ML20008E424

Preliminary Hazards Summary Rept for Yankee Rowe.Includes Revisions Through 581201

ML20008E424
Person / Time
Site:	Yankee Rowe
Issue date:	12/01/1958
From:	YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20008E423	List: ML20008E422 ML20008E424
References
NUDOCS 8101070133
Download: ML20008E424 (292)

Text

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YANKEE ATOMIC ELECTRIC COMPANY

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LICENSE APPLICATION

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s TECHNICAL INFORMATION t

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HAZARDS

SUMMARY

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I ACKNOWLEDGEMENT l

This report was prepared as a joint effort by the personnel of Yankee Atomic Electric Company, Westinghouse Electric Corporation, and Stone &

Webster EngineeIing Corporation.

Special technical assistance was provided by Professor James M. Austin I

and Dr. Theos J. Thompson of Massachusetts Institute of Technology and Dr. Shields Warren of the New England Deaconess Hospital.

I

1 Hydrogen analyses of the pressurizer vapor during steady vented operation have chown from 325 to $10 ce(STP)H /kg condense.te which corresponds to hydrogen 2

pressure in the vapor of from 0 91 to 1.h3 psi. These results confirm the adequacy of the theoretical approac'.., considering the sensitivity of the calcu-lattens to the assumed vent flow rate.

~

C.

Hydrogen Concentration in the Carbon Steel Base Metal Only steady state values of the H e neentration distribution need be 2

calculated, using the above sources of H, sin e these are the worst cases, i

2 and are shown to be less severe than those for the pressure vessel.

1 1.

Steady State Concentratien From Diffusion Eates

• he steady state concentraticn of H at the inner surface of the 2

carbon steel (where it is the maximum) is calculated from ll.h%.X/D C

=

where H en ntration in carbon steel at inner surface; ppm C

=

g 2

%=H fluxduetocorrosien,ce(STP)/cm-hr 4

2 X

. vessel thickness, em

=

diffusioncoefficient,em/hr D

=

The values of the quantities and results of the calculations are given in Table III-1.

It can be seen that the calculated H neentration is 2

1cwer in the pressurizer than it was calculated to be in the pressure vess?l.

2.

Steady State Concentrations From H S lubility 2

-At cperating conditions, it was shoin that s18nificart H pressure can 2

exist in.the pressurizer, and hence it must be shown that these pressures do not result in significant dissolved H at the arbon steel surface.

2 III-5

l

'Iable III-1 I

,ueady State H Concentration Calculations Parameters and Ileculte 2

For Yankee Prescurizer C_og ating Crnditiens Shutdown Crnditions e:, em 7.95 7.95

/

riF/"r)/c/-h1 3.?6 x lo-(0.1062)(C )~1' ' (3. % x 10~ )@)

l g

l D,"

em /hr 0.j56 8.61 x 10~

an 0 0$$

1.f5 r

C.

ppm (Pressu m Vessel) #

0. 4.2 1.99 l

~

. ?O )/ i

2 uhrve 200'C em /hr

>)

P 5.05 l

2

-7820/hT 2

t D

n.32 x 10 xe belev 200'C em /hr i

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I (b)

See WCAF-2855 for rationale (c)

Frrm WCAP-2855 I

9 P

111 6 1--

c For this purpose, use is rnade of Sievert's equation relating H 2 solubility to H Pressure.

From WCAP-2855 2

59.4 e -6;60/Fr 1/2 C

~

p

=

o with

~

hydrogen partial pressure, atmospheres p

=

hydrogen solubility in iron, ppm C

=

g At the maximum calculated hydrogen pressure of 80 psi, C 0 713 ppm.

=

g and at normal operating conditions with venting (p % 2 psi) C 0.11 ppm.

=

g 3

Transient Calculation of Hydrogen Concentratien During Shutdown It was noted in WCAP-2855 that the hydrogeri concentration in the steel at the inner surface (maximum) can be conservatively estimated in the transient condition from ct (t )dt C

(1)

=

(t-t )l/2 (7f D) /,,Jo

~

where / (t )-is the H flux into the metal as a function of time.

2 To estimate C after a four month shutdown, we must make use of the g

fact that % (t') decreases as C increases (as noted above). ' Assuming g

/ (t )

/ss + (k ~ ss) e.

(2)

=

= steady state flux (u 0.047 % )

where

%(

initial flux (

9 36 x 10~

ccH (STP)/cm -hr)

=

2 it can be shavn that (OD)!

! + (pt

%ss) e ~

k ~ !

!e~Idy

= 2 %ss

~

C t

y O

(3)

IIT-7 J

In order to evaluate k, we assume that the H flux into the 2

steel after four months is 0.15 % i. Fran this, and Equation (2)

~1

~

From Equation (3), C, = 0.84 ppm k is found to be 7 72 x 10 hr after four months. The original assumption that p=0.15piafter four months can be checked by evaluating f = 0.1062 C-1.63 with C = 0.84 ppm.

It in found that f, which is the fraction of corrosion hydrogen entering the steel, is 0.141 when C = 0.84 ppm, in good agreement with the assumption that f = 0.15 This is a slightly dif-ferent approach to transient calculations than vac presented in WCAP-2855 but it is felt to be more accurate.

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2 IV.

Effects of Hydrogen An analysis of the possible effects of hydrogen on the mechanical properties of the A302 B steel was reported in WCAP-2855 for the reactor vessel case. The same criteria apply to the pressurizer.

The maximum hydrogen levels from corrosion in the pressurizer material are lover, particularly at operating conditions, where in the reactor vessel the steady state concentration maximum was 0 32 ppm, and la the pressurizer, 0.0855 At shutdown conditions the maximum attained at steady state would be only ppm.

1.65 ppm, with a buildup to ~.84 ppm expected for a h month shutdown.

Because significant H pre sur an exist in the p~ essurizer at operating con-r 2

ditions, the amounts were calculated.

With normal venting, the H pr ssure is l

2 about 2 pai corresponding to only 0.11 ppm.

Even if not vented, the maximum H2 pressure is 80 psi, corresponding to a hydrogen concentration in the steel of 0.713 ppm.

None of these conditions are cause for any concern with respect to degradation of mechanical properties of the steel or integrity of the pressuri-cer.

I i

IV-1

V.

FATIGUE LIFE CONSIDERATI0'G l

1 An analysis has been made of the effect of the postulated cracks in the base metal on the fatigue life of the pressurizer during its subsequent operation.

The fatigue evaluation has been conducted assuming various strength reduction factors to determire the taxicum existing crack depth which can be tolerated in the vessel vall.

The spherical and cylindrical portions of the vessel in both the steam and water phase region of the pressurizer were evaluated in this analysis. The mnvay flange, transition sections between the heads and shell, and the nozzle penetrations were given apecial consideration since they have a stress concentration factor associated with their e eometry.

From this evaluatien it was found that the highest stressed area was around the nozzle penetration in the cylindrical portion of the vessel.

This region was then used to complete the fatigue analysis.

'Ihe following pressure and te=perature transients occurring si.maltaneously were utilized in this investigation:

No. of A Temperature Transient Cycles Duration Steam Region Water Region 1 Pressure

("F)

(*F)

(psi) 1.

Heat-up (100*F/hr.)

60 5.66 hrs.

-5

-5 2000 2.

Cool-down (100*F/hr) 60 5.66 hrs.

5 5

-2000 3

+10% step change in 5,000 100 sec.

11 12

- 120 power 200 see.

-6 23 80 h.

-10% step change in 5,000 25 see.

-6 13 100 power 250 sec.

11 13

-100 5

Loss of load & loss 10 15 sec.

-32 100 500 of flow 30 sec.. 100 100 500 10 min.

27 27

-1070

6. -Reactor scram 100 15 sec.

5 5

60 30 sec.

-3 14 40 7

Control rod position 22,000 1 min.

-3 5

-3 5 50 changos 1 min.

35 35 50 8.

On-off heater cycles infinite 2 hrs.

+ 25

i The values listed in the "o Temperature" and ".3 Pressure" columns are the maximums l

that occur during each transient. The "3 Temperature" column contains the actual difference between the mean temperature and inside surface temperature of the vessel wall.

The above transients chosen for this analysis are conservative estimates covering i,

all modes of normal and emergency operation believed to be significant in the operation j

of the plant. The specified number of occurrences for each transient is a conservative prediction of the number of occurrences that could be anticipated in the remaining a

operational life of the pressurizer.

These estimates are based on the number of transients the plant has experienced to date.

s j

An added degree of conservatism has been introduced into the analysis by the i

assumption that the accident transients will occur at the end of core life.

At this j

time during plant life there is a reduced plant operating temperrture relative to the pressurizer temperature which results in larger transient temperature changes.

The stress fluctuation at the tip of the postulated crack which were assumed to exiat in the highest stressed areas of both the water and steam phase regions of the vesuel vere then calculated for each transient condition.

These stresses were computed "a a function of the fatigue strength reduction factor (K) associated with the postulated crack depth, the pressure fluctuations, and the temperature changes.

These stress fluctuations are listed in the following tabu 3ation.

pressure Stress Thermal Stress Transient psi Steam Region Water Region Cycles psi psi 1.

Heat-up 16 700 K

-1330 K

-1330 K 60 2.

Cool-down

-16,700 K 1330 K 1330 x 60 3

+10% power step

- 1,000 K 3480 K 3760 K 5,000 i

665 K

-1885 K 7210 K 4.

-10% power step 835 K

-1885 K 4080 K 5,000 i

-835 K 3330 K 4080 K 5

Loss of load & loss 4,180 K

-10,000 K 31,500 K 10 of flow 4,180 K 31,500 K 31,500 K

-8,950 K 8,480 K -

28,000 K i

6.

neactor scram

-500 K 1,570 K 1570 K 100 330 K

-950 K 4400 K 7

Control rod position 420 K

-1,100 K

-1100 K 22,000 changes

-h20 K 1,100 K 1100 K 8.

On-off heater cycles

+210 K infinite

~

Various values fer the fatigue strength reduction factors vere then applied on a i

trial and error basis to establish a peak combined strass fcr each transient condition.

t For each value of assured strength reduction factor, the minimum and maximum values i

of fluctuation stress verc deter-ined by combining the transient stress conditicns in a tanner to establish the maximum stress difference for the predicted number of cycles.

In this nanner, it was determined that a fatigue strength reduction factor of 9 25 may i

be tolerated at the tip of a crack and still raintain the cumulative fatigae usage factor of the water phase region of the vessel below the a' lovable value of 1.

Likewise, a fatigue strength factor of 15 5 can be tolerated at the tip of a crack in the ctea-phass region of the pressurizer.

From reference (1) it was deter =ined that the depth of a crack on the vall surface i

required to develop the fatigue strength reduction factor of 9 25 in the water phase region of the vessel is 0 35 inches.

Similarly the surface crack depth permitted in the steam phase region of the pressurizer is in excess of 0.6 inches. 'Iherefore, it is concluded that a fatigue problem associated with the cbserved cracks in the Yankee F.ove pressurizer is unlikely.

(1)

Innger, B.

F., " Design of Pressure Vessels for Low Cycle Fatigue", ASE Paper 61-WA-16 presented at the ASE Winter Annual Meeting, New York City, Nove=ber 26, 1961.

. _ _. _ ~ _ _ _ _

4 h

4 VI.

Integrity Analysis _

4 A.

Transition Temperature Approach As part of the engineering analysis of the Yankee plant, the hydrostatic c

test temperature was determined by use of the transition temperature approach.

t l

Essentially the hydrostatic test temperature is at the Fracture Transition Elastic temperature (FTE) which is accepted as being NDTT +60 F (nil ducti-lity transition temperature +60 F). The nil ductility t ransition tempera-

. ture is determined by the drop veight ta nt, ASTM-E-208.

However, accepted l

engineering practice is to use a Charpy V-notch fix of 30 ft, lbs. to deter-mine the NDTT of SA302 Grade B because of the established correlation be-tween NDTT and Charpy V-notch as stated in Section III and PB 151987 (Tente-l tive Structural' Design Basis for Reector PNssure Vessels and Directly Associated Components).

The B & W resistance veld clad was applied to SA302 Grade B plate material d

for the Yankee pressurizer shell course sections and the top and bottom heads.

B & W specified a hydrostatic test temperature of 70 F for the Yankee pres.

surizer based on the transition temperature approach. This hydrost

'c test 0

temperature was determined adding'60 F to the temperature at which 30 ft. lbs.

was obtained during notch toughness tests, i.e. FTE.

Considering'the same conservative criteria for possible increase in tran-sition temperature due to hydrogen in WCAF-2855, an additional 40 F should f

be added t.o the FTE before operating stresses are applied. The recommended minimu:.1 pressurization temperature for the Yankee pressurizer is FIE +h0 F or 110 F.

The actual operation of the pressurizer provides a minimum pres-surization temperature of approximately 212 F.

l B.

Fracture Mechanics Analysis The likelihood of fast fracture originating at a postulated large crack

-in the vessel wall has been evaluated. JBased on the evaluation below, it.

[

has been concluded that fast fracture of the Yankee pressurizer is not pos-sible.

VI-l

Regardless of how the crack is postulated to form, even a crack k inches lorig, which is entirely through the 2 inch thick vessel head at the connec-tion to the cylindrical section, vill noc propagate at any of the stress and temperature conditions that the vessel experiences.

4 Trwin'sb criterion for fast fracture from a t hrough-crack has been applied

~

in this analysis:

,(psi) in (1)

K

=

1 - } ({g)2 where K = the crack tip stress intensity factor (a material factor), psi M.

a = half crack length (set a_ equal to the vall thickness), in.

s tensile stress normal to the crack, psi

= yield strength, psi l

ya l

l If the calculated value of K is less than the actual K, fast fracture vill not occur.

For a through crack in a vessel vall of length twice the vall thickness, the apprcpriate value of K corresponds to the condition of plane

=

stresa.

If the temperature is high enough, the fracture made is entirely by cheur, that is, the failure is ductile rather than brittle. For SA 302B Charpy V-notch data ebows duct 13e behavior above 200 F.

The temperature of the pressurizer is well above this while it is under load.

Peak values of K, the plane stress intensity factor, are not available, but lover tempera-ture data /2, shown in Figure 1 (open circles) can be extrapolated to show the trend. The peak value (the curve levels off) is probably above 100 Kai / W.

l l

l The pressurizer functions such that saturation conditions exist at all l

tenperatures. The maximum saturation pressure of 2500 psia corresponds to a temperature of 668 F, and a vall stress of about 20,000 psi.

Under peak i

(

conditions, the wall stress may rise to 25,000 psi, but at the same tem-(

perature:

668 F.

i j

I I

l VI-2

(; '

p~

d o J '9 Bal y' ' '7/1/56

. C TABLE OF CONTENTS Pave 1

NUCLEAR REACTOR DESIGN 100:1 100 GENERAL 100:1

l '

101 CORE DESIGN 101:1 Mechanical Design of the Core 101:1 General 101:1 Fuel Assemblies 101:1 Control Rods 101:2 Core Structure 101:3 l

Control Rod Drive Mechanism 101:4 Thermal and Hydraulic Design of the Core 101: 5 General 101: 5 Thermal Design 101:6 Hydraulic Design 101:7 Table 1 - Mechanical Design Data -

Initial Core 1018 Heat Output as a Function of Time Af ter

);

Shutdown, Decay Heat 101:9 Nuclear Design of the Reactor Core 101:10 General 101:10 Table 2 - Nuclear Design Data - Initial Core 101:11 Temperature Coefficient of Reactivity 101:12 Table 3 - Temperature Coefficient of Reactivity 101:13 Pressure Coefficient of Reactivity 101:14 r

Table 4 - Pressure Coefficient of Reactivity 101:15' Doppler Coefficient of Reactivity 101:15 Table 5 - The Dcppler Coefficients 101:16 Void Coefficient of Reactivity 101:17 1

Effects of Plutonium Buildup 101:18 Steady State Effects'on Reactivity g

and Flux Distributions-101:19 i

g L

Moderator Temperature Coefficient 101:19 e

Table 5a - Yankee Core Temperature i

Coefficients at Beginning and End of Life 101:21 Doppler Temperature Coefficient 101:22 Void' Coefficient-101:22 Delayed Neutron Fraction and Neutron Lifetine.

101:22' r

O r

B:la l

7/1/58 1

,-()

Page Table $b - Variation of Delayed Neutron Fraction and Prompt Neutron Lifetime During Core Life 101:23 Effects of Nonuniform Plutoniu=

Jistriaution 101:23 Table 5c - Kinetics Parameters for End of Life Core Assuming Triple Pu Concentration 101:24 Conclusions 101:24 102 CRITICALITY CONSIDERATIONS 102:1 Cold Reactor at Beginning of Life 102:1 Table 6 - Reactivity (Control Rods Withdrawn) 102:1 Hot Reactor with Chemical Neutron Absorber 102:1 Xenon Transients 102:1 Handling of Fuel 102:4

,em

's i,

B:2 l

7/15/57 l

t O

Page 103 CRITICAL EXPERIMENTS 103:1 General 103:1 Part Core Critical Experiments 103:2 Initial Loading of the Reactor 103:3 Reactor Start-up 103:4 Rearrangement of Fuel Within the Reactor Core 103:6 104 CONTROL 104:1 General 104:1 Control Rods 104:5 Chemical Control 104:8 105 REACTOR CORE EVOLUTION 105:1 200:1 2 -PLANT-DESIGN 200 GENERAL 200:1 201:1 201 REACTOR PRESSURE VESSEL 202 MAIN COOLANT SYSTEM 202:1 Function 202:1 l

Genera 1' Description 202:1' l

Basis for. Design 202:1 203 INSTRUMENTATION AND CONTROL 203:1 Function.

203:1 General Description

. 203:1 Basis for Design 203:4 f

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---+.is

+g gy e

Ba 12/1/5p s

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Page 204 PURIFICATION Sll TEM 204:1 Function 204:1 General Description 204:1 Basis for Design 204:1 205 CHARGING AND VOLUME CONTROL SYSTEM 205:1 Function 205:1 General Description 205:1 Basis for Design 205:2 206 PRESSURE CONTROL AND RELIEF SYSTEM 206:1 Function 206:1 General Description 206:1 Basis for Design 206:1

()

207 DECONTAMINATION SYSTEM 207:1 Function 207:1 General Description 207:1 208 WASTE DISP 03AL 208 :1 Function 208:1 General Description 208:2 Basis for Design 208:2 Gaseous Wastes 208 7 Combustible Solid Waste 208:10 Noncombustible Solid Waste t

208:10 Equipment Capacities and Ratings y,

gy q'sO c.

208:11

. I d' 209 SHUTDOWN COOLING SYSTEM Y./g?S 209:1

?c. <?k," b g-)g

~

N, '<g:' ffd Function (i

e w

209:1 1

!?

\\'_

4 209:1 General Description j

Basis for Design D/

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209:1 t

B:4 2/27/57 0

zm 210 MONITORING AND ALAHM SYSTEM 210:1 Function 210:1 General Description 210:1 211 RADIATION SHIELDING 211:1 Function 211:1 General Descrip tion 211:1 Basis for Design 211:3 212 CHEMICAL SBUTDOWN SYSTEM 212:1 Funetion 212:1 General Description 212:1 Basis for Design 212:1 0

213 VAPOR CONTAINMENT 213:1 Function 213:1 General Description 213:1 Details of Vapor Container 213:3 Vapor Container Tests 213:4 Air Pressure Test 213:4 Leakage Detection Test 213:4 Leakage Rate Test 213:5 Continuous Leakage Indication During Operation 213:5 214 VAPOR CONTAINER VENTIL&TT'S SYSTEM 214:1 Heat Removal Gystem 214:1 Heating System 214:1 Purging 'fentilation 214:1 V-Plant Stack 2142

B: 5 2/27/57 (m

U

_P_an 215 FUEL HANDLING SYSTEM 215:1 Function 215:1 General Description 215:1 216 STEAM-ELECTRIC PLANT 216:1 Funetion 216:1 General Description 216:1 Grading and Fencing 216:1 Footings and Foundations 216:2 Structural Steel 216:2 Siding and Roof Deck 2162 Floors 216:2 Turbine Generator 216:3 Condenser 216:3 Lubricating Oil System 216:4 Circulating Water System 216:1+

Condensate and Feed Water System 216:5 (s wl Service Water Supply 216:5 Cooling Systems 216:6 Secondary Plant Drains 216:7 Compressed Air Systems 216:7 Piping 216:7 Controls and Instruments 216:7 General Description of Electrical Equipment 216:8 Main Transformer 216:9 Station Service Transformers 216:9 Fire Alarm 216:9 Miscellaneou. Electrical Eruipment 216:9 Air Conditioning 216:9 Ventilation Systems 216:10 Heating Systems 216:11 Sources of Steam for Heating 216:11 Drainage Systems 216:11 Potable Water 216:12 pd-

B:6 2/27/57 O

Page 217 FIRE PROTECTION SYSTEM 217:1 Yard Hydrants 2171 Transformer Fire Protection 217:1 Lubricating Oil Room 217:1 Area Beneath T:1rbine Generator 217:1 Interior Hose Station Protection 2171 i

Source of Water 2172 Portable Extinguishers 217:2 Codes and Standards 2172 218 CORROSION CONTROL SYSTEM - HYDROGEN INJECTION 2181 Function 21821

' (3

~>

General Description 218:1 Basis for Design 2181 219 SAFETY INJECTION - SHIELD TANK CAVITY SYSTEM 219:1 Function 219:1 General Description 219:1 Basis for Design 219:1 220 SAMPLING SYSTEM 220:1 Function 220 1 General Description 220:1 Basis for Design 220:1 221 VENT AND DRAIN SYSTEM 221:1 Function 221:1 General Description 221:1

B: 7 2/27/57 f)

\\_/

EGfe 222 COMPONENT COOLING SYSTEM 222:1 Function 222:1 General Description 222:1 Basis for Design 222:1 3

SITE 300:1 300 GENERAL 300:1 Location 300:1 Access 300:1 Population 300:1 Land Use 300:2

{ ';

PublicWaterSupplie(1 300:3 Site Layout 300:3 301 METEOROLOGY 301:1 Pollution C1tnatology of the Deerfield River Site - Report by Professor James M. Austin 301:1 Meteorological Measurements Program 301:24 302 Hydrolocv 302:1 Plant Site 302:1 Deerfield River 302:1 Drainage Area 302:1 Improvements 302:1 River Flow 302:2 303 GEOLOGY 303:1 Plant Site

'303:1 Foundation Design 303:2

( ))

304 SEISMOLOGY 304:1 105 PREOPERATIONAL RADIATION MONITORING 305:1

f i

B:8 2/27/57 i

i O Pace 4

PLANT OPERATION 400:1 i

400 GENERAL 400:1 401 PLANT ORGANIZATION 401:1 i

402 PERSONNEL TRAINING 402:1

)

i Plant Supervisory and Technical Training 402:1 Plant Ope:ator Training 402:1 i

403 INITIAL PLANT INSPECTION AND TESTING 403:1 Manufacturer's Cc=ponent Tests and Inspections 403:1 Plant Systems Inspection and Tests 403:1 Nuclear Star ~ up and Tests 403:1 404 NORMAL OPERATIL MOCEDURES 404: 1 O

409 ExERGENCr PROCEDs.

405:1 406 REFUELING PROCEDURES 406:1 407 RADIOLOGIC _A_L HEALTH AND SAF M 407:1 I

408 PLANT MAINTENANCE AND ACCESS 408:1 409 ROUTINE TESTING PROGRAMS 409:1 Main Coolant System Hydrostatic Testing 409: 1 Steam Generator Testing and Inspection 409:1

?

Radioactivity 409:1 Instrumentation and Control 409:1 Water Chemistry 409:1 Fuel Inspection and Testing 409:2 Performance 409:2 New Core Nuclear Testing 409:2 O

Vapor Container Leakage. Rate Testing 409:2 410 PLANT SECURITY AND SPECIAL-NUCLEAR MATERIALS TRANSFER AND ACCOUNTABILITY 410:1

.. _ - _ _ _. _ _ =

B:9 7/15/57

()

Pace 5 ACCIDENTS AND HAZARDS 500:1 j

500 GENERAL 500:1 501 REACTIVITY ACCIDENTS 501:1 Stact-up Accident 501:1 Cold Water and Boron Concentration Accident 501:2 Loss of Chemical Neutron Absorber 501:3 Continuous Rod Withdrawal at Power 501: 5 502 CHEMICAL ACCIDENT 502:1 503 MECHANICAL ACCIDENTS 503:1 i

Loss of Coolant Flow Accident 503:1 Loss of Water Accidents 503:2 General 503:2 Small Break 503:3 t

Medium Size Break 503:3 Large Break 503:4 Criticality of the Core During Blowdown 503:4 Decay Heat 503:5 Vapor Containment 503:5 Missile Protection 503:9 Loss of Turbine Load 503:10 1

504 CORE MELTDOWN 504:1 General 504:1 i

. Mechanism of Core Meltdown 504:1 Rate of Core Melting 504:1

^

Criticality Consideration Following Melting of the Core 5D4:2

' Fission Product Activity at the End of Core Cycle 504:3 Tasie 7 - Fission ProductLGamma Activity Ft11owing Shutdown 504:4

-( ) ~

Fission Product Release to Vapor Container 504:4

j Bs10 7/15/57 Pace i

50 5 HAZARDS FROM REACTOR ACCIDENT 50 5:1 Maximum Credible Accident 50 5:1 1

Hypothetical Accident 505:1 506 CONCLUSIONS 506:1 O

i B:11 7/1/58

[v')

Following Figure Pace 1

GENERAL VIDI 0F REACTOR PLANT 100:2 2

SIMPLIFIED FLO'd DIAGRAM - PRIMARY AND SECONDARY SYSTEMS 100:2 3

REACTOR CORE CROSS SECTION 101:1 4

REACTOR SECTION 101:1 5

FUEL ASSEMBLY SECTIONS AND DETAILS 101:1 6

CONTROL ROD DRIVE MECHANISM 101:4 7

REACTOR TEMPERATURES - INITIAL CORE 101:7 8

TEMPERATURE COEFFICIENT OF REACTIVITY vs TEMPERATURE 101:14 8A TEMPERATURE COEFFICIENT OF REACTIVITY vs BORON CONCENTRATION 101:14 9

REACTIVITY EFFECT OF VOIDS IN REACTOR CORE C'

(N0 CHEMICAL POISON IN COOLANT) 101:17 b) 9A UNIFORM VOID COEFFICIENT vs BORON CONCENTRATION 101:18 9B AVAILABLE EXCESS REACTIVITY vs CORE LIFE 101:19 90 THERMAL NEUTRON FLUX vs RADIUS AT BEGINNING AND END OF CORE LIFE 101:19 10 EFFECTIVE MULTIPLICATION FA': TOR vs TEMPERATURE 102:1 11 XENON POISON AND RAT 2,OF CHANGE OF k FOR START-UP AFTER MAXIMUM XENON OVERRIDE 102:2 12 REACTIVITY vs NUMBER OF FUEL ASSEMBLIES (NO CONTROL RODS) 102:4 13 EFFECTIVE MULTIPLICATION FACTOR vs NEUTRON ABSORB 3 CONCENTRATION 104:6 14 REACTOR CORE EVOLUTION 105:1 (3

v

l B:lla 7/1/ 58 h

Following Figure Page 15 SECTIONAL VIEW - REACTOR VESSEL AND CORE 201:1 1

16 CONSTANT T PROGRAM 203:2 avg 17 REACTOR CONTROL SYSTEM - CONSTANT Tavg PROGRAM 203:2 1

18 MONITORING AND ALARM SYSTEM 210:1 19 MAP OF PROPERTY OWNED 300:1 20 GENERAL AREA MAP 300:1 1

5 W

O 1

1 1

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i 4

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f-4 1

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B:12 2/27/57 O

ro11 ovine Figure Pare 21 BORING AND SEISMIC SURVEY PLAN 302:2 22 TYPICAL SOIL CONDITION 303:1 23 LARGE GLACIAL BOULDER 303:1 24 LOG OF BORINGS 303:1 25 SOIL BEARING VALUES FOR FOOTINGS 303:2 26 YANKEE PLANT ORGANIZATION CHART 401 1 27 START-UP ACCIDENT WITH CONTINUOUS ROD WITHDRAWAL 501:1 23 REACTOR POWER VS TIME AFTER STEP CHANGES TN REACTIVITY 501:2 29 MAIN COOLANT FLOW VS TIME 503:1 30 REACTOR TEMPERATURES VS TIME AFTER LOSS OF POWER TO ONE PUMP 503:1 31 REACTOR TEMPERATURES VS TIME AFTER LOSS OF POWER TO TWO PUMPS 503:1 I"l 32 REACTOR TEMPERATURES VS TIME AFTER LOSS OF POWER TO FOUR PUMPS 503:1 33 REACTOR TEMPERATURES VS TIME AFTER LOSS OF POWER TO FOUR PUMPS WITH SIMULTANEOUS SCRAM 503:2 34 REACTIVITY VS HEIGHT OF WATER 503:5 35 TOTAL DECAY HEAT GENERATION (BETA AND GAMMA) j AFTER INFINITE TIME OF OPERATION AT A l

GIVEN POWEB 503:5 i

36 PRESSURE RISE IN VAPOR CONTAINER VS ELAPSED TIME AFTER RUPTURE 503:6 j4 37 PRESSURE IN VAPOR CONTAINER FOLLOWING A NUCLEAR ACCIDENT - NO INSULATION ON VAPOR CONTAINER SHELL 503:6 i

38 CORE MELTDOWN VS TIME AFTER RUPTURE - 1/10 FT2 RUPTURE 504:1 2

39 CORE MELTDOWN VS TIME AFTER RUPTURE - 3 FT RUPTURE 504:1 40-FISSION PRODUCT GAMMA ACTIVITY AFTER INFINITE TIME OF OPERATION AT 392 mv POWER 504: 5 n

41 GAMMA DOSE RATE FOR OUTSIDE EXPOSURE 505:2 42 INTEGRATED GAMMA DOSE FOR OUTSIDE EXPOSURE 505:2

3:13 7/1/56 Following Q

Drnvine Pace 646 J 420 OVERALL COMPCSITE FLOW DIAGRAM 200:1 646-J 421 MAIN COOLANT SYSTEM 202:1 646-J 423 PURIFICATION SYSTEM 204:1 646 J 430 CHARGING AND VOLUME CONTROL SYSTEM 205:1 646-J-422 PRESSURE CONTROL AND RELIEF SYSTEM 206:1 9699-RM 41F FLOW DIAGRAM RADI0 ACTIVE WASTE DISPOSAL 208:2 646 J 425 SHUTDOWN COOLING SYSTEM 209:1 646 J 426 CHEMICAL SHUTDCWN SYSTEM 2I2:1 9699 FM-1A ARRANGEMENT VAFOR CONTAINER, PLAN "A-A", "B-B" 213:1 9699-FM-1B ARRANGEMENT VAPOR CONTAINER, PLAN "C-C", "D-D" 213:1 9699-FM-10 ARRANGEMENT VAPOR CONTAINER, ELEVATION "E-E",

"F-F" 213:1 9699-FM-11A VAPOR CONTAINER - SPHERICAL ARRANGEMENT 213:3 9699-FM-12A VAPOR CONTAINER - PENETPATIONS 213:3 9699-FM-19A PROPCSED FUEL HANDLING ARRANGEMENT 215: 1 9699-FM-18A MACHINE LOCATION PLAN - OPERATING FLOOR 216:1 9699-FM-18B MACHINE LOCATION PLAN - GROUND FLOOR 216:1 9699-FM-18C MACHINE LOCATION - ELEVATION 216:1 9699-FM-18D MACHINE LOCATION PLAN - FEZZANINE FLOOR 216 1 9699-FM-22A CIRCULATING WATER SYSTEM - PLAN 216:1 9699-FM-22B CIRCULATING WATER SYS E - SECTI0NS 216:1 w

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216:8 9699-FE-1A MAIN ONE LINE DIAGRAM occ< tio C.

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Bt14 7/15/57 Following O

Drawing Pace 646-J 431 CORROSION CONTROL SYSTEM 218:1 5599-QM-1 SAFETY INJECTION - SHIELD TANK CAVITY SYSTEM 219: 1 9699-QE-1 POWER SUPPLY - SAFETY INJECTION SYSTEM 21933 646-J 428 VENT AND DRAIN SYSTEM 221:1 646-J 424 COMPONENT COOLING SYSTEM 222:1 9699-FY-5A EXCLUSION AREA PLAN 410:1 (v~';

s

10022 2/27/57 The reactor is controlled at operating temperature by 21+ neutron absorbing control rods.

In order to control excess reactivity in the cold reactor made available by its negative temperature coefficient, a neutron absorbing chemical compound is dissolved in the coolant moderator.

Inherent stability and control is provided by the negative temperature coefficient of the reactor, and because the coolant moderator is operated near boiling temperatures.

During large transient increases in re-caused by boiling, decrease re-activity, the moderator voids, Doppler coefficient acts to limit activity.

Simultaneously, the more rapid reactivity transients.

A general view of the Yankee Plant and a simplified flow diagram of the primary and secondary systems are shown in Figures 1 and 2, respectively.

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i GENER AL VIEW OF RE ACTOR PLANT O

REV. 7-15-57

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