ML20199K699

Safety Evaluation Concluding That Request to Use YAEC-1339, Yankee Atomic Electric Co Application of FIBWR2 Core Hydraulics Code to BWR Reload Analysis, at Vermont Yankee Acceptable

ML20199K699
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Site:	Vermont Yankee
Issue date:	01/20/1999
From:	NRC (Affiliation Not Assigned)
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NUDOCS 9901260489
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9 . . . . . ,d SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION HELATING TO REPORT YAEC-1339 APPLICATION OF FIBWR2 CORE HYDRAULICS CODE TO BWR RELOAD ANALYSIS l VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION POCKET NO. 50-271

1.0 INTRODUCTION

By letter dated April 14,1997, Vermont Yankee Nuclear Power Corporation, licensee for l Vermont Yankee Nuclear Power Station (VYNPS), requested that the staff review and approve i the use of their FIBWR2 computer code for transient critical power ratio (CPR) analysis (Reference 1). The licensee responded to a staff request for additional information on September 23,1997, (Reference 2). The method will be applied to VYNPS by including it as a

reference in Section 6 cf the Vermont Yankee technical specifications (TS) (References 3 and

! 4). FIBWR2 is an extension of the approved FIBWR code and contains the following new features:

a. transient capability

b. non-equilibrium fluid model by using two continuity equations and forcing the vapor phase to saturation at the system pressure

c. the ability to model modern fuel designs

d. improved coding allowing for user defined CPR correlations and more efficient execution 2.0 DISCUSSION OF NEW CODE FEATURES I FIBWR2 uses four field equations as the basis of an inhomogeneous, non-eouilibrium fluid l

dynamics model. This approach is well established and has been used for many years. This i

model allows for accurate representation of sub-cooled boiling phenomena and predictions of phasic velocities to accuracies dictated by the accuracy inherent in the semi-empirical void-quality-slip relationships. These equations are integrated using an explicit first-order method which is widely used in two-phase codes and is also well established. FIBWR2 retains all of the features of the previously approved FIBWR code and will be used as a functional replacement for FIBWR. The solution methodology assumes the following:

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a. all channels experience the same overall pressure drop;

b. the flow is one-dimensional, but not restricted to co-current upflow;

c. an average pressure is used to evaluate density; and

d. the steady state void and pressure drop relationships are valid during transient simulations.

In addition to the additional physics available in FIBWR2, the code's modeling capabilities were i extended. FIBWR2 allows modem fuel features such as part length rods and internal water l channels to be modeled. Also, FIBWR2 can accommodate a large number of parallel fuel channels including active channel, water tube and bypass flow paths.

3.0 FIBWR2 VALIDATION FIBWR2 was compared to both analytical results and experimental data. Void profile, flow decay transients, and osciliatory transient results were used as analytical benchmarks.

Experimental benchmarks consisted of comparisons to critical heat flux data from the 16 rod electrically heated test section identified as 16R-2 (References 5,6,7). FIBWR2 compared well to the analytical benchmarks. The void profile and decay transient results justify use of the void-quality-slip models in FIBWR2 and the flow model for non-oscillatory transients. The FIBWP2 calculated oscillatory transient results, although reasonably accurate, exhibited a reduction in amplitude and a phase shift which are typical characteristics of solver's with first order accuracy such as Euler's method (the solver in FIBWR2). The oscillatory results show typical agreement, but are insufficient to qualify the code for oscillatory transients.

The experimental benchmarks serve as integral benchmarks for the applications of FIBWR2 which are considered in this review. The 16 rod data was taken in test assemblies with pin designs typical of BWR fuel. Comparisons were made to 50 steady state tests with a mean of 1.007 and a standard deviation of U.0186. The approved RETRAN and TCPYA01 method when applied to the same test set leads to a mean of 1.006 and a standard deviation of 0.0182.

Comparisons were made to 20 transient tests to evaluate the ability of FIBWR2 to predict the onset of boiling transition. The root mean squared differences between the test results and FIBWR2 calculations was 0.458 seconds which is comparable to the results using the currently approved RETRAN and TCPYA01 method. A sensitivity study on the effect of void model, time step size and nodalization leads to the following conclusions:

a, there is a negligible effect on the calculated time of boiling transition; and

b. the time step and nodalization proposed in Reference 1 are adequate for BWR problems.

As discussed in reference 1, a noding study may need to be perfonned to ensure convergence.

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4.0 FIBWR2 HOT CHANNEL METHODOLOGY ,

The proposed FIBWR2 methodology is evaluated by comparing results using the new code to results from the previously approved RETRAN and TCPYA01 methodology (Reference 8). The licensee proposes to replace RETRAN and TCPYA01 with FIBWR2, but the application methodology used for RETRAN and TCPYA01 remains unchanged from Reference 8.

Comparisons for the Vermont Yankee Nuclear Power Station were made for the following l events:

a. generator load rejection without bypass;

b. inadvertent high pressure coolant injection;

c. loss of stator cooling; and l

d. loss of feedwater heating. l l Some differences are expected due to the different models used in the two methods. The l FIBWR2 results consistently underpredict the RETRAN calculated critical power ratio (CPR) with the maximum difference being 0.059. The FIBWR2 predicted CPR values for the l l decreased subcooling events are more consistent with the RETRAN results with the largest

difference being -0.007. The differences in the CPR values for the pressurization transients can be attributed to the improved void fraction predictive capability in FIBWR2 and the staff l considers the predicted differences to be acceptable. '

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5.0 CONCLUSION

S I-The staff has considered a request from the license to use the FISWR2 codes in its approved BWR hot channel methodology (Reference 8). The benchmarking and hot channel comparative results presented in Reference 1 are sufficient to qualify the code for CPR analyses. Therefore, the staff concludes that the request to use YAEC-1339, " Yankee Atomic Electric Company Application of FIBWR2 Core Hydraulics Code to BWR Reload Analysis," at

! Vermont Yankee is acceptable. A TS amendment is necessary prior to using this methodology.

An amendment is being prepared separately to incorporate the use of this approved Vermont Yankee methodology in the TS.

Principal Contributor: A. Ulses Date: January 20, 1999 i

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6.0 REFERENCES

1. Letter from James F. Duffy (Vermont Yankee) to USNRC, ' Submittal of YAEC-1339,*

April 14,1997.

2. Letter from James F. Duffy (Vermont Yankee) to USNRC, " Response to NRC request for additional information regarding YAEC-1339," September 23,1997.

3. Letter from Donald A. Reid (Vermont Yankee) to USNRC, " Technical Specification Proposed Change No.192," August 20,1997.

4. Letter from Bernard R. Buteau (Vermont Yankee) to USNRC, " Marked up Technical Spacification pages for Proposed Changes 192 and 193," September 18,1997.

5. GEAP-1022-11,

• Deficient Cooling,11th Quarterly Progress Report, January 1 -

March 31,1972," General Electric Co., April 1972.

6. GEAP-1022-12, " Deficient Cooling,12th Quarterly Progress Report, April 1 -

June 30,1972," General Electric Co., July 1972.

7. GEAP-13295, " Transient Critical Heat Flux - Experimental Results," General Electric Co.,

September 1972.

8. YAEC 1299P," Methods for the Analysis of Boiling Water Reactors Transient Critical Power Ratio Analysis RETRAN-TCPYA01," Yankee Atomic Electric Company, March 30,1982.

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