ML20095J700

Generator Tube Rupture Event - Retran Calculations

ML20095J700
Person / Time
Site:	Ginna
Issue date:	06/30/1984
From:	Tessier J, Wei T ARGONNE NATIONAL LABORATORY
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17254A929	List: ML20095J700
References
CON-FIN-A-2311 ANL-LWR-NRC-84, ANL-LWR-NRC-84-3, NUDOCS 8408290363
Download: ML20095J700 (70)

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ANL/ LWR /NRC 84-3 June 1984 GINNA SGTR EVENT - RETRAN CALCULATIONS by J. H. Tessier and T. Y. C. Wei Light Water Reactor Systems Analysis Section Reactor Analysis and Safety Division ARGONNE NATIONAL LABORATORY 9700 South Cass Avenue Argonne, Illinois 60439 Prepared for:

Division of Systems Integration Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D.C. 20555 1

NOTICE: This informal document contains sp\0NAL preliminary information prepared primarity p g**o l

, for interim use by the Office of Nuclear g r, i

Reactor Regulation, Nuclear Regulatoru o o Comission (NRC). Since it does not constitute a final report, it should be N

• Q A cited as a reference only in special l circumstances, such as requirements for N O' ed'y regulatory ndeds.

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-i-TABLE OF CONTENTS Page List of Tables........................................................... 11 L i s t o f Fi g u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 1 EXECUTIVE

SUMMARY

......................................................... 1

, 1.0 I NT RO D U CT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.0 PL' ANT M0 DEL......................................................... 4 J.0 I N P O CO M P AR I S O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1 t=0 D e c k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

, 3.2 t= 4 2 mi n s Dec k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ . . . . . . . . 13 4.0 P AR AMETR I CS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 4.1 Westi nghouse Opera tor Gui del i nes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 4.2 S tu c k O p e n P 0 RV . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4 4.3 Fai l ure tc Te rmi na te HP I . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 5.0 SU MMARY AND CO N CLU S I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 0 Acknowledgements......................................................... 61 References............................................................... 62 4e b'

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-ii-List of Tables Table Page 4.1-1 O p e ra to r Ac ti o n s . . . . . . . . . . . . . . ' . ' " ' " " * " ' ' * " * * * * ' " ' "

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-111-List of Figures Figure - Page 2-1 . RETRAN Model Vol umes and J uncti on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2-2 RETRAN M o del H ea t S l a b s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3-1 ' R CS P re s s u re vs . Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4

3-2 P re s suri zer Level v s., Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3-3 Pres suri zer Tempera tures vs . Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3-4 Core Exi t Tempe ra tu res v s. Ti me. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3-5 Reactor Yessel Upper Head Tempera tures vs. Time. . . . . . . . . . . . . . . . . . 20 3-6 R CS Av era ge Tempe ra tu re v s. Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3-7 Loop "A" Col d Leg Tempera ture vs . Time . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3-8 Loop "B" Cold Leg Tempera ture v s. Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3-9 " A" S team L i ne P re s s u re vs . Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3-10 "B" S team Li ne P res s ure v s. Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3-11 "A" Steam Generator Narrow Range Water Level vs. Time... .... .... . . 26 3-12 "B" S team Genera tor Narrow Range Level vs. Time. . . . . . . . . . . . . . . . . . 27 3-13 Steam Generator Tube Rupture, Safety Injection, and Charging F l ow v s . T i me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 3-14 Reactor Vessel Upper Head S team Vol ume vs. Time. . . . . . . . . . . . . . . . . . 29 3-15 Reactor Coolant Loop Fl ow Ra tes vs. Ti me. . . . . . . . . . . . . . . . . . . . . . . . . 30

. 3-16 Loop "B" Reactor Vessel Inlet Flow Rate vs. Time.. . . . . . . . . . . . . . . . 31

[ 3-17 RCS H ot Leg Tempera tures v s. Ti me. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 3-18 Reactor Vessel Downcomer Temoera ture vs. Time. . . . . . . . . . . . . . . . . . . . 33 3-19 " B" S team Li ne Wa ter V ol ume v s. Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 3-20 "B" Steam Generator Safety Valve Flow Rate and Flow Area v s . . T i me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 3-21 "A" S team Line Pressure vs. Time (0-20 Mi nutes) . . . . . . . . . . . . . . . . . . 36 3-22 "B" S team Line Pressure vs. Time (0-20 Mi nutes) . . . . . . . . . . . . . . . . . . 37

N_ , .

tv- I List of Figures (cont'd)

Figure Page 3-23 "A" and "B" Steam Generator Levels vs. Time (0-20 Minutes)....... 38 3-24 RCS P re s sure v s. Ti me (0 - 20 Mi nu tes ) . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 3-25 Pressuri zer P ressure vs. Time (0 - 4 Mi nutes ) . . . . . . . . . . . . . . . . . . . . 40 3-26 Core Power and RCS Average Temperature vs. Time (0 - 4 Minutes).. 41 4.1-1 RCS Vol . 61 Pressure /SG B Vol . 68 Pressure. . . . . . . . . . . . . . . . . . . . . . . . 45 l 4.1-2 U pper Head Vol . 19 Va por V o1 ume . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 4.1-3 Na rrow Range P re s suri zer Leve1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 4.1-4 S I Fl ow / B rea k F1 ow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 4.1-5 SG'8 Leve1........................................................ 47 J

4.1-6 Core Exi t V ol . 13 Tem pe ra ture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 4.1-7 S G A V o l . 58 P re s s u r e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 4.2-1 RCS Vol . 61 P ressure/SGB Vol . 68 P ressure. . . . . . . . . . . . . . . . . . . . . . . . 48 4.2-2 SG Tube Downhil l Vol . 44 Quali ty. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 4.2-3 Upper Support Vol. 20/ Active Core Vol,1/SG Tube, V o l . 4 3 Q ua 1 1 ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 4.2-4 Upper Head Vol. 19 Vapor Vo1ume.................................. 51 4.2-5 Guide Tube Vol . 14/ Bundle Top Vol . 15 Quali ty. . . . . . . . . . . . . . . . . . . . 51 4.2-6 G ui de T u be V o l . 16 Q ua l i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2 4.2-7 P re s s u ri ze r Mi x ture Lev e1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 4.2-8 Ves sel B I nl et J un 65 F1 ow. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 4.2-9 P re s s u ri z e r P O RY F1 ow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 4.2-10 S I Fl ow / B re a k F10w . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 4.2-11 Ve ssel Onwacomer V ol . 18 Tempera ture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 4.3-1 R CS ( V o l . 61 ) P re s s u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 4.3 "B" S team Li ne (Vol . 68 ) Pres sure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57

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-4.3-3 HP I , Chargi ng and Tube Rup ture _ Fl ow Ra tes . . . . . . . . . . . . . . . . . . . . . . . . 58 4.3-4 "B" S team Gene ra tor S a fe ty Val ve F10w. . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 4.3-5 Reactor Vessel Downcomer (Vol . 18 ) Tem.pera ture . . . . . . . . . . . . . . . . . . . 59

, 4.3-6 Reactor Vessel U pper Head (Vol . 19 ) S team Vol ume. . . . . . . . . . . . . . . . . 59 O

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EXECUTIVE

SUMMARY

A simulation ' of the Ginna Steam Generator Tube Rupture (SGTR) event of January - 25, .1982 was performed utilizing the latest EPRI-released version of RETRAN, RETRAN02/M0003, in conjunction with RETRAN02/ MOD 03A input decks ob- .

tained from the Institute of Nuclear Power Operations (INPO) and modified by ANL. The RETRAN02/M0003 results agree well With the INP0 RETRAN02/M0003A calculations. A reasonable match is therefore obtained between calculations and the measured data from the actual event. Where differences between the

, two calculations have occurred, they can be explained in terms of code model differences between the M0003A and the MOD 03 versions and in terms of sensi-tivities in the INPO calibration to data. In addition, three parametrics were performed which included variations on operator actions and further equipment failure. Results of _ these parametric calculations demonstrated; that oppor-tune timing in conforming to recent operator guidelines would prevent filling the disrupted steam generator solid and alleviate concerns about loading ques-f tions; that additional failures occurring in the PORY line downstream of the PORV would not necessarily lead to significant core damage; and that sufff-

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cient thermal margin exists against pressurized thermal shock conditions even if there was a further continuation of safety injection flow. Furthermore, I the parametrics have contributed to the understanding of the thermal hydrualic phenomena that occurred during the actual sequence of events during the Ginna SGTR incident.

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1.0 INTRODUCTION

The Institute of Nuclear Power Operations (INPO) recently performed a simulation [1] of the Ginna steam . generator tube rupture (SGTR) event of January 25, 1982 [2] using an interim version of the RETRAN02/M0002 code

[33. In this report results are presented for a comparison of calculations utilizing RETRAN02/ MOD 03, the most recent release of the code, and the RETRAN02/M0003A input deck furnished by INPO with the actual event. In addi-ti on, parametric calculations were carried out varying the scenario of the actual event in terms of operator actions and mechanical failures. These served to increase the understanding of the thermal hydraulic phenomena which occurred during the incident. The three parametrics performed were:

a) A duplication of the event with a variation in the operator ac-ti ons. The latest Westinghouse operator guidelines for SGTR, E-3 (July 5, 1982) were followed to determine their efficacy in pre-venting the ruptured generator from going solid.

b) The pressurizer PORY (pressure operated relief valve) is assumed to stick open during the operator's efforts to depressurize the primary

', side and the downstream block valve is presumed to concurrently fail in the open position. This examines the ability of the SI (safety injection) to maintain the system in a stable condition.

c) Finally, the safety injection (SI) system was presumed to be left on beyond the point of termination in the actual sequence of events.

The quasi-steady cooldown rate in the downcomer obtained in this parametric could be of significance to pressurized thermal shock problems.

Thi s document details the parametric calculations and discusses the results obtained, as well as those calculations performed for comparison with the INPO computations. Section 2.0 describes the INPO plant model used, the various input modifications which had to be made for successful execution and the coding changes to the RETRAN02/M0003 source program required to correct for code deficiencies. The result of the comparison against the INPO calcula-tions and concurrently the Ginna data are described in Section 3.0 while the parametrics are presented in Section 4.0. Conclusions are drawn in Section 5.0.

1 .

2.0 PLANT MODEL The Ginna SGTR everit started with the plant at normal operating conditions with- the primary side entirely in single phase, with the exception of the two-region pressurizer which was in thermodynamic equilibrium, and a

. secondary side which was in two phase steaming off into the turbine. Upon tube rupture, the primary side commenced depressurization with a loss of inventory through the rupture. Flow through the rupture was choked. The secondary side of the SGB began to respond in the manner of a two region nonequilibrium pressurizer model (with a mixture level) as- the steam generator began to fill up. Complications in the thermal hydraulic response were introduced because of system feedback effects with turbine load reduction, reactor scram and safety injection ini tia tion. Fowever, until the primary system had depressurized such that the relatively stagnant upper head region reached the saturation temperature the entire- primary loop was governed by single phase hydraulics. The outsurge from the two region pressurizer can be treated by a nonequilibrium pressurizer model. Bulk flashing such as that l which ultimately occurred in the upper head is also a phenomena simulated by nonequilibrium pressurizer models. Natural convection occurreo on the primary side during the flashing period as the pumps coasted down and the operators initiated manual depressuriza tion. The pressurizer rapidly refflied during the manual depressurization leadf ig to possible nonequilibrium conditions. On the secondary side, the tubes did not uncover so dryout is not a concern and the heat transfer is the normal two phase heat transfer. While pressure measurements were available on the secondary side, data are lacking on the flow through the various valves as the ruptured B steam generator filled solid. (Data are limited in general as the Ginna plant is not instrumented as an experimental facility). With the filling of SGB, single phase choked flow through the safety relief valves from the basically incompressible volume occurred. The primary side also tended to single phase again during this period of filling SGB as the head region steam bubble began to collapse with the continuation of SI flow. Heat losses and thermodynamic nonequilibrium had to be considered during the bubble collapse. Finally, the termination of the SI did not lead to the introduction of additional thermal hydraulic phenomena not previously discussed.

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Plant models developed to simulate the Ginna SGTR event have to envelope

. all these varied thermal hydraulic phenomena. The Institute of Nuclear Power

, Operations developed such a plant model using engineering judgement where necessary to compensate. for the limitations of the data' availability, di s-cussed earlier, and has obtained reasonable agreement with the event after calibrations.

h INP0 provided RETRAN02/M0003A decks (t=0 and t=42.5 minute decks) and restart information which used the volume / junction nodalization shown in Fig.

2-1. Figure 2 presents the heat slab nodalization used. The Fig. 2-1 nodalization was used from time = 0 to 42.5 minutes at which point renodali-zation was performed. The renodalization mainly took the form of moving the 5

nonequilibrium pressurizer model from the disrupted steam generator (SGB) dome

. to the corresponding steam line. This apparently was done, at least in part, in order to avoid numerical problems with the nonequilibrium pressurizer model when complete filling occurred. There was also. a change in the volume split

'between volumes 19 and 20 when complete draining or filling took place. In summary, at transient time equal to 42.5 minutes, the RETRAN model was revised S- by INPO to enable treating tne SGB (steam generator B) steam line as a non-equilibrium pressurizer volume during the period that it filled with liquid. l This was considered the most realistic modeling available and permits

! selection of a spray option (with condensation calculated) and variation in the inter-region heat transi'er coefficient through appropriate input changes made in restart decks. It was also necessary to change the geometric model of the . steam line to ensure that the liquid flowing from the steam generator, into the line, always entered the liquid region in order that excessive condensation due to homogeneous mixing of liquid and steam did not occur when

.- the spray option was turned off. This required placement of the junction connecting the steam generator and line at an artificially low elevation, but the effect of this on the calculation is judged to be acceptably small. The

] impetus for incorporating the non-equilibrium model of the steam line was to provide added means for controlling the calculated pressure levels to match measurements. For further details regarding the nodalization reference should

. be maoe to the INPO draft report [1].

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l The aforementioned renodaliza tion, and other INPO revisions, cannot be accomplished in a restart deck. Consequently, an entire new input deck (t =

42.5 ' mins deck) was required, initialized to non-steady condi tions. *These ini tial conditions were the end-point values of plant parameters calculated

- for the initial portion of transient (t=0 deck). It is noted that RETRAN does not permit supplying inputs for all . required variables, e.g., heat conductor temperatures, surface heat fluxes and values of slip in junctions having two-phase flow. Such Avantities are normally computed for steady-state conditions

- by i teration. However, for initialization in a dynamic condition, the steady-

, state initialization option is bypassed (word number 25, JSST, on card number 01000Y is set equal to 1) and quantities that are not input are computed without a staady state search. This- approach is reasonable if time deriva-tives in the relevant equations are acceptably small. This apparently is true for this application since there is no evidence of significant discontinuities in calculated results at the time of transition to the new input deck.

In order to match the measered data with the limited instrumentation capability available INPO had to undertake fairly extensive calibration ef-forts. Among the more significant adjustments made are renormalizations of upper head " valve" areas, steam generator B S/R valve arear, MSIY flows and PORY areas. The INPO draft document provides additional information [1]. In E general the RETRAN02/M0003A decks obtained from INPO, once modifications were

! made for the RETRAN02/ MOD 03 version, were used "as is" to produce tne results documented in this report. Minor alterations had to be made in order to

affect the changes required for the three parametrics and these are discussed

in the respective sections. More extensive alterations had to be made to

! compensate for code modifications to obtain the comparison results with the INP0 computations for the actual Ginna event. These input deck al tera-tions/ code model modifications are listed below.

For the t=0 deck:

a) the input for the upper head heat slabs had to be altered to accom-J modate for the changes to the non-equilibrium pressurizer heat loss model; i

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9 b) the steam. generator dome to bundle junction had to be forced 'into a.

zero slip calculation to avoid non-physical results. This was the original M0003A option;-

while it proved infeasible to compensate for the change 'in bubble

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c) rise model it should be noted that M0003 normalizes the Wilson -

, bubble rise correlation to the rise velocity computed in the steady l state initializer. This was not done in MOD 03A.

For the t=42.5 minutes deck:

d) the Wilson bubble rise model- yielded a bubble veloci ty of zero except for the inttial value; e) job failures occurred when the calculation called for closure of a previously opened SRV; i

f) non-physical results were calculated at the time of SI termination (two phase flow with slip in a junction connected to a non- ,

equilibrium volume).

j Each of these problems were resolved af ter consultation with EI person-i 1 nel. Some were circumvented by changes to the input deck whereas others involved FORTRAN source changes. In addition to these problems, M0003 j stability required smaller integration time - steps than those used in the INPO i calculations (MOD 03A). This increased the demand on computer time and caused

} a need for more restarts to complete these calculations.

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3.0 INPO COMPARISON ,

-In order to simulate the measured system response during the Ginna SGTR event, INPO [1] divided the transient into two parts: one from initiation of tube rupture to commencement of manual depressurization (t=42.5 minutes); the second from t=42.5 minutes to shortly after SI termination at which point the INPO computation as terminated. Discussion of the ANL comparison with INPO is therefore similarly divided into two parts. Section 3.1 detafis the comparison from t=0 to t=42.5 minutes and Section 3.2 concludes with the latter part of the transient from t=42.5 minutes to SI termination.

3.1 t=0 EDeck 4

During the period from t=0, the point at dich tube rupture occurs, to t=42.5 minutes when operator-initiated depressurization using the pressur-izer PORY commences, the reactor system goes through a turbine load reduction phase, reactor scram and a tripping in of the safety injection (SI). Concur-

. rently the condenser dump valves are cycled, the various feedeter pumps are manually controlled and the isolation procedures are followed by the operators in order to minimize dose rates and to reestablish control. Refereice should be made to the INPO draft report for a detailed chrcnicle of the actual Ginna SGTR event.

The t=0 deck takes the event from the initiation of tube rupture to the time- when operator initiated depressurization through the pressurizer PORY at t=42.5 minutes occurs. At this point the problem is renodalized and calcu~

1ation continued with the t=42.5 minutes deck. In addition to the splicing of results made necessary by this renodalization procedure at 42.5 minutes there is an additional splice ; necessitated by INPO's further recalibration during the time period 112-180 seconds. With the differences in code models between RETRAN02/M0003A, the code version used by INPO, and RETRAN02/M0003, the version utilized by ANL to produce the results presented in this report, it can be seen that there are possibilities for discontinuities at the times of 112 seconds, 180 seconds and' a2.5 minutes. While some smoothing could be rationalized and was indeed used at these junctures, the result do show some discontinuities at these points. These will be discussed in perspective. The

\

perspective of this document is to concentrate on the differences between the INPO calculations and the ANL computations and to understand them in terms of modifications in code models between the two code versions 'and also of sensi-tivities in the INP0 calibration to the data from the original event.

Figures 3-1 to 3-26* present the comparison between the INPO and ANL resul ts. This set of graphs represents the . entire set presented in the main text of the INPO draft report. The ANL curves have been traced onto the INPO figures. Reasonable agreement has, in general, been achieved between the INPO RETRAN02/M0003A calculation and the ANL RETRAN02/M0003 computation. Where differences have arisen they cu be attributed to three or four modifications as discussed in subsequent paragraphs.

In order to compute the narrow range pressurizer level adjusted for instrument error at nonsaturation conditions the input deck had to be altered to incorporate a stand alone control block model. An error was discovered in the transcri,ption process which affected the adjusted level. As the control

• block is a stand alone model utilizing input thermal / hydraulic (T/H) conditions from the main RETRAN T/H calculation it is completely ignored by the main computation. However, Fig. 3-2 for the pressurizer level shows that the error ler.ds to an underprediction of the instrument adjusted level.

In RETRAN02/M0003A the nonequilibrium pressurizer model had no heat loss associated with the volume boundary. Two-sided heat slabs could be

attached to the nonequilibrium volume by the user but the volume does not "see" tne neat slabs. The heat slabs however do "see" the volume. This model was altered in M0003 with the slabs and volume interchanging heat in an energy conservative manner. However, the use of an adiabatic boundary condition is now necessary on the non-pressurizer volume side of the heat slabs with Mod

03. These alterations imply that the heat transfer in the upper head region, which is modelled by INPO using the non-equilibrium pressurizer model, particularly that incurred by the " fictitious" conductor (slab 20) cannot be duplicated by the ANL M0003 calculation. Consequences of this model 1

• These Figures are grouped at the end of Chapter 3.

I 1

g? ~

1-_

modification can be seen in Figs. 3-1 (RCS pressure),' Fig. 3-5 (vessel upper head temperatures) and Fig. 3-14 (vessel upper head steam volume). There are certainly effects on othe: parameters-but those should be of lesser importance particularly for the parameters on the secondary side. Heat losses from the pressurizer pe: se are treated through the use of a control block and there are no heat slabs associated with the steam generator domes. Thus even though i the non-equilibrium pressurizer model is used for the pressurizer and the steam generator domes this model alteration should not directly affect those volumes.

A modification was made to the bubble rise model between the M0003A and the M0003 code versions. The MOD 03 model now normalizes the bubble rise velocity computed by the Wilson bubble rise correlation to the' velocity com-puted by the initializer to obtain steady' state. This may sound inconsequen-tial .as the Wilson correlation is only used in the nonequilibrium pressurizer model and pressurizers are normally initialized with no bubbles. However, INPO chose to use the non-equilibrium pressurizer model in the steam generator

. domes.- Upon switching out of the automatic initializer and proceeding into 3 the transient MCD03A would use a different bubble rise velocity from MOD 03.

The implications of this difference is an alteration in steam generator level behavior. Figures 3-11 (A SG water level), Fig. 3-12 (B SG water level) and tim corresponding figure for the first 20 minutes of the transient, Fig. 3-23, show the difference in initial swell which can be attributed to this model

$ modi fica tion. The INPO calibration of steam flow, etc., has to be recali-brated to data to account for the al teration in code model. This change should 'also affect steam generator pressures '

Finally, the effect of the splicing discussed earlier can be seen in Figs. 3-14, and 3-24 to 3-26. To reiterate however, in general the ANL compu-tations using RETRAN02/M0003 compare reasonably with the INP0 calculation using RETRAN02/ MOD 03A. Where differences do occur attribution can be made to model alterations and calibration sensitivities.

i

. .. 1

. .

j 13 l 3.2 t=42 mins Deck This portion of the -calculation ~ spans the transient time from 42.5 to 80.0 minutes ' af ter tube rupture. These results are also depicted in Figs.

3-1 through 3-20 ' with the initial portion discussed earlier. This latter portion of the Ginna calculation begins 'with initial opening of the pressur-izer PORY to accelerate RCS depressurization and concludes shortly after SI

, termi na tion. Throughout this period, decay heat removal and RCS cooldown occurred by continued. injection 'of the relatively cold SI water and feed and bleed operation of the intact steam generator (SGA) as modelled in the' input decks. Since there was no measure of the SG steam and feedwater flows, they were adjusted in the model to provide correspondence with measured pressure

and water level data.

As the aforementioned figures show, there is excellent agreement between ANL's calculated results and those obtained by INPO. The plots for primary side parameters overlay nearly exactly, .except for some of.the fine structure in the oscillatory behavior of certain parameters. In addition to the agreement shown for RCS temperatures, pressures and flows, there is also essential overlap in the curves of tube rupture flow and upper head steam-formation for the two sets of calculations (Figs. 3-13 and 3-14).

The only differences of note, albeit small, are those in parameters

~

calculated for the faulted steam generator (SGB). INPO, in their calcu-

la ti ons, used a complicated program for the area of the SGB safety relief valve (SRY) to control calculated pressure to match measured data. To do this, several restarts were made wherein both open and close pressure dependent trip setpoints and open and close valve area table entries were changed. Effectively, this process varied the valve area with time, opening or closing it contingent on calculated pressure levels. The trip setpoints were not the normal plant values, but were selected to be close to the running i

pressure levels as indicated by data. Thus, small differences in calculated pressure levels caused the valve behavior to be different in the ANL calcu-la ti on s.

f

1. ,,,,,.-,y_

_ _ . _ , . , . y ,_. , . . _ - _ - , , , , . , .

, . . , , . _ _ . , . .,._._.,v

14 4

l The first departure from INP0's results is evident in Fig. 3-2 which shows three openings and closings of. the SGB SRY in the time period of ~ 51 to 55 minutes whereas only a single cycle was calculated by INPO. This is also i manifested in the sawtooth nature of the ANL curve of SGB pressure (Fig. 3-10)

' duri ng this time period.- . These ' additional ' openings of 'the relief valve released more-liquid from the system to ' the atmosphere causing a slight delay

( ~ 1.4 minutes) in the calculated time to completely fill the B steam line with liquid as shown in Fig. 3-19.

' After the secondary side became solid, the calculated SGB pressure

. levels became very sensitive to the flow resistance out of the system at the SRY. In the INPO calculations, the valve area, and hence its resistance, was varied often using pressure trips and area tables in the manner described above, attempting to match pressure data. Using INP0's trip setpoints and valve area tables in the AHL calculations proved unworkable because the timing for switching between valve opening.and closing modes is critical to obtaining the proper areas versus time. This timing of actuating the open/close trips -

is not deducible from the reported results and even small calculated pressure differences soon resulted in erroneous trip times and attendant valve area values, and subsequent large deviations in pressure levels. It was necessary, therefore, to change the method of programming the valve area with time in the ANL calculations. The INPO curve of valve area shown in Fig. 3-20 was used to derive numerical values. which were entered in an appropriate RETRAN table and the trips were changed such that af ter t = 55.42 minutes all areas were obtained- from that table. Also, the time at which the sudden large area increase occurs was delayed to be coincident with the later time to fill the steam line with liquid as calculated by ANL. This approach, while not precise, yields levels of agreement with INP0's results considered adequate as examplified by the comparisons of SGB pressure levels and SRY flow rates shown in Figs. 3-10 and 3-20 respectively.

4 In sumary, the results obtained by ANL for this phase of the Ginna event using RETRAN02 Mod 03 also show excellent agreement with and confirm INP0's earlier calculations. Although some differences were encountered, they are not sufficient to negate this conclusion. Also, the available measured l

da ta is represented quite well by the calculated system respvises lending l

l

-- . _ . - _ - __,_ --.._- , _ - _ . _ . _ , - . - _ _ - ~ . _ . , - . , - _ . _ - , _ . _ . . , . ~ _ - . - . _ . - . _ . - , _ _ . . . , _

i ,

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15 credence to INPO's overall conclusions regarding the transient plant status following the actual steam generator tube rupture event.

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Rapid in-Surge Of Continued Riso in Pressurizer ,

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42

\

4.0 PARAMETRICS Three parametric analyses were conducted in order to further clarify the understanding of the thermal hydraulic phenomena which occurred during the actual sequence of events. As discussed earlier these are: simulation of the most recent Westinghouse operator guidelines; further mechanical failure as embodied in the assumptien of a non-responding block valve; and a variation in operator action . with continuation of the SI flow. It should be understood that for. the most. part the tables / control blocks, made necessary by the INPO calibration to the data obtained from the actual event, .were not altered for these parametrics. Section .4.1 details the operator guidelines computation while Sections 4.2 and 4.3 describe the stuck open PORY and SI continuation parametrics, respectively.

- 4.1 Westinghouse Operator Guidelines Upon comparison of the actual event sequence of ths Ginna SGTR incident to the latest Westinghouse Operator Guidelines for SGTR (E-3, July 5, 1982) it is to be concluded that the operator actions conformed with the E-3 guidelines. However, implementation of certain procedures were delayed enough

, to negate actions takert to prevent the ruptured steam generator from going solid. For this parametric the event was reanalyzed using the INPO t=0 deck with operator initiated PORY depressurization moved up, by approximately 20 minutes, to 25 minutes after initiation of the transient. Judging- from the chronicle of the Ginna event, this scenario should provide sufficient time for operator action and, in addition, all Westinghouse guideline conditions for depressurization had been met at this time. Modifications were made to the deck -to follow the latcet hestinghouse operator guideline from the point of p pressurizer PORV depressurization and continuing on until the calculation was terminated at the time when the operator could energize the pressurizer heaters and reestablish pressure control. No alterations were made to most of the numerous tables / control blocks made necessary by INP0's calibration to the actual Ginna event. Chief among these assumptions during the relevant phase of the incident is the use of upper head "nive" area tables to simulate time dependent form factors. As these areas are held constant during the period of interest to the parametric considered here, the utilization of the tables

. _ .  :, - , _ . . ~ . .. O _.-._ , .-.___ ,__.,_- ..._... _ .,,,. , _.. m _ ,...__._..-_,__-,-,,.m.

43 i

1 could be justified on the grounds of simplification of a multidimensional geometrical problem. In addi tion, no modifications were made to the tables / control blocks for the behavior of the A steam generator (intact generator) which implies -that the procedure for utlizing the intact generator

- as a- heat sink follows that of the Ginna- event exactly. Modifications, however, had to be made to the pressurizer heater control blocks.

Figures 4.1-1 to 4.1-7 illustrate the system response. While the transient is plotted for the period from 1000 seconds after tube rupture on, a ttention should be focused upon the PORY depressurization and post-PORY depressurization period, namely from 1500 seconds to 2400 seconds. The event sequence of the period prior to 1410 seconds corresponds exactly to that of the actual Ginna incident. Table 4.1-1 shows the timing of the various operator actions required by the guidelines.

M

. .- l 44 Table 4.1-1. Operator Actions Event Time (secs)

Charging on 1410 Open pressurizer PORY 1500 Cycle PORY Same sequence as in actual Ginna incident. Just moved initiation up.

to 1500 seconds.

SI flow termination 1720-1765 (45 second ramp)

Letdown established 1850

, The reactor coolant system (RCS) pressure trace, Fig. 4.1-1, shows g tne four operator initiated pressurizer PORY openings, the PORY block valve closing and the termination of the safety injection. Figure 4.1-2 depicts the

, collapse of the upper head bubble with the closure of the PORY block valve.

Volume 20, the volume directly below volume 19 and physically a part of the reactor vessel upper head region, also undergoes flashing and fu'rther complete steam bubbles collapse during this' period. This volume was modeled as a homo-geneous volume by INPNO. As volume 19 does not completely empty, this is a physically consistent picture. The narrow range pressurizer level instrumen-ta tion (uncorrected for non-saturation conditions), Fi g. 4.1-3, shows a re-

. filling to ~ 65% where a leveling off takes place. At this point the operator could re-energize the heaters and reestablish pressure control. From Figs.

4.1-1 and 4.1-4 it can be seen that a quasi-steady state has been -reached with the tube leakage flow now of a negligible proportion and the difference in primary to secondary pressure attributable mainly to the hydrostatic head of a ruptured steam generator filled with substantially more liquid than at normal operating conditions. However, Fig. 4.1-5 which is the narrow range B gener-

. ator level, shows that there is considerable margin to filling up the ruptured generator. In the context of filling the steam dome, a narrow-range measure-ment of ~ 210% is to be considered as full. The core exit temperature of Fig.

4.1-6 shows that the degree of subcooling has decreased to 10*F, but that is due to the use of the actual operator actions for steam generator A (intact generator) during the actual event. Figure 4.1-7 depicts the A generator pressure which is an indicator of the generator behavior and can be compared wi th Fig. 3-9.

i I

45

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TIME (SEC)

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54 4.2 - Stuck Open PORY i

For this parametric it is postulated that when the pressurizer PORY

failed to close as it did during the actual event, the ~ downstream block valve i failed in the fully open position'. Thus unlike the actual series of events, j and additional small break LOCA compounds the original SGTR. At this point the Westinghouse SGTR operator guidelines call for the LOCA guidelines. In this parametric the SGTR guidelines are followed to this time and then the calcuation is continued without any operator intervention until primary side

. recovery commences. The computation indicates that this occurs within a few

- minutes at which point the analysis is terminated with the level in the vessel slowly rising. Figures 4.2-1 to 4.2-11 show the response of the system. The calculation is performed with the t=42.5 minutes INPO deck so the time origin f is t=42.5 minutes of the -actual Ginna event when the operator initiated system depressurization actions in order to reduce tite primary / secondary pressure di f ferential .- .The only change made to the deck were alterations to the PORY area cards. None of the numerous time tables were modified which imples, for

[ example, that the intact generator (SGA) is operated exactly the way that it was during the Ginna event. The RCS pressure trace (volume 61), Fig. 4.2-1, graphs the four pressurizer PORY openings and then at ~ 150 seconds a pla-teauing which can be attributed to flashing in the tubes (volumes 43 and 44) of the disrupted steam generator (SGB). Figures 4.2-2 and 4.2-3 for the steam generator tube quality confirms this. With isolation the disrupted generator becomes a region of low flow and stagnation. Reverse heat transfer across the disrupted generator tubes tends to hold the pressure up. Figures 4.2-4 to

• ~

4.2-6 show that the head (volume 19) had completely voided at ~ 160 seconds and tha t flashing in the guide tubes (volumes 14 and 16) had already ccm-menced. The flashing in the steam generator tubes and the vessel head region F causes the pressurizer to fill as depicted in Fig. 4.2-7 at ~ 185 seconds.

Wi th the filling the PORY (junction 122) begins to discharge single-phase liquid. There is an initial surge out of the generator into the vessel with the flashing, as can be seen in Fig. 4.2-8, but mora stable conditions occur

within minutes with the SI/ charging flow dominating over the PORY flow from a now solid pressurizer. Figures 4.2-9 and 4.2-10 show these flows. Reverse leakage occurs in the disrupted generator and the level in the vessel head is i slowly ri sing. As can be seen in Figs. 4.2-5 to 4.2-6 the guide tubes had

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55 refilled at ~ 160 seconds when the SG tubes had commenced flashing. From Figs. 4.2-3 and 4.2-5 it can be concluded that no significant core uncovery

' occurred during this period and the calculation was terminated at this point.

The minimum downcomer temperature in Fig. 4.2-11 appears to have decreased by approximately 10*F during this period as compared to the actual Ginna event.

Further simulation may require renodalization to compute bubble collapse in the disrupted steam generator tubes and modifications regarding the tables and control blocks for. the intact steam generator behavior.

4.3 Failure to Terminate HPI For this case, ANL reanalyzed the Ginna event, as it occurred, but in this analysis it was assumed that the operator did not terminate operation of the high pressure injection system (HPI). The objective of this analysis wa: to determine the consequences in the primary and secondary systems when failing to terminate HPI.

In the actual Ginna event, the operator secured the HPI pumps at one hour and twelve minutes after tube rupture. This was considerably later than permitted by procedures with the delay largely attributed to operator's concern for potential core uncovering due to upper head void formation. More-over, if RCS depressurization had begun when the governing criteria were met

.it is likely that filling the "B" steam generator and line would have been avoided (see Section 4.1). All of this is to say that there actually was a significant time delay in HPI termination during the Ginna tube rupture event .

which was accommodated without serious consequences to the plant. Securing the HPI pumps reduced the ongoing cooldown of the RCS and the discharge of radioactive water to the environment; some release continued because the charging pumps remained on and their flow exceeded the letdown rate. INPO estimated that the "B" generator SRY did not completely close until three hours and two minutes after tube rupture; at the time their analysis was concluded at one hour and twenty minutes, the estimated mass of water released through this SRV was 64,000 lbm. l l

. :,L . . . - _ - - . . - - -L . _ _ - _ , - _ _ . _ - . . - . . - - , _ _ , . - - - - --

56 For the purposes of this analysis, INP0's modeling of operator control of the intact steam generator (feed and bleed) was unaltered; the only

-changes made in the model were to inhibit tripping the HPI and to maintain a constant flow area for SGB SRY equal to that assumed when HPI was . terminated i in the actual event. The calculation was continued for eight minutes beyond

~

f actual HPI termination to show the response . trends in the primary and second-ary systems. The salient results of this calculation are shown in Figs. 4.3-1 through 4.3-6. These graphs begin at the time the INPO deck was re-nodalized; .

all results up to the time actual HPI termination occurred are identical to the original calculation.

As shown by comparing Figs. 4.3-1 and 4.3-2, primary system (RCS) pressure remains above that of SGB by approximately 300 psi. ' This pressure

_ differential maintains flow through the ruptured tube into SGB and attendant release through its SRY. These flow rates are nearly steady at approximately 600 gpm, essentially the sum of HPI and charging flow rates. Figures 4.3-3 4

and 4.3-4 depict these continued flows.

l The sustained injection of relatively cold HPI water into thi RCS

, causes its moderate rate of cooldown to continue. For example, examination of the calculated fluid temperature in the reactor vessel downcomer, as shown

in Fig. 4.3-5, gives an estimated rate of approximately -70*F/ hour during the end period of the calculation; the results also indicate a slow reduction in cooldown rate as anticipated. Based upon a limiting acceptable rate of

-100*F/ hour, these results show that a certain thermal margin still exists even for continued operation of the HPI system. It is also noted that the continued injection of cold water causes eventual elimination of the upper j head void at the time the calculation is ended as shown in Fig. 4.3-6.

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SUMMARY

AND CONCLUSION A simulation of the Ginna SGTR event of January 25, .1982 was performed utilizing' the latest EPRI release version of RETRAN [33, RETRAN02/ MOD 03 in conjunction with RETRAN02/ MOD 03A input decks obtained from INPO and modified by ANL. The RETRAN02/ MOD 03 results agree well with the INPO RETRAN02/ MOD 03A calculations [1]. A reasonable match is therefore obtained between calculations and measured data from the actual event [2]. Where differences have occurred, they can be explained in terms of code model differences between the M0003A and the M0003 versions and in terms of sehsitivities in the INPO calibration to data. In order to match the limited thermal hydraulic data available, INPO had to resort to a number of calibration adjustments, based on engineering. judgement, in its application of the thermal hydraulic models available in the RETRAN02 code. To further the unders'tanding of the thermal hydraulic phenomena which occurred during the actual event, three additional parametric calculations were performed which included variations on operator actions and further equipment failure. The three parametrics performed demonstrated; that opportune timing in conforming to recent operator guidelines would prevent the filling the disrupted st,eam generator solid and alleviate concerns about loading questions; that failures in the PORY line downstream of the PORY would e not necessarily lead to significant core damage and; that sufficient thermal margin exists for pressurized thermal shock si tua tions even if there ydts a further continuation of safety injection flow. While there is a significant out-of-containment loss of ECCS inventory in the third parametric where SI flow was not terminated, timely conformance to the operator guidelines, as evidenced by the first parametric, would prevent such a condition from occurring. The parametrics have contributed to the understanding of the thermal hydrualic phenomena that occurred during the actual sequence of events during the Ginna SGTR incident.

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61 Acknowledgements We are indebted to R. Eliasz of Rochester Gas and Electric (RGAE) and to the staff at the Institute of Nuclear Power Operations (INPO), E. Winkler and R. Wyri ck, in particular, for their . cooperation and efforts in transferring the INPO decks to ANL and in providing ' additional clarifications. M. Paulsen and J. McFadden of Energy Incorporated (EI) are also to be thanked for their suggestions on RETRAN questions.

This report was prepared by ANL staff in partial fulfillment of a project under the direction of the U. S. NRC Division of Systems Integration, R. J.

Mattson, Director; B. Sheron, Branch Chief for Reactor Systems; N. Lauben, Section Leader; J. Guttmann, Project Manager.

ANL staff who provided input to this report were J. H. Tessier and T. Y.

C. Wei, authors; and K. Rank and M. Mehaffey, report preparation.

1

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p 62 References ,

1

1. INPO staff, " Thermal-Hydraulic Analysis of Ginna Stean Generator. Tube Rupture Event," Institute of Nuclear ' Power Operations draft report (September 1983).

2. NRC staff, "NRC Report on the January 25, 1982 Steam Generator Tube Rupture at .R. E. Ginna Nuclear Power Plant," Nuclear Regulatory Commission Report, NUREG-0909 (April 1982).

3. J. H. McFadden et al., "RETRAN-02 A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," Electric Power Research Institute Report, EPRI NP-1850-CCM (May 1981).

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