ML20147F518
| ML20147F518 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/30/1978 |
| From: | Cheng H, Diamond D BROOKHAVEN NATIONAL LABORATORY |
| To: | Fieno D Office of Nuclear Reactor Regulation |
| References | |
| BNL-NUREG-24903, NUDOCS 7810200010 | |
| Download: ML20147F518 (51) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ - _ , '. O INTERIM REPORT Accession flo. Contract Program or Project
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Reactor Core Safety Analysis Group Subject of this Document: Core Analysis of Peach Bottom-Erurbine Trip Tests Type of Document: Infonnal Report Author (s): Hsiang-Shou Cheng and David J. Diamond Date of Document: September,1978 Responsible NRC Individual Mr. Daniel Fieno 7 and NRC Office or Division: Core Performance Branch Div. of Systems Safety U.S. Nuclear Regulatory Comm. Washington, D.C. 20555 This document was prepared primarily for preliminary or internal , use. It has not received full review and approval. Since there ' may be substantive changes, this document should not be considered final. Brookhaven flational Laboratory Upton, NY 11973 Associated Universities, Inc. I for the U.S. Department of Energy Prepared for l U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Under Interagency Agreement EY-76-C-02-0016 HRC FIN !!o. A-3004 IllTERIM REPORT 7 5 / C, 7 C ' C Q l 0 -/t i L
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BNL-NUREG-24903 INFORMAL REPORT LIMITED DISTRIBUTION CORE ANALYSIS OF PEACH BOTTOM-2 TURBINE TRIP TESTS S Hsiang-Shou Cheng David J. Diamond Reactor Core Safety Analysis Group Thermal Reactor Safety Division Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 l i Manuscript Completed September 1978 Date Published September 1978 Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Contract No. EY-76-C-02-0016 FIN No. A-3004
1 4 4
^
NOTICE This document contains preliminary information and was prepared pri- . marily for interim use. Since it may be subject to revision or cor-rection and does not represent a final report, it should not be cited as reference without the expressed consent of the authors. 9
TABLE OF CONTENTS \ Page ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l*
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . 2
} 2.0 ANAUSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5 l i 2.1 Test Conditions . . . . . . . . . . . . . . . . . . . . . . 2.2 Reactor Model . . . . . . . . . . . . . . . . . . . . . . . 11 2.2.1 Core Representation. . . . . . . . . . . . . . . . . 11 2.2.2 Feedback Model . . . . . . . . . . . . . . . . . . . 11 l l 2.2.3 Delayed Neutron Data . . . . . . . . . . . . . . . . 17 l 2.3 Steady State Calculations (Initialization). . . . . . . . . 17 2.4 Time Dependent Core Inlet Conditions. . . . . . . . . . . . 20 3.0 DISCUSSION OF RESULTS. . . . . . . . . . . . . . . . . . . . . . 32 3.1 Neutronic Response. . . . . . . . . . . . . . . . . . . . . 32 3.1.1 Reactivity Behavior. . . . . . . . . . . . . . . . . 32 3.1.2 Power Behavior . . . . . . . . . . . . . . . . . . . 32 3.2 Thermal-Hydraulic Response. . . . . . . . . . . . . . . . . 52 3.3 S e n s i t i v i ty S tu d i e s . . . . . . . . . . . . . . . . . . . . 56
4.0 CONCLUSION
S AND RECOPMENDATIONS . . . . . . . . . . . . . . . . 67 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 l -i-
i ABSTRACT
. An analysis of recent BWR turbine trip experiments is presented. This was
- done as part of the NRC technical assistance program at BNL to further validate-e the methods being used to audit calculations submitted as part of the licensing l process. The analysis was performed by means of the two-dimensional (R,Z) core dynamics code BNL-TWIGL in conjunction with the system transient code RELAP-38.
Using the measured power history in_the core, RELAP-3B first calculates system thermal-hydraulics to provide the time-dependent core inlet conditions as input-to BNL-TWIGL. BNL-TWIGL then performs a space-time analysis of core neutronics and thermal-hydraulics with feedback. The calculated power history was then com-pared to the measured _ as the basis of qualific,ation. Good agreement was obtained, thus confirming the validity of the basic models of the two codes. t 4 l
. l
1.0 INTRODUCTION
l The analysis of turbine trip events for boiling water reactors (BWRs) is i an important part of the BWR licensing process. Theoretical analysis of this ; i
- type of transient using various computer codes has been widespread but has lacked validation from full scale tests on operating reactors. The recent
- turbine trip experiments performed at the Peach Bottom II reactor in April . ;
1977(1) can be used as benchmarks. The tests were performed near end-of-cycle 2 l with three different initial power, control rod pattern and flow conditions. ]- The reactor was not. scrammed due to the turbine trip (as it would be in normal 3 operation)', but was scrammed on a high-flux signal. These tests can be used not l } only to validate computer codes' currently used in BWR safety analyses, but also 1 to quantify the conservatism of previous calculations. 4 The object of the present work is to qualify the BNL capability to calcu-late a turbine trip.' event using the test data. This involves both the system code RELAP-3B(2) and the core dynamics code BNL-TWIGL.(3) Specifically, we 1 i wish to verify the system thermal-hydraulic model of RELAP-3B and the core l nuclear-thermal-hydraulic model of BNL-TWIGL. This report documents the portion l } of work involved with BNL-TWIGL; namely, the calculation of the core dynamic re-i sponse of the three turbine trip experiments. The calculations involving l < l i RELAP-3B are documented in Reference 4. I
- in order to facilitate the analysis, we adopted a calculational Scheme as illustrated in Fig. 1. Starting with the measured power history (powtr vs. tims) .
in the core as input, RELAP-3B calculates the system thermal-hydraulics, in- ! cluding that of the core. The turbine trip is the initiating event in this calculation. This provides the time-dependent core inlet conditions (i.e., pressure' vs. time, flow vs. time, and- temperature vs. time) as the boundary ? l
MEASURED (POWERvs,T ) y NO YES RELAP-3B : AGREE? : STOP u CORE INLET B.C. f . P ig vs. T W ig vs. T T in vs. T v BNL-TWIGL v [ ( CALCULATED POWER vs T
)y ~
4 l Figure 1 Calculational Scheme Employed in the Present Work I l L -- - - _ _ _ _ - - _ _ _ - - - - _ _ _ - - - - - - - - - - - - _ - - _ - - - - - - - . - -
i l l conditions to BNL-TWIGL, which then performs a space-time analysis of the core , l neutronics and thermal-hydraulics. The calculated power history in the core is then compared to the measured one as the basis of qualification for the two l codes. Good agreement was obtained for the power history as well as system pressure responses, thus confirming the validity of the transient model of BNL-1 TWIGL and RELAP-3B for calculating turbine trip events. , J l An alternative way of doing the analysis would have been to use an arbitrary ! (but reasonable) power history as the initial input to RELAP-3B, The calcula- ; tional scheme would then be expected to converge to the measured power history. Since the mea sured power was already available, it was more efficient to proceed as outlined aLove. Section 2 provides a brief description of the test conditions, and also presents the model of the experiments set up with BNL-TWIGL. The results of the calculation and their comparison with test data are tresented in Section 3. Important conclusions and recommendations are given in Section 4. 1 1 l 1 I
. I I
1 l l
k 2.0 ANALYSIS-2.1 Test Conditions The turbine trip 'ransient tests (1) were carried out by manually tripping the turbine through turbine stop valve closure. In order to obtain significant core neutron and heat flux. response, the anticipatory scram from turbine stop valve closure was disabled and the high flux-scram set point was adjusted to ) protect the core. Table I lists the initial power and flow conditions, as well l as the flux scram set points for the three tests. Control rods were used to achieve a critical reactor with a desired power distribution prior to the turbine trip. Figures 2, 3 and 4 show the initial ,
)
control rod patterns for the three turbine trip tests, respectively. For each j test quite a few control rods were involved, making it important to properly l represent the control rod distribution in the analysis. , For turbine trip transients, the axial power distribution is of interest j because of its shape change during the transient. The radial power distribution is of secondary importance,'since it changes little 'during a turbine trip.(5) The initial core average axial power profiles are derived from process computer output of in-core detector signals. They are shown in Section 2.3, where the calculational initialization is discussed. An important aspect of the turbine trip tests is reactor scram due to the high-flux set point. Fig. 5 shows the measured control rod insertion behavior
. as a function of time after the trip signal. Note that there is some spread in the scram speed among various control rods. The solid line is the average rod insertion behavior and was used in the calculation. Note also that there is a delay of 0.15 sec.to 0.25 sec prior to actual rod motion.
This scram delay time is an important parameter for any over-power transient such as the turbine trip. , Table I Peach Bottom-2 Turbine Trip Test Conditions Core Flow in cram Core Power Sett{ing 6 (MWT) (% Rated) (10 lbs/hr) (% Rated) (% Rated) Turbine Trip 1 1562 47.4 101.3 100.3 85 Turbine Trip 2 2030 61.6 82.9 82.1 95 Turbine Trip 3 2275 69.1 101.9 100.9 77 O e
.. . . . . . . - _ . . . _ - - - _ - a
t 59 55 - 10 20 10 51 48 48 48 48
. 47 8 28 32 28 8 43 48 48 48 48 48 48 39 10 28 8 8 28 10 35 48 48 48 48 48 48 31 20 32 8 32 20 27 48 48 48 48 48 48 23 10 28 8 8 28 10 19 48 48 48 48 48 48 15 8 28 32 28 8 I 11 48 48 48 48 7 10 20 10 3
2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 n ROD IN ROD OUT n INDICATES THAT THE CONTROL ROD IS INSERTED n NOTCHES FROM THE BOTTOM (FULLY INSERTED: n =48) Figure 2 i Initial Control Rod Pattern for
- Turbine Trip Test 1 4
f 4 59 55 14 12 14 - 51 48 22 22 48 47 8 12 16 12 8 l 43 48 22 44 44 22 48 ! 39 14 12 12 14 35 22 44 16 16 44 22 i 31 12 16 16 12 : 27 22 44 16 16 4
'4 22 23 14 12 12 14 19 48 22 44 44 22 48 l
15 8 12 16 12 8
- II 48 22 22 48 7 14 12 14
- 3 ,
. 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 i n - R OD IN ROD OUT n-INDICATES THAT THE CONTROL ROD IS INSERTED n NOTCHES FROM THE BOTTOM (FULLY INSERTED: n =48) i Figure 3 Initial Control Rod Pattern for Turbine Trip Test 2
I 59 55 10 12 10 51 14 40 38 14 47 8 8 8 8 8 43 14 36 48 48 36 14 39- 10 8 ^ 8 10 35 38 48 16 16 48 40 31 12 8 8 12 27 40 48 16 16 48 38 23 10 8 8 10 19 36 14 48 48 36 14 15 8 8 8 8 8 ll 14 38 40 14 7 - 10 12 10 3 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 n
~
R O D IN ROD OUT n INDICATES THAT THE CONTROL ROD IS INSERTED n NOTCHES FROM THE BOTTOM (FULLY INSERTED:n =4 Figure 4 u e r p Te g.
I I I I I I I TYPICAL RODS NEAR THE AVERAGE SPEED l.O -
//
/u'/ -
z O / /, ,// // / to 0.8 - FASTEST ROD ////k
//// SLOWEST ROD -
W /// Z //// a O
// //
/ / //
T O.6 w
,/
////
/ /
/
k' O / z /,/ /
,/
o / / p O.4 '/ / / -
//
o // / 4 // ,' T // / --- OBSERVED ROD MOTION L'- // / O.2 -
//
f
/ ROD MOTION ASSUMEDIN -
,,/4,/ CALCULATION
/
/
' I I
' I I I I O.0 1 O.0 0.5 1.0 I.5 2.0 2.5 3.0 3.5 4.0 TIME (s) .
Figure 5 Measured Control Rod Insertion , . Characteristics
l 2.2 Reactor Model 2.2.1 Core Representation: The calculational model mocks up the reactor core with axial and radial reflectors in (R,Z) geometry, as shown i in Fig. 6. The finite cylinder geometry is a good approximation, since the control rod patterns involved are all nearly symmetric. l There are five radial scram zones within the active core to repre-l , sent the initial control rod pattern and the subsequent scram bank simula-tion. The entire reactor, including the reflectors, is partitioned into 11 radial channels and 26 axial planes for thermal-hydraulic representation, and 36 radial mesh points and 61 axial mesh points for the neutronics. The axial reflectors are assumed to be a mixture of 60% stainless steel and 40% water. The water of the bottom reflector is at a hot oper-ating condition with no steam voids, while that of the top reflector is assumed to consist of 70% steam voids. The radial reflector is pure water at hot operating condition without voids. The piecewise control density distribution for each of the 5 radial scram zones is presented in Figs. 7, 8 and 9, respectively, for the three turbine trip tests. These distributions were calculated from the actual control rod patterns (see Figs. 2, 3 and 4) by simple volume weighting. 2.2.2 Feedback Podel: Thermal-hydraulic feedback and reactor scram l l are of great importance for predicting the transient behavior of a turbine trip. The feedback model in BNL-TWIGL takes into account the effect of steam voids, moderator temperature, fuel temperature (Doppler) and control rods on the two-group cross sections for each material. Specifically, the following expressions are used:
l l 1 1
)
v I 4 6 8 9 14 18 23 27 30 34 36 +-- R ADIAL MESH TOP ', l' , 61
, i i _
~
14 1 14 14 14 1 14 14 1 14 14 : 14 14 (4
~
8 8 8 8 8 8 8 8 11 Il 12 5
, 8 8 8 i 8 8 8 8 8 11 11 12
- 51 8 8 8 8 8 8 8 8 12 il 11
- 4g 8 8 8 8 8 8 8 8 il Il 12 - p 7 7 7 7 7 7 7 7 il il 12 F 45 7 7 7 7 7 7 7 7 Il ll 12 i
h 43
~
7 7 7 7 7 7 7 7 il il 12
- 41 6 6 6 6 6 6 6 6 11 11 12 h 39
~
6 6 6 6 6 6 6 6 ll 11 12 37 6 6 6 6 6 6 6 6 11 11 12 -
- 35 i 5 5 5 5 5 5 5 5 10 10 12 -
33 5 5 5 5 5 5 5 5 10 10 12
~
4 4 4 4 4 4 4 4 10 10 12 - 3I 4 4 4 4 4 4 4 4 10 10 12 - ' 4 4 4 4 4 4 4 4 10 10 12 - 4 4 4 4 4 4 4 4 10 10 12 F 3 3 3 3 3 3 3 3 9 9 12 L
; 21
, 3 3 3 3 3 3 3 3 9 9 12 r g 3 3 3 3 3 3 3 3 9 9 12 -
2 2 2 2 2 2 2 2 9 9 12 - 2 2 2 2 2 2 2 2 9 9 12 - 13 2 2 2 2 2 2 2 2 9 9 12 -
~ '
l i I I I I I I 9 9 12 - I I I I I I I I 9 9 12 - i i - 12 ; 13 13 13 13 13 13 - 13 3 13 l' 13 13 i l 80T70M R t . AXIAL M ES H Figure 6 The (R,Z) Reactor Model for BNL-TWIGL
-. . . . - . . . . . - . . - - . . ...~. -
t CONTROL FRACTION CONTROL FRACTION 9ppp p .- p -p p .O P .- l o m# mm o om A m- e o ' 1 o , i . . . o . . . . o moOm wom om 4 om ^Xm b mmb - - CONTROL FRACTION' g I p ,o ,o p p - Om - sgm - i g -o m A m m o No oI a o , zH Im - zam - m .,
, i i i m
o - 6 m8m _ mg5 -
-=6 -
wa om-
;; r. , m m A -
m m-
. "Q g . >x -
Ch "E mm -
' & $ *, N 5 CONTROL FRACTION CONTROL FRACTION f}
-c
-zim m-9 9 p P oma m 9 .-
m o p poPP-om bm m o
= mE5 -
o , i i . i o . . . . i N - - mhNm om m ma m mA - w S Imm KI N bO
- ~
N5 o rg j OO zI g _ z m u mH ag5 - Ng5 - m m
, , e . - - -- _-_ - __ ~v.- ,_ _ . , . _ ,
l CONTROL FRACTION CONTROL FRACTION ooooo- ooooo- ' o io b-6 6'o 6 iu b b6 o o i i i i- i o i i i i .
-o moo m wom gm om %m 4 xm4 - -
CONTROL FRACTION gIDo $% m5m 9 9 .o 9 p .- p om A m m o yo or g o i . . i i m zym - gym-o p; u35
^
-- 0 5 -
?r a m8m ox
_I xm4 N- N-
*Q",P r I K mm ; 'h 4 -
CONTROL FRACTION CONTROL FRACTION Pa ~
- N5
#A "E
~
ojm z m-pp,o,op-o m a m m o p.o p p p .- oN A mm o o o my mE5 - i i i i i i e i i i o s- Wo om - I m8m ox
' om j x m3 y
x mA - D 7 gI 4 S mm - N O ~ N5 0 0 o yg zI m - mH m a-c bSD -
^
S-
~
II)iil l! lljJ)lj' N-N O
.o .
Oo. I I T T pm i C Pm C A A R o,m i
. RF Pm F
L O Oh
. L O ob i
R R T pm , T PN i N N O ' O - - _ - - C Po C po mg aN oN A m m _O w om3 oom* Esy -~ 8Mm rDQIH s_ mg%$ Noy _ m ypr mOgm m N N ,o i O .o . O I I T T C Pm . C pm~ . A A R R F pm F o6 i L L O ob . O pA i l R R T pN T N Pm i N O O - - C Po - - - - ~ C po - - omA m m _o w omA mg sN oo mIOo% O~ 8mmxm57as-mQg N S mom"E N0z" _ N6, O I _ T C o6 A R Fo6 , L O ob , R T u N oi O C o'o 'J- - - - - oN A mm5M nomm Im5y _G mox>KNozmo
,24* e S' y o8 ;" Q o 1 a" 7 s U32* T4 #n w
,lj,II1[il-l lll ,
BIj(M) Ij(M,4,T')"I$(M>")+ m g
- I'k - k R) f (1)
BIj(M)
+
BT m
* (Tm-TmR) .
where 2 Ij(M,a)=ay(M)+bj(M)*a+cj(M)*a. (2) In the above expressions, the superscript g is the group index, the sub-l script j denotes the cross section type (diffusion, absorption, fission, etc.), and M represents the material composition. flote that we have se-lected the reference void fraction to be zero. This is immaterial so long as the relevant coefficients are consistently defined. The parameters a, b, and c used to obtain the void dependence in , e Eq. (2) are input for both controlled-and uncontrolled situations. There-fore, the resulting Ij from Eq. (1) is for cither controlled or uncontrolled composition. These data are then used in Eq. (3) to obtain cross sections at every point in the reactor that represent the proper control density as well as thermal-hydraulic feedback: uc I j(t) = Euc + f3 (t) Ij - Z j (3) s where the superscripts "uc" and "c" represent the uncontrolled and controlled data, respectively, and f (t) is the time-dependent control density at the 3 location of interest simulating the control rod movement due to reactor scram. The superscript "g" has been dropped for brevity. 2
.- - .. - ~ . -. - ..
i f 2.2.3 Delayed Neutron Data: The transient calculation calls for i an appropriate set of delayed' neutron data representative of the E0C 2
' condition. Table II lists the delayed neutron data for the six precursor groups used in the present analysis (6) which is'a weighted set for the fissile U-235 and fertile U-238. This.is a good approximation because l
the uranium isotopes still dominate the reactor composition at EOC. How-ever, the effect of plu'tonium buildup due'to burnup is taken into account j by normalizing the S $
's such that the total 8 = 64 is equal to the E0C B (0.00546) as reported by GE.(7)
{_ 2.3 Steady-State Calculations'(Initialization) 1 l Since the transient response'of the turbine trip depends on the initial con-ditions, it is essential to obtain the same initial conditions as the actual i j test, especially the axial power distribution. This was done by means of an . initialization procedure illustrated in Fig. 10. i The main unknowns for BNL-TWIGL are the 2-group cross sections and their feedback coefficients corresponding to the core conditions at the time of the tests. . the core was close to the end of cycle 2 (E0C 2) with'an average ex-posure of about 15,000 F'WD/T. Since the exposure distribution in the core was not known by us, we employed an iterative procedure to get at the exposure distribution, i and hence the cross section data. It starts with an initial guess on the ex- ', posure distribution.in 8 axial and 2 radial zones. This permits determination of-the 2-group cross section data for the core utilizing the exposure dependence ,
- of the cross sections calculated with HAMMER (8) for a typical fuel rod.* A l
.The actinide concentratic HAMMERwereprovidedbyGEbg)ateachexposurewhichwereneededasinputto for a BWR/4.
4
- -v -- v ,
Table II Delayed Neutron Data 4 Delayed Group Yield Fraction Decay Constant i 8 4 Aj (sec-I) 1 0.000207 0.0127 2 0.001163 0.0317 3 0.001027 0.1150 l . 4 0.002222 0.3110 ) l . 5 0.000699 1.4000 i 6 0.000142 3.8700 Total: 0.005460 t e l 4 l : I i
INITIAL GUESS EXPOSURE DISTRIBUTION
-y ADJUST TYPICAL FUEL R0D : EXPOSURE Z = I(E)- DISTRIBUTION
\ A y
CORE ! 2-GROUP I's FEEDBACK COEF. CONTROL BNL-TWIGL < R0D ! PATTERN v CALCULATED l AXIAL POWER PROFILE l l l r ' STOP \ YES AGREE? - NO MEASURED AX1AL POWER PROFILE ; l Figure 10 i Initialization Procedure for Obtaining I
- j. Initial Conditions
steady-state calculation by'BNL-TWIGL yields the core-average arial power
-profile which is compared to the measured profile. If the agreement is not satisfactory, the exposure distribution is readjusted based on the observed discrepancy in the axial power shape. The new exposure distribution is then used to revise the cross section set. This procedure is repeated until an ,
i acceptable agreement in the axial power profile is obtained. , l l The turbine trip test #3 was selected for obtaining the cross section ~ set representing the core at the time of the experiment via the initialization i procedure. This same set was then used to calculate the initial conditions of 1 the other two turbine trip tests without any further adjustment. Table III ] lists the initial core inlet conditions used in these calculations. Figs. 11, l
- 12 and 13 show the comparison of the calculated and measured core-average axial ;
- i
, power profiles for the three tests, respectively. The overall agreement is sufficiently good to ensure the correct initial conditions for subsequent tran- l sient calculations l 2.4 Time-Decendent Core Inlet Conditions The time-dependent core inlet conditions are input tc BNL-TWIGL by means of table functions (i.e. , inlet boundary conditions vs. time) using a 41-point re-presentation. These transient boundary conditions were obtained from RELAP-3B . calculations of system thermal-hydraulics, as illustrated in Fig. 1. Because of the importance of the transient pressure response, the calculated . core exit pressure responses are plotted against the test data and shown in Figs. 14 15 and 16 for the three tests, respectively. The good agreement indicates the adequacy of.the RELAP-3B model. 4
TABLE III Initial Steady-State Core Inlet Conditions Pressure Inlet Flow Inlet Temperature (psia) (lb/ft -sec) 2 (OF) Turbine Trip 1 1011.21 313.66 533.1 Turbine Trip 2 994.12 256.69 526.1 Turbine Trip 3 1008.00 315.52 529.2 l 1
l l Peer h Sottom 2 Turbine Trip Test L. Initial Power L..-- , i LEGEND 1 B BNL-TWlGl. / A N l t l 4 14 " D = Tes t Dat a / -' a4 x , 10 a /' ' e "81 { / { ,..[
!m o.l f/ t .
g/
-/ \p
\ \
3 j 02- / j l 00 . CD && S.2 a3 04 09 as 07 Os 04 10 J Fractional Core Hetsht 1 i Fig.11 Initial Core Average Axial Power Profile of Turbine Trip 1 i l Peach Bottom 2 Turbine Trip feet 2. Initial Power i to - a LECIND B = BNL-TWlGL i.4 D= Test Data .g 12 [ '... l f . is-e .. U l , j Os.
"a, z
as - D t'
- l 4 64-f as K . ! l 00 as om na 64 as os a7 as as t.0 j Fractional Core Height 1
) Fig. 12 - Initial Core Average Axial Power Profile of Turbine Trip 2 Peach Bottom 2 Turbine Trip Test 1 Initial Power
^
le LEG EN D l B = BNL-TWIGL i D= Test Data j
, < h, 5
/. I g 10
- f \
- 6. I pi ,
g as , / \; 5
- t l
"] D[ -
\! '
! 6
'*i / f b
i 02 01, al 03 03 64 05 a' 47 08 60 10 Fractional Core Height Fig. 13 - Initial Core Average Axial Power Profile of Turbine TriD 3 1
1060 i i i i j i l
'\
I I
/
1050 - I
/
TEST DATA l \ I g
] 0%>
1040 oooo RELAP-3B l0- 0l
/l g l l '
7 II
^ I 3 1030
{l f j E. / V w
/ \l g /
IlV g1020 - jo - 0 'f e 1010 - o-o m-o-o-o-o- # lf lo - 1000 - 990 - I l l l l
. O.0 0.2 0.4 0.6 0.8 1.0 1.2 TIME (s)
Figure 14 Transient Core Exit Pressure Response, Turbine Trip 1
._._._7_ ,
W 1060 i i i i ii i i i i i 1050 - ---- TEST DATA n - II OO d)
~
oooo RELAP-38 II o o hl Id 2 1030 R\["l
- - o l \
b~ I\ l \ N o 1020 -
\\v \lif l lI -
u) v \ (' o \l\U . E 1010 - Io
/o -
/
/8 -
le 1000 - jo -
- lo -
990< mo-o +o-o-<d i I i I iI i I i l i
^
O.0 0.2 0.4 0.6 0.8 1.0 1.2 TIME (s) Figure 15 Transient Core Exit Pressure Response, Turbine Trip 2
1070 i i i i i
'\ 10 I
;\ \$ oo c
1060 - I O - ll O TEST DATA { g \q 1050 - oooo RELAP-38 l $3 I l \l I I I Io I)Il t v ll \ I E 1040
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- I Il N I
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1010 - lo - U o-o-o-o-om-o-op_O / 1000 - - I 990 I I I I O.0 0.2 0.4 0.6 0.8 1.0 1.2 TIME (s) Figure 16 , Transient Core Exit Pressure Response, Turbine Trip 3 Inlet flow also varies with time during the turbine trip events. In gen-eral, the variation is within 6% of the initial value. Figures 17, 18 and 19 present the transient inlet flow responses of the three tests, respectively. Inlet temperature varies by less than 1 0F during the transient and may be assumed to be constant for practical purposes. The variation of inlet temper-i ature was taken into account in the present analysis, nevertheless. . 3 h 4 4 e a - a 4 .a4 _ 44 4 .%., e y - v -- 4 d f J e O. L 1. I li i i ro
\,
p e 4 i, D s i O E i 1 N 8. M 1
' 0 e=4
' Cf. 4 C. - 3 .- e C 1-3
@ ) [ e a 1
! c.
= -n C7) L. <
i -
*e .C. 3 l 8
h LL-s H l, C%
- O I i
m ! C l a
.I en
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. O O O O O O OO
- CD W v N O CD (O ro - to to r0 to N N
- (gu s/ql) Mol.d J.37NI 2
1
3IO i i i i i 290 - - c7 C ' 4 270 - - v e 3 l o_J 250 - - i E$ i l-
$ 230 - -
a 1 210 - - I I I 190 I I O.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (s) Figure 18 ; Transient Inlet Flow Response of Turbine Trip 2 b- -
---"--------A _ _ _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ _
- i i)i1 0
- - _ - 3
. - i '
5 2 e s i ' 0 n o _ 2 p s e3 R _ p _=- . wi
) or -
. ( s 9 1 lT F e E e r t n ei i I 5 M 1 I i u g l b I u nr T F t T _ nf i eo . s n a r i I 0 T 1 i I 5 _ O .
- - - _ - O.
0 8 0 6 0 0 0 0 0O
. 4 2 0 8 6
. 3 3 3 3 3 2 2 AN C.<e oJIL WoaZ.- t N.
j i v 3.0 DISCUSSION OF RESULTS 3.1' Neutronic Response
- 3.1.1. Reactivity Behavior: The overpressurization caused by the
- turbine trip collapses steam voids, and thereby introduces' positive re- ,
j activity in the core because of the negative void feedback. In short, i the void reactivity is the driving force of the transient. This is illus-i trated in Figs. 20, 21 and 22 for the three turbine trip tests, respect-i ively. t l The dominant role of the void reactivity is clearly seen. It is re-sponsible for the initial power excursion until reactor scram is activated. i .The negative scram reactivity then turns over the power burst and eventually l l quenches the excursion. The moderator temperature feedback is negative and 4 l i significant throughout the transient. Its neglect will lead to overpre-4 ,
- diction of the power peak. The Doppler reactivity is negligible throughout the transient. In summary, a correct calculation of the void.and' scram
- f. ,
_ reactivity is the key to a successful prediction of the transient response i of a turbine trip. 3.'1.2 Power Behavior: The transient behavior of the total nuclear power in the core is shown in Figs. 23, 24 and 25 for the three turbine trip I tests, respectively, along with the measured ones for comparison. The over-all agreement is quite good. The notable discrepancy for all three cases ,
.is the overprediction of power around 1 sec owing to the overestimat.e of core pressure by the system' transient code RELAP-3B (see Figs. 14, 15 and -
16). Another noticeable discrepancy is the rising part of the power burst for_the' turbine trip tests 1 and 2. The calculated power exhibits a slower l 1 1 l
- n. - , , - -
Peach Bottom 2, Turbine Trip, Test 1 i I LI9I S T= Tot al om
- D= Doppler ,/
f 'bs.'. 5 = Scra m /
"*' V. Vo gd '
- M= Mod. Te m p. \
2 f s
,/ \
} o.aur a N7 o
I N oooo = , . . . . ..o N..... p w N y aa. c ..g _u
- N x
i g N ,s 7 i
.o. N N
-acus .
)
** I oo e 6. u. u a ia Time (s)
Fig. 20 - Transient Reactivity Behavior of Turbine Trip 1 { Peach Bottom 2, Turbine Trip, Test 2 4 000 - T= Total D= Doppler ,/ 7 - Ql r ' S = Scra m / ooco - V= Vold * { . M= Mod. Tem p. s 1 3 '
$ aa -
- . .. x l x N 3--
N f f "' ss r
=
- N
-com .
N N N
-"" ao n .. .. a. a b Time (s)
Fig. 21 - Transient Reactivity Behavior of Turbine Trio 2 Peach Bottom 2, Turbine Trip, Test 3 LE91h.D \ To Total D= Dopple r nom
- S = Scra m '[-%v V= Void M= Mod. Tem p. *
/
N nous - , i
$ I
=
x o-j ND.s....
' c .. -g __h ._ ._.o
..I y
l l N N s
\y I
! x i ,
-oam d
's c j f 5
-acti ao 0.3 04 08 08 IS IS Timets)
Fig. 22 - Transient Reactivity Behavior of Turbine Trip 3
l l l l l Peach Bottom 2. Turbine Trip. Test t t%D y LE91 8 ' l B BNL-TwlGu2* 1 DMH) l D = Tes t Data - auno J $ l t
=z l \
% amoa y i l llBa
- e 's .
ma -
.\x. c6
- l u u a< u is in u Time (s)
Fig. 23 - Total Power Behavior of Turbine Trip Test 1 Peach Bottom 2. Turbine Trip. Test 2 ano - LEGI S D B.BNL-TalGUM DMH) D= Test Data p 1 E c eaa. t a
$ tonoa
%D.
'l 60 -
00 02 64 M 98 La J Time (s) Fig. 24 - Total Power Behavior of Turbine Trip Test 2 Peach Bottom 2, Turbine Trip. Test 3 4W10 . l Lf.GIh.D N' i
- 9. BNL-rwlGUM DMH)
D = Test Data EXLO
- C J
- i i mo m
t e1 j D ;
'- '"'i / \ l
- /
\ i j s ;
m01 x'
,1 ,
ao u e4 a. u in a Ti m e(s ) Fig. 25 Total Power Behavior of Turbine Tri: Test 3 j l l l
1 i rise than the test data. This is probably attributable to the slcwer rise of core pressure calculated.by RELAP-3B (see Figs.14 and 15). q The power behavi'or is sensitive to the assumed value of the direct
. moderator heating'(DMH) and scram delay time (SDT), as will be discussed l
1 later in Section 3.3. The results presented in this section were ob- < t tained with a DMH of 2% for all three cases, and a SDT of 0.17 sec for
- Tests 1 and 2 and 0.20'sec for Test 3. This disparity in SDT is still
- not_ resolved at present. Sensitivity calculations by varying the initial
[ j rate of change in core pressure clearly indicate that the correct SDT for
- Test 3 is about 0.20 sec. However, similar sensitivity calculations in-i dicate that Tests 1 and 2 call for a SDT of 0.17 see to match the test i data. It seems logical to assume that the SDT would be subject to a sta-l tistical random variation of some sort, which may be a function of time.
I j In short, the SDT may vary slightly within a small spread (statistical i uncertainty) as a function of time. l } A more stringent test for the calculational model is the comparison of l power behavior at different axial levels in the core. This is shown in
- Figs. 26 through 29 for Test 1; Figs. 30 through 33 for Test 2, and Figs.
q 34 through 37 for Test 3, respectively, at four different levels (A, B, C and D) in the axial direction. Levels A, B, C and D are 1.5 ft., 4.5 ft., 7.5 ft., and 10.5 ft. from the bottom of the active core, respectively.
. Both the calculated and measured curves are radially averaged power histories at each level. Each of the curves shown was obtained by inte-
~
~
grating over all radial region power traces. Again, the agreement is satisfactory, further confirming the validity of the calculational model. 1 In all three tests the BNL calculations consistently overpredict the power peak in the bottom section of the core and underpredict the peak in ! Peath Bottom 2 Turbine Trip Test 1 - A Level Peach Bottorn 2 Turbine Trip Test I - Il 1.crel 4%9 WEG-1ECEND IECEND
.m . H = BNL-TWIGL p D = BNL-TWIGL D = Test Data D = Test Dat a ,
7m0- .O D n ;- n > 2 uO - J q sono -
- I 3 b
- 500-E 5
tt 300 0 , t* u amo- u e e 5 k g sun- g 200e-30n.0 l} no00 : TO - @ 5 00 , , , , . 00 , , , , , Os 02 04 Os as as t2 00 02 as Os os to a2 Time (s) Time (s) Figure 26 Figure 27
, Power Behavior at Level A Power Behavior at Level B y of Turbine Trip Test 1 of Turbine Trip Test 1 Peach 13ottom 2 Tut bine Trip Test 1 - C Level Peach flottom 2 Turbine Trip Test 1 - D level amo- arc o - --
LEG END LEGEN D H = HNL-TWIGL B = DNL-TWIGL O. D= Test Data . D = Test Dat s ( 4000- 4000 - E E E E M wo - M woo-t e alt:0 - 2000 - Em0 l4 g 30D0 I ~3 a a 00 . . , , , 00 , , , , , 00 02 04 08 OS AD B2 00 02 04 Os Os ID 12 Time (s) Time (s) Figure 28 Figure 29 Power Behavior at Level C Power Behavior at Level' D of Turbine Trip Test 1 of Turbine Trip Test 1,
Peach llottorn 2 Turbine Trip Test 2 - A level Peach Bottom 2 Turbine Trip Test 2 -- B 1.evel nuG- - soon LW[ND LEGEND f N~ B = BNL-TWIGL D = Test Data f jj-B a BNL-TWIGL D = Test Dat a f
.Li(
[ 4000-
'RO S - [
^ [ m W %
S E N- E s 3 0_ / g amo. g h iwo- h "~ ' E E 30t.0 14 son 0 :? sno .
....n 3 Do , , , , ,
~
3 00 , , , , , 00 02 04 OS 08 to 12 04 02 04 08 08 to 42 Time (s) Time (s) Figure 30 Figure 31 i . M' Power Behavior at Level A Power Behavior at Level B of . Turbine Trip Test 2 of Turbine Trip Test 2 Pearli flottorn 2 Turbine Trip Test 2 - C Level Peach Bottom 2 Turbine Trip Test 2 - D level s000 .- ac0-LEGEND :D. LG END H = HNL-TWIGL ! i B = HNL-TWIGL D = Tes t Dat a .. D= Test Data .. o 4t0 0 -
= eg
=eg 4004 -
3:10-E S { { 320-ano- . E ,. E "^
/. r.
ateo i1--=-~~.. 100 0 .,. . g D D oo-.- , , , , , 00 , , , , Do 02 U4 08 08 1.0 12 00 02 04 06 08 to 32 Time (s) Time (s) Figure 32 Figure 33 Power Behavior at Level C Power Behavior at Level D of Turbine Trip Test 2 , of Turbine Trip Test 2
.. _ . . . _ m . - .. . .m . ..m.
Prat ti iksitoin 2 Tur tzitie Tr ip Test 1 A I.es el Pem h lkitt oria 2 Turt>me Trip Test 3 H Lesel s,3 o ) . . . . . L!X;END
' LitEND ~
if H=B5L-ThIGL g H4'Hi1[TulGL I D -' Te t luta , D = Test Data l\ emoI +t l It i) g t \ 5 'l i .- !! j ==. I - j .f$ r j t: wo - > I .amo- 5 \ s \ t \ g g amoJ \ uma p s x - 4 .co . r x x . ee. --- , , . . . Y. 3 oo . . . . CD 02 04 Se 08 50 12 00 02 04 06 09 10 12 Time (s) Time (s) Figure 34 Figure 35 , Power Behavior at Level A Power Behavior at Level B M of Turbine Trip Test 3 of Turbine Trip Test 3 Peat h Ihntom 2 Tur turie Ti sp Test 3 - C Level Peacti Ikittorn 0 Turbme Trip Test 3 - D 1.es et 64W:0- - M60 --- -- --- I (7;f?D , LLMEN!) D H = HNL- TWlGL m H = BNL-Tu lGl.
,,,_ D= Tnt tbta , , _ D= Test tbta
~
3 4mo- ,l i
\
~g eooo- ;
lr t ;
-{
, t f 5 ! \ 5 g an - g ase- <
b ! % t
- A y \
p amo- N qx p 2000 - ( + we a > NN >, s' s 3 nuoa N () 3 eo --
, - - . , - ., . oo -
r v---,- , , m 60 04 Ge 06 08 8J 32 00 02 04 06 OB 10 12 Tiene(s) Tiene(s) Figure 36 Figure 37 Power Behavior at Level C Power Behavior at level D of Turbine Trip Test 3 of Turbine Trip Test 3
the top section relative to the measurements. This trend may be due to the fact that a constant DMH fraction was used throughout the core. Since the DMH is caused by neutron slowingdown and prompt gamma heating in the moderator, the presence of voids will affect the DMH. Thorlaksen has shown that the DMH decreases linearly with increasing void fraction.(10) This means that the DMH would decrease with increasing core height. Since the presence of DMH leads to an instantaneous negative feedback (without time delay), which tends to reduce the power rise, the inclusion of a void-dependent DMH would improve the agreement between the calculation and measurement (cf Section 3.3). All the calculated radial region powers were found to show more or less the same transient behavior, which indicates that the radial power shape changes little during a turbine trip. This is also confirmed by the experimental data. However, this is not true for the axial power shape, which shifts significantly during the transient. This is clearly reflected in the dif-ferent power behavior exhibited at different axial elevations, as shown in Figs. 38, 39 and 40 for the three tests, respectively. This means that a space-dependent core model is required to predict accurately the correct power behavior during a turbine trip transient. 3.2 Thermal-Hydraulic Response l The major thermal-hydraulic variables of interest are the void fraction, l l fuel temperature and coolant temperature. Both the overpressurization and scram affect the content of steam voids in the core. As a consequence, the core-average l void fraction changes substantially throughout the transient. This is shown in i I
Peach Bottorn 2 Turbine Trip Test 1
" 7 u:gno l As A Level
-.;#:!B;:l l D= D Levei fg 4
2 oa i / A i //
// 4 3 .. .
j .\ N g =o . .. N N N'\
.N e m)
"a u i< u Time (s) as io a Figure 38 Different Power Behavior at Different Axial Levels of Turbine Trip Test 1 Peach Bottom 2 Turbine Trip Test 2
"*j ugno e A= A Level *
!:!B;:l M
*#8 - D= D level // 1 O
1 ll k. g
-a .
j; }* e j . I t j
}
_., N y 4 u 60 u 64 QA u 1.0 7La Tim e(s) Figure 39 Different Power Behavior at Different Axial Levels of Turbine Trip Test 2 i Peach Bottom 2 Turbine Trip Test 3 mmo e;e U::l f\ no - C = C t.evel I D= D Level i l, 1 v
\
% smo. 6
\ l \
b l N l
!6 y y\\ ,
\
h( s\ \,\'s
- l l
N81 ' N
\\ i )
! j \ N ,'N '
_a NN u: u u o. u o. in a Time (s) Figure 40 Different Power Behavior at Different = Axial Levels of Turbine Trip Test 3 Figs. 41, 42 and 43 for the three tests, respectively. This void behavior corresponds to the pressure behavior which is driving the transient and the reactor. scram. The -fuel temperature does not change appreciably during the transient be-Figs. 44, 45 and 46 show the cause the power excursion is not severe enough. transient behavior of the core-average fuel temperature for the three tests, re-spectively. These behavior are attributable to the power behavior shown in Figs. 23, 24 and 25. The coolant temperature changes little throughout the transient because of the relatively short time span of the transient. 3.3 Sensitivity Studies It has been found during the course of the present work that the calculated turbine trip response is sensitive to the direct moderator heating.(DMH) and . scram delay time (SDT). The direct moderator heating is defined as the percent-age of total fission power that is directly deposited in the moderator (inside the channel box) due to gamma heating and neutron slowingdown. The scam delay time is defined as the time delay between the scram set point and the cual < scram rod movement into the core. Rigorous calculation of the DMH in a BWR is scarce in the literature. Most safety analyses for a BWR assume a constant value of 1 to 4% for DMH. Since the presence of steam voids will affect neutron and photon streaming as well as I L neutron slowingdown, the DMH is expected to be a function of void fraction. A calculation by B. Thorlaksen for a BWR has shown a linear relationship between the DMH and the void fraction.(10) For an average void fraction of 0.4, his calculation yields a value of about 1.5% for the DMH inside the channel box. t ____ _ - - - - - - - _ - - - - - _ - _ _ _ _ _ _ _ _ .
l l 4 l j l Peach Bottom 2. Turbine Trip. Test I san . A em y S IM " i j a ,- -
. l O
Y 0D0' 6tes -
&lm a
&JM , . . . .
00 E6 l.0 AJ Le SA 10 Time (s) Figure 41 Transient Behavior of Core Average Void Fraction for Turbine Trip Test 1 Peach Botton) 2, Turbine Trip. Test 2 s as u.. i . h w. g E OJ. '
. s t
l w' 024 -
- an-00 64 n.0 LA 10 to 10 Time (s)
Figure 42 l Transient Behavior of Core Average Void Fraction for Turbine Trip Test 2 Peach Bottom 2, Turbine Trip, Test 3 , sm .
- 62. *
- w. I 2 .
l
*^'
x 3
- w. \ g s~-
bas 4
,n .0 u 0, .. a.
w a i. i. i. a0 . Time (s) ! Figure 43 I Transient Behavior of Core Average Void Fraction for Turbine Trip Test 3 l
j Peach Bottom 2 Tut bine Trip, Test 1 ii.a. 4
,/
11 % 0 ]
,, 0 ,
F / 1 um0 j i _. ! N j _. j N s _ .. naeso . BP00 . . G4 09 LO Ib Le LS 10 Time (s) Figure 44 Transient Behavior of Core Average Fuel Temperature for Turbine Trip Test 1 Peach Bottom 2, Turbine Trip. Test 2 ma L10a0 9 F f usao- \
$ l E =0 l b
6- 1
} u460-4 1J200
- 00 0.S (4 54 1D 4.3 30 Time (s)
Figure 45 Transient Behavior of Core Average Fuel Temperature for Turbine Trip Test 2 Peach Bottom 2, Turbine Trip. Test 3 l 14M4 - A 4480 0 -
=0 / \
3 '"$8 [
? i N g c=0,i !
\
- g t I ! \
2 "i
\l
- / N
=0, j i i !
0.D 02 C4 08 de to 11 Le L8 La 30 Timeis) 1 Figure 46 Transient Behavior of Core Average Fuel i Temperature for Turbine Trip Test 3 l l l l ; l
N l 4 ; i : l A constant DMH of 2% was assumed in all the calculations _ presented so far. ! In order to.see .the sensitivity, calculations were also done with a DMH of 4% l- -. . - - - for th.e turbine trip Tests l'and 2. The results are shown in Figs.' 47 and 48 7 for Tests 1 and 2, respectively. The sensitivity of the power behavior to the-DMH is quite evident. The higher the DMH, the lower the power peak c,e to the stabilizing'effect of the DMH. The feedback due to DMH has no time delay. This , l behavior has been well explained _in Reference 11. 'The importance of.the DMH in l - a pressure transient warrants a careful study of DMH,in a BWR. Moreover, the I
- dependence of the DMH on the void fraction should be taken into account in the l calculational model for a turbine trip event.
The scram delay time (SDT) is difficult to measure accurately. Most estimates j indicate that SDT may range from 0.1 see to 0.4 sec. Sensitivity calculations suggest that the correct SDT for these turbine trip tests is in the neighborhood of I 0.17 see to'O.2 sec. Fig 49 illustrates the sensitivity of the power behavior to ! the SDT for the turbine trip. Test 3. The importance of the_SDT is as much'as the DMH, so long as a turbine trip is concerned. A careful evaluation of SDT is also , i i warranted. I l j i 3 1 1 i !R I i ). r t :
Peach Bottom 2, Turbine Trip, Test i 250.0 LEGEND 2 2 = 27. DMH 4 = 4% DMH 200,0 -
. A.
n
~i p 150.0 -
E c 100.0 - 0 0 50.0 e E 0.0 , , , , . 0.0 0.2 0.4 0.6 0.8 IJD 12 Time (s) Figure 47 S.ensitivity of Transient Response to Direct Moderator Heating (DMH) for Turbine Trip Test 1 Peach Bo ttom 2, Tu rbine Trip, Tes t 2 300.0 LEGEND 2 = 2?. DM H k ggg, 4 = 47. DMH \ 200.0 - . 2 : 6 120- - m / ,
$ 100.0 -
*0.0 -
'- .y 9
1 0.0 . 0.0 0.2 0.4 l 0.6 0.8 1.0 12 ' Time (s) Figure 48 l Sensitivity of Transient Response to Direct Moderator Heating (DMH) for Turbine Trip Test 2 n> ~_,
/
. / x _a. . e o a ,
-a 9 = . "- . . ' ,9 i
E to
- 7. ,' /
, u U
-T
' ,' j' c AM
....- -/ e ou n.
O
'...=**. ...*......-
rfs m-w G C C. g
.......N g-w-
e m ..,~~~..f
/ A C .e Cf. .
e$) C c e U m 4 .- e o C x -s a . u m-n= C W 5 C
- .- w eu
-a
- - .- u
~
, - m
~ l so
==
- o .E-G-~ >,-
a N
~
n >. 1, .,. .- s,
>e aC 1
~Q .-r==
- .o
-0m
! a e, i w C
o v v A e
.s *
} %
o We w.8 a sz m !W
, . ,3 ~
a > i
< r,n
- o,nw w c ._a 9 J.
e oc C. Il ll
; - c3 .
i O l m 6 1 1 1 i~~ l I i I o o e o o q c c q c c c o o o o o o O LO O LO O C O C v M F N N *-* ~ (paaua %) .lamod 4 i
4.0 CONCLUSION
S AND RECOMMENDATIONS It is concluded that a combination of BNL-TWIGL and RELAP-3B is capable of predicting accurately the transient response of a turbine trip in a BWR. The good agreement with test data suggests that the system thermohydraulic model of RELAP-3B and the core dynamic model of BNL-TWIGL are adequate. For BNL-TWIGL this means the validity of core neutronics and thermal-hydraulics, as well as the feedback model and scram model. For turbine trip events, the physical parameters of major importance and sensitivity to' core dynamic response other than void and scram reactivities are the direct moderator heating and scram delay time. A careful study is warranted to quantify more precisely these two parameters. In particular, the void de-pendence of the direct moderator heating should be modeled in the core thermal-hydraulic representation. A more extensive sensitivity study is recommended. This should involve a more severe transient such as a turbine trip without bypass from full power at E0C condition. This is often referred to as a " Licensing Basis Anticipated Tran-sient" for a BWR. Calculations should be done to check the sensitivity to un- ) certainties in cross section feedback coefficients, scram speed, delayed neutron fraction, and other nuclear and thermal-hydraulic input. l l l b L_____
REFERENCES U 1 L. A. Carmichael and R. O. Niemi, " Transient and' Stability Tests at Peach.
- Bottom Atomic Power Station Unit 2'at End of Cycle 2, EPRI NP-564, Elec-tric Power Research Institute (JQne 1978).
- 2. " User's Manual for RELAP-3B-M00 110: A Reactor System Transient Code" . > ,
~BNL-NUREG-22011, Brookhaven National Laboratory (1977).
- 3. D. J. Diamond, Ed., "BNL-TWIGL, A Program for Calculating Rapid LWR Core ,
Transients", BNL-NUREG-21925, Brookhaven National Laboratory (Oct. 1976). 4
- 4. M. S. Lu, C. J. Hsu, W. G. Shier, H. Connell and M. M. Levine, " Thermal-
> Hydraulic Analysis of Peach Bottom-2 Turbine Trip Tests", to be published, Brookhaven National Laboratory (1978).
- 5. H. S. Cheng, M. S. Lu and D. J. Diamond, "A Space-Time Analysis of Void Reactivity Feedback in Boiling Water Reactors", BNL-NUREG-23501, Brookhaven National Laboratory (Oct. 1977); also, Nucl. Tech. (to be published).
! 6. R. Sher and S. Fiarman, " Analysis of Some Uranium Oxide cnd Mixed 0xide l Lattice Measurements", EPRI-NP-691, Electric Power Research Inst. (Feb. 1978). j 7. R. C. Stirn, " Generation of Void 'and Doppler Reactivity Feedback for ]. Application to BWR Design", NE00-20964, General Electric Co. (Dec. 1975).
- 8. W. Rothenstein, Y. Barhen and E. Taviv, "The Revised HAMMER Code", EPRI-
- NP-565, Electric Power Research Inst. (1977).
- 9. D. F. Ross, Jr., Nuclear Regulatory Commission, private communication, (Sept. 1977).
I 10. B.-Thorlaksen, " Analysis of Control Rod Ejection Accidents in large Boiling i Water Reactors", Risd Report ik). 344, Danish Atomic Energy Commission, ResearchEstablishment,Risd(Nov.1976).
- 11. W. Frisch, S. Langenbuch and P. Peternell, "The Significance of Fast Moderator Feedback Effects in a Boiling Water Reactor During Severe Pres-
- sure Transients", Nucl. Sci. & Ena., 64,843-848(1977).
1 4 J t '
l ACKNOWLEDGEMENT 1 l l The authors would like to thank William Bornstein for his assistance in modifying the BNL-TWIGL code and for providing graphical output for use in the analysis. The cooperation of the Licensing Code Applications Group which did the RELAP-3B calculations is also greatly appreciated. This work was sponsored by the United States Nuclear Regulatory Commission. t l
Distribution List D. Fieno, NRC (5) S. Weiss, NRC (8) M. Dunenfeld, NRC P. Check, NRC K. Kniel, NRC D. Ross, NRC D. Eisenhut, NRC ' R. Mattson, NRC V. Stello, NRC W. Minners, NRC J. Telford, NRC L. Tong, NRC T. Murley, NRC S. Hanauer, NRC NRC Public Document Room NRC Bethesda Technical Library B. Zolotar, EPRI G. Sherwood, GE C. Eicheidinger, W-F. Stern, C-E J. Taylor, B&W W. Mechadon, Exxon J. Rahmstahler, INEL R. Brodsky, DOE ACRS (15) W. Kato, BNL RSP Group Leaders (8) RSP Division Heads (4) RSP Library RCSAG (10) H. Richings, NRC F. Coffman, NRC C. Berlinger, NRC M. Fleishman, NRC Z. Rosztoczy, NRC F. Odar, NRC P. Norian, NRC H. Denton, NRC ^ S. Levine, NRC R. Minogue, NRC J. Naser, EPRI B. Sehgal, EPRI J. E. Wood, GE M. Lu, BNL C. Hsu, BNL
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