ML20085F182
| ML20085F182 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 12/31/1966 |
| From: | SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
|
| References | |
| FOIA-91-17 NUDOCS 9110220071 | |
| Download: ML20085F182 (68) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
\,
G SAFEGUARDS REPORT TOR Tile SAXTON RL : TOR OPERATING AT 35 MWt 9
December, 1966
,' 7
.yui .
- ~ ,1
~s
.AN y .
k, o110220071 > t o a , .;
FTIR F-OIA f l g
fiSTo 91-17 pi g S
j
e b
TABLE OF CONTENTS Pare I. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . 1-1 A. Program Objective and Scope . . . . . . , . . . . . . . . . 1-1
- 8. Program Description . . . . . . . . . . . . . . . . . . . . 1-1 II. POWER ESCALATION PROGRAM . . . . . . . . . . . . . . . . . . . . 11-1 111. PLANT MODIFICATIONS . . . . . . . . . . . . . . . . . . . . . . 111-1 A. Pressurizer Safety Valve Anti-Simmer Devices . . . . . . . 111-1 B. Reactor ' trip Circuit Modifications . . . . . . . . . . . . 111-2 C. Variable Frequency Motor Generator Set Modifications . . . III-4 IV. NUCLEAR AND THERMAL AND llYDRAULIC EVALUATIONS . . . . . . . . . IV-1 A. Nuclear Evaluation . . . . . . . . . . . . . . . . . . . . IV-1 B. Ibermal and Hydraulic Evaluatior. . . . . . . . . . . . . . IV-3 V. FUEL ELEMENT MATERIALS EVALUATION . . . . . . . . . . . . . . . V-1 A. Introduction . . . . . . . . . . . . . . . . . . . . . . . V-1 B. Fuel Performance . . . . . . . . . . . . . . . . . . . . . V-1 C. Zircaloy Creep Rate . . . . . . . . . . . . . . . . . . . . V-2 D. Zitcaloy Corrosion and Hydriding . . . . . . . . . . . . . V-2 E. Summary and Conclusions . . . . . . . . . . . . . . . . . . V-3 F. References . . . . . . . . . . . . . . . . . . . . . . . . V-4 VI. ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . VI-l A. General . . . . . . . . . . , , . . . . . . . . . . . . . . VI-1 B. Reactivity Incidents . . . . . . . . . . . . . . . . . . . VI-l C. Mechanical Incidents . . . . . . . . . . . . . . . . . . VI-3 D. Hypothetical Accident . . . . . . . . . . . . . . . . . . VI-6 VII. INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . VII-l VIII. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . VIII-l IX. APPENDIX . . . . . . . . . . . . . . . . . . . . . . . . . . . . IX-1 i
a
LIST OF TABLES Table No. Title Pare l IV-1 Core Nucle Parameters . . . . . . . . . . . . . . . . . . IV-2 IV-2 Thermal and liydraulic Design Parameters . . . . . . . . . . IV-6 i l
IX-1 Fuel Rod Parameters Rabbit Capsule Irradiations . . . . . . IX-3 l
IX-2 Dimensional Measurements on Fuel Rod Samples from l the Rabbit Capsules . . . . . . . . . . . . . . . . . . . . IX-5 l l
IX-3 Summary of Fuel Capsule Irradiation Experiment i Parameters . . . . . . . . . . . . . . . . . . . . . . . . IX-8 IX-4 Diameter and Length Measurements on Fuel Rod l Sample from CVTR Capsule A-2 . . . . . . . . . . . . . . . IX-10 IX-5 Diameter Measurements on Fuel Rod Samples from CVTR Capsule A-4 . . . . . . . . . . . . . . . . . . . . . IX-12 IX-6 LRD In-Pile Assemblies in Saxton . . . . . . . . . . . . . IX-14 IX-7 Summary of Zircaloy Clad in Saxton . . . . . . . . . . . . IX-15 11
1.15T OF TIGQRES Figure fio . Title Page 111-1 Safety Valve Anti-Simmer Device . . . . . . . . . . . . . III-5 III-2 Modification of Saxton Reactor Trip Circuits . . . . . . III-6 III-3,a Variable Frequency Set Generator Field Transfer Scheme One-Line Diagram . . . . . . . . . . . . . . . . . III-7 III-3,b Variable Frequency Set Generator Field Transfer Elementary Diagram . . . . . . . . . . . . . . . . . . . III-8 IV-1 Change in Saxton Hot Channel Factors . . . . . . . . . . IV-9 VI-1 Continuous Rod Withdrawal . . . . . . . . . . . . . . . . VI-8 VI- 2 Primary Coolant Flow Coastdown Following Loss of Pump Power . . . . . . . . . . . . . . . . . . . . . . VI-9 VI-3 Loss of Load Incident - Steam Generator Steam Temperature . . . . . . . . . . . . . . . . . . . . VI-10 VI-4 Loss of Load Incident - Steam Generator Steam Pressure . . . . . . . . . . . . . . . . . . . . . VI-11 VI-5 Loss of Load Incident - Primary Coolant Temperature . . . VT-12 VI-6 Loss of Load Incident - Nuclear Power . . . . . . . . . . VI-13 VI-7 Loss of Load Incident - Pressurizer Pressure . . . . . . VI-14 VI-8 Loss of Load Incident - Pressurizer Water Volume . . . . VI-15 VI-9 Whole Body Dose from Activity Confined in Containment . . . . . . . . . . . . . . . . . . . . . . . VI-16 VI-10 Thyroid Dose from Plume . . . . . . . . . . . . . . . . . VI-17 9
iii
k
- 1. INTRODUCTION A. PROGRAM OBJECTIVE AND SCOPE The objectives of the Saxton power Escalation program are to gain operating experience with a pressurized water reactor plant at ,.
the operating conditions of temperature and pressure being planned for large scale reactor plants presently in the construction and design stages and to verify the ability of the system to operate under such conditions.
The scope of the program covers the re-evaluation of the Saxton plant for operation at power levels up to 35 MWt, actual operation at these higher power levels, and evaluation of plant performance.
Some of the specific areas to be evaluated include reactor trans-1ent response and behavior of in-core materials of construction.
B. PROGRAM DESCRIPTION The general method used for the power increase can be divided into three phases: (1) plant preparation and checkout; (2) startup to 23.5 MWt utilizing the variable frequency set to provide power for the main coolant pump; and (3) power escalation from 23.5 MWt to 35 MWt.
The power will be increased according to the general procedure out-lined below with detailed procedures to be prepared prior to the power increase:
- 1. Startup to 23.5 MWt will be accomplished according to the pre-sent operating instructions, staying within limits set forth in the presently existing license and technical specifications.
- 2. The reactor will be operated at 23.5 MWt equilibrium condi-tions prior to the power escalation.
- 3. Measurements will be made prior to the power increase to verify that pertinent reactor parameters are within normal operating limits.
4 physics parameters will be measured periodically throughout core life up to the time of the power escalation, providing a well-founded base for comparison of design values with mea-sured values to verify predictions of reactor behavior.
- 5. All important plant perameters will be closely monitored and compared to predicted values during the power rise. Departure 1-1
O' from the predicted values by a value larger than a predeter-mined tolerance will require a reduction in power until evalu-ation of the variation has been made by the responsible Westinghouse and SNEC personnel. (
- 6. Power will be increased in two or more separate steps allowing enough time between the steps to evaluate reactor physics behavior and the reactor and secondary plant behavior.
- 7. When all parameters have been shown to be within acceptable limits and after the flux distribution has been evaluated, power will be increased to 35 HWt. Again, all important plant parameters will be closely monitored and compared with predic-ted values during the power rise. Departures from the predic-ted values by a value larger than a predetermined tolerance will require a reduction in power until evaluation of the vari-ation has been made by the responsible Westinghouse and SNEC personnel.
- 8. The reactor will continue to operate at 35 MWt (or the maximum specific power of 19.1 kw/ft).
e 1-2
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ . _ _ _ _ . _ . _ -J
- 11. F0WER ESCALATION PROGF#1 In order to implement the increase in reactor power and to maintain a l DSB ratio greater than 1.3 based on the W-3 correlation, which is the I presently accepted Westinghouse design criteria for departure from nucle-ate boiling, the reactor inlet temperature will be reduced as power is j increased. This reduction in inlet temperature, coupled with a main cool-ant pressure increase, will increase the subcooling sufficiently to keep the calculated reactor transients within acceptable W-3 DNB ratios. ;
Part of the scope of the proposed test is to simulate future water reac-tor operating conditions. In order to do.this, it is necessary to increase the main coolant operating pressure above the present 2000 psi cperating pressure, but to less than the 2500 pai design pressure. The pressure which has been selected for the increased power operation is 2200 psi, a pressure which can be achieved with the relief valves in the plant, and one which is representative of the generation of pressurized water rr. actors now being designed and built. The resulting mean clad tenperature at the hot spot in the core is expected to be 713'r with the reactor operating within the proposed license limitations for 19.1 kw/f t and at 2200 psi.
The calculation of specific power has several uncertainties associated with it. Each of these uncertainties is added to the calculated number before comparing tne specific power to the licensed value. Experience has shown that this number is consistently high but is, nevertheless, used for purposes of license compliance. If these uncertainties are all maximized and combined by direct addition, the maximum calculated mean clad temperature at the hot spot would be 721*F.
During reactor operation at the proposed license level of 19.1 kw/ft and 2200 psi, approximately 107 fuel rods can be expected to operate with a mean clad temperature at some point on the rod greater than 700'F. All special test subassemblies shall be evaluated individually. In the worst case (if all of the uncertainties in the power calculation were consid-ered to be maximum at the same time) the same 107 fuel rods could be expected to operate with a mean clad temperature greater than 707'F, but with the maximum het spot, mean clad temperature still not exceeding 721*F.
For a maximum overpower condition of 114%, the overpower trip set point can be 107%. The 7% difference is caused by consideration of all of the pertinent instrumentation errors and allowances. If calibration curves of nuclear instrumentation errors versus rod positions are drawn and power range channels are set intentionally high by an amount equal to the maximum negative error that could occur during a rod withdrawal acci-dent, then at the same maximum overpower condition of 114%, the over-power trip set point could be increased to 109%.
i 11-1 s
g, -
- 1 r
Holsture carryover through the moisture separators in the steam generator to the turbine inlet will be evaluated during normal 23.5 MWt operation and continuously during the initial rise to any new power level. P-imary evaluation vill be made by monitoring noise and vibration of the turbine.
The primary coolant will be monitored for fission product activity dur-ing rise to any new power level.
I l
t P
I II-2 I
111. PLANT MODIFICATIONS A. PRESSURIZER SAFETY VALVE ANTI-S1HKER DEVICES In order to more closely simulate the operating conditions of the ,
newest generation of pressurized water reactors, it is necessary to increase the main coolant operating pressure for Saxton during the high power operation from 2000 psia to 2200 psia. The higher oper-ating pressure will allow a higher maximum clad surface temperature and. consequently, a higher maximum mean clad temperature. The Sax-ton primary system was designed for 2500 psia and 650'F so that the increased operating pressure will be well below the system design pressure.
The code safety valves on the pressurizers will be modified by the addition of an anti-simmer device which will eliminate the simmer-ing which presently can occur on these valves at about 300 to 400 psia below the valve set pressure. The addition of the anti-simmer device to the safety valve will not affect the set pressure or oper '
ation of the valve.
Figure 111-1 shows a schematic of the anti-simmer device installa-tion. One air loading motor is used in each device to_ add approxi-mately 200 psia to the-spring force of the safety valve. The air supply to each device contains two separate three-way solenoid valves for redundancy. A pressure signal to the solenoids approach-ing the safety valve set point will cause the solenoid to release the air pressure from the anti-simmer air diaphragm so that only the spring force is holding the safety valve closed. Pressure signals to the solenoids are taken from two separate taps from the pressur-iter.
In the event of loss of electrical power or pressure signals, the solenoid will failsafe and release the pressure from the anti-simmer air diaphragm. A relief valve is-also placed in the air supply to each diaphragm to prevent excess pressure buildup.
111-1
____________j
W B. REACTOR TRIP CIRCUlT MODIFICATIONS {
- 1. Loss of Load At the increased power operation, it is desirous to install two independent reactor trip signals which would be activated on complete loss of load. The pressurizer safety valves remain available for retteving system pressure, but maintenance consid- ;
erations preclude their operation if it can be avoided. For i this reason, the following trip circuit additions are proposed.
- a. Install two pressure switches (63/AST1 and 63/AST2) in the turbine auto stop oil syatem to sense decrease in auto stop oil pressure whenever the turbine is tripped. Contacts from these switches shall be added to the reactor trip brea-ker circuits directly and via auxiliary relay contacts (63AX/AST1 and 63X/AST2) as phovn on Figure 111-2. Cir-cuitry shall be arranged such that actuation of one pressure switch will trip one reactor trip brecker diiectly via the-breaker's shunt trip coil and the other reactor trip breaker via an auxilfary relay acting un the second breaker's shunt trip coil. The second pressure switch circuitry will be the same except it will trip the second breaker directly and the first breaker via the second pressure switch's auxiliary-relay circuit. Both auxiliary relay circuits shall act upon both reactor trip breaker undervoltage trip devices.
- b. Install a multi-contact position switch assembly, adjustable over full travel (3 3/4 inch PRV) on the four-inch pressure regulating valve in the main steam line to sense closing of the valve beyond approximately 50%. Contacts from this switch shall be added to the reactor trip breaker circuits directly as shown in Figure 111-2.
- 2. Reactor Coolant Pump Power Supply At the increased power operation, the reactor coolant pump will be supplied with power from the variable frequency motor genera-tor set (See Section III-C). In order to assure that a loss of coolant flow could not occur from a lese than normal coolant flow rate due to drift in the frequency control, the following reactor trip additions are proposed.
Install two underfrequency relays ( $/UF1 and 81/UF2) in the variable frequency supply set to trip the reactor if the fre-quency should drop to 60 cycles per second. (Normal Operating frequency is 63 cycles per second.) Contacts from these relays 111-2 i
shall be ad6ed to the reactor trip breaker circuits via auxili-ary relay contacts (81X/UF1) and (81X/VF2) as shown in Figure III-3. Circuitry shall be arranged such that actuation of one underfrequency relay will trip both reactor trip breakers via the auxiliary relays acting on the breakers' shunt trip coil.
The second underfrequency relay will operate in the same manner l as the first. Both auxiliary relay circuits shall act upon both reactor trip breaker undervoltage trip devices, i
l This arrangement will prevent reactor trip for normal small flue-l tuations in frequency while assur4ng that a loss of flow could i
not occur from a significantly reduced flow coadition.
- 3. Permissive Switch in order to over-ride the above reactor trips during low power operation or during startup, install a manually actuated per-missive switch (69). This switch is under strict administrative control and would be used to over-ride the above trips only when the reactor is to be operated at a power of less than 23.5 MWt.
l I
1 I
i l
l l
111-3 l
. . - _ _. _ . . _ _ _ _. .,___. _ _ . . _ _ , - . - , . . . . . . _ .__ . _ , . _ . . ~ . - . . . . _ _ . . , . ,
C. VARIABLE FREQUENCY MOTOR GENERATOR SET MODIFICATIONS 4
During the increa" ' power operation of Core 11, the reactor cool-ant pump will be supplied with power from the variable frequency ;
motor-generator set. A description of the motor-generator set is given in Section 218 of the Saxton Final Safeguards Report. The motor-generator set will be used to assure an increased flow coast-down of the reactor coolant pump in the event of a loss of AC power '
to the plant. The modifications to be made to the plant electrical system to provide the increased coastdown are described below.
Trip circuit additions to assure normal motor-generator set fre-quency are described in Section III-B-2. A one-line diagram of the modifications is shown in Figure 111-3a. .
Measurements of the reactor coolant pump flow coastdown with the coastdown energy of the generator have been made at Saxton and are the basis for the curve used in the loss of flow accident analyses presented in Section VI-C. In order to assure that the generator coastdovu inertia will be available in the event of loss of power to the motor-generator set, the caciter field supply of the genera-tor will be automatically transferred from the A-C driven exciter to the 125V Station Battery Bus. The transfer will be accomplished using a live transfer scheme in which the generator exciter field supply will be connected to the station battery prior to the open-ing of the tie to the decaying normal exciter field supply.
The signal to accomplish the transfer will be initiated by a loss of voltage to exciter field control (See Figure III-3b). Redundant under voltage relays (Item 27-1 and Item 27-2)-are provided to ini-tiate the transfer signal to the transfer equipment. Redundant transfer equipment (Items T-1 and T-2) are also provided to assure the completion of the transfer. The connection of the generator exciter to the battery bus is paralleled through the transfer equip-ment and the connection to the normal AC supply is in series through the equipment so that either set will perform the necessary transfer functions. The power required to operate the transfer equipment will also be taken from the 125V battery.
I l
l l
l l
l t s l
t III-4 l
. _ _ ,_ _ _. . _ . ~ . . _ . _ _ _ . _ _ . . _ . . _ , _ _ _ _ _ _ , . _ _ __ __
E t
N *-
l%
w M
M e
g
(.
- o
,a _ _O_ ,
" C M
- p m a
" g
- m E
~~~
@g ,. e -; _
- s
's s M -
_; a i s ss O
. N sg l .3 w
- 5. h
<m g
/
.J w N
85 #G W
- : : : : : ,, 8n;!! ,E a v a
i i,,- c wa g
I -
]ll>
- J d
- 5<
~
j;l U P~
< l J $q e il e m >
l l -- + l/) I,l' y y U l > Q u.
6 / .>
/ l
- b <
w A f l
- / N E .n
/ \*
/ < .
r__~_,* .i 5 e
5
(. 0-Mm g a E s
R 2 u
.Q
- t*
M .: 4 4
.t i
4 4
3 ,.
= h n avvis e
,,..w .n.tou gI tlg I ,
...e 1s i
-- .;hnd;. TI: .
3 k }5 N 12 *
- i. l t ,1 )i l, E
- "l
. g a s
..i a t,- - ma y = .%niih[,i
! I a
D a
ol . t ..gg; e y
[ I !wl.1 us.) %} "
,=ca e ; -+.w .
g a
h a !. a + 2 a -
1l B
_ .. =
~T . EO a )a
!!y E
e g 8
E E -h ri e a B .i I v
44V1%
a - --
= = = G;=a t . g i .i s
- x. ui + wh- E I 2
- =
1 e
'N,h'g g s
t 5 a 4a+R =
, t _l #
@g ,5 ft i si' e.
g< o
= g1 a .
sji-i sg =
a h g: ,
1 gu = g Ie-. -i 5 ! ! w I m
=
y g a
s 4p .g. =g . g i
I ll
- g. !. a 7X
-e_ .5 4"E Ei,s
-l
- e,+42&JA.me---.AA M 4 4 ,*ML.W.* ,a ..J4h5es . JLJg a_.:ia i a d h .nW-J wmh'ed m a.4Aua- u 4 # m ,4. . ak4*p,- hii A6J -e m.-4ha Cd4+=*.e4.b d qa ---.& E.ma-sw,A-am,,._m.waewemea m.i.,-s.- e m ee .
O t
. . 4 '
', ce o . . + .
o g, ,..- .
, a:C o
gg
==
H N m .. . c3 Q
.. r" ::3 * , W
=* 3 '
to C3 W f f"/*o' Q* ,"
- g j.
C3 84 g
> .r.. w f
'r,f';.' .t e,J '
- r. ..
, 6'.
CD O
w
+. ,'e .
N
= &
Z
.* .= = = s
< /.b Q
g w
- e - -. ..
e .., .-
Y ?g jj. , , # .r .. g. o o, . ';; u it -
W m w 't 4% g J", ,'S <
w N F *'
e L.L.
It) e=
d
.I.
M
- '*?.- . Z
- 4 q
'I , .
n *'
.. >~
O e . 11 si '
b o
- 3I
- u Pt P-W
' (' y a
. . O M .J H *g a:(
e
= .
W'Wtl 83 k M w
- g . == 5mE Q w 2 R,_i w E
o C i
65 < K "
)
.{ . ., .
. ^
. g -{W" m A g $ ,',',.I*..',t.'%..=.i,.,.. .. .
.,'\I ' -x W
a' *d,I' ,
a l Q E i
.n- I
- Z
. s$: w
= a ,
4.
~ v. . I g .
a
~
g s'c L ._J w wg W; .tf ~~ i r E a
F .s
,,, a.
Pm
.s,, ~)
v} .'.
me w
J F 40 FM ,
- y s],
g === CQ -
q*y.,%p p.'g y' ' .cq e6 .
>o ~
. .. . 2,. M I{'Ih$,kl,131.'d.hh?,'h'e.2O . Ih 5 I
I i
- III-7 L
.m _ _ ____ .__ _. __ _ _
[**;i 'rsike 6 656 (as egg ,,t es tit atte a sit j- y ,. _
- tg
- l l *pni ses e i toa sw c aunts ' s ise in use N ,.'j u i.
f x
3
- *{ '
74..na, .6*s e.i .i se.k te ou a, .
. . . . .. ). . . . . l_) .
'O t+MtH+-4 H a O
.w i .
{
E.
7 6
i
-l I f i
2 :0
,, L9.J
. 1 k 4 A H--
C
, f. *~ "T* * : *:
'C * *-1,* O h 3
U 5 ~, .. .. g 7 g ** .k 8
(
e ./ . 'A.. ..
3d.e g
, +4, .. y..
- s. 7J
- w
- 8 3=
ae = .
=
I .* C H t- t H
--g e 3 3 C C C q . j._.o
, s. . . -
ett. 8attp I O g-.
r> y=
=.
E u .-
33 al f.ggy-up E
5 ll-
. 3 .: p s; ; a;k V , ,
p I'.:g .
8' g- c . gi l**Iae re
.a .i -
a s
u-
.= I t _1.: - g.a p..g 6 E IE --a R . II E- ".
_ gA kl f - E *, I A= g a
o i.'
e k.t %. f!,
a r*a r
,. (
Ep JO EE k[t 1
l
- s. i 3- = r . .
.- , u y .. .
E E o Y s
e l u es %
.=
(.)
2 A : 3 A Y h h E r-r : :t">
k,1 $
9 r :
=
> r x
- .a'.
b.
B 4
III-8
I l
IV. NUCLEAR AND THERMAL AND HYDRAULIC EVALUATION l
l
. A. NUCLEAR EVALUATION l
! 1. Summary of Core 11 Parameters i
During the initial startup and low power operation of the Core II, a series of measurements were made to determine the operating chatacteristics of the core and to compare I the experimental results with the design values. The following table presents some of the parameters measured for Core II and compares them to the design values pre-sented in the Safeguards Report for Core II.
l l
IV-1 4
. . . . . , . . < _ , - . . , - . , . , - , , , . , . . , , , . , . . . . _ ,, _ .o,, . ,,,__m, ., ,__,.m. , ,.,.,._w.,,._., oym..._ ,,.m... . , . , . ..
Table IV-1 Core Nuclear Parameters l
Parameter Experimental Design Value l
l Reactivity Coefficients Power Coefficient - 0.99 x 10 -4 ok/ k/% power - 0.8 x 10
-4 ok/ k/% power at 20 MW at 20 MW 1.05 x 10 -4 ok/ k/% povar - 1.0 x 10 O Ik/% power at 12 MW at 12 MW Moderator Coefficient - 3.0 x 10
-4 ok/
k/F' - 3.3 x 10
-4 ok/ /F' at 507'F, 800 ppm at 507'F, 800 ppm
-4 Ak/
2.6 x 10 k/F* - 3.0 x 10 0 /k/F*
at 520*F, 1250 ppm at 520'F, 3250 rpm Pressure Coefficient 3.0 x 10-6 ok/ k/ psi 3.5 x 10 -6 ok/ k/pai i
0.106 ok/ k 0.103 + .013 Ak/ k Installed Excess Reactivity I
hot clean at power hot clean at power Section 2 gives more details on the measurements to b-taker. during the power escalation to evaluate any changes in the core parameters that might be caused by the increased power level.
- 2. Core II Nuclear Hot Channel Factors Hot channel factor measurements have been made for Core II to verify the core design and to assure compliance with the technical specification limits. Figure IV-1 presents thg estingted change in the nuclear hot channel factors, Fq and F3p, as a function of hours of operation at 23.5-MWt. The base point for the curve is based upon measure--
ments on March 1, 1966.
IV-2
This curve shows an F of 2.86 as the b(st cstimate and avalueof3,09usedinthesafetyevaluagionwhichin-cludes the instrumentation errors. For F the best entimate was 2.17 with the saf ety evaluatkNn using 2.34.
This curva is the basis for the selection of the safety evaluation nuclear hot channel factors for the increased power operation of Core 11. It was assumed that power operation at 35 FNt would commence af ter the core had experienced a burnup equivalent to 4800 hours0.0556 days <br />1.333 hours <br />0.00794 weeks <br />0.00183 months <br /> operation at 23.5 FNt. This burnup vos cocched in the fall of 1966, with the exact time depending on tht- load factor at which the plant operates. With the above assumption, the hot channel factors for the start of the increased power operation are Best Estimate Safety Evaluation F"4 2.49 2.69 s
F 1.89 2.04 s
The above safety evaluation nuclear hot channel factors were combined with the engineering hot channel factors described in the thermal and hydraulic section to arrive at a total hot channel f actor for core performance evalua-4t tion. Using these numbers, a core power level of 35 FNt
.f/ will produce a maximum linear heat rating of 19.1 kw/ft in the hottest fuel rod. Hot channel factor measurements "j will be made during the initial period of increased power operations and the plant will be operated so that neither the power icvel of 35 MWt nor the maximum linear heat rating of 19.1 kw/ft will be exceeded at steady state.
B. THERMAL AND HYDRAULIC EVALUATION
- 1. General The thermal and hydraulic evaluation of Core 11 for opera-tion at 35 MWt is based on the same ground rules established for operation at 23.5 MWt. The evaluation of the Core II DNE conditions is now made using the Westinghouse W-3 correla-tion rather than the W-2 correlations used in the original design and evaluation of Core II. The details of the W-3 correlation are given in WCAP-5584, "DNB Predictions for an Axially Nor Uniform Heat Flux Distribution."
IV-3 e %=
The only change in the thermal and hydraulic criteria in the minimum DNB ratio for core optration and design tran-
! sients. The minimum DNB ratio for the W-2 correlations l was 1.25 and the new minimum ratio for the W-3 correlation in 1.30. These ratios correspond to the so'irce statistical point in both correlations: that is, at this DNB ratio, l there is a 95% probability that DNB will not occur with a confidence level of 95%. The higher minimum number for i
the W-3 correlation is due to a little more scatter in i the experimental data upon which the correlation is based.
l
- 2. Engineerinn Hot Channel Factors A description of the engineering hot channel f actors is given in the Section 111 of the Core 11 Safeguards Report.
Evaluations performed on Core II after the submission of the Safeguards Report have shown that a reductign in the enthalpy rise enginee tng hot channel factor, F I*
pogsible. Theheat41uxengineeringhotchannekHfactor F has not beaa changed and remains 1.045 as defined in t0e, Core l' onfeguerds Report.
The changes to F 4H are as follows:
I a. Statistical subfactor for pellet diameter density, enrichment and eccentricity and for fuel rod diameter, pitch and bowing: The previous design number was 1.14.
Based upon measurements of the as-built Core II, this factor has been reduced to 1.08. The major improvement is realized through the use of the new, improved spring clip grid design described in the Core 11 Safeguards Report.
- b. Inlet Flow Ma1 distribution Subfactor:
This factor is based upon previous analyses of in-core
' measurements and remains the same as before, 1.07.
l l c. Flow Redistribution Subfactort l
l This factor is changed from a previous value of 1.05 l to a new value of 1.02. The change is based upon re-vised computer code calculations which determine the effect of flow redistribution out of the hot channel.
IV-4
e
- d. Flow Mixing Subfactor:
This factor has been increased from a value of 0.95 to 0.96 based on an evaluation of the effect of the '
new grid design.
In summary, the new factor ist Subfactor yalue_
Pellet diameter, density enrichment and eccentricity Fuel rod diameter, pitch and bowing ,j Inlet Flow Ma1 distribution 1.07 Flow Redistribution 1.02 Flow Mixing 0.96 TOTAL F 3g = 1.13
- 3. Thermal and Hydraulic Design Parameters Table IV-2 presents the thermal and hydraulic parameters for the operation of Core 11 at 35 MWt.
IV-5
Table IV-2 Thermal & Hydraulj L Desian Faram,eters
. Total Core Total Heat Output 35.0 MWt Tctal Heat output 0 119.5 x 10 Btu /hr Heat Generated in Fuel 97.4 %
System Fressure - Nominal 2200 psia System Fressure - Minimum - Steady State 2150 psia Syates Frassure - Minimus - Transient 2050 psia Totil Flow Rate
- 3,21 x 100 lb/hr Effective Flow j Rate for Heat Transfer 2.74 x 10I lb/hr Flow Area'for Heat Transfer Flow ** 2.53 ft Average Velo' city Along Fuel Lode ** 6.01 ft/sec Coolant Temperatures i
Wom,inal Inlet '
,460 F
.Maximu~m Inlet Inc1uding Instrument Er,rors and Deadband 485 F Avdage Rise in Vessel. -
32.F
' Average Rise in Core 37 F ,
&verage in Vessel 496. F Average in Core, 499 F Rest'. Transfer
' Act;ive' kest Tran'ef er Surface Area .. .
of Fuel, Rods 510.8 ft 2 Average West Fluu 254,000 Btv/hr it Average Therma? Output
- h.83kv/ft Masteue clad Surface Temperature '
at Nominal Fressure . '454'F ,
- Assuming constant voluet pump.
- s Assuming all montrol rods out, followers in.
IV-6
.c -
.w-- -3.=_r-c.--.m..rry-e-v -,r, - ~ 4. , mv e --..w..~-,.
f Table IV-2 (Cont'd.)
Central Core Region (U02 - PUOy Fuel)
Fq Heat Flux Hot Channel Factor 2.81 F
3g Enthalpy Rise Hot Channel Factor 2.30 Nominal Outlet Temperature of Hot Channel 578 F Haximum Outlet Temperature of Hot Channel 583 F Haximum Outlet Enthalpy of Hot Channel 590.5 Btu /lb Saturation Enthalpy at Kinimum Steady State Pressure 689 Btu /lb Maximum Heat Flux 640,000 Btu /hr ft Maximum Thermal Output 19.1 kw/ft W-3 DNB Ratio at 100% Power Nominal Conditions 1.$2 W-3 DNB Ratio at 2050 paia, Tg , max.
114% Power 1.32 l
l l
l l
l l
IV-7
. _ _ _ . - _ _ _ _ _ _ . _ . ~ . _ _ _ _ . . . . . _ . _ _ _ _ _ . _ . _ _ . _ _ _ _ . . . . . _ _ . _ _
i
- 4. Fuel Central Temperatures Maximum fuel central temperatures for the core with a power level of 35 M'4t and a maximum linear heat rating of 19.1 Kw/f t have been calculated. With the core loading of Core II, the hottest fuel rod is a zircaloy clad, pelletized Pu02 - UO2 rod. The maximum central temperature would occur for 19.1 Kw/ft.
condition and 35 MWt operatien combination at the start of the high powsr operrtion. Using the basis outlined in the Core II Safeguard Report, a maximum central temperature of 4465'? is calculated which is below the fuel melting temperature. At the maximum over power condition of 114% power (39.9 MWt) the Linear heat rating is 21.8 Kw/f t and the associated maximum central temperature 19 4825'T which is also below the fuel melting temperature.
- 5. Fuel Clad Temperatures At 19.1 Kw/ft (the maximum linear heat rating for 35 MWt) and 2200 psi, approximately 107 fuel rods can be expected to operate with a mean clad temperature at some point on the rod greater than 700'F. In the worst case, (if all of the uncertainties in the power calculations were considered to be maximum at the same tire) the same 107 fuel rods could be expected to operate with a mean clad temperature greater than 707'T but with the maxitum hot spot mean clad temperature still not ettceeding 721*F.
l
- IV l l
_ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ ,___ _ . _ _ _ _ _ _ . . _ . _ _ . . _ ~ _
- nt
- pj"p*Hf filt!!E j!i 93 -
3 E tit h yt::
- nt :st+ tin :tg u.. jt.. tg .4 -tt da+ tt!t g:: ;.." n:t r1p.. :in m ;L.}q. v. u. ..
egij. 33 ,t 2.Im:li.3.::
- d! II?"li+n.
mt thun n::un J.'.Fi!!!! tftt: un t:l;.; T.u ... n,.. n ti j;r.f.. :::te3.r o t
.ptt
- = tjr 1: q.. 3: ;;p..h.u:tt.:i.t.y a up.. it:;,;;..:.}g :u.1.r
- p.m-y a..;.u.h 4... m 7.p
- .. p..n m m..: .y o
..: ,m, . ngi.m
- ht: ;;}m...:. ..;;.. ..n.;.m...:. ;m. . 1 a, a .. g ,s t.... .n.a.. i.l..t. . . .
7 ...
4 . .
m f -
- a -
o
.m : .:lya ..r;{ ~
w mt m: Jg: ulgamt aan a en m an-lim u i.ll;!if rNi t lih: nlr. m- gm: nim
..-+ n* tmwtu . kn{pn 1.i L.r! t~ h::.
nn ene 4 tt it ;t.
- 1. ra!{jih~h: s1U"m: .a:ler:.- md du+ nn tm
- n. pf:"il.I
- n en E -
. nun : nWm. n:: gp
- a tn .Lu n ;m t
m mi + r#j{m! e r1y;n? : i.t.n, f i ,it,g;}~..: ; p.e. n..p. y :.g: ::ti_ md_ _
- .g3
- t.gj{;p.t a..n. h.: :: . d.gi{h.t r pm: ,t.;. m. .. .
_. m...t. :..m.
q.y1 n.:.
- .. w.:j:3;m.
a .i.i,,un u.;;n. _i; ;r}!.-
3 1
a u"' a mt
~
sg un i f:t!nn .:tut t.t .hgrJ t+ !!::
. . . . . . . rm tu{mt us t:::a. t;. :4>j1!g-gi na
- n. ~2 :u!- n:. .:::::lu::.
m: i "l . m:.-y.!gute!
mq #j +rb t yp t
!u nn W;ip:m: il::;- ;:n.;.p w::p:::p_: ;m . nrata...t, mtu nr!?! gh.qi gp; ptun : ".i
- niliu. d gt:ia...
m Et N.
ilH .st:i~~1ad g ;
': !i . .. t :gl:"tn r t.u t .m , ,"' ga 8 upn.r.g y
t r ...
.. t_ _ a.m;m..
.ru..; .:I. .. p. a t.:.t p.. _
... . ..Fn.n
- r. . '.t II N
- I;2 pUl?U o
m
.m ;.. ::. - - - "
t{jgt;n :-
tdlji IlaI"!!i.I:;t t.l.fl__.4du 24 }.I rn .n: :n utj 14 ::a
- m. mr mt m: mil!"L{: dj12m 11:. t-
.- nhnru::t Qil.]f.4 t a;g;+i{.1t:
t
- . . -r! t t t ;*
q!';
?h.lfi.i " 1. h .. . d
- m= un. _gh.1.. g pt~.m...{t:ngn. up: n.g
- n. . .t r m- p.: .. t.t ,. h_g:j
< t:g p..a.7:.i.g
- l+.. . r. :.;g mE.tm
.t.=r,u.
traf;t
.t... ::
- g
~
- m.n{
. . I_.L;p t t+u. n. ..;.r.9a.a..t. n..P}I
=. . . I 'i . E. .
9
- "..att'+j
- u. a..H.T. ..
.ut n :p "m.,m.'
.it a n. . :, u. . n. . . 4 ap i;t g .. g. n
.::n m; i i;4 ui;' p *.mt iIE42 2* -g hp"m:
n:ggp: : u: g- -gg;g . . ._ lt 1-u &n r': Tr;
- ;; uw ar ;r:
":r! 3 Spa -2;.n n:: n:t j;.;q;g.fa 5Hm .."q;;; bgj
- + .HF: +;--n-
[.;p dyyd .:p1 :::
g g,l u.
nt. . mA. t Fi~!nd.
nfi jtti. gi mt.. . . 3 I.pt"' _1g:,3,. j!m 1 m:t-:: : : 2., .m tn .
t;. . . .y. .n.{i t., ttilh1,.3 4E.,h. 'il l.il
+.. f . .fg1 J..EF.". ~.. g-4e a!ti 5
. .i ,.,fi,,!i,,'
- n. r rw un . , . ta..
3.trji*:I. :t :t..b.:. !rpp..::
3,q
- n:
" C
- g
- t:a
.:7 spg r" m" n" .b:i.d. m zg;tu.t
.i en, pu : run
- tp* r.: c
- da e.a s :a im au ,9cn .i 3.fp m 1: fttt +n :n:;L. .a f .rt : - t:!t r& +ljr+d +p 1
..l3.:.Ig:u-
- tn
- . . p+t (t
.:-: : rm 1"'~. t. .
i.n. -
gjity ph:2);;jj. t;gya ! ;f :t -
..Ibt k: -nEtr.t.t t"i.l.t.~9:th.
~ ..
ed.t : -
lif[ p.t,r. }# Ei.}I',,,F, t_i,n.hn t -.tH ".
.p. ::a. . .
th.,. :.i. .!. !. .y e :*.u.~.f;;t.t. l.i.3. [id. ..a;,- f[t. t: u..n .l .-I 4.. i .
- rg4! t2.rljy -
. . m. . . t. . . . -
Uf tvi i t t e- - - d}.j p+::r: h#
.:r a ~cc t d: ~it'j!m** cir..li~- . ;{. t_ l::t;;l *h:t: u ifr: ni ilti u. gr ,.ch 1,gt 13- i tr 3 ...
- t. 4-}t.} o nijdk":: n ..n 2Hf-i:: n : it}J +Clit -~m t *.f- ttg t 1ttetq;; . :r ,, s ;*it i g((
- . .t t -+ - o tm.. .nme e n4 ,: . I I.fr iD,v gf}.._}y{3.fj}} .3m 2
t.tt w!n:e
- Ju. t,.pt i.t. y. . *.'.t!.. jit,: ..!!{1 .}t.*. 3.m. !,y .l. .
r[3 I l.i.'%.. r.'i. '
. t
\ -*
'1k.II.I.
1 14.
ll.. fiA..4.I I. l. ..3_M 2 ..
g h
If *ifNihN h? NNkb f - ht f!+ b! hftNh$h I* i bhf hke NN I b.
it-r , . m
.. . # np y3'h 1... . m ;e.=.;".:.it :.
!.i.l.t t fl.yo m ,.c.
ti+t .. 2mfttt tt ..4. l h l^p. . m. . ~ .. J
, . tti.e.r:n..'t.a. t :n+'e:
r,;g .,et!!p.s.i ig.a.;n- .
t ti c.i.r.,: m..!:4::t.. . *tt. t.I-. . f.t.i._ .
4J aC
_o t..u. *4.}
.. .c-
- u. ct 4 i..r .1, . e. :
o u o ~ .
.t { j
..- o.,
i:. N.y 5... .+..,
4 1;
1 }j -{ .g.
4.t.. .JJ{h . t.tjjji,j, t.i.j*d.
- :. d j.. .;)ta .y.l.i.17 41 1 ..n,t";j} gr*1 .'.A '.!.r. c. .i{'** J{. 4
- T1.
,{h3k, fit -- !.dLl 1 + e
.,,, p 4 _ .
J'.t'. ,, ...~ . . _.. y 1 . pu .
[ lf.f.
- ._. 4 g'
}}}, Ng.* m J l
$. I . dif b3t ' tt g
gI.
g
. e
,g31 h3'lU $9tk;iii ilhiH !N..kY35M- l
. t di b $[ IEI3 $.NMi N $tI n d.i4NI15l$ e n'
iS!$N IN IU .NIN$5N .y! w
- 5 LEEN r k;,,gtd:NNgWnfi =pw 1p4bt.. nt t i
9siLh 3m.1..i M3 H~otm rm.. , a wnm i. r11 i .g d
g, M
IihiiNI 9$lsOSE' E!!$* $i$$ifi lMM N @ It4 $liS $${f 0 EiiMi WM i !H! Its .M hfs iWih Mi f WW!"in!!!i E.3 lie gl<
C l 4 4di ff#iht = a u$
a n
1 E!%I n - .
!!ilMi !ni s {! HO M..Eo. i$ . -
i ' . liifNi
~
t
.4 ~3} MI-dt'Ti
~ . ~. . m f.fk g
CW o
il]'
i.b l .i lihai .tP Mi..R.ft --
M ..
- .. .g I
s N! f*' Tk *N !
M e
[ -;h' N Ib
!N y gf h .Y.3.El g 1
i:n}.j:n t hg y t rj 11tt t
- irl W ta. 1:mt:g ;.t; A 14 i fig t g!' f3 At :tm:
a:((.. t'tt . {f..;
a e:*t -
- i}itt 41+t..1"21 .
tvu ., lul1;tt: :. LQ- ; t c : .
et +- r .e tal;g.h"t .:m f et,t! ' 3'::tt : t! r"ijuj r - .
a I U::: -.:lg. a ftrt }il.tfn:1. W
- .gnu~ fy.
2 +.1.
vt:
- . r
- : .d-;nm:.alc.
t.
in . . .:
t I"*tt '"pt t::.3. . d.g +*+ll- ;1r. :.t;tL. . lutti..j .*1(t:f u t-m-u.
d c ".3 .. * :a -
d..t..:. J. ftjt. ,.}-..rM. f.sg.m.m.
i -
tp p.j;_.I.p,tr.
.11 ti t :4t.h....ff". ....[1 u.hko{ }{!.t. _h.
- n. 7.f. d. 4.{yj ..- jr"14" 1 E .
it j 1.;.*. .t.NiijT.i..tt.yh.i_ A 1.y .+1 .h.i. l. .
3.g;Q,t.
-i;r:1 14 trit t >
t 7 ra:vt!
+
3 m nu .t'+ilin4 t
grya4r. =mtr. :;r.ttt n* n:41:. .:jjjng hi{j:tigett:{g{t
.;:: un ijf'(m4:: n h.n:4
't u 'E . -d: im nr s.i :;:m- t-m j+it. a =
.:= am tn :. . .- .
u
.m ; *m . .
- .. ih,th...
- m m .2... k. .'4.' [n.*.!*ki.ih [}f.!.
{1 tko d.i..{g-} i.:. ._1. . .:tJ...
~
- 4. . . i.*.1 #-..:}{ t!. i.t.' , .. h..i!*.::3[ ..... h,t.'. jjiu.j h. . - - - -k; ["r.t.((3_,, . . p . .
h N'M d n fI5' $l kt5hNb iIfh
- I- :
yv ht7 m NbbkNNh,k
.w NbN:hk, u . .. . .u yt1m.
^
h:1 . -
3._2 ~c. .r..~ir..4.
tu nt..p. r.4.
.r. .u. y;y1..gr..y.,.m. .. .
m,gic. .a.si: In ..d..m.
- - t.gi.,g,ret atfut :t .
.t .
r + :2.p--Q}jt
- tr 4: >
tyt 2t gir: g:.qttJu: :n c tt:.'
ut"un im.p.m t ": t, witm :[Yjj81i tit.:- it 4.(1+1 ..
{i gb4 ;".p g.: : }i-
+
m dq ; 3s:
4tp mt I
' 9turem.m.r: : .-. n . . .
.g -:. .tu 3 i c ,ip :'3
~aj 'h
-+IN((J{h. f . hJ s. Ni.Etil #h'e+
- m. m ma p .. tt
- af 4 .I.IIIfiI i
~ ; :d Ir I. - 4 it. 'i .jfN ig;;- n f{jt}e$ ft; ..~ihgi hili
.N-m .. " . i
.E: .
m m . .
4 I.j. ;d O
- v. . ~ ~
tMFOjgj q=;; E' [it((:((h. !}yi.f:.]~ 3.1~ xn i
l :
d iiii fi#*IId ,:@. 33 .la .- Y.ii.@.uM n
(
w* nauir'It o l+N* hF3h m I' Jt} " mt t
..u m! " tM
'y{"*t4nt 1 -
a f"ifr
- p. ~.: a:*ti n ,13 i ,w t
i- . Y23P.14tht;ti ' fH. ..I.y: *t t.a., t*Wi7
- v '
im $f t m..r
-El *mrliir:j M' .i iF..J.t S".h* *H.' w. .
- j
- .f f.pi i{
@um.ft
.. r h. ;,:$b m.
-.= a-- t* .l{pt}t h.i E_T*j.tza.13 m .-
.[ ft.In:tm.:.ditti it{q, i t. ;-.2 l
.. jt --{: +t:t t;;* 4 O
- f * . =t t; +1 t;u
- I t tg :
-- C
- j12-3 4 g=:q. -tt ., o 1 tem:.ym.r r nd 8
- rtm
- dx w:: re m. tr a
e .tu. :,1 tru.m
- t z.m 4g
. tmt ctmt1 h: a .
4 njib In3..;f gE..y -m te . n4t M, .'"~' , . .
996'i N H3HVH' a g 4. 2 1 uJ:' H
- 'm r i. t .: ky:
r L 0fi k *4 bg i=s:] r: iipm 3:...
Egypt :- - n 2::. w - '"l
. 'it ::2' .
-.g.
+!!.!.I.3s
- + *. t* .. - - -
+
7:l s t..gp,
- fi. .at y. tf.r.r;.;,) .q.
+
j ., m: Q. -.
+o
,l !..t_rr..: ... . :. .
.g ...s 4, n g.. o Gi'. ,' T("
[.ik::
g--~Mh.
{. ."'
$*!I I'3$"$1I!**dhh 'TI" Ihh im ' t'"y r W t " " -*n 4:
- a, ..
+ }iiI* '
- i i:n#jM.mgi:ut+
,;; ~ Cy; "tt " v: y? ;.. '"+ : .,
- llitttt::4;t!P ' -
,. o c?t. .
e O e O e O e o m 80 5 5 m me e e e
- O. O. e .
e4 *"4
' D 3 10 H71 N NI 20NVH3 N
IV"9 i
k S
i =
1 1
+ .
V. FUEL ELEMENT MATERIALS EVALUATION A. INTF.0 DUCTION The problems associated with raising the power levels of Saxton fuel 1 elements from 16 kw/ft to 19.1 kw/ft peak and operation with mean clad temperatures of 721'F peak have been analyzed from a materials standpoint. Three posaible problem areas are discussed in relation to fuel element integrity: 1) fuel performance *,.2) Zircaloy creep,
- 3) corrosion and hydriding, and 4) fuel swelling, in addition, relevant Westinghouse experience with high power Zirca-loy fuel rods is presented in the Appendix. Results of irradiation experiments in the WTR, Saxton, NASA Plumbrook, and GETR reactor are given which provide useful information on such parameters as speci-fic power density, fuel-clad gap, diametral expansion, fuel melting, burnup, corrosion, coolant chemistry, etc.
Based on Westinghouse experience with fuel elements operating under comparable or more severe conditions than for the Saxton Power Esca-lation grogram, it may be concluded thct the Saxton fuel elements g
can safely operate under the new design conditions from a materials
- standpoint.
b B. FUEL PERFORMANCE f
Operation vf the Saxton fuel elements will not result in center melting either at nominal peak (19.1 kw/ft) or overpower conditions.
Maximum fuel temperatures for the Pu0 2
-UO 2
Pellet rods are tabula-ted in Section IV-B.
"estinghouse experience with extreme power ratings in fuel elements,
- f. Appendix, Section 1X) shows that even gross center melting een sh'.ely be accommodated. Thus, design limitations based upon fuel ,
melting are conservative.
I increased thermal expansion of the fuel pellets with the increased power levels will not jeopardize fuel cladding integrity. Axial expansion is minimized by dishing of the pellet endfaces. Initial (cold) diametral pellet-to-clad gaps of 0.005 inches and 0.0065 inches for the mixed oxide and uranium dioxide fuel roda, respec-tively, will be sufficient to prevent appreciable clad plastic deformacion (in excess of 1/2% strain) or failure from radial expansion.
Westinghouse experience with vibratory-compacted fuel (cf. Appendix, Section V1) indicates there are no problems from fuel expansion.
d V-1 l
l l
t Fuel swelling due to internal accunulation of fissien products with burnup will also be safely accommodated by the initial fuel-clad gaps. Bases upon $nitial fuel densities of 94%, and anticipated
"'"# ** * ""III~
peakburnupsofSggg00 MWD /MTUfuelswellinE0"111 ciently low rate (0.0016 siV_ per 10 f/cc} to limit swell-ing and subsequent clad fa. lure. YAlthough che control rod followers will experience a peak burnup of ~50,000 MWD /MTU, the anticipated fuel swelling can be accommodated by the initial fuel void volume and fuel-clad gap (a relatively large fuel-clad gap will exist near the end of life since the power rating of these fuel rods will be relatively low).
Higher peak burnups and average fuel temperatures will increase the amount of fission gas released by the fuel. However, the con-servatively designed fission gas plenuma in the Saxton rods will limit the accumulated internal gas pressure to less than the exter-nal coolant pressure.
)
Based upon the installed excess reactivity and measured boron worth, '
the expected life of Core II will slightly exceed the design life of 8400 hours0.0972 days <br />2.333 hours <br />0.0139 weeks <br />0.0032 months <br /> of operation at 23.5 MWt (See Figure Il-1 of the Core II Safeguards Report). Based upon this life, the maximum fuel rod internal gas pressure is less than the coolant system operating pressure of 2200 psia proposed for the 35 NNt operation of-Core II.
C. ZIRCALOY CREEP R ATE The maximum mean clad temperature of 721*F in the 19.1 kw/f t rods is well below-the critical region in which the permanent diametral creep rate of-Zircaloy increases exponentially. It was calculated that tube-reduced Zircaloy would not f ail for 7.4 years at the max-imum (16,000 psia tensile or compressive) expected design stress level 1. ? c 4surized water reactors. This calculation was based on the ultr1-ecuservative assumption that the maximum clad stress Aists throughout life.
It is concluded that no problem exists with creep in the fuel rods during the high power tests since stresses will be compressive and strains will be limited by the fuel pellet.
D. ZIRCALOY CORROSION AND HYDRIDING Excessive corrosion and accompanied pickup of hydrogen leading:Lo loss of cladding ductility is not a-problem with Zircaloy fuel rods and is not expected to be at' higher power levele and temperaturcs.
Zircaloy-2 and Zircaloy-4 have been irradiated in Saxton at surface temperatures up to approximately 642*F with nucleate boiling.
V-2
~ - .
- . . - _ _ . _~ --- . -_ - =- - - - - - - - . -
Experience to date with Zircelay-clad fuel rods operating in the bor-ated environment shows:
a) Satisf actory in-pile performance of Zircaloy-clad rods. A lus-trous black oxide was present in all cases, and there were no adverse ef f ects related to exposure in the Saxton chemical-shim environment.
b) Good agreement of weight gain values with published out-of-pile results. There is no adverse effect on the corrosion rate due to irradiat$on or to the borated coolant.(2) Hydrogen absorp-tion by the cladding also agrees well with published out-of-pile results and anticipated pickup levels.(3)
ThepossibleextentofZircalg-4corrosionandhydridingcanbecom-a combination of models representing puted using the RB-COM code (
forced convection (RUST) and nucleate boiling (BROIL) heat transfer conditions. Predicted curves for hydriding and corrosion for Saxton as a function of time agree.well with actual in-pile data from this reactor. Based upon predicted values for similar or even more severe conditions than scheduled in the Saxton Power Escalation Program, increases in corrosion weight gains and hydriding of the Zircaloy clad will easily be within tolerable limits.
E.
SUMMARY
AND CCNCLUSIONS The performance and integrity of the Saxton core fuel elements under increased power conditions have been analyzed from a materials stand-point. The major possible problem areas of fuel performance, Zirca-loy clad creep, and Zircaloy corrosion and hydriding have been examined. No problems are expected during operation of the Saxton fuel elements at the proposed power levels. The relatively modest magnitude of the proposed increase in specific power together with the initially conservative fuel element designs will permit expected changes in fuel behavior. Increased clad temperature and heat flux will not introduce problems of creep or corrosion and hydriding in the cladding.-
These conclusions are substantially supported by Westinghouse irra-diation experience with high power density fuel elements in pres-surized water reactors.
V-3
4 REFERENCES
- 1. M. L. Bleiberg, et al., " Effects of High Burnup on Oxide Fuels,"
WAPD-TM-1455, (March 1962).
Fuel Rods
- 2. D. R. McClintock, "The Performance of Zircaloy-Clad UOin a Bo
- 3. D. R. McClintock, "Ef f ect of Surf ace Treatment on Zircaloy-Clad 8, No. 1, Fuel Rods in a Chemical Shim Environment," AUS Trans.
pg. 17., June 1965.
- 4. K. C. Thomas, D. B. Scott, and R. J. Allio, "A Computer Method for Predicting Corrosion and Hydriding of Zircaloy under Heat Transfer Conditions." 14th Annual AEC Corrosion Symposium, Augusta, Georgia, May 1965.
V-4 I
i
~ - -__ ,
VI. ACCIDENT ANALYSIS A. GENERAL The increased power level (35 MWt as compared to the 23.5 MWt licensed operating power) and the corresponding changes in operating conditions and instrument settings require that some of the inci-dents previously reported (sn the Saxton Final Safeguards Report and in the Safeguards Report for the Saxton Reactor Partial Plutonium Core II) be re-evaluated. Information in Section 502 of the Final Safeguards Report relating to the possible causes of incidents and the safeguards provided applies to the incidents analyzed for this report and will not be repeated.
B. REACTIVITY INCIDF.NTS
- 1. Pod Withdrawal Incident An uncontrolled rod withdrawal is assumed to occur from an electrical or mechanical failure in the nuclear instrumentation and control systems or by operator error. In this unlikely event, an electrical interlock ensures that only one of the two control rod groups would be withdrawn. Assuming that the most reactive control rod group is withdrawn at its maximum rate (1.5 inc! ' per minute) in its maximua worth region, the reactivity addition rate is limited to less than 7.2 x 10-5 2Lk/sec. In Section VI of the Safeguards Report for the Saxton Reactor Par-tial Plutonium Core II, rod withdrawal trantients are presented for cold and hot suberitical, and full power operation initial conditions. These analyses were based on a conservative high insertion rate of 2.5 x 10-4 tik/sec and illustrate the effec-tiveness of the overpower trip in terminating a rod withdrawal transient. Startup frcm the hot or cold suberitical condition is of less consequence than rod withdrawal from power, so rod withdrawal from power is re-analyzed. The principal change affecting the transients starting from suberitica) is the increase in the overpower trip setpoint from 115% of 23.5-MWt to 107% of 35 MWt.
The additional energy generated and, hence, the increase in fuel temperature in reaching the higher trip level is not sig-nificant. It should be noted that the analyses presented for rod withdrawal from suberitical in Section VI of the Safeguards Report for the Saxton Reactor Partial Plutonium Core II arc highly conservative in that no credit is taken for either the startup rate trip (set at 2 decades / minute) or for the reduc-tion in overpower trip setpoint (to 5 MWt) during zero power operation.
VI-1
I.
The transient for rod withdrawal from power is shown in Figure VI-1 based on the following conservative conditic.us t a) initial power leve) is 103% of the nominal (35 MWt) power to allow for calorimetric errors.
b) initial primary coolant pressure is at its minimum value of 2150 psia to allow for instrument errors.
c) Initial coolant inlet temperature is at its maximum value of 485'F allowing for instrument error.
d) Minimum expected absciute value of negagive fuel tempera-ture (Doppler) coefficient: -1.0 x 10- tsk/*F.
e) negative moderator tem-Minimumexpectedabsolutevalueo{21k/*F.
perature coefficient: -?. 0 x 10-f) Reactor trip initiation due to overpower at 114% of the 35 MWt power (7% over the 107% reactor trip setpoint to allou -
for instrumentation errors).
With the nuclear flux and hot spot heat flux peak at 120% and 114% of their nominal full power values respectively, the mini-mum DNB ratio (c61culated using the W-3 correlation) is 1.31 and occurs 3.7 seconds after initiation of the incident. This indicates that the power level reactor trip protection would prevent core damage in the improbable event of an uncontrolled rod withdrawal.
- 2. Steam Line Break Incident Rupture of a secondary plant steam line is reflected into the primary system as a step load increase and results in a decreas-ing coolant inlet temperature. The negative moderator temper-ature coefficient causes reactivity and power to increase. For large breaks, the control capability of the plant is exceeded and the reactor protection system will automatically initiate a reactor trip by either the overpower or low pressure condition.
Following trip, heat extraction exceeds heat generation and the coolant temperature decreases further. Due to the negative mod-erator temperature coefficient, the shutdown margin is reduced until heat removal is terminated.
The consequeaces of a steam line break for the proposed 35 MWt operating conditions are not changed from those reported VI-2 I
a- ,
1
previously in Reference (1), (2) and (3). The lower secondary temperature, 420'F for 35 MWt operation as compared to 490' for 23.5 MWt operation) will result in a smaller steam mass flow rate through the break and will reduce the rate of cool-down in the primary cystem. Since the average coolant temper-ature is reduced to 530'r for the proposed operating condi-tions, the moderator temperature coefficient is less negative than that assumed in the reference analyces. Both of these effects will result in a less severe reactivity transient. The effect of the increased power level is to increase decay heat follovir.g reactor trip which reduces slightly the primary sys-tem cooldown. The operating conditions for 35 MWt result in less reduction in shutdown margin during blowdown of the steam generator contents.
C. MECHANICAL INCIDENTS
- 1. Loss of Flow Incident A loss of coolant flow could result from loss of electrical power to the reactor coolant pump motor or from mechanical fail-ure in the pump motor or coupling between the pump motor and pump. A mechanical failure causing sudden seizure of the pump motor is not considered credible. Following loss of coolant flow, coolant temperature will increase because of reactor trip circuit delays which allow continued power generation while flow coastdown in occurring. If the heat generation is not termina-ted rapidly enough to prevent DNB, clad failure can result.
Power generation during a loss of coolant flow incident is ter-minated by an automatic reactor trip initiated from either a low voltage signal on the reactor coolant pump-bus, a low frequency signal on the MG set control or from a low flow signal.
In the event of a loss of the station 480 V auxiliary electrical system, reactor coolant pump coastdown is hxtended (see Section IV) by automatic switching which transfers the field supply of the generator to the station battery. This extended coastdown (1) Saxton-Final Safeguards Report (2) Saf eguards Report for Phase I of Saxton Nuclear- Experimental Corpor-ation Five-Year Research and Development Program (December 1961).
(3) Saxton Nuclear Experiment Corporation - Safeguards Report for the Saxton Reactor Partial Plutonium Core II (March 1965).
VI-3
____-_-__m_______-__----:<---- ------ -- -- " - - - - - - - " ' - -
_ _ _ _ _ _ _ . _ _ _ _ _ - _ _ _ = _ _ _ _ _ _ _ _ _ -
l Ao oriown vu rigure \l-2. the loss of fluw tran61 eat . . u . . .i this coastdown was analyzed using the following conservative initial conditions:
a) Power: 35 MWt x 1.03 - 36.05 MWt b) Inlet Temperature: 480*F - 485'F c) Primary Pressure: 2200 psia - 50 psia = 2150 psia d) Initial flow rate corresponding to MG set low frequency reactor trip setpoint: 60 cps.
The minimum DNB ratio in the transient is 1.31, occurring 1.4 seconds after initiation of the incident. Hence, there is no core damage for loss of power to the pump motor.
- 2. Loss of Load Incident A loss of load could result from either turbine trip or steam line valve closure. Protuctive trips which sense " auto-stop" oil pressure and main steam line pressure regulating valve closure would trip the reactor immediately and there would be no consequence.
Loss of load without immediate reactor trip will result in heatup and pressurization of the primary system, and the reac-tor would be tripped by the hot leg temperature trip. The primary oystem pressure could increase to the safety valve setpoint which would cause steam release and limit primary pressure.
The losa of load incident is analyzed with the following assumptions:
- a. Initial power was assumed to be 103% of 35 MWt to allow for calorimetric errors.
- b. No immediate reactor trip.
- c. No pressurizer spray action.
- d. No pressurizer solenoid relief valve action.
VI-4
9
- e. Maximum expected value of the moderator reactivity coef fi-cient: 2.0 x 13-4 tsk/'F.
- f. Maximum expected abtolute "" " * "" #"" " Y coefficient: -1.65 x 10"5 dik/*F.
- g. No hot leg temperature trip.
The transient response is shown on Figures VI-3 through VI-8.
Figure VI-6 si.ows che power decrease resulting from the nega-tive temperature coefficient. Figure V1-7 shows that the first safety valve opens in about 33 seconds to limit pressure to 2520 psia, a value well below the setpoint of the second safety valve (2575 psic) . The maximum rate of steam release is 7720 lb/hr, well below the capacity of a single safety valve (20,000 lb/hr).
The above conservative analysis demonstrates that the plant is protected for this incident.
- 3. Loss of Coolant Incident Loss of coolant through a rupture in the primary coolant sys-tem resu)ts in decreasing water level and decreasing system pressure. The reactor is tripped when the pressurizer water level decreases below 20 cu. ft. or when the system pressure decreases below 2050 paia whichever occurs first. The two safety injection pumps are started when system pressure decreases below 1000 psia to deliver borated water to the main coolanc system.
Section 504-A of the Final Safeguards report analyzed the con-sequences of loss of coolant caused by rupture of the largest pipeline connecting to the reactor coolant-system. That analysis, which is based on a reactor coolant system average temperature of 530'F, shows that the water addition by safety injection prevents the core from being uncovered,thereby pre-venting core damage. The reactor coolant temperature will be reduced to 500*F for the proposed 35 MWt operating power level.
The effect of this reduction in average coolant temperature is a reduction in the saturation pressure of the system which in turn speeds delivery of borated safety injection water to the system. Hence, the analysis in the Final Safeguards report is conservative for the proposed 501'F average coolant temperature operating condition.
VI-5
4 D. HYp0THETICAL ACCIDENT The existence of a situation with consequences more sever. than those resulting from previously discussed credible incidents is postulated. This hypothesized accident is defined as an inctantan-eous blowdown of the contents of the reactor coolant system to the containment vessel resulting in meltdown of one hundred percent of the reactor core and subsequent release of fission products. Even though this situation is not considered credible, it is presented as representing an upper limit on the accidental release of radio-active material from which consequences to the public can be con-servatively assessed.
The ef f ect'of flashing of the contentc of the primary coolant sys-tem is analyzed in Section 506 of the Saxton-Final Safeguards report.
In that analysis, the peak containment pressure reached 30 psig with a containment vapor temperature of 250*F. Since the reactor coolant average temperature is to be reduced from the present 530'F to 500*F for the proposed operating conditions, the pressure transient will be less severe, resulting in a peak of 24 psig with a contcinment vapor temperature of 205'F. This is considerably less than the 30 psig containment design pressure.
The of f-site doses are calculated for one hundred percent core melt-down. The total core fission product inventory is computed based on operation at 35 MWt. The quantities of radioactive fission products released to the reactor containment during meltdown are estimated based on fission product release fractions reported in TID-14844.
These fractions are:
100% of the noble gases 50% of the halogens 1% of the solids Fifty percent of the iodine isotopes released to the containment are aesumed to plate out on walls and structures within the containment.
The remaining activity is assumed to be mixed uniformly throughout the containment volume and the direct gamma dose is calculated at various distances f rom the reactor containment.
In addition to the dose from activity confined in the containment, the dose to a receptor in the plume of fission products leaking from the containment is evaluated. The most significant aspect of this leakage is the thyroid dose caused by inhalation of the iodine iso-topes in the plume.
A constant containment leak rate of 0.4% per twenty-four hours (the rechnical specification limit) is used with Sutton's meteorological model f or determining activity concentrations downwind from the VI-6 1
I --___- -_----- -_ -_---
l containment. The parameters used with Sutton's equation were deduced from measurements at the site and are representative of stable wind conditions. This model and the parameters are discussed in Section 103 of the Saxton-Final Safeguards Report. Two and twenty-four hour thyroid doses are computed based on inhalation of air containing the maximum (plume centerline) activity concentration.
Figures VI-9 and VI-10 show the two and twenty-four hour direct gamma dose and the two and twenty-four thyroid dose as a function of dis-tance from the reactor containment.
The following table summarizes the computed doses, l
l l Time in Distance Which from l Dose Dose Accumulated Containment i
25 Rem-whole body 2 hrs. 270 M 25 Rem-whole body 24 hrs. 375 M 300 Rem-thyroid 2 hrs. 240 M 300 Rem-thyroid 24 hrs. 720 M l The doses from this hypothetical accident comply with the limits set 1
forth in 10 CFR 100 and du not represent undue hazard to the general public. The doses in any credible accident would be substantially less than those reported abovc.
I VI-7
-5 FUE! TEMPERATURE COEFFICIENT = 1.0x10 ok/'F
~0 MODERATOR TEMPERATURE COEFFICIENT = -2.0x10 ok/*F TRIP LEVEL 114%
Reacter Trip 120 - Initiated i
1 100 -
NUCLEAR FLUX 80 -
HOT SPOT HEAT FLUX m
i 60 - PRIMARY COOLANT
$ PRESSURE e
g .2170 e
- o. g) 40 -
.2160 ,5g h
m
, 2150 20 -
0 i i i 0 5 10 15 TIME,, SECONDS CONTINUOUS ROD WITHDRAWAL
~4 (REACTIVITY INSERTION RATE = 2.5 x 10 Ak/ SECONDS)
FIGURE VI-1 VI-8 s
f PRIMARY COOLANT ROW C0ASTDOWN
~
( FOLE0WI?G'LO33 0F PUMP POWER 100 --
80 -
(
~
l
= so -
l ;5*
I ag
! ti 40 -
l L
e 9
20 -
I I I I .I I I O -1 2 3 4 5 6 7 TIME - SEC FIGURE'VI-2
N.Ilh[ i$! b ! N k.b} h~ h. h[r
- - dN - h. - bb h N 00 h_-M h N I}k- h ! U d N N N N N b j h -. MIIi[NN M hII.[
hk NN Ni ' !}.b NN iin##Etsi HEMESEe- Engs EBU ##giWisi M @l liH H R!i M M li?
575 s~
im %++t:m :n..m m 4 m.:.+#. _: -...tm. f . m .2thn-nu m+.m .
v*"w/. n.a"%.t.~.. .n+u
,tf! -ti. L; ei ,
u:1 ti .
r u". : : .t._r.
1 . . . t
@ i.n.i.E{iEil. [E. m
. . .. FEE$~ijd s [3f. II.S.,j(.t@
_ 4 _
$,{6tE_E .f.dI, ((E Hli i.1H.@.
. . !H.
. {@ {h..l.ill.
!!fdski M $$$ .._.,;-A v
NN$$h hhE -
g -
N ;
I$ b NN NE uNIl h -
Ol. ^ h 3lb b ljli %.
,fn.m E. t.)uH..My,tt.:..t= ,iii gg .u.. E. _ w ,i EI,T g}}. :r. _j}N
. . ~ _
g.
- m. . _
.ta ..
- -@"i..
.{. @g.H.I;t_t,j@
. . 3, .
d_-3p ..__. {W}h
{ p. t.
1.}e! t 3 550 N$IN5 NN NNNNNN NMt b N bb M NN M M .h !!b NN N
-~~ ---
- - - - - - - = ----
S$555 N5ENN. .E
} Nil-:$ .f@SS NS $} $. :$! !!N((N
@! SN.Y NI! . -
UN Nd -
b E_EEE.EE.EE..E. - _
.j$. .,),@_ lei. @E."jyy H._L!"
- j. - .y{ @jn..i. .- Qi.! JN..!!ylf{yj
. y $ Ql!. jy [U. 'n.yy _
O m u
- m mtini
- * 5 fj+gini";*n p:T*p?
- m mit:rt :1 xm
- tthau. & + m J{4.'.g I
{ 1+1' titi {ti: .n :!tti .{Ti+!
+MT :1 tt t:m*ttti tel l r
'H 'Ht l'it'{ f't 'ti . .::1W.... -
c g. g_.
3-(gg .s .. .. . w m ...
w .+*4. e n -
g Eli__EfifE nii : +1n S E $$$I t;NilxifN!FIEi f3 feil [w10 fNA UE ...ff E fi$ IIl,t i!Il Eli
=.anesit's
~~ ** " ~ ' '
f,g^as NitaEng ^ * "^
N gp* gmiga '"'
- ^ gD'M'N#^1 1#1"!'F uh s^i 5 "
'"1 N$El E.hMNETEM$$klN Mbik! hh hk N[l- M W'{'$[N
^
525
^ -
- h. .M hN N$$$hES$$ IN.} INM f!N NN $!IN N!} NN NM hM M N} M l GTiH~EMuna EE!! nWN $$ us si M Mi inM #H n e M $ M
=
88 Bfiia/NNEMN MHal WWiiE EIElin it/r ern 4r+ +t;t:p:d;t t+ t4 gII.gpt uti e t$ nitu E!tei nuii r #j mi M W $
- i mt:tB .+.f.:t= 1:ci im n tgIttt tg tu' ' ini n;t-S it:t _
,p< n J. -
.. lit Ef} P.t tt$' .
hjh.. !N t.tr hp. t]N t. . .
$$hN t'DIh} $hNNEN $h }N!ifh kNff hM .hNh hh $h}kf E EEhEEEE#s Nsas iinE iaMD Mi E ii8 MHEMN ND H Tfi4NU a e g E5!/MM fn* !EGE2E iisE 91* 51M E E !g M Hi f#1 M IN E M itti t++1 lit M
I.Q fytil If t*;} l.it'N["H P ::=En .- :
w $$_'[.fEU-!.E.i.
_ n . - . ..u .-.m t4(.*.,ij u ! Uhii .t! m: tih m{. tti tr i gU"' .H.}}. fif^t t}tt itt w e. - . ..~ . _
3_ z .; .+t : : -
tu d tit:it i.EdMEbMI T:Ei:30itt' Nnb4 =
tu' ++ .
Wh-tt th. [+hl h+
.rf it! E-ik*m-:1 475 1 w- -
4.--
-.k~
- m. .a _..+~.* .en. ::.a*:2:.m.ri.,.#@[-i.
+i A
~
~
m 3. . B.n :: :
- - *+t _.
- , 4.gOU. J.g.}7.:t. . -
[i^rML; ntp .n i.3-E. .
,3.$_
- i. E.+~i..!n!.E .d=E
- m ++
L- - .~4-.+++rH +~ +,g, - * * ~
++n t.4 m- -
4 -
=. +* in :
[ l+ititi4: EE"U~'I tutl.. Ed . titiht+$Il'kg*i$t . g+C .. t: N'~~
m 4wt:. .: g.'xmtyggg gtm g rr+
++_
g 4-tt*rrn+ *+t Efi- IfP.gt.Fu.*imigy tti utt .t 4m:::t tt=t . . _ .
r4u u.4.U1; tH+tip:3._i.tt:Et Un n4, t
_ _ 5 ,- . ._._
.g _,_". .
+..- -
n
- ' el
.mgt
++
.x e~;t u4,-- .m _B:
IEEW: :.
a5. n*:. r*J..EE..i..ii,_i.Hi,t-. &., W.
++
Tif' --t .t m._44;,#g + . _ ..
.!"'./.
e.. + .+*- .+ ..+ &m +
$ :ttN:h*I I;f 4~++,4 4.s -+ % -+_
.. g_ - +g +'
- iti itini tiftitti iii!E uL C' O. _'I$j+3 .;_'[ it M:t{$HEJiw &J7 ;..!E -OTT 450 -. -. ..
mt3 E++ **
l
.+,*-_ _ t
_gg gg e!
1_ .,,.
.+[.-- :+r..q o
~
- I:p::: ntt:nt =- -
m & .p.4 : . ..in 2.+.. H:;gggt
-n- .
m m 1.g. nt
+
,t*:ut t m++;-: ~m
- t. : ++++ ++r t+~ut: It: *+t tr ++ m ixttit
- tm:in
.n:4
_.i r.n:= u.+44&w m:
4 .gn m:ttii H:Att- i~$ :m .++t E *E mtmt ut: im.im;:t:n$,!$' m: mt
+ -++++" = T an
- a. .- u4 U:$ iigH.EEK. nn f. .u--r.Em::::
ch , -_ _ , =,
g . i5+-
++*
= "*T--I_Eg -
i nir' -- -
3 #
- ;w: r- an +e'"*** .ggm: Imy.
ib._.+,
qn: +.~ .atut 5.+ ._ 4 u.4 .+:t ttti.g wt.
au
&- .mu+_ . _
Let:tm _ tm>m gy:rjt _
m ;= -. ---- -- E J m m. .~_, n, ,,, . , . + +#4 ~ e 1IIIE.3IF.ilt f::d= nt_:IEi-tit.J 51 4..+.+-+ttH MEM:E BEllii_. guid; 4+++++r E 425 g m ,+._ 4+ s g
- n y "
4 e4++~ e.* h. &_.n4 mitt:nu uwJattttmt..tt::xm
,p4+w. s+ a+w *ptg4 .
4 g .,-a ,,
p nnmt _ 4.t =-
.uz4nt 3g -.42. 1a 3++ +t.i--
- 1. 1
- ,; u+.g gy+"
- 1. +
- it- tt- tre-ree y
=
m jgg j@'4 ,g+mrdp-
++ -+- g g-+i~g_
- ,+
0 40 80 120 160 200 240 TIME, SECONDS I
LOSS OF LOAD INCIDENT FIGURE VI-3 VI-10
- ' $ $f N!kNN h h k k b h Nkhh h k k h b h .
NN !k Nkl h ja O tm n is imdi W sinMit fik nu is itN..j!!i itui 84j M n . t $.1_!H:1M.U
!g.} U..t.i..n.i{il.:E.i.t
_ .u. . E,E. a!.H. . N_F..El.ll.{_i. U. .N. U.I.I.A (N. .
.id.n,
. - }}$ f,I!..!?{N.l1N_
l h1 : t- it b ON
- nu
-+t2m Ea= m:nE i:ttE *-
imtxtmm:r
~+
miit+-pgt;m :jgtit eer n!l it :
}jde 7;1pa
!! inijp
~-
,a f41 u !gipi; t a j:n: t: artg7:
ggo ..
mi.nti.jgg.
m e _fiilt.iiE.RE..*mrj@M_Efj *iEffii.
. _ i.E{ lij _ ?,jEi.i.@ $.$3 M)$-j.ili
- jj[i ji.E@
inh h
55$.IIiinidhi d n$E$N$5NENMb!i NSNA !N.$ $ I I M,il - .-
l nu w+.i cm. n.;x m: **
-* 5
- ==- r- n- + t i et==r=m -55====:
- m me, ift. 5.i,n n %. yHL. [yu: Iy yJi:: ID'I y}u:te i'U
. f
. ~._m. me 55d'.
n . .n 2
.r
.4p
.t
.w 3 .
Ei!stdili !$dJii it:ttn in :u..
. . m. :t
- ::*43*ti!EE7::.: + :dl stii5N,1 litfEN dn.riEE Hn sin l EE5@EM iitrEEEME5I@$5ESE5$M !Si NE EEifE mi ES lili Illi annW na11 5 Win!N sEni!!!#E HE His si! IW HEim ME Iti! M es im nu MiMi EH gene M9n inti!E Riisi $1 fif Wh by Ri ll 4i li uns sun en smhm ses M E s i!il itali tn ni it 11 M
- 5 MEE MMIME MM Elib fiini Ehll !!E EN id il!! Ni a i ID 3 M f En eini REMM iki !!Init0 !D1E Mnn HiiWi Mi !ffi If si s M -
2 W a R$ E fi! E#f !WiFiidi iiij #W !!REl un H E#- d i : l L E 1000 nun maai#i R M91Rd Ek as EMI M .-
M M in ---- - - - -
4 9 4 -
3 [fi h.-l kN! nil NINl Iii' NI fl -NIRNi
$3 I$5 Ei $$$$I [Si-EI IE ji.
Nb g EEI kh NN fk IS$N bII 3ttN I fi IN hiN N ib $
$ M iinWiE Ni M r.ii#u i 1 Mi! !!4 iti :
11 i H l# til 1 11 f" g ,,, Mala! an.mi W#ti t#i M e19 sh #s a mi W cil 'il Wi!pnM #
- !!Ennlin $i im Ru@Ek eij n!i in lii 11 18 ' ti M1 hitM im E &
snnni !E !n &E !EiW Eill I5 1]il im Hil IlMMMifRi a iMM llW l N$$II$NNN kEh (5$NM E N I N N{ ((55[$ lE hN b :
l 800 WMEs5stiinEE ;31 Mblitsa!1 = - :
u.,. L .. 2.
1W.s@ R1 -. .- - -.
$i't idi I,dd:
4 . .
..j.
..hM tEf$.$ I.:-Ei$35t3 $_HE fl i b.D_bihi!Edii .k.
ni i .
f$_i _$$
EW fiEM !fliWWWEEWiiME E 'a ai#RE utM min siWiimnii MPTiEM it illi sii E WE #tMaf# :
- e We MM i;s) PRE sihis Leistgiisg9fi g nti #.ii 700 ..,,,- 1
. .s. -
p . .. ._- g ,, 4,_ .. . .. .---- -
2-1y M5#Mt
~..
td M. , i:hg+ t[ "n Mitti-+ bkNE-Eb: I
- ttt 3h._ I :
== tg- .ufittgfig+:jg-=1 tz e-Er+.
- ht:tli .nz:: .th gtttz:li : ;
t#4a:4ta :44#nat:. .44m mt ::dp
+
- tr e.f._t
_4 ..
nn t : : : mtm_u.. t . :
- = -++ f+ r ut ;.t:+.::m:t- nr. -t + #:4g ::4:::tt*m::t 4tt:4p r d
+
nn :Bh; FdttI:E'.
t+ 2:tt 4L : :i: :x: i.-:it1
- + :r- 4.a -_:T: + -+. ur :: .N[i.
600 :~ ++ t" +r+- .:x:m m gp prgt it7 Tttm p +...4 u-i 1.4_au u +. _-+.. 4_4 *"
t-r::rx7--t r:"+'d 1:::*t+*~ "
T,- .
+. a.. c..C
+ -4.++ ++*+ mt e "
+- -+ *+ -
- 4. 4: :g*- ;:-x: :
te t7:titpt#: m+$::C m.41 .a_ .
.$ 5.Ittt ui# + rt" -
"r -
. t+t-t"+"'- ^+
+-. +
A
. 4a.4 .-w4. . . +444 u. 4-4+. + 4. _4 45- 4 4 -4
- ,,l 'T**t+' + * *" .*t---**tt *+'t -+++-g; ; jg; - -
--4" t
"'+ ' ' ~ ~' ""
500 i
b 40 60 120 160 200 340 TIME', SECONDS IMSS OF' LOAD INCIDENT FIGURE VI-4 VI-ll
600 ..g g4 ug-4" m p9. 4ttttti13tt:4*" :
t= ttj-pp : :dt t: :tt :-t::t :. :
t1 4 1-p)
- .. y :# lt: A r:1 u.+_#q .
.uw m ++. ++_ #....m :
.-# d 1 c E$ik h[.h s!I$I $11 k: :hE Ed- $k5[ hhhiE f.M M ik h ERffl 31}ffEEE *"' E = l MEM EMRMn $! $kd},ff
@@g@j $$ "" ***; b+tMU'W9 [jjjghy$%M { :)}. *. e h .
J 3,3 1 _i i!
_ _ 4$$M 4dN_N_.L.4 gat.:j{Mlkg[.h_ iE.1L
@.@.j f. J{ - M iM *+
. .[:
.NL, a L '
L fif=" "#ffffTt WMMW M- - -Men R$i R$ LITr. IE._
( - -
- _4_
EEEMEEEEpsi
@ MM @ @ E# E!! @ ik h j[ l :M ild
== EEEHBap2EM EWWE fu b M MJ i lili 1. I. 1(_fi__.
hi 550 unsame "~ s>
- "t -
="gs#u --
g~g#ydt4ty++t M : g-g- -4t ytl t
4 j- ~ t "navnWHUM$$n mal R #1 9i$ $W u T g t A EliB! sRM MtW# WW M WW lt GU M i 1 1 1 En! M uitMMEM Eu~ 4 til b f
~
g EM EM IE13MWe b '
g if (R i 1 I
~~
525 MM !isift fE
~
$ l: Yffi ts : '
Lij~ 0 1
@ WEi S.s!i !it:i :tsiit :sk EE _
t EESE $ill W5_ Ifitif 'E _ i!
h h[h NN k 7t -t .:;: ;s NJ N ~
500 - -
@g]g;;; ;;;;g =E=a : E :_: : y :
p RMN56 $43} nf ~. : -
Eli T g MP" g.j vn g. :g M ?- -:-
% .m
. .g .. :: :nt .ut- :
g _: p::. ,
_ : t, :
na}.. ,Ot
+4 ,m --. 44+. w..;g. 4n ..+ g :_ ..,__ ---
tititttrtWEam=tn: d. nit:: .4- -
[ ji 475 T 0 40 80 120 160 200 240 TIME, SECONDS LOSS OF LOAD INCISUI FIGURE VI-5 i
VI-12
f
.fj too M d# .
n m lit milii tt i rift hitti #nu i!EM W.# Witi !.m I :::: I[}'.In. 1,?!. _Td M !h riF'..2'. d[..i*:lie .I~ in .t:g!TN!Elit ..
. . . . . . _ a. :.
-_N THitti: : I.,M
!w}
Il 15
!!h. nii.t :H imiEW #.1 ft M4 .im iciiti Wi! !!%
e, r j
... ..4,. . . , ..
- . '.6 .+..,
- ..$.1,.-{g'n.. ; .
." i"- "-" * '4~ 7 . .
- ?:
fNf f.}i,
- t. ::...}E. .- ,. .
. 1. g;.$,. '.th.
.. ...t..'... ..
_... ., 77' M 1. .
T
.1 T.t' tM: ..[
"; Ig.: 4g
. (t..
L::d
.ij .,
4... .4 ij.; I.
'at-d.
f:-: .:t be
- t
- f..&.i
.... .. t: c t ..a
.uTg.. t 90 .
.+ .. t . .
';tt .t, w
77,d,.
.,; k: _t:.. . .;-
- ..27,.,.
7 ..y n
- - .. . 12-;
.1 a g ::
..i.. . .
, . m
./2. _. .,.._,..
. .... ._T. ' . , " .
2
. . .""::-dc. ..*..'.". w : :,.
- .:.g.-.:.,g ::.t .::
pn 4 i -
r-- --. t.
4,,,, q.* ; ... .. .:
d tic: g -
.4 .
4..
'-+-
- c: g .:, .! r;; -: +:
+ ni r. ::a ..4..
242 1;.:; .. *t rk
- h:.
d I=:i 3
' wn:: N !b :
N.C!st d'N -(Ei :i !EIj!'i ttf8 EIpW;ihn:lidi! ,
[
T' '
"W+T E.sisis...iE!.!
_ .. , .-. . ..i t.t.I .t, ! 4
,w;gy'{if....}-Ty r .
r fim [..l.h.}M}H..
- _xw.. 211:~:
f.h i.y[fhib hh ;lh
. r
"~
- " j[
[] '#f g .I h'l.
Eg }.1iUj ~z. :F. : ':
fJ g..n
'. 7. .
ft. -itf: s 1:5?i,-}]
f {j} :
- . a n.t +
fi'i l
- MiIii O( I E.T.' P'
- I WN N. ::- , :
4:~.ih. 9E!A5 .f Wi
~.,
li
.j}.M
.i.ra [.:E lx:i
. .k.M' .,
... ..~ 2
- 1 -
5
- m. }r1R3
..p. .,;.
1(i:lEj.. ,..fjjR._::}-
o.. ._
$ ,), :
3- 3m
[h._h, . Y..y~.1 n ..
gl:#il
.3_:3 'f} }'t1 I.' l,f. F
.M Fr :f r
- 1 f- @j y 4$ Lt E}l jp- .4
- .
JEj :'$ j [
N- $F i s . .
Gk N.b~.f fi
}.Ih. f. ~d- $
Eh _h E'_ ;
m f
@:M E pi fit R t N::igHnMO Tit i 4t! M
" " Win E in r
&W E.4 LW F f ;?M b A iiM:1 1 lin M!
- -. }u.
.; gr :-:t r -g -
if} ie 44?"r4'Em.
l-im... .. . . . . .
in .ilI Ma =it :-l- ****l1I.
ut I :--l r i
a T 3I "}fL 5: :~Ni
= m .[!M. -EE MN! EI 'i- - if iII di[E: ti '
Old ...
aq -
_~ , , ..
1 h_3 $.
3 .
t[ : E:
IE iit d +I,l}f}
.. : :: n p m ,
J -y y-ec.:. w.x4
. : t :
..4_- :_t- ;
r..: -1:::.
- 3. j ;g -r I
.: ft::
g:,: .,
- l. .g. {. .j
- c .'.. g t; ,
-4. :
tg yt
.,xt ,
i t '
l 15MMlif.
[g ..N 40 5': !I[INllN} .}.i$ ( Ul 'Nl 3g ;;1 n n; g -yr .,utacil 6., :
- 7 ; n ;. , ,
T :+cEh,g; w .*w : 2:.: -:c:: s : ;;:Ni -
.UE i$ili!Ihh N !}Nb '[ .
{M} !!@[ T 'Eil f5
- iigEBe;fiiin st F M2A wgh yim N iNUdliMi!!Y$il N I Ni@ ;iM. E:!: -h Md
- :'11**d "J 4i..i'.i .
- 1 !AF:+ o .
- 4 :
i MmW :m44*t: n -
. . . . . = --
-.'hj-. .>4-.: ,; ;?i4 -M+** '
+ .
4 g nh = .:
j E":
4: dec i "!O3: : M iiiiF4ii .iii4i ! -
l
.+
. .. 4 Yi. -
m- "* ! i _
20 ~.
m ;.
- 3. .t_.
it
! :.c:
1 ll . , ,
ii , ,
- .:n: :a: 3 10 O' 40 40 ;120 140- 200 240 TDIE SRCouBC LOSS OF LOAD INCIDDIT FIGURE VI-6 VI-13
_ - _ _ _ - = _ _ _ - _ - - . .
o M UMi# gi[p
^
31't E..U$..i.s..i..;t;;
it ti i . .. fjf.jlii.;ti n.kll.i.E.IL..iyi .. .
t2t t M,#t;t; i,iE.$...ii,n;+.
5 v+t
~ , , . , , . ..,. L E$.2 .,
x* : i[:
1*;tt it-
-m #_. tt9n.gtgtg di , A _
E!i
_t .
N IE! EE i: M is! iElili i$leil ilisil Esi! stisE iis 9tiffiiHi aEtti sa tat fi!!!UE1 1 $; ;
=E sii!E g. i.s, !E.i..jn:i !E2
- - . -- .. r!jW...s,1 nut
,. t* as_pgi_r;nse=_4T**
.. 4 .
m ,Eiq:u:t I::=m pW:$
$ ig1 Tr[ :E qi q u2 b
- =r ==::n m +
h tj
-+
h h tt iNx t'}
d t xt ht t~ hh -* + + Nd-hkf _k}mrt!! ~+4
~
tt-t 1:k + y -
g a
tri:g;
- til g t.npt
- + 1g :gti 4:
lig; 7tti ~:Jt
- ::r :- ni;-}Hg a :tt t t.:7.;t" lr di tri: : ___ i 1.~tt uit g. f.ifpilm. at{7M' . .1 ~
- :in : - -
.:in. _
= E. ...~
t:tf- t. ii litMt.1h M[. fEIIIII: Edl.: Efit'% xp :t b4 ~+4
_E:E '. ffk.ifinr*e ~
331 *.3:1 :tti r:
15fE h!j . ! Ni si !!!IsN Idik !E}*M$$$0i?itNEM5NE if$$$3N MEHd "
EEn# !i nit!is$ LEON !O jNi i AEWiW*H% 4:# EMEiidiin : a . I o
- =1 I:t-p: +t ? tt: P. j.- . ._;L-
- :+: -: t+
p#t~: -t t-3 t: x; Mit: ;-: :+ r ::t++ 1 ar t :tn r. . ;. It}-+:,,,. t,e -jg{:ntig,.it tr 1.; r; I_#4:.a
- 4 4 :1 : : :: ::tt: -"{ ..
=
T L + 1 jt1@ T:7tU E.+ Nyi.i
~ .
3 f[j-:$_fM&.Ej@}.
g .
.k~. .i.jf.
Q-g.]TE43i.iH.
y... , , }- {:Et . + ;
g:g
._j +
24 4"
..e 4- +.. +_. .o 42
]II E. T-..
t . .
L
- n:
-t -
- +:: _ i: :. ntt tt gtt gg m+u n
- + --~ *
.A_ 2 tid; ttt it:t itii n tJ wg...I.- tt#n I_: :.Jt.t'4_
~
t:
ti t t..ct {i t-4: : ikt 71;:.t;ti Mit . . p :t ttit -
7:T 14a.t;+.1, x
+ us -
i!II h!! !!$i !!il Itih I kE$I5}!i fin 5 N$ b$15 lt' .1_ .31
~
iSIN iil 1,:; ?
. c- 'o pg gij: t;; .r f;iIfjf::t.ni +
ik 'm ? p : x;tt d,i 4ttt k*g md t.:n ::1:
a; 4, {t"[E!It:ttt-! .,g. ,44 i. _4 p;+s+
t e.4 tt::. 4: tp x4 : :jy{x} :tu .t$gg+
. 44tti tr :T n.n tingtME ditir;:- di _..- v.
- O Ni siill IEf!NiliE! $i }$53 $$lifiNE!Il$ MEM Ell 5!i l!EEN E11 w f;u'+; l_{ yit :
.MaisiliEEdlil I!!I[N us .. .
- .- . mm y _ _ . - g8 -
tiN I i k!h g.
$3 f.3 l .e :
itf3U difi!{i !h.E 1.un .fi!i Ith! M*_Likf! :{~MN :: : M.
E1 [ !EffE E!!i lf_! T: ! H E. es;
- In!;33g j 2.,tggttj qs. mt.g ug.g. jp.. gg gn.gs utg u a_rg;t.
- e :
s![
- I.!t... .
a a..g..M..a..
. - n ,. - .
, g kh,iin hNN Nil hh k i li$'~ i ihIhN !h! !h!h! MIf th! k}lI NIhh hin kN!}!! h hN 1 M VI iti! id ifJi4 15 Jil!!L $ le!F giti signis iTuli e a italit 111% si i y
- iM0! !!Dini lit is li!! ;;gi m ga tigig gglig a rij g! ggi ;gg g[#g114 s ing jij it; x, ,F1: ;f ni!4ilisi !$ !$ !N Ill! Illilid list $1 uli unili! !!!! i.il 0 11!R EM iHim its !s 1$ ii!L j! t':1 J 8 su 11!!!!!!M lin it sia inig! mijin ini s*13 gpi!! Et iij m! Imirti % !FtHIis .
Eli MMM'!!$ uli li lii !i!! Hli !!!! tit in iiiG;;;"" Met a mi liti d in !!H!1 fR!!!! t M fi BE ini iin El #E im $U 11 W ni in) in!$#1 ini i~ cii if inq iptmn@gadiblijw: +
timet. ngi iHMig i W;sE
~
i!!ni!!ninniit WE!!Ilit eiliMit int ,
8 8 8 n
8 n
m <
N N N N VISd 'XEDSSZHJ ll22DISSSZEd
?
h N$ NNh ![*k t:.h hh hkh!f k., ,b h.wb b,ia llM., t h h.hk.
EdME LEEMBitn;i fat a!fi
, w
. ~ ..
$im,t Un l,tN N.Il$ lih fid. int . Itk IN Th. IN. h.1 :
sn = c. .t.s..i_ TIE.a..s,.an..n.i:rgs.
t= -- - .
ii_ g_i,si n
,n.g,n.uia,.e,,ng-.a,.na. ,... s,ylt,.
e . p,t.
.g,- gi ..
Si.ij.in_i$.._;*In m _F.I$. .Q
.- . &. Q*.l. . . _-2k._$fe:u'u' -." w $a.. $,&&_$_$$.&. . .&.&n 59$5$1FeilNMM3$INit
- nt +MIr++
}NNMNM Nh*M$EdN$ NN$3 NNN
--2"M
+ ++- i' E m: 5!! 3 **-.T
- t +t:
1+ -
m k
"n"-$# :t ii.md4-iti.UliIlli-3dit1 N+ -.' .d ttrsE tltM++it$a1':' .
N+FN .N 1G-N' +
Ei..=..ts. :.mtHE:= =,t mr
. . ...m .gggg g3*:O"f3 3.E. ;ttf E ur 5mtnt:: lug;f.. qf,t : ng.
=.-
mi nrr++m_ _ _ +~+
7tg t
.n im2x:.'
w_
-: +
it++ ++
- p .:
{*}i.
, t1
%g g. nm.g1g:tm p4t -g.
- nt r m : t:trtitt m'i : tit mu t au . d ut *" -" -
4:E -
+1;- .
N
" EbNi[fMN NEkbNS.NdN5 NkUlf N$ $N INN E $lf s;EsNiN sufn M sn unEEMuttfii*d HEHE !# n$56
=n-
=n. 7t+<#gg: _: httr
.+ m +
p:+ g.
.n ,,i
+
>fEh::mgjnctm n tur:-mtL m mmtm:
nni:+ titr.Et :t 4++
c:%
g .
i#Q3 u.1 ut .. 3.m 1
a:
N 40 NNINN N5 Nmv -
h NNNN N$!$ NNh EN -: h$b if$b -
Tii NINi h I/l$NM"ME k!b -$NN $$f}II::NNkt Nh- $ !'kN fb1 $E U
5 imus ,igniaM EM W41 1#W un9 5*15MainMNT
,c a . -- . - .
w ,. .. .
wf im _k.h.ImNtilf[Hli ib.-(IET$1f!5 n?fM..E::
- E N.$I! aiu
- -wh..< . . .
~
~ trrt +n , - -- O_ DEN. :
- w
'EiFMU 'f ::,tn IS EN NNi+itfNNN$ b3tl+3!hi t u$~$$i;t
.itritit NM- : :T_
iin Miu! 2 Bli lift UN40 R$5 Wil & EEsu NEs3-fiiriisifl f@i.?
Me itWini N us eMEi fitfig lui 801E Me unNR f tisyF Wi sa klI a# un Rif 54 a WifFi i fMi!E#i Meis4R 3 20 K..l.ii IH en mi.g. 1ptug s 4.
i .. ,1 4h p-n _
(+*:#.
+* . =.#t.v
.w a r**: a.e. p. L.u.,mai,.
r .
0 40 80 120 160 200 2o0 TI'ME, SECONDS t
i i
LOSS OF LOAD INCIDENT l
i l
FIGURE VI-8 i
i 1
l l
l l
r l
l I VI-15
J' J W M-- '"-- 2: 22 p.:d_.Q- [.2 :M
.E~ i -- -_
i ___ .
1 -~-- ,
5 ~'--
10 ~
-l'- WHOLE BODY DOSE FROM ACTIVITY :::::
~~~~~~
CONFINED IN CONTAINKENT _. --
\
1 4 -
10 - - . 4 fHi- '
Y Y --____.
ch . ,.
2 !!OURS
-f3 3 ,
10 ___b;i
-A i
1 l
i ,
m ti m .
g ___(__.2,g -_- 24 HOURS _
L Y< ._.
3 o ( .---_
m I
\,\ ,
\
W M
S 10 2 !
i I
A g.
(T i
\( s '
1 .
t i I I 16 lm% 'l 2I f .
\ \..; -- --- '-- -- -- -- -- ---
-- - .a -- .-- -- .-- .
i N N
s 1 > 'l \ '
\
lo - x, 1 s
\, t+- 2%-- -- ,
\ \
%< ( '
'N s
% N O s 10 x, .
s,.
m
.'s 's J L, .,
N N i
b N s EXCLUSION s RADIUS
's s '
ss s
NN %,
-1 I t ! I I i 1 l' l : .
40 100 200 300 400 500 604 700 000 900 1000 1100 DISTANCE fit 0M CONTAINER (METERS)
FIGUR'E VI-9 l
VI-16
I LJ P'] L 1If ] I T EF1 )1 s -' " ,
] T l' 11 -
- r. LL . w 11 ! ' L'I ' E im.
g 10 , i n ._ a , z _ . 1 _ a
'b F iihi'U.2.. h i U
~
i i if i n 7 5
. ii. ii. ilt! $ e t
- La <! Fjfb u # f 7 JIi u - -
i : n 7 n iE ';F
]} .,Il M dj : , P' !]
--r' Ek i @. .i d? N' /si.i H. L!
ith.k.
- C l' .
ad ar 4 -
ib ? 1 1
n md sam m e srs v 3
m en e s
(:
...r .
$ 1
- m. .t u v , , o u e s m
}A tt: W
!" 1 THYROID DOSE FROH PLUME o MF i Ik
' ~
J.
j J. IEd i L j ,
t Ait' .
,i
""W r.j ji.3.d9. d. @.I.iT.M, i I. . \.. E..i!.T..3.i. n* t :.~2 u.1 . L u
g.: 1.; .3 .W:; : ;m -
m- r .:a ;.-:
_ -.'.~' .u,'...-.m. yi.
- 7-L gd[ E(.I m e::r.u:. .:n- :.m NN...4 w :: W44 :=i :==
m & q ,. .
4 ... w ... .~.
, 44 ..
....4
..., + . ... ....-4 4 ,
%. . .I,I, 11, .
_s ._ 4 , . . .
..w
.. . L. .. ., .. . . -_ ,. .. . . J . - .
o., t1 ..... m.
E.. .
. . ..m. . - i
, ... 4,., .
.. 4
-A.. .
10s WM 3_x ,
Tu W h :
n 1
+ +
. 1 i
m u 1 3m . 4 . r - t .
l'.
i sfLin , . \ i *. ed 2 9 . i
[. j y s i.! L ( ?{ dft' hF t( f. ; W 9 .yi m. W i
I} :
4' i i 14 Msppb.J ni. p
' '- " '"" " W e r f@ c h p 1 .g h I r : -
}
} ,a 300 REM m m h. . m- A g.
- i o a '
l I s- !p \ r -
_ , iH i: o a .
Til i 1 .i jl'i SU L ,' y 'F d
\ ' fi- .m-I 4 i-.L U f l{F t l M U kT T h"11 :
l 5 . ki i kh 5 p c. :q .a
~
A .e ; .t 3: .r: JJ q
=.m t.
w m 1.3f--Mu &mdL :r.m ;% J. : d'n Q:. . m u !;J. A itd. l !'$' n.:
l.gm.:
.w -
L
=.
+
u pp, e.l, i..o,,r_ .
.y 1
1 - .i k. .,
.4
- o. m4
. ,4 _a,,
4- p-m .., ..
4 e 6+ .u-
- p x
+ +oi ...o 5..
,4 . ,t t a qs m....
_=...
i.
4 t- 4 6
{-l.
.,4 4 , h te , ...e+i .I. .
$ $w ..2+al E i 46 ., , . (44 - &+- .$h;. a. . n, 9,4 ,4
~~
' s i ii 0
h ..
. X -hY! ', t ,
w vi
. 6 i
4 m &
r a . . ,
a4 3
- x. a ,
a m .
n
- ) .i ;
p, -
N ip l' .
@ . (h I- 4 @. #i I" > .
1{ $ 4 qa Ql .44! y J %
E O bl , E I i [ I '
'h N.
v.* .e.=u
& '4'.J
,,] g. t as
. l e.
i
.d.d.o.
lt
~h -
. k (ek-I -
,h. ) .3 p
t 14 Ill. 4 i .).
- I.-6s b. ' .
an,,,as } i t. . q i tt y L -
!4 tJ h ,
' qr f \ ta .i 4 @ V 4 hh s+ t il @ . 1, :tH 1 0: e i h !@E:-U t! i
$k .
Y f.t N r.M ftN Nd o i' h O! $ 1?) M T Ik h p_
q I
g .p, ,
._. . . . p.4:
g . .
..p pi
- w. . . t.w
&' "7'
.g,..... . - . .
y., , :. , ::....g
. . .a m. . : .x.. u...a . .- -.m.
_.. . 1. . 4 .p.g _n. ;u._t.
..._= . , ;
.. . . 2 n.
4-.p... , . , 4.., . . ..,.n.
,y,... , 4
'...a. ,..m. ,..,
,. (. .4 4 . . , - .
-bb ..,.,
l .J.,,. . a : ..,..
g .r,6 .. .. , .1,'L...,a. p,. ..
.+ .
,.e r ...a. , . ., . .m _, ... . .-, .
j 9...9, .,o.. .. , , . .... n.. , ,a.
- - , + , + 4 4 v ..,. ;. .. ,
4 .p , in 6. u 4 4 eg 44 .. . ,
10 A
.m I 4
x.1 '
w! T.
4 1 i i il i1I s.
e ..
FF. .19 i j4 .1'l gii j j i ;],' I iij. p; i }
id. e _
E it: '1 j i ).f ( In r sj si i9 ,
S i] =t 2 ei j\ gj j a i i;; .t
..n.
l kNh h -b{ %N b t M lhAl El -
ll '.{ 14ki k l hMj 6'( ;i L j f.. .
I h{ 59 T y INi 41 i i h' 3; (
$5MMN. m W "
M ' '
'k :i bd-t
.sa i
i W ,
n hl r!t Pt , a
- s 3 , , .
- e, e , b - - - >
2 HOURS-a ;-4 4 % " "
W 24 HOURS N. " '
n m W, s 6 +ee Ap a -
A , , .. . . _.w. - r Git: IJ Tii SFj kis n@ i si n M 6t MM ii.4% MiW $&#@Hi %W i w m.; : x.:r q,h 4
- u. p..
bj)gp :sW, 1.
'*"t y h1h~': a y}.,,.h12 .x.42.7 _;
g l nh
.. ?:' "/:::!';p -O# ..,. :- '"
1
.,.e.n,e. i.ee,., ... ., ++ . 4 eg 4,4, .
.eee n l.a. 44e w., . + ,.p. gq ,
4,,e
.+90.P.",as n..h.t +
.a. < e+e ,41 , g,.,0 av w
- a ,e % -
. -o.g .eee 9,
.u.sep.se
- c. ,p' <. ,ti..' - -a +
9at- 4 9- .%.9 v.H.m. . ..4. = '.,e4,,4,
. .a, = . . .
,*. e .,4
=.4 4., &,-4'4 n q+.,
. .u m 4+. .$.>. ,.,.m.-
0 4 , +. .
p,,,i..s ..9 10 1I. 1 i s i i lU i t te ati r. f - t i 3
!4 !] 41.i L !
6U e ; +4 . . 4516 h; i ,I .
- i. i : . I., i ' ,1 g _ J]66 eij 'L 4 { ,f . s I M'$ Ofr,.[. T 1 Ir W15' ' E 58 .8Mi ilY'j I M
- NL bL '
1h l+;9 Hh h i f N : N$9 $ I N 1b !!f? * '
'N N IElM I l'~ $.95i d . .l. iINM k kd d '
l NN MWl ' 5i ; '
! '.hs?55 -N, il@ :M ii i NES 4 2L' I% 'Y 'L E
-)LM5 I % s i j
[mM r -
f ?
i: it__i j t t j : T 'G 7 plin, 'h qi
. a 3'
> m.'p.
4
' ' u.l ' e W U'ii i i $ I iY 'i -. T
.f m' '
q l 1.
a'
+
m w a. ny.
o i s-<m..e o m 4
i , ,i i ,n ri,1 4 m. I.
"1 q s
! n .
-- -- . e ; p g2- -;, 7 ;q -
l' spe l l i U , : .: ,
, . ._m..
EXCt.USION:- .; .-_ - -r -- , .
p n
] *._-
RADIUS _-, Y I
.I. .. .
.. m sc -
s a
w W"
10' " ++' " - " ~ " - " " "" ~ " -
100 1000 10,000 100A00 MSTANCE FROM CONTAINMENT (METER $)-
- FIGURZ VI.lo VI-17
, _ . - _ . _ - . = . _ . . . - ~ _. - . . _ _ . . _ , _ _ - . _. _ __
r 4
VII. INSTRUMENTATION in order to provide experimental measurements and continuous monitoring of conditions in the core, the Saxton Reactor was provided with a rela-tively extensive in-core instrumentation system. The system is capable of measuring outlet temperature, flow rate, pressure drop and the mag-nitude and axial distribution of the neutron flux. No modifications of the system are required or planned.
4 V II-1
VIII. CONCLUSION The plant modifications in conjunction with the changed operating condi-tion and protection device setpoints will prevent core damage in the credible incidents. Analysis of a hypothetical accident much more severe
' than the credible incidents has shown that the plant design and the avail-able separation distances result in off-site doses which comply with the siting regulations (10 CFR 100) when the operating power is 35 MWt.
l l
l 4
VIII-1 4
+ , . . - - . , , , . , . . , . . . ,e, , - , + . . . . , , - . . - , , , . . , , , - ---,,-s,- .. w - , , , , , , . .
t n>
IX. APPENDIX WESTINGHOUSE EXPERIENCE WITH g HIGH POWER LEVEL FUEL RODS i
t D
4 6
4 N
IX. APPENDIX WLSTINGHOUSE EXPERIENCE WITH HICH POWER LEVEL FUEL RODS Westinghouse experience with high power level fuel rods is illustrated in this section. Emphasis is given to the experiments developed with Zircaloy clad fuel rods. The main experiments with stainless steel clad fuel rods are only briefly summarized.
A. IRRADIATION OF SIX CAPSULES CONTAINING SAMPLES FROM THE CVTR CORE IN THE WESTINGHOUSE TEST REACTOR (MARCH AND NOVEMBER 1960)
- 1. General As part of the CVTR Research and Deve}opment Program, a series of capsule irradiation experiments (31 was devised to define more clearly the then al performance capabilities of sintered UO Pellets contained in Zircaloy-2 cladding.
2 The philosophy of the program was to carefully control irradia-tion test in order to obta1n unambiguous thermal performance data for use by reactor designers. The parameters to be evaluated at various fuel rod power levels included the effect of the initial cold diametral fuel-to-clad gap on UO3 surface and center temperatures and on the cladding stress, 8ue to the fuel-clad differential expansion.
The capsules were designed to minimize errors and potential problems which would lead to difficulty in interpreting the experimental results, and to minimize variation in both radial and axial thermal neutron _ flux.
l 2. Description of Experiment Six capsules denoted'R-1, R-2, R-4, R-5, R-6, R-8 and R-ll, each containing three fuel rods, were irradiated at futi rod power levels of from 11 to 24 kw/ft (360 to 785 w/cm).
The fuel rod configuration used was a Zircaloy-2 tube contain-ing a column of UO2 Pellets, having a fuel length of about_5 inches. All UO2 pellets used were right circular cylinders 3 l 0.430 inches in diameter with a nominal density of 10.3 g/cm l
4 IX-1
(94% of theoretical). The inside and outside diameters of the Zircaloy-2 cladding were varied to obccin the required cold diametral gaps; however, the cladding wc11 thickness was maintained at 0.032 inches in all cases. A 0.080 inch exial plenum was provided in all fuel rods to accommodate axial thermal expansion of the UO Pellet column relative to the 2
Zircaloy-2 cladding.
Cold diametral gaps of 0.006, 0.012 and 0.02f '.oshtJ were selected. Three fuel reds, each with a different cold diametral gap, were irradiated simultanecasly to eliminate possible variations in fuel rod power level. Different fuel rod power levels were obtained by using U07 jellets of differ-ent U-235 content.
An irradiation time of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> was chosen to allow fuel redistribution due to either aiatering of the UO I th'#
2 time and temperature depcodent phenomena. Gross thermal cycling of the UO, fue) ,as not des
- red. The 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> irra-diation time was $hrit enough to praclude the possibility of '
a major change in 'ineraal neutron ilux because of a reactor trip or significant Lnanges in control rod positions.
Table 1 summarizes tbo baPic fuel rod parameters used for the six rabbit capsule irradiations.
- 3. Post Irradiation Examination After irradiation, the rabbit capsules were examined in the WTR hot cells. The capsules were disassembled and the fuel rod samples removed. Length and diameter measurements were ,
taken on all the fuel rod samples to detect any deformation of the Zircaloy-2 cladding which may have occurred because
- of interaction with the UO r because of internal gas 2
pressure buildup as the result of excessive fission gas release.
- a. Fuel Rod Dimensional Changes The diameter and length of all the fuel rods were measured after irradiation to establish whether swelling or other deformation of the cladding had occurred during irradiation.
2 IX-2 i
i i
Table IX-1 ecM;NAL SURTACE HEAT
[P.ABBIT COLL _'NEAR FLUI AT !
CAPSULE F11r L ROD rt'EL !!AMETPAL POWER F%TR FUEI, RCD ELHBEF T TER EMPICf' MENT CLEAFJO2 LEYEL LVIT"JT 0.D. l Etu 3 1 U-235 Inches C-/fi V/cm Mr. Ft R-1 1-1 ?.6 0.006 266 1-2 2.6 0.012 11.0 1 0.5 360 283 1-3 2.6 0.025 2TT R-2 2-1 2.6 0.006 286 i
2-2 2.6 0.006 11.0 1 0.5 360 286 w
M 2-3 2.6 0.025 2TT La Bk k-1 3.8 0.006 kl7 s I
k-2 3.8 0.012 16.0 1 0.9 525 h12 k-3 3.8 0.025 bo2 - 1 R-6 6-1 52 0.006 kTO
- 2 5.2 0.012 18.0 1 1.0 590 h6k ,
4 6-3 5.2 0.025 -65h R-8 8-1 5.2 0.012 h6k i 8-2 52 0.012 16.0 1 1.0 590 k6k I 8-3 5.2 0.012 k6k i
i-
! B-11 11-1 T.6 0.006 62k 11-2 7.6 0.012 2k.0 1 1.2 785 618 ;
11-3 T.6 . 025 602 i [
l l
Roter . .UOp pellets 0.k30 inch diameter, nemical 9h5 dense O/U ratio 2.00-2 1 Zir aley citading dimensloos varied to give reguired cold diac-tral 6$earsneer Irradiated h0 hours in the VTR rabbit tu:e fac411 ties.
1 The following measurements were madet
-Overall length: two readings for each measurement;
-Diameters at the following positions:
One-inch from top: two measurements, 90' apart, Center: as above, One-inch from bottom: as above.
The diameter measurements taken 90' apart were in agreement with each other within the recognized pre-cision limits (t 0.0005 inches).
Table 2 summarizes the pre- and post-irradiation measurements for the fuel rod samples from the rabbit capsules.
No significant dimensional ' changes occurred during 1rradiation of the fuel rod samples. In some cases; 1.e., fuel 4:ods with initial diametral gaps of 0.006 inches operated at the higher fuel rod power levels j (fuel rods No. 4-1, 6-1 and 11-1, as listed in Table l 1), the radial thermal expansion of the UO9 fuel I
relative to the cladding should have resulted in a zero diametral gap during irradiation. No significant diametral changes were noted on those fuel rods I
indicating that the interfacial pressures between the
.UO, fuel and the Zircaloy-2 cladding were not suffi-cient to plastica 11y deform the clad.
l b. Metallography of Fuel l Selected cross sections of UO samples taken from all 2
fuel rods were pressure mounted in an epoxy resin for subsequent metallographic preparation. The radii corresponding to various microstructural features such l as equiaxed grain growth, columnar grain formation, and changes associated with the solidification of molten UO 9 were measured to establis'a radial temperature pro-files. No fuel rod (0.006, 0.012 and 0.025 initial diametral gap fuel rods)' experienced center melting while operating at .1, 16 and 18 kw/ft. Only the fuel IX-4
Table IX-2 i,i.eune.at e. mim 1. run. =>r ren tw=. ni mm wnm
.eremaio. e 0 w ,s ei ei j
a wis c e.ie me. a.: ..t '
.6 nu ee e,e tre i ,,, ee rins.i
.e e s.
t<.s t i,.ei.es ca..e ere t.e t n. ee e = >.e e = ee im ee i.e .ee i. e twe: had p ue+e r 11 t.t b.)
etmagth a 6.612 6.641 *Dael 6.6st 6 614 0 001 6.6.9 *6. 61(,
- e e.0. WV
, d 0.540 Catf e p. tor 0, $^1 0.%1 0
, seDieneter at
- 0.%01 0.601 0
- g* 0.%01 0. var + 0. te n 0.%" 0%&
' 0'IN *
- .6 0-
. 0 %.i . - 0 $.et
. eb.m.u .
Peel aos tour 12 ft
, -- 6.P Length 3 6416 6 616 0 nd/ 6.6018 6.609 .b dW4 .6l1 se e4. 6pg see.g cop g C.$0% 0. W's 0 6 $01 fi $01 0 0.%01 edel 0 i t en.e te r d' O. %% 0. SM ti 0 501 0. W + 0. W1 0 %D6 l
0.%6 OMM 0. m 0 t.#, o m- -
e0.uv..t_0. %i -
0 0. SM 0. lt4 0 twel had 6mwr 11 P.) 6-!
tengt h 1 6 616 6.636 0 4.641 634 +0.Mt 6.00 6.609 0.008 eg 0.$16 0 419 e0.001 0.$19 0.$19 0 0418 0.519 *0.006 bn ter ter e 0416 0.$49 *0.001 0.519 0.110 *0.00A s# 0.%39 0 539 0 i . 0.110 0 136 0 0.$16 *0.001
.3 0.ist 0416 0.516 0 rown <e,s.ie no. n.6 .4 u e,. 7,et ce r,e t o. i c>.,,, e lhthQ _ IMtes 3 DCMe I f.f l.f o Ite hre lhthee ere eut c .se Jhtbee Inttr e l ht b* e k Pwn 6 4 Omwr 0
64 twl 11 3 tent a 6.6t60 6 At10 .0.001 6,6160 6.63t0 0.007 6.6810 6.6 t30 0 0 0.6993 0.499% *0.0010 0 $03% 0 . % 36 e0.0001 D A964 0Ae66 0. 000t talemeter ej 0AM 0. 50M + 0. 0010 0.10 % 04M6 *0.0001
,Jm Oda .0.Ma 0 4995 0.1007 e0.129
, O d m O d u , .0. m , DJm 0. m, 0. 0a ,
tuoi he poter 67 &c-t 14 4 te ng th 1 6.6160 6.61 M 0.003 6.6100 6.6096 0.0006 6.6160 6 Alto *0.001 e 0.$0 p 0 403) .0.0001 0403) 04036 *0.0001 04032 blame te r 4, 0.%0 16 90.0006 04039 4.50%1 *0.0017 0446 0.9%8 e0.000t 04M1 0.5066 + 0.0W4 s CAMS 0.5%6 +0.0011 0.5063 0d%) *0.0002 03011 0. S t .1 *0.0010 Puol be powr 61 b) 11 3 lar.gu 6.6099 6.60f0 0.00t 64060 6.6050 .ri.tc) 6.4010 6.6060 0.001 4 0 9197 0.5161 +0.0010 blameter e l 0.5030 0 1036 e0.0006 0.11$9 041% +0.0006 0.1166 edit) *0.0009 04M1 0404% +0.0006 0d161 04160 ej 04163 0.$214 *0.0010 0 403% 0.SMS *0.0040
. 0.0004 0 5119 0. 46 *0 0009
- Langve.
DveraL1 two rootsage.length, ove restlage nede for each metourement. Dumber listee le the everage of the se biometers ter esen elens.or measurement two readings were maae anah 90s as. art , sember ateut to the tverage 6f ben meneurweente.
6 .
meneuremente mese & laen free twel toe top 3
s, . .e.emeesio .e.e et e..t,er 4
3 measurements maae 1 naan free tws! roe bottom see Durre formee eart64 eleasseabit flied 6ff tr16r to adlag length eessureesets.
l IX-5
._ . _ _ _ . ~ _ ..-
tod with 0,025 inches initial diametral gap expeitonec' melting in the central region. But no significanc diam-etral change was measured, as shown in Tabic 2. The fuel rods with 0.006 and 0.012 inches initial diametral gap and which operated at 24 kw/ft experienced no center melting. It can be concluded that the surf ace and center temperatures of sintered UO,, fuel can be lowered by decreasing the initial diametral ga and thus allow-ing reactor operation at 24 kW/ft without central melting.
B. IRRADIATION OF 'WO CAPSULES CONTAINING SAMPLES FROM THE CVTR CORE IN THE WESTINGHOUSE TEST REACTOR (MARCH-JULY 1962)
- 1. General Capsule irradiations of U03fuel rods: vere performed _to evaluate tne effeet of fuel ro8 power level on cladding dimensional changes and fission gas release.
Two capsules, designated A-2 and A-4, were irradiated in the Westinghouse Test Reactor. On capsule contained three fuel fuel rods with a 38 inch fuel length and was irradiated at peak fuel rog.gower levels of 19 kw/f t to a maxitnum fuel burn-up of 3,450 . The other capsule contained four fuel rods with 6 inch b l lengths. peakfuelrodpowery,gvelsof22.2 ;
kw/ft were measured with irradiation to 6,250 MTU'
- 2. Description of Experiment To evaluate the effect of the initial cold diametral fuel-t'-
clad gap on the radial and oxial thermal expansion of the UO 2 fuel relative to the cladding, a range of gaps was selected so that high interfacaal pressure would exist during operation in some of the fuel rods while a finite hot gap should exist at all times in others when operated at the same power level.
Cold diametral gaps of 0.002, 0.005 and 0.012 inches were used. The Zircaloy-2 cladding dimensions were varied to obtain the desired diametral clearances while keeping the wall thick-ness constant at 0.032 inches and the UO2 pellet diameter con-
' stant at 0.430 inches, i
IX-6 4
4 x
The various initial cold diametral gaps used in the fuel rods also resulted in different UOyfuel surfaces and
- center temperatures allowing the f ractional fission gas release from UO L be measured at different average U0,
, 2 '
fuel temperatures.
Two capsules. A-2 and A-4. were irradiated in the WTR. ,
Table 3 summarizes the parameters used for the design of <
the fuel rods for the two capsule irradiations.
- a. A-2 Capsule Design capsule A-2 was designed primarily to evaluate UO thermalexpansionrelativetotheZircaloy-2cladbing with different initial pellet-to-clad gaps. The three ;
fuel rods contained a 38 inch long column of UO, pellets. The fuel enrichment was varied along the length of the fuel tods to maxiaize the length of fuel operating at high temperatures.
All UO, pellets used ucre right circular cylinders 0.430 inches'in diameter with a nominal density of 10.3 g/cm ,
(94% of theoretical). The inside and outside diameters of the Zircaloy-2 cladding were varied to obtain the various cold diametral gaps; however, the cladding wall thickness was maintained at 0.032 inchen in all cases.
- b. A-4 Capsule Design Capsule A-4 was designed primarily to evaluate the
- fission gas release from sintered UO2 in fuel rods operatitig with high UO 2center temperatures.
The various initial cold diametral gaps used would result in different UO fuel surface and center temperatures.
2 The fuel rod configuration used was a-Zircaloy-2 tube containing a column of seven UO, pellets, 0.860 inches long, giving a fuel length of aBout 6 inches. A 0.100 inch axial gap was provided in the fuel rods to accom-modate any axial expansion of the UO2 P"11** C 1"*"
relative to the cladding.
Capsule A-4 contained four fuel rods with fuel of the same enrichment. The fuel rods had the same initial pellet-to-clad gaps as the A-2 capsule fuel rods.
IX-7
-. _ .. _ . _ . . . _ _ _ . ~ . _ _ _ . . . ._..__.__.u_.___._ _ _ . . _ . . _
l i
i l
[
L i
Table II-3 SIDOULRT OF FUEL CAPSUIE IRRADIATION EXPERIMNT PARADE ERS Scuminal
. Initial Peak Fuel Rod Fuel Rod Maximum Surface Beat Capsule Fuel Bod Fuel Colum Fuel Diametral Power Level Nominal Flux at Fuel Pod Mr Number Imagth Enrichment Cleerence (Actual) 0.D. 0.D.
2 x 10 s inebes 5 U-235 inches Kv/ft w/es Inches hr rt Y
-A-2 2-1 38 Variable 0.002 19.0 62h 0.k96 k96 2 36- 0.006^ 19.0 62h 0 500 k95 2-3 36 5 7 to 0.012 19.0 62h 0 506 kBS 8 55 Ak k-1 '6 k.5 0.002 22.2 T2T 0.k96 SSO b-2 6 k.5 0.006 22.2 727 0 500 576 k-3 6 k.5 0.012 22.2 727 0.506 568 kb 6 k.5 - 0.012 22.2 727 0.506 568 Note: UO2 pellets 0.k30 inch diameter, nomical 94% dense. 0/0 ratio 2.00 - 2.01 1r-2 eledding dimensions varied to give ir.itial diametral clearances
l
- 3. _ Post Irradiation Examination After irradiation the A-2 and A-4 espaule fuel rods were examined in the WTR hot cells.
The capsules were disassembled and the fuel rod samples re moved. Diameter and overall length measurements vera taken on all the fuel rod samples to detect any deformation of the Zircaloy-2 cladding which may have occurred because of interaction with the UO 2 r because of internal gas pres-sure as the result of high fissien aus releases.
l
- a. Capsule A-2 ruel Red Dimensions 1 Hessurements The post irradiation diameters of the three A-2 capsule fuel rode were measured at various positions along the lengths of the fuel rods. The overall fuel rod lengths were also measured. Table 4 summarises the diameter and length measurements made on the A-2 capsula fuel rods.
Within the accuracy of the measurements no significant diameter or length changes occurred during the irradia-tion of fuel rode A-2-2 and A-2-3 which had initial cold diametral clearances between the UO, pellets and the
.Zircaloy-2 cladding of 0.006 and 0.012 inches, respec-tively.
In the case of fuel rod A-2-1, which had an initial cold :
diametral gap of 0.002 inches, fuel rod diameter increases were found near the lower end of the fuel rod and some elongation of the fuel rod has occurred.
The Zircaloy-2 cladding used to fabricate this fuel rod had an inside diameter of 0.432 inches and 0.032 inch nominal wall. The UO, pellets used in all fuel roda were 0.430 inches in diameter.
These diameter changes are attributed to deformation of the Zr-2 cladding by the UOy pellets as they thermally expand-radially against the cladding.
Fuel rode with an initial 0.006 inch cold diametral gap or greater showed no measurable diameter change when irradiated at comparable power levels.
IX-9
,wm ---., , ,--r ,,,-,c , .,-r.-+3 --er +---m-et--*-m
- P +- e t * ~ - * - '
- v r -v7-
t r
L P
l I
i l
I Table IX-4
. I DImmmat AsD LEsc'11r w.astmpearts on rJE1 mod sArizs Pnost crm cAPsuiz A-2 :
i
- I I" "
Ftsel Dieuseter Change overall Length over Entire _ (3) over 20- I Rod Diesmetral stresured Pro Precision in Pre- Poet-Wember oep (1) g i o.oco5 Diameter ,frrediatice Irradiati m Cherwe tenat5 In. of W sth lo n. lo (inchee) (in.) (1a.) ~ mils inches (inches) (inches) (in.)
fe{ ,,
5 o
A-2-1 o.cor di o.6963 o.6958 -0.5 -o.0005 ft 0.b960 g.W o.b958 v.&960 ,
4.2 -0.0002 Lo.943 ko.960 +o.017 o.k3 o.67 I 43
+2.2 +o.00E2 i en o.b9h3 o.69T3 +3.o +o.0030 A.2-2 ^ C.006 ~ 0.5019 0.5o1% -o.5 -0.o005 [
t o.5o2o o.5o12 -o.8 -0.0008 60-955 '0 959 +0 00* 0 C9 4 o.5021 0.5019 -o.2 -o.0002 43 o.50th o.5 azo +o.6 +o.o006 t
I l A-2-3 o.c32 c.5c66 c.5052 - -1.6 -o moth o.5666 o.5064 -o.2 -o.0002 6o.965- h0.971 4.006 o.15 -
o.5061 o.5065 +o.h +o. cook o.5051 o.5051 o o (1) 4 - stee.orensent ease ik inches free tep of fuel rod q1- Steeserement ande 26 inches from tcp of fuel rod i
4 - ste=surement made 32 inches free top of ftsel red
.A 3 h - Steeserveent made 38 inches from tep of ftsel rod (2) For each diameter measure: ment both pre- and post-trradiatica, two* reedings were taken 90* epart. i Deber listed is the everage of bcth seasurements.
]
(3) Interaction between 130 2 pellets and Zirealcy-2 eledding in fiael rod A-2-1 vos gnet enough to i cease radial cladding eeformation over about 20 inches of length, i
I i
. -- t
- b. Capsule A-4 Fuel Rod Dimensional Measurements Table $ summarized the diameter and length measurements made on the A-4 capsule fuel rods.
Within the accuracy of the measurements, no significant dimensional changes were noted except in the case of fuel rod A-4-1 which had an initial 0.002 inch cold diametral pellet-to-clad clearance. The diameter of this fuel rod increased slightly due to the radial thermal expansion of the UOy pellets against the clad-ding.
- c. MetallographY of puel Selected cross occtions of UO, samples taken from the A-2 and A-4 capsule fuel rods *were pressure mounted in an epoxy resin for subsequent metallographic preparation.
The radius corresponding to various micro-structural features of the U0 9auch as equiaxed grain growth and columnar grain inf5rmation was measured to establish radial temperature profile.
No evidence of melting in the UO, can be seen in any of the camples irradiated in the A-2 and A-4 capsules.
- 4. Conclusions The results from the A-2 and A-4 capsule irradiation experi-ments enable some general conclusions pertsiding to the design of sintered UO fuel rods.
2 The results obtained from the examination of the Zircaloy-2 clad UO fuel rods' indicate'that extended operation at 2
fuel rod power levels of 18-22 kw/ft can be achieved with-out failure or fuel rod dimensionai changes if the initial fuel-to-clad gap ta large enough to accommodate the relative radial expansion of the UO 9fuel against the cladding. The initial diametral gap between the UO, and the cladding selected for the Saxton rods will not fesult in cladding diameter increases due to thermal expansion of the UO3.
(
1X-11 i
l l
,r l' .lf- ll!!!
. ..i .' l 1 ! f-l i 301 e r) 656 3"0 0
1 0
51 00 010 e e s 11 1 00 000 h t e 00c 000 000 c e h 000 000 c. 0 n e
e nc 000 000 000 300 i
l i
+ - + 5 C ( + *+ + - - - - 0 0
n 0 i
0 eI 351 51 301 g s 6 56 +
m l
- i 111 000 000 010 e c
0 (m +++ + - - - - + - + i s
. i t e 6 r
- e A
e s
h E
A ) 0 L93 7k5 9 Ts s 656 656 A
C t h e
%569Mn m26 'u 1 00 : 00 555 GG0 555 n
s) c k. i. a.
555 A c3n C00 C0G 000 a
- i. ?( i C00 t
(
u -
r N
O P
w T s g
S ) n I s 0oo 777 h L kk5 i i
3 I
7 e kkk 111 000 66'6 000 666 000 . e e) b 999 d l
A r2e 555 555 555 n
- k. k. L. o .
nE 0 P( ci
{
000 O00 000 000 d
o l
r n i
o o w
0 r e t l
e R u t
a ,
l b
a L E e .f id e adf a c .
T 'L F
Poo r is t ;
f n i t ;
3
'v rd c1 e i
=_
ee et e -
a t r ;e 0: to t e +
r eu ms gp 3 i2) j3 dd6 ddd
.{ 3 t f
b r o
rw r
is ae) dda ddd a E
w ie1 D) (:
som i c
rr r
p r-ee F
- fef -
L S t e fh hnh d at A o E
M cew nc-1 e
m tb i n t s ef u 1 e1 :
s mo e
a p re d a eee ug N.
A lC o )
ddd nna eem e l
sa ar I Cl s 2 u ee D
lrh ae 2 0
6 0
2 1 1 sss s mva at c 0 0 0 0 ttt n e na e r i en 0 0 O m 'e s t mi 0 ti i a(
ni ID ea uum al w
e r
sss l
a e
md ae it aaa f ds ee o i WMM hl
=== ecar 6
ee rm b d
or g 3 1 ou Re 1 2 3 E d A FN l m b
L L- L- L- ) ) )
eu uN A- A A A 1 2 3
( ( (
F 1
- 1! i wxeM" fa Illl IeI! ji ._
1 C. LRD IN-PILE TESTS PROGRAM IN THE SAXTON REACTOR
- 1. General The purpose of this program has been a) To perform in-pile proof tests to verify technical feasibility of prototype designs, materials, and fabrica-tion variables proposed for use in a large plant chemical shim environment, and b) to perforu fuel and cladding experiments aired at reducing overall fuel cycle and plant costs. ,
A series of subassemblies, which in most cases represents a combination of these objectives, has been irradiated in the Saxton Reactor. The present status of each experiment and significant results to date are detailed in TS.ble 6.
Inthasubsequentsectiggyyg)performanceofZircoloyclad t
fuel rods is examined.
- 2. Description of Experiments and Post frradiation Examination of High Power Level Zircaloy Clad Fuel Rods Table 7 summarizes the Zircaloy portion of the LRD irradia-tions program.
The type of Zircaloy used as cicdding material, the peak power level and the peak burt.up are reported for each experimental fuel rod.
- a. Zircalov Clad Fuel Rods from Saxton Modified 3x3 Subassembly No. 503-4-23 Evaluation of the in-pile performance of the Zircaloy clad fuel rods irradiated as part of 3x3 subassembly No. 503 23 was completed in the period April-June 1965. Two as-pidled and three autoclaved pre-oxidized Zircaloy clad fuel gods designed to operate at 16 kw/ft (530,000 btu /
hr-ft ) were irradiated._ The rods operated as part of the Saxton core for a' total of approximately 58 effective full power days at a maximum clad surface temperature of 640'F.
During this time the rods achieved a burnup of approximately 3000 MWD /HTU.
IX-13 >
t I
lI I I ,'
i l. I. I. 15.
i i j j j s l}, 1]. l l l'
, ,' , r i s.
l
' }
j iei s
} .;I J
i i
.l'is i y
, 1(i I
e
- 18 'li'all' l *
13 ,
llL I3, i'*3l' l !
!. !J l' ll !
, . I '1 1. 'j ' -] 4 a ';
3
- y 1
, .1, il ll s'e!lil l
'. .,}, ell'lls) i,ll
} i.,
u ]g. % i
, % % ". ' . %% 5 -
llL : i I i i i ii liii i i . i 1.!
! ! I I I i ii till 1
-) >
s m l.
u 11 ,
i i , , , ,, .... ,
l ill
! lLi 1 ,
1 i i . $1:11 All I i ! ! ! !! !!!! !
141 1 1 1 1- t i !! til! 'l
{l I i ia i 13, lI'g.t lls!i,y!,i 3 II t '
lj 83 i il1ll l ,
gi 11
{Iij
- i. 4 s,ii l, 8 It i. ,
i,i i i. '
ij 3 i' il aj. ,,
i y- ' l. g i i, * . y .it s *<
- l 3., ,
T !! 11.
~ . .. ..
1! 1 is11!!:IIill!11:l1 i ll
. 1.111liiLl!
1 3 3 3 3. } ,i l i 3, 3 it :1 ul: 11: 11: 1 lia} 1: I u i i l i l i i T'.-14
+
w.,-,-...gm. ., g e, 9,m.,..J..y,,..,,,g, g_e,, ,,,,,ymme,,,,,,r,w,,-,%%,..f,',y ,..,,,...,%m,g,,gg,,,,m,%.,,,..pf5,-w w w,% ,M I ' ,, p-~.,,#.,q.- . . +
_ _ _ . _ _ _ _ _ _ . _ _ _ . ~ . . _ _ _ _ _ _ _ ._ . _ _ _ _ . _ . _ . = _ _ _ _ _ - . _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . . . _ . _ _ _ - - - - - ~ _ - .
Y . 1 . '/
8thet&RT OF 1Wif 8r CIAb IW RAIT05 lorfste full tower teek teek tuk me, et
$essee Assembly he.
Aeeambir htt , tire, k.As listerial Ceeditleg Treateeet Aars Eme. A4 _ ha t Flug,, m 1 (av/ft) letu/hr/ft) selDhrrV)
St.2 unsated terwee 1M 13.4 450.000 9.$00 Mbe.fi )m) 4 i
(pl. free) , tre-etldited)
Ameealed Foresee 190 4.$ 1$0,000 3.400 MbeeH ls) 4 in)
(s4. free) Fre-estdised , , ,
Aeonated A tacleve $4 16.0 350. W 3.900
$0be.33 1st 3 It 2 (pt. free) tre-estitend >
emmaaled As.pte kled la 16.0 350.000 8.900 MS-4 23 3a) 3 t ra t (pl.freel 14.0 2.900 M be.te ts) 3I *) St-4 101 C.W. Autecieve 16 445.000 tre-estsisee Fernece 244 U.S 450.000 12.300 M be 14 3:3 1(b) tr.1 Assosted .
. (si. free) Pre-estdiend H.S It. M0 M be.to )m) II 'I te.2 Aenealed Pereste Pre-estdised 346 450.000 ,
(ut. free)
I8) ,8t.4 101 C.W. Foresee 16 H.S 450.000 2.600 S03-4 24 3sl A tre-eatolsed 101 C.V. Butf. Pret 6 99 14.0 441.000 3. M0 M beatl 33 1 tr4 Autecten treealdised 101 C.W. As.Sett. Prof. 99 14.0 443.000 3. M0 M be.23 ):3 1 86 4 Fureate 3M 13.8 e M.000 11.600 M be.25 an3 1('I tr.3 Anno 61'e d *
(Pt*ftee) 6.800 30b4 2$ 3a3 I IU tv.t Annuted Autotleve 1M 16.0 150.000 (pt. free) tr.4 101 C.W. Aeteeleve ill 12.0 400.000. i.000 Sob 10-1 tst 3 tre-esidised Autocleve H.S 40.000, 8.000 Mb101 N9 $ tr-4 103 C.W. ill Pre-esidised St.1 101 C.W. Auteeleve 139 14.0 463.000 8.(00 10910-1 tal 3
- tre-esidised tr.3 101 C.W. Amtesten 15) 14.0 445.000 8.000 M b10-1 tst 3 (si ftes) Pre-eeldtsed tr.4 101 C.W. . Autoelen ill -14.0 445.000 8.000 Mb10-1 tal 3 Pre-estdised
- tr4 101 C.W, ' Ae.f tekled 1$5 14.0_ 44l.000 4.000 S0b10-1 tot 3 tr.4 101 C.W. Pureece ill 14.0 46).000 4.000
$01-10-1 tst 3 Pre-estitsed tr4 101 C.V. Antecieve US 84.0 465.000 8.000 50>10-1 tst 3 Pre-eatetted l oe O.D. 6 1.D.
3 tr4 101 C."J. Auteeleve 15$ 16.0 350.000 8.000 M b 10-1 tut i
tre-esidised 1e7758e i
i
! (e) Centete looes estde U02 I"I
- Ib) Preetemely trtedteted to Assembly Ite. Mbe.31.
l l (t) Provisuetr tiredteted to Aaeambly me. Mbe.11. (Centales il all die. defect.)
(d) Cestetes 15 ett diametes defect.
(e) Prevleuely trtediated to Assamely Dee. M b4.!! and .24 (f) Preetowely irredteted to Aeeembly us. S?be 23.
f r:-15 1
The main purpose of the experiment was to determine the effect of preirradiation surface treatment on Zircaloy- ,
clad fuel rods when exposed to nucleate boiling heat transfer conditions in a chemical shim PWR environment.
The post irradiation examination indicated satisfactory in-pile performance of both the pre-oxidized and as-pickled fuel rods irrespective of surface treatment prior to irradiation. No dimensional changes or abnormalities were observed on the fuel rod surfaces.
- b. Saxton In-Pile Defect Test 3x3 Subassembly No. 503-4-24 1
Visual examination of the intentionally defected Zircaloy clad rods war completed at the Post Irradiation Facility hot cells. The examination yielded the follow-ing observations:
J) No fretting was observed.
- 2) Crud deposition was very light.
- 3) No indication of attack of the cladding or fuel was observed in the vicinity of the defect.
l Therefore, even if a defect were to occur no further undesirable problem, such as defect enlargement, clad bursting, etc., is expected,
- c. Saxton Special 9x9 Assembiv No. 503-10-1 j The operating characteristics of this assembly are shown l in Table 7. In-core examination was successfully performed with the aid of a boroscope upon completion of the crud test. The assembly has experience approximately 8000 MWD /
MTU. No indication of failure, cracking, or attack of either the Zircaloy or stainless steel cladding was observed,
- d. Saxton Advanced Fuel Assembly No. 503-4-25 The operating data are shown on Table 7. The present general appearance of the subassembly is satisfactory. No fretting, cracking, or attack of the grid or cladding mate-rial was observed.
IX-16
-e - , - - , . w , - , m,,.
- . _ . . - - _ . . . . . , . . - _ . . . _ . _ , . ~ . . . . - . _ - - . - . - . - . - - - - - . -
P D. SAXTON SPECIAL 2x2 STAINLESS $ TEEL FUEL RODS SUBASSEMBLY No. 503-9-1 Four fuel rods have optrated successfully in the Saxton core at a peak power rating of 25 kw/ft to a 'surnup exceeding 6500 MWD /MTU, There was no indication of cracking or swelling of the clad and the overalt appeasante of the rode Vas excelaent.
E. NASA-PLUM BROOK REACTOR-HIGH POWER *HIGH BURNUP-IRRADIATION PROGRAM UO, fuel capsules are 'geing irradisted in the NASA - Plum Brook Reuctor as part of the High-Power, High-Burnup Irradiation Program.(7)
Fuel pins containing 0.300 inch diameter pellete 96% dense with a 6 inch fuel column are clad in Type 304 stainless steel. The cap-sules .*re being irra.fint ed at mean linear heat ratings of 20 to 47 kw/ft, to a maximum burnup of 80,000 MWD /MTU. Four capsules have been irradiated to 10,000 MWD /MTU at a peak power rating of 49.6 Lw/ft. Three of these capsules experienced no clad deformation even though they were exposed for a long time at a very high power level causing large fragmentation in some fuel pellets and center cavities with a diamuter as large as 20% cf the fuel diameter in others. All of the feel nellets had experienced center melting.
In the fourth capsule, over 75% of the cross sectional area of the pellets melted due to exposure. at extremely high power levels.
Part of the clad melted (possible because molten uranium was momen-tarily in contact with the cladding). The center of the pellets shifted about 13% of the fuel radium toward the molten clad zone and an internal cavity, whose diameter was about 45% of the fuel diameter was formed. However, na expulsion of uranium into the coolant or excessive clad deformation occurred, i
Three other capsules were irradiated in the Plum Brook Reactor in aprogrrmdecignedtomeasygythethermalconductivityofUO,at i
tcLyeraturea up to 2300'C. The UO fuel columno were 4-172 inches j long and 1-1/4 inches An diameter. Theyweresuccessfullyirradiated l at red powers of 22-25 kw/ft.
l F. FUFL PIN IRRADIATION IN THE GETR Four vibratory compacted and two pelleted fuel pins w irradiated in the GETR st peak rod power of 21 kw/f t.ggy successfully The pins
- were 5.2 inches long, had an active fuel dianeter of 0.56 inches and veri 304 stainless steel cladded. The pelleted UO was 88.3% dense v!.1. the vibratory compacted UO was fr m 80.4 to2 86.7% of theoreti-cal density. Residual diametral 2expansion was observed in both pellet rods, which had extremely small diametral gaps of 0.001 inch.
(
IX-17 l
. - - __m _ , _ _ - - __ - , . _ __~ _ ._ . _
Ilowever, no diametral increase was observed in the vibratory compacted rods, operating at comparable power ratings, The difference is attributed to the ability of the vibratory compacted fuel to utilize internal volume space to compensate for thermal expansion.
i
.I IX-18 L
es
l l
l 5 REFERENCES I
(1) Christensen, J. A., Allio, R. J. and Biancheria, A., "Helting )
Point of Irradiated Uranium Dioxide," WCAP-6035 (February ;
1 1965)
(2) Lyons, M. F., Coplin, D. H.. Pashos T. J., and Weidenbrum, B.,
"UOy Pellet Thermal Conductivity fron Irradiation with Central l Helting " ANS Trans. , Vol. 7, No.1 (June 1964), p 106-107 l 1
(3) Duncan, R. N., " Rabbit Capsule Irradiation of UO "2 CVNA-14i, j (June 1962) ,
(4) Duncan, R. N., "CVTR Fuel Capsule Irradiations," CVNA-153 f (August 1962) I (5) "LRD Quarterly Progress Report, January-March 1965." VCAP-3269-12 (6) "LRD Quarterly Progress Report, April-June 1965," WCAP-3269-13 (7) Schreiber, R. E., " Irradiation of Refractor Fuel Compounds, UO,, and UC, at High Specific Power to High Burnup - Post-1rfadiation Examination of Capsule 1" (NASA Report to be issued, 1966)
(8) Balf our, M. C. , Christensen, J. A. , and Fenari, ".i. M. ,
"In-Pile Measurement of UO Thermal Conductivity," NASA-2 CR-54740, March 1966, also WCAP-2923 (9) Balfeur, M. C., and Ferrari, H. M., " Irradiation of Vibratory Compacted UO 2 Fur.1 Elemencs," ANS Transactions, Vol. 8, No. 1 June 1965 IX-19
_. --- -.- . .