BWR transient analysis payoffs from LOCA research

Nuclear Engineering and Design 122 (1990) 349-355 North-Holland

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BWR transient analysis payoffs from LOCA research G.E. Dix a n d B.S. Shiralkar General Electric Nuclear Energy, 175 Curtner Avenue, San Jose, California 95125, USA

The extensive reactor safety programs directed toward Loss of Coolant Accident (LOCA) responses have provided both advanced technology and successful programmatic approaches that have significant applicability to current reactor transient studies. Extension of the benchmark TRAC-BWR computer program is proving to be highly successful for the full range of BWR transient phenomena. Work in progress will extend this technology to accurate Engineering Analyzers and Training Simulators that operate on low cost desktop workstations. The payoffs of these developments for improved reactor design, analysis and operation are summarized.

Introduction Research on Loss of Coolant Accidents (LOCA) dominated nuclear reactor safety studies for a long period of time. Tong [1] estimated that more than $500000000 has been spent to solve the LOCA problem. However, following assessments of the TMI-2 accident, worldwide research efforts moved toward precursors for severe accidents, including more probable transient events, and operator guidance. It is important in this transition that we do not lose the valuable results and lessons-learned from past LOCA studies. These include directly applicable technology as well as technical and programmatic approaches to address such complex problems. The costs for future technology improvements must be reduced since other $500 000000 solutions cannot be supported, This paper summarizes important lessons-learned from the Boiling Water Reactor (BWR) LOCA technology program, and the important payoffs achieved by extending this technology to BWR transient analyses. The significant impacts of microcomputer advances are also highlighted.

2. LOCA research background The general LOCA research program extended over a long period, with significant expenditures starting about 1965. In the early days, LOCA models were formulated with very simplified assumptions to allow for bounding design analyses within the existing knowledge of controlling phenomena and computer system

constraints at that time. Supporting research generally focused on the phenomena that were controlling in those bounding models. This resulted in islands of information that often could not be consistently integrated. The lack of consistency frequently resulted in additional conservatism being introduced into the design and licensing of nuclear reactors, as new research results became available. By 1975, General Electric identified that a more fundamental, integrated approach to LOCA research was appropriate for the BWR. A long range plan to develop realistic LOCA analysis capability was initiated which ultimately involved regulatory agencies, utility research organizations and BWR system vendors worldwide. The process started by identifying all potentially controlling phenomena and assessing the state of understanding for each. Appropriate experiments were then defined to expand those areas where existing knowledge was inadequate. The TRAC computer program, Pryor et al. [2], was adopted for this realistic analyses development because of the extensive modeling detail and robust numerical procedures. Separate BWR versions of TRAC were developed from that basis. However, to avoid the prior problems of directing research on the basis of perceived uncertainties in a specific model, the developments were planned to allow the supporting experiments to include all controlling phenomena to the degree practical. This required both small scale separate-effects experiments and large scale simulation experiments where the controlling phenomena produced system responses representative of fullscale reactors. The costs for these realistic experiments required integrated support by BWR vendors as well as

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utility and regulatory research groups worldwide. The synergy between parallel development of very detailed benchmark models and large scale simulation experiments resulted in rapid technology advances and a thorough understanding of BWR responses under LOCA conditions. The primary experimental and modeling results for LOCA applications have been reported previously. Dix [3], Tasaka et al. [4], Myers [5], and Sutherland et al. [6] summarize key experimental resuits, while Shiralkar et al. [7], Yahagi [8], Alamgir [9] and Spore et al., [10] identify the modeling advancements. The continuing payoffs of these prior developments for realistic transient analyses are summarized in this paper.

3. Best estimate LOCA capabilities To properly characterize the extensions for transients, it is useful to briefly describe the analysis approach and capabilities for BWR LOCA conditions. Key components of a BWR steam supply system are illustrated schematically in fig. 1. In normal operation, approximately 15% of the flow is vaporized as it passes

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through parallel enclosed fuel channels. The remaining 85% of the flow is pumped back to the inlet plenum to be circulated, with returning feedwater from the turbines, through the fuel channels again. For most possible LOCA conditions, the forced circulation pumps would be tripped off, and the control blades rapidly inserted. Therefore, realistic LOCA predictions require accurate methods for rapid flow and power transients. BWR LOCA events would include a rapid depressurization period (due either to large breaks, or to automatic depressurization valves that open for small breaks), followed by a period of liquid inventory depletion until the safety systems refill the vessel. Several different flow regimes, including co-current and counter-current fows of vapor and liquid, would occur in different fuel bundles during these periods of LOCA transients. The elevations of two phase levels in the fuel bundles are very important during these transients. Below the twophase level, liquid is the predominant continuous phase; whereas, above the two-phase level, vapor is the predominant continuous phase. Two-phase levels would drop slowly in the bypass region between bundles and within the counter-current flow bundles. Subcooled emergency cooling liquid would be injected at various locations, producing significant gradients in local subcooling and condensation effects. These complex flow and distribution phenomena impose requirements that benchmark BWR LOCA modeling include detailed three-dimensional and multichannel features, with a full range of accurate two-phase flow regime and level tracking capabilities. The success of TRAC-BWR to accurately predict these phenomena has been demonstrated in the data comparisons reported by Myers [5], Sutherland et al. [6], and Alamgir [81. The 30 ° Steam Sector Test Facility (SSTF) provided some of the best BWR LOCA response data (Myers [5]), These tests simulated the response in full scale of a BWR during the refill/reflood portion of the LOCA transient. SSTF had 58 channels representative of the various power levels and radial locations in a BWR. Tests showed that after initiation of ECCS in the upper plenum, the peripheral channels experienced counter current flow limitation (CCFL) breakdown, with large amounts of liquid drainage. Most of the channels were in the counter-current flow regime with two-phase levels. High power channels exhibited co-current flow as steam from the lower plenum and leakage flow from the bypass flowed upward through those bundles. The primary data from this facility were limited to the steady state end conditions. One clear indication of the bundle flow regimes was provided by comparing pres-

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sure differential measurements (density head and flow losses) across the upper portion of the bundles (AP 1 in fig. 2) with the pressure differential across the orificed bundle inlet region (AP 2 in fig. 2). Under co-current upflow conditions, the measured pressure differentials in both regions were large and in the- same direction. For counter-current flow conditions, the smaller and reversed flow losses at the inlet region significantly reduced that pressure differential. For complete downflow condition, the significant flow losses through the orificed inlet region actually reversed the direction of the overall pressure differential for that region. These pressure differentials are illustrated for the three flow conditions by the associated steady state data measurement bands in fig. 2. A severe evaluation of TRAC-BWR capability to predict the flow conditions in these SSTF tests was

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made by starting the TRAC-BWR predictions from an initially static condition. This technique forced TRACBWR to solve a complex transient to establish the final steady state flow conditions for all bundles (and assured that the predictions were not favorably biased by selected initial conditions). The resultant predictions in fig. 2 indicate the complex flow regimes predicted during this arbitrary transient and, more importantly, indicate the ability of TRAC-BWR to achieve the same steady state flow conditions as measured. These different flow conditions were predicted for correct radial locations as well. The Full Integral System Test (FIST) facility (Sutherland et al. [6]) provides additional challenging data for model assessment. This facility included a complete BWR system scaled to one full size fuel bundle. Figure 3 illustrates typical measured temperature transients for different fuel rod simulators at the bundle mid-plane from this FIST facility. The TRAC-BWR hot-rod model predictions are seen to match the highest measured temperature, while the predicted planar average temperature reasonably represents an average rod performance. Separate comparisons confirmed similar accuracies for associated predictions of system flows and inventories, two-phase level in the bundle and heat transfer above the mixture level. With the availability of the TRAC detailed benchmark model and supporting large scale simulation experiments, appropriate simplifications were defined to provide a fast running design model. General Electric led an effort to develop such a model, called SAFER (Shum et al. [11]), that provides realistic BWR-LOCA predictions with some conservatism, as illustrated by comparison with best-estimate TRAC-BWR predictions in fig. 4 for a large break in the recirculation line. The simplifications in this model were carefully selected to

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4. E x t e n s i o n s

for transient capabilities

A primary difference between LOCA analysis and other BWR transients is neutronic feedback during transients. Transients can include rapid flow decays (e.g., pump trip) and power peaks (e.g., turbine trip without bypass), slow changes in overall conditions (e. g., feedwater heater failure or ATWS), or flow and power oscillations (instabilities). Instabilities can include both core-wide or regional behavior. Core-wide instability is the usual mode, where neutronics feedback forces the flow and power oscillations for all fuel bundles to be in-phase. However, it is possible when local regions are sufficiently less stable to create higher harmonic modes, with regions oscillating out of phase (balancing flow continuity from the inlet plenum). Field experience confirms that the least subcritical firstharmonic (i.e., half-core) is the most likely mode for these regional oscillations. The latter transients are very demanding for analysis, since both detailed kinetics nodalization (one-bundle radial node size), many thermal hydraulic groups (10 to 20) and many axial nodes are required for accurate predictions. Comparing the full range of BWR transient analysis requirements with the benchmark modeling features

included for LOCA indicated that relatively few additions were necessary. The vessel component models and the two-phase flow and heat transfer modeling were sufficient without further development or experiments. Implementation of improved neutronics and balanceof-plant (BOP) models were primary requirements to give TRAC-BWR capability for transients. Using the successful benchmark strategy of the LOCA development, a full 3D neutronics model was implemented into TRAC-BWR. This model was kept consistent with a thoroughly verified steady state BWR core simulator model (Crawford et al. [13]) so the extensive reactor field data could be directly utilized. The existing BOP models in TRAC-BWR were found to have sufficient generality that only minor extensions and optimizations were required to match available data. The primary experimental support requirements are for confirmation of the coupled neutronics models with available reactor transient data. From the LOCA development lessons-learned, the first priority for transients has been to complete and verify accurate benchmark modeling capability. The modeling additions are now effectively completed. Reactor transient data comparisons are still in progress. Figure 5 illustrates the type of excellent predictions being achieved (Andersen and Shaug [14]). In this case TRAC-BWR accurately calculated the steam line dynamics and pressure wave propagation to predict the resultant power peak for this turbine trip test. As previously discussed, BWR transient analyses cover a broad range, including low probability events (e.g., ATWS, Stability and Seismic) and also more frequent flow, power and pressure transients. Applications range from Benchmark analysis to Engineering Analyzer and Training Simulators. Each of these application areas has different simulation requirements, as summarized in table 1.

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G.E. Dix, B.S. Shiralkar / BWR transient analysis Table 1 Simulation requirements Benchmark analysis

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Event sequences match Alarm triggers match

The approach for LOCA analysis of developing a detailed benchmark model and a separate simplified model was effective to provide the different accuracy and computation speeds required. However, it was resource intensive to develop and maintain completely separate models. Therefore, the approach used for BWR transients has been to provide alternative detailed and simplified analysis options within the same TRAC-BWR model. For example, a compatible 1-D neutronics model has been included as an analysis option. With accurate benchmark transient analysis capabilities available, the next steps will be to provide simplified modeling options and coding revisions that increase computation speeds while still maintaining acceptable accuracies. The first goal is to provide Engineering Analyzer capabilities for use in operations support, Emergency Procedure Guideline evaluations, and post-event analysis at reactor sites. Transient analysis times a factor of five longer than real-time are considered acceptable for these applications. Subsequently, TRAC-BWR capabilities for direct application in reactor Training Simulators are planned. This will require computation times approximately one-half of real-time. As discussed below, progress in software and rapid improvements in computer performance should allow these integrated transient analysis goals to be achieved with TRAC-BWR in the near future.

5. Transient applications with TRAC-BWR The availability of benchmark TRAC-BWR transient capabilities has allowed for many unique system analyses. A few examples of interest are summarized here.

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5.1. Seismic effects analysis A seismic event recently occurred at a reactor site with two operating BWRs. During the event one of the reactors experienced a scram due to an apparent power increase indication from an average power range monitor (APRM). However, the second reactor did not experience a scram. It was not certain whether the seismic oscillations had caused sufficient void and power oscillations to produce a real power trip or not (i.e., which plant had experienced an erroneous instrument signal). The identified mechanism for void and power oscillations was the varying vertical acceleration associated with the seismic event. Hence, the flexibility to input a time varying gravitational field into the TRAC-BWR model, allowed Shaug et al. [15] to directly analyze the effects of the site frequency spectrum. As indicated in fig. 6, the calculated power peaking during the seismic event was well below scram levels. The response difference between the two reactors was therefore concluded to have been caused by a spurious signal in the plant that scrammed. 5.2. Anticipated transient without scram (A T W S ) For an ATWS to become a severe accident, both backup rod insertion and borated solution injection must fail. For limiting plants with small residual heat removal capacity, Emergency Procedure Guidelines would require operator controlled depressurization and actuation of a low pressure core spray system. Such ATWS analysis therefore requires calculation of complex thermal hydraulic regimes, with power generation during depressurization and reflooding. Figure 7 shows typical TRAC-BWR predictions of Andersen et al. [16]

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for reactor power during the first 1000 s of such a degraded transient. It was found that long-term core power reduced to appproximately 3.5% of rated, which is within the capacity of the Residual Heat Removal System for many plants. Moreover, the peak cladding temperatures were low enough to preclude fuel damage. These findings show this event sequence, previously identified by IDCORE [17] as a risk dominant sequence for BWRs, could have much less severe consequences than previously thought. 5.3. Advanced reactor design

GE Nuclear Energy is actively engaged in the design of advanced BWR systems which rely on inherent passive safety. TRAC-BWR is being used to help optimize the configuration of these reactor designs. In addition to accurate modeling, modularity is a key feature of TRAC-BWR that is critical to this design process. With flexibility to represent a wide variety of components and configurations using basic component models, TRAC-BWR is ideally suited to explore various design concepts. Gravity driven emergency cooling systems (GDCS) analysis requires an integral model, consisting of reactor vessel and containment, to realistically predict the coupling of vessel and suppression pool conditions with the emergency coolant flow. Figure 8 illustrates typical calculated pressure responses in the primary vessel and containment wet-well air space for a bottom drain line break. The differential between these pressures controls initiation of GDCS. TRAC-BWR analyses have been used to optimize vessel inventory and depressurization rate to prevent core uncovery for the smallest downcomer volume in this advanced design. Future test programs will further verify these TRAC-BWR analyses; however, the extensive background validations provide very high confidence in the

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6. Microcomputers for reactor transients

The rapid improvements occurring in microcomputer performance will have a significant impact on the costs and availability of realistic reactor transient computations. Desktop engineering work stations are now being utilized for design studies (Dix and Congdon [18]), including use-of TRAC-BWR. As processor speeds increase, workstation systems will be able to provide the Engineering Analyzer and Training Simulator capabilities defined in table 1. Preliminary evaluations indicate that a TRAC-BWR based Engineering Analyzer would require processor capability of approximately 3 million floating point operations per second (MFLOPS). With further simplifications in modeling detail, a TRACBWR based Training Simulator would require processor capability of 5 to 10 MFLOPS. Recent processor tests by Dongerra [19] indicate the limits of current worksta-

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G.E. Dix, B.S. Shiralkar / B W R transient analysis tion capabilities at a b o u t 1 M F L O P S . A s s u m i n g a reasonable microprocessor d e v e l o p m e n t trend of nearly d o u b l i n g processor speed each year, T R A C - B W R based Engineering Analyzers a n d T r a i n i n g Simulators could be practical using desktop engineering workstations by the 1990-92 period. Availability of such low cost a n d p o r t a b l e systems will provide new opportunities for i m p r o v e d local training, p l a n t event evaluations a n d on-line system monitoring.

7. Conclusions The extensive B W R L O C A research p r o g r a m has provided m u c h more t h a n the L O C A specific technology that was its focus. The experimental base a n d detailed b e n c h m a r k models are f o u n d a t i o n s for extensions to the full range of B W R transients. The programmatic lessons-learned from the L O C A studies are also i m p o r t a n t to the success of subsequent transient developments. In particular, focusing first on a c h i e v e m e n t of detailed b e n c h m a r k prediction capability can improve the probability of success a n d reduce the overall resources required. It is far more reliable to specify appropriate modeling simplifications o n the basis of large scale experiments a n d detailed b e n c h - m a r k model analysis rather t h a n on engineering j u d g e m e n t alone. T h e final payoffs of this work will come in the near future. A d v a n c i n g m i c r o c o m p u t e r capabilities will allow these detailed a n d accurate models to be used with Engineering Analyzers a n d T r a i n i n g Simulators based o n low cost engineering workstations. W i d e s p r e a d availability of such tools should greatly improve general u n d e r s t a n d i n g a n d operation of nuclear plants.

References [1] L.S. Tong, Thermal hydraulics for reactor design and safety, Acceptance Presentation for Technical Achievement Award from ANS Thermal Hydraulics Division, Washington DC, (1986). [2] R.J. Pryor et al., TRAC-P1A. an advanced best estimate computer program for PWR LOCA analysis, Los Alamos Scientific Laboratory, May 1979 (NUREG/CRA-O665, LA-777-7S). [3] G.E. Dix, BWR.loss-of-coolant technology review, Proc. 2nd Int. Topical Meeting on Nuclear Thermal Hydraulics, Santa Barbara, California (1983). [4] K. Tasaka, et al., ROSA-III experimental program for BWR LOCA/ECCS integral simulation tests, (Final Report), JAERI 1307 (1987).

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[5] L.L. Myers, BWR Refill-Reflood Program final report, N U R E G / C R - 3 2 2 3 , E P R I / N P - 3 0 9 3 , GEAP-30157 (1983). [6] W.A. Sutherland, Md. Alamgir, J.A. Findlay and W.S. Hwang, BWR FIST phase II test results and TRAC-BWR model qualification, NUREG/CR-4128, EPRI NP-3988, GEAP-30876 (1985). [7] B.S. Shiralkar, J.G.M. Andersen, A.B. Burgess and S.A. Wilson, Evolution of LOCA analysis at General Electric, Proc. Int. Nuclear Power Plant Thermal Hydraulics and Operations Topical Meeting, Taipei (1984). [8] K. Yahagi, Improvements of BWR LOCA/ECCS analysis in Japan, Trans. of the Thirteenth Water Reactor Safety Information Meeting, Gaithersburg, Md. (1985). [9] Md. Alamgir, BWR Refill/Reflood Program task 4.8 TRAC-BWR model qualification for BWR safety analysis - final report, NUREG/CR-2571 (1983). [10] J.W. Spore et al., TRAC-BDI: an advanced best-estimate computer program for boiling water reactor loss-of-coolant accident analysis ( N U R E G / C R - 2 1 7 8 , EGG-2109, Volumes 1-4) October 1981. [11] F.D. Shum, S. Itoya, K. Tominaga, K. Tasaka, W.S. Hwang, H. Aoki and B.S. Shiralkar, An improved model for BWR LOCA analysis, Second Prec. of Nuclear Thermal Hydraulics. ANS., New Orleans, LA (1984). [12] B.S. Shiralkar and J.G.M. Andersen, Best estimate LOCA analysis applications and benefits, Invited Paper, Topical Meeting on Anticipated and Abnormal Transients in Nuclear Power Plants, Atlanta, GA (1987). [13] B.W. Crawford, et al., Use of operating plant data for validation of the General Electric three-dimensional BWR simulator, Trans. Amer. Nucl. Sec. 44 (1983). [14] J.G.M. Andersen, J.C. Shaug and B.S. Shiralkar, TRAC development at General Electric, Prec. Fourteenth Water Reactor Safety Information Meeting, Gaithersburg, MD, NUREG/CP-0082, Vol. 5 (1986). 115] J.C. Shaug, J.T. Tatsumi, J.G.M. Andersen and B.S. Shiralkar, Analysis of fluid dynamics in a boiling water reactor in a varying acceleration field, Forum on Unsteady Fluid Flow, ASME Winter Annual Meeting, Boston, MA (1987). [16] J.G.M. Andersen, L.B. Claassen, S.S. Dua and J.K. Garrett, Analysis of anticipated transients without scram in severe BWR accidents, EPRI NP-5562 (1987). [17] Technology for Energy Corporation, Risk reduction potential, IDCORE Technical Report 21.1 (1984). [18] G.E. Dix and S.P. Congdon, Nuclear interactive evaluations on distributed processors, Int. Reactor Physics Conf., Jackson Hole, WY (1988). [19] J.J. Dongerra, Performance of various computers using standard linear equation software in a Fortran environment, Argonne National Laboratory, Technical Memorandum No. 23 (1987).