ML20244D858

Responds to Request for Addl Info Re Plant Hatch diesel-generator Loading Design

ML20244D858
Person / Time
Site:	Hatch
Issue date:	06/08/1989
From:	Hairston W GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-331, NUDOCS 8906200016
Download: ML20244D858 (14)

Text

,

Gerwewe, comw.y.- '

..h ' [33 I!t(l.:7 sVN Avffi.Vi! i

&  :,  ; n;.vu Gm., acn l y, ' ..

• Adehow C4 G3 M " ]

vmqAnw+ '

' t.c owmen cemer newny . 3' FmtCP&v Dox L ?f.

Bounf. tam A:abwen '420!

. iWep%ne if:S SU; %81 ,

":o L; hf rr, t .. l c t /p;~

w. c. natrsion, m Gereof Vege P'etident hJC%JCpOl60M l

HL-331 0022V X7GJ17-H000 l

J '

June 8, 1989 .

U. S. Nuclear. Regulatory Commission l

' Attention: Document Control Desk Hashington, D. C. 20555 ,

l j

PLANT HATCH - UNITS 1, 2 .i NRC DOCKETS 50-321, 50-366 'I OPERATING LICENSES DPR-57, NPF-5 1 REQUEST FOR' ADDITIONAL INFORMATION )

.ON PLANT HATCH DIESEL-GENERATOR LOADING DESIGN 1 Gentlemen:  ;

i In discussions with NRC staff concerning Plant Hatch diesel-generator l operation, several questions were posed by the-staff regarding loading of ;l the diesels under various postulated single failure scenarios. . This  !

inquiry resulted from Georgia Power's January 27, 1987 -submittal i requesting changes to the-plant's Technical- Specifications. The enclosed l information is being provided as a response to those questions.. ,

a He hope this information will be useful to you. Please contact this office at any time if you have questions. ,

i l

Sincerely, j

&l)- h W H. G. Hairston, III l

GKM/eb j Enclosu es:

1. Responses to Questions on Plant Hatch Diesel i Loading Design 1

2. Bechtel Report on Shutdown Capabilities with One Diesel Generator /1 t

.p6! j c: (See next page.) L, g j 1

~8906200016 890608 PDR ADOCK 05000321 p PDC

U. S. Nuclear Regulatory Commission June 8, 1989 Page Two c: Georaia Power Comoany Mr. H. C. Nix, General Manager - Hatch Mr. L. T. Gucwa, Manager Engineering and Licensing - Hatch GO-NORMS U.S. Nuclear Reaulatory Commission. Washinaton. D.C.

Mr. L. P. Crocker, Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaion II Mr. S. B. Ebneter, Regional Administrator Mr. J. E. Henning, Senior Resident Inspector - Hatch l

l l:

l

ENCLOSURE 1 RESPONSES TO QUESTIONS ON PLANT HATCH DIESEL LOADING DESIGN The following information is provided in response to NRC questions on Plant Hatch emergency loading design. The questions were generated in a series of phone calls between December 8, 1988 and February 7, 1989.

ILRC Ouestion Assume there is a design-basis loss of coolant accident (LOCA) in Unit 1 and a loss of offsite power (LOSP) for both units. A single failure takes out one of the Unit 2 dedicated diesel generators ([DGs], either 2A or 2C).

Meanwhile, the swing DG (1B) and the two dedicated Unit 1 DGs (1A and IC) stay with Unit 1, since it is having the LOCA. The LOSP will cause a main -

steam isolation valve (MSIV) closure (reactor vessel isolation). Can Unit 2 be shutdown with the suppression pool temperature (SPT) remaining within applicable limits? Conformance could be shown by either demonstrating that one dedicated D/G is adequate (which may only leave one residual heat removal [RHR] pump and one RHR service water [RHRSH] pump for suppression pool cooling [SPC]), or by showing there is adequate time to re-align DG 1B to the Unit 2 emergency bus and restore a full train of SPC (2 RHR and 2 RHRSH pumps). For either the situation where the swing diesel is re-aligned to Unit 2, or the unit is shutdown with 1 DG, verify the suppression pool temperature remains within limits and that equipment loading is within limits.

1 GPC Resoonse Regulatory Guide (RG) 1.81, " Shared Emergency and Shutdown Electric Systems l for Multi-Unit Nuclear Power Plants", provides guidance on acceptable  !

designs for plants with shared essential equipment. The RG states, ". . . a suitable design basis for multi-unit nuclear plants is the assumption that an accident occurs in only one of the units at a time, with all remaining units proceeding to an orderly shutdown and a maintained cooldown condition." Regulatory Position 2C of RG 1.81 also applies:

1 "Onsite power capacity should be provided to energize sufficient Seismic Category I equipment to attain a safe and orderly cold shutdown of all units, assuming the loss of offsite power and the most severe (in terms of power drain) design basis event and a single failure in the onsite electric system."

The above scenario of a full LOSP, failure of a dedicated DG, and only 1 i

dedicated DG on Unit 2 would be considered an emergency situation. This would cause the operator to use symptom-oriented Emergency Operating Procedures (EOPs) to bring Unit 2 to an orderly shutdown. (Plant Hatch E0Ps l

0022V l HL-331 E-1 l

I ENCLOSURE 1 (Continued)

RESPONSES TO QUESTIONS ON PLANT HATCH DIESEL LOADING DESIGN GPC Response (Continued) are developed from Emergency Procedure Guidelines [EPGs].) In this case, the applicable- suppression pool temperature limits would be based on the heat capacity temperature limit (HCTL) of the suppression pool. The HCTL is related to the b11h pool temperature but includes consideration of local pool temperature limits. Enclosure 2 presents a Bechtel analysis showing a single DG with one RHR and one RHRSW pump is sufficient. to bring the reactor to shutdown and remain within the HCTL. Enclosure 2 was based on realistic assumptions (e.g., the reactor is initially at 1001, power). Loading on the DG is not above its capEbility. Subsequent to the issuance of Enclosure 2, in 1982, four. changes have occurred: (1) The later revisions of the E0Ps have a different HCTL curve (the potential impact is discussed below), and the' 200*F per hour cooldown rate may not be entirely appropriate; (2)

Contrary to the last statement in Enclosure 2, Plant Hatch E0Ps presently do include guidance on appropriate operator actions; (3) Depending on the event (i.e., sufficiently high reactor pressure to automatically open an SRV), low low set logic may initiate a 150 psig blowdown of the reactor vessel rather than manual. operator action being required to do so; and (4) The operator is not instructed on time durations to begin or maintain blowdown. Rather the operator is instructed to use the appropriate cooldown rate to maintain the HCTL.

The specific scenario with one RHR and RHRSH pump was not analyzed in Reference 1, which showed our conformance to NUREG-0783 local pool temperatures during safety relief valve (SRV) discharge. A set of transients and small break LOCAs were analyzed which was expected to result in maximum local SPT, with an assumed initial pool temperature of 95'F. The Reference 2 analysis showed that the elimination of this operating SPT limit did not alter the conclusions of Reference 1.

The original design bases of the plant as delineated in the FSAR do not require GPC to consider transients or accidents on Unit 2 for the scenario of 1 DG remaining and an accident in Unit 1. However, none of the specific events analyzed in Reference 1 had less than 2 RHRSH pumps, and with one DG available, only one RHRSW pump would be available. The HCTL presented in Enclosure 2, and the current HCTL curves (derived from Appendix C calculations for Revision 3 of the EPGs) consider bulk-to-local temperature i differences. The bulk-to-local temperatures when following symptom-oriented EPGs may be different than that calculated in the Reference 1 event-oriented analyses by taking credit for different procedures (e.g., depressurizing through different SRVs). However, following the EPGs and the HCTL, local temperatures should remain below the NUREG-0783 limits. Note that Revision 4 of the EPG's will modify the HCTL curves and will shift the focus from maintaining a subcooled pool to protecting the containment from overpressure conditions. In essence the Reference 3 analyses, which 0022V HL-331 E-2

ENCLOSURE 1 (Continued)

RESPONSES TO QUESTIONS ON PLANT HATCH DIESEL LOADING DESIGN GPC Resoonse (Continued) provides justification to eliminate the local SPT limit, will be used.to raise the curve. This report is under review by the NRC and should be approved about the time GPC formally implements Revision 4 of the EPGs.

NRC Ouestion Assume there is a LOCA and LOSP on Unit 1, with a single failure of diesel battery 1 A. This causes the loss of Division 1 essential equipment. The remaining essential equipment from Division 2 appears to be deficient (e.g.,

1 RHRSW pump is available . on emergency bus 1G due to DG loading limitations). Provide evaluation or calculations for acceptability of these circumstances.

GPC Resoonse The LOCA analyses presented in the FSAR are generally divided into a short term and long term response. During the short term, the analyses consider the. performance of the emergency core cooling systems (ECCS) and conformance to fuel cladding temperature and metal-water reaction limits. For Plant Hatch, the short term analyses cover the first 10 minutes following the LOCA and do not take credit for operator action. During this time, a portion of the decay and sensible heat has been transferred to the containment. For the design basis accident (DBA) double-ended recirculation line break, for example, the core has been reflooded for several minutes, fuel temperature has decreased, and a substantial amount of energy has been transferred to the containment. Requirements for adequate core cooling are now less than they were immediately following the LOCA since the vessel is reflooded and fuel decay heat is decreasing rapidly. After 10 minutes, the design basis takes credit for the emergency diesels' loading being monitored and controlled by the operators. Therefore, credit can be taken for operator action. General Design Criterion (GDC) 38 of 10 CFR 50 Appendix A states the safety goal of containment heat removal systems "shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and maintain them at acceptably low levels." i For the limiting failure of the E emergency bus, concurrent with a DBA LOCA l and LOSP, the FSAR of both units assumes 1 RHR pump, i RHR heat exchanger, i and 2 RHRSH pumps are available for long term containment cooling. Peak I (bulk) suppression pool temperature for this case is about 200*F and occurs about 5-10 hours after the LOCA. This case was reanalyzed with our proposed 0022V HL-331 E-3 l

ENCLOSURE 1 (Continued)

RESPONSES TO QUESTIONS ON PLANT HATCH DIESEL LOADING DESIGN GPC Resoonse (Continued)

Technical Specifications amendment request of May 13, 1988, to delete the 95'F suppression pool temperature limit with basically the same result.

Full flow Net Positive Suction Head (NPSH) requirements for all safety systems pumps are met, even assuming 0 psig in the wetwell air space and conservative inputs like the RHR/RHRSH heat exchanger design fouling factor. Clearly this is beyond the GDC requirements, but within the guidance of Safety Guide 1.1, "NPSH for ECC and Containment Heat Removal System Pumps."

The postulated failure of Diesel Battery A causes loss of power and/or control power to the following equipment:

RHR Pumps A and D*.** (one in each loop)

RHRSW Pumps A and C* (one entire train)

Core Spray (CS) Pump A Plant Service Hater (PSW) Pumps A and C*

• Loss of_ control power only.

• • Unit 2 only. In Unit 1, a control power auto-transfer scheme maintains control power to the breaker which allows for RHR pump D to operate automatically.

The operator can restore power to those pumps which have lost control power only by manually closing the appropriate breaker. In the 0-10 minute time interval (core cooling phase) the FSAR requires 1 CS pump,1 RHR pump in the unbroken loop, and 1 PSW pump to be operational.

If the break is in recirculation loop A, then RHR pump B and CS pump B will provide adequate injection for the short term. If the break is in recirculation loop B, then RHR pump C and CS pump B will provide adequate injection for the short terc. CS pump B, RHR pump B, and PSH pump B are loaded on the "G" 4160 V switchgear, and RHR pump C is loaded on the F 4160 V switchgear,'which are supplied by emergency diesel generators C and B respectively.

For the containment cooling phase (long term), operator action will be necessary to align the loads on the two operating DGs as reported in Table 8.3-14 of the Unit 2 FSAR. A possible alignment is shown in Tables 1 and 2 for Units 1 and 2 respectively. To obtain this alignment, the operator would have to perform the following manual actions:

I 0022V HL-331 E-4

ENCLOSURE 1 (Continued)

RESPONSES TO QUESTIONS ON PLANT HATCH DIESEL LOADING DESIGN GPC Response (Continued) o Manually close the 4160 V feeder breakers for RHR pump D (if LOCA is in Unit 2), and for RHRSH pump C (regardless of which unit the LOCA is postulated to occur in).

o Manually open the PSH crosstie valves for RHRSH pump motor cooling.

o Open the RHRSH crosstie valves.

These actions are called out and/or required by the plant procedures except for manually opening the PSH crosstie valve, which is implied in the procedures (operators are to check for adequate pump motor cooling). It should be noted that the above alignments allow the DGs to operate within their continuous rating of 2850 kW. Emergency Operating Procedures prevent the operator from overloading the emergency buses.

Initially, operator focus is primarily on adequate core cooling (which in the long term can easily be met with any one low pressure pump) and-containment pressure. In this postulated scenario, these parameters would be under control in the first ten minutes, so the operator would then focus on containment cooling. The above loading schemes are consistent with FSAR requirements leaving one low pressure pump for injection, as well as one RHR and two RHR pumps for SPC.

NRC Ouestion Review drawings and check that, for each unit, plant service water (PSH)

Division 1 supplies all Division 1 protected equipment and PSH Division 2 supplies all Division 2 protected equipment. Specifically, we have noted that on Unit 2 the CRD pump 2A room coolers (Division 1) are cooled by Division 2 PSH (Reference FSAR Figure 9.2-6, Sheet 3 of 3). Also, drywell chiller 2B (Division 2) is cooled by Division 1 PSH (Reference FSAR Figure 9.2-6, Sheet 2 of 3).

0022V HL-331 E-5

ENCLOSURE 1 (Continued)

RESPONSES TO QUESTIONS ON ,

I PLANT HATCH DIESEL LOADING DESIGN GPC Resoonse The essential equipment has been reviewed for PSH supply, and Division 1 and Division 2 equipment is properly separated. In other words, a complete failure of either division of PSH, still leaves sufficient essential equipment to cool the core and to bring the unit to a safe shutdown condition. The CRD pump room coolers and drywell chillers are not essential equipment.

Note that the original loop-selection system design of the RHR low pressure coolant injection (LPCI) design was modified in the 1976-1977 time frame.

Changes were made to the power supply of the RHR pumps (as well as other changes). These changes switched the divisions of RHR pumps D and C to Divisions I and II electrically but RHR pump D still receives Division II PSH cooling watv while RHR pump C still receives Division I PSH cooling water. However, the LPCI modification received extensive review and approval by the NhC (References 4 and 5), and was the subject of revised failure mode and effects analyses (FMEA).

References:

1. NEDC-24371-P, "Edwin I. Hatch Nuclear Power Station 1 and 2 Suppression Pool Temperature Response," October 1981.

2. Letter, R. P. Mcdonald to U. S. NRC, " Request to Revise Technical Specifications: Suppression Pool Temperature Limit," dated May 13, 1988. (This submittal included GE Report EAS-19-0388, " Elimination of the High Suppression Pool Temperature Limit for Plant Hatch Units 1 and 2," as Enclosure 4.)

3. NED0-30832, " Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge with Quenchers," December 1984.

4. Letter, George Lear (NRC) to GPC, " Issuance of Operating License Amendment 31 to Edwin I. Hatch Nuclear Plant Unit 1," dated March 30, 1976.

5. Letter, George Lear (NRC) to GPC, " Issuance of Operating License Amendment 48 to Edwin I. Hatch Nuclear Plant Unit 1," dated January 4, 1978.

0022V HL-331 E-6

'l L- ,' 4 TABLE 1 UNIT 1 POSSIBLE LOAD DISTRIBUTION OF EMERGENCY BUSES (LOSP,10-60 min POST-LOCA, DIESEL BATTERY 1 A FAILURE)

Bus 1E Bus 1F Bus 1G Pumo Service Pumo ho Pumo ho Pumo ho CS -

18 - 1275 RHR ID 1125 1B - NR RHRSH ,

1C 1195 18 - NR 10 1195 PSH 1D 645 1B - NR 600-V loads 20 635 Total' horsepower' 0 hp 2985 hp 3105 hp Total kilowatts 0 kN 2475 kW 2575 kN Note:

NR - Operable, not running.

Efficiencies used Diesel Generator in FSAR Tables: Ratinas (1C)

CS 0.90 2850 kN continuous RHR 0.90 ---

2000 hr RHRSH 0.90 3250 kW 168 hr PSH 0.90 3250 kN 30 min 600-V loads 0.90 Unit 1 RHR Pump 1D has a control power auto-transfer switch that will transfer control power for the pump breaker from diesel battery 1A to diesel battery 18 on loss of voltage at diesel battery 1A. This modification was not made on Unit 2.

All values were extracted from FSAR Tables 8.4-9 through 8.4-12.

0022V 4

_ _ _ _ _ _ _ _ _ __ d

g TABLE 2-UNIT 2 POSSIBLE LOAD DISTRIBUTION OF EMERGENCY BUSES (LOSP,10-60 min POST-LOCA, DIESEL BATTERY 2A' FAILURE)-

Bus 2E Bus 2F Bus 2G.

j_'. Pumo' Service Pumo ho Pumo ho Pumo ho CS -

28-1000

'RHR- 2D 1080 2B -~NR RHRSW 2C 1220 2B NR 2D 1220 PSW 2D 600 :2B - NR.

600-V' loads 187 1009.

Total horsepower 0 hp 3087 hp 3229 hp Total kilowatts 0 kW 2477 kN 2613'kN Note:

NR - Operable not running.

Efficiencies used Diesel Generator in FSAR Tables:- Ratinas (2C)

CS 0.93 2850 kW continuous RHR 0.93 3100 kW 2000 hr RHRSH 0.93 3250 kW 300 hr-PSW'- 0.93 3500 kW 30 min 600-V loads 0.93 All values were extracted from FSAR Table 8.3-14.

0022V

ENCLOSURE 2 Becitel Power Corporaton -

sog ,s - co . . s 15740 Shady Grove Rcad Gatthersburg, Maryland 20877 301 - 258 3000 December 17,15)v Mr. L. 3. Long Southern Company Services. Inc.

. P. O. Box 2625 Sirmingham Alabama 35202 E. 1. Hatch Nuclear Plant Units 182 3echtel Job 6511 Shutdown Capabilities with one Diesel File A29.2/9810/B-S$-11751

Dear Mr. Long:

We have completed the analysis which investigates our plant shutdown capabilities utilizing only those components su), plied by one diesel generator.

The results of the analysis have shown that there are two viable methods of cooldown which can be utilized, assuming a loss of ofIsite power with only one diesel operational. They are as follows:

1. Rapid depressurization - This mode is presented in GE's REDO-24372

" Minimum Systems Required for Safe Shutdown During a Fire in Edwin I.

Hatch Nuclear Power Station Units 1 and 2". System IV assumes that RCIC is operable along with SRVs, and one loop of RER in 1.PCI. The RCIC is used to maintain the reactor in het shutdown conditions.

RCIC alone will not prevent relief valve discharge and cycling until approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after scram. Therefore, per EPG it is necessary to prevent excessive SRV cycling by manual actuation of one SRV to depressurite the vessel to approximately 150 psi below the minimum SRV setpoint. RCIC and one SRV are assumed to initiate when reactor water level drops to level 2, the automatic trip level of RCIC. If the EPG is followed. RCIC may be initiated prior to the level trip. Once RCIC and one SRV are initiated, the reactor pressure is reduced rapidly. However, the reactor water level remains above the TAT through the transient. This is because the RCIC flow exceeds the SRV flow at approximately 1825 seconds.

The peak pool temperature for this case is approximately 155' T which provides sufficient margin to the local poc1 temperature limit. The suppression pool pressure is approxir.stely 8 psig. which is well within design limits.

~ - ----____mm___ __

~ ENCLOSURE::ec 2 (C0T.) htel Pbwer Corporation a

' Mr. L. 3. Loss Decod er 17, 1982 3-SS=11751 At 4000 seconds the reactor pressure is appresimately 144 pois. < Tbs depressurisation rate is also very slow. The reactor water level.is at 32 f t, which is a below t.ormal water level but above the fuel sons. Hence, it is necessary to manually actuate an additional SRV to allow operation of LPCI to restore resetor inventory. Shutdown Cooling may be initiated when reactor presevre is below 135 peig and reactor water level is returned to normal or above.

The drawback of the method above is that high thermal stresses (due to the rapid cooldown rate) result in the reactor vessel during depressurisation.

2. Controlled depressurisation - This method utiliseu either the IPCI or RCIC System and the manual operation of a SEV for reactor level, pressure, and temperature control and the RF:R System'in the suppression pool cooling mode (with one RRR service water pump) to control the suppression pool temperature. This mode is based on the utilization of the torus heat capacity temperature limit which is presented in.the BWR Owners Group Emergency Procedures Guidelines.

The scenario is as follows: at t=0 a loss of offsite power occurs automatically causing-a turbine trip -reactor scram, containment and reactor isolation and diesel start signals.

One diesel starts. one diesel is assumed down for maintenance and one diesel either fails to start or swings to the other unit.

The SRVs automatics 11'y prevent overpressurisation. In order to prevent excessive cycling, one or more $RVs is manually actuated to depressurize the vessel to 915 psia within two sinutes of t=0. This

- is conservative since the low low set systen will automatically actuate to control pressure even sooner.

The RER heat exchanger is put into operation in the suppression pool cooling mode at t=10 min. The RCIC/HPCI suction is assumed to automatically shift suction from the condensate storese tank ta the suppression pool on high level.

Manual depressuritation begins at t=20 minutes after the operators have ascertained that the second diesel cannot be started. A

' cooldown rate of 200* F/hr has been shown to be adequate to reach cold shutdown without overheating the suppression pool as is shown on Figure 1. The 200* T/hr cocidown rate is in excess of the 100* F/hr normally allowable cooldown rate but is far below the cooldown rate utilized in NED0-24372 and is therefore considered seceptable in this off design circumstance.

EIM 10$

Dc.to _ -

T. M. nitton - !

ENCLOSURE 2 (CONT.)

BechtelPower Corporation 3

Mr. L. 3. Less '

December 17, 1982 3-55-11751 The following majer s'guipment is assumed to be available regardless of which amargency diesel generator is operational

1. 11 SRVs

2. Suppression pool with minimum water inventory.

Condensate storage tank with 100,000 gal 3.

4 '. 1 RER pump at 7700 spa

5. 1 RER service water puer,i at 4000 gym

6. 1 RER heat exchanger, started at 10 min.

7. 1 plant service water pump and ECCS corner room cooler

8. RPCI and/or RCIC initially operates to restore reactor level. At some tima (assumed to be within one minute) either RPCI or RCIC (selection is random) isolated due to high room temperature since

• one division of coolers has no plant service water for cooling.

The following start conditions are assumed:

1. River water temperature is 95' F.

2. Suppression pool temperature is 95' F

3. Condensate storage tank tenparature is 95*

4. Reactor is initially at 100%

It must be understood that if the 'l' diesel is the only available diesel the essential 600 V electrical buss associated with the cooldown components to be used must be manually transferred to the 3 diesel.

To the best of our knowledg's th'a plant Ratch operations department does not at this time have in place operating procedures which provide the operator with the necessary guidance to utilise these methods. They would have to ba written prior to credit being taken for their use.

i If you have any questions or cessants, please advies.

Very truly yours, h*

Robert A. Glasby Project Engineer RAG KLE DED 1f f

All with Enclosure M :;7 - ;---

cct J..R. Jordan y L. T. h ews 3,gp[0 g 7 gg R. C. Nix -

R. L. Sumner

3. K. Saxley ,

W. F. Garner T. M. Milton .

-i e

,. . ENCLOSURE 2 (CONT.) .. . -- . l

- , 4 TORU6 TEMPERATURE RISE SHUTDOWN WITH 1 RHR SW PUMP

~ . - - . . . . . - - . _ . . .

l a

- i

...-_.s..',.,-- ..

. /' *, HEAT CAPACITY TEMPERATURE LIMIT

. 100 \ -

...',/.. /y '

g,o  !

, /y',M

. s IF SUPPRESSION POOL CANNOT EE - , .-

f ,Q i MA!NTAINED BELOW LIMIT, MAIMTAlW n RPV . PRESSURE SELOW LIMIT, g .g f, b'-

5

\

. . _. .% y b.D . /,,/,e no .,

,/.,,*_..

3

.s .h.

/ o< * **/.*'s_, .- - - - .

,y,

.a 16 0 \ ,- -

ie f .,. ' ' , y. , '.<.

,. -/ p/ , - ,

- - - - . - . ~ . - - - -- -

iso . ppd s. .% - . ..

. r. ,. . .. ; .

<,*g 'd g . s e' .

....._.5, -.....-.- . .

go '

. m' ,

.. _... %. s %

130 - - -.-- - - . - .

N--% - - -

'4* 's .

% ,% s 120  ;

N s e "ef0-3 __..._.3 11 0 - - - -

- - - - - , @j, .

, j 8J i

o goo -- .. ...; '. ..: . . . . . . . . . . . - 4's 4 4 y . .. - -. . . . .

. st? N .

s '

90 o too 100 800 400 soo 6 ors 100 . 800 900 1000 R E ACTOR ' PRESSURE ,'PRX (PSIA) i FIGURE i

~

,,-[-$4

^

scene#

1 t a/'

l-e e .

N. a . C\i --- - . . _ . .