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UNITED STATES g
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JAN 1 S Ysi3 MEMORANDUM FOR:
D.B. Vassallo, Assistant Director for LWRs, DPM FROM:
R.L. Tedesco, Assistant Director for Reactor Safety, DSS
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR BABC0CK AND WILCOX'S STEAM GENERATOR ANALYSIS CODE TRAP-2 (TAC-3784)
Report
Title:
TRAP-2, FORTRAN Program for Digital Simulation of the Transient Behavior of the Once-Through Steam Generator and Associated Reactor Coolant System Report Number: BAW-10128 Responsible Branch and Project Manager:
LWR-4, J. Wilson Requested Completion Date: 7/10/79 Review Status: Under Review Th'e Analysis Branch has ccmpleted preliminary review of BAW-10128 describing the TRAP-2 code which is used by B&W to analyze postulated steam and feedwater line breaks. The code was approved with the methodology described in B-SAR-205 for analysis of the mass and energy released to the containment. This review is limited to the application of the code for the analysis of secondary system r
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breaks as they affect the primary system.
Enclosed is a list of the additional information required before we can complete our review.
9_htw R.L. Tedesco, Assistant Director for Reactor Safety Division of Systems Safety
Enclosure:
As stated cc:
R. Boyd Z. Rosztoczy W. Butler S. Hanauer P. Norian J. Mazetis R. Mattson
.W. Jensen K. Kniel T. Novak (S. Newberry M. McCoy S. Varga
'F-Tdar L. Phillips J. Wilson S. Salah
Contact:
W. Jensen, NRR, X27911 k.
8201210394 810403 PDR FOIA MADDEN 80-515 PDR
BA!!-10128 O
TRAP-2:
FORTRAN Program for Digital Simulation of the Once-Through Steam
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Generatc> and Associated Coolant System 1.) BAW-10128 states on Page iii that TRAP-2 is used for analysis of main steam line breaks and main feedwater line breaks. Provide a discussion of any additional proposed usage of the TRAP-2 code for transient or accident analysis.
2.) Provide the results of noding and time step studies which indicate.
that the solutions obtained by the TRAP-2 code are convergent.
3.) We understand that the RADAR code will be utilized to predict the minimum DNBR for the core hot channel.
Discuss how the TRAP-2 output will be utilized by the RADAR code and provide any conser-vative assumptions that will be applied to the TRAP output in generating input for RADAR. Discuss how hot channel flow reduction factors will be applied.
4.) Our previous evaluation of the RADAR code dated September 19, 1974 was limited to the use of the code for loss of flow transients.
BAW-10064 describes RADAR and states that the code is designed for the analysis of slow reactor transients. Justify the applicability of RADAR for the analysis of core response during steam line breaks and feedwater line breaks.
k 5.) Provide comparisons between the predictions of the RADAR code with those of the THETA-1B code for the hot channel transient for a double ended steam line break and a double ended feedwater line break.
6.) Following a steam line break, the increased level and decreased secondary temperature causes rapid cooling in the adjacent primary loop. This coolant may mix imperfectly in the lower plenum with the hotter fluid from the intact loop.
a.) Discuss how imperfect mixing will be treated in selecting input for the hot channel analysis utilizing the RADAR code.
b.) Provide a sensitivity study showing the effect of imperfect mixing on the core reactivity and resulting power transient which may occur following a main steam line break.
c.) The fluid leaving the core will also mix imperfectly in the upper plenum.
Provide and justify the method used to establish the temperature of the fluid in the upper plenum above the upper core support plate.
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7.) The TRAP-2 noding diagram shown in Figure 1-1 of BAW-10128 indicates that the core is not modeled as a separate fluid volume. Justify by maans of a sensitivity study that conservative results are obtained for the hot channel analysis using this approach.
8.) Figure 1-1 indicates that the upper plenum and upper' head region are described with a single node, thus, the region of relatively stagnant flow above the upper plenum cover plate may not be adequately described.
During a rapid plant cooldown the reactor coolant in this upper volume may flash and act to pressurize the primary system. Provide sensitivity studies which provide the effect of modeling this upper volume on the core DNBR following a main steam line break.
9.) Section 1.5.9 indicates that the static pressure and enthalpy calcu-lated for the upstream control volume will be used with the Moody correlatio'n to compute critical flow. This correlation is a function of the stagnation conditions at the break and use of the static conditions may result in the calculation of flow rates that are too low.
Discuss how the TRAP code or code input will be modified to utilize stagnation conditions for the Moody correlation.
10.) Justify that the two node fuel pin conduction model in the TRAP-2 code provides adequate noding for analysis of secondary system breaks.
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11.) The TRAP-2 code provides two options for prediction of departure from nucleate boiling in the core, a.) Discuss which option will be utilized for analysis of secondary system breaks.
b.) Frovide a sensitivity study showing the effect of delayed nucleate boiling in the average core analyzed in TRAP-2 on the hot channel analysis. This study should be performed in conjunction with the core noding sensitivity studies requested under Items 6.6 and 7.
Discuss the conservatism of using higher or lower than expected core heat transfer in the TRAP-2 code to establish the fluid conditions that are input to the RADAR code.
12.) Section 1.6.6 describes the core power calculation in TRAP-2 utilizing a point kinetics model, a.) Justify that this model is capable of describing local power peaks in the core that might be produced by a stuck control rod or by i
non-uniform mixing in the core of cooler water from the loop con-taining tne postulated steam line break.
b.) A recent CE paper entitled " Design Analysis Using Coupled Neutronic I
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and Thermal-Hydraulic Models" by S.G. Wagner, et. al., was presented at the Topical Meeting on Advances in Reactor Physics in Gatlinburg, Tennessee (April,1978). This paper presented an evaluation of steam line break analyses using a 3-dimensional coupled thermal-hydraulic
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neutronic code. The paper shows that cross flow in fuel bundles (open channels) inserts additional reactivity. Provide justifi-cation for not considering this effect in the TRAP code.
13.) Page 1-40 indicates that a slip flow model is utilized in the steam generator for steady-state operation and a bubble rise model is utilized for transient analysis.
a.)
Provide the slip flow equations utilized for steady-state operation.
b.) Describe how switching is accomplished between the two models at the start of a transient.
c.)
Provide the results of an analysis demonstrating that the translent model is capable of maintaining the steady-state level.in the absence of a break or utilizing a break of negliable size.
14.) Equation 1-88 is used to calculate the flow weighted enthalpy for each.
secondary steam generator node. The actual enthalpy for the secondary nodes will be lower than those predicted by equation.1-88 since the steam will travel faster than the liquid. Discuss how secondary control volume mass and energics are calculated in the TRAP-2 code to account for slip between the steam and water phases.
Discuss how these values of mass and energy are made to correspond to those of the design for the steam generator.
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15.) Page 1-31 states that a multiplfer is applied to the secondary side heat transfer coefficients to prevent discontinunities in switching from steady-state. Provide the maximum value of this multiplier that is applied for the analysis of B SAR-205.
16.) Section 1.9 describes two options that can be used in calculating the bubble rise velocity. Discuss which model is used for analysis of feed-water line breaks and steam line breaks.
Provide the results of a sensitivity study of the effect of bubble rise velocity in the steam generator on the hot channel analysis for a main st,eam line break.
17.) Provide a comparison to experimental data demonstrating the accuracy of the bubble rise model used in the TRAP-2 code for analysis of steam l
generator blowdown.
18.)
Page 1-52 states that the boron concentration in the core that is used to compute core reactivity is based on the average concentrations in the. upstream and downstream control volumes. Since no core node is l
provided, the baron concentration would be based on the average of the l
upper and lower plerum volumes. Justify that this approach is adequate for detemining the boron concentration in the core for steam and feed-water line breaks.
19.) Provide a comparison of the TRAP-2 code predictions with applicable reactor transient data.
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/e 20.)
Following a steam line break, feedwater flow will increase for the affected steam generator. This will occur when the pressdre in the steam generator with the break is reduced so that the total feedwater flow is diverted.
In addition the reduced pressure will cause flow through the feedwater pumps to approach runout conditions and flashing of the heated feedwater will cause the feedwater inventory in the lines to flow into the affected steam generator.
Discuss how these processes are analyzed in the TRAP-2 code. Discuss the affect of increased feedwater flow on the core temperature transient.
21.)
Flow through the feedwater isolation valves will be reduced during the time when the isolation valves are closing. This reduction in flow will be a function of the losses in the partially closed valve relative to the total line losses. Discuss how flow through partially closed isolation valves will be calculated in the TRAP-2 code.
22.) Provide the equations and assumptions used to model the pressurizer in the TRAP-2 code.
Include the treatment of the spray and heaters.
Discuss the accuracy of the equilibrium pressurizer model in TRAP-2 for analysis of feedwater line breaks when the system pressure is calculated to increase and cause a surge into the pressurizer.
Provide comparisons to any appropriate experimental data.
23.)
Provide and justify the assumptions used to calculate reverse heat
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. flow from the intact loop steam generator into the primary system.
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This discussion should address both the liquid and the steam regions.
24.) Provide and justify the criterion for selecting the maximum time step for use in the core kinetics calculation.
25.) To permit the NRC to perform audit calculations, provide the area, thickness and location of the primary and secondary metal heat slabs used for analysis of B-SAR-205. Justify the omission of any metal not considered in the analysis.
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