ML17298B693

Forwards,For Review as Proposed FSAR Changes,Proposed CESSAR Changes Re Justification for LPSI & HPSI Pump Flow Reductions,Supplementing

ML17298B693
Person / Time
Site:	Palo Verde
Issue date:	12/19/1984
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Knighton G Office of Nuclear Reactor Regulation
References
ANPP-31535-EEVB, NUDOCS 8412260218
Download: ML17298B693 (34)

Text

REGULATORY FORMATION DISTRIBUTION SY M (RIDS)

ACCESSION NBR;841 FACIL:STN"50"528 STN 50 529 STN"50 530 AUTH'AME VAN BRUNTzE E

RECIP,NAME, KNIGHTONgG+N, 2260218 DOC ~ DATE: 84/12/19 NOTARIZED; NO Palo'erde Nuclear Stationi Unit li Arizona Publi Palo'erde Nuclear Stations Unit 2i Arizona Public Palo, Verde Nuclear Station~

Unit 3i Arizona Publi AuT'O'O'R'. AFF ILI AT ION Arizona Public Service Co, RECIPIENT AFFILIATION Licensing Branch 3

DOCKET ¹ 05000528 05000529 05000530

SUBJECT:

Forwards for review as proposed FSAR changesrproposed CESSAR changes re justification for LPSI 8

HPSU pump flow reductionsisupplementing 801218 ltd DISTRIBUTION CODE:

8001D COPIES RECEIVED;LTR ENCL-SIZE'ITLE:

Licensing Submittal:

PSAR/FSAR Amdts 8 Related Cor respondence NOTES:Standardized plant.

Standardized plant Standardized plant.

05000528 05000529 05000530 RECIPIENT ID CODE/NAME NRR/DL/ADL NRR LB3 LA INTERNAL: ACRS 41 ELD/HDS3 IE/DEPER/EPB 36 NRR ROEiMeL NRR/DE/CEB 11 NRR/DE/EQB 13 NRR/DE/MEB 18 NRR/DE/SAB 24 NRR/DHFS/HFEB40, NRR/DHFS/PSRB NRR/DSI/AEB 26 NRR/DSI/CPB 10 NRR/DSI/ICSB 16 NRR/DSI/PSB 19 NRR/DSI/RSB 23 RGN5 EXTERNAL: BNL(AMDTS ONLY)

FEMA-REP DIV 39 NRC PDR 02 NTIS COPIES LTTR ENCL 1

0 1

0 6

6 1

0 1

1 1

1 1

1 2

2 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 3

3 1

1 1

1 1

1 1

RECIPIENT ID CODE/NAME NRR LB3 BC LICITRAzE 01 ADM/LFMB IE FILE IE/DQASIP/QAB21 NRR/DE/AEAB NRR/DE/EHEB NRR/DE/GB 28 NRR/DE/MTEB 17 NRR/DE/SGEB 25 NRR/DHFS/LQB 32 NRR/DL/SSPB NRR/DS I/ASB NRR/DSI/CSB 09 NRR/DS I/METB 12 NR AB 22 REG FI 00 I/MIB DMB/DSS (AMDTS)

LPDR 03 NSIC 05 PNL GRUELiR COPIES LTTR ENCL 1

0 1

1 1

0 1

1 1

1 0

1 1

2 2

1 1

1 1

1 1

1 0

1 1

1 1

1 1

1 1

1 0

TOTAL NUMBER OF COPIES REQUIRED:

LTTR 53 ENCL 45

I'I

~

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li C

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0 ill tl

Arizona Public Service Company r

ANPP-31535-EEVB/TFQ December 19, l984 Director of Nuclear Reactor Regulation Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3

Docket Nos.

STN 50-528/529/530 Post-FDA Proposed CESSAR Changes File:

84-056-026 G.1.01.10

References:

(A)

(B)

(c)

Letter from E. E.

Van Brunt, Jr.,

APS, to G.

W. Knighton, NRC, dated December 18, 1984;

Subject:

Post-FDA Proposed CESSAR Changes.

Letter from A. E. Scherer, CE, to D. G. Eisenhut, NRC, dated December 5,

1984 (LD-84-070);

Subject:

CESSAR Amendment 10.

Letter from A. E. Scherer, CE, to D. G. Eisenhut, NRC, dated December 5,

1984 (LD-84-071);

Subject:

High Pressure Safety Injection Flow.

Dear Mr. Knighton:

Reference (A) requested that a number of previous proposed CESSAR changes be reviewed as proposed PVNGS FSAR changes.

We wish to supplement reference (A) with those proposed CESSAR changes which we had understood to be reviewed on the CESSAR Docket (STN 50-470).

These changes were transmitted b'y references (B) and (C), and are attached for your use.

This request is necessary, since the attached CESSAR changes pertain to justi-fication for Low Pressure Safety Injection and High Pressure Safety Injection pump flow reductions.

Please contact Mr.

W. F. Quinn of my staff if you have any questions on this matter.

Very truly yours, APS Vice President Nuclear Production ANPP Project Director EEVB/TFQ/mb Attachments 84i22602i8 84i2i9 PDR ADOCK 05000528 A

PDR pgyI

r r, t>>

~'

I

'Mr'. G.

W. Knighton

~

Post-FDA Proposed CESSAR anges ANPP-31535 Page 2

cc:

E. A. Licitra (w/a)

A. C. Gehr (w/a)

R. P.

Zimmerman (w/a)

~ ~

pl ~

I

+

~

ANPP-31535 STATE OF ARIZONA

)

)

sate COUNTY OF MARICOPA) k I, A. Carter

Rogers, represent that I am Nuclear Engineering Manager of Arizona Public Service

Company, that the foregoing document has been signed by me for Edwin E.

Van Brunt, Jr.,

Vice President,

Nuclear, on behalf of Arizona PublicService Company with full authority so to do, that I have read such document and know its contents, and that to,the best of my knowledge and belief, the statements made therein are true.

A. Carter Rogers Sworn to before me this day o

, 1984.

otary Publ My Commission Expires:

.My Commission Expires April 6, 1987

~

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p FJIp

'H HP ~ p V

HI I'>>

~,,I

CESSAR ANEl'IDNEHT 10 EXCERPTS FROH DECE%ER 5, 1984 LETTER FRQH A.

E,

SCHERER, CE, TO D, 6, EISEi'IIIUT, NRC (LD-84-070),

TABLE 5. 3. 3, 3-1 5AF":'< l'<JECT 0't Pl.tMP5 Alt<i.".UM )EL.'VER 0 /LOS To~r

(~">>u>>ng One Emergency Qe~e~a<>r RCS Pressure si 1775. 0 165O.O

)440. Q 1270. 0 1095. 0 865. 0 605. 0 310. 0 200. 0 l'5o.a 100. 0

50. 0 0

'o'4 Rate per ln Ai A2 0

0 50.0 SO.O 100. 0 lpp. p 125. 0 125. p 150. 0 150. 0 175 0

175 Q

2OO O

2OO O

225. O 225. 0 234. 0 234 p

58t,o~~ ~p~~~

I~.o~~

i%4~~~

isa~

25@4-&a~,O~

Bl 0

50. 0 lOO. O 125. o 150. 0 175. 0 200. 0 225. 0 234. 0

~ O &38-.9 243. 0 246.0 250. 0 B2 0

Qo P

100. 0 125. o 150. 0 175. 0 200. 0 225. 0 234. 0

~O~

243. 0 246. 0 250. 0 injection Point Al is assumed to be attached to the broken pumo discharge 1 eg.

n

~

'1

TASLt'. 3. 3. 3-2

'( 'ARA)4j~'2 p jn 1

I I

~

%(

<8 ~ a~ (

~ ~q ~ c S/1AL'REAK cCCS P~RFOR.'gHC=

~ v~;

O~ud ll

'I C Reactor Power Level (102".. of Nominal }

Average Linear Heat Rate (102, of Hominal)

Peak Linear Heat Rate Gap Conductance at Peak Linear Heat Rate Fuel Centerline Temperature at Peak Linear Heat Rate Fuel Average Temperature at Peak Linear Heat Ra te Hot Rod Gas Pressure Moderator Temperature Coefficient at Initial Oensity System Flow Rate (Total)

Core Flow Rate Initial System Pressure Core Inlet Temperature Core Outlet Temperature Low Pressurizer Pressure Scram Setpoint Safety Injection Actuation Signal Setpoint Safety Injection Tank Pressure High Pressure Safety Inaction Pump Shutoff Head Low Pressure Safety Injection Pump Shutoff Head

'Ia 1 ue 3876 5.6 15.0 1497 3681 2319 1187 0.0 164.0xl0 159.lxl0 2250 565 523 1600 1600 608 nl '.s kw/ft kw/ft btu/hr-ft psia

~

/0C lbs/hr lbs/hr psia F

=F ps ia os la osia Dslg pslg

O q

I LU I~

Cl) 40 E,O 80 T fi'lL AFTER RUP l URE SECONDS C-E 8%678

~ I 1,0 x DOUBLE Ei!DED GLltLLOTINE OPEAI'AFETY IIIJECT OII FLOI<

NTO INTACT DISCHARGE LEG 0-GL

20 LjP Cu bG 1".O TlfiE AFTER RUPTUPE, SECONDS l,C x BOj'BLE EHOED GUILLOTINE BREAK IW f'Ul".f'ISCflARGE LEG F

I l

TAT

C-E Power Systems Combustion Engineering. Inc.

1000 ProsPect HillRoad Windsor. Connecticut 06095 Tel. 203/688-1 91 1

Telex; 99297 POWER....=

SYSTEMS STN 50-470F December 5,

1984 LD-84-071 Hr. Darrell G. Eisenhut, Director Division of Licensing U;S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

High Pressure Safety Injection Flow

Dear Nr. Eisenhut:

In an effort to provide a suitable technical specification margin for High pressure Safety Injection (MpSI) pump performance for the first System 80'"

plant, a re-analysis of the most limiting small break Loss Of Coolant Accident (LOCA) has been performed.

This small break (0.05 ft cold leg) was selected as the basis for determining the effect of reduced HPSI pump delivery for the following reasons.

(1)

Large break LOCAs are not influenced by HPSI flow.

(2)

This break size and location (0.05 ft cold leg) is the most limiting small break.

(3)

Reduced HPSI pump performance has no impact on the consequences of the non-LOCA Chapter 15 safety analyses.

A comparison of the previous peak clad temperature and two-phase mixture height in the core is attached (Figures 1

and 2).

Also attached is a

CESSAR change that is provided for your review.

It will be incorporated into CESSAR in the next amendment.

A review of Figures 1

and 2 indi cates tht the maximum peak clad temperature for this break size increased from 1557'F (from previous CESSAR analyses) to 1630'F.

This increase is attributed to the slightly longer period of core uncovery resulting from the decrease in HPSI

'flow delivered.

This small break analysis is still conservatively bounded by the tIIost limiting large break LOCA peak clad temperature (2169'F occurs in a 1.0 ft double-ended cold leg guillotine break ).

In summary, a

CESSAR change is forwarded to reflect a reduced HPSI pump flow.

This change was necessary due to as-built conditions in the first System 80 plant.

A re-analysis of the most limiting small break LOCA demonstrates that system performance remains well within the acceptance criteria of

Mr. Darrel1 G. Eisenhut

.... December.5,

1984, LD 071 Page 2

10 CFR 50.46.

Additionally, the higher resulting peak clad temperature remains at least 500'F below the limit case large break LOCA.

The attached change will be included in a future amendment to CESSAR.

If you have any questions or comments, feel free to call me or Hr. G. A. Davis of my staff at (203) 285-5207.

Very truly yours, COMBUSTION ENGINEERING, INC.

A. E. Scherer Director Nuclear Licensing AES:las Attach.

cc:

P. Moriette

FIGURE 1

-- 0-,05-FT -BREAK REDUGED-liPS-I-PUfiPDELIVERY=

PEAK CLAD TB1PERATURE 4.

The four safety injection tanks (SITs) are piped so that each SIT feeds a

single cold leg injection point.

Thus:

a.

for a break in the pump discharge

leg, the SIT flow credited is 100K of the flow from three SIgs.

The remaining SIT is assumed to spill out the break.

b.

for breaks in other locations, the SIT flow credited is lOOX of four SITs.

Table 6.3.3.3-1 presents the high and low pressure safety injection pump flow rates assumed at each of the four injection points as a function of reactor coo1aot system pressure,

~N~<~<p) go,~~

Core and System Parameters

6. 3. 3. 3. 3 The significant core and system parameters used in the small break calcula-tions are presented in Table 6.3.3.3-2.

The peak linear heat generation rate (PLHGR) of 15.0 kw/ft was assumed to occur 15K from the top of the active core.

A conservative beginning-of-life moderator temperature coeffi-cient of 0.0 ~o/'F was used in al,l small break calculations.

The ECCS performance analyses as performed, do not account for steam generator tube plugging which may occur over the plant's lifetime.

(7)

The initial steady state fuel rod conditions were obtained from the FATES computer program.

Like the large break, the small break analyses employed a hot rod average burnup which maximized the amount of stored energy in the fuel Since the small break analysis used a higher PLHGR than did the large break analysis (15.0 kw/ft vs 14.0 kw/ft) the fuel rod parameter values given in Table 6.3.3.3-2 differ from those on Table 6.3.3.2-2.

Because the large break results are always more limiting than the small break results, the small break analysis is run at a higher PLHGR to prevent requiring a reanalysis should the large break results improve.

Since the small break results are goverened mainly by the core liquid level transient (see Results Section below) which is a function of the total core decay heat generation

rate, the higher PLHGR does not significantly affect the small break results.

6.3.3.3.4 Containment Parameters 6.3.3,3.5 Break Spectrum Six breaks were analyzed to characterize the sm~ll break spectrum.

Five

breaks, ranging in size from 0.5 ft t~ 0.02 ft were postulated to occur The 0. 5 ft break was also analyzed for the in the pump d>scharge leg.

large brea)>ipectrum (sect break size One break, ion 6.3.3.2) and is defined as the transition equal in area to a fully open pressurizer safety

/ N E'ER 7 (8) s/n c x C~

6.3-28 The small break analysis does not credit any rise in containment pressure.

Therefore, other than the initial containment pressure, which is assumed to remain constant, no containment parameters are employed for this analysis.

The initial containment pressure was assumed to be 0.0 psig.

INSERT A for the six break spectrum analysis identified in paragraph 6.3.3.3.5.

Table 6,3.3.3-1A presents the safety injection (SI) pump flow rates used ia an alternate analysis of the limiting small break LOCA, the 0.05 ft~ break iin the reactor coolant pump discharge leg.

This break was reanalyzed to demonstrate the acceptability of a small reduction in the SI pump flowrate.

INSERT B

The 0.05 ft2 break which was determined to be the limiting break size and the most sensitive to the SI pump flow capacity was also analyzed using the reduced SI pump flow discussed in paragraph 6.3.3.3.2.

6. 3. 3. 3. 6 valve, (.03 ft ) was postulated to occur in the top of the pressurizer.

2 Table 6.3.3.3"3 lists the various break sizes and locations examined for this analysis.

h Res ults The transient behavior of important NSSS parameters is shown in the figures listed in Table 6. 3. 3.3-4.

Table 6. 3. 3. 3-5 summarizes the im Extant results of this analysis.

Times of interest for the various breaks ana yzed are presented in Table 6. 3. 3. 3-6.

A plot of peak clad temperature (PCT) versus break size is presented in Figure 6.3. 3.3-7.

The o.05 ft break results in the highest clad temperature ~iLCip of the small breaks analyzedg&RM JH8=<Tg<.

0 inc>p,~

I break spectrum is the 0.2 ft break with a PCT of 10304F.

It is important to note the differences in the transient behavior of these two break sizes, because each characterizes differe~t controlli~g features of small breaks.

The larger breaks (between 0,2 ft and 0.5 ft ) temperature transients are terminated by the action of the safety injection tank~ (SIT) whereas the temperature transients for the smaller breaks

(< 0.05 ft ) are terminated solely by the high pressure safety injection pump (HPSIP) prior to the2actuation o] the SITs.

For the intermediate break sizes (approximately 0.2 ft to 0.05 ft ) both the SITs and HPSIP play an important part in terminating the transient, with the HPSIP becoming more important as the break size decreases.

As shown in Figure 6.3.3.3-7, PC( as a function of break size remains fairly constant until the 0.2 ft

)reak.

Then the PCT rises for the 0.05 ft and then falls for the 0.02 ft break.

This rise and fall in PCT can be adequately predicted by observing the transient behavior for breaks less than or equal to 0.2 ft The peak clad temperature is predictably affected by:

1)

Time of initial core uncovery, 2)

Depth of core uncovery, and 3)

Duration of core uncovery.

2 As the break size becomes progressively smaller than 0.2 ft, the inner vessel two phase level follows a definite pattern:

1)

The time of initial core uncovery is later, 2)

The depth of core uncovery is less, 3)

The time of core uncovery becomes

longer, and, 4)

The actuation of the SITs is later during the period of core uncovery and eventually does not occur.

6. 3-29

INSERT C

The.05 ft~ case yeilds a peak clad temperature of 1557'F based on the SI pump

--- -"flow capacities of Table 6.3.3.3-1 and 1630'F based on

%he -SI pump -flow capacities of Table 6.3.3.3-1A.

In either case the result is more than 500'F higher than the other small break cases presented yet more than 500'F below the limiting large breaks reported in Section 6.3.3.1.

This trend continues until the core does not2uncover at a/1.

For System 80 this occurs for a break size between

0. 05 ft and

0. 02 ft (and for all smaller breaks).

As the. break size decreases, both the later time of initial core uncovery and its shallower depth tend to mitigate the temperature transient.

However, the increased duration of u~covery acts in th~ opposite direction.

In progressing from the 0. 2 ft break to 0. 05 ft break the increaseg duration dominates and therefore the peak clad temperatures ris~.

This trend continues until a break size is reached, typified by the 0. 05 ft

break,

~here the three parameters are balanced.

For breaks smaller than this, the increase in time to initial core uncovery and the shallower depth dominate causing less severe temperature transients.

This t~end continues until the core does not uncover as typified by the 0.02 ft break.

Thus, by analyzing several break sizes over this range, the behavior of PCT versus break size can be adequately determined.

To demonstrate the conservatism associated2with the small break ECCS perfor mance results provided herein, the 0. 05 ft break was reanalyzed using a

more realistic measure of the decay heat generation rate.

As required by Appendix K to 10CFR50, the spectrum analysis employed a decay heat generation ra]e equal to 120K of the standard AHS curve.

The reanalysis of the 0.05 ft break used a decay heat generation rate equal to 100K of the AHS curve.

This one change reduced the peak clad temperature ~ morc W~

S'gg'F'.

3. 3. 3. 7 Instrument Tube Rupture In addition to the ~ small breaks discussed

above, the rupture of an in-core instrument tube was consider~d.

A break, equal in size to a completely severed instrument tube (0.003 ft ) was postulated to occur in the reactor vessel bottom head.

Following rupture, the primary system depressurizes until a reactor scram signal and safety injection actuation signal (SIAS) are generated due to low pressurizer pressure at 1600 psia.

The assumed loss of offsite power causes the primary coolant pump and the feedwater pumps to coast down.

After the 30 second delay required to start the emergency diesel and the high pressure safety injection pump, safety injection flow is isitiated to the reactor vessel.

At this time an emergency feedwater pump is also

started, providing a source of cooling to the steam generators.

Due to the assumed failure of one diesel, only one high pressure safety injection pump and one emergency feedwater pump are available.

(Four SITs and one low pressure safety injection pump are also available but do not inject due to the high RCS pressure.)

The steam generator secondary sides also become isolated at this time.

The primary side depressurization continues accompanied by a rise in secondary side pressure until the secondary side pressure reaches the lowest set point of the steam generator safety relief valves.

The primary system pressure continues to fall until it is just slightly greater than the secondary side pressure.

At this point, the flow from the one operating HPSIP (66.3 ibm/sec) exceeds the leak flow (26.4 ibm/sec).

Therefore the

6. 3-.30

I

Table 6.3.3.3-1 SAFETY"INJECTION PUMPS MINIMUM DELIVERED FLOM TO RCS (Assuming one Emergency Generator Failed)

Flow Rate Per Infection Point" (gpm)

RCS Pressure

~si 1700 1581 1483 1349 1199 993 782 605 310 200 130 100 50 0

Al

.5 51.25 76.75 102.75 128.75 155.25 181.50 200.0 225.0 234.0 581.0 1282.0 1884.0 2357.0 A2

.5

51. 25 76.75 102.75 128.75 155.25 181.50 200.0 225.0 234.0 581.0 1282.0 1884.0 2357.0

.5

51. 25 76.75 102.75 128.75 155.25 181.50 200.0 225.0 234.0 240.0 243.0 246.0 250.0 82

.5

51. 25 76.75 102.75 128.75 155.25 181.50 200.0 225.0 234.0 240.0 243.0 246.0 250.0

• Injection Point Al is assumed to be attached to the broken pump discharge leg.

TABLE 6. 3.3. 3-2 GENERAL SYSTEM PARAMETER AND INITIAL CONDITIONS SMALL BREAK ECCS PERFORMANCE ANALYSIS

~uantit Value Units Reactor Power Level (102" of Nominal)

Average Linear Heat Rate (102; of Nominal)

Peak Linear Heat Rate Gap Conductance at Peak Linear Heat Rate Fuel Centerline Temperature at Peak Linear Heat Rate Fuel Average Temperature at Peak Linear Heat Rate Hot Rod Gas Pressure Moderator Temperature Coefficient at Initial Density System Flow Rate (Total)

Core Flow Rate Initial System Pressure Core Inlet Temperature Core Outlet Temperature Low Pressurizer Pressure Scram Setpoint Safety Injection Actuation Signal Setpoint Safety Injection Tank Pressure High Pressure Safety Injection Pump Shutoff Head Low Pressure Safety Injection Pump Shutoff Head 3876 5.6 15.0 1497 3681 2319 1187 0.0 164.0xl0 6 159.lxl0 2250 565 623 1600 1600 600 Mwt kw/ft kw/ft btu/hr-ft"-'F oF oF psia do/'F lbs/hr lbs/hr psia oF psia psia psia pslg Pslg (a)

VQ~ ~~4~ ~ ~~~~@'

o os~

~ XZ~~ ~at ~M~ ~

7M/z 4. 3.9.3-/A

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TABLE 6. 3. 3. 3-5 FUEL ROD PERFORMAtlCE

SUMMARY

SMALL BREAK SPECTRUM Break Size ft 0.50 ft /PD 0.35 ft /PD 0.20 ft /PD 0.05 ft /PD 0.02 ft /PD 0.03 ft /HL e'3 o a'P/<8 Maximum Clad (a)

Surface Temperature

('F) 954 932 1030 1557 995 1012

/$90 Peak Local Zirconium Oxid.

(")

<. 0020

<.0015

<.0041

<.8825

<.0011

<.0011 Q I +/A6 Hot Rod Zirconium Oxid.

(cl )

<.0003

<.0002

<.0007

<.1430

<.0003

<.00004

(.cosy (a )

Acceptance Criteria is 2200'F.

(b)

Acceptance Criteria is 17'.

(c)

Acceptance Criteria is 1.0, Hot rod oxidation values are given as a conservative indication of core-wide oxidation.

(g 8rcak'g~A~ A. S1'~~~ M~W 7~)c

d. 3.9. 3-I 8~~ ~Wg~~ ~~g ~ zz ~g ~rW~

7M/<

6. 3. 3. 3 - ( 8

TABLE 6.3.3.3-6 TIMES OF INTEREST FOR SMALL BREAKS (Seconds)

Break Size

~IO 0.50 ft /PD 46.5 0.35 ft /PO 50.0 c~)

0.20 ft /PD 62.0 2

C+~

0. 05 ft /PO 208. 0 gc) 0.02 ft /PO 492.0

<.)

0.03 ft /HL 585.0 y @gal/P8 4/2' C~)

LPSI Pum On 158.0

~

244 445 a.

a.

a.

SI Tanks On 142.0 204.0 400.0 b.

b.

b.

Hot Spot Peak Clad

~T.

Il 160. 0 235.0 442.0 2010.0 437.0 540.0

/QQO a.

Calculation terminated before time of LPSI pump activation.

b.

Calculation terminated before initiation of SI tank discharge 4.

cf ~<

y>.z~ a ~r~ ~--4 ~-<

lt 6'- 3.z. 3- /

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8&/c-1 z.z.3-/A.

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1 r