ML20213A598

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Forwards Results & Conclusions of Investigations Proposed in 861205 & s Re RELAP5YA.Small Break LOCA Demonstration Case for One PWR Will Be Submitted Prior to Completion of NRC Review & Approval,Per 870205 Discussion
ML20213A598
Person / Time
Site: Yankee Rowe, Maine Yankee, 05000000
Issue date: 04/21/1987
From: Papanic G
YANKEE ATOMIC ELECTRIC CO.
To: Mckenna E
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
FYR-87-42, MN-87-50, NUDOCS 8704280091
Download: ML20213A598 (19)
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Text

{{#Wiki_filter:. Telephon2 (617) 8778100 TWX 7103807619 l YANKEE ATOMIC ELECTRIC COMPANY \ W> 3 1671 Worcester Road. Framingham, Massachusetts 01701 

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AN KEE April 21, 1987 FYR 87-42 MN 87-50 United States Nuclear Regulatory Commission Document Control Desk kauhington, DC 20555 As untion: Ms. Eileen M. McKenna Project Manager Project Directorate #1 Division of PWR Licensing-A

Reference:

(a) License No. DPR-3 (Docket No. 50-29) (b) License No. DPR-36 (Docket No. 50-309) (c) Letter, YAEC to USNRC, FVY 83-4, dated January 14, 1986 (d) Letter, YAEC to USNRC, MN 86-127 or FYR 85-098, dated October 16, 1987 (e) Letter, YAEC to USNRC, MN 86-138 or FYR 66-107, dated November 4, 1986 (f) Letter, YAEC to USNRC, MN 86-150 or FYR 86-118, dated December 5, 1986 (g) Letter, USNRC to YAEC, NYR 87-28 or NMY 87-27, dated February 9, 1987

Subject:

RELAP5YA

Dear Ms. McKenna:

In Reference (c), Yankee Atomic Electric Company (YAEC) submitted RELAP5YA, a computer program for light water reactor system thermal-hydraulic analysis, to the NRC for review and licensing approval. References (d) and (e) provided our responses to 25 questions requiring resolution from the NRC review of RELAP5YA. As a result of the investigations performed to address these questions and provide additional code benchmarking analysis, we identified the need to modify our PWR Small Break Loss of Coolant Accident (SBLOCA) approach to better represent observed and expected core liquid inventory behavior. Our program and schedule was provided to you in Reference (f). On February 5, 1987, we met with members of the NRC staff to describe and discuss the modeling progress achieved, and the results obtained (Reference (g)). Approaches involving improvements to the downcomer region hP lok o

United States Nuclear Regulatory Commission April 21, 1987 Attention: Ms. Eileen M. McKenna Page 2 nodalization and the RELAP5YA interphase drag models had been developed. Comparisons to the semiscale S-LH-1 small break tests and additional code benchmarking capability with the LOFT Test LS-1 were presented. The staff encouraged the course of the improvement program and recommended that we , pursue additional refinements in the modeling representation for each of these tests. The schedule for proposing the final model for PWR SBLOCA analysis was maintained as March 31, 1987. Subsequent to this meeting, we discussed with you the results of investigations into the modeling of accumulator performance, particularly in regard to the LOFT Test L5-1. In addition to the refinements we had 

                       ' identified for accumulator representation, we had been informed of a RELAPS MODI coding change expected to improve modeling accuracy for accumulator performance prediction. We agreed to extend the schedule for our work to April 17, 1987 to allow us to implement and evaluate this additional model change in RELAPSYA.

. The purpose of this letter is to report the results and conclusions of l the investigations proposed in References (f) and (g). The identified workscope has been completed. We judge that the RELAP5YA modeling changes made to ~ better represent the interphase drag phenomena provide the improvements necessary to establish our licensing methodology. Attachment A describes the changes in modeling concepts and the results of comparisons to separate effects tests and integral experiments. However, as discussed in the > Attachment, certain recent input sensitivity studies resulted in code failure midway through integral test simulation. We believe the failure may be attributable to instability in the code numerical solution which becomes apparent as these types of refinements in physical phenomena modeling are introduced. 4 Attachment B provides a summary of other efforts associated with the workscope, including a discussion of the revised downcomer nodalization, LOFT L5-1 accumulator modeling representation changes, implementation and testing of the accumulator model change in RELAP5YA, and the LOFT Test L5-1 

      ,                 benchmarking results associated with each of these changes. The analysis of the LOFT Test L5-1 with the RELAP5YA code as described in Attachment A is also summarized.

We are encouraged by the improved results observed thus far with the RELAPSYA code version revised for interphase drag model enhancements. Therefore, we plan to seek a solution to the numerical instabilities observed in certain cases. Additionally, we plan to improve our representation of the Semiscale Test S-LH-1 through addition of passive metal structures. As discussed on February 5,1987, we will submit a SBLOCA demonstration case for one PWR prior to the completion of your review and approval. We will keep you informed of our progress on each item. At present, we expect to complete the solution to the code execution problems by June 1,1987. At that time, Yankee would request a meeting with the USNRC to present our completed model.

United States Nuclear Regulatory Commission April 21, 1987 Attention: Ms. Eileen M. McKenna Page 3 We trust that you will find this information satisfactory, however, should you have any questions please contact us. Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY George / ic, J Senior Project Engineer Licensing GP/dlb Attachments

ATTACHMENT A Modifications to RELAPSYA Interphase Drag Models The interphase drag models in RELAP5YA provide the interphase momentum conductance term (Fyy) required for the solution of the phasic momentum equations. The method used in RELAP5YA Version 18G for calculating F (Reference (A.1)) is similar to that in RELAP5/ MOD 1 (Reference (A.2))yy, The . features of the method are: (a) Use of four separate flow regime maps for Vertical Components (VRT), Horizontal Components (HRI), Annulus Components (ANF), and Pump Components (HMF). (b) Mechanistic calculation of FIJ for geometrically well-defined flow regimes and interpolation in mixed regimes. 20 (c) Use of an artificial high value for FIJ (10 N-S/H$) at void fractions below 0.001 or above 0.999, and exponential interpolation to provide transition to this value from two-phase regimes. (d) Averaging over adjacent junctions by volume and length-weighting. (e) Arithmetic averaging between old and new time values. Due to the spatial and temporal averaging scheme employed, and due to the artificial high value of 10 20 N-r/F6 used at the single phase transition, this method generally overestimates FIJ. It is believed that this would lead to underestimation of core liquid inventory in PWR SBLOCA calculations. The modifications to the drag models which have been performed for the current investigation contain the following features: (a) For all regimes, Fyy is calculated without spatial or temporal averaging. (b) For Horizontal Components (HRT) and Pump Components (HHF), the high value of 1020 N-S/F8 at the single phase limit has been reduced to 105 N-S/td. Otherwise, the previous models are still in use. (c) The Vertical Components (VRT) and Annulus Components (ANF) are treated , similarly. For these components, there are no artificial limits at the single phase transition. The limiting values are calculated by natural extension of the two-phase models. Also, the two phase models have been modified by the addition of flow-dependent regime transitions and a mechanistic treatment of entrainment at high void fractions. The modified code (Version G03) was assessed against two separate effects tests (FRIGG Loop Tests and GE Level Swell Test 1004-3). The results showed improved calculation of transient void distributions, especially at low flows when interphase drag has maximum impact. Figures A-1 and A-2 show a l

comparison of the transient void fraction at two vessel elevations for the GE level swell test. It can be seen that the new code version (G03) provides a better calculation of the level swell and subsequent drainage than the previous code version (18G). Additional assessments were performed against two integral tests (Semiscale Small Break Test S-LH-1 and LOFT Intermediate Break Test L5-1). The results of the LOFT LS-1 calculations are discussed in Attachment B. The results from the Semiscale simulation showed an improved calculation of transient void distributions and loop scel clearing effects. However, beyond the time of break uncovery, the system was calculated to depressurize more rapidly than observed in the test (Figure A-3). This caused early accumulator injection and core refill. Thus, core heatup was not calculated. These results are very similar to RELAP5/ MOD 2 calculations of this test (Reference (A.3)). The calculation of Semiscale Test S-LH-1 with the new RELAPSYA version did indicate that accumulator water was able to penetrate the downcomer and refill the core, as observed in the test data. However, the early accumulator injection due to faster depressurization prevented core heatup during the boil off period. The Semiscale input deck used for the RELAP5YA simulation did not contain any of the passive metal in the system. These are currently being incorporated into the input model. This should provide an additional heat source to the fluid and decrease the depressurization rate. It renains to be seen whether the calculated system pressure response can be improved to be closer to the data and if the thermal-hydraulic behavior of the core will then be properly calculated. An additional calculation was made for the Semiscale test (without parsive metal) by varying the break discharge coefficient. The results were similar to the previous calculation up to the time of break uncovery. Beyond this time, the calculation again showed a faster depressurization than observed in the test, but also failed to compute shortly thereafter. This is currently being investigated. REFERENCES A.1 R. T. Fernandez, et al., "RELAP5YA - A Computer Program for LWR System Thermal Hydraulic Analysis, " YAEC-1300 P, January 1983 A.2 V. H. Ransom, et al. , "RELAP5/ MOD 1 Code Manual," NUREG/CR-1826, March 1982 A.3 G. G. Loomis and J. E. Streit, "Results of Semiscale Mod-2C Small Break (5%) LOCA Experiments S-LH-1 and S-LH-2," NUREG/CR-4438, November 1985

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ATTACHMENT B RELAP5YA SBLOCA Modeling Improvements and Assessment B.1 MODIFICATIONS TO DOWNCOMER NODALIZATION In RELAP5YA simulations of Semiscale Test S-LH-1 and LOFT Test L5-1, the calculations showed that most of the injected ECC water was distributed in the loops instead of penetrating the downcomer and quenching the heated core. This was contrary to the observed phenomena in these tests. In order to improve the code calculations in the post-ECC injection period, the downcomer region was nodalized with two parallel downcomer channels connected by n cross-flow junction. The donored void fraction at the cross-flow junction was controlled by a correlation that was developed to predict ECC penetration behavior ( Reference B.1). This method was assessed against LOFT Test L5-1 and a plant calculation. This methodology and the preliminary assessment against LOFT Test LS-1 was presented to the NRC staff in the meeting on February 5,1987. The approach showed improved ECC penetration behavior in the plant application, but did not improve the LOFT L5-1 calculation. We believe that the difference is mainly due to the particular configuration of the broken loop in the LOFT facility. The approach was not assessed against Semiscale Test S-LH-1. The work on downcomer nodalization was not continued further, because of the more encouraging results obtained with modifications to the interphase drag models described in Attachment A. B.2 MODIFICATIONS TO ACCUMULATOR MODEL In the integral tests simulated with RELAP5YA, the accumulator flow was calculated to be intermittent, contrary to the measured accumulator flow in the experiments. To improve the calculation of accumulator flow, two changes in the accumulator modeling were pursued - a) improvement of the numerical solution scheme in the RELAP5YA accumulator model, and b) improvements in nodalization. The improvements in the numerical solution method consisted of a more implicit treatment of the momentum flux in the accumulator component model of RELAP5YA. This change was reported by EG&G to provide smoother accumulator response. The code updates to modify the accumulator model were obtained from EG&G (Reference D.2). The improved accumulator calculation method was tested against Semiscale Test S-LH-1 and LFOT Test LS-1. In both simulations, the accumulator flow exhibited the same oscillatory behavior seen in the previous RELAP5YA calculations. The nodalizatica improvements consisted of a better representation of the accumulator surge line and associated loss coefficients, and a better i w 

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rspraxntatica cf tha exit etendpipe within th2 accumuistar va sal. Additionally, the surge line was modeled separately using standard - RELAP5YA components instead of lumping it within the accumulator component. This allows a more. implicit treatment of the momentum flux in the surge line, and hence, should provide a more accurate calculation of-the accumulator pressure. The information on the surge line geometry *and dimensions ~of the accumulator vessel were obtained from EG&G. (Reference B.3). The modified nodalization was tested against LOFT Test LS-1. . An improved calculation of accumulator response was obtained, as shown in Figure D 1. Further details of the LOFT LS-1' simulation are presented in the next section. 

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B.3 LOFT TEST L5-1 ASSESSMENT A preliminary assessment of RELAP5YA against LOFT Test L5-1 was presented to the NRC staff meeting on February 5, 1987. The calculated results generally showed inability of ECC water to penetrate the downconer and refill the core,. contrary to the experiments results. Two additional LOFT L5-1 calculations have since been performed and are discussed in this section. The first calculation used the same RELAP5YA version as used previously (Version 18G), but the LOFT input model was revised. The versions consisted of a) improved accumulator nodalization as discussed in Section B.2, b) improved representation of fluid inventory in the steam generator secondary side (Reference B.4), and c) realistic accumulator water temperature (previous calculations used water. temperature of 2000F to be consistent with Response II.2 of Reference B.5). The calculated system pressure response is shown in Figure B-2, and agrees well with the data. The heated core was calculated to quench, as shown in Figure B-3. However, the accumulator also injected about 400Kg more water than reported in the test data. Also, in spite of the additional accumulator injection, the system mass inventory did not recover until about 300 seconds'into the transient (Figure B-4). This indicates that in spite of improved input modeling, the calculation continued to distribute more water in the loops rather than into the vessel, as shown by the experimental data. The second calculation used the modified interphase drag model (RELAPSYA Version G03), described in Attachment A. The input model was sim7.lar to that used in the first calculation, except that the accumulator water temperature was reset to 2000F and the amount of accumulator water injection was limited to the amount reported in the test data. The system pressure response is compared in Figure B-5. The 3 calculation shows a faster depressurization rate beyond 80 seconds. This caused early injection of the accumulator. However, core heatup was also calculated to begin earlier than observed in the test. Figure B-6 shows that the calculated rod temperature at this elevation exceeded the measured temperature. Superheating of the steam leaving the core was also calculated. Figure B-7 shows that the amount of water injected by the accumulator in the calculation was limited to that in the test. 3

Figure B-8 shows that the calculated primary system mass is lower than reported in the test data. However, this was apparently adequate to quench the heated core. Overall, the calculation shows that'the revised interphase drag models provide a better calculation of fluid distribution in the system. The cause of faster depressurization and early core heatup in the calculation are being further investigated. REFERENCES B.1 G. P. Alb and P. L. Chambre, " Correlations for the Penetration of ECC Water in a Model of a PWR Downcomer Annulus," Nuclear Engineering and Design, 53, pp 237-248, 1979 B.2 Letter, Edna C. Johnson (INEL) to L. SCHOR (YAEC), " Transmittal of updates to RELAP5/ MOD 1/ Cycle 029," ECJ-24-87, dated March 17, 1987 B.3 Letter, S. Modro (INEL) to L. Schor (YAEC), " Accumulator "A" for L5-1," q SMM-08-87, dated March 16, 1987 B.4 Donald B. Farrel et al., " Experimental Data Raport for LOFT Intermediate Break Experiment L5-1 and Severe Core Transient Experiment j L8-2,"NUREG/CR-2398, November 1981 B.5 Letter, YAEC to USNRC, FYR 85-121, dated November 1, 1985 i i J l r I e I I. 4 i I l

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