ML20135B814

Proposed Changes to Tech Spec 5.2.2.a,eliminating Min Control Rod Withdrawal Rate Limit

ML20135B814
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	09/03/1985
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
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ML20135B784	List: ML20135B781 ML20135B797 ML20135B814
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NUDOCS 8509110193
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{{#Wiki_filter:--- _ _ _ .g: N-ATTACHMENT 1 Consumers Power Company Big Rock Point Plant Docket 50-155 Proposed Technical Specification Page Changes September 3, 1985 (- I l 3 jr1263!INM'i P 1 Page TSOB0885-0005C-NLO2 _______-J

44 5.2.2 control Rod Performance (a) Control Rod Performance The following limits shall apply to any control rod which can be withdrawn. It shall be permissible to tag and valve out the hydraulic drive water to a fully inserted control rod which is defective or does not meet these limits provided the remaining rods do meet the limits. The following tests shall be performed at each major refueling ] l but not less often than once every 20 months, and prior to i startup following any outage greater than 120 days in length. g (1) Withdrawal of each drive, stopping at each locking position to check latching and unlatching operations and the functioning of the position indication system. (ii) Scram of each drive from full withdrawn position. Maximum scram time from system trip to 90 percent of insertion shall not exceed 2.5 seconds. (iii) Insertion of each drive over its entire stroke with reduced hydraulic system pressure to determine that drive friction is normal. (iv) Continuous withdrawal and insertion of each drive over its stroke with normal hydraulic system pressure. (b) Core Shutdown Margin Verification The reactivity of the core loading shall be such that it is always possible to maintain k at less than 0.997 with the most valu-ablereactivityworthEktrolbladecompletelywithdrawnfromthe core. The core shutdown margin shall be verifica by a demonstra-tion that the reactor is suberitical with the most valuable reactivity worth control blade fully withdrawn, plus an immedi-ately adjacent blade withdrawn to a position known to contribute 0.003 k or more to the effective multiplication. In the event f that the maximum reactivity condition occurs at a temperature greater than ambient, the demonstration will either be performed at that temperature or a suitable additional margin will be demonstrated at ambient. This verification shall be performed prior to start-up after any shutdown in which the system has cooled sufficiently to ll Proposed 4/15/85 t Proposed 9/03/85 l TSB0885-0005D-NLO2

1 ATTACHMENT II Consumers Power Company Big Rock Point Plant Docket 50-155 EVALUATION OF ROD DROP ACCIDENT IN SUPPORT OF DELETION OF MINIMUM 23 SECOND WITilDRAWAL CRITERIA September 3, 1985 23 Pages TSOB0885-0005C-NLO2

I' i Standard Review Plan - 15.4.9 Acceptance Criteria I j - 1.. Reactivity excursions should not result in radially averaged fuel rod j enthalpy greater-than 280 cal /gm at any axial location in any fuel rod. II. The maximum reactor pressure during any portion of the assumed i excursion should be less than the value that will cause stresses to l exceed the " Service Limit C" as defined in the ASME Code. i j III. The number of fuel rods predicted to reach assumed fuel failure thresholds, and associated parameters such as the amount of fuel i reaching melting conditions, will be an input to a radiological i evaluation. The assumed failure thresholds are a radially averaged J, fuel rod enthalpy greater than 170 cal /gm at any axial location for i j zero or low power and fuel cladding dryout for rated power-initial i conditions. l i ii i l-i. 4 4 l 4 1 J I l, f = + i i l l .TSOB0885-0005C-NLO2

j \\ 2 ( TABLE OF CONTENTS SECTIvN TITLE PAGE I Methodology Used to Meet Acceptance Criterion I 3-4 II Methodology Used to Meet Acceptance Criterion II 4-5 III Justification for Meeting Acceptance Criterion III 6-8 Appendix A 9 Appendix B 10-14 References 15-16 ( L IC0984-0006A-RE01

3 l - ( SECTION I: METHODOLOGY USED TO MEET ACCEPTANCE CRITERION I .The peak deposited enthalpy resulting from a rod-drop accident was calculated according to the format outlined in Reference 1, pp.11-12. Although Reference 1 shows a generic calculation of peak deposited enthalpy, it demonstrates the format followed for rod-drop accident analyses. The basic assumptions made in the Exxon Report XN-NF-78-51 (Reference 1) are that: 1. A central core region control rod becomes decoupled from its drive mechanism and remains fully inserted in the core while the original withdrawal sequence continues. 2. At the worst point in the withdrawal sequence, the stuck rod becomes ( - freed and falls out of the core, producing a prompt critical reactor which results in a rapid power increase that is effectively reduced to operational level through the Doppler feedback mechanism. As is shown in Figure 3.2 of Reference 1, scram bank insertion does not significantly affect the transient power level. However, the insertion of the scram bank, along with the Doppler feedback mechanism, does serve to bring the reactor down to a subcritical stage after Doppler feedback has reduced the power level to about 10 x (Maximum designed thermal power), or 2400 MWt. 3. The reactor remains at the suberitical stage due to the full insertion of the scram bank. P l u-IC0984-0006A-RE01

4 C The X-TRAN computer code (Reference 3) was used to generate the graph of peak deposited enthalpy versus control rod worth for 4 different axial x radial peaking factors (Figure 1.1 of Reference 1). The base values for the Doppler coefficient and delayed neutron fraction used to generate Figure 1.1 are -6 c =-9.52 x 10 ak/k-F,andlI= 0.00525, respectively. Figures 1.2 and 1.3 D of Reference I supply information necessary to adjust the peak deposited enthalpy (from Figure 1.1, Reference 1) according to different values of the Doppler coefficient and the delayed neutron fraction, respectively. Specific analysis of a rod-drop accident was performed for Cycle 20 at Big Rock Point (Reference 4), and indicate that the total peak fuel enthalpy during the power transient is well under the specific energy design limit of 280 cal /g for UO fuel (Reference 6). A total peak fuel enthalpy of not more 2 than 179.95 cal /g would occur in the hottest core region during the power transient resulting from a rod-drop accident that occurs according to conditions outlined in Exxon report XN-NF-78-51. Appendix A contains a copy of Reference 4, and shows the conservative measures taken for the Cycle 20 rod-drop accident analysis. These measures include selection of a Doppler coefficient, and the assumption of an initial fuel enthalpy of 18 cal /g, which produce a slightly higher total peak fuel enthalpy than should actually occur. SECTION II: METHODOLOGY USED TO MEET ACCEPTANCE CRITERION II Some conservative assumptions regarding heat transfer and reactor vessel thermodynamics were made, and are listed below. 1. The energy generated in the core during the rod-drop transisnt power is instantaneously transferred from the fuel to the coolant at various times t, n=1,... 13. This is very conservative since energy transfer delay time due to thermal resistance of the fuel, gap and cladding is considered to be zero, s IC0984-0006A-RE01

5 2. The vessel is assumed to be enclosed by an adiabatic boundary, ie, no heat escapes from the reactor vessel during the rod-drop power transient. The system is thus considered to contain an adiabatic-reversible heat transfer mechanism for transient power. 3. Assumption #2 implies conservation of mass within the system, ie, initial coolant mass at the beginning of the power transient is equal to final coolant mass. The pressure relief safety systems therefore do not apply. 4. The mixture enthalpy (hm) as calculated in Table B-II is equal to h, the g fluid enthalpy. This is conservative since mixture quality q = (h,- h )/(h -h ). See Appendix B for details. g g f The maximum reactor pressure during the power excursion resulting from a rod-drop accident was calculated for a 10 mk control rod according to the format listed below. Appendix B contains details of the calculations involved at each step. C (i) The transient power curve (Figure 3.2, Reference 1) was integrated in increments from t, = 0.55392 see to tn, n=1.. 13. The initial time t is taken as the point where reactor power increases above 240 9 MWt. This resulted in reactor core energy generation En" various time periods of the transient power level. (ii) The enthalpy of the mixture (h,) was calculated as initial enthalpy plus Ah,. The quantity Ah, is the enthalpy increase of the system due to the energy deposited from the rod-drop accident. (iii) Mixture enthalpy of the system is taken as fluid enthalpy (h ), and g Reference 7 is used to calculate the corresponding saturation pressure and temperature, P and T,, respecuvely. n k IC0984-0006A-RE01

6 CONCLUSION As is shown in Table B-II, the vessel pressure increased toev 1466 psia at t l3 = 0.75 sec, producing a total pressure change of 116 psia due to the rod-drop power transient. Additional vessel pressure increase due to the power transient beyond 0.75 see is insignificant, and was not calculated since the main steam bypass valve is fully open 0.2 see af ter a pressure increase of 50 psia is attained (see Reference 8). Therefore, the maximum vessel pressure during any portion of the power transient resulting from a rod-drop accident is considerably less than the design pressure of 1715 psia for the BRP reactor vessel. This is conservative relative to maximum design pressure according to " Service Limit C" of the ASME Code (Reference 2). SECTION III - JUSTIFICATION FOR MEETING ACCEPTANCE CRITERION III A generalized cladding failure threshold of 170 cal /g was established from a series of tests performed by J.E. Grund, et. al., in 1969/1970 (Reference 5), and given in General Electric report #NED0-10527 (Reference 6). Iloweve r, specific tests and analyses were performed for different fuel types in Reference 5, where GEP pellet fuel most closely matches Cycle 20 fuel at Big Rock Point. From Table III of Reference 5, the clad thickness for GEP fuel is 0.032 in., which compares quite well to the clad thickness of 0.034 in, for Big Rock Point !!-type fuel. Test results for GEP pellet fuel are shown in Table C-I, Reference 5, where CDC transient test No. 487 indicates cladding failure occurred at 205 cal /g. This test result should correspond more closely to the cladding failure threshold for !!-type fuel. The generalized failure threshold of 170 cal /g (Reference

6) is derived from tests on different fuels, where clad thickness varied from 0.014 in. to 0.032 inches.

L IC0984-0006A-RE01

7 ( Reference 15 displays results of transient tests conducted on 0.3125 in, o.d. pellet fuel with a clad thickness of 0.020 in. and clad material consisting of Zr-2. Loss of clad integrity for this down-scaled fuel rod occurred at fuel enthalpies ranging from 223 to 256 cal /g. Although the higher clad failure threshold is expected due to the smaller fuel diameter, this demonstrates the dependence of clad failure during transients on parameters such as fuel diameter and cladding thickness. The total peak fuel enthalpy from a rod-drop accident at Big Rock Point was calculated in Reference 4 as approximately 180 cal /g (See Appendix A), using some very conservative parameters. This peak fuel enthalpy should occur in a region consisting of the four assemblies clustered around the dropped control rod. If the generalized clad failure threshold of 170 cal /g is assumed, a limited amount of fuel rod perforation may occur in the region of the four clustered assemblies, if peak fuel enthalpy approaches 180 cal /g in certain rods. The radiological consequences of such an accident can be gauged by comparison to the "ma<imum credible accident" (MCA) considered in Section 13, Volume 1 of the Final liazards Summary Report (Reference 13). Specifically, the case of a 10% core meltdown resulting from core spray reduction during the MCA is employed. The immediate concern of an accident which causes fuel rod perforation is release of gaseous fission products contained in the fuel rod plenum of irradiated fuel. For rod locations under going clad melting, the fuel pellets would fall out of the core, preventing these pellets from reaching melting temperatures. An overall stunmary of the radiological consequences of a 10% core meltdown is presented in Figures 13.5-13.15, Reference 13, and indicate that radiation doses are insignificant at the distance of a small population zone, and are not considered hazardous at distances near the plant boundaries. The only precaution that may be necessary is non-use of milk products for approximately 3-4 weeks due to ground deposition of Iodine-131 in an area of less than 1 sq. mi. near the plant site, as described on p.27, sec. 13, Reference 13. This implies that k IC0984-0006A-RE01

8 the radiological consequences of a rod-drop accident would have an insigni-ficant impact on the area near the plant boundaries, since highly localized clad melt temperatures would be generated, and would not approach 10% core meltdown. This assumes that fuel melting temperatures are not attained in the rod-drop accident. Therefore, if a 170 cal /g clad failure threshold is assumed for current BRP fuel, the radiological impact of a rod-drop accident described by Exxon Report XN-NF-78-51 (Reference 1) and calculated in Reference 4 would still be considered negligible. ( ( IC0984-0006A-RE01

9 APPENDIX A Determination of Peak Fuel Enthalpy for the BRP Reactor The calculation of peak deposited enthalpy using the format of Exxon Report XN-NF-78-51 relies on determination of 5 neutronic parameters. These are:

1) Calculation of the hot standby dropped rod worth for a particular rod which will cause the reactor to reach a prompt critical stage. The reactor is assumed to be near a critical state initially (p.1, Reference

1) with a corresponding reactivity less than the delayed neutron fraction (p.3, Reference 1).

2) Calculation of the axial x radial peaking factor, excluding xenon, Doppler, and thermal-hydraulic feedback mechanisms.

3) Calculation of the Doppler coefficient.

4) Calculation of the delayed-neutron fraction in hot standby mode.

5) Selection of an appropriate assembly local peaking factor.

Calculation of parameters (1), (2) and (4) was performed using the GROK computer code (Reference 11), and apply directly to Cycle 20 operation at Big Rock Point. Since peak deposited enthalpy for a simulated Cycle 20 rod-drop accident was calculated in Reference 4, the details are not shown here. Instead, a copy of Design Review B*C*20h840828 has been included. ( IC0984-0006A-RE01

10 APPENDlX B Determination of Incremental Vessel Pressure Calculations listed below are ordered to correspond to the format steps of Section II. (i) Equation of curve used to approximate reactor core power of Figure 3.2, Reference 1: log P = mt + b, t, 5 t i t and t7<t<t12' 5 => P = C10*', where C=10, and m= slope of log P. b Equation of energy used to calculate reactor core energy of Figure 3.2, Reference 1: t E= P dt = C10*'dt; for regions 1 and 2 as defined below 'l [10*'" - 10*li]; for regions 1 and 2 as defined below.

> E

9 mln(10) IC0984-0006A-RE01

r 11 Results of graphical analysis of Figure 3.2, Reference I are shown in Table B-I, utilizing the equations below: Region 1 Region 2 0.55392 5 t 5 0.59373 0.60863 $ t 1 0.65 M3 = 62.61384 My = -40.85246 bg = -34.68306 b2 = 27.39672 -34.68306 27.39672 C1 = 10 C2 = 10 EI= C' [10*l "-10mit0]' E = t t7)' t {10m2 n 2 -10m2 mt n(10) n I n m2 n(10) 1 n=1,2,3,4,5. n=8,9,10,11,12. For the time interval t5' '7, a e nstant p wer level was assumed. Ws power level was calculated by using the power equations in Regions 1 and 2 at times t5 *"d '7, respectively, t find an average. Thus, P =b 3 326.0, normalized, gs,ty 2 -34.68306 62.61384(0.59373) where P3 = 10 10 , Region 1 7.39672 -40.85246(0.60863) P2 = 10 10 , Region 2 The truncated-triangle approximation to the logarithmic power increase of Figure 3.2, Reference 1 is thus conservative due to the constant peak power assumption for t5 # ' # '7 The constant power assumption for is likewise conservative. t <t$t33 12 k. IC0984-0006A-RE01

12 TABLE B-I Time, t atn

• Normalized
  • Normalized
  • Energy Generated, (sec)

(sec) Power, Pn Energy, En(10-2) En (MJ) t .55392 0 1.0 0 0 o t .56 .00608 2.403 .973 2.335 g t .57 .01608 10.158 6.352 15.245 2 t .58 .02608 42.950 29.097 69.833 3 t .59 .03608 181.60 125.263 300.631 4 t .59373 .03981 5 214.966 515.918 t .60 .04608 326.0 419.368 1006.483 6 t .60863 .05471 7 700.706 1681.694

• 8

.61 .05608 299.72 744.531 1786.874 t .62 .06608 117.00 938.777 2253.065 g t .63 .07608 45.674 1014.605 2435.052 10 t .64 .08608 17.830 1044.206 2506.094 33 t .65 .09608 6.960 1055.761 2533.826 12 t .75 .19608 5.000 1105.761 2653.826 33

• (P =1 => 240 W t)

O D = E '#" ( t:tn )

• • (En tot I.

IC0984-0006A-RE01

r 13 TABLE B-II Mixture Saturation Saturation Time, tn Atn AEn

Enthalpy, Temperature,

Pressure, APn (sec)

(sec) (MJ) hm(BTU /lb) Tn ( F) "n (psia) (psia) t .55392 0 0 592.3 582.32 1350.0 o t .56 .00608 2.335 592.3 582.32 1350.0 0 g t .57 .01608 12.910 592.4 582.39 1350.8 0.8 2 t .58 .02608 54.588 592.7 582.62 1353.1 2.3 3 t .59 .03608 230.798 594.0 583.57 1363.1 10.0 4 t .59373 .03981 215.287 595.2 584.45 1372.3 9.2 5 t .60 .04608 490.565 598.0 586.49 1393.8 21.5 t .60863 .05471 675.211 601.8 589.21 1423.1 29.3 7 t .61 .05608 105.180 602.4 589.64 1427.7 4.6 g t .62 .06608 466.191 605.1 591.55 1448.5 20.8 9 t .63 .07608 181.987 606.1 592.26 1456.2 7.7 10 t;; .64 .08608 71.042 606.5 592.54 1459.2 3.0 t .65 .09608 27.732 606.7 592.68 1460.8 1.6 12 t .75 .19608 120.000 607.3 593.11 1465.6 4.8 13 Description of Variables: (i) at * 'n ~ lo' "*II n (ii) AE =E - E,,g, n=1.. 13. 9 n (iii) Mixture Enthalpy, h, = (h,/ =to *I I" )' "' '" I " I I I " I t " tot k IC0984-0006A-RE01

{ 14 mixture enthalpy, h,/,g, = hg = 592.3 BTU /lb. The coolant mass t total, M =167,395 lbs., calculated for the BRP reactor vessel at g 0% void, 582.32'F and 1350 psia (See References 7,12). For t, < t 5 t33, h, is conservatively taken as h, since h, > h. g g (iv) Both T and P, were calculated using reference 7, and correspond to n mixture enthalpy, since h, a b,.. (v) AP = P,- P,,3 I I ( l k IC0984-0006A-RF.01

[ 15 ( REFERENCES l

1) " Control Rod Drop Accident Analysis for Big Rock Point", Report No.

XN-NF-78-51, Exxon Nuclear Company, Inc., January, 1979.

2) "ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components", 1983, Division I, subsection NC.

3) "XTRAN-PWih A Computer Code for the Calculation of Rapid Transients in Pressurized Water Reactors with Moderator and Fuel Temperature l

Feedback", Report No. XN-CC-32, by J.N. Morgan, Exxon Nuclear Company, Inc., September, 1975.

4) Design Review B*C*20*840828, UFI-740/22*13*32; by S.M.Soltis, August.28, 1984.

5) " Subassembly Test Program Outline for FY 1969 and 1970", J.E. Grund, j

R.L. Johnson, K.O. Johnson, R.V. Miller, R.T. Johnson, B.E. Norton; j IN-1313, Idaho Nuclear Corp., Idaho Falls; August, 1969 (IDO-17277).

6) " Rod Drop Accident Analysis for Large Boiling Water Reactors", C.J.

Paone, R.C. Stirn, J.A.Woolley; Report No. NEDO-10527, 72 NED 18; Atomic Power Equipment Department, General Electric Company; March,1972. (

7) "ASME Steam Tables - Fourth Edition," C.A. Meyer, R.D. McClintock, G.J.

Silvestri, R.C Spencer, Jr., Copyright 1979.

8) " Sequence of Pressure Relief During Transient", Letter f rom S. A.

Bartosik to S.M. Soltis regarding procedure and results of " Turbine Bypass Valve System Functional Test", implemented 7/9/84 at BRP; August 21, 1984.

9) " Nuclear Reactor Analysis" J.J. Duderstadt, L.J. Ilamilton, The University of.lichigan, Copyright 1976 by John Wiley & Sons, Inc.

10) " Nuclear Engineering llandbook", liarold Etherington, Editor; Copyright 1958 by McGraw-liill Book Company, Inc.

11) " Flare - A Three-Dimensional Boiling Water Reactor Simulator", D.L. Delp l

ct.al., Report No. GEAP-4598; July 16, 1964.

12) " Big Rock Point Plant Technical Specifications" Docket 50-155, License DPR-6; Consumers Power Company; Amended May 18, 1984.

13) " Final llazards Summary Report for Big Rock Point Plant", Volumes I and II, Consumers Power Company; November 14, 1961.

l

14) " Big Rock Point Fuel Data Book", According to Operational Reactor Physics Procedure ORP-B-05, updated 10/19/83; Consumers Power Company.

'k I l l IC0984-0006A-RE01

[ 16 C

15) "The Response of UO, Fuel Rods to Power Bursts. Detailed Tests on 5/16-Inch OD, Pellet Fuel, Zircaloy Clad Rods."

W.G. Lussie; IN-ITR-112, Idaho Nuclear Corp., Idaho Falls; January,1970. l l i l t i I l ( l l l l l k IC0984-0006A-RE01

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well ^$ $ Wef.S 3 Y,3s and 3G, fggmeret SYtf 37 as l'on cerrative m ercares were ta k*n ist the sefcctson.f lo v -p ower Doppler c.*e ffic ten t, a nd a rseanbly Io c< / a pseking fu t.r. rt<. tor.p.uo., Appler ese.nicia,,t .F -9.at1 xto ' * % / r is taken from p. I o I th e Bot. <ose for d s fa et et o % vcid a. d d a fuel te s.,p e << t a *. o f / 317 'c, s'>'- re f. 5. 7 s e. a s s embly /se<l /calciny fact'or o f /. t 7 is tales from fie.he *t r o f t e F. C, a hl'h is docu mentafion ef phyr.hs a. con s tants So e //3 tyre 4 ae l. F ic h e # ts Q f re F. C is a tis ti,9 .f local pea kis 9 facf.rr Jar H 3 a. t o '. vo N, s,J ma-feses th e. assa m e d c ore condi ti n s in re f t. e

8scnaottWolag )1 t Cach mea r 4

f. I o f 3.

L ale al< 6an je Cxil.s e f peok deposite A enthafyy, and total deposited suthatty By definition, re<ctor is in tA< pr yt critic <l 6 s t** te wite n f1P abere f = N ' N* k4 ) IN ) a.nl b delayal ne fry (re 9. s Fra eta,.. \\< < r;c = o. 9 89 931 lis,'n Aafa fre m 4,*cl.c. e,2 + fic h e. 14 : srEP 3f_ $ fff 3C $ 1CP gQ S ftp 3 ; Pe a k lla+ nanJb godOceth,%"[/k 6.ilf

o. S t

o. 4'6 c.fr

' l lot B e st Gig en.Valas, h in L122700 _QJJ6 7ff

o. 9 ff// 7

c. 997 98 9

= & N*** 0 9 7 6.'l l.L-A,.'J) B L 2_. _0' $ 90S30 pslayed Alcato'on f~ra e tion, fs,t

c. o o S 6 Si

c. 0 0 S71'l c.ccS7f/

o. c c S 791 keacts'vit, f

-o.co 7260 o.co3/2 0 o.cogjos

o. co gi t f

Rod Drop f)c < 16 t fa, a,., e ters (As apropriate. pr* meters fer stop 37 s f tAs .a ith.fra u si s.*g u e o.c <. a re. listeA below. t) rask lpr s COdbt Ao } 4 *rt/s (jo 2 ak) ti,3 3 li) l1nial1fadial kicing hetor g.1953 iii) Dsy/er Coe Sficie,< t ('% /.r) - 9 24 7 no " iv) Total Delayed Lcation fr<ction, Ps c

o. c os 73.2.

v) /?csen6(y l-c<<t /c aking factor

t. t 7

• >ts i n, t h e b < s,'< r. 2 w. r i I, a c $., e*,.,,, ' g, a - w I y))

kon

[. l troce so rty, spy ll tta cin n,. t n' 2-p, 2 o f.1 fse red uortA e*xver.rian fr om '/. ' N to fo% k pr,ce e / 5 as F.//su.s

• Rad uortA, io%k=RU,,,,,a,'Y-[k,,-{!*]

Tb s, f*r ctep 3 7 : [o. 't Y % * /N) ~[* 177199)- { 10)

• Y. 59 t i '< k - Y.31 m k L(sing Sig.

t. t. I r e f. I,

inte rp ola tiox be 1< se e,, tig e <n t x radi<l PF c arves a f rou a>ud

3. 70 yie t}s a

p. a. e. ( y'/. L in er e eak detosited eau t/s/ y) ef affroum a tely f

ta s. s o *

• f

n ferpofs ficn in.s a ct a fou rol u ertf eeylo n.

s A

• uld le e ela tiveif tu a r.< C*.

ys;ne fig. 12 e f

• re h /, p. A. e. una.f4Ner.! CYr, Dotf er.

f lAs,Ag Sa y. l. 3 o f re f. I, p. d e. <>, a lt of fie r = o. 9 79,P,.c. I Ad.i utiny ti, e f.d. e. for tA< faff er e e }.c 'cie n t l a ri and k7afeA iteatrae fra ct ion : Ad] <.s ted f. A e. =. (13 c.zo)(l. ov.r)d 0 9 79) 'Ij f % i 31. <ta. c 'Ify futtler o d.]a s fn en t a f the p.d? e.. ts a escus< f for loca i f ealciny e f fee ts in <a a ss en Hy ajie(dr :

1., f. - AJJ u e t<d f. P. f. e

( I 31. Y.2) (f. t 7) * *Ih E cl.97 **l/9 t ini tial f.<el en thalpy a F 18**)$ as A s s a ><io,y *n r< 5. I rugger ts, lu ta l d op.. sri e d e n th < f y in t/se la <f t o r

• s u liief frm rad.Jr.y s ecidenl a s de scribed in a

altael,,.,e,it I, is oppra xirra te // 1 1 9 1.s* ' y9 paak.

{o o

• *20derogag A t%cA n e.g n 3 CA =e k /,:rt p ro ebr< / esc ribo / in a t t<< h ane,,t I I.)

A ss He c a3,ee wi tA tA e r a 4 -d r*? a cc ede < t <,, /r't <**n.r po s tufs fed by re f e ct os se / ?

Yes, tsu p a.)

Has the-frofer GAOIl esee m t< tape beca A freAa. [ af a ? utilize.d in 1, n h a,y t e. ca/ cal ation s on

Yes, m.rz 3)

A ee. tAe e de n la cio,, s . pe r fo r,,,e / i,, a t t a d,,, e o, t ,2. e eer ee t ? rt cg '/) Ate & va lu e.s n uu med Fer f4a Deppfer e ce 9Sicieas t, a sse,n &ly loc.ai peJiny factor, and initial fuel antAalpy corr < restive ? %. rtutt f --}}