1. The modifications to the input values used for determining Estal heat releases during the reflood transient in the WREFLOOD code were bussd on the numerical solution of the conduction equation for the release of catal heat. Care has taken to ensure that the input values employed wara consistent with technically appropriate and recognized standards for thm determination of total metal heat releases for Lower Plenum and Lower Downcomer metal components during the reflood portion of the Large Break IDCA.

Attachment 2

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i Figure 1 contains four curves corresponding to various functions of integrated (total) metal heat releases during reflood as a function of tiSo after Bottom-Of-Core Recovery. The curve labelled " INTERIM REFLOOD"

raflects the thick metal heat release calculated by solving the conduction l

l cquation using the INTERIM REFLOOD code. The curve labelled "WREFLOOD W/

STANDARD ALPHA" illustrates the thick metal heat release calculated by WREFLOOD using the standard decay constant with the value of[ ]">

csiculated for Prairie Island based on standard input calculation tothodology. Comparison of these curves clearly demonstrates the t

Attachment 2 REFERENCES

1. " Westinghouse ECCS Evaluation Model 1981 version," WCAP-9220-P-A, Revision 1 (Proprietary), February 1982.

2. Bordelon, F. M., et al., " SATAN VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8302 (Proprietary), June 1974.

3. Kelly, R. D., et al., " Calculation Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary),

WCAP-8171 (Non-Proprietary), June 1974.

4. Bordelon, F. M., and Murphy, E. T., " Containment Pressure Analysis Code (COCO)," WCAP-8327 (Proprietary), WCAP-8326 (Non-Proprietary),

June 1974.

5. Bordelon, F. M., et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary), and WCAP-8305 (Non-Proprietary),

June 1974.

6. Young, M. Y., et al., "BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," WCAP-9561-P-A (Proprietary), March 1984

7. Young, M. Y., " Addendum To: BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients (Special Report: Thimble Modeling in Westinghouse ECCS Evaluation Model)," WCAP-9561-P-A, Addendum 3 (Proprietary), 1986.

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'tXHIBIT 1 k UNITS 0 STATE 8 l) f NUCLEAR REGULATORY COMMISSION -

f semenesseroes.o.c. asses *

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...** DEC211E3 Mr. 'E. P. Rahe, Jr., Manager Mucle'ar Technology Division .

P. O. Box 355 Pittsburgh, Pennsylvania 15230

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Subject:

Acceptance'for Referencing of Licensing Topical Report

• i WCAP-9561, "8 ART A-1: A Computer Code for 8est Estimate Analysis of Reflood Transients" We have completed our review of the subject topical report submitted January 15, 1980, by Westinghouse Electric Corporation letter NS-TMA-2169.

We find this report is acceptable for referencing in Itcense applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation which is enclosed. The evaluation defines

. the basis for acceptance of the report.

We do not intend to repeat our review of the matters described in the report l and found acceptable when the report appears as a reference in license .

applications except to assure that the material presented is applicable to the specific plant involved. Our acceptance applies only to the matters described in the report.

• 1 In accordance with procedures established in NUREG-0390, it is requested

• that Westinghouse publish accepted versions of this report, proprietary and non-proprietary, within three months of receipt of this letter. The accepted versions should incorporate this letter and the enclosed evaluation between the title page and the abstract. The accepted versions shall include an -A (designating. accepted) following the report identification symbol. ._

Should our criteria or regulations change such that our conclusions as to the acceptability of the report are invalidated Westinghouse and/or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the topical report without revision of their respective documentation. -

5facerely. .h G. Sedh .

QCecil0. Thomas, Chief Standardization & Special

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Projects Branch Division of Licensing .

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Enclosure:

As stated . .

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Attachment 2 I-EXHIBIT 1 1-During blowdown and reflood, stored energy is released to the fluid from metal components and walls in the downconer and lower plenum. This heat release raises the temperature of the liquid entering the core, which in turn affects core fluid and heat transfer conditions. .

E

• The present WREFLOOD code uses a staple, exponential decay heat release model for two components defined as thin and th::k metal. The decay constants used ensure a conservative amount of hot metal neat release to the fluid.

It is desirable to obtain a more accurate measure of hot metal energy release when using the BART code. Accordingly, the following model has been added to WREFLOOD. #

The various components in the downconer and lower plenum are arranged into groups of slabs, cylinders, and hollow spheres. As many as ten different geometries are allowed in both downconer and lower plenum. Table 5-1 shows a typical breakdown of components for a PWR. .

The surfaces of the various compc ents are assumed to remain wet during the transient, if the quality in the downcomer and lower plenum is less than 1.0, so that the heat transfer is conduction limited. When the quality is 1.0 and

) during refill, the walls are assumed to be adiabatic.

l The metal heat flux int'o the fluid is calculated by numerically solving the conduction equation. An energy balance is then performed in WREFLOOD to calculate liquid temperature in the lower plenum and downconer. ~

5-10. BART Blockage Effects ,

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To account for blockage in BART. the basic equations in BART were modified to accept a source term, S, representing the exit of steam from or entry of steam to the flow channel due to flow redistribution'. For single-phase, flow, the conservattoe equations are as follows.

75908:1b/032884 5-25

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uwNsYAb Attachment 2 NUCLEAR REGULATORY COMMISSION -

), mammeorose.o.c.nemes Wooooo

• E. P. Rahe, Jr., Manager

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Nuclear Safet Department AUG 25 5 -

Westinghouse lectric Corporation i Box 355 -

Pittsburgh, Pennsylvania 15230-0355 .

SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT WCAP 9561 ADDENDUM 3. REVISION 1 The Nuclear Regulatory Comission (NRC) staff has completed its review of Topical Report WCAP 9561. Addendum 3. Revision 1 " Thimble Modeling in Westinghouse ECCS Evaluation Model," which was submitted with your, letter dated July 24, 1986. We find the report to be acceptable for referencing in license applications to the extent specified and under the limitations 611neated in the report and the associated NRC evaluation, which is enclosed.

The evaluation defines the basis for acceptance of the report.

We do not intend to repeat our review of the matters described in the report '

and found acceptable when the report appears as a reference in license applications, except to assure that the material presented is applicable to the specific plant involved. Our acceptance applies only to the matters described in the report.

In accordance with procedures established in NUREG-0390, it is requested that

• Westinghouse publish an accepted version of this report, proprfetary and

• non-proprietary,within three months of receipt of this letter. The accepted version shall incorporate this letter and the enclosed evaluation after the title page. The accepted version shall include an -A (designating accepted) following the report identification symbol. - ,

Should our criteria or regulations change such that our conclusions as to the acceptability of the report-are invalidated Westinghouse and/or the ,

applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the. topical report without revision of their respective documentation.

Sincerely.

r har Division of PWR Licensing-A

- W Assis ctor. *

- Office of Nuclear Reactor Regulation

Enclosure:

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EXHIBIT 2 From table 1-2 it can be seen that including the effects of empty thimbles I results in a small penalty, due to slightly lower flooding rates caused by the filling of the thimbles.

The effects presented in this table are considered typical of all plants using the 1978, 1981, and.1961 + BART evaluation models, e

6

. 1. 6 Compensati.ng Effacts A review of the metal heat transfer calculation in the version of WREFL

' used in the 1978 and 1981 models indicated that this calculation was releasing an overly ' conservative amount of heat in the downconer and lower plenum to the E water wh,1ch is flooding the core, lowering its subcooling and reducing heat transfer. This conclusion was reached by comparing the heat release i

calculated with the 1978 and 1981 model to the heat release calculated by the model used in the BART evaluation model. The reason for the differenc between the two models is that the older WREFL000 version (prior to BART) uses specified inputs to simple exponential functions to calculate metal heat.

i ,- release (2] ,

while the WREFL000 version used with SART uses a more accurate conduct;1on solution E to calculate the heat release.

Calculations were perforised with the 1981 model with revised metal heat input

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which resulted in total metal heat re' lease closer to (but still conse what the more accurate BART model version would predict. It was found that this offact more than compensa,ted for the penalty due to thimble filling (see '

Table 1-2).

1.7 Conclusions and Recommendations The results presented above indicate that, for the 1978 and 1981 versions of the Westinghouse ECCS evaluation model, sufficient margin exists in the current calculation of metal heat transfer to compensate for the effect of thimble filling. It is concluded that no further analysis is required for plants using these models.

Because these evaluation model ve' rsions are being replaced by more advanced models (BART and SASH), it is recoma* ended that any future calculations using the 1981 model incorporate the additional thimble 9

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Although this will result in a small penalty, it is anticipated -

i that analyses using the 1981 model will be requested only for plants dich '

exhibit substantial margin to the 2200*F limit. Therefore, the existing conservative metal, heat input can be retained.

1 -

Because the evaluation model using SART already contains the more accurate

metal heat release model, the effect of thimble filling cannot be offset in

' the same way as the 1981 and 1978 models.

In addition, the BART calculations are further impacted by changes in hot '

assembly power, described in the next section.

A discussion of the impact of the thimble filling effect on 8 ART analyses will be presented following Section 2.

1.8 References .

1.

' Westinghouse Emergency Core Cooling Systen Evaluation Model - Modified October 1915 version", WCAP-9168. Section 2.2. ,

2.

' Calculational Model for Core Reflooding...." WCAP-8110. Section 2.4.6.

3. '8 ART-A1...."WCAP-9N1-P-A,pg5-25.

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