ML20087H571
| ML20087H571 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/31/1984 |
| From: | Kunita R CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20087H552 | List: |
| References | |
| NF-1583.04-1, NF-1583.04-1-A01, NF-1583.04-1-A1, NUDOCS 8403210080 | |
| Download: ML20087H571 (41) | |
Text
.. .
e NF-1583.04-1 VERIFICATION OF CP&L REFERENCE BWR THERMAL-HYDRAULIC METHODS USING THE FIBWR CODE AMENDMENT 1 RESPONSE TO NRC QUESTIONS MARCH 1984 i
APPROVED BY: M. I, b1C(44/MJ R. K. Kunita Principal Engineer-Surveillance and Accountability CAROLINA POWER & LIGHT COMPANY 411 FAYETTEVILLE STREET HALL RALEIGH, NORTH CAROLINA 27602 8403210080 840315 PDRADOCK05000g
.P
, INTRODUCTION
- This amendment to Topical Report NF-1583.04, " Verification of CP&L Reference BWR Thermal-Hydraulic Methods Using the FIBWR Code," is provided in response to requests for additional information as conveyed by the letter to Mr. E.January dated E. .Utley,'
~17, CP&L, 1984. from Mr. Domenic B. . Vasallo, NRC Division of Licensing,
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Question 1 Please. identify any changes or modifications over the Electric Power Resear Institute(e.g.,
models version of the FIBWR code. c conservation equations, For these changes, if any, the modified models, etc.) should be described. void fraction, and subcooled boiling
Response
There code. have been two modifications at CP&L tothe Both of these modifications involved the EPRI version of the FIBWR without altering any of the existing models or programming. addition of capabilities consisted critical powerof the addition
/ heat of the GEXL correlation as a new option for theThe first cha flux calculations.
form of a separate subroutine and uses The correlation was inserted in the user-supplied correlation. a call statement reserved for a of the CP&L Topical. The verification of this change is included as part The simulator, nodal exit secondPRESTO-B.
modification was made to facilitate the FIBWR interac Equations for determining the effective inlet and form Carolina" loss& coefficients Power Light Company for PRESTO Topical, are described in Section 4.3.2 "A Description and Validation of Steady-State Analysis Methods for Boiling Water Reactors" (CP&L File NF-1583.01, Regulatory February 1983), currently under review by the Nuclear Commission.
Values for each of the variables in these equations are either input to or calculated by FIBWR and are, therefore, stored in the FIBWR common arrays at the end of the execution.
programming Equations 4.3.2 and 4.3.3 of NF-1583.01 This modification involved into FIBWR so that just prior to the end of execution, the PRESTO loss coefficients would be calculated and printed on the summary page of the FIBWR output.
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I 4 Question 2
- Define the " slow' transients" in Item 2 on Page 4.
will.be.-analyzed by the FIBWR code for the hot bundle analysis.Specify which transien
--_Reaponse The steady-state FIBWR code can be used to analyze the type of transients in which surfacethe heatmoderator lto the deposition rate in the coolant due to conductdon from the clad
. generation rate of the fuel..can be' conservatively represented by the heat the average core enthalpy changes slowly are the inadvertent pump HPCI start and loss of feedwater heater transients.
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n M: Question 3 I
Describe; how the FIBWR code - is coupled to the neutronics code - for power
. distribution - calculations - (e.g. , the iteration between the neutronic and l thermal-hydraulic calculations and the differences in input -to FIBWR between the neutronic-type calculations and the CPR-type calculations) Also describe .
- inthe ' coupling Item 4.on Page between
- 2. the FIBWR code and the plant process computer as men
~ Response The for theFIBWR code neutronics calculates the core flow' distribution in order to define input code.
flow curve from FIBWR to.be inputThis is done by either defining a leakage flow / total input to PRESTO to match FIBWR bypassinto PRESTO or adjusting loss coefficients-
~ conditions.- calculations ~ for specific operating inlet and exitFIBWR' 1 flow distributions loss ' coefficients for PRESTO areby also used tothe weighting calculate orifice effective , tie plate, 'and grid strap losses by. the relative flow through each component.
A description of the neutronics code' coupling with FIBWR is given in Section specified4.3.2 of the Carolina in Question 1. Power & Light ~ Company Topical NF-1583.01 bypass flow calculations from the input used in the CPR-type calcul .
Future 1 applications ~ may involve calculating input process computer analogous . co -the effective inlet constants for the plant
- PRESTO. loss coefficients used by methods. :These and areconstants' are currently determined by the vendor's hydraulic -
supplied for each Brunswick unit at the beginning of the operating cycle.
Should CP&L decide to use FIEWR for process computer applications, further documentation will be submitted to the NRC at that time.
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y,,,,.m.i-n,v ,,e,ww.e,-w*-.,- .r *yv-p e r m v c er*~ t--w rer - r i m v - *w e vv =w w me w n+-m m v -< 1, v s-
' Question 5 Explain the applicability of the spacers and lower and upper tie plate form loss coefficients from Reference 3. Table 5.1 (i.e. , EPRI-NP-1923) to the Brunswick core thermal-hydraulic analysis (cf. Table 2 of the CP&L Trpical) .
Response
The EPRI-NP-1923only.
geometry-dependent loss coefficients in Table 2 of the CP&L Topical are all The lower tie plate, upper tie plate, and spacer loss coefficients in EPRI-NP-1923 were derived for fuel assemblies with geometries identical to those at Brunswick and, therefore, should be applicable for Brunswick that core thermal-hydraulic analysis. Equal in importance is the fact the ability to match Brunswick hydraulic conditions.the CP&L FIBWR
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Question 6
. Explain how the water _ rods are modeled hydraulically. Are they different from the active flow channels?
Response
The channel.
-flow water rods.are represented as a separate flow path parallel to the active .
- of the water tubes: Section 4.7 of EPRI-NP-1924-CCM gives the following description pressure drop of the"FIBWR calculates active coolant parallel the to water tube flow consistent with the the tube.
The entrance and exit elevttions
-each channel. are input with refereitce to the start of the heated region of The water tube is assumed to start at or above the bottom of the reference length, but may extend up into the upper unheated region.
The FIBWR code models the water tube _ pressure drop in a similar manner to the active
- coolant, two-phase with ' theloss local exception thatmodels multiplier the homogeneous void relationship and are used should the water tube
' experience bulk boiling. No friction loss multiplier or two-phase corrections to allow to the acceleration the water -pressure drop are included; if the flow rate is low enough tube to boil, significant. the pressure loss components will not be I
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_ Question 7 .
Only the form loss coefficients for water rod entrances are given in Table 2 Explain why the exit form loss coefficients are not given. .
Response
data provided by the GE document specifiedTheinexit NRC loss Quest cuafficients used'are those given in EPRI-NP-1923.
the FIBWR input for the water rod form 1 css coefficients is given below:The method 1.
A single-bundle model representing an 8x8 retrofit assembly was establishedspecified conditions in orderintothe setGE thedocument.
power and active flow equal to the nominal input were taken from EPRI-NP-1923. Initial form loss coefficients 2.
Several FIBWR cases were then run, varying the water rod entrance loss coefficient until the water tube flow natched the value given in the GE document.
The entrance loss coefficient was selected over the exit coefficient at the exit;because flow into the water tube is, more restricted than flow bottom. therefore, most of the water tube pressure drop occurs at the 3.
Loss coefficients derived in Step 2 were verified by running an off-nominal case and confirming the agreement between the FIBWR-calculated water tube flow rate and the GE data for the off-nominal condition.
4.
Finally, an approximation was needed for the water tube entrance form loss coefficient of calculating theanratio 8x8Kstandard fuel assembly. This was done by based on the respective water tube geometries,whereKist$ep/Kahik$elaverageofthelossthroughthe tube entrance holes given by the equation:
K= A REF
~
2h2 g
i= I il -
4 and A
nRjF = vater tube flow arca, number of water tube entrance holes,
~A ~
= area of watei tube entrance holes, and Kf=square-edged holes). water hole loss coefficient (1.5 was used for all Values for 4 ,
(CESTAR II), ShEtion 2.
n, and Ag were obtained from NEDE-240ll-P-A-6 The cmpirically-derived value for K from Step 2 was multiplied by the geometry ratio defined above to lIrkbe at an approximation for the standard 8x8 water tube entrance loss coefficient, which design.corresponds with the emp'.rical metnod used for the 8x8 retrofit
Question 9 Provide the detailed derivation of the leakage coefficients .(C1,C2.C3, and C4) .for bypass flow paths in Table 4.
the Justify the parameters needed in derivations,
~different paths. such as pressure differentials and flow fractions of
.those documents. If Ceneral Electric Company information is used, provide
Response
bypass flow paths wereThe leakage coefficients C1 C2,C3, and C4 used by FIBWR calculated by an iterative method. In the first iteration, the process described in section 5.1.4 of EPRI-NP-1923 was e appli to bypass flow fractions from open literature and initial guesses at pressure differentials to calculate an initial set of' bypass coefficients. These coefficients were used in FIBWR to generate a new set of pressure drops for the core support plate and channel dependent leakage paths.
During the second iteration, the pressure differentials generated by Electric in the document described in NRC Questions values for C1,C2,C3 'and C4.
Lower Tie Plate Flow Holes (Path 9)
From GE Report NEDC-21215 (Reference 6 of the CP&L Topical):
W = 259.6[ 4 P (Eq. 1) where-W = flow through the two lower tie plate bypass holes (Ib/hr) f = density of coolant (lb/ft )
oP = differential (psi) pressure across lower tie plate (LTP) holes Equation 1 is of the form W = C1 oP with + C3aP C1=259.6f+C2aP C2 = C3 = C4 = 0 Assuming a density of.47.12 lb/ft ,
plate flow holes (path 9) becomes: the value of C1 for the lower tie . t C1 = 259.6 (47.12)b = 1782.0 '
and their relative bypass flow fractions.The remaining leakage coefficient Control Rod Paths (paths 1,2, and 5)
Flow fraction for path 1: 0.297 Flow fraction for path 2: 0.048
' Flow fraction for path 5: 0.003 Total control rod flow fraction: 0.348
i-
-Flow fraction for path 9: 0.346 Number of control rod paths: 137
-Number of LTP paths: 560 AP control rod paths: 24.0 psi
-oP path 9: 10.0 psi
.(560) C19 AP 9 = (137) Cl er O cr
.348
-(560) .346 (1782) (10)b = (137) C1C# (24)b C1 = 4729.0 El = C3 = C4 = 0
~
Instrument Tube Paths (path 3)
Flow fraction for path 3: 0.002 Flow fraction for path 9: 0.346 Number of Instrument tubes: 31 Number of LTP paths: 560
'AP path 3:
'AP path 9:
24 psi 10 psi 560 C19 6P 9 346 = (31) C13 aP3 560 (1782) (10) = (31) C1 3 (2M C1 = 120.1 CE = C3 = C4 = 0 Shroud Path (path 4)
-Flow fraction for path 4: 0.001 Flow fraction for path-9: 0.346 Number'of shroud paths: 1 Number of LTP paths: 560 AP path 4: 24.0 psi AP path 9: 10.0 psi 560 C19 oP = C14 'oP 3 9 4 560 1782 (10)b = C1 4 (24)
C1 1861.7 '
.Ck==C3=C4=0 9
g , -..rr, -,---.-r ..----,,-,y - . , - - , -wa ---,+-n m----,-m-r-- ,--c,--, a-o,,- , - - - - -
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. Spring-Plus Paths (Path 10)
^ Flow fraction for path 10: 0.005
, Flow fraction'for path 9:
0.346 Number.of spring plugs: 77-
~
_ Number of LTP paths: 560 4P: path 10: 24.0 ps'i AP path 9i '10.0 psi 560' C19 AP = 77 C1 0 6 9 10 10
=560 0.005 3 1782 (10)b = 77 C110 (24)b C1 = 120.9 b3=C3=C4=0
-Fuel Support Paths (Path 6)
Flow fraction for path 6: 0.015 Flow fraction for path 9:
0.346 Number' of fuel ' support paths: 560 Number of LTP paths: 560 AP._ path 6: 10 psi ,
-AP path 9: 10 psi 560 C1 ' AP = 560 C16 OI6 3 9 9
, -(560) 6 _1782-(10)
= _ (560) C16 (10)b Cl 77.25 CI=C3=C4=0
=
~ Finger. Spring Path (Path 8)
The flow Brunswick 2,' Cyclefractions5. provided by GE represent core average end-of-cycle for A more detailed model of the finger spring leakage path was made'using the Cycle 5 fuel mix to represent finger spring deformation in:
fuel assemblies which.have been incore longer than one cycle. The leakage , _ . . .t
._ coefficients defined developed'from a flowintest - EPRI-NP-1923, using fresh fuel.section 5.1.4.1 for path 8 _ were in the CP&L model'to represent new fuel. The .These s ne coefficients are used
- i. core verage finger spring leakage fuel: coefficient can be expressed as a~ weighted mixture of new and used
.(A) Cigyg =-(B) C1 ,,+ (A-B) Cl 3 used where- 'C1 5 06
. ( oP g). = 702(AP y).
06 C1 g = 702( AP ).
C1 ,,= 702 (9.1).2106 .
g C1 = 1117.7 g
and CI is defined as follows:
AVG
-Flow fraction for path 8: 0.280 Flow fraction for path 9: 0.346 Number of finger spring paths: 560 Number of LTP paths: 560 AP path 8:
OP path 9:
9.5 10.0 560 80 C1pP9 = 560 C1pPg 560
- 1782 (10) = 560 C1 8 (9.5)b C1 AVG = 1479.5 A = 560 B = 160 560 (1479.5)
Cl
= 160 -(1117.7) + 400 (C1 )
used
- 0
- The final interation involved a verification of the parameters used to generate- the leakage coefficients against output from a FISWR execution.
Agreement between pressure drop val 6es shown below eliminated the need for an additional iteration.
- oP Location Path 6 . Iteration #1 Iteration #2 10.0 Path 9 9.81 10.0 Path 8 (avg) 9.81 9.5
> Path 8 (new fuel) 9.1 9.36 C're Support Plate 24.0 9.04 23.93 With the exception of the control rod dependent path, flow fractions calculated leakage from the FIEWR output compared favorably with the GE values.
coefficient The results shown in Table 3 of the CP&L Topical.for this one path was adjusted uce the down
, . .x The leakage coefficients C1.C2,C3, and C4 define the bypass flow as a function of pressure drop- for given fuel and lower internal geometry. The bypass flow fractions provided by General Electric represent core average values, and variations for specific bundles are expected. However, we have confidence to evaluate bypassthat flows the constants from pressure calculated with these flow fractions drops typical can b conditions . of Unit 2 operating .
Because of similar physical characteristics and pressure drop of Unit 2 as well as to Brunswick Unit 1. ranges, we also believe that these co
'002
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- Question-10 Provide the values of A and B in Equation 4 used for ~ the Brunswick analy Also provide.the values of ELDE and ELDG, which are inputs to the FIBUR code .
as -adjustment.
(See Pages 6 factors for the upper tie plate and grid spacers, ,
through 115 of EPRI-NP-1924-CCM.) Explain why therespectively.
modified homogeneous model for the
'EPRI-NP-1924-CCM versus Equation 5 of the CP&L Topical) is not use .
- Response
- The values of A and B used in Equation 4 are those given on Page .-
US B of 103 the - General Electric Company - Report ~ NEDE-24011-P-A-6-US (CESTAR II). The spacer and' upper tie plate two-phase multiplier adjustment factors (ELDG, ELDE
~
.respectively) used in the Brunswick analysis were input as: ,
ELDG = 1.0 ELDE = 1.0 The selected option for the homogeneous model having the form:two-phase form loss multiplier was the modified 4
2-Phase Local = 1 + (E) (X) ([l/ [g - 1)
(Equation 4-24 of EPRI-NP-1924-CCM) where E = Adjustment Factors (ELDG, ELDE)
X = Equilibrium Flow Quality
/"1, /"g = Saturated Liquid and Saturated Vapor Densities
, Respectively However, since the value of E was selected as 1.0, this equation reduces to:
2 4
-Phase Local " ()( I! 8 ~ 1)
(Equation 5 of the CP&L Topical) which is the homogeneous two-phase form loss multiplier.
The use of a 1.0 adjustment factor is consistent with the Yankee Atomic methodology demonstrated in EPRI-1924-CCM and is recommended for design purposes by R. T. Layhey and F. J. Moody in Reference 8 of the CP&L Topical.
5
y f_ Question 11 .a
.' process computer outputs were used.In benchmarkingthe , theplant core press
. process computer models use the same iterative calculational techn
~ FIBWR code, but with different and sometimes less' detailed models In light of this, explain the usefulness of these benchmark comparisons .
_ Response The plant process ' computer. pressure drop model uses cycle-specific , input
. coefficients core pressure determined drop. by the vendor's detailed hydraulic codes to calculate Although the process computer itself may use less detailed component representations than FIBWR, it is designed to produce the same
-pressure drop / flow characteristics as its more complicated parent code. The value of benchmarking to the plant process computer is the extension of the indirectly cal::ulated byFIBWRthe vendor'svalidation by a-comparison of FIBUR-cor rops detailed hydraulic models. These
. effectively. reflectcomparisons are also' useful in demonstrating the ability of the FI ano ther. -- changes- in the core loading pattern from one cycle to I
^
e 126
~
. Question 12 Discuss if there were any corrections to the FIBWR calculated pressure drop across the core support plate to natch the exact locations where th pressure tap measurements were made (cf. Table 6).
calculated and measured AP's are different, If the locations for low-flow situations that in high-flow where the static head becomes more importantthe situations. than er
-_ Response' The measured core support plate pressure drop [m) is determined from pressure FIGURE taps located in the Standby Liquid Control System illustrated in 12a.
'Using the notation in FIGURE 12a, P, , PS- 4 (Eq. 1)
.and, assuming constant density, f5=f3+f1 4= (Eq. 2) 2 + / (1 + h) (Eq. 3) where: /* = density of coolant h
1
= elevation of tap 2 above tap 3
= elevation of tap 3 above pressure transmitter both static and dynamic pressure losses occur.As coolant flows from ,
point 3 can be expressed as: Therefore, the pressure at 3=E'2+/'h+AP ggy (Eq. 4) where:
-AP g/ h = static losses between points 3 and 2
= acceleration) pressure drop associated with core flow (friction, local, subs'ituting t
Eq. 4 into Eq.2 5 = '2 + [ h +f 1 + A Pgg (Eq. 5)
AP m canEq.
. substituting be 5expressed and Eq. 3 in intoterms of its pressure drop components by Eq. 1:
AP oP" ,oP2+/*h+[l+oPI 0" [( h+1)
= .
(Eq. 6)
-Thesupport
-the core static terms plate.all cancel out leaving only the dynamic losses through The FIBWR core support plate pressure drop (oPp )
of EPRI-NP-1923 as the differential pressure between the inletis defined on page 5-18 to the. orifice
and the top'of the core _ support plate. This involves both static and dynamic losses and can be expressed as:
. app , APgy, + [X (Eq. 7) where: AP f7 = flow losses through core support Y = and elevation between the center line of the inlet orifice the top of the core support plate. (Sce FIGURE 12a)
The difference between the FIEWR calculnted pressure drop and the core support plate pressure drop from the pressure caps is:
correction term = 4P - AP
= AP y + fX - oP ggg
=f X FIGURE top of the fuel12-b gives the location of the core support plate relative to the support piece.
The location of the inlet orifice can be scaled from FIGURE 12-c~ to give a value of X equal to 4 inches.
UsingtypicaloperatingconditionsinaBWR,['=46.17lb/ft.
The correction term [X becomes (46.'17 lb./ft ) (4.0 in) 1728 in jfg 3
= 0.11 psi.
We consider _0.11 psi to be negligible and, therefore, no correction term was applied _to the figures in Table 6 of the topical report.
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_ Question 14 used in Figures 6 and'7 data comparisons. Provide the documents os
Response
The local channel conditions and critical power ratios in Figures 6 and 7 the Topical were obtained from the following documents: of
- 1. General Electric Company:
" General Electric Company Boiling Water Reactor Reload 1 Licensing Amendment Plant Unit 2," NED0-24029, June 1977 (Attachment 14-1)for Brunswick
- 2. General Electric Company:
" Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 1. Reload 2"; NED0-24239; January 1980 (Attachment 14-2)
- 3. General Electric Company:
for Brunswick Steam Electric Plant" Supplemental Reloed Licensing Submittal Unit 2, Reload 2"; NED0-24150; October 1978 (Attachment 14-3)
- 4. General Electric Company:
" Supplemental Reload Licensing Submittal for Brunswick Steam Recirculation Pump Electric Plant Unit 1, Reload 3 (With Trip)"; Y1003J01A52; (Attachment 14-4) January 1983 5,
General Electric Company:
" Supplemental Reload Licensing Submittal for Brunswick Steam Recirculation Pump Electric Plant Unit 1, Reload 3 (Without Trip)"; Y1003J01AS3; (Attachment 14-5) January 1983
- 6. General Electric Company:
" Supplemental Reload Licensing Submittal for Brunsuick Steam Electric Plant Unit 2 Reload 4 (Without Recirculation Pump Trip)"; Y1003J01A37; June 1982 (Attachment 14-6)
- 7. Cencral Electric Company:
" Supplemental Reload Licensing Submittal for Brunswick Stu o i.lec tric Plant Unit 2, Reload 4 (With
~
Recirculation Pump ' trip)"; Y1003J01A45; June 1982 (Attachment 14-7) 8.
Letter to Mr. J. D. Martin of Carolina Power & Light Company From Mr.
Unit L. M. Quintana of General Electric Company; " Additional Brur.swick LMQ83-101; September 2, Cycle $ Confirmatory Load Line Limit Analysis In 27,1983 (Attachment 14-8)
The cover page from each of the above documents, accompanied by the pages containing the Documents local channel 4 through 7 are drafconditions, t are provided as a t ta chmt.n t s.
been published at the time that versions of reload submittals which had not purposes.
supplied toAttachment Carolina Power &8Light is a formal transmittal of data which was p information purposes only. Company by General Electric Company for in the formal transmittal differed from the initial data which was us FIBUR benchmark.
This new data point has been evaluated using the CEXL
correlation illustrated ininFiguresFIBWR,14-a andand the'resulting 14-b. change to Figure 7 of the topical is Figure 14-a is a reproduction of the original Figure 7. and. Figure.14-b shows the same figure with the new 7x7 data point. '
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i GENERAL ELECTRIC E i
BOILING WATER REACTOR I RELOAD-1 LICENSING AMENDMENT FOR BRUNSWICK STEAM ELECTRIC PLANT [
UNIT 2 .
?l e
e t
4
,g, ,,,_y-ng,+ -ww--w ** - - , . . , . - . - . .
' F *
- r P 1
. -- ~ ~ ~ --tItP 6 - - J :~ ~ ', ".~
Attachee:t 14-1 ~ 7- - -
.- NEDO-24029 .
Table 4-2;, ,, ,[.
SUMMARY
OF RESULTS OF LIMITING ABNORMAL ,
OPERATIONAL TRANSIENTS e Maximum ACPR EOC-2 7x7 8x8 Turbine Trip w/o Bypass 0.17 0.i; Load Rejection w/o Bypass 0.18 0.24 Loss of 100*F FW Heater 0.13 0.14 Inadvertent HPCI Pump Start 0.08 0.10 Rod Withdrawal Error (RBM set at 106%) 0.20 0.17 Feedwater Controller Failure - 135?. NBR.run out capacity 0.05 0.06 .
Table 4-3 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS
~
e 7x7 8x8 Peaking Factors (Local, Radial, Axial) (1.24, 1.24, 1.40) (1.22, 1 345, 1.40)
R-Factor 1.100 1.098 Bundle Power (MWt) 5.294 5.738 Nonfuel power Fraction 0.04 0.04 l
Core Flow (M1b/hr) 77.0 77.0 Bundle Flow (103 lb/hr) 126.3 117 7 Reactor Pressure (psia) 1035 1035 Inlet Enthalpy (Btu /lb) 526.9 526.9 Initial MCPR 1.26 1 30 4
1.
I
==r , l Atttchment 14-2 HEDO 24239 SONEn258 CLASSI JANUARY 1960 I TW%DN $$lI$M5 NES $b 5b'o D N Y=N=Hh1E b55'64"w*bY $ 'ki5(SY U S ? N W$hh
?j r
SUPPLEMENTAL RELOAD 2 LICENSING SUBMITTAL FOR 2, BRUNSWICK STEAM ELECTRIC -
PLANT Is it UNIT 1 n
. RELOAD 2 0 u
I I li lt I
i
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~~.
Attachsert 14-2 NEDO-24239
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) .
Shutdown Margin (ak) pre (20*C, Xenon Free) -
l 600 0.045
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2) r EOC1-2000 BOC3 to mwd /t to EOC3-2000 mwd /t i EOC3
-8.50/-10.62 -8.70/-10.87 void Coeffictent N/Aa (c/1 Rg) 41.6 41.6 Void Fraction (%)
/
-0.225/-0.214 -0.220/-0.209 Doppler coefficisnt
- N/A (c/*F) 1374 1374 Average Fuel Temperature ('F)
-37.93/-30.34 -36.72/-29.38 Scram Worth N/A ($)
Figura 2a _ Figure 2b Scram Reactivity . _
- 7. RELOAD UNIQUE C2 TAB TRANSIENT ANALYSIS INITIAL CONDITION PARAM _
F8x8R_ ,
> 8x8. 8x8R_ I I
50C3 to 50C3 to BOC3 to EOC3-2000 EOC3- 20C3-2000 EOC3-E0C3-2000 EOC3-2000 mwd /t to 2000 mwd /t to 2000 mwd /t to EOC3 mwd /t EOC3 mwd /c _
EOC3 mwd /t_
1.22, 1.22, 1.20, 1.20, 1.20, 1.20 Peaking factor 1.54, 1.45, 1.52
~
(local,' radial, 1.34, 1.40 1.47, .
1.40 1.40 1.40 1.40 1.40 asial) 1.40 1.051 1.051 1.051 1.098 1.098 R factor 1.051 .
5.960 6.260 6.558 6.178 6.487 -
Bundia Power (MWt) 5.703 112.1 114.3 112.1 115.4 113.0 g
- Sundle Flow 114.2 (103 lb/hr) .
1.24 1.30 1.24 1.32 1.25 ,
Initial MCFR 1.30 ,
"N = Nuclear input data ,
A = Used in transient analysis
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OCTOBER 1978
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SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 2 a
4 I
~
?'IPal b FIFCTi..J. .
Attachment 24-3
- 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3 3.2.1.5 and 5.2)
EOC Void Coefficient N/A* (t/5 Reg) 8.10/10.13 Void Fraction (5) 41.76 Doppler Coefficient N/A (t/%0F) 0.1938/0,1841 Average Fuel Temperature (OF) 1538 Scram idorth N/A ($) 38.85/31.08 Scram Reactivity vs Time Figure 2 e
Exposu re * *
- EOC EOC EOC Peaking factors (local, radial and axial) 1.24/1.234/1.40 1.22/1 324/1.40 1.22/1.460/1.40 R-Factor 1.100 1.098 1.051 Sundle Power (MWt) 5.267 5.646 6.220 i Bundle Flow (103 lb/hr) 125.82 116.24 117.03
- Initial MCPR 1.25 1 32 1 32
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
~
( None
. ./. ~ ..
l- *N Nuclear Input Data A: Used in Transient Analysis -
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,- Attachsent 14-4 a
Y1003J01A52 Rev. 0
'4. CALCUI.ATED CORE EFFECTIVE HULTIPLICATION AND C0ffrROL SYSTEM WORTH .-
NO VOIDS. 20*C (3.3.2.1.1 AND 3.3.2.1.2)
BegInning of Cycle, k,7g Uncontrolled 1.11; Fully Controlled 0.958 Strongest Control Rod Out 0.939 R. Maximum lucrease in Cold Core Reac t tvi t y wi t h E.spii~ist i- 0.0 in Cvele. AK
- 5. STANDR,Y 1.IO,UID CONTR01. SYSTEM SilUTDOWN CAPABil.lTY ( l.1..'. ! 3) .
%u t d.w: " u .: i n .'en np_m t .'0
- C , ).e n. n F s . /
600 0.031 i
- 6. kELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5'AND 5.2) (REDY EVENTS ONLY)
EOC '.
Void Fraction (%) 41.3 Average Fuel Temperature (*F) 1302 Void Coefficient N/A* (C/% Rg) -8.33/-10.42 Doppler Coefficient N/A (c/*F) -0.232/-0.220
' Scram Worth N/A ($) -46.31/-37.05
- 7. REI,0AD UNIQUE CETAB TRANSIENT ANALYSIS INITI Al. CONDIT,10N PAPAMETERS (5.2)
Bundle Fuel " "" "" "
R- Power Hundle Flow lisit f al Design I.oca l R.id l .n l Axlal Fac t o r (tnt ) (103 lb/lir) Mr'PR B'H: 4 to EOC 4 -'
P8x8R 1.20 1.57 1.40 1. 0'a i fa.n71 109.9 L.20
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~
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._ - 4 Attach:st 14-5
. .g Y1003J01A53 Rev. 0
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTFA WORTil - NO VOIDS, 20*C (3.3.2.1.1 and 3. 3.2.1. 2)
BOC k g Uncontrolled I.I14 Fully' Controlled 0.958 Strongest Control Rod Out 0.954 R, Maximum Increase in Cold Core Reactivity 0.0 with Exposure into Cycle, Ak
- 5. STANDBY LIQUID CONTROL SYSTEM SilVTDOWN CAPABILITY ( 3. l . .! .1. 3 )
Shutdan Mard n ( .. ?
ppm _.'O Y, Xenen Ftce) 600 0.031
- 6. REIAAD UNIQUE TRANSIENT ANALYSIS INPLfr (3. 3.2.1. 5 'and 5. 2)
(REDY EVENTS ONLY)
EOC 4-2000 mwd /T EOC 4 Void Fraction (%) 41.3 41.3 Average Fuel Temperature (*F) 1302 1302 Void Coefficient N/A* (C/% Rg) -8.12/-10.15 8.33/-10.42 Doppler Coefficient N/A (C/*F) -0.219/-0.208 -0.232/-0.220 Scram Worth N/A ($) -46.31/-37.05 -46.31/-37.05
- 7. RELOAD-UNIQUE CETAB TRANSIENT ANALYSIS INITIAL. CONDITION PARAMETERS (5.2)
Fuel Peaking Factors Bundle Power Hundle flow inittal Design (Local, Radial, Axial) R-Fac t o r (MWL) (103 lb/hr) MCI'R 1:0C 4 to EOC 4-2000 mwd /T _
P8x8R 1.20 1.53 1.40 1.051 6. ',16 111.0 1.21 8x8R 1.20 1.55 1.40 1.051 6.614 110.3 1.21 8x8 1.22 1.42 1.40 1.09H 6.04% 109.6 1.21 M
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! 7.
RELOAD-UNIOUE CETAB TRANS!ENT ANALYSIS _ INITIAL CONDTTION PARAMETERS (S.2.2)
I
'! Peaking Facturs l
Fuel Exposure (Local, Radial. Bundle Power Bundle Flow Initial Design (.Wd /ST) Axial) R-Pactor (M'.*t ) fl03 lb/hr) MCPR 8x8 EOC (1.22,1.33.1.40) 1.093 5.686 113.5 1.29 ECC-2000 (1.22,1.45,1.40) 1.098 6.187 110.2 1.18 j 8x5R ECC (1.20,1.46,1.'0) 1.031 6.210 11".1 1.30 EOC-2000 (1.20,1.59.1.40) 1.051 6.757 110.8 1.19 PS::SR EOC (1.20.1.43,1.40) 1.051 6.092 115.6 1.33 EOC-2000 (1.20.1.57,1.40) 1.051 6.664 112.1
-- 1.21 --
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorizacion: No .
Recirculation Pump Trip: No Rod Withdrawal Limiter: No
~s Thermal Power Monitor: Yes f
,(,,) Measured Scram Time: No Exposure Dependent Limits: Yes Exposures Analyzed: EOC and ECC-2000 :CTd/ST
- 9. CO' RE-WIDE TRANSIENT ANALYSIS REST *LT! (S.2.2.1) aCPR Exposure h h/A (Uncerrected) pg, Transfert (MWJ/ST) (2 N3R) j] NBR) a Ex9 3x R P2x92 Fesconse
? .
Load ECC 582 128 0.22 0.23 0.26 Figure 2a
, Rejection Withouc EGC-2000 447 117 0.11 0.12 0.14 Figure 25 Svpass Loss of BOC-EOC 121 122 0.13 0.13 0.13 Figure 3 100*F Teedvater
- j,. Heating ,
Feedvater EOC 127 110 0.04 0.04 0.05 Figure 4a i
Controller i Failure ECC-2090 113 110 0.04 0.05 0.05 Figure ab
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. .e Y1003J01A45 Rev. 1
- 5. STANDBY LIOUID COMTROL SYSTE!! SitUTDOWN CAPABILITY O.3.2.1.3)
Shutdown Marlin (ak) 223 (20*C. Xenon Free) 600 0.036
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT ( 3. 3.2.1. 5 ANL S.2. 2)
E0C Void Coefficient N/A- (c/0 Kg) -7.!t!-9.93 Void Fraction (*.) 41.9 Doppler Coefficient N/A (:/* *F) -0.219*-0.209 Average Fuel Te=perature (*F) .312 Scram Worth N/A (S)** ---
- 7. RELOAD-tlNIOCE GETAB TPANSIENT ANALYSIS INITI AL CCDITION PAR.L"ETERS (S.2.2's
'eaking .* actors fuel (Local, Radial, dundle Fever Bundle Flow Initial
~~
Design Exoosure Axial) R-Factor (MWt) (103 lb/hr) .v.C P R - 8x8 EOCS (1.22, 1.43, 1.40) 1.098 6.114 110.'s 1.20 8x8R EOC5 (1.20, 1.57, 1.40) 1.051 6.689 111.2 1.20 h P8x8R EOC5 (1.20, 1,56, 1.40) 1.051 6.620 112.4 1.22
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization: No Recirculation Pump Trip: Yes , Rod Withdrawal Limiter: No - Thermal Fower Monitor: Yes . Measured Scras Tine: No Exposure Dependent Limits: No l*#M" *N = Nuclear Input Data ^ A = L' sed in Transient Analysis
- **Ceneric exposure independent values are used as ci.en in " General Electric Standard Applicatio. for Reaccar fuel," NEDE-2eGil-P-A
- January 1982.
6 2
. _ - - -. . -- - - , - . . - - , - . . ~ - . - -
Attcchacnt 14-8 GENER AL $ ELECTRIC NucttAn ENEncy eusiNtss oPERADONs CfNERAL ELECTRIC cCW#At#
- 175 (URTNER AVENUE
- SAN Jose, CAtlFCWNIA 9519$
September 27, 1983 cc: JH Craven LMQ:83-101 *JS Dietrich CR Dietz
*J'd a Gitnick RA Hanvelt
*RT Hill RG Matthews
*JP Rea
*AF Wenger
*EB Wilson
*with attachments Mr. J. D. Martin Fuel Department CAROLINA POWER & LIGHT COMPANY P. O. Box 1551 Raleigh, NC 27602
SUBJECT:
- Additional Brunswick 2 Cycle 5 Confirmatory Load Line Limit Analyses Information
REFERENCES:
- 1) " Load Line Limit Analysis Reverifi' cation for Brunswick Steam Electric Plant Unit 2, Reload 4 (Without Recircu-lation Pump Trip)", NED0-22088, July 1983.
- 2) " Load Line Limit Analysis Reverification for Brunswick Steam Electric Plant Unit 2. Reload 4 (With Recircula-tion Pump Trip)", NED0-22228, July 1983.
Dear Jack:
. Draf t copies of the referenced documents were niven to CP&L for review on August 9, 1983. These documents report the results of analyses perfonned by GE to confirm the applicability of the original ~
Brunswick 2 load line Ifmit analysis (LLLA) to Cycle 5. Attached for your information are the GETAB transient analysis initial input condi-tion parameters for the events analyzed to support operation above the 100P/100F load line in Cycle 5. Attached as well are eight figures with ~ the transient plots for those events. Please let me know if CP&L has any questions or comments about the referenced documents or the attached information. We are delaying final printing of the Cycle 5 LLLA documents pending conclusion of CP&L's review. Very truly yo rs,
.u k i. - '
,g L. M. Quintana Fuel Project Manager Brunswick 1/2 H/C 174; (408) 925-2026 rem Attach.
. . . _ _ _ . _ _ _ . _ . . _ . . _ _ _ _ _ _ _ , , . _ ~ _ _ _ -
h Attcchrcnt 14-8 ' Brunswick 2 Cycle 5 GETAB Transient Analysis Initial Condition Parameters Peaking Factors Bundle Bundle Fuel (Local, Radial, Power low Initial Design Exposure Axial) R-Factor 1MWt)_ (10{lb/hr) MCPR 100% Power - 94% Flow - with RPT (Events Analyzed: TTNBP, Fb'CF, LFLH, Inadvertent ilPC1) P8x8R EOCS 1.20,1.54,1.40 1.051 6.536 105.4 1.22 8x8R EOC5 1.20,1.56,1.40 1.051 6.628 104.2 1.20 8x8 EOC5 1.22,1.42,1.40 1.098 6.041 103.9 1.20 7x7 EOCS 1.24,1.28,1.40 1.100 5.446 116.6 1.19 100% Power - 94% Flow - without RPT (Events Analyzed: LRNBP, FWCF, LFLT, Inadvertent HPCI) P8x8R EOCS 1.20,1.44,1.40 1.051 6.113 107.9 1.31 8x0R EOC5 1.20,1.46,1.40 1.051 6.229 106.5 1.28 8x8 EOCS 1.22,1.34,1.40 1.098 5.714 106.0 1.27 7x7 EOCS 1.24,1.25,1.40 1.100 5.309 116.7 1.22 l 85% Power - 61% Flow (Events Analyzed: LFWH, Inadvertent HPCI) l P8x8R EOCS 1.20,1.64,1.40 1.051 5.905 67.53 1.18 l 8x8R EOCS 1.20,1.64,1.40 1.051 5.902 67.2 1.18 l 8x8 EOC5 1.22,1.50,1.40 1.098 5.410 66.8 1.37 ! 7x7 EOC5 1.24,1.36,1.40 1.100 4.928 74.5 1.16 1 M**.W :.* o. a l l t
CPR COMPARISON GCXL QUAunCATION 1.35 - 1.54 - 1.33 - i 1.32 -- 1.31 - %o 1.30 - o 1.29 - +o
' 1.25 - +
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. _ _1.31 _ _ . _ _ .
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Question 15 Discuss results. the sensitivity of the axial node sizes on the the rmal-hydraulic The Topical' Report presented results using one channel to represent one fuel bundle. fuel bundles) are If " collapsed," channels (one channel representing several intended for future analysis, discuss hcw it will be approached and the sensitivities on hot bundle parameters.
Response
Sensitivity studies were conducted to determine the effects of changing the axial node size and the number of fuel regions on hot bundle parameters. The 75-channel, high-power, high-flow case from Brunswick Unit 1, Cycle 3 described in the topical (Figure SC) was used as the base condition for these studies. The 75-chancel case depicts an eighth-core symmetric model, each channel representing which representeight assemblies, with the exception of those along the diagonal, four assemblics. This model is useful for evaluating the effects run. of radial peaking factors in adjacent channels, but can be expensive to The preferred to common geometrymethod for core-wide and orifice analysis in to lump channels according type. In doing represented as three channels: this, the 75-channel case can be peripheral 8x8. central 8x8, central 8x8 retrofit, and for hot channel work.A fourth channel depicting a single hot bundle can be added Carolina Power & Light This compressed core representation will be used by core-wide and hot Company for most hydraulic applications and produces model as shown in Table 15-A. channel results very similar to those of the eighth core A of compressed changing the core model axial nodewith a single hot channel was used to study the effects size. A 24-node axial power shape was reduced to twelve, eight, and nodes, respectively. six nodes by averaging over two nodes, three nodes, and four nodes; and the A FIBWR model was run, varying only the number of axial results of these executions are shown on Table 15-B. The hydraulic 0.8 percent parameters change in appeared the hot very insensitive to the axial node size with a char.nel void fraction and only a 0.2 percent change in the core pressure drop and the minimun critical power ratio. CP&L will use a 24-node power shape for most FIBWR applications to remain consistent with the neutronic codes. A more detailed study was performed on the effcets of collapsing chaanels together. Four cases examined are described below: Case 1 - An eighth-core symmetric model of Brunswick Unit 1, Cycle 3 with the hot channels representing four assemblies. Case 2 - A four-channel compressed model of Brunswick Unit 1, Cycle 3 with a single assembly as the hot channel. Case 3
-A two-channel compressed model of a full-core assemblies. of 8x8 retrofit A hot channel represents 280 assemblics.
CLase 4 - A two-channel modt. of a full-core of 6x8 retrofit assemblies. A single into one assembly channel. is the hot channel, and all remaining assemblics are conpressed l
?.
In all four cases, the core pressure drop and the radial power of the hot channel was fixed and the code solved for core flow and hot channe The results shown.on Table 15-C indicate that the hot bundle parameters are defined primarily by the power produced in that channel and are independent of how many channels are represented or how the remainder of the core is modelled. 9
-%=.--
I e 126
TABLE 15-A HOT CHANNEL NOT -
# OF PEAKING CORE FLOW CHANNELS FACTOR P (N1b/hr) MCPR 73 1.3952 20.7577 116.58 1.3502
- 4 1.3952 20.7472 11G,42 1.3498 TABLE 15-8 TOTAL
# OF CORE NOT VOID BYPASS NODES P FRACTION FRACTION MCPR - -
24 20.7472 0.4495 0.1130 1.3498 12 20.7443 0.4482 0.1131 1.3496 8 20.7512 0.4463 0.1132 1.3519 6 20.7800 0.4459 0.1133 1.3528 TABLE 15-C HOT HOT CHANNEL PEAKING CORE FLOW CASE FACTOR P (M1b/hr) MCPR 1 1.3952 20.7577 116.58 1.3502 2 1.3952 20.7577 116.44 1.3498 3 1.3952 20.7577 116.47 1.3499 4 1.3952 20.7577 116.47 1.3499 6 O
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