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BWR Methods assessment and applications
Nuclear Engineering and Design 83 (1984) 219-223 North-Holland, Amsterdam
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9. B W R M E T H O D S A S S E S S M E N T A N D A P P L I C A T I O N S G.E. W I L S O N E G & G Idaho, Inc., Idaho Falls, ID 83415, USA
9.1. Introduction Following the accident at Three Mile Island Unit-2 (TMI-2), it became obvious that the ability to analyze abnormal and anticipated plant transients needed to be improved. A vital part in the improvement process is the assessment of new analysis methods and their applications. Both topics, assessment and applications, as related to boiling water reactors (BWRs), are discussed in this section. At the ANS Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors at Jackson, Wyoming [9-1], 12 papers addressed BWR methods assessment and applications. Specific subjects covered in the meeting presentations included: US Nuclear Regulatory Commission (NRC) research [9-2], plant transients data bases [9-3], methods assessment with subscale experimental data [9-4], methods assessment via applications to plant data [9-5-7], methods assessment via comparison of two codes [9-6, -8], and methods application to specific issues [9-9-13]. Paper contributors included the NRC, the utilities, the Electric Power Research Institute (EPRI), the national laboratories, the reactor vendor, and various foreign organizations. In this reviewer's opinion, the subjects covered at the Jackson conference were sufficiently balanced to give an overview of the current state-of-the-art of BWR methods assessment and applications, with one exception; that is, thermal-hydraulic analyses of anticipated transients without scram (ATWS) were not reported. ATWS is included in the overview of BWR methods assessment and applications presented in the following discussion. The overall objectives of the improvements in operational transient analytical methods (and thereby research, assessment, and applications) were well summarized by Beckner et al. [9-2] as follows: (1) Transient analyses are generally required to be best estimate calculations, rather than bounding
or conservative calculations. Uncertainties in modeling phenomena or plant boundary conditions cannot, therefore, be dispensed with simply by making conservative assumptions. (2) Operator actions and balance of plant systems, including nonsafety systems, play a very important role in transients and cannot be ignored or treated in a simplistic manner as is often done in design basis accident (DBA) analyses. (3) Different thermal-hydraulic phenomena occur during these transients. Low flow rates or nearly stagnant conditions result in nonequilibrium and separated flow regimes (slug, stratified, countercurrents, etc.) which are more difficult to model than the primary homogeneous flow characteristics of large-break loss-of-coolant accidents (LOCAs). (4) Analyses of ATWS require more detailed neutronics models and more accurate thermalhydraulic conditions to properly calculate reactivity feedback. (5) The long duration of these transients require more efficient running computer codes to make lengthy calculations practical. Beckner et al. indicated the Two Loop Test Apparatus (TLTA) and Full Integral Simulation Test (FIST) subscale experimental facilities have or are contributing data for the assessment of the BWR Transient Reactor Analysis Code (TRAC), R A M O N A , and the nuclear plant analyzer.
9.2. BWR methods assessment studies The need for an assessment data bank containing both experimental and plant transient data is well recognized. One such bank is being developed by the utilities' research organization, as described by Bailey et al. [9-3]. This data bank uses D A T A T R A N as the data base management system. Through D A T A -
0 0 2 9 - 5 4 9 3 / 8 4 / $ 0 3 . 0 0 © Elsevier Science Publishers B.V. ( N o r t h - H o l l a n d Physics Publishing Division)
220
G . E . Wilson / 9. B W R methods assessment and applications
TRAN, the data from various sources can be uniformly structured, stored in a central repository, and later retrieved. Plant transient data are being archived as they become available from the utilities. The bank currently contains the BWR Peach Bottom 2 turbine trip tests. Future incorporation of Grand Gulf BWR power ascension data is planned. This reviewer notes that the objectives of the EPRI data bank are similar to those of the existing NRC experimental data bank, and exchange of data between these two systems should be considered. The elements in a successful code assessment study using subscale experimental data were typified by Alamgir and Sutherland [9-4]. The data base for this study came from a small-break LOCA test in the FIST facility. FIST is a single-bundle integral system capable of full-power steady state operation and realtime simulation of transients. The test simulated a 0./)5 ft 2 break in the suction side of one recirculation pump, coupled with failure of the high-pressure core spray (HPCS) system. The assessment study was a pretest prediction of the subject test. Alamgir and Sutherland indicated that success in such studies is a direct function of the understanding of, and attention paid to probable system behavior in the system definition modeling. The general quality of this assessment (and the other papers reviewed in this section) is exemplified by the midplane temperature comparisons shown in fig. 9-1. From these, and other assessment results, the authors conclude the BWR T R A C code captured the controlling phenomena very well. They also noted, that because vessel stored heat in a BWR is significantly less than in FIST, the refill/reflood process in a BWR will be more effective than that shown in the FIST facility. This reviewer notes that this may not be true for all subscale facili1000 I / 800I-
I t -- - - -- Average measured midplane temperature .... Minimax measured midplane temperature BWR T R A C c a l c u l a t i o n
800-
I ~
/~ /
J/
P" 4 0 C -
2C
0
I
'
- -
"~
"
I 100
I 200
I 300
I 400
500
Time (s) Fig. 9 - t . C o m p a r i s o n of B W R T R A C
pretest prediction with
data from a BWR FIST small-break test without HPCS.
ties for transients where the ambient heat transfer is proportionally higher than a BWR. Several different approaches to the analysis of transients were covered by Yokobayashi and Takahashi [9-5], House et al. [9-6], and Alamgir and Sutherland [9-7]. In all cases, the codes were assessed by application to full-scale BWRs. BWRDYN [9-5] is a "components" code. Components in this context denotes a code in which the calculational algorithms are structured to describe large regions (such as a pressure vessel) or systems (such as a feedwater control). In the case of BWRDYN, the resulting thermodynamic and control system equations are integrated by the Euler method, and those describing the steam line flow dynamics by a fourth-order Runge-Kutta technique. Components codes have the distinct advantage of being fast running. For example BWRDYN has a stated ratio of calculational time to transient time in the order of 1/12. This feature is a valuable code attribute, particularly useful to scopingtype studies. Components codes are generally less useful in situations where detailed local behavior is necessary or where the facility or transient phenomena being simulated differs singificantly from that assumed during development of the code. BWRDYN was assessed with data from two transients in the T O K A I - 2 BWR: 20 cm change in water level setpoint and turbine trip. For the setpoint change transient, the reported results show good code/data agreement for feedwater flow, water level, neutron flux, and recirculation flow. In the turbine trip comparisons, steam dome pressure and recirculation flow were well calculated. The water level comparisons early in the transient were not as good as those of the setpoint-change case. From these results, Yokobayashi and Takahashi concluded the code well simulates plant behavior. Assessment of R E L A P 5 / M O D I , Cycle 14 with Peach Bottom turbine trip Tests T T I , TT2, and TT3 was reported by House et al. [9-6]. In contrast to BWRDYN, RELAP5 is a one-dimensional, twophase, nonequilibrium, building block code. That is, any system can be modeled to whatever degree of detail is desired using volumes, junctions, and heat slabs. This code can also simulate control systems, again with a building block approach. The code version used in this study required updates to model the jet pumps and to improve steam separator performance. This study served a dual assessment function in that, in addition to the RELAP5/plant data comparisons, results from simulations with the R E T R A N code were also given. In one calculation, the
G.E. Wilson / 9. B W R methods assessment and applications
R E L A P 5 and R E T R A N models were as nearly identical as possible. In three others, the RELAP5 model was modified to an optimum configuration to account for differences in the codes. From the RELAP5/RETRAN comparisons, the authors conclude both codes give similar calculational and run time results. The calculational results compare well with the plant data. Other conclusions related to desired RELAP5 modeling and initialization improvements, and to code-to-code modeling conversions were reported. Run times were not reported. However, based on this reviewer's experience and his perceptions as to the level of detail contained in the models, he would estimate the computer time was in excess of the transient time by at least one order of magnitude. This perception can be contrasted to similar information for BWRDYN, already discussed, and BWR T R A C , to follow. However, some caution is warranted in such a comparison because it is unlikely the RELAP5 and BWR T R A C models were optimized for run time. A full-scale assessment of the reactor vendor's version of BWR T R A C with Peach Bottom turbine trip Test TT1 was presented by Alamgir and Sutherland [9-7]. BWR T R A C is a building block code similar to RELAP5, except the code is multidimensional and is BWR specific. The subject study was a nonneutronic analysis because the measured core power was input as a boundary condition. The authors stated that a two step assessment scheme (thermal-hydraulics first, then neutronics) was the preferred method of analysis. The reported study constituted the first step. Code/plant data comparisons of pressure throughout the system indicated the base case calculation captured the fundamental mode of acoustic oscillation and pressure wave propagation. The trends in the collapsed downcomer water level were also correct, and considering certain modeling limitations, the absolute values of this parameter were considered satisfactory. Two sensitivity studies were reported which give further information relative to steam line nodalization and the steam bypass valve loss factor. The ratio of computer time to transient time for the base case was 126. The authors concluded the code realistically predicts the pressure transient in the main steam line and the reactor vessel. This reviewer notes that the steam dome and turbine inlet pressures calculated by R E T R A N , RELAP5 (see previous paragraph), and BWR T R A C were quite similar. Thus, there is some evidence that for this transient, multidimensional modeling was not significantly important for adequate calculational
221
results. Further aspects of this perception are discussed in the conclusions section. The comparison of balance of plant (BOP) modeling with R E T R A N and the modular modeling system (MMS) was reported by Arndt et al. [9-8]. R E T R A N is a building block code, whereas MMS treats the various BOP components as modules (that is, tends to be more of a components code as previously defined). The modeling used in the study was based on the Fermi 2 BWR and was restricted to the BOP. The simulated transient was structured to address questions concerning feedwater pump suction pressures during a turbine trip. One important finding during the study was that both codes require substantial user time to establish the initial steady state conditions. This reviewer would expect this condition with the building block codes (RETRAN, RELAP5, BWR TRAC), but was somewhat surprised to find the same requirement for a components code. The reported results indicate that for the two types of codes the trends in heat transfer and water level in the heaters are similar, but there are differences in amplitude and phase. Reactor feed and heater feed pump suction pressures were in better agreement. The authors found that MMS is a powerful tool capable of saving significant labor and computer costs. However, MMS must be skillfully applied to provide reliable results.
9.3. BWR methods application studies The discussion in this section relates to methods application studies where code assessment is, at most, a secondary objective. The core kinetics aspects of the partial rod insertion incident at Browns Ferry (June 1980) was studied and reported by Kitayama et al. [9-9] using the T O S D Y N code. T O S D Y N was developed by the authors to model the reactor core with three-dimensional neutronics and parallel-channel thermal-hydraulics. In the study, the core was modeled assuming half-core symmetry to reduce computer costs. To the extent possible, rod positioning was programmed from the plant data. The results from the study led the authors to conclude that, even though the rod insertions were highly unbalanced, from the critical power ratio (CPR) view, the core responded homogeneously during the transient. In this reviewer's opinion, this conclusion may have significant implications in the analysis of ATWS, as discussed later. The capability to model the BOP with T R A C BD1/MOD1 was presented by Weaver et al. [9-10].
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T R A C - B D 1 / M O D 1 is a building block code for BWRs, and the user can model BOP units to any degree. However, in the interest of user convenience, component modules have been developed for such items as the turbine and feedwater heaters. The features of these modules were presented. Initial developmental assessments of the models indicate they were qualitatively correct. An alternate approach, relative to frequency domain analysis, to meet the objectives of stability analysis was presented by Hornyik and Naser [9-11]. This technique used the advanced (highly nonlinear) modeling capabilities of the R E T R A N code and has the advantage of being amiable to periodic perturbations. Periodic perturbations are easier to use than pseudorandom switching which must be measured and input as a boundary condition. In addition, frequency domain analysis has limitations for nonlinear systems where large perturbations near the point of unstability can be expected. The authors concluded the proposed technique potentially offers stability analyses which are more general. This study provided a dual product in that the R E T R A N model of the Peach Bottom BWR was benchmarked with the data from plant stability tests. Good qualitative agreement was obtained between the calculated and measured parameters. Once additional planned work has eliminated some discrepancies (such as smaller calculated neutron flux) further validation of the stability analysis method can be effected. Probabilistic risk assessment (PRA) techniques provide a means of determining the system success criteria needed to safely shut down a BWR after a transient initiating event. Often in PRAs, bounding assumptions (that is, best case/worst case) are made because of the lack of information or to expedite the initial analysis. In this way, the most probable dominant sequences can be quickly determined and the efficiency of the study can be assured. However, to confirm the validity of the assumptions underlying the dominant sequences, further mechanistic analyses are important. Moore et al. [9-12] reported such an application using the R E T R A N code. Of five candidate key sequences, two (station blackout and loss of feedwater coupled with loss of high-pressure systems) were selected for further evaluation with R E T R A N . In both cases, the R E T R A N analyses modified the conclusion of the original PRAs relative to success criteria. The authors concluded the coupling of mechanistic thermal-hydraulic analysis with the P R A methodology results in a more realistic evaluation of success criteria. Similar conclusions (to
be published in late 1983) have been formed in the ATWS Severe Accident Sequence Analysis Program being conducted at the Idaho and Oak Ridge National Engineering Laboratories. The application of RELAP5/MOD1 Cycle 13 to five similar Ioss-of-feedwater (LOF) transients in a typical BWR/4 was covered by Lu et al. [9-13]. The authors gave details of the sequence of events calculated by the code for all five cases. The core uncovered on a core average basis in only two of the transients (those assuming loss of high-pressure coolant injection coupled with loss of the reactor core isolation cooling, with and without the automatic depressurization system). These results did not preclude the possibility of localized core uncovery which would require a more detailed model to determine. The authors concluded their analysis demonstrates the applicability of the code and model to a wide range of LOF transients of interest in the licensing process.
9.4 Conclusions
The following discussion presents this reviewer's conclusions and perceptions relative to the state-ofthe-art as reflected by the Jackson conference and his own experience. Priority transients of interest (that is, dominant sequences) have or will soon be defined, on a first cut basis, by such methods as PRA and failure mode effects analysis (FMEA). Because these techniques often initially assume bounding conditions to achieve efficiency of study, further refinements in the P R A s / F M E A s are necessary to define the dominant sequences correctly. These refinements can be effected via feedback from mechanistic thermal-hydraulic and neutronic analyses. Significant progress has been made in the last 3 years in developing, assessing, and applying advanced thermal-hydraulic and neutronic analysis methods (codes). Similar progress in the development of necessary data bases has been made. The availability of subscale experimental data is good and is improving. The availability of plant data is not yet as good and further work is necessary. This is particularly true of BWR containment behavior. With two exceptions (as subsequently discussed), the papers in the Jackson conference indicate the basic needs for thermal-hydraulic and neutronic analysis methods are being met. Both fast running components codes and building block codes, which are slower but give more detailed local behavior, appear able to generally simulate plant behavior. Desirable refinements in these codes are known and
G.E. Wilson / 9. B W R methods assessment and applications
the planning for the next 2 years shows promise these refinements will be developed. However, it should be noted that a common industry standard defining the required adequacy and sufficiency of code performance does not exist. It seems unlikely code development will be terminated until such a standard is formulated. Results presented at the conference indicate that in transients, which do not exhibit significant multidimensional behavior, the general thermal-hydraulic results from all the advanced codes are quite similar. Thus, it is incumbent upon the analyst to choose carefully the most efficient analytical tool. The need for such a careful selection is expected to decrease with time because options which increase running time are currently being developed for the slower running building blocks codes. The two exceptions, mentioned previously, are associated with containment analyses and with the analyses of transients which exhibit significant coupled multidimensional neutronics and thermal-hydraulic behavior (such as A T W S and core damage transients ). Presentations on these subjects were noticeably absent in the conference. Relative to containment analysis, plant data for the development/assessment/application process, is not readily available. It is also this reviewer's perception that containment code development (prior to containment failure) has a lower priority than other methods development. Successful resolution of several of the more important BWR transient questions will require good containment analysis. Thus, this is an area which needs immediate attention. tt is this reviewer's opinion that a single code, which combines adequate multidimensional neutronics, multidimensional thermal-hydraulics (including boron injection and mixing), control system, and BOP analysis capabilities, does not now exist. Even more basic is the question of just how much detail is required in the neutronics simulation of such a code. Kitayama et al. [9-9] indicated that even though the rod insertions in the Browns Ferry A T W S were highly unbalanced, the core responded homogeneously. This may indicate that a code combining all the features just mentioned is not absolutely necessary for
223
adequate simulations of most ATWS. That is, it may be more efficient to develop the general core neutronics behavior in a separate components code (having multidimensional neutronics and simple thermalhydraulics), and then approximate this behavior with a good thermal-hydraulics code which has only onedimensional neutronics simulation. Further work is required to establish the optimum analytical methods to adequately and efficiently simulate the subject transients.
References [9-1] ANS Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors, Jackson, Wyoming, September 26-29, 1983 (Plenum Press, New York, 1984). [9-2] W.D. Beckner, F. Odar and L.H. Sullivan, PWR and BWR Anticipated and Abnormal Plant Transient Research Sponsored by the US NRC, ibid. [9-3] P.G. Bailey, G.A. Cordes and R.K. House, The EPRI Plant Transient Data Base, ibid. [9-4] M. Alamgir and W.A. Sutherland, FIST Small Break Accident Analysis with BWR TRAC, ibid. [9-5] M. Yokobayashi and Y. Takahashi, Verification Study of Transient Analysis Code BWRDYN Using Startup Test Data of Tokai Unit 2, ibid. [9-6] R.K. House et al., Application of RELAP5 Code for Simulation of Three Turbine Trip Transients at the Peach Bottom Unit 2 BWR, ibid. [9-7] M. Alamgir and W.A. Sutherland, Peach Bottom Transient Analysis with BWR TRAC, ibid. [9-8] C.R. Arndt, E,M. Page and T.L. Tederington, Balance of Plant Modeling with RETRAN and MMS, ibid. [9-9] K. Kitayama, S. Tsunoyama and S. Ebata, Analysis of Incomplete Control Rod Insertion by Three-Dimensional Kinetic Code Developed for BWR Simulation, ibid. [9-10] W.L. Weaver, M.M. Giles and C.M. Mohr, Balance of Plant Modeling in TRAC-BD1/MOD1, ibid. [9-11] K. Hornyik and J.A. Naser, Stability Margin of BWRs Operating with High Power/Flow Ratios, ibid. [9-12] E.V. Moore et al., Use of RETRAN for Determination of Some BWR Success Criteria, ibid. [9-13] M.S. Lu, M.M. Levine and W.G. Shier, BWR/4 Loss of Feedwater Transients Analysis, ibid.
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9. B W R M E T H O D S A S S E S S M E N T A N D A P P L I C A T I O N S G.E. W I L S O N E G & G Idaho, Inc., Idaho Falls, ID 83415, USA
9.1. Introduction Following the accident at Three Mile Island Unit-2 (TMI-2), it became obvious that the ability to analyze abnormal and anticipated plant transients needed to be improved. A vital part in the improvement process is the assessment of new analysis methods and their applications. Both topics, assessment and applications, as related to boiling water reactors (BWRs), are discussed in this section. At the ANS Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors at Jackson, Wyoming [9-1], 12 papers addressed BWR methods assessment and applications. Specific subjects covered in the meeting presentations included: US Nuclear Regulatory Commission (NRC) research [9-2], plant transients data bases [9-3], methods assessment with subscale experimental data [9-4], methods assessment via applications to plant data [9-5-7], methods assessment via comparison of two codes [9-6, -8], and methods application to specific issues [9-9-13]. Paper contributors included the NRC, the utilities, the Electric Power Research Institute (EPRI), the national laboratories, the reactor vendor, and various foreign organizations. In this reviewer's opinion, the subjects covered at the Jackson conference were sufficiently balanced to give an overview of the current state-of-the-art of BWR methods assessment and applications, with one exception; that is, thermal-hydraulic analyses of anticipated transients without scram (ATWS) were not reported. ATWS is included in the overview of BWR methods assessment and applications presented in the following discussion. The overall objectives of the improvements in operational transient analytical methods (and thereby research, assessment, and applications) were well summarized by Beckner et al. [9-2] as follows: (1) Transient analyses are generally required to be best estimate calculations, rather than bounding
or conservative calculations. Uncertainties in modeling phenomena or plant boundary conditions cannot, therefore, be dispensed with simply by making conservative assumptions. (2) Operator actions and balance of plant systems, including nonsafety systems, play a very important role in transients and cannot be ignored or treated in a simplistic manner as is often done in design basis accident (DBA) analyses. (3) Different thermal-hydraulic phenomena occur during these transients. Low flow rates or nearly stagnant conditions result in nonequilibrium and separated flow regimes (slug, stratified, countercurrents, etc.) which are more difficult to model than the primary homogeneous flow characteristics of large-break loss-of-coolant accidents (LOCAs). (4) Analyses of ATWS require more detailed neutronics models and more accurate thermalhydraulic conditions to properly calculate reactivity feedback. (5) The long duration of these transients require more efficient running computer codes to make lengthy calculations practical. Beckner et al. indicated the Two Loop Test Apparatus (TLTA) and Full Integral Simulation Test (FIST) subscale experimental facilities have or are contributing data for the assessment of the BWR Transient Reactor Analysis Code (TRAC), R A M O N A , and the nuclear plant analyzer.
9.2. BWR methods assessment studies The need for an assessment data bank containing both experimental and plant transient data is well recognized. One such bank is being developed by the utilities' research organization, as described by Bailey et al. [9-3]. This data bank uses D A T A T R A N as the data base management system. Through D A T A -
0 0 2 9 - 5 4 9 3 / 8 4 / $ 0 3 . 0 0 © Elsevier Science Publishers B.V. ( N o r t h - H o l l a n d Physics Publishing Division)
220
G . E . Wilson / 9. B W R methods assessment and applications
TRAN, the data from various sources can be uniformly structured, stored in a central repository, and later retrieved. Plant transient data are being archived as they become available from the utilities. The bank currently contains the BWR Peach Bottom 2 turbine trip tests. Future incorporation of Grand Gulf BWR power ascension data is planned. This reviewer notes that the objectives of the EPRI data bank are similar to those of the existing NRC experimental data bank, and exchange of data between these two systems should be considered. The elements in a successful code assessment study using subscale experimental data were typified by Alamgir and Sutherland [9-4]. The data base for this study came from a small-break LOCA test in the FIST facility. FIST is a single-bundle integral system capable of full-power steady state operation and realtime simulation of transients. The test simulated a 0./)5 ft 2 break in the suction side of one recirculation pump, coupled with failure of the high-pressure core spray (HPCS) system. The assessment study was a pretest prediction of the subject test. Alamgir and Sutherland indicated that success in such studies is a direct function of the understanding of, and attention paid to probable system behavior in the system definition modeling. The general quality of this assessment (and the other papers reviewed in this section) is exemplified by the midplane temperature comparisons shown in fig. 9-1. From these, and other assessment results, the authors conclude the BWR T R A C code captured the controlling phenomena very well. They also noted, that because vessel stored heat in a BWR is significantly less than in FIST, the refill/reflood process in a BWR will be more effective than that shown in the FIST facility. This reviewer notes that this may not be true for all subscale facili1000 I / 800I-
I t -- - - -- Average measured midplane temperature .... Minimax measured midplane temperature BWR T R A C c a l c u l a t i o n
800-
I ~
/~ /
J/
P" 4 0 C -
2C
0
I
'
- -
"~
"
I 100
I 200
I 300
I 400
500
Time (s) Fig. 9 - t . C o m p a r i s o n of B W R T R A C
pretest prediction with
data from a BWR FIST small-break test without HPCS.
ties for transients where the ambient heat transfer is proportionally higher than a BWR. Several different approaches to the analysis of transients were covered by Yokobayashi and Takahashi [9-5], House et al. [9-6], and Alamgir and Sutherland [9-7]. In all cases, the codes were assessed by application to full-scale BWRs. BWRDYN [9-5] is a "components" code. Components in this context denotes a code in which the calculational algorithms are structured to describe large regions (such as a pressure vessel) or systems (such as a feedwater control). In the case of BWRDYN, the resulting thermodynamic and control system equations are integrated by the Euler method, and those describing the steam line flow dynamics by a fourth-order Runge-Kutta technique. Components codes have the distinct advantage of being fast running. For example BWRDYN has a stated ratio of calculational time to transient time in the order of 1/12. This feature is a valuable code attribute, particularly useful to scopingtype studies. Components codes are generally less useful in situations where detailed local behavior is necessary or where the facility or transient phenomena being simulated differs singificantly from that assumed during development of the code. BWRDYN was assessed with data from two transients in the T O K A I - 2 BWR: 20 cm change in water level setpoint and turbine trip. For the setpoint change transient, the reported results show good code/data agreement for feedwater flow, water level, neutron flux, and recirculation flow. In the turbine trip comparisons, steam dome pressure and recirculation flow were well calculated. The water level comparisons early in the transient were not as good as those of the setpoint-change case. From these results, Yokobayashi and Takahashi concluded the code well simulates plant behavior. Assessment of R E L A P 5 / M O D I , Cycle 14 with Peach Bottom turbine trip Tests T T I , TT2, and TT3 was reported by House et al. [9-6]. In contrast to BWRDYN, RELAP5 is a one-dimensional, twophase, nonequilibrium, building block code. That is, any system can be modeled to whatever degree of detail is desired using volumes, junctions, and heat slabs. This code can also simulate control systems, again with a building block approach. The code version used in this study required updates to model the jet pumps and to improve steam separator performance. This study served a dual assessment function in that, in addition to the RELAP5/plant data comparisons, results from simulations with the R E T R A N code were also given. In one calculation, the
G.E. Wilson / 9. B W R methods assessment and applications
R E L A P 5 and R E T R A N models were as nearly identical as possible. In three others, the RELAP5 model was modified to an optimum configuration to account for differences in the codes. From the RELAP5/RETRAN comparisons, the authors conclude both codes give similar calculational and run time results. The calculational results compare well with the plant data. Other conclusions related to desired RELAP5 modeling and initialization improvements, and to code-to-code modeling conversions were reported. Run times were not reported. However, based on this reviewer's experience and his perceptions as to the level of detail contained in the models, he would estimate the computer time was in excess of the transient time by at least one order of magnitude. This perception can be contrasted to similar information for BWRDYN, already discussed, and BWR T R A C , to follow. However, some caution is warranted in such a comparison because it is unlikely the RELAP5 and BWR T R A C models were optimized for run time. A full-scale assessment of the reactor vendor's version of BWR T R A C with Peach Bottom turbine trip Test TT1 was presented by Alamgir and Sutherland [9-7]. BWR T R A C is a building block code similar to RELAP5, except the code is multidimensional and is BWR specific. The subject study was a nonneutronic analysis because the measured core power was input as a boundary condition. The authors stated that a two step assessment scheme (thermal-hydraulics first, then neutronics) was the preferred method of analysis. The reported study constituted the first step. Code/plant data comparisons of pressure throughout the system indicated the base case calculation captured the fundamental mode of acoustic oscillation and pressure wave propagation. The trends in the collapsed downcomer water level were also correct, and considering certain modeling limitations, the absolute values of this parameter were considered satisfactory. Two sensitivity studies were reported which give further information relative to steam line nodalization and the steam bypass valve loss factor. The ratio of computer time to transient time for the base case was 126. The authors concluded the code realistically predicts the pressure transient in the main steam line and the reactor vessel. This reviewer notes that the steam dome and turbine inlet pressures calculated by R E T R A N , RELAP5 (see previous paragraph), and BWR T R A C were quite similar. Thus, there is some evidence that for this transient, multidimensional modeling was not significantly important for adequate calculational
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results. Further aspects of this perception are discussed in the conclusions section. The comparison of balance of plant (BOP) modeling with R E T R A N and the modular modeling system (MMS) was reported by Arndt et al. [9-8]. R E T R A N is a building block code, whereas MMS treats the various BOP components as modules (that is, tends to be more of a components code as previously defined). The modeling used in the study was based on the Fermi 2 BWR and was restricted to the BOP. The simulated transient was structured to address questions concerning feedwater pump suction pressures during a turbine trip. One important finding during the study was that both codes require substantial user time to establish the initial steady state conditions. This reviewer would expect this condition with the building block codes (RETRAN, RELAP5, BWR TRAC), but was somewhat surprised to find the same requirement for a components code. The reported results indicate that for the two types of codes the trends in heat transfer and water level in the heaters are similar, but there are differences in amplitude and phase. Reactor feed and heater feed pump suction pressures were in better agreement. The authors found that MMS is a powerful tool capable of saving significant labor and computer costs. However, MMS must be skillfully applied to provide reliable results.
9.3. BWR methods application studies The discussion in this section relates to methods application studies where code assessment is, at most, a secondary objective. The core kinetics aspects of the partial rod insertion incident at Browns Ferry (June 1980) was studied and reported by Kitayama et al. [9-9] using the T O S D Y N code. T O S D Y N was developed by the authors to model the reactor core with three-dimensional neutronics and parallel-channel thermal-hydraulics. In the study, the core was modeled assuming half-core symmetry to reduce computer costs. To the extent possible, rod positioning was programmed from the plant data. The results from the study led the authors to conclude that, even though the rod insertions were highly unbalanced, from the critical power ratio (CPR) view, the core responded homogeneously during the transient. In this reviewer's opinion, this conclusion may have significant implications in the analysis of ATWS, as discussed later. The capability to model the BOP with T R A C BD1/MOD1 was presented by Weaver et al. [9-10].
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T R A C - B D 1 / M O D 1 is a building block code for BWRs, and the user can model BOP units to any degree. However, in the interest of user convenience, component modules have been developed for such items as the turbine and feedwater heaters. The features of these modules were presented. Initial developmental assessments of the models indicate they were qualitatively correct. An alternate approach, relative to frequency domain analysis, to meet the objectives of stability analysis was presented by Hornyik and Naser [9-11]. This technique used the advanced (highly nonlinear) modeling capabilities of the R E T R A N code and has the advantage of being amiable to periodic perturbations. Periodic perturbations are easier to use than pseudorandom switching which must be measured and input as a boundary condition. In addition, frequency domain analysis has limitations for nonlinear systems where large perturbations near the point of unstability can be expected. The authors concluded the proposed technique potentially offers stability analyses which are more general. This study provided a dual product in that the R E T R A N model of the Peach Bottom BWR was benchmarked with the data from plant stability tests. Good qualitative agreement was obtained between the calculated and measured parameters. Once additional planned work has eliminated some discrepancies (such as smaller calculated neutron flux) further validation of the stability analysis method can be effected. Probabilistic risk assessment (PRA) techniques provide a means of determining the system success criteria needed to safely shut down a BWR after a transient initiating event. Often in PRAs, bounding assumptions (that is, best case/worst case) are made because of the lack of information or to expedite the initial analysis. In this way, the most probable dominant sequences can be quickly determined and the efficiency of the study can be assured. However, to confirm the validity of the assumptions underlying the dominant sequences, further mechanistic analyses are important. Moore et al. [9-12] reported such an application using the R E T R A N code. Of five candidate key sequences, two (station blackout and loss of feedwater coupled with loss of high-pressure systems) were selected for further evaluation with R E T R A N . In both cases, the R E T R A N analyses modified the conclusion of the original PRAs relative to success criteria. The authors concluded the coupling of mechanistic thermal-hydraulic analysis with the P R A methodology results in a more realistic evaluation of success criteria. Similar conclusions (to
be published in late 1983) have been formed in the ATWS Severe Accident Sequence Analysis Program being conducted at the Idaho and Oak Ridge National Engineering Laboratories. The application of RELAP5/MOD1 Cycle 13 to five similar Ioss-of-feedwater (LOF) transients in a typical BWR/4 was covered by Lu et al. [9-13]. The authors gave details of the sequence of events calculated by the code for all five cases. The core uncovered on a core average basis in only two of the transients (those assuming loss of high-pressure coolant injection coupled with loss of the reactor core isolation cooling, with and without the automatic depressurization system). These results did not preclude the possibility of localized core uncovery which would require a more detailed model to determine. The authors concluded their analysis demonstrates the applicability of the code and model to a wide range of LOF transients of interest in the licensing process.
9.4 Conclusions
The following discussion presents this reviewer's conclusions and perceptions relative to the state-ofthe-art as reflected by the Jackson conference and his own experience. Priority transients of interest (that is, dominant sequences) have or will soon be defined, on a first cut basis, by such methods as PRA and failure mode effects analysis (FMEA). Because these techniques often initially assume bounding conditions to achieve efficiency of study, further refinements in the P R A s / F M E A s are necessary to define the dominant sequences correctly. These refinements can be effected via feedback from mechanistic thermal-hydraulic and neutronic analyses. Significant progress has been made in the last 3 years in developing, assessing, and applying advanced thermal-hydraulic and neutronic analysis methods (codes). Similar progress in the development of necessary data bases has been made. The availability of subscale experimental data is good and is improving. The availability of plant data is not yet as good and further work is necessary. This is particularly true of BWR containment behavior. With two exceptions (as subsequently discussed), the papers in the Jackson conference indicate the basic needs for thermal-hydraulic and neutronic analysis methods are being met. Both fast running components codes and building block codes, which are slower but give more detailed local behavior, appear able to generally simulate plant behavior. Desirable refinements in these codes are known and
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the planning for the next 2 years shows promise these refinements will be developed. However, it should be noted that a common industry standard defining the required adequacy and sufficiency of code performance does not exist. It seems unlikely code development will be terminated until such a standard is formulated. Results presented at the conference indicate that in transients, which do not exhibit significant multidimensional behavior, the general thermal-hydraulic results from all the advanced codes are quite similar. Thus, it is incumbent upon the analyst to choose carefully the most efficient analytical tool. The need for such a careful selection is expected to decrease with time because options which increase running time are currently being developed for the slower running building blocks codes. The two exceptions, mentioned previously, are associated with containment analyses and with the analyses of transients which exhibit significant coupled multidimensional neutronics and thermal-hydraulic behavior (such as A T W S and core damage transients ). Presentations on these subjects were noticeably absent in the conference. Relative to containment analysis, plant data for the development/assessment/application process, is not readily available. It is also this reviewer's perception that containment code development (prior to containment failure) has a lower priority than other methods development. Successful resolution of several of the more important BWR transient questions will require good containment analysis. Thus, this is an area which needs immediate attention. tt is this reviewer's opinion that a single code, which combines adequate multidimensional neutronics, multidimensional thermal-hydraulics (including boron injection and mixing), control system, and BOP analysis capabilities, does not now exist. Even more basic is the question of just how much detail is required in the neutronics simulation of such a code. Kitayama et al. [9-9] indicated that even though the rod insertions in the Browns Ferry A T W S were highly unbalanced, the core responded homogeneously. This may indicate that a code combining all the features just mentioned is not absolutely necessary for
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adequate simulations of most ATWS. That is, it may be more efficient to develop the general core neutronics behavior in a separate components code (having multidimensional neutronics and simple thermalhydraulics), and then approximate this behavior with a good thermal-hydraulics code which has only onedimensional neutronics simulation. Further work is required to establish the optimum analytical methods to adequately and efficiently simulate the subject transients.
References [9-1] ANS Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors, Jackson, Wyoming, September 26-29, 1983 (Plenum Press, New York, 1984). [9-2] W.D. Beckner, F. Odar and L.H. Sullivan, PWR and BWR Anticipated and Abnormal Plant Transient Research Sponsored by the US NRC, ibid. [9-3] P.G. Bailey, G.A. Cordes and R.K. House, The EPRI Plant Transient Data Base, ibid. [9-4] M. Alamgir and W.A. Sutherland, FIST Small Break Accident Analysis with BWR TRAC, ibid. [9-5] M. Yokobayashi and Y. Takahashi, Verification Study of Transient Analysis Code BWRDYN Using Startup Test Data of Tokai Unit 2, ibid. [9-6] R.K. House et al., Application of RELAP5 Code for Simulation of Three Turbine Trip Transients at the Peach Bottom Unit 2 BWR, ibid. [9-7] M. Alamgir and W.A. Sutherland, Peach Bottom Transient Analysis with BWR TRAC, ibid. [9-8] C.R. Arndt, E,M. Page and T.L. Tederington, Balance of Plant Modeling with RETRAN and MMS, ibid. [9-9] K. Kitayama, S. Tsunoyama and S. Ebata, Analysis of Incomplete Control Rod Insertion by Three-Dimensional Kinetic Code Developed for BWR Simulation, ibid. [9-10] W.L. Weaver, M.M. Giles and C.M. Mohr, Balance of Plant Modeling in TRAC-BD1/MOD1, ibid. [9-11] K. Hornyik and J.A. Naser, Stability Margin of BWRs Operating with High Power/Flow Ratios, ibid. [9-12] E.V. Moore et al., Use of RETRAN for Determination of Some BWR Success Criteria, ibid. [9-13] M.S. Lu, M.M. Levine and W.G. Shier, BWR/4 Loss of Feedwater Transients Analysis, ibid.