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, LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF: | , LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF: | ||
| Line 58: | Line 57: | ||
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t1 133 core flow passa'ges and the two reactor vissel outlets was measured by ; | t1 133 core flow passa'ges and the two reactor vissel outlets was measured by ; | ||
sampling the 50 2 concentration in the air at these locations. The ratio of ;i the S0, concentration in each flow passage ta the concentration in the inlet ! | sampling the 50 2 concentration in the air at these locations. The ratio of ;i the S0, concentration in each flow passage ta the concentration in the inlet ! | ||
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.f [ ]andatotalfractionofinlet flow from two cold ley,s which _ | .f [ ]andatotalfractionofinlet flow from two cold ley,s which _ | ||
exits from the core channels on the opposite half of the .x e of - | exits from the core channels on the opposite half of the .x e of - | ||
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: z o . o a | : z o . o a E | ||
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Latest revision as of 00:27, 18 February 2020
| ML20002A228 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 10/31/1980 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML19260G378 | List: |
| References | |
| NUDOCS 8011050331 | |
| Download: ML20002A228 (15) | |
Text
{{#Wiki_filter:l l s l l l l ,t i Calvert Cliffs Unit 1 Cycle 5 NRC Reload Ouestions Response (Answers on CESEC Ilodel Used in SLB Analysis) October 31, 1980 9 9 8 0110 50 ~ 2
I p .
~
i , LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF: A. MAKES ANY WARRANTY OR REPPESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF,OR FOR DAMAGES RESULTiNG FROM THE USE OF, ANY INFORMATION, APPARATUS,
- METHOD OR PROCESS DISCLOSED IN THIS REPORT.
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1 Question I Provide the vessel cross flow mixinn data and describe the experiments from which the cross flow factors were derived.
Response
The RCS node scheme used for the Calvert Cliffs Unit 1 Cycle 5 steam line break analysis represents the reactor vessel by two parallel sets of nodes, each set of nodes representing one-half of the reactor vessel water volume between the reactor vessel inlet nozzle and tt reactor vessel outlet nozzles. See Figure I-1. This node scheme was chosen to represent in-complete mixing in the reactor vessel of coolant entering the vessel from the two steam generator loops. Mixing within the reactor vessel is repre-sented by inlet plenum and outlet plenum cross flows between the two sets of nodes, as indicated in Figure I-1A. (Figure 1-1B gives the association between node number and RCS region) . 1 The mass flow between the j and k node of Figure I-1A is denoted by W j,k. The mixing flows W2,15, W14,3, W6,19, and W18,7, are calculated usin;; constant, e:yeriuentally deteruinud .miains paranieters, F anu F g, 7 defirad such that: - 2,15
=FW7 2,3' W =FW
- 14,3 1 14,15' 6,19 " 0"6,7' and 18,7 " 0 18,19' where W 2,3' 14,15' #6,7' ^" 18,19 ar f und fr m the solution of the time dependent conserv, tion of mass and energy equations.
The values of the mixing parameters used for the analysis were F = 7 F0"- ' s These parameters ar'c calculat<! from experimental data obtained from a 0.248 - scale model of the Omaha PWR (Reference 1-1). The design of the experiment was based on the laws of similitude, using air as the working fluid. In the experiment, 50, gas was injected into one of the four reactor vessel inlet air streams. The distribution of S0 2 I""I '"
,, , _ . . _ _, _. .- , ,,T- -- # - --TF r -
g rrN 8"'
i! t1 133 core flow passa'ges and the two reactor vissel outlets was measured by ; sampling the 50 2 concentration in the air at these locations. The ratio of ;i the S0, concentration in each flow passage ta the concentration in the inlet ! was used to determine the flow distribution fraction between an inlet and any core flow passage or core outlet. These experimental results yielded a fraction cf inlet flow from one steam generator looi which reaches the outlet to the hot leg of the oppos'_tc loop
.f [ ]andatotalfractionofinlet flow from two cold ley,s which _
exits from the core channels on the opposite half of the .x e of -
~ ~
.F and F ## "" l 7 0 F = ,
I , i - and ' F
- O
Reference:
^
I-1. II. L. Crawford and L. J. Flanigan, " Final Report on Studies of Flow in a 0.248-Scale Model of the Omaha PWR," Battelle Memorial Institute, Columbus, Ohio, April, 1970. 0 9 8 .
. ' -~
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$950
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@2 E iARGING -
SAFETY INJECTION FUMPS AND TANKS -NARGING a .
NODE i ?HYSICAL'CESCRIPT:CN 1 COLD IEG 2 UPSTREAM HALF CF I:lLET PLENUM (SEFORE FLC'4 MIXING) 3 COWNSTREAM HALF 0F INLET PLENUM (AFTER FLCW MIXING) 4 BYPASS FLCN 5 CORE 6 UPSTREAM HALF 0F OUTLET PLENUM 7 COWNSTRER4 HALF 0F OUTLET PLENUM 8 HOT LEG 9 STEAM GENERATOR INLET PLENUM
. 10 UPSTREAM HALF 0F STENi GENERATOR TUSES 11 00'dNSTREsi HALF 0F STER 4 GENERATOR TUBES 12 STEAM GENERATCR CUTLET PLENUM 13 SA'ME AS l' IN OTHER STEAM GENERATOR LOOP -
14 SAME AS 2 IN OTHER STEAM GENERATOR LOOP 15 SAME AS 3 IN OTHER STEAM GENERATOR LCOP 16 SAME AS 4 IN OTHER STEAM GENERATOR LOOP
.l7 SAME AS 5 IN OTHER STEAM GENERATOR LOOP 18 SAME AS 6 IN OTHER STEAM GENERATOR LOOP 19 SAME AS 7 IN OTHER STEAM GENERATOR LOOP 20 SAME AS 3 IN OTHER STEAM GENERATOR LOOP 21 SAME AS 9 IN c! DER STEAM GENERATOP LOOP 22 SAME AS 10 IN OTHER STEAM GENERATOR LOOP 23 snag As 11 In 07aga s73;3 ggnggA7ca tcop 24 SRiE AS 12 IN OTHER STEAM GENERATOR LOOP 25 - REACTOR VESSEL CLOSURE HEAD 26 SURGE LINE :
27 PRESSURIZER o, ~ , r----[YtIU5 cas co, PHYSICAL J1ESCRIPTinti 0F PRIMARY tonLMIT !rJ0 n edes I-1B ! nano clan FOR SLB MALYSIS tlxle ce Pvacr Nani em s ee
"*w m, he
'fd T 'T N
I Question II Provide the basis for the selection of the temperature used for the moderator reactivity calculation.
Response
In the steam line break analysis, moderator reactivity increases due to moderator cooldown in the presence of a negative itTC. The method of cal-culating moderator reactivity is chosen to provide an upper bound on moderator reactivity increase during the steam line break transient. This upper bound on moderator reactivity increase is achieved, in part, by using a lower bound on moderator enthalpy as input to the moderator reactivity calculation. In this analysis, the lower bound on moderator enthalpy was obtained by using the enthalpy that the coolant from the ruptured steam generator loop would have if it advanced from the reactor vessel inlet to the core mid; lane without mixing with coolant from the intact steam generator loor. 0 4 a 4 e _;q
- ~- _ ks Q . .m'w . ,e _ gy mm__,n., m - -- M '-f;_
Question III , Describe the upper plenum coolant flow input fraction model. Ilow does this model differ from flows predicted via pressure differentials? Provide the basis for the selection of the upper plenum to upper head flow f raction and discuss if it varies with time. Address the conservat{ve nature of this codel versus a flow distribution based on pressure differ al.
Response
The flow from the region above the core to the closure head node is modeled in CESEC as a fixed, input fraction of the total flow leaving the nodes which represent the region above the core (upstream half of the outlet plenum). The flow between the closure head and the downstream half of the outlet plenum is determined from the solution of the conservation of mass and energy equations with the further constraint that this flow is equally divided between the two parallel nodes used to model the downstream half of the outlet plenum. In terms of tl.e nodal diagram shown in Finure I-l of the Response to Question 1 , W is_ the flow between the j'" and k th nodes: 3,k 6,25 " H 6,7 + 6,19}'
"18 25 " H 18,19 + 18,7).
and' 25,7 " 25,19' where Fg is an input constant, W 6,7' "18,19, nd W 25,7 " 25,19 are f und from the solution of the c(nservation of mass and energy equations, and W6,19 *" 18,7 are determined from W 6,7 and klt yg, respectively, using vessel cross flow mixing data (see Response to Question It. The factor, Fg , is not included as a source of conservatism in dm input for dds analysis (refer to the answer to Question 2 of Reference III-1 for a list of con-servatisms used). Fg is determined from an analysis based upon full flow, full power steady state pressure differentials, and was held constant throughout the analysis. During a steam line break transient, this fraction is expected to vary
, due to dcasity variations and due to variation of flow from the closure head to the outlet plenum. "This variation of F would affect the enthalpy of the closure head fluid and, inparticular,aftersaNurationisreached, it would affect the , . RCS pressure. Since safety injection flow rate is a function of RCS pressure, the variation of F couldg affect the amount of safety injection horon reaching the core. An evaluation performed for the Calvert Cliffs Unit 1 Cycle 5 steam line break analysis has shown that the effect on boron reactivity is insigni- l ficant. For example, this evaluation showed that a two order of magnitude de-crease in F causes only several hundreths o.f a percent op difference in the boron H
reactivity calculated during the transient.
Reference:
111-1. Informal transmit,tal from C. Brinkman to M. Conner, October 17, 1980. l l l m - e~ ~~ . , ~ - ~. > wes==swM --meer-wa% ww- C
t guestion IV Provide the detailed flow equations utilizcd in CESEC-SLII. Describe the equationa used to model the reactor coolant pumps. Is a two-phase degrada-tion modc1 uti1ized in CESEC-SI,li?
Response
The equations used in the steam line break analysis to model the reactor coolant pumps are given in the response to Question 3 of Reference IV-1. The two-phase degradation model was not utilized in this analysis. The fluid in the RCS cold legs remained sub-cooled during the steam line break transients analyzed for Calvert Cliffs Unit 1 Cycle 5.
Reference:
IV-1. Informal Transmittal from C.15rinkman to M. Conner, October 17, 1980. 9 l 1 l l 1
, . . _ _ - - . . . - -. . _ . . .,- ~ . -
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i
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Question V Describe in more detail the bases for your heat transfer mechanism from primary to the secondary side of the steam generator. In particular, explain the physical basis for the mass dependency and its acceptability based on applicable experimental data.
Response
The bases for the . heat transfer mechanism from the primary to the secondary side of the steam generator, including the variation in heat transfer as steam generator liquid inventory varies between 5000 and 2500 lba, are
- described in the Response to Quertion 3 of Reference V-1. It was conser-vatively assumed that the full heat transfer area exists for steam generator Avoidance of singularities in liquid invento ries _ greater than 5000 lbm.
the solution of the conservation equations requires a lower limit to be set for the ligt id mass in a steam generator. An evaluation of the sensitivity of total peak reactivity to steam generator inventory shows that a 2500 lbm uncertainty in inventory translates into less than a 0.05 %Ap uncertainty in I reactivity during SLB events. Thus, a lower limit on steam generator liquid inventory of 2500 lbm was used to avoid singularities in the solution of the conservation equations.
Reference:
9 V-1. Informal transmittal from C. Brinkman to M. Conner, October 17, 1980., . < 9 6
.s .-
4 * . F ___s..., . . . . . . . .
,,,5 . .
t Question VI Provide the details, including equations, of the stean generator dryout nocel. What assumptions are made in the secondary side phase separation model? Ca t CESEC-SLB model phase separation in the primary system? If not, why is it acceptable? How were the 5000 lba and the 2500 lba of secondary side liquid inventory limits determined for the heat transfer coefficient? Provide the tube area for Calvert Cliffs. Provide graphical output of the steam generato - primary to secondary heat transfer coefficient as a function of time for the Calvert Cliffs SLB analysis.
Response
The details of the steam generator dryout model are provided in the response to Question 3 of Reference VI-1. In the steam line break analysis, complete phase separation is assumed in the steam generator. This assumption is implemented by specifying a steam generator exit quality of 1.0. This assumption maximizes energy removal from the steam generators via the broken steam line, since a lower exit quality would result in less energy removal during the blowdown. Maximizing energy removal from the steam generators, in turn, maximizes RCS cooldown during the steam line break and thus produces the greatest potential for a return to power. Phase separation is modeled in the pressurizer; however, phase separation is not bodeled in the remainder of the primary system. A discussion follows on the effects of phase separation in the primary system during the steam line , ' break transient presented in the Calvert Cliffs Unit 1 Cycle 5 submittal. During the transients, after the pressurizer emptics, the analysis indicates that voids laitially form in the reactor vessel upper head. Voids form at this location because, as the cooldown progresses, the upper head temperature decreases more slowly than temperatures at other locations in the RCS due to the relatively low flow circulation through the upper head. Phase separation in the upper head is not expected to alter the location at which the voids form and is not expected to remove voids from the upper head. Later in the transient, after flow in the intact steam generator loop stagnates and after RCS temperature in that loop falls below the intact steam generator temperature, voids form in the stean generator tubes. Phase separation in the steam gen-erator tubes is not expected to alter the location at which these voids form and is not expected to remove voids from the stean generator tubes. Conse-quently, phat.e separation is not expected to affect the reactivity calculation. The secondary side liquid inventory limits are discussed in the response to Question V. The steam generator tube, area for Calvert Cliffs is 90,600 ft 2per steam - generator. Graphical output of the steam generator primary to secondary heat transfer coef-ficient as a function of time for the Calvert Clif fs SLB analysis is shown in Figures VI-1 and VI-2. Referenc_e: VI-1. Informal transmittal from C. Brinkman to H. Conner, October 17, 1980.
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3 gg stfon VI_I_ i Provide details of the improved safety injection (SI) model. Ilow does the new model differ from the one used in the old version of CESEC? Provide i the equations utilized in this improved model and describe the differences ]. between this model and the handling of SI in the old CESEC versions. J
Response
The safety injection (SI) model present in the version of the CESEC code which was used for the steam line break analysis for Calvert Cliffs Unit 1 Cycle 5 incorporates the same boron transport equation present in the earlier version of CESEC. The difference between the two versions is that the newer version provides a more detailed model of boron transport, as well as the transport of the mass and energy of the SI fluid itself. , This is achieved through a more detailed RCS nodal scheme and solution of
- the mass and energy conservation equations at each node for the flows used !
i in the baron transport equation. In addition, though not credited in the Cycle 5 analysis, the SI tanks are mo' deled in the newer version. See the
. Response to Question'3 of Reference VIl-1 for the equations used in the S1 model in CESEC.
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Reference:
] VII-1. I. formal transmittal from C. Brinkman to M. Conner, October 17',1980!
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