ML20002D113

Forwards Request for Change to Tech Specs of License DPR-6, Permitting Insertion of Fuel Bundle Containing New Zirconium Alloy.New Alloy Shows Improved Corrosion Resistance & Hydrogen Pickup Properties Over Zr-2

ML20002D113
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	02/19/1965
From:	Haueter R, Schmidt W CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Boyd R, Doan R US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8101190464
Download: ML20002D113 (19)

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2 3 1965

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• JACKSON, MICHIG AN fI g

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Pebruary 19, Dr. R. L. Doan, Director Be: Docket 50-155 ('

Division of heactor Licenair4 lileCopg United St.ates Atostic Energy hission Washington, D. C.

80545 Dear Dr. Doan Attention:

Mr. 90Eer 8. Boyd Tran==ttted herewith are three (3) executed and nineteen (19) coctormed copies of a request for a change to the Technical spec-l ifiestions of License DPR-6, Docket No. 50-155, issued to Consumers Power Ocurpeny on May 1,19t%, for the Big Rock *oint Nuclear Plant.

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l This change vill permit insertion into the reactor a l

fuel bundle containing a nov sirconism alloy (Zr + 1.15 v/c Cr)

I which shows significantly imme corrosion resistance and hydro 6en pickup properties over Er-II.

Your early consideration of this request vill be appreciated since the modifiestioca to the therw.1 shield support assembly an pro 6rsaain6 rapidly. We expeet to be starting up again during the early part of March 1965 ami app-oval of this change will permit insertice of this developmental fuel bundle during the prosent shutdown.

Yours very truly,

[

Robert L. Baueter Assistant Electris Production Superintendant - Enclear D

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iREQUEST FOR CHANGE TO THE N

TECHNICAL SPECIFICATIONS OF LICENSE NO. DPR em cuf fo

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-ph' Docket No. 50-155

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For the reasons hereinafter set forth, it is requested that the Technical Specifications of License DPR-6 issued to Consumers t

Power Company on May 1, 1964, for the Big Rock Point Nuclear Plant, be.-

changed as follows:

t I.

Section 5 A.

In Section 5 1 5, delete the entire section and replace it with.

the following:

"5 1 5 General Core Composition The data in this section present general design features of the

' original, research and development fuel and reload fuel that shall' make up the physical composition of the core.

(a) Enrichment of Fuel, approximate weight percent U-235 from 2~ 6 to h.5, inclusive.

(b) General Core Data:

Number of Fuel Bundles in Core 86 Total' Nominal Weight UO in 86 Bundles, Lb 29,000 2

l Moderator to Fuel Volume Ratio 2.65 Equivalent Core-Diameter, Inches 76 (c) Fuel Bundles:

The general dimensions and configuration of the three types of fuel bundles shall be as shown in Figures 5 2, 5 3 and 8.1 of these specifications. Principal design

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features shall be essentially as follows:

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. Fuel Bundles.

'Criginal Reload-Research & Development General.

Geometry, Fbel. Rod Array ~

,12 x 12-11 x 11 11 x ll Rod Pitch,; Inches

0 533 0 577 0 580-u Standard ~ Fuel' Rods per Bundle

' 132.

109 109.

. Special Fuel Rods - per - Bundle -

-12*

12 12 Spacers per, Bundle.

3 5

T-Fuel-Rod Cladding Material ~

304 SS Zr-2 304 SS, Zr-2 L Inconel 600 and/or'Incoloy 800 St'andard Rod Tube Wall, Inches 0.019-0.034 0.010 to.0.030, Inclusive Special. Rod. Tube. Wall', Inches 0.031 0.031 0.010 to 0.030,-

Inclusive Fbel Rods

' Standard Rod. Diameter, Inches' O.388-0.k49 0.425 Special Rod. Diameter, Inches 0 350 0 344 0 320 UO2. Density, Percen+, Theoretical 94 t 1 94' i l~

90 to 95, Inclusive Active Fuel Length, Inches Standard 70 70

~68 to 70, Inclusive Corner 59 Fill Gas Helium Helium Helium

• (4 Special Fuel Rods at Bundle Corners Are Segmented)

(d) Channels:

Number of 304 SS and/or Zr-2 88 Wall Thickness, Inches :

304 SS 0.075 L

Zr-2 0.100

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~Inside Width, Inches :

i-304 SS 6 57 i- '

Zr-2 6.54 fr

3 Length, Inches:

30h SS 79-5/8 5

Zr-2 79-3/h (e). Total Weight Supported by Core ; Support Plate:

86 Puel Bundles @ hh0 Lb/ Bundles, Lb 37,8ho 88 Support-Tube-and-Channel Assemblies

@100Lb/ Assembly,Lb 8,800 86 Orifices @ 10 Lb/ Orifice, Lb 860 2 Channel Plugs G 10 Lb/ Plug, Lb 20 1 Flow Distributor Assembly, Lb 2,500 Total Weight, Lb 50,020 (f) Special Features of Reload Puel :

The principal design features of the reload fuel are tabu-lated below. The design is very similar to the Zircaloy-2 clad development fuel now operating in the reactor. Each bundle utilizes two enrichments as well as reduced size corner rods to minimize local power peaking.

The higher enrichment fuel rods are located in the center of the bundle.

Puel rods are held in position by five wire and spring type spacers located along the length of the bundle.

These spacers serve to minimize deflection and vibration of the fuel rods.

The above design details are shown in Figure 5 3 Puel Outside Rods Inside Rods Large Small Rod OD, Inches 0.449 0.hh9 0 3kh Pellet OD, Inches 0 373 0 373 0.275 Cladding Thickness, Inches

• 0.03h 0.03h 0.031 UO2 Ground UO Ground UOp Ground 2

F el Material Pellets Pellets Pellets Nominal Cladding-UO Oap, Inches 0.008 0.008 0.00T 2

Puel Enrichment, Percent h.2 2.6 2.6 UO Density, Percent 9h 9h 9h 2

Number of Segments per Rod 1

1 1

Active Puel Length, Inches 70 70 70

• (Clad Is P.

standing)

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Fuel-Outside Rods Inside Rods-Large Small

- Weight of UO2 per Rod, Lb

. 2.846~

2.846

-1 547 Water-to' Fuel Ratio 2.68 2.68 2.68 Number of Rods per Bundle 37 72 12:

- Weight of UO2 per Bundle, Lb 328.8 "

I-B.

Add a new section, "5 2 3."

5 2 3. Principal Calculated Nuclear Characteristics of Reload Fuel Until-'a significant number of reload fuel bundles are-inserted into the reactor-core,'the principal calculated nuclear char-acteristics of the 84 '- 86. bundle core vill be essentially the same as those noted in Section'5 2.2.

For reference' purposes, the nuclear characteristics of the reload fuel bundles ara presented below:

(a) Reactivity (kco )

b Temperature 68 F,:Zr_ Channels 1.275 572 F, Zr Channels

].303 572 F, Zr Channels +

20% Void 1.296 (b) Moderator Temperature Coefficient (d kerr/k r; per F) e 77 F

-3 Start of Cycle

-0 72 x 10

-3 End of Cycle

-0.14 x 10 (c) VoidCoefficient(dk

/keff per Unit Void Within the Channel) eff 68 F 572 F

-0.08

-0.09 (d) Doppler Coefficient The Doppler Coefficient is dependent on moderator to fuel volume ratio.

Since the reload fuel and original fuel have essentially the same moderator to fuel volume ratio, the Doppler Coefficient for the reload fuel vill be the same as noted in Section 5 2.2 -

(b)."

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. DISCUSSION 1

Theabovelproposedchan6evillenableConsumers* Power

. Company =to : insert two reload: fuel' bundles into-the reactor core after

'the current shutdown for thermal: shield modifications..One bundle vill

. contain fuel rods clad withla new zirconium alloy. The other reload-bundle vill be used to balance the power distribution of the core.

.. HA,MRDS CONSIDERATICNS -

' The reload bundles : described above' and shown in Figure

~; 5 3 vill utilize -the: same -hardware as the present Phase I and Phase II:

' developmental bundles',,except'for the spacers. Eight Phase I bundles

utilizing this hardware have been in the reactor since April 1963, and

.have

exposuresofapproximately4000Mvd/T. Fifteen Phase II bundles utilizing this same hardware have been in the reactor since May-1964, and have exposures of up'to 1600 Mvd/T. Experience with this hardware Jhas been excellent. The decision to utilize it for planned reload _s bundles is a result of this excellent performance.

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The spacers shown in Figure 5 3 are of the wire and

. spring type design currently. considered by General Electric for their commercial: fuel. The springs provide the necessary lateral pressure

to minimize fretting wear of the zircaloy cladding.

The dual: enrichment utilized in these bundles is ex-pected to provide a si nificant improvement in cote peaking factors 6

at a slight enrichment penalty. The effect upon other nuclear and thermal hydraulic factors is not significant.

Based upon the above considerations, we have concluded-that the requested changes to the Technical Specifications do not present significant hazards considerations not-described or implicit

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in the Hazards-Summary Report.

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Section 8 Add a'new section, "8.1 3."

-"8.1 3 Special Developmental Puel Design Features Initially, two relc d type fuel bundles will be inserted into the reactor to evaluate an improved zirconium alloy. This alley (Zr-Cr) contains 1.15 weight percent chromium in reactor grade

. zirconium. Only one bundle (designated B-1) will contain fuel rods clad with the Zr-Cr alloy. However a second bundle similar in design, but containing fuel rods clad only with Zircaloy-2, will be used to balance the power distribution within the core.

Other than for the differences in fuel rod cladding, these two bundles are identical in mechanical design.

The drawing for these two bundles is shown in Figure 5 3 The principal design differences of these two bundles are tabu-lated below.

Each bundle will contain fuel rods of two different rod diameters and UO2 of two different enrichments.

The location of the various types of fuel rods within the bundles is shown in Figure 5 3 Figure 8.3 shows the location of the special Zr-Cr rods in the bundle. A total of 18 rods (2 dummy rods and 16 fuel rods) will be clad with the Sr-Cr alloy and 8 rods (2 dummy and 6 fuel rods) will be clad with Zircaloy-2 which was produced in a manner identical to that used for producing the Zr-Cr.

The dummy rods will be regular size tubing, but void of fuel. Precut tensile specimeas will be strapped to a tantalum rod inside of these rods.

The tantalum rod is required to prevent flux peaking at the loca-tion of the dummy rods. The bundles have been designed in such a manner that all of the fuel rods can be removed.

Puel Bundles B-1 Regular Reload Puel Array 11 x 11 11 x 11 Spacers per Bundle 5

5 Number of Puel Rods 117 121 Number of Dummy Rods 4

0 Number of Rods Clad With Annealed Zr-Cr Alloy 18 O

Annealed Special Zr-2 8

0

_ Cold Worked Standard Zr-2 95 121 Weight of UO per Bundle, Lb 317 5 328.8 2

FIGURE 8.3 LOCATION OF EXPERlHENTAL ZR-CR AND Z9-2 CLAD FUEL RODS IN B-l e 'OOO@OOOM o

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The ' nuclear characteristics' of Bundle B-1, ~ because.it contains

,the 41 dummy. rods,.willibe slight 1p.different than the regular

(. y bundle..The ' presence of the h dummy rods results in the fol-

. lowing changes ~(no other parameters.are affected):

1.

.Th'eflocal peaking factor-using zirealoy channels vill be

. reduced from l.228 to l'.22h.

2. Ir i_ vill befreduced.by 2.7 percent (i.e.' 21k x - 100. = 2 7%).

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The heat flux vill be increased by.0 7; percent.

These changes, however, vill:not result in any significant dif-ference in the performance and lifetime of -the two bundles."

. DISCUSSION l

The above proposed change vill enable Consumers Power I.

Company.and General' Electric Company to-obtain irradiation experience on

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a new~and promising zirconium alloy developed for use as a nuclear fuel.

. cladding. The objective of the proposed test is to confirm under actual

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power reactor operating conditions the lover corrosion and hydriding I-rates of the new alloy as compared to Zirealoy-2. 'This new alloy will r

be in a' fuel' bundle which will be inserted into the reactor core at the end of the current shutdown-for. thermal shield modification.

L

.A.

Background Inforrration An improved zirconium alloy has been developed under the Specific Zirconium Alloy Design Program, AEC Contract AT(Oh-3)-189, P.A.'24.

This alloy is the result of an intensive AEC/ Euratom-sponsored program conducted at the General Electric Atomic Power Equipment Department (GE-APED). - The ' proposed alloy contains 1.15 a/o Cr and has been found i

to exhibit improved corrosion resistance and smaller post-exposure-l.

hydrogen content by-comparison to Zircaloy-2. The mechanical properties l

of the Zr-Cr alloy are.quite similar to those of Zircaloy-2.. Based upon the development work performed.to date, this alloy offers future promise cf utilizing zirconium cladding at higher surface temperature

- (i.e.,' higher heat flux and higher power density).

B.

Description of Proposed Program It.is proposed under this program to design and fabricate 26 fuel rods.

v Of the 26 fuel rods,16 vould be clad with the new alloy, 6 would be.

clad with Zircaloy-2, and the remaining h would be 2 unfueled tubes of

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.the'new alloy e,nd 2 unfueled tubes of Zircalo'y-2..The nuhber of-fuel

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rods of-each type.vas. selected to' determine in-reactor rate's of cor-rosion.of Lhy'drogen uptake under both heat. transfer and 'nonheat trans-

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. for conditions (fueled "and unfueled : tubes).

It also will yield a

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ramparison.of the performance of the new. alloy with respect to Zircaloy-2. LThe 26 fuel 'rodsLvill be incorporated in a regular reload fuel bundle fabricated for_ the Big Rock Point Nuclear Plant. ~This-fuel bundle will 'be :1rradiated' at Big' Rock Point;at a location and

at conditions which will not interfere with the present program. At appropriate, scheduled shutdowns of the plant, rods will'be. removed (an'd replaced) from the fuel bundle _and shipped to Vallecitos -for

post-irradiation examination. Rod removal and replacement-are ten-tatively scheduled at exposures of _ about - 5,000 Mwd /T-and 10,000 Mwd /T.

C. ' Technical Information-

'An extensive summary of the Specific Zirconium Alloy Design Program--

is available as progress reports to the USAEC.

(See GEAP-4368 and GEAP-kh8h ) Some of the pertinent information from this program is summarized below:

1.

Full-scale vendor fabrication of tubing and rod material for the Big Rock Point fuel rods was accomplished without incident.

2.

The fabricability and quality of the material vere found to be essentially identical to that of Zircaloy-2 material processed side by side with the Zr-Cr from ingot to final fuel rod. The material of both compositions met ASTM,' GE-APED and special test specification for quality.

3.

Tests on the Zr-Cr alloy established microstructure,. tensile properties, and nuclear properties for the mill products.

4 '.

Corrosion tests verified that the mill product shows lower cor-rosion rates thar. Zr-2, particularly under conditions _ of over-

-temperature, 5

This material compared favorably with laboratory pilot material in check tests at 932 F.

6.

The Zr-Cr alloy was. tested in water at 550 F and 680 F, as was

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I the case in the early. laboratory tests in 572 F steam.

The per-formance of the Zr-Cr: alloy in water was excellent.

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.9' e7. Welding tests revealed no problems as the alloy. could lie welded -

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~ exactly like-Zr-2. The-welds appear more corrosion resistant than the base material for both alloys.

78. ' Assembly.of-the Big. Rock Point Zr-Cr fuel rods was performed

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exactly ~1ike assembly of the Zr-2 rods. All rods passed quality Lcontrol tests including helium leak r,ests and x-ray radiography.

1. -Fabrication of Tubing and Rod

A 1000 lb ingot ~of Zr '+ 1.15 Cr vas melted and processed by the material supplier at'fu11' scale to provide tubing and~bar_ stock-(for end plugs). A Zr-2 ingot was processed right.'along with

.the Zr-Cr.

No ' differences in behavior were observed in fabric-

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ability. Both alloys were solution treated and quenched in.the extrusion billet stage to_ ensure homogeneity. Two methods of cold reduction, drawing and rocking-to-size, were used on one half of'each alloy. The final annealing temperature'of 1400 F was selected, based on laboratory tests, to. optimize Zr-Cr cor-

-rosion resistance and to give a uniform fine grain size for both alloys.

All dimensional specifications were verified. The certification test data for the ASTM corrosion test are given in Table I.-

TABLE I ASTM Three-Day 750 F Steam Corrosion Tests Zr-2 Tubing 16.2, 17.8, 17 1, 15 5, 16.0, 16.2, 14 9, 16.2 Bar 18.6, 18.6, 17.6, 17 9 Average: 17

• 1 mg/dm Zr-Cr Tubing 13 4, 14.2, 14.6, 16.3, 17 3, 13.4, 16 3, 12.2 Bar 13 2 12.6, 12 9, 15 8 2

Average: 14 2 mg/dm ACCEPTABLE LIMIT: 22mg/dm 2.

Inspection at GE-APED-i Special reinspection of all of the material was performed at l-GE-APED. = Visual examination, ultrasonic tests, and dimensional checks rejected a few individual tubes'of both alloys as less than perfect. No serious lack of~ quality was observed.

Chemical l

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check analyses were made for each lot of the Zr-Cr material. The results were within specification

Material Supplier

• GE-APED *

~ Specification

.Cr 1.18-1.29%

1.11-1.23%

1.00-1 30%-

Fe.

275-300 Ppm 50-800 Ppm 1000 Max-N1.

< 20 X 40 70 Max

'02 1010-1150 794-1251 900-1500 N

35-36 14 45 60 Max 2

H2

9-11 11-1k 25 Max Hf.

_62-70

< 100 100 Max B

< 0. h

< 0.2 0 5 Max All minor impurities analyzed well below limits.

• At least three analyses in.this range.

3 Tensile Properties Tensile' tests on both bar stock and tubing of Zr-Cr and Zr-2 were performed at several test temperatures. The room tempera-ture results are compared to.those for Zr-2, Table II.

TABLE II Room Temperature - Tensile Test Data Tubing Zr-Cr Values Material Zr-Cr Zr-2 Supplier GE-APED Average Average 0.2% Yield.

(54.5)(h7 5)

(59)(h6.9) 50 1 3 57

• 5 (h6.5)(h6.5)

(48.0)(h7.h)

Urs (Th.2)(72.8)

(80.1)(75 5) 75 !2

-76 1 3 (72.1)(73 8)

(76.3)(75 9)-

%l Elongation (26.5)(30 5)

(28.8)(28.6)

'30 1 3 27

• 3 (335)(330)

(29 2)(33 8)

Bar Stock O.2% Yield (h3 0)(h3 2)

(h9 3)(48.h) h5 t 3 50

• 2 (39 5)(hh.6)

UTS (Th.1)(Th.3)

(75 5)(75 9)

Th ! 1 75 ! 1 (736)(731)

% Elongation (25 0)(25 0)

(23 5)(23 5) 25 t 1 25 *2 (25 0)(25 0)

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11 h.

Microstructure

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Metallography was performed on both bar stock and tubing samples of Zr + 1.15 v/o Cr.

Both the Zr-2 and Zr-Cr material had the desired fine grain'cize (ASTM 8 or finer). The desired, fine, uniform distribution of intermetallies with no stringers was revealed. As ' expected from earlier studies ar.d from the ' pertinent phase diagrams, the Zr-Cr microstructure contains more inter-metallic second phase.

(Iron, chromium, and nickel have very limited solid solubility in zirconium, <0.01 v/o. Zr-2 contains a total of 0 30 v/o Fe + Cr + Ni.

The Zr-Cr alloy contains 1.15 v/o Cr.) Probably-because of more second phase, special grain growth studies showed the Zr-Cr alloy more resistant to grain growth. No growth was observed in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> up to lh50 F for Zr-Cr.

Slow grain growth is observed at ih00 F for Zr-2.

5 Nuclear Properties The microscopic neutron cross section for Zr + 1.15 Cr is calcu-lated to be 0.256 barns as compared to 0.210 barns for Zircaloy-2.

The macroscopic section for Zr + 1.15 v/o Cr is calculated to be 0.0105 versus 0.008h to 0.0091 for the Zr-2 composition range.

(The stainless steel value is 0.260. )

Relative reactivity checks on samples of the Big Rock Point Zr-2 and Zr-Cr tubing were made in the GE-NTE reactor. The ratios of reactivity, expressed as equivalent ppm of boron, confirm the ratios of the calculations of cross section. The values were 41.0 ppm boron equivalent for Zr-Cr and 36.4 ppm boron equivalent for Zr-2.

The Zr-2 value is in agreement with all previous samples for General Electric Zr-2 fuel cladding. These test results also en-sure that the material contains no spurious impurities of high capture cross section.

6.

Additional Corrosion Testing Special 175-hour cerrosion test runs in refreshed flowing steam are being run at 573 F and 750 F.

Runs at 932 F are complete. Actual samples from the early laboratory pilot melts were run along with the material supplier's material. The results are given in Table III.

The material supplier's material-performed as expected from the laboratory tests.

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TABLE' III

Corresion. Data 932 F Weight Gain'(eg/dm ) After Alloy'-

60 Hours 75 Hours 135 Hours 175 Hours 235 Hours.

LZr-2 Material' Supplier's Tubing-..

5h

106 139

'* Lab Melt Sheet 68 129 Zr-Cr-Material Supplier's Tubing

.37 60

- 84

• Lab Melt Sheet 53 80 Average Values Tabulated Averag6 Devintion i 10%'

• These values for these times agree with data taken during the 7000 hour0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> =

. run.on laboratory material. The long-term corrosion rates previously reported were determined from the 7000 hour0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> run.

All early. laboratory material was tested in steam. Additional

. tests have now been performed in the WR water loop. This ex-reactor loop simulates reactor water chemistry, velocity, and ~

s temperature. The results are given in Table IV.

Some test samples were also run at 680 F in water. These results are-also given in Table IV.

TABLE IV Corrosion Tests ~1n Water 550 F - WR Icop Water 2

Weight Gain (mg/dm ) After Alloy 1000 Hours 2750 Hours

~3870 Hours Zr-2 (12)*-

(18)*

(22)*

Zr-Cr**

11 16 16

.572 F - Static Autoclave Water 51 Hr 1080 Hr 1750 Hr 2860 Hr 3750 Hr Zr-2-(Commercial) 9 ~

18 20 22 2h

-Zr-Cr**

6 lh 16 20 22-U 680 F.- NR Loop Water 75 Hours

Zr-2 (10)*

I

'Zr-Cr**

9 Average: Values Tabulatedi

? Average -Deviation i 10%

• Lustman & Kerze,'Page 633

• 4 Laboratory Melts

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13 The tubing order from the material supplier also incl'uded some larger diameter tubes (0 502 ID) capable of test in the BWR cor-rosion loop as heat transfer surfaces (heat flux 250,000 Bta/Hr/

Ft ).

These heat transfer corrosion tests are in progress.

Nothing of significance occurred to shut down the test during the first 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />.

The test is scheduled to run 7800 hours0.0903 days <br />2.167 hours <br />0.0129 weeks <br />0.00297 months <br />.

As part of the procedure for fuel rod fabrication, all cladding is given a lk-hour 750 F steam treatment. The Zr-Cr fuel tubing received this treatment as did the Zr-2 tubing. The Zr-2 tubing met the visual standard after testing. A uniform black ZrO III" 2

typical of the expected weight 6ain of about 12 mg/dm was ob-served. For the Zr-Cr alloy, a weight gain of only about 6 mg/dm is expected. At this low weight gain, ZrO2 films on any zirconium alloy are still in the thin interference color thickness range.

The Zr-Cr fuel tubing showed these expected interference colors typical of thin films of about the weight gain expected for IL hours at 750 F.

7 Welding In preparation for fuel rod assembly, welding development tests were performed.

Samples of material, identified only as "zire,"

were submitted to three separate velders in three different GE-APED welding shops. Welding parameters including veld box pump down, electrode, spacing, power, and speed of travel were requested. The range of parameters actually used was very wide and bracketed the normal Zr-2 fuel rod welding parameters.

Two tensile specimens cut from the welded material from each of the three shops were broken at room temperature. All but one broke with good ductility (28-61% R. A. ) outside the veld. The one sample which broke in the less ductile veld had been velded at very high power so that the veld zone was as wide as the gauge section of the sheet tensile specimen.

Samples were corrosion tested. Weight gains for the total samples were acceptable. The corrosion films in the weld region looked blacker and more uniform than those on the base metal. This is always observed for Zr-2 also.

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Meta 11ography of the welds' revealed that the oxide fiIm over the 4

weld was indeed thinner than over the base metal. The' micro-

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structure was typical of zirconium alloy welds.

No crack, voids,.

or microporosity was evident.

Nitrogen analyses of samples cut from the veld regions gave 26 I.

5 ppm, 31 i 8 ppm and 25 t 6 ppm,respectively,for the three welder sources. - These. values are well within the specification of less

-than 60 ppm..

'r The over-all' finding was that very acceptable 5 elds could be made under conditions even less stringent than those for normal Zr-2 fuel' rod-to-end cap welds. The tests were on sheet butt welds.

The fuel rod end cap joint is designed to give a much easier weld. Production veldc are fully controlled and automatic. For these reasons, the fuel rod welds were made exactly the same for

~ Zr-2.and Zr-Cr rods.

l All production velds were given helium leak tests and x-ray radiography. The Zr-Cr rods and the Zr-2 rods passed both of these tests without any indication of flaws.

HAZARDS CONSIDERATIONS

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All tests performed to date on the Zr-Cr alloy have shown that, perfonnance vise, this alloy is at least as good as Zr-2 except for corrosion resistance and hydrogen pickup properties where the Zr-Cr alloy is expected to be significantly better. The use of four dummy rods in this bundle'will.not have any significant effect upon its performance or ll ^

lifetime. A hazards analysis o

'he possible effects of the Zr-Cr alloy rods leads us to believe that the addition of this bundle to the reactor core does not present any significant hazards not described or implicit l

in the Hazards Summary Report.

CONSUMERS POWER COMPANY lb ffnsFY Vice President-l Date: February 19, 1965 Sworn and subscribed to before me this 19th day of February 1965 h e --M. Na -ad Notary Public, Jackson County, Michigan My commission expires February 16, 1968 i

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