ML20003C303
| ML20003C303 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/31/1981 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20003C301 | List: |
| References | |
| 80NED283, NEDO-24279, NEDO-24279-01, NEDO-24279-1, NUDOCS 8102270611 | |
| Download: ML20003C303 (22) | |
Text
._
NEDO-24279 80NED283 Class I January 1981 VERMONT YANKEE NUCLEAR POWER STATION PROPOSED STABILITY AND RECIRCULATION PUMP TRIP TESTS l
l t
I l
i I
l l
l NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GEN ER AL $ ELECTRIC Do22 706// . - . -
DISCLAIMER OF RESPONSIBILITY This document was prepared by or for the General Electnc Company. Neither the General Electnc Company nor any of the contnbutors to this document:
A. Makes any warranty or representation. express or implied. with respect to the accuracy. completeness. or usefulness of the information containedon this docu-ment. or that the use of any information disclosed on this document may not sninnge pnvately owned rights; or B. Assumes any responsibility for liability or damage of any kind which may result from the use of any Information disclosed in this document.
CONTENTS Page SMDfARY vil
- 1. INTRODUCTION 1-1
- 2. STABILITY TESTS 2-1 2.1 Test Description 2-1 2.2 Test Acceptance Criteria 2-4 2.3 Safety Analysis 2-5
- 3. TECIINICAL SPECIFICATION Cl!ANGES 3-1
- 4. REFERENCES 4-1 APPENDIX A - RECIRCULATION PMIP TRIP TEST DURING ThE VERMONT YANKEE STABILITY TESTS A-1 4
lii/1v
k r
- NEDO-24279 ILLUSTRATIONS Figure Title Page 2-1 Vermont Yankee Stability Test Conditions 2-3 1
2-2 End of Cycle 8 Generator Load Rejection, Without Bypass (70% Power) 2-7 2-3 Cycle 8 Loss of 100'F Feedwater lleating (70% Power) 2-8 5
I 2-4 Cycle 8 Inadvertent Startup of IIPCI Pumps (70% Power) 2-9 2-5 End of Cycle 8 Feedwater Controller Failure (70% Power) 2-10 l
4 d
. e v/v1
.u --_._ _,. _ _ . - . _ - - . . . _ . . _ _ . . _ . . . _ . _ . . - __ _ _ , ~ . _ _ _ - _
_.- . _ . _ _.._ _ ___ _ . ._ . ___ __ _ _ .. _ ... . _ _ _ _.m m_.__._,
NED0-24279 l
l SLTfARY I
i j A stability test and a recirculation pump trip test will be performed at Vermont Yankee Nuclear Power Station during Cycle 8 to obtain data for licensing support
- and model qualification. This document describes the proposed tests, presents j
results of safety analysis to show that the tests can be performed while meeting I
all applicable safety criteria, and identifies the Technical Specification f changes necessary to permit performance of the tests.
vil/viii
NED0-24279
- 1. INTRODUCTION It is proposed that a stability test and a recirculation pump trip test
- be conducted at Vermont Yankee Nuclear Power Station during Cycle 8. The objec-tives of the stability test are to:
(1) provide test data for qualification of Vermont Yankee stability performance; (2) provide data for high decay ratio plant operating characteristic assessment; (3) provide test data for additional stability model qualification at natural circulation, high power conditions where decay ratio is expected to be highest; and (4) provide support for low pump speed startup.
Stability tests will be performed at four points (Figure 2-1) on the minimum pump speed and natutal circulation lines. The approach will be to perform tests with vessel and core pressure perturbations introduced through the tur-bine control system. The resulting neutron flux response of the core will be measured and used to determine a core transfer function. A digital data acquisition system will be used to introduce the pressure perturbation and record the system response.
Similar tests have been performed in the past at Peach Bottom Atomic Power Station, Unit 2. As with the Peach Bottom results, the Vermont Yankee results will be used for licensing support and model qualification.
OThe recirculation pump trip test is described in Appendix A.
1-1/1-2
- 2. STABILITY TESTS General Electric Company, in cooperation with the Vermont Yankee operating staff, will perform a series of four stability tests at Vermont Yankee Nuclear
( Power Station during fuel Cycle 8. The proposed power-flow conditions of the four tests are shown in Figure 2-1. These points are selected along either the minimum pump speed or the natural circulation line. Data f rom test point VPT1
,t (minimum pump speed) will allow determination of the sensitivity between min-imum pump speed and natural circulation on stability of the system. Data from tests points VPT2 and VPT3 will be extrapolated to estimate a power level for test point VPT4 (in no case to exceed 70%). Test point VPT3 will also allow ,
measurement of the natural circulation condition decay rntro at the 100%
rated rod line point. Test point VPT 4 will be at the highest achievabic decay ratio subject to the following restrictions: (1) power not to exceed 70%; (2) APRM power oscillations not to exceed 115%; and (3) decay ratio not to exceed 1.0.
2.1 TEST DESCRIPTION At each of the test points, small pseudo random binary perturbations will be introduced into the turbine control system in a manner that will produce variations in the reactor vessel pressure. The pressure perturoations will be a series of steps (or near steps) of approximately 10 psi in magnitude.
The Nuclear Steam Supply System response will be monitored through the APRMs, the LPRMs, system pressure, feedwater and recirculation flows and temperatures within the nucicar steam supply system. The signals which will be recorded are listed in Table 2-1. The temporary test instrumentation will not be seismically qualified, but will be installed so as to not degrade the seismic qualification or performance of existing plant instrumentation. The transfer function relating neutron flux response to core pressure will be the principal result determined from the test data, and will be calculated at the completion of each test to establish the decay ratio for that test point. The decay ratio information will then be used to determine the progression to the next test psint.
QJustification for limit cycle operation is given in References 2 and 3.
2-1
NEDO-24279 Table 2-1 VERMONT YANKEE STABILITY TEST SIGNALS Number of Signals Test Signal Description 80 LPRM 4 APRM 4 Jet Pump Pressure Differential 2 Jet Pump Loop Flows j 1 Core Pressure Differential 1 Reactor Vessel Pressure 1 Reactor Vessel Pressure Differential 2 Core Exit Pressure 2 Reactor Feedpump Flows 1 Reactor Feedwater Temperature 2 Recirculation Loop Drive Flows 2 Recirculation Loop Temperature 1 Total Core Flow (Minimum Filtering) 1 Total Steam Flow 1 EPR (Electrical Pressure Regulator)
Pressure Controller 2-2
NEDO -242 79 180 Q PROPOSED TEST POINTS 160 -
/
/
/ /
IM -
/ /
/ /
REPOSITIONED ! #"
/ l 120 -
REPOSITIONED
/ '
/
APRM ROD BLOCK /
9~
LINE AND ROM !'
b UPPER LINE / APRM E [ / SCRAM LINE 2 /
J j 100 g
/ APRM ROD BLOCK LINE
$ I E l/
i g _____J e .0 -
4 _ _ _ _ _ _
VM4 RATED F LOW CONTROL LINE 60 -
5 VPT1 i
VPT3' VPT2()
l 40 -
f NATURAL CtRCULATION l /
l
[ LINE 20sPUue SPE ED LINE 20 -
I l
l 1
0 ' '
O 20 40 60 80 100 120 140 RATED CORE FLOW (%)
Figure 2-1. Vermont Yankee Stability Test Conditions 2-3
NEDO-24279 2.2 TEST ACCEPTANCE CRITERIA Level 1 and Lrvel 2 criteria have been established to assure that plant response to the stability tests will be within acceptable limits. If the Level 1 criterion is exceeded, the test must be suspended immediately until corrective action can be taken.
If the Level 2 criterion is exceeded, the responsible test engineer must evaluate the situation and determine whether the test should be suspended and/or corrective actions should be taken.
- 1. To avoid inadvertent high flux scram, the following criteria are to be met:
Level 1 (maximum allowable)
Maximum allowable APRM response to pressure perturbations is 20%
of rated power.
Level 2 (expected results)'
Nominal APRM response is not expected to exceed 115% of rated power.
- 2. To avoid any significant impact of possible oscillatory behavior, the following criterion is to be met:
Level 1 (maximum allowable)
, If decay ratio reaches 1.0 and limit cycle behavior occurs, the l
APRM oscillation must not exceed 15% of rated power for the duration j of the data acquisition.
The power level shall be monitored by a visual and audio device con-l
! necting to the APRM signals to ensure that the criterion will not be violated, i
- 3. To avoid exciting a resonant response in the reactor control systems, the following criteria will be met:
l i
2-4
.- ._ . _ _ _ . _ . . .__ ~ - - -- . . - _ - - . _ _ . ._- . _ - - . _ - . - . . _ = . . .
NEDO-24279 Level 1 (maximum allowable) [
] The decay ratio must be less than 1.0 for each process variable outside
, the core that exhibits oscillatory response to pressure perturbations.
Level 2 (expected results)
Decay ratios are expected to be less than 0.5 for all process variables 4
outside the core.
- 4. Offgas release will be continually monitored during testing and
~
the following limits will be applied:
Level 1 (maximum allowable)
The steady-state SJAE E6 offgas and the reactor coolant solubles (primarily iodine) must remain within the site administrative limits.
Level 2 (expected results)
'The steady-state offgas radioactivity releases, as normally measured by the SJAE E6 offgas and reactor coolant fission products (primarily iodines), are not expected to be influenced by the proposed testing.
However, a detailed review of fission product releases will be accom-i plished p ior to the tests to establish Level 2 guidelines for the responsible test engineers.
2.3 SAFETY ANALYSIS Those . parts of the reload license analysis (Reference 1) which may be affected by operation at the special test conditions have been re-analyzed at point VPT4 (maximum power, minimum flow expected during the test):
- 6. ' UNIQUE TRANSIENT ANALYSIS INPUTS
- EOC8 Void Coefficient N/A (-c/% Rg) . 9.75/12.19 Void Fraction (%) 47.4
- These' sections refer to corresponding sections in CE' report Y1003J1A02, July 1980 (Reference.1).
2-5
NEDO-24279 EOC8 Doppler Coefficient N/A (-C/*F) 0.228/0,217 Average Fuel Temperature (*F) 1042 Scram Worth N/A (-$) 36.27/29.02 Scram Reactivity vs Time see Figure 2A (Ref. 1)
Thermal Power (MWt) 1115 70%
Steam Flow (Mlb/hr) 4.33 67.3%
Core Flow (Mlb/hr) 17.3 36%
Dome Pressure (psig) 983 Turbine Pressure (psig) 964
- 7. UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS
- Core Power (MWt) 1115 Core Flow (M1b/hr) 17.28 Reactor Pressure (psia) 987 Inlet Enthalpy (Btu /lb) 483.3 8x8 8x8R P8x8R Peaking Factors (local, 1.22,1.47.,1.40 1.20,1.52,1.40 1.20,1.52,1.40 radial, axial)
R factor 1.098 1.052 1.052 Bundle Power (MWt) 4.168 4.496 4.499 Bundle Flow (Klb/hr) 40.65 41.38 41.49 Initial MCPR 1.25 1.26 1.26
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTZ Pcwer Core Flaw $ h/A SL V f. cts Flant M eg Meg 1 t!) (I WBF) (1 N!N) Q M M PnB 9:RR TM lle.ponse Load Fejection SOC-EOC 70 36 102 71 1126 1136 0.05 0.06 0.06 Figure 2-2 w/o typene Less of 100*F tat.FtJC 70 36 82 0? 94 3 -
- 0. I ' O.13 0.13 Figure 2-3 feedwas ar Beet Las Inadvert eet sor goc 70 36 89 88 985 -
0.18 0.19 0.19 Figure 2-4 NFC1 5 tart Feedwater SOC-EOC 70 36 84 75 990 1005 0.08 0.08 0.08 Figure 2-5 Cowt roller Failure
- Sze footnote on page 2-5.
to 2-6
_ _ . _ _ _ _ _ _ _ - ~ _ __ _ _ _ _ _ _ .__ .. . . _= _ _ _ _
1 NEUlf'.T! $ "X l - lI. . .f. ti eFrC3 v-RI'[ tp;[g
.* nyt -y . '.l t of t tux ,
,.- .vr ie nw
('.Lr,* su_
. cg i iOK ' t' jtlu g, i r * , ,
l 4* . .t., fwv; F60w
+
i N ).
2 t pe. .
- - ~
p 4,
[
Si, 1, I r.2 ?.,
3 m* '
N9 ta- '
% I
-r t
t :.L .\
'~ ~3N~ _ i 1 -
Nm. 7 1
f
\q i
M ;'
g, * i 3 g _. g, 3u 3% , -
t1 3 1-
- 9. g, 0* 37, W* O. q. B* 1;. 16.
ggg (SEE1 ggg tSEC1 2:
tn 8'
~ ,, ,
I v N
1 I I LEVEL t!@-fYF-SEP-SKIRT 1 vf'IFTWilv!T) U i 2 VES#f t 3h *f l ow r. e Ltn rcpCTIv!Tr l 3 ilt
- AW L t?4
,_ />4 e tat ff.u.Tivl t 2-m1 -
sec Tuu atu'btrJ 7 con.Tmm i M .
a j/ ~~
u
-(
Q ,,,-
- = O-mu<c. - 3 - --
.;_;.a
^
Ei@ r"~g -1 3 ~
2~ x_
C3 n.-
3 a -i i .: .i.
N PE) 3. . _
-i - s N
@fE) $ s c',* . .... >
..... . .~
,Td, . i n,'9. 4. ts. 12. I'b. 0. 0.9 .1.i .~ l' . , .
TIME ISEC1 Tit 15tCi u
b Figure 2-2. End of Cycle 8 Generator Load Rej ection, Without Bypass (70% Power)
=
0 m2? _
..~ . . . .,.
m.o ,.
M, d_-:-
- =m' 'n gg h e c '_,'
P
]
ca., -
L
- u. -
y.-.. w sx Li e,
.. - .? '
. c, -; <. '2 1 g
-~,, < :. n
- - - .: u o r o E M- ' '
- a u O W -
n
.-- l
, .. a g ! -r s
- w ce f , C s .
4 O
o
. ,f ,. e
. = , , - > =
l : u
~
- o
$ ,I M c
{
. -E 3
- ,t- a T
. . . .o . . o o
9 N
3 S S i i 2 tsl51tG'Od43111 A113e3d ."
o
.O H
5 *
- g O D u, 2 -
L ,- g
- [
c ~~ l MT$
eoo 0
k-s 3 A
h-r: . Wgid ,
m
- ' m s G r.- F g o u - - -
T,:-
e t' bylA
=.,W i pg u K
=cc= u,-
a; J,
. -E.1
-~r g
-~ g,, .
n u
a_- u N i
m D M f' - M ,e W W &
8:
, 8 "
m F e
9 .
o, O
... 5 b- .
S 8-S. R R R 4, (G31th A IN3'M3dl 1
I i
1,> .
. I TUTR**4 5 TUX I W EL d $ RISE (% !)
2 Ad SpurE T AT Flux l. 2 F t it r m. .t rttu
~
150. 3rm y rym 150.
3 '"I!' 'I-"I-MLN -
4 Litt hviLc4 tb g100. 100.
I 1S 19 I E -
. . b
- c. _
(( * $0. 50.
. 1,: c 3 3 3 ,
- pj .
"'-** I 1 I O.
- D. h > -'
' ^ ^ *
- 0. 20. 40. 00. 80. G. 20. 46 0 . 60. 60.
TI E ISECl TIE 15ECl .
Z fn Cf N O I '
u I LEVEll! D -REF-SEP-SMIRT I VOID Trit!VITT e" 2 VFS5EL SIE A*" t.0 4 2 Co n fR 'F KTIVITT Ig' 3 Tttn4!T 5 'E Aw t N I* 3 y_ qre, PE411 v l Y T 4 Ftt,aTriW.4 -
4'.c:.4. 7.i..L.II e IT r i , t
_ 1 100' 0' * * "
P.s) 5, N ,-
M I I 1 E. .l.
M u &
- g -
FM a. :. .>.... . a, .... ... ..
D. 20. 40. fio. 80. O. 20. W. [0 IC.
TIE ISECI T 11. tSt el Figure'2-4. Cycle 8 Inadvertent Startup of HPCI Pumps (70% Power)
_ . - . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ . _ _ _ . . _ _ ._ ._,m _ _ ,
[ t a
l l Af tlift<% d iIV l ltV*^'ff Pglrq cter (pgll 2 .m v t r ,< nt re: i
, i. H .r. us gg .} f(* 'r,i '!* I 4% yg, *
- l .s t agp e , *4 r
# 'I' i-
- k .
I c ,
" r :.. . s e 5
w.
- 12. 150.
L,
- n %
, .= ,
~
3 so. ' so. \
c
- s it w- 2 3 3 7; *: _
y_ m 3 '4 m_ ,
t 4
., D. ' - - ' * -50. - t- - ^ -
t i 0. 10. 20. 30. 40. O. 10. .m . E. 40.
i TIME ISECl TI4 (5001 1
! Z M
- t:t O
+ N e a
a
' .H N 4 o 7 N
- 1LEVELt!VjH-F1CF-SEP-5%IAT l Citt FTd,W U 2 VEstit S $ (M t N ' .* f;F f14 .hitItv!Tf 3 it tw.r. :fWLN ,. 7. . i t = rre .tiv1'T i
125*
1 iilli(4.MD li OT- -
I* 4 :oi&(~6 ; ..Ililli!
75 y ,- ,
O. h8~ ~I.~E- *.
- C3
" / E 25' v
- \- 9 -
1 +-
- 63) g'
- w i
, V N. k ..
E'
), -
I
- - 25. - '- + 2. - r- - i _ ._ __J l ._ _-t .
O. 10. 20. D. 40. O. 8 1 .
I* .~ 4 .
y"d T!K-(SCCI 11"t 1
L3.,
Figure 2-5. End of Cycle 8 Feedwater Controller Failure (70% Power) ii
l l 10. LOCAL ROD WITHDRAWAL ERROR TRANSIENT
SUMMARY
The rod withdrawal error was analyzed with the special RBM setting (Section 3.2) at the natural circulation condition. The ACPR was 0.13.
- 11. MCPR OPERATING LIMIT
- For the duration of this test, the MCPR operating limits shall be:
(a) with forced recirculation flow:
Ex8 8x8R PPx8R Kg x normal limit (b) without forced recirculation flow:
8x8 8x8R _P8x8R 1.30 1.31 1.31 The part b limits include a 0.05 adder to account for lack of operating experience at high power, natural circulation conditions and the lack of a data base to benchmark the calculations at this condition.
- 14. LOSS-OF-COOLANT ACCIDcNT RESULTS*
The flow-biased APRM rod block normally provides the LOCA margin required for operation at reduced flow. This assurance will be diminished when the APRM rod block trip is increased and special restrictions on MAPLIICRs are required for operation in that mode.
Af ter adjustment of the APRM Rod Block setting, the MAPLHGR shall be 80% of the limits specified in the Technical Specifications.
- See footnote on page 2-5.
2-11/2-12
- 3. TECHNICAL SPECIFICATION CHANGES Due to the special nature of this test, the following changes to the Vermont Yankee Technical Specifications will be required:
(1) To minimize the risk of inadvertently tripping the plant during the conduct of these tests, Vermont Yankee requests relief similar to that provided to perform various required Surveillance Tests under Technical Specifications. Specifically, relief must be granted to permit bypassing of any affected trip function for those moments when the test instrumentation is installed or removed at the station.
The required trip functions shall be otherwise operable during the test.
(2) Three of the test points are at the natural circulation condition; it will be necessary to modify the Technical Specifications to allow operation without forced recirculation flow.
(3) Test point VPr4 is above the APRM flow-biased rod block line and very close to the APRM flow-biased scram. Higher settings for these two trips will be required to conduct the test. The APRM flow-biased rod block line equation will be changed to SRB 1 0.66W +
75%, and the APRM flow-biased scram line equation will be changed to S < 0.66W + 85%. It will also be necessary to remove the require-ment for peaking factor set down to assure that the test point can be reached. These changes are possible because no credit is taken for the flow-biased APRM scram and rod block in any safety analysis except LOCA. Special LOCA limits are to be imposed when the flow-biased APRM scram and rod block is raised to compensate for the change in trip setting. Since the fuel cladding integrity criterion was changed from MCHFR (a local phenomenon) to MCPR (a bundle phenom-enon), the peaking factor setdown requirement no longer serves any useful function. Violation of the 1% cladding plastic strain limit l
will not occur for scram at 120% during any transient initiated from
! the operating limit LHGR, (4) To reach test point VPT4, it will be necessary to raise the Rod Block Monitor setting. The Rod Block Monitor trip setting equation
- vill be changed to 0.66W + 75%. Special evaluations of Rod With-l .rawal Errors with the increased setting show that the consequences of this transient are bounded by the proposed MCP" limits.
1 (5) At natural circulation, the standard Kf adjustment factors to the MCPR are quite large and could prevent reaching test point VPT4.
Following the recirculation pump trip, the MG set controller will l be placed on the minimum setpoint, and administrative controls will be utilized to prevent the inadvertent starting of the recirculation pumps. Special MCPR limits, based solely on limiting transients from test point VPT4, will be observed in lieu of the Kf adjustment.
Hence, the Kf factors are not necessary during testing at natural circulation.
(6) Special MAPLHGR limits will be observed in lieu of the standard flow biased APRM rod block setting.
3-1/3-2 r
NED0-24279
- 4. REFERENCES
- 1. Supplemental Reload Licensing Submittal for Vermont Yankee Nuclear Power Station Reload No. 7, Y1003J1A02, July 1980.
- 2. Letter, R. E. Engel to Darrel Eisenhut, NRC, April 4, 1977.
l l-l l
i l
s 4-1/4-2
NEDO-24279 APPENDIX A RECIRCULATION PUMP TRIP TEST DURING THE VERMONT YANKEE STABILITY TESTS A.1 INTRODUCTION A recirculation pump trip test will be incorporated into the Vermont Yankee stability test program. The objective of the additional test is to provide Vermont Yankee Nuclear Power Corporation data for qualification of operational transient computer codes. A description of the test is as follows.
A.2 TEST DESCRIPTION The recirculation pump trip test will be performed after the stability test at VPTl has been done and before reach'.ng the test point VPT2. Following comple-tion of the stability test at VPT1, the control rods will be inserted to reduce the power to a lower level along the minimum recirculation flow line such that test condition VPT2 will be reached if the recirculation pumps are tripped.
But, rather than trip the pumps at this point, the following procedure will be followed: (1) Increase the core flow as close as aractical to 100% rated flow condition; (2) establish thermal hydraulic steady state; and (3) trip the pumps from this operating mode to obtain the recirculation pump trip test data. Both MG set drive motors will be tripped manually and simultaneously from the control room.
A.3 ADDI-fIONAL INSTRUMENTATION REQUIREMENTS In addition to the test signals delineated in Table 1, the following signals l
l will be required:
Number of Signals Test Signal Description 1 Total Steam Flow 4 Steam Line Pressure Differential (for flow measurement, 4 lines) 1 Narrow Range Vessel Water Level 2 Recirculation Loop Dr'ive Flow (both loops) (the same as those for the stability test)
A-1
r.
l l
NED0-24279 l
Number of Signals Test Signal Description
- 2 M-G Set Generator Speed (both loops) 2 M-G Set Generator Trip Signal i A.4 SAFETY REVIEW Automatic Recirculation Pump Trip systems have been installed at several BWRs to mitigate the consequences of both accidents (ATWS) and transients. The effect of a recirculation pump trip is a large, fast, negative reactivity insertion due to the void reactivity effect. Pump trips have occurred at BWRs in the past, inadvertently, as a result of tests, and as a result of turbine trips or load rejections. No adverse safety consequences have resulted from any of these trips. The natural circulation condition is the least stable condition on the BWR-power flow map; however, the analyses presented in Reference A-1 show a decay ratio of less than 1.0. In addition, the special monitoring.during the test will preclude the existence of any unsafe oscilla-tions following the pump trip. The technical specification changes required to permit continued operation at. natural circulation are discussed elsewhere ir this report.
A.5 REFERENCES
, A-1. " Supplemental Reload Licensing Submittal for Vermont Yankee Nuclear Power Station Reload No. 7," Y1003J1A02, July 1580.
J c~ A-2 i '
.