ML20003A666
| ML20003A666 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/22/1969 |
| From: | Walke G, Wall H CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P, Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8102040610 | |
| Download: ML20003A666 (17) | |
Text
r CORSumCES y(
c Power l
e/a. /
Company Generaf Omces; m2 West M6chigan Avenue, Jackson, Michigan 49201. Area Code 517788 0550 n.'i Deeember 22, 1969 d
- /
p.et-m Cy-Dr. P. A. Morris, Director Re: Docket 50-155 Division of Reactor Licensing DPR-6 ZEK United States Atomic Energy Commission Proposed Tech Speo Washington, DC 205h5 Change 19
Dear Dr. Morris :
Attention:
'Mr. D. J. Skovolt Transmitted herewith are three (3) executed and thirty-seven (37) conformed copies of a request for a change to the Tech-nical Specifications of License DPR-6, Docket No 50-155, issued to Consumers Power Company on May 1, 1964 for the Big Rock Point Nuclear Plant.
This proposed change (No 19) will enable Consumers Power Company to insert into the reactor at Big Rock Point a fuel design designated as "EEI-UO2-Pu02", which will permit the irradiation of plutonium-uranium mixed oxide fuel. The purpose of this irradiation is to provide needed data on the operating characteristics of mixed oxide fuel with a statistically significant number of fuel rods.
It is our intention to insert "EEI-UO2-PuO2" fuel into the Big Rock Point Reactor during our next refueling outage which is currently scheduled for February 1970. We would, therefore, be most appreciative of an expeditious handling of this Request for a Technical Specifications Change so that we might receive approval before February 1,1970. We recognize that this is a contracted schedule for a Technical Specifications Change.
By way of explana-tion, we would like to point out that there are four parties involved in the various contract negotiations - USAEC, EEI and two utilities.
It was easier to resolve the technical issues than the contractural issues.
Yours very truly, d-<
p m,
Nuclear Fuel Management Administrator GJW/dmb c.d gg owpy-
~
~
i DEC2 91959 i
j ym s 3 :y)
Q yo@ Qb stcutme cy, occm cttM Q 7r g l
l T41\\
1.
Re$V=l
'a i.ek sr,1*-?:*::! " A #
AF Consumers Power Company Docket No. 50-155 Request for Change to the Technical Specifications License No. DPR-6 For the reasons hereafter set forth, it is requested that the Technical Specifications of License DPR-6 issued to Consumers Po er Company on May 1,1964,= for the Big Rock Point Plant be changed as follows:
A.
In Section 5.1.Sa, change to read as follows :
" Enrichment of Fuel", approximate weight percent U-235 from 2.6 to 5.2, inclusive. Approximate weight percent of plutonium (fissile Pu-239 and Pu-241) 1.0 to 10 in normal (0.7 w/o U-235)
UO
- 2 B.
In Section 5.1.5b, change to read as follows :
P us UO -Pu0 in 84 bundles.
l Total nominal weight of UO 2
2 2
C.
In Section 5.1.5 add Figure 5.9.
D.
In Section 5.1.5, replace the present table of fuel bundle parameters with the following table:
2f fil e
v
9 J. g i s *@fe*
g*
e
==e e r ~* e 7.,,
F1II. Pl?:DtIt 5.15 (Cer* 1)
Pescarch and Develorment Origina l R e loa d P e loa d Reload Centermelt Centermelt EEE Ce ne rs i (A )
(%C)
U)
(E -C)
"D" Fuel Intermedi+te A har:ced
" W ified E C" kT *Pu0p Ceomet ry, Fuel R od A r ra y 12 x 12 11 x 11 9x9 9x9 11 x 11 8x8 7s 7 9m9 9m9 Rd Pitch, Inches 0.533 0.577 0.707
- 0. 7f" 0.550 0.807 0.t 21 0.707 0.707 13 74 70 109 3e 29 52 0
Standard Fuel Rods per Bundle 12{
10}
8 Special Puel Rods per Bundle 12 73 11 12 28' 20' 297 81 35 Spacers per Bundle 3
5 3
3 7
5 5
3 3
Fuel Pod Clndlins Nterial 30455 Z r -2 Zr-2 Z r-2 304SS,Zr-2 Zr-2 Zr-2 2r-2 with various Zr-2 Inconel f,no and/or initial mechanical lacoloy 600 prope rt ies StsndJrd Rod Tube Vall, In.
0.019 0.034 0.040 0.040 0.010 to 0.030 0.035 0.040 2r-315-ISn Inc lus ive 0.040 Specia l Rod TuSe Wall, In.
0.031 0.031 0.040 0.040 0.010 to 0.030 0.c35 0.010
- 0. 0<.0 0.040 loclusive Fuel P Ws Standard Rod Diameter, In.
0.388 0.449
- 0. % 25 0.5625
- 0. 42 5 0.570 0.700 0.5625 Special Rod Diameter, In.
0.350 0.344 0.%25 0.5625 0.320 0.570 0.700 0.5625 0.5625 y
6 Tuel Stacked Density, Percent 94 + 1 94 1 1 Pellet 40-95 Pellet 94 Pellet 90-95, inclusive 94 Pellet 94 Pellet 94 Pellet 62 Theoretica l 85 Pewdered P5 Pewder
% Powder Ant t va M Lerch. Inetce S t a nd.e rd R od 70 70 69.75 70 68 to 70 66-67.3 e 5-66.3 70 70 Specia l Rod 59 (Corner) 64.6 Central 64.9 Centra l 04.9 cent rel, 66.6 Removable Till Cas Hellian Helium Helium Helim Heli m Helium Helium Helium Fe ' i ten 1 Four special fuel rods at bundle corners are segmented.
2 Reload B, C, E and E-C fuel bundles may contain (in the corner regions of tbc bundle) four 2r-2 tubes having encapsulated cobalt ta rgets sealed within.
3 R e lre d E a nd E -G fuel bundles have a special central fuel rod to which the bundle spacers are fixed. In addition, two of the interior bialdte f ue l ri d e a re remova b le a nd ma y centain UO -Puo fuel.
y y
4 Special rods have depleted uranim.
5 In addition to special rode for reload E, reload E-C has four gadclinia containing rods.
6 With 37. dishing on selected rods.
7 U%.Puo fuel rod stack density will vary f rom 74 to 92 percent theoretical by using annular, dished, or non-dished tellets in selected rods.
8 64'UO -rug rr s similar to standard UO rods, 4 removable Puo rods. N gadolinia containing rods, 4 cobalt corner rods and 1 emrty (water-filled during overstion) y y
2 y
y spacer rod 9
l2:=
r--
s-P00R ORIGINAL
/~1 n_
G D.
e CillAljdV!dik!aNCildl 2 (GL DN5 i
"F
- t l
I r
+r I g.
yn
!..~!
bh sfll H..YY.
j g[
j
- j w 5 6->c -c
,t 1
. ii JL JL u?JC J o
__[
l!
If
^ ^ Or Or ^$7r ;
a
'=== h=!
L4 S y44c i ja E
l I
L-x u. -:
V N
JL A u
]4.a[m-h
' ^ ar 7c,^
f KQf
.0
'%_!44
'7 g
l w n. -t -sre a a
c
'I J
k l
h x
!,l2 I
l Y{
~
Thlhhhh I
m)d m
!?
l}
I o !l
,i ae u
Qa t.,l h
il g
j
! ;j m
k l
?
m u i d l.1lGXMLEKK}nl 1
- i ce-e is
!!E;
?
s!
5 e
we b
[y.c %gwg[d\\
l a
~; y-g 2f I
' i f._
I
2.
II. Discussion - EET UO -Pu0 Bundles 2
2 A.
Procram Description 1,
The EEI Program for the thermal reactor utilization of plutonium inc.ludes a g.
't a test of three Pu0 -UO C ntaining, prototype bundles. The program 2
2 M
objective is to design and test bundles which are interchangeable with y
regular Big Rock Point reactor bundles. These bundler are intended to demonstrate behavior and performance lifetime of Pu0 -U0 fu 1 bundles 2
2 relative to UO I""I*
2 B.
Fuel Description The EEI Bundle, like the "E-G" bundles, are designed to operate for four cycles and achieve an average burnup of 20,000 MWD /T. The design has five different types of plutonia rods. Four types are used to provide an acceptable power distribution and the fif th type provides a test of 80 percent fissile plutos.ium. Four cobalt rods with 35 gm Co/ft were retained for consistency with the "E-G" design. One spacer-capture tube will be filled with water at the center of the assembly. Eight UO -
r ds augment control in a manner which matches the "E-G" deskgn.
Gd 0 23 Eight removabic rods are included in the design - four cobalt corner rods and four plutonia-containing fuel rods. The performance of the 2
fuel will be monitored through examination of the removabic fuel rods.
The bundle design is physically the same as the Reload "E" and "E-G" fuel. The only differences are:
1, four removable fuel rod positions are used (instead of two).
2, the central spacer rod contai.s no fuel and is perforated to permit water ingress.
3, eight gadolinia rods are included (instead of four) to match "E-G" poison reactivity control.
The enrichment distribution and local peaking factors are arranged so that established "E" and "E-G" Technical Specifications apply.
The position and number of gadolinia containing fuel rods has been changed as their reactivity worth is affected by the presence of plutonium. The gadolinia-containing rods do not contain plutonium.
.t Figure 5.9 and Figure 1 shows that fuel rod types, positions, and the Ie enrichment distribution within a bundle. Four plutonium enrichments were selected to give adequate power distribution. The fifth plutonium containing fuel rod type contains plutonium of which 80 peccent is fissile.
The other four plutonium containing rod types centain about 90 percent fissile plutonium. The 80 percent fissile plutonium is deployed in four of the removable rod 'ocations.
3.
The non-plutonium containing rods, i.e.,
four cobalt, eight gadolinia,
and the spacer rod, are mechanically identical to "E-G" design except that tre spacer rod is empty and perforated.
The plutonium containing rods are also mechanically identical to "E-G" p
UO c ntaining rods. The Pu0 -UO e nt ining r ds are identified by 2
2 2
serial numbers on the lower end plug.
d M
The Pu0 -U0,, rods all contain cold pressed and sintered fuel pc11ets 2
of annular Elesign prepared from mcchanically blended, ceramic grade UO and Pu0 p udere. The annular hole is 0.150 inch diameter and the 2
2 fuel matrix density is 92 percent. The only rod-to-rod variation is the plutonium enrichment,which is identified by varying the upper end plug diameter, The thermal performance of this fuel will be similar to low density UO 2 fuci except that the annular feature causes lower fuel center tempera-tures relative to solid pellets of the same density. The plutonia fuel in al'. three bundles will operate well below melting at 122 percent 2
overpower (500,000 BTU /hr-f t ). The peak fuel temperatures at 500,000 and 410,000 BTU /hr-ft2 are 4606 F and 3840 F, respectively. Since all fuel is 92 percent dense, the thermal conductivity integral has been reduced from the "E-G" standard. The corrected integral and equation a re :
g2805 C Kdt = 85.5 w/cm.
N
- C 0
31E
. 6 x 1 8 G + 460[ w/cm K = 60 1+T
+
This corrected integral was derived for low density UO,,,
Previcus s ubmitta ls (13) have documented the observation that UO -Pu0 I" 1 C "~
2 2
taining small amounts of Pu0 has essentially identical thermal perfor-2 mance.
C.
Nuclear Desinn The nuclear characteristics of the Pu0 -UO bundles were calculated 9
2 using standard GE nuclear methods. The enrichments were selected to give the powe distribution shown in Figure 2 The peak rod power is pur-
[
posely Jc,ated in the removabic fuel rod positions. The highest local
- I peaking '.ctor is 1.287 which is less than the 1.3 peaking on the plutonia 7{
rods in the "E-G" bundles. The local power distribution becomes less peaked with exposure as illustrated by Figure 2.
The reactivity values and power coefficients for the "E-Pu" des,1gn are shown in Table 1.
These coefficients are essentially the same with the exception of the void coefficient. Insertion of only three bundles will have an insignificant effect on the core void coefficient.
r 4
4 The isotopic content of the plutonium used in these bundles is as follows:
"80*/" Fissile "90*/," Fissile Pu-238 0.268 0.104
- s Pu-239 75.356 86.919
't Pu-240 18.238 10.162
_g
- f Pu-241 4.956 2.532 Pu-242 1.182 0.283 D.
Thermal Hydraulic Analysis The thermal-hydraulic characteristics of the "E-Pu" bundles are essen-tially identical to the reload E-G fuel with two plutonia fuel rods in the removable rod positions. The local peaking factor was reduced to 1.287 in ihe "E-Pu" design as compared to 1.3 in the previous plutJnia fuel rods now operating in "r G" bundles. If necessary, these bundles will be placed in core positions that have radial power factors similar to the sixteen bundles now containing plutonia. The resultant thennal-hydraulic performance provides additional margin from the mini-mum critical heat flux ratio (MCHFR) limit, 1.5 at 122 percent everpower, due to the reductions of water quality in the bundle.
Ccre thermal-hydraulic analyses have been performed on predicted core configurations which indicate that all license 1Laits will be met.
During the refueling outage, these analyses will be performed on the finally-selected core configuration.
E.
Special Handlinn Procedures The three bundles will be shipped to Big Rock Point in a regular RA-1A container which is being licensed separately. Each bundle will be enclosed in a sheetmetal container which provides secondary containment -
during shipment. These containers will not be opened until they are inside the Big Rock Point containment vessel. Once removed, the bundles will be handled in an identical fashion to UO f"*1*
2 F.
Accident Analysis 1.
Reactivity Excursion Analysis a.
Postulated Reactivity Accidento
[
The Big Rock Point reactor operates with one specified rod 4 L withdrawal pattern. The rods ar? grouped in banks of two or more; all the rods in a bank are withdrawn together, with a procedural Ibait of two notches between any two rods
~
in a bank. This sequencing prevents large rod worths; however, an operator error or series of err' rs can result in larger worths. The possible rod drop situations and rod stre gths when the core is critical and at hot standby are:
Li,
' \\.
5.
Case 1: In-sequence potential of.008 Ak for drop
'from full-in position to drive position.
Case 2: In-sequence potential of.021 Ak for drop
- s from full-in to full-out.
^t L!
Case 3 : Out-of-sequence potential of less than
- f
.021 Ak for drop from full-in to full-out.
Case 4: Maximum theoretical worst case of about
.045 Ak.
Case 1 requires the following equipment malfunctions and operator error:
a) Rod becomes uncoupled from drive.
b) Drive is withdrawn (in-sequence), but blade hangs.up temporarily. Operator does not notice that blade is not following.
c) Rod then unexpectedly releases and drops from full-in to position of the drive due to gravity.
Case 2 requires an additional operator error of withdrawing the drive completely rather than concurrent with the bank.
Case 3 consequences are less than those for Case 2.
C :sc 4 is considered hypothetical as it requires still further compounding errors beyond those enumerated above.
Case 2 at the hot standby condition was used for this analysis.
These are the same conditions used by DRL for their analysis of the centermelt fuel (1).
b.
Kinetics Calculations The most important parameters in a nucicar excursion kinetics calculation are:
- 1) Quantity of reactivity insertion
[- !
- 2) Rate of reactivity insertion I!b
- 3) Specific power distribution
- 4) Doppler coefficient
- 5) Resonance neutron flux distribution
- 6) Initial power
d i
6 0
The only significant difference between the " current"*
core and the "EEI Plutonium"** core is in the specific J
Power distribution. The plutonium bundles have the samu power producing capability as standard reload fuel and
, f-
- peaking factors that are very similar to the standard 1
However, the plutonium fuel is of an annu-reload fuel.
,g lar design which reduces the mass of fuel that contains
- [
the energy generated during a transient. The effect is 1
to raise the plutonium fuel energy density in any given accident by 13%. The effects on mass of fuel above given energy levels are shown below:
.021 Ak Rod Drop at Hot Standby
~
" Current" "EEI Plutonium" Core Core Peak Enthalpy (cal /gm) 450 450 Mass of Fuel (kg) above:
425 cal /gm 1.0 1.0 1
330 cal /gm 26 26 265 cal /gm 37 49 j
230 cal /gm 58 67 As can be seen there has been an increase in the mass of fuel above 265 cal /gm and the mass of fuel above 230 cal /gm. It should be noted that these increases will occur only if pla-to:;ium is loaded irradtstely adjacent to a centermelt bundle.
If all of the EEI plutonium bundles are loaded next to a centermelt bundle, the figures above would still apply.
c.
Primary System Intecrity As discussed at length in prersious applications for this plant, the integrity of the primary system depends upon the severity of any steam explosion. The severity cf o steam explosion
's depends upon the following factors:
- 1) Time of fuel failure 7b
- 2) Mechanism of fuel failure
- Currently licensed core
,---~w
7.
- 3) Amount of fuel failed
~~
- 4) Energy in the failed fuel
'f'
- 5) Heat transfer rate to the coolant
-L Q[
- 6) System geometry As has been shown in previous applications a severe steam explosion will result only if there is a significant quantity of promptly dispersed fuel in the moderator. For material to be promptly dispersed it must attain an energy density on the order of 425 cal /gm or more. The above tabic demonstrates there is little material in this range for all considered con-ditions.
A large quantity of data has been obtained recently in the SPERT IV Capsule Driver Core (2-8). These data and earlier data indicate that fuel subjected to a transient energy depo-sition of 275 cal /gm or less remains intact (is not dispersed) after the transient. This also applies to fuel that has sig-nificant burnup (even though the cladding may fail). This is consistent with the latest calorimetric data for UO2 (9-10) which indicates incipient melting occurs at an energy level of 269 cal /gm. Recent tests with physically blended mixed-oxide fuels have given no indication that this type of fuel behaves differently from conventional uranium fuels (11-12).
The results of tests run at 225 and 274 cal /gm with the mixed oxide fuel were virtually identical to results obtained with uranium fuels tested at these icvels.
In the previous license for plutoniem fuel (13), the above information was not available and more conservative assumptions were made as to failure threshold. In light of the new test data, a conservative threshold for dispersal of mixed oxide fuels, as with urania fuels, is 265 cal /gm, as used in the supporting evidence for Change 18 to the Big Rock Point Tech-nical Specifications, the same as uranium fuels. This analysis was based on that fact.
Even if one promptly dispersed all of the fuel above 265 cal /gm,
+,
the energy in the dispersed fuel would amount to only 61.5 NW-sec. This is below the 64 HJ-sec that was considered tolera-
}r ble in the DRL evaluation of the centermelt license. An
- [
evaluation of the consequences calculated by DRL for a 64 HR-sec j
deposition indicates that they are conservative by approximately two orders of magnitude.
As evaluated in the license application for Change 17 to the Big Rock Point Technical Specifications, the bone dose at the site boundary does not change due to the addition of plutonium l
to the core. This is so because plutonium is a non-volatile solid and the fuel vaporizations must occur to release non-
~
8.
volatile solids. However, none of the plutonium is calculated to vaporize as a result of the postulated.021 tak rod drop accident. Nor is there calculated to be any vaporization in the case of a complete core meltdown.
, e-
'I d.
Conclusions
_ g It is concluded that the results of a postulated reactivity accident are slightly more severe in the "EEI Plutonium" core than in the " current" core. However, the results are still within an envelope considered acceptable in granting the license for the " Current Core".
It is also concluded that there is no
- danger of breaching the primary system due to a credible reac-tivity accident with either core loading.
2.
Loss of Coolant The loss-of-coolant accident was discussed at length in conjunction with Change 14 which allowed insertion of reload "E" fuel. The addition of these tundles to the core will not increase the severity of the postulated accident. As mentioned above, in discussion of core thermal hydraulics, these assemblies will be placed in core location with lower power factors in order to readily meet thermal limits. In addition, the annular fuel will operate at a lower bulk average fuel temperature relative to solid peligt fuel for a' given linear power. At full power (410,000 BTU /hr-ft ) the peak fuel temperature in annular fuel is 38620F compared to 44000F for solid j
fuel and the fuel volume is 10 percent less so less specific heat is available in the fuel. The results 01 any postulated IDC accident will be less severe because of the reduced bundle stored energy.
III. Conclusioni i
Based on the above analyses and comparisons with E and E-G fuel, the following conclusions concerning the EEI UO -Pu0 fuel bundles are made:
2 2
1 1.
Fuel rod and bundle mechanical design is essentia 11y identical to E /E-G.
2.
The local power factor is slightly higher for some UO -Pu0 9
2 rods than the UO r ds in the E/E-G design. The local power 2
factor is slightly lower than UO -Pu0,ll be located in radialrods previously inserted 9
in the core. The plutonium bundles wI positions so that the peak rod power will not exceed the design J!
peak power for the E/E-G fuel.
- L 3.
Peak fuel temperatures in these bundles will be less than the solid and dished Pu0 -UO Pellet containing rods previously 3
2 inserted in the core.
4.
The 0.150 inch annulus selected for these pellets is conserva-tive in that a 0.200 inch annulg has shown good structural integrity by previous analysis.-
J 9.
5.
The results of a postulated reactivity accident are slightly more severe in the core loading with these three EEI bundles.
There is no danger of breaching the primary system dup to a credible accident with this core loading.
"9
'I
-6.
The bone dose at the site boundary after a postulated accident
- {
does not change due to the addition of these Pu0 -U02 coataining 2
if bundles.
7.
The severity of a loss-of-coolant accident is probably less in this core because of the lower heat content in annular fuel.
Based upon the above considerations, we have concluded that the use of three E-Pa fuel btmdles in the Big Rock Point reactor does not present a significant change in the hazardous considerations described or implicit in the Final Hazards Summary Report.
CONSUMERS POWER COMPAh7 By Vice President Date: December 22, 1969 Sworn and subscribed to before me this 22nd day of December 1969 uw 1
A wt t )
Notary Public, Jackson County, Michigan My Commission Expires January 15, 1972 2f iL e
+-
-s y-
10 Figure 1 BIG ROCK POINI - EEI PHACE II EPU BUNDIE DESIGN Gap
' t.
'1
~.
~~~
- s.
(
s ',
AF Co 1
I 5*
2 2
.l 6
i I
N 2
4 M
4 3
3 l
l!
i l
l 3
3 l
M I
I l
i 4
i 4
i j
Octant Symmetry j
l l
l 1 __.
l
- Special, removable rod in four places
- Rods h
wlo Pu Total Pu Fissile Fraction U
Comments 8
1 1.624
.90 Natural 150 Mil I.D. Annulus i
20 2
2.550
.90 Natural 150 Mil I.D. Annulus 20 3
9.072
.90 Natural 150 Mil I.D. Annulus 12 4
5.500
.90 Natural 150 Mil I.D. Annulus 8
- . t 5
2.551
.80 Na tural 150 Mil I.D. Annulus 3.47.
1.0 w/o Gd 0
=7 8
'. ' Gd 35gm/ftC$b!1t 4
Co Saturated wo tr rod 1
W l-Fuel Dencity: 927. Theoretical L
w
k 11.
Figure 2 BIG ROCK POINI - PLUIONILE BUND 12 DESIGN Hot Operating.(25% Voids-No Control) fg Power Distribution (Norm to 76 Rods)
- t Top - 0 Days-
, g Bottom - 311 Days @ Rated Power
- I!
I 1
1.231 1.287 1.218 1.186 Co 1.058 1.162 1.102 1.093
.43 1
.816 1.097
.341 j
4
.851
.848
- 1. 13 7
.574 i
.929
.989
.938
.961 1.022
.981,
.920
.980 l
.915
.966 W
t 1
i
~.
i I
I I
~
_., l k
6 "
r 75L ee
_m-
, _ _ _ _ _ _ _ _ + - _-
._~
12 L
l TABIZ I i
l COMPARISON OF PRINCIPAL CAlfUIATED NUCIZAR CHARACTERISTICS l
l OF "E-PU", AND i
I
- i REIAAD "E" AND "E-C^
l di ____________________________________________________________________
i Reactivity (kJ l
"E -Pu"
'L9" "E"
l 68 F 1.160 1.208 1.268 572 F, 0 voids 1.168 1.203 1.280 572 F, 257. voids 1.158 1.183 1.262 Ternperature Coefficient: Ak f g f; per F @ 77 F Start of cycle
%.30x10 M.27x10~0 +0.38x10-4 Void Coefficient: Ak/k ocr unit void within channel Cold (68 F)
-0.050
-0.08
-0.07 Hot (572 F)
-0.084
-0.12
-0.11 Doppler Coefficient ok ff4fg per F Fuel Temp.
Modera tor,
-3
-5
-5 1.3x10 68 F 68 F-0 voids
-1.35x10 1.3x10 5
-5
-5 l
a..
1323 F 572 F-0 voids
-1.05x10 1x10 1x10
-5
-5
-5
- - 1323 F
. 572 F, 257. voids-1.25x10 1.2x10 1.2x10 9 t' T Sm.
L
=
-m--,
+-m
--,.e w
w
.e m
I~g w
p y 7--4 p-w
-r-r
i 13 References
~
1.
" Safety Evaluation by the Division of Reactor Licensing, Docket No.
<g 50-155, Consumers Power Company, Proposed Amendment No. 1".
't;g 2.
IDO-ITR-100, " Transient Irradiation of 1/4 Inch 0.D. Stainless Steel gr Clad Oxide Fuel Rods to 570 cal /g UO ", October,1968.
2 3.
ID0-ITR-101, " Transient Irradiation of 0.466 Inch 0.D. Stainless Steel Clad Oxide Fuel Rods to 300 cal /g UO ", November,1968.
2
'4.
IDO-ITR-102, " Transient Irradiation of 1/4 Inch 0.D. Zircaloy-2 Clad Oxide Fuel Rods to 590 cal /g UO ", November,1968.
2 5.
IDO-ITR-103, " Transient Irradiation of.3125 Inch 0 D. Zircaloy Clad Oxide Fuel Rods to 450 cal /g UO ", January,1969.
2 6.
IDO-ITR-104, Ibe response of UO Fuel Rods to Power Bursts, 9/16 Inch 2
0.D., Pellet and Powder Fuel, Zircaloy Clad", April,1969.
7.
In-1302 (IDO-ITR-106), "The Response of UO Fuel Rods to Power Bursts,
_2 Detailed Tests on 5/16 Inch 0.D., Powder Fuel, Zircaloy Clad Rods",
June, 1969.
8.
In-ITR-107, " Behavior of 5 Inch Long,1/4 Inch 0.D., Zircaloy-2 Clad Oxide Fuel Rods Subjected to High Energy Power Bursts", August, 1969.
9.
R. A. Hein, P. N. Flagella ; "Enthalpy Measurement of L'0 and Tungsten 7
to 3260 K", Annual Meeting of Am. Cer. Soc., April 20-25, 1968.
l
- 10. ANL-7527 "Argonne National laboratory, Reactor Development Program Progress Report, December,1968", January 29, 1969.
11.
D. L. Fischer, et al. ; " Nuclear and Fast Transient Aspects of Plutonium Particle Size in Thermal Recycle Fuel", Transaction, ANS Annual Meeting, June,1969, Volume 12, No.1.
12.
R. L. Johnson, SPERT, Personal Communication to R. W. Friis, General l
Electric, San Jose.
- 13. Change No.17 to Technical Specif
- Jtion for Big Rock Point.
i.
- .f6.
e
~
~
I Ll
~I W
E.
sr EdOM:
DATE OF DOCUMENT DATE RECEIVED NO.:
~
%1 12-22-69 12 M 9 3966 g
gg LTR.
MEMO:
ORT:
CTHER:
X CRIG.a CC:
OTHERi Dr Peter A. Morris 3 signed & 37 con.d ACTION NECESSARY O
CONCURRENCE O
LATE ANSWERED CLA248 F:
NO ACTION NECESSARY O CoMwtNT O
avi POST OFFICE FILE CQDEs N
DESCRIPTION: (Must Be Unclassified) 50-155 (IMPUT)
REG NO:
REFERRED TO DATE l
RECEIVED BY DATE__
Ltr trana the following:
Zimann 12 @ 69 w/9 cys for ACTION DISTRIBUTION:
a ENCLOSURES.
PRJPOSED CHNIGE RJ.4UEJT N3 Regulatory file 4ft--
'd 1 notarized 12-22-69 to Tech Specs to AEC PDR cuth insertion into reactor a fuel Cgl e (2) d2:ign designated "Ji:hI-U3 -Pu) which 000 fR.: P 506 A) -
p will permit irradiation of plubnitan-I*
I E
uranium mixed oxide fuel........
(40 cys ree'd)
M*b1D*
[V l.iOY C. */
0
' t t r-kEMI.R KS:
D. The-naan iVi h i
Boyd
--..+
=6-3p.yispter,rnn Aun R u a_ yn h t t=n DTIE(Laughlin) 33IC(Buchanan) 3966 fod "s.....u.....,....,s... rs o....
u.s. ATouic mtRoy coiannon MAIL CONTROL FORM FORM ^EC-82es
<e-e m p
!t I
=-
25 Sm.
L b
l i