ML20037A926

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Safety Evaluation Supporting Application for Revision of License DPR-2 Re Cycle 7 Fuel
ML20037A926
Person / Time
Site: Dresden 
Issue date: 05/14/1969
From:
COMMONWEALTH EDISON CO.
To:
References
NUDOCS 8008130145
Download: ML20037A926 (52)
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EXHIBIT I DRESDEN NUCLEAR POWER STATION DESCRIPTION AND SAFMTY E7ALUATION REPORT OF CYCLE 7 FUEL This report provides technicalinformation in support of the attached application for revision of Isresden Operating License DPR-2, as amended. It is not in-tended that the material contained berein constitutes C

" Technical Specifications" in the sense of the Licens-ing Regulations (10 CFR, Part 50, Section 50.36).

May 14, 1969 i

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Regulatey File C).

TABLE OF CONTENTS 1.

DESCRIPTION OF PROPOSED AMENDMENT TO APPENDIX "A" OF DPR-2 1

2.

PHYSICAL CHARACTERISTICS OF THE REIDAD FUEL............

3 3.

NUCLEAR CHARACTERISTICS OF RELOAD FUEL AND C O RE.................

13 3.1 Power Distribution 13

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3.2 Temperature and Void Coefficients.....,...

26 3.3 Reactiv'.ty Control and Shutdown Margin...

26 3.4 Burnable Poison Behavior............'.

30 3.5 Nuclear Calculational Methods for UO and 2

PuO -UO Fuel..................

31 2

2 4

3.5.1 Fuel Element Nuclear Calculations 31 3.L.2 Three-Dimensional Power and Burnup Distributions................

32 e

3.5.3 Moderator Temperature Coefficients 23 4.

THERMAL AND HYDRAULIC CHARACTERISTICS 35 5.

ADDITIONAL EXPERIENCE WITH GADOLINIA-URANIA.

.?......................

37 5.1 Dresden Operating Experience with Gadolinia Poison......

37 5.2 Big Bock Point Operating Experience........

37 5.3 UNC PLATR Facility Experiments 38 6.

SAFETY CONSIDERATIONS 39 6.1 Potential Hazards Associated with the Handling of Unirradiate.1 Type VII-Pu Assemblies........

39 11 i

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6,2 Effect of the Delayed Neutron Fraction of Plutonium Fuel...................

41 6.2.1 -Refueling Accident 41 6,2.2 Core Kinetic Behavior............

42 7.

REFERENCES.....................

46 TABLES 1.

Isotopic Composition of Plutonium for Dresden Plutonium Recycle Assemblies..............

4 2.

Dresden I Fuel Element Data for Cycle 7.........

11 3.

Assembly Average Lattice Data -Unirradiated Types I, III, III-F, V, VI, VII, VII-Pu...........

14 4.

Lccal Peaking Factors for Dresden Types VII (VI),

Type VII-Pu, Type VII-Pu-L, and Type VII-Pu-I Fuel Assemblies....................

21 5.

Predicted Limiting Void Reactivity Coefficient during Cycle 7.....................

27-Calculated # eft or Dresden Cycles 6 and 7 45 6.

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FIGURES-1.

Fuel Rod Arrangt. ment for Fuel Type VH-Pu Plutonium Reload Assembly '...............

5 2.

Fuel Rod Arrangement for Fuel Type VII-Pu-L Plutonium Reload Assembly...............

6 3.

Fuel Rod Arrangement for Type VII-Pu-I Plutonium Reload Instrumenied Assembly..............

7 4.

Assembly Drawing -Type VII-Pu 8

5.

Cycle 7 Preliminary Locations for Assemblies SA-1, PF-10, A-465; Plutonium-Bearing Assemblies; and o

Type VII-I Assemblies 10 6.

Preliminary Dresden Nuclear Power Station Cycle-7 Lo ad ing........................

19 7.

Relative Power in Dresden Plutonium Recycle Assem-bly (Type VII-Pu) Hot, 20 v/o Steam, Beginning of Life...

22 8.

Relative Power in Dresden Plutonium Recycle Instru-4 mented Assembly (Type VII-Pu I) Hot, 20 v/o Steam, Beginning of Life....................

23 111

9.

Relative Power Distribution in a Dresden Plutonium Recycle Fuel Assembly (Type VII-Pu) Rotated 180*

from Normal Orientation................

24

10. Relative Power Distribution in a Dresden Plutonium Recycle Fuel Assembly (Type VII-Pu) Rotated 90*

from Normal Orientation................

25

11. Dresden I-Calculated Temperature Coefficient is Temp:rature 28
12. Dresden I Temperature Coefficient vs Exposure at 68*F..

29

13. Calculated Effective Delayed Neutron Fractions (Seff) for Type VII and Type VII-Pu Assemblies 43 9

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1. DESCRIPTION OF PROPOSED AMENDMENT

'IO APPENDDC "A" OF DPR 2 The purpose of this amendment is to request authority to insert.

eleven (11) plutonium-bearing assemblies (Type VII-Pu) and 85 Type VII fuel, equivalent to Type VI fuel

  • into the Dresden I reactor at the beginning of Cycle 7.

The use of 11 plutonium-bearing assemblies is part of a joint pro-gram between Commonwealth Edison and United Nuclear designed to evalu-(

ate the use of plutonium as a fuelin boiling water reactors. The physical, nuc1 car, and thermal-hydraulic characteristics of the plutonium assem-blies are discussed in Sections 2 through 4 of this Exhibit. Section 5 discusses additional expeSence with gadolinia-UO. The safety considera-2 tions associated with the utilization of these plutonium assemblies are discussed in Section 6.

--Extensive experience has been attained in recent years in the handling and use of plutonium as a fuel for thermal reactors. A Plutonium Utilization Program has been in progress at the Hanford Laboratories since 1956 to demonstrate plutonium recycle. Extensive handling and f tradiation experience has been obtained under this program from operations of the PRTR (Plutonium Recycle Test Reactori and the PRCF (Plutonium Recycle Critical Facility). A plutonium recycle zone was utilized in the EBWR (Experimental Boiling Water Reactor) at Argonne National Laboratory to demonstrate the utilization of plutonium in a hiling water power re-actor. Also, the Saxton Reactor has operated with a p'atonium recycle core (Core U), and performance was within expected operating limits.

  • Type VII and VII-I fuel assemblies are identical to Type VI and

-Type VI-I fuel assemblies except for a few minor mechanical design changes. These changes do not influence the nuclear er thermal-hy-draulic characteristics of the fuel. All descriptions and data pren for Types VI and VI-I fuelin Reference 6 apply respectively to Types VII and VII-I fuel.

n.

t Extensive critical experiments have been performed -s which have established the ability of present analytical methods to calculate plutonium-bearing lattices.5 Plutonium depletion is normally treated as a part of uranium depletion. Approximately 40?o of the total power produced in a Dresden Type VI assembly over its lifetime is generated by plutonium iro-topes bred in the assembly. At end of life, a Type VI Dresden assembly generates approximately 657o of its power from plutonium isotopes.

In UNC's PLATR (Pawling Lattice Test Rig), substitution criticals are being performed on detailed mockups of the Type VII-Pu assemblies to measure power distributions and reactivities as a check against

-ca cu a ons. In addition, tests in the UNC PTF (Proof Test Facility) on l

l ti the finished fuel assemblies will be performed to verify the uniformity of reactivity worth of individual plutonium-bearing assemblies.

Fuel rods fabricated with plutonium isotopes were first introduced into Dresden Unit 1 at the beginning of Cycle 5. At that time, four plu-tonium rods were inserted into previcusly exposed assemblies. The 11 plutonium-bearing assemblies to be inserted into Cycle 7 will contain a total of 99 plutonium-bearing rods. Approximately one-quarter of the fuel

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rods in the plutonium-bearing assemblies will contain mixed plutonium-uranium oxide. The remaining fuel rods wi11'contain slightly enriched uranium c::ide.

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2. PHYSICAL CHARACTERISTICS OF THE RELOAD FUEL It is proposed to reload the reactor with a maximum of 96 fuel as-

-.semblies at the start of Cycle 7. The reload batch will consist of a max-imum of 85 Type VII fuel assemblies and 11 Type VII-Pu fuel assemblies.

The 65 Type VII fuel assemblies are expected to consist of 80 " normal" uranium assemblies (Type VII) and five instrumented or source-bearmg assemblies (Type VII-I). The Type VII and VII-I assemblies are essen-tially identical to Type VI and Type VI-I fuel assemblies previously described.8 The 11 type VII-Pu plutonium fuel-assemblies are expected to consist of three " normal" plutonium assemblies (Type VII-Pu),

'six " low gadolinium plutonium" assemblies (Type VII-Pu-L), and two

" instrumented" plutonium assemblies (Type VII-Pu-I). There is no ax-ial variation in fuel enrichment or poison cancentration in any of the 96 fresh assemblies.

The fuel rod arrangement and loading of the plutonium assemblies

..are shown in Figs.1, 2, and 3 for Type VII-Pu, Type VII-Pu-L, and Type VII-Pu-I, respectively.

An assembly drawing of Type VII-Pu fuelis shown in Fig. 4.

As shown in Fig.1 the plutonium-bearing assembly Type VII-Pu configuration consists of a 6 x 6 matrix of 36 fuel rods. Twenty of the 36 rods are " normal" uranium rods containing uranium enriched to 2.34 w/o U ss. Six of the 36 rods are " power peaking correction" rods 2

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2 U as,w/o Fissile Pu, w/o Total fiss., w/o No. of Rods 1.77 1.77 6

o 2.34 2.34 20' Pu 0.71 1.78 2.49 9

2.34 2.34 1 (55g Gd O )

2 3 Average: 1.84 0.45 2.29 36 Fig.1 - Fuel Rod Arrangement for Fuel Type VII-Pu Plutonium Re-load Assembly 4

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Us23, w/o Fissile Pu,w/o Total fiss., w/o No. of Rods

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2.34 2.34 20 Pu 0.71 1.78 2.49 9

2.34 2.34 1 (27.5g Gd:Of Average: 1.84 0.45 2.29 36 Fig. 2 - Fuel Rod Arrangement for Fuel Type VII-Pu-L Plutonium Reload Assembly 5

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2 U ss, w/o Fissile Pu, w/o Total fiss., w/o No. of Rods 1.77 1.77 6

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Average: 1.82 0.46 2.28 35 Fig. 3 - Fuel Rod Arrangement for Type VII-Pu-I Plutonium Reload Instrumented Assembly 6

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containing 1.77 w/o U235 Nine of the 36 fuel rods contain 2.3 w/o plu-tonium oxide in a matrix of natural uranium oxide. The remaining rod is a " fuel-poison" rod containing 1.74 w/o Gd O in a matrix of uranium 2 3

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oxide enriched to 2.34 w/o U2as. The isotopic composition of the plu--

tonium is given in Table 1 and is the same for all Pu assemblies. The fuel pellets in all of the Pu-U oxide rods have been prepared by mechan-ically blending, then sintering.

Three plutonium-bearing rods in these assemblies are removable.

A spring-loaded lock mechanism is provided to allow the lifting handle to be rotated to enable access to the three removable rod positions.

The fuel Type VII-Pu-L (Fig.2) is identical to the fuel Type VII-Pu (Fig.1) except for a lower gadolinium content in the " fuel-poison" rod.

In the Type VII-Pu-L assembly, the " fuel-poison" rod contains 0.87 w/o Gd 023 in a matrix of uranium oxide enriched to 2.34 w/o U235 Three Type VII-Pu assemblies will be placed in the interior of the core, while six Ty?e VII-Pu-L assemblies will be placed toward the periphery of the core, as shown in Fig. 5.

The Type VII-Pu-I (Fig.3) assembly consists of a 6-X 6 matrix of rods, 35 of which are fueled.

Twenty of the rods are "

uranium rods containing uranium enriched to 2.34 w/o Uza5, normal" six of the rods are " power peaking correction" rods containing 1,77 w/o U235 The remaining nine fueled rods contain 2.3 w/o plu-tonium oxide in a matrix of natural uranium oxide.

The remaining rod location contains an instrument guide thimble.

Two Type VIl-Pu-I assemblies (Fig. 3) will be utilized in instru-mented locations in the core, as shown in Fig. 5.

Relevant dimensions for rods in the plutonium-bearing assemblics are shown in Table 2.

8

. \\.L TABLE 1 -ISOTOPIC COMPOSITION OF PLUTONIUM FOR DRESDEN PLUTONIUM RECYCLE ASSEMBLIES *

(Types VII-Pu; VII-Pu-L; and VII-Pu-I)

Composition of Plutonium, a/o 238 Pu 0.4 tu Pu 71.3 Pu "

20.6 2

241 Pu 6.1 242 Pu 1.6 Percent fissile = 77.4 a/o C

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  • Per analysis forwarded with Form AEC-388 SS Material Transfer i

Form dated June 28, 1968 signed by V. D, Donihee. Transfer from HVA to SNM-871 No. I.

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3. NUCLEAR CHARACTERISTICS OF RELOAD FUEL AND CORE The nuclear characteristics of the Type VII and Type VII-I fuel are identical to Type VI and Type VI-I fuel.

The basic physics lattice data for Type VII and Type VII-I fuel _are summarized in Table 3 with those types of fuel which are scheduled for use during Cycle-7.

l l

Nuclear characteristics of the plutonium-bearing fuel assemblies -

l (Type VII-Pu, Type VII-Pu-L, and Type VII-Pu-I) are also shown in l

Table 3. The nuclear characteristics of the plutonium-bearing fuel as-i (

semblies are very similar to the Type VI and Type VII uranium fuel.

The cold, beginning-of-life, infinite multiplication factors for the Type VII-Pu and Type VII-Pu-L assemblies with and without control-rods are close to those for Type VI fuel.

The hot, operating reactivity levels are lower for the plutonium assemblies due to a larger negative void coefficient.

A representative core loading is shown in Fig. 6. Location of plutonium assemblies and other special assemblies are shown in Fig. 5.

Deviations from this pattern may be necessary at the time of refueling.

The loading will be within the framework of the "1-in-4" scatter load pattern employed in previous Dresden fuel cycles.

The nuclear calculation methods used are described a.id discussed at the end of this Section.

3.1 POWER DISTRIBUTION The overall gross core power peak for Cycle 7 is calculated to be approximately the same magnitude as it was in Cycle 6.

cor Cycle 7, the predicted overall power peak (radial X axial X local) at any time during the cycle will normally be less than 3.0 and will not exceed 3.24, wPich is the value established from the license r quirement, so that the maxinum heat flux is 360,000 Btu /hr-ft at 700 'tw(t).

In addition, the predicted SICIIFR at 1251 power at any time during the cycle will be >1.5.

Thermal-hydraulic characteristics are given in Section 4.

13

p i

4 l

TABLE 3 - ASSEMBLY AVERAGE LATTICE DATA -UNIRRADIATED TYPES I, III, III-F, V, VI, VII, VII-Pu i

Fuel Type I Fuel Type III Fuel Type III-F j

-Temperature, F 68 546 545 68 546 546 68 546 546 Void Fraction (within flow channel) 0 0

0.20 0

0 0.20 0

0 0.20 h

j k=, Uncontroned 1.132 1.151, 1.145 1.154 1.160 1.150 1.210 1.201 1.197-km, Controlled 0.946 0.884 0.852 0.970 0.904 0.873 1.009 0.927 0.902 Control Worth, i

ak./k.

0.164 0.232 0.256 0.159 0.221 0.241 0,166 0.232 0.246 2

2 l

M. cm 09.8

63. b 75.4 38.8 61.9 74.1 39.0 61.7 77.0 Burnable Poison 3

0.055 Worth, akm/k.

0.048 i

)

4 L

e 0

4 i

i 1

i t

i j

TABLE 3 -(CONTINUED) j Fuel Type V, Group 5*

Fuel Type V, Group it Fuel Types VI, VIIt Temperature, *F 68 546 546 546 546 68 546 546 Void Fraction I

(within flow channel) 0 0

0.25 0

0 0.25 0

0 0.20 koo, Uncontrolled 1.125 1.093 1.063 1.191 1.172 1.154 1.210 1.203 1.201, koo, Controlled 0.967 0 905 0.845 1.024 0.974 0.917 1.009 0.924 0.906 l

Control Worth,

{

Ak /koo 0.140 0.172 0.206 0.140 0.169 0.208 0.166 0.23 2 0.246 l

M, em 38.6 60.8 72.0 40.6 61.8 76.0 39.1 60.8 71 2

2 i

-Burnable Poison Worth, Akoo/koo 0.131 0.080 0.055 l

  • Type V, Group-5 defines the reload fuel with the greatest Gd:O loading.

3 l

TType V, Group-1 defines the reload fuel with the least Gd O loading.

2 3 l

31ultigroup k. definition used for Types VI and VII.

5 4

l

p TABLE 3 -(CONTINUED)

Fuel Fuel Types VI, VII*

Fuel Types VI-I, VII-It

. Types VI-I, VII-I*

Temperature, *F 546 68 546 546 546 Void Fraction (within flow channel) 0.20 0

0 0.20 0.20 k, Uncontrolled 1.220 1.277 1.301 1.288 1.314 k, Controlled 0.906 1.059 0.999 0.970 0.970 Control Worth, Ak /k.

0.257 0.171 0.23 2 0.246 0.262 M *, c m 71.6 39.9 61.0 72.4 72.4 2

Burnable Poison Worth, Ak./k.

58

  • TRILUX 8 k (four-factor) definition used for Types VI, VII, VI-I and VII-I.

tMultigroup k definition used for Types VI-I and VII-I.

5

O.

TABLE 3 -(CONTINUED)

Fuel Fuel Type VII-Pu*

Type VII-Put Fuel Type VII-Pu-L*

Temperature, 'F 68 546 546 546 68 546 546 (hot, no (hot, (hot, (hot, no (hot, power) power) power) power) power)

Void Fraction (within flow channel) 0 0

20 20 0

0 20 k, Uncontrolled 1.215 1.206 1.186 1.204 1.?18 1.210 1.189 k., Controlled 1.017 0.936 0.896 0.896 1.020 0.940 0.898

~

Control. Worth, Ak./km 0.163 0.223 0.245 0.256 0.163 0.223 0.245 2

2 M, cm 38.2 59.2 69.8 69.8 38.2 59.2 69.8 Burnable Poison Worth, Ak./km 0.043 0.041

  • Multigroup k. definition.

5 tTRILUX >8 k (four-factor) definition.

U O

e

=

p t./

TABLE 3 -(CONTINUED)

Fuel Fuel Type VII-Pu-L*

Fuel Type VII-Pu-It Type VII-Pu-I*

Temperature, F 546 68 546 546 546 j

(hot, (hot, no (hot, (hot, power) power) power) power) i Void Fraction (within flow channel) 20 0

0 20 20 I

km, Uncontrolled 1.207 1.269 1.262 1.233 1.245 k., Controlled 0.898 1.054 0.965 0.920 0.920 Control Worth, Akm/k.

0.256 0.169 0.23 5 0.254 0.264 2

2 l

M, cm 69.8 38.8 59.3 69.7 69.7 Burnable Poison Worth, Ake/km l

  • TRILUX,8 km (four-factor) definition.

8 tMultigroup km definition.

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Fig. G - Preliminary Dreniten Nuclear l'uwer Stat 6on Cycle 71.tmlin;:

a e

The plutonium-bearing fuel assemblics have uranium rods in all peripheral positions of the assembly to prevent power peaking at in-terfaces between adjacent plutonium and uranium assemblies. The local power distributions within the VII-Pu plutonium-bearing assem-bly and the VII-Pu-I instrumented _ plutonium-bearing assembly are shown in Figs. 7 and 8, respectively.

The power peaking within-VII-Pu-T.

b plutonium bearing assemblies'is the same as those shown in Fig.7.1 The local power peak in the plutonium-bearing assembly, Type VII-Pu, occurs at the beginning of the cycle and is calculated to be -

1.28.- This is not in excess of the local peaking factor for Type VI fuel (Table 4). The local peak in Type VII-Pu-I is lower than in Type VII-Pu

' due to the absence of the gadolinia-containing fuel rod. :With burnup and '

depletion of gadolinium in the Type VII-Pu, the local peaks will decrease from their initial values.

The Type VII uranium assemblies are essentially identical to the Type VI assemblies and will have the same power distributions as the Type VI assemblies.A A design feature is provided to assure proper orientation of Type VII and Type VII-Pu fuel assemblies. This design feature con-sists of a projection on one side of the lifting handle which points toward the control rod. In addition, the fuel assembly number, which is en-graved on top of the lifting handle, is read from the control rod side.

Refueling procedures and verification techniques assure that the as-semblies are inserted properly.

In the unlikely event that a Type VII-Pu assembly is disoriented,-

the maximum local power peak would occur if the assembly were loaded in a manner so that it is rotated 180 degrees from the normal'orienta-tion. As shown in Fig. 9, the beginning-of-life power peak is 1.32 for this improper orientation compared with-1.28 for a properly oriented-assembly, an increase of only 3% If the assembly were loaded im-properly' with a 90 rotation from normal, the maximum local power peak would be 1.30 (Fig.10).

Disorientation of a VII-Pu-I or VII-Pu-L assembly would'have tht-same or less effect than the disorientation of a VII-Pu~ assembly-because VII-Pu-I and VII-Pu-L assemblies in nofmal orientation'have weak to average power densities which are eaual to or less than the.

VII-Pu assemblies in normal orientation.

~

20

C:ntrol rod position j

f

~-

~

1.04 1.15 1.09 1.14 1.02 A

(1.04 1.03 0.89

( 1.28 }

0.89 1.05

~

V.

1.15 0.89 0.74

(~ 1.01 (1.11)-

0.92 V

A A

{O

~

1.09

( 1.28 }

( 1.01) 0.24i 1.02:

0.88 V

V

\\d A

A A

1.14 0.89

(

)

(

)

1.05 0.92 0.88 0.93 86 w

-(-

2 U ss, w/o Fissile Pu, w/o Total fiss., w/o No. of Rods 1.77 1.77 6

).2.34 2.34 20 0.71 1.78 2.49 9

2.34 2.34 1 (55g Gd O )

2 3 Average: 1.84 0.45 2.29 36 Fig. 7 -Relative Power in Dresden Plutonium Recycle Assembly (Type VII-Pu) Hot, 20 v/o Steam, Beginning of Life

  • The circles are the symbols which represent the different composition rods.

The number in the circle is the relative power.

21

Ccntrol rod position j

Q f

'1.12

)

1.08 1.03 1.08 1

~

n 0.97 0.97 0.84

[1.21 )

0.84 1.01 N'

v 1.08 0.84 0.72 1.11 0.91 U

A

~(q A

A 1.03 i 1.21) 1.01)

(0.98)

(1.14) 0.92 V

V V

V y

m A

0.84

'1.11 I

1.02 0.97 1.01 0.91 0.92 1.02 0.92)

(.

U ", w/o Fissile Pu,w/o Total fiss., w/o No. 'of Rods 2

1.77 1.77 6

2.34 2.34 20

(

0.71 1.78 2.49 9

v I

1 Average: 1.82 0.46 2.28 35 Fig. 8 -Relative Power in Dresden Plutonium Recycle Instrumented Assembly (Type VII-Pu-I) Hot,20 v/o Steam, Beginning of Life

  • The circles are symbols ehich represent the different composition rods.

The number in the circle is the relative power.

22

'?,

(

TABLE 4 - LOCAL PEAKING FACTORS FOR DRESDEN TYPES VII (VI), Type VII-Pu, Type VII-Pu-L, AND Type VII-Pu-I FUEL ASSEMBLIES Basis -

Number of Rods Type per Element Peaking Factor BOC VII and vi 36 LS8 VII-I and VI-I 35 1.17 VII-Pu 36

- L 28 VII-Pu-L 36 1.28 I

VII-Pu-I 35 1.21 C

e 9

i l

23

Control rod posihon

,)

f j

1.13 1.'03 1.03 1.12 e

A A

A I.13

(

(

)

0.87 1.06 1

f r

R

'l.03 i 1.14 0.2

  • (1.00)

( 1.19) 0.97-vv e-A n

1.03

( 1.17)

(1.00) 0.70 0.80 0.98 V

V y

4 A

~

1.12 0.87 (1.19) 0.80 0.89 0.85 V

%W (1.06 0.97 0.98 1

0.96 IT2st w/n Fissile Pu. w/o Total Fiss., w/o No nf Rods 1.77 1.77 6

2.34 2.34 20 A

{

} 0.71 1.78 2.49 9

V 2.34 2.34 1 (55g Gd O )

2 3 Average: 1.84 0.45 2.29 36 Fig. 9 -Relative Power Distribution in a Dresden Plutonium Recycle Fuel Assembly (Type VII-Pu) Rotated 180* from Normal Orientation (Void Fraction = 0.20; Temperature = 546 F; Exposure = 0 Mwd /STM)

  • The circles are symbols which represent the different composition rods.

The number in the circle is the relative power.

24

Control rod position j

(

1 Gk 1

1.22 1.12 1.13 j

1.081 O

($

1.19 0.93

( 1.30 0.87 0.96 0.91]

%BY 1.06 I

0.72

'0.82 1.'00

~

O

. O (O 1.02

( 1.12) 0.2

( 0.98 )

1.17) 0.95 y

V V

U A

I[N.05)( 1.08) 0.81 1.00 A

1.07

( 1.23)

V V

V l

0.99 0.89 0.90 0.98 U235, w/o Fissile Pu.w/o Total Fiss., w/o No. of Rads 1.77 1.77 6

2.34 2.34 20 A

f

)

0.71 1.78 2.49 9

V 2.34 2.34 1 (55g Gd 0s) 2 Average:

1.84 0.45 2.29 36 Fig.10 -Relative Power Distribution in a Dresden Plutonium Recycle Fuel Assembly (Type VII-Pu) Rotated 90 from Normal Orientation (Void Fraction = 0.20; Temperature = 540 F; Exposure = 0 Mwd /STM)

  • The circles are symbols which represent the different composition rods.

The number in the circle is the relative power, 25

~

)

7,

~

3.2 TEMPERATURE AND VOID COEFFICIENTS The temperature and void coefficients for the plutonium-bearing i

]

Type VII-Pu assemblies are more negative than for the Type VII and Type VI assemblies for both cold and hot operating conditions. The moderator coefficients for a core containing plutonium-bearing assem--

blies will be less positive than for an equivalent all-uranium core. The limiting overall core moderator coefficients c_ontihue to occur at the end of the fuel cycle, under cold (68*F) conditions.

Moderator coefficients were determined at EOC-7 conditions for an average core exposure of ~12,700 Mwd / ton from two-dimamslonal multi-group calculations. Temperature coefficients were determined as a func-tion of core temperature at the EOC-7 average core exposure. The effect 1

of the presence of control rods was conservatively neglected in the tem-perature coefficient calculations. The effective moderator temperature coefficient at EOC-7 as a function of water temperature is shown in Fig.11, and as a function of exposure at 68'F in Fig.12. The modera-tor void coefficients at EOC-7 are given in Table 5.- The void coefficients are negative, and the nat effect of moderator heating is less than $1,00 i

3.3 IPEACTIVITY CONTROL AND SHUTDOiVN MARGIN The infinite multiplication factors under controlled and. uncontrolled conditions are presented in Table 3 for each type of fuel to be loaded into Cycle 7. Inspection of the table shows that the fuel and control rod char-acteristics combine to produce the most reactive core at ambient tempera-ture. Also, in Cycle 7, the effective reactivity level of the mixed core con-taining these fuel assemblies will decrease with exposure. Thus, the most reactive state of the core will be at ambient t'emperafure at the beginning of Cycle 7.

The license requirement for cold shutdown margin is that the core must be at least 1% suberitical with the worst control rod stuck out when the core is in its most reactive state.

Cold shutdown calculations were performed for BOL Cycle 6 and for Cycle 7 with a 96-element reload. The Cycle 6 data provided a basis for checking and normalization of the res.11ts, since experimentally de-rived cold shutdown data for BOL Cycle.6 are available.

The calculational results showed that with the core reload con-figuration displayed in Fig. 6, the license requirement for cold shutdown 26

_r.y 8

-rwe

-ww.-,.

-,v

-e, eeo m,

,m-

  • -o-

,~.n-=

L l

4 TABLE 5 - PREDICTED LIMITING VOID REACTIVITY COEFFICIENT DURING CYCLE 7 (End-oi-Cycle 7)

Void Fraction Moderator Void Coefficient, (Interior to Flow Channel)

Temp, *F Ak/k/%' Void -

0 68

-1.3 x 10-4 10 68

-2.0 x 10-4 i

4 9

O t

27

O 46.0

+5.s Q,-

l

+4.0

- AN EOC7

=

O I

+3.0 x

N 4

h

+2.0 i

"o S

+1.0

.g k

I 0

EOCG

.S 5

]

-1.0 o

o j

-2.0 s

Noa

-3.0 i

cb

-4.0 1

-5.0

-6.0 0

100 200 300 400 500 600 Temperature, F Fig.11 -Dresden I - Calculated Temperature Coefficient vs Temperature 28

_________________J

+10.0 g

l l

+8.0 l ~' '

l l

l

.l l

+ 0.0 I'#

Generalized ass'ys, l

de:

- + 5.2 x 10- s

=

+4.0 III-F, V, VI, VII l

l calculated for M

l l

mixed core ::.0 Measured r

l l

EOC 7 b

Type I %

y

+2.0 3

OC 5 l

f I

s BOC 4 0

measur 5 /

l l

' [ ing all

"~

q 0

f omeasured l taln measured

[

l Type VII-Pu I

c!

assemblics

-2.0

'O I

l ol

~

(

- EOC 6 calculated curve l

l Ol

-4.0 g

o Experimental data, BOC 4, BOC 5, BOC 6 l

g gl l

(Refs. DNPS 80-5-65; DNPS 106-5-67; j

g g[

l o

l j

-6.0 and DNPS 119-6-68) gj l

.2 l g

O Data for Type I fuel (TID-7672, p.80) l

-8.0 1

I c

I 2l l

l v

s

?!

I 25 -

-10.0 l

l l

i

-12.0 EOC EOC EOC Avg. core exposures at I:- 4 6

7

-14.0 O

2 4

6 8

10 12 14 3

Exposure, Mwd / ton x 10 l

.g Fig.12 - Dresden I Tempera'ure Coefficient vs Exposure at 687 l

L-----------_-----.

margin will be met. That is, with any peripheral rod stuck out of the core, the shutdown margin is in excess of 1?o. Greater shutdown mar-gins are obtained with any of the nonperipheral control rods stuck in the "out" position.

3.4 BURNABLE POISON BEHAVIOR Type VII, Type VII-Pu, and Type VII-Pu-L iuel utilize gadolinia as a burnable poison. The gadolinia, which acts as a highly self-shielded poison, is contained in a cingle UOrbearing fuel rod in each of these F

assemblies.

In the Type VII and Type VII-Pu assemblies, the gadolinia is con-tained in UOg (2.34 w/o enriehed) pellets and has a concentration within 8

the UO matrix of 1.74 w/o (55.0 g gadolinia per rod or 0.183 g/cm ).*

2 In the Type VII-Pu-L, the gadolinia is contained in UO (2.34 w/o enriched) 2 pellets and has a concentration within the UO matrix of 0.87 w/o (27.5 g 2

3 gadolinia per rod or 0.0915 g/cm ). In Type VII and Type VII-Pu-L, the concentration and geometric configuration of gadolinia is such that it is t

l57 effectively depleted in the high cross section isotopes, Gd ss and Gd after 1000 Mwd /T exposure.

('

d In the Type VII-Pu, the concentration and geometric configuration of gadolinia is such that it is effectively' depleted in the high cross sec-l57 tion isotopes, Gd'55 and Gd

, after ~5500 Mwd /T exposure.

As shown in Table 3, the reactivity worth of the gadolinia for the Type VII assemblies is ~5.5?o ak/k at 68*F. The reactivity worth of the gadolinia for the Type VII-Pu and Type VII-Pu-L is ~4.37o Ak/k at 68'F. Because the proposed loading pattern contains only 80 Type VII and nine Type VII-Pu and VII-Pu-L gadolinia-bearing assemblies, the effective poisoning of the core is only ~1.2?o Ak. Evaluation of the gadolinia burnup in Cycle 7 indicates that with the proposed loading the core will have maximum reactivity at the beginning of the cycle. That is, the positive contribution to core reactivity due to gadolinium burnup is always less than the negative contribution of fuel burnup throughout the core. The net change in reactivity with time always will be negati/e.

  • Additional information and experience with UO -gadolinia 2

is given in Section 5.

30 e

3.5 ' NUCLEAR CALCULATIONAL METHODS FOR UO AND 2

PuO -UO FUEL 2

2 3.5.'1 Fuel Element Nuclear Calculations The nuclear constants of the individual foel assemblies are calcu-lated using both zero-dimensional and one-dimensional reactivity and burnup formulations. The neutron spectrum is recalculated before cach discrete time step. The calculation method for Type VII uranium assem-blins is the same as used for Cycle 6 (Reference 6) in which the multi--

group Amouyal-Benoist method is used to calculate thermal flux self-shielding averaged over the thermal neutron spectrum. Thermal neutron parameters in Type VII-Pu assemblies are calculated using a code equiva-f lent to the THERMOS code with the Nelkin scattering kernel to determine spatial and energy dependent spectrum-averaged properties.

Effective burnable poison thermal cross sections are obtained from three-dimensional Monte Carlo calculations performed at several burnable poison depletion states. A nonlattice peaking factor is applied as a correction to the homogenized model to account for different thermal

. flux levels between the water outside the shroud and the water in intimate contact with the fuel pins. The ncalattice peaking factor is obtained from two-dimensional diffusion theory calculations.

Nuclear consta'nts derived from the above calculations include M,

2 k., and number densities of all important isotopes as a function of burn-up. In addition, the power-dependent xenon worth, required in the three-dimensional burnup calculatims, is ccmputed as a function of burnup.

The calculation method for uranium assemblies was checked ex-tensively against light water reactor cold critical and operating data, and against Monte Carlo calculations, with excellent agreement.

The calculational method used for Type VII-Pu assemblies has d

been extensively checked >8 against experimental data for mixed-oxide lattices with a deviation between calculated reactivity and experimental data of 0.6% Ak/k over a range of water-to-fuel volume ratios from 1.1 to 8.2. This calculational procedure also predicted the reactivity levels in the EBWR Plutonium Recycle Experiments to within

  • 0.3% Ak/keff.3 31 i

~

3.5.2 Three-Dimensional Power and Burnup Distributions l

Three-dimensional power and burnup distributions were calcu-lated for Cycle.7 starting with burnup distributions projected for the remaining fuel at the end of Cycle 6. In these calculations, nuclear.

2 constants such as k=, and migration area. M, for each fuel assembly type were allowed to vary as a function..ae local burnup and local coolant void fraction, for all fuel assembles. in coupled nuclear /

I thermal-hydraulic reiterative calculations. The local reactivity was also made to vary with local power -

.eration -reflectin' the effect-g of fuel, Doppler temperature coefnelent, and equilibrium xenon con-centration which vary with power generation.

4 In this calculation model, neutrons are absorbed within the node boundaries in which they are thermalized, since node dimensions are in excess of.5 in., which is several times the mean free path of thermal 4

neutrons. The fast neutrons are allowed to migrate to adjacent nodes and beyond until they are thermalized.

Allowance is made for the presence or absence of a control-rod at each node. Reflectors are treated by calculating an albedo at the-4

-('

per fission applicable to PuO -UO fuel.

core boundaries. Allowance is also made for using the correct Mev 4

2 2

Burnout heat flux ratios are determined for all nodes from power i

generstion and quality data generated in the above calculations,'taking in-l to account the power-dependent coolant flow ~ distribution, and the depend-ence of burnout heat flux on local quality and mass velocity.

The validity of the above calculation m'ethod has been checked -

by comparison with fuel cycle lengths and power distributions from Dresden I and an operating PWR." Fuel cycle lengths calculated by this method are within 5% of those actually achieved in Cycles l' i

~

and 4 of Dresden I, and also within 5% of the cycle length expected -

for Cycle 1 of the Trino Vercellese (SELNI) PWR reactor. Dresden I control rod critical positions were predicted to within 0.2% Ak/k for beginning of Cycle 6 operation.

Radial power distributions calculated for Cycle 1 of Trino at about 3000 Mwd /MTU.show excellent agreement with experimental power distributions derived from Mn" activations. The average de-32 4

e em

T viation between calculation and experiment for each fuel element is less.

~

than' 2%. In addition, calculated axial power. distributions for Dresden I, Cycle 4, are in very good agreement with flux wire ' measurements.'

3.5.3 -Moderator Temperature Coefficients The moderator coefficients for uranium assemblies at EOC-7 were determined for a generalized uranium fuel a'ssembly represent- -

ing Types'III-F, V, and VI fuel elements following burnout of burnable poison, at EOC core-average burnup. These calculations were per-

-formed using a two-dimensional diffusion theory program with cross-section input derived from the zero-dimensional reactivity and burnup calculations. The'effect of.the presence of control rods was conserv-atively neglected in the calculation of the coefficient.

The justifications for using the generalized fuel assembly rep-resenting Types III-F, V, VI, and VII fuel, and for neglecting the con-trol contribution are as follows:

1.

At EOC 7, essentially all gadolinia is' depleted from all ele-ments, and 65% of the core is comprised of Types III-F, V, VI, and VII elements. These four element types are geo-metrically similar and have the same initial enrichment.

As shown in Fig.12, the calculated temperature coefficient' for the generalized element is a nearly linear function of ex.

posure in the range of interest. A scatter-loading refueling pattern is employed which tends to average out the nuclear properties of the fuel in the core. Data for' Type I fuel (1.45% initial enrichment) given in' Fig.11'(Reference 11) show the temperature coefficients for this fuel to be nearly the same as a function of exposure as for the generalized fuel element. The temperature coefficients expected for' Type III-B fuel elements should be no more positive than those shown in Fig.12 for Type I or the generalized fuel element. This is because Type III-B (which has an initial enrichment intermediate between Type I and the generalized fuel element) contains erbium, a slow-burning poison which -

tends to make the moderator coefficients less positive."

Therefore, for the purpose of calculating moderator coeffi~

cients, the entire core is reasonably represented as fueled.

entirely with the generalized fuel type.

33

______m_

_-. _ _ _ _ _ - _ __._ _ _ -.-_mme_.__

m

\\'

I' i

4. THERMAL AND HYDRAULIC CHARACTERISTICS The thermal and hydraulic characteristics of the Type VII uranium-assemblies are identical to those of the Type VI assemblies.

The thermal and hydraulic characteristics of the plutonium-bear-ing assemblies (Type VII-Pu, Type VII-Pu-L, and Type VII-Pu-I) have been examined. These assemblies meet the existing license require-ments for the Type VI (Type VII) assemblies which are:

1.

The maximum heat flux at design power shall not exceed 2

(

360,000 Btu /hr-ft,

2.

The minimum critical heat flux ratio shall be greater than 1.5 when calculated at 125% design power.*

The hydraulic characteristics of the plutonium-bearing assem-blies (Type VII-Pu) are essentially the same as those of Type VI (or

' Type VII) fuel.

The maximum allowable hot spot factor for the plutonium as-semblies is 3.44, which is the ratio of allowable-to-average heat flux (360,000/104,444). This hot spot factor is composed of engineering nuclear factors. The engineering hot spot factor is 1.061 based on plutonium and rod diameter tolerances. Using this engineering factor, the allowable nuclear hot spot factor is 3.24 (3.44/1.061). The nuclear factor consists of radial, axial, and local power factors that are func-tions of core location, control rod pattern, and fuel burnup. At the be-

  • Based on " Burnout Limit Curves," from Reference 15 and average quality over the cross section of the assembly.

35

2.

The contribution of controlled nodes to the overall coefficient is small, particularly at EOC conditions, because the fraction of the core contrdiled at EOC conditions is relatively small; also, the flux square importance weighting required to obtain the overall coefficient would further reduce the contribution from the controlled ncdes, since the flux is depressed in these nodes by virtue of the presence of control. Inclusion of the control contribution would tend to make the computed coeffi-cients more negative (or less positive).

The temperature coefficient computed in the above manner at 68'F as a function of exposure is presented in Fig.12. The slope of the line drawn for the generalized uranium assembly is +0.40 x 10-3 (Ak/k)eff/

'F per 1000 Mwd /T advance in average core exposure. This slope is in good agreement with that obtained from EOC-4 and EOC-5 coeffi-cients reported in previous submittals,14 which yield a slope of +0.45 x-ts 10-3 (Ak/k) eft / F. Also shown on Fig.12 are the EOC average core ex-posures for Cycles 4, 6, and 7. The advance in average core exposure from EOC 6 and EOC 7 is small because the equilibrium cycle is being approached. Only a small increase of +0.4 x 10-3 in moderator coeffi-cients from EOC 6 to EOC 7 would be expecte'd if all assemblies in Cy-I- '

cle 7 were Type VII. uranium elements.

The average moderator temperature coefficient for a core contain-ing'all uranium assemblies at EOC-7 condition is +5.3 x 10-s Ak/k-F.

The moderator temperature coefficient calculated for an uncontrolled Type VII-Pu assembly is -1.0 x 10-8 ok/k-F at 4000 Mwd /T and has a L slightly negative slope as a function of exposure.

The core-averaged temperature coefficient (68 F) for Dresden I with 11 plutonium recycle assemblies is +5.2 x 10-8 Ak/k-F at EOC 7.

This compares with+4.9 X 10-5 5k/k *F predicted for EOC 4 and +5.2 X 10-3 Ak/k-F predicted for EOC 5.13,14 34

--.=.

ginning of life, the local peaking factor for the plutonium assemblics is 1.28, which allows a radial x axial nuclear factor of 2.53. Detailed nu-t' clear analyses using three-dimensional core representation, allowing for' fuel burnup distribution, nonuniform void distribution, and planned control rod withdrawal patterns have shown that the nuclear peaking factors can be maintained below the designated limits.

Critical heat fluxes are calculated using the Janssen-Levy critical heat flux design limit curves" for the average quality over the cross section of the assembly and the assembly flow rate. Preliminary studies indicate that the critical heat flux ratios (CHFR) would be greater than the license limit of 1.5 at 125% design power. The actual CHFR's to be obtained during operation will be recalculated during the cycle fuel-management, and the control rod patterns will be adjusted to maintain ratios greater than the license limit of 1.5.

The centerline fuel temperature is less than the melting point of the (Pu-U)O mixture (melting point 5050'F) at the maximum allowable 2

2 heat flux of 450,000 Btu /hr-ft (19.4 kw/ft). For (Pu-U)O mixtures with 2

less than 5 w/o PuO, the thermal properties of the mixture are very 2

nearly the same as UO. The thermal conduc'tivity of fuelused in these 2

I, calculations is base _d on data presented in Reference 16. The fuel-clad 2

gap conductance was conservatively taken as 1000 Btu /hr-ft

  • F. The peak fuel temperatures at design and 125% design power conditions are 3940'F and 4930*F, respectively.

Previous operating cycles have shown an increase in core pressure

~

drop during the cycle which results in a 4 to 5% decrease in total recircu-lation flow. This increase in pressure drop is caused primarily by crud

~ buildup on'the inlet orifices of the fuel assemblies. Experience on Cy-cles 4, 5, and 6 has indicated that cleaning all of the orifices during re-fueling results in reproducing design flow rates at the beginning of the subsequent cycle.

Crud buildup has different effects on the assembly flow rates de-pending on the orifice configuration. In general, the flow rate increases in the assemblies with large orifice diameters located in the inner flow region of the core at the expense of a decrease in flow in the assemblies i

{

with the small orifice diameters located in the peripheral flow region.

~

These variations in assembly flow rates including a 5% decrease in total recirculation flow during the cycle have been shown not to produce heat -

fluxes or critical heat flux ratios outside of the license limits.

36 L

t i

I -.-

w

.5.

ADDITIONAL INFOR'!ATION ON AND EXPERIENCE WITH GADOLINIA-URANIA

,5.1 DRESDEN OPERATING EXPERIENCE WITH GADOLINIA POISON Gadolinia-bearing " fuel-poison" rods in four special-Type III-F assemblies, which were introduced at the beginning of Cycle 4, now have been irradiated for a full additional cycle (Cycle 5) in the Dresden I station.

No operational difficulties were encountered with the additional exposure as of the end of Cycle 5 In addition, Type V assemblies have performed as predicted for one complete cycle and a portion of a second cycle.

No failures have been found in any of the gadolinia-urania rods.

The nuclear behavior of gadolinia has been well predicted. The correctness of design methods for calculating gadolinia effects have been verified by the fact that the Cycle 5 control rod withdrawal pat-terns agreed with predictions and Cycle 6 beginning-of-life reactivity leval was predicted within 0.2% Ak/k. Cycle 6 operation shows excel-

~

lent agreement between predicted imd observed control rod movements.

5.2 BIG ROCK POINT OPERATING EXPERIENCE Six rods containing axially distributed gadolinia in urania have been irradiated in the Big Rock Point reactor. Sixty-three percent of the Gd O -UO Pellets used in these rods have gadolinia concentrations 2 3 2

equal to or greater than that of tM gadolinia in the Type VII and Type VII-Pu assemblies. As of June 1968, these assemblics had been operated to an exposure of 14,900 Mwd /E.TU without evidence of failure.

37

5.3 UNC PLATR FACILITY EXPERIMENTS Critical" tests were performed on gadolinia using the void sub-stitution method at UNC's PLATR facility (Pawling Lattice Test Rig) which included critical tests with and without Gd O rods included in 2 3 the mocked up assemblies. - These tests yielded a verification of the :

calcuhted Gd O reactivity worth.

2 3 b

o O

O 9

38 l

_ J

0

\\

\\

\\

6. SAFETY CONSIDERATIONS In considering the safety of a reactor containing plutonium-bearing assemblies, two areas of concern are: (1) the potential hazard associated with the accidental dispersal of the plutonium from unirra-diated fuel assemblies, and (2) the effect on the kinetic behavior of the core due to the reduced delayed neutron fraction in plutonium fuel.

The hazards resulting from a failure of irradiated PuO -UO fuel 2

assemblies is not significantly different from the hazards associated

(

with the failure of irradiated UO fuel assemblies.

2 6.1 POTENTIAL HAZARGS ASSOCIATED WITII THE HANDLING OF UNIRRADIATED TYPE VII-Pu ASSEMBLIES The Type VII-Pu fuel assembhes will be subject to the same stringent integrity specifications and quality assurance checks em-ployed for Type VI fuel. The Type VII-Pu fuel assemblies will be shipped to Dresden in sealed shipping containers which conform to DOT regulations.

When the individual shipping containers are opened, the assemblies will be surveyed and smear tests taken.

The Type VII-Pu assemblies will then be stored in the New Fuel Storage Vault until loading.

In the unlikely event that a Type VII-Pu assembly is dropped or that radiation surveys indicate a possible leak in a plutonium assembly, procedures outlined in the "Dresden Radiation Control Standard" will be in effect until it is determined-that no radio-active hazard is present.

Additionally, the air in the New Fuel Storage Vault will be sampled weekly for alpha activity while the Type VII-Pu fuel assemblies are being stored.

39 O

O g

[.

At the time of refueling, three fresh Type VII-Pu assemblics will be transferred into the fuel storage pool, channeled, placed in a fuel transfer basket, and transfered to the reactor canal for loading. From the moment that the Type VII-Pu assemblies enter the fuel storage pool, the assembucs will be subject to the same handling procedures as em-ployed for the potentially more hazardous irradiated Dresden fuel as-semblies.

The problems associated with radiation protection while handling l

fresh fully encapsulated plutonium fuel are not qualitatively different -

i from those which exist with the presently licensed uranium fuel. As noted in Reference 18:

" Plutonium fuels have been stored and handled in the same man-ner as uranium fuel, and irradiated fuels have been routinely han-died for special examinations and core changes without difficulty.

No unusual procedural controls have been made necessary, nor has any specialized operator training been required specifically l

as a result of using plutonium fuelin the PRTR.

"The PRTR experience has shown that the effects of plutonium fuel failures are no different than those for irradiated uranium g

fuels. Emissions have been virtually limited to fission gases N

with no evidence of particulate washout. Alpha contamination, usually of primary concern in fabricating plutonium fuels, is of little concern in reactor operations, as gannaa contamination governs procedures for almost all maintenance work."

In addition, the fuel in the Type VII-Pu assemblics will be in the form of sintered pellets rather than the vibratory compacted form used In PRTR. The sintered PuO -UO pellets are inherently less susceptible 3

2 to particulate leakage or washout.

Experience with PuO -UO in the Saxton Reactor Core II," the 2

2 20 22 EBWR, the PRTR,21 and Dresden Cycle 6 have shown that fresh PuO -UO fuel assemblies can be handled safely.

2 2

Radiation levels during fuel handling of Type VII-Pu assemblics will be larger than for uranium assemblies due to the 0.06-Mev gamma activity from the Pu decay dauchters and the 0.7-Mev gamma activity 241 from residual fission products (Zr", Nb", Ru*, and Rh ") carried along 8

40 6,

with the high-exposure plutonium during reprocessing. The gamma s

plus neutron dose rate from Pu at the surface of a clad rod is expected to be ~25 mrem /hr. Perscnnel monitoring procedures will be follow-ed and protective shielding used as~ required.

6.2 EFFECT OF THE DELAYED NEUTRON FRACTION OF

. PLUTONIUM FUEL The fresh Type VII-Pu assemblies contain approximately the same quantity of plutonium as contained in Type III-F, Type V, or Type VI fuel after irradiation to 20,000 Mwd /STU. The isotopic ~ com-position of the plutonium in the fresh' Type VII-Pu assemblies is also approximately the same as in the irradiated uranium assemblies.

The calculated delayed neutron fraction Seff or a fresh Type VII-Pu f

assembly (0.0065) is the same as that calculated for a Type VI assembly after 10,500 Mwd /STU, and is greater than that for hvghly exposed urani--

um assemblies because a larger fraction of power in the Type VII-Pu assembly is produced in uranium.

6.2.1 Refueling Accident k

The analysis of the refueling accident parformed for Cycle 6 load-ing*3 was repeated for a Type VII-Pu-I assembly refueling accident. In/

this analysis, the Type VII-Pu-I assembly was inserted into a vacant fuel position at the maximum hoist design rate of 12 in./sec. A fuel assembly.

reactivity worth of 1.5%2k/k was used.

Two control rods, next to the vacant fuel position, were assumed to have been inadvertently withdrawn to give a'near-critical 2 x 4 uranium fuel assembly array. Negative Doppler feedback was assumed. As in the previous analysis, no negative reactivity feedback effects from clad heat-ing or void forination'were considered. The resulting excursion will be terminated (as with a Type VI-I refueling accident) by the overpower scram system without fuel or clad inelting.

Based on these assumptions, the calculations indicate that the maximum radially averaged fuel temperature in any fuel rod would be less than 2850*F, and the corresponding maximum central fuel temperature would be less than 3275*F. Clad temperatures will not exceed 2250'F. The effect of reduced on the refueling accident re-41 H.

l f

i, sulted in only a 76*F rise in the center fuel temperature. The #eff or the 2 x 4 assembly array is 0.0066 vs 0.0070 originally used for the Cycle 6 reload accident with uranium fuel.

6.2.2 Core Kinetic Behavior During normal operation, the rate of reactivity change is adjusted by the operators to maintain the reactor period within the normal opera-tional limits. No additional hazards are introduced by maintaining the present reactor period limits during Cycle 7.

The response of the Dresden I reactor to an uncontrolled reactivity addition (accident situatio 1) is determined by: (1) the magnitude of the reactivity chnge, (2) the effective neutron lifetime, and (3) the magnitude of negative feedback mechanisms (prompt acting Doppler coefficient, moderator temperature, and void reactivity coefficients).

For an accident involving a reactivity addition less than # eft, the reactor response is governed by the relative magnitudes of the neutron lifetime and the inherent feedback mechanisms. The average neutron lifetime is a function of the lifetime of prompt and delayed neutrons, i.e.,

\\ ;~

f = f* +pegg T where f* = the prompt neutron lifetime, sec (slowing down time plus thermal lifetime)

Y = average of the delayed neutron mean lifetime, sec.

A reduction in the value of the average neutron lifetime produces a more rapid response to a given Ak/k change.

The values of f* and Yare essentially equal for Type VII-Pu and Type VI and Type VII uranium assemblies. The variation of #eff with exposure is shown in Fig.13 for Type VII and Type VII-Pu assemblics.

Initially #eff of the Type VII-Pu assembly is equivalent to uranium as-semblies exposed to ~10,500 Mwd /STU and remains only slightly smaller than the higher exposure uranium assemblies for the remainder of life.

The overall kinetic response of the Dresden I reactor is approxi-mately equivalent to a BOC 7 fuel loading where the Type VII-Pu fuel I

I assemblies are replaced with uranium assemblies exposed to ~10,500 Mwd /STU 42 l

l

0.0090 t

t 0.0080 i

o Type VI and VII assemblies 0

0.0070 cf Type VII-Pu assemblies t

l 0.0000 w

l 0.0050 0

10,000 20,000 30,000 Average Assembly Exposure, Mwd /STM Fig.13 - Calculated Effective Delayed Neutron Fractions (Fert) for Type VII and Type VII-Pu Assemblics 43 l

Since the average burnup of uranium assemblies discharged from Cy-cle 6 is greater than 10,500 Mwd /STU, the' kinetic response of the core is equivalent to a Cycle 7 loading in which only 85 (96-11) fresh uranium ass' mblies are used.

e A comparison of calculated values of #eff at the beginning and end of Cycles 6 and 7 (Table 6) indicates only a slight reduction in the core-averaged #eff. The decrease in #eff noted for Cycle 7 is due mainly to the higher average exposure of the uranium assemblies rather than to the presence of the 11 Type VII-Pu assemblies.

Thereforc, ne significant changes in the dynamic response of the reactor.should occur as a result of the introduction of plutonium assemtlies.

Moreover, the increased moderator temperature coefficient and the increased prompt Doppler coefficient (approximately 15% greater than uranium fuel) result in a stronger acting shutdown mechanism.

As noted in Section 6.2.1, the refueling accident for an array of one Type VII-Pu assembly and seven uranium assemblies yields es-sentially the same results as for an all-uranium array. Other accidents involving the entire core would be less affected by the presence of plu-tonium since the ratio of 11 Pu assemblics to 464 assemblies in the core

\\

(0.024) is considerably less than 1 in 8 (0.125).

f l

O f

44

_____-__2__

______m_______

TkBLE 6 - CALCULATED Sett FOR DRGSDEN CYCLES 6 /d?D 7 Core Averaged Serr Serr, Pu Assemblies Beginning of Cycle 6 0.0069 End of Cycle 6 0.0064 Beginning of Cycle 7 0.0068 0.0065 End of Cycle 7 0.0062 0.0062 4

e 4

F 45 w-

+

s-*-,yr*

m, y,,-

s<m-

,--y,wv--

-, g n,

,-,,--.,y,g,.

g

--.,,,e

I

7. REFERI OES 1.

Uotinen, V. O. and Williams, L. D.: Experiments and Calculations for H 0-Moderated Assemblies Containing UO -2 w/o PuO Fuel 2

2 2

Rods, BNWL-SA-1107 (May 1967).

pug -UO Fueled Critical Experiments, WCAP-3726-1 (July 1957).

2.

2 2

3.

Kier, P. H.: Analysis of the Initial Critical Experiments of the EBWR Plutonium Recycle Program, ANL-7368 (Aug.1967).

(,

4.

Celnik, J. et al.: Evaluation of Plutonium Recycle Nuclear Calcu-

\\

lation Methods by Comparison with Experimental Data, UNC-5168' (Feb. 28,1967).

5.

Calabro, R.: Test of Plutonium Recycle Nuclear Prediction Methods for Heterogeneous Lattices and Reactor Experiments, UNC-5211 (Apr.15,1968).

6.

Description and Safety Evaltiation Report of Proposed Change No.14 for Dresden Unit 1, Cycle 6, Docket No.50-010.

7.

Honeck, H. C.: THERMOS - A Thermalization Transport Theory Code for Reactor Lattice Calculations, BNL-5826 (1962).

8.

Orr, W. et al.: Nuclear Design of the Saxton Partial Plutonium Core, WCAP-3385-51 (Dec.1965).

9.

Goldstein, L. et al.: Calculation of Fuel-Cycle Burnup and Power Distribution of Dresden I Reactor with the TRILUX Fuel Manage-ment Program, Presented at 1967 Annual Meeting of ANS, San Diego, Calif., June 11-15, 1967.

'46

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TAllLE !! - (CONTINUED) s s,

Claditin:.

Iterut.tr Itivis Spe cial Corner it. win

't.

Wall Thiit-

. it.sl Cun.

Numter m u fuel u o 00 Fwl Numtxt w a ruct w n 10 Futt nens,in.

in ur.ition Enrirbme ni Dia mi ter ite quirest l Comiw+1tton Enrichment Diameter Fuel Type Materi.il

.OD. 6n.

Ite quirist Compi ition-g Vil {plut.. mum Zr-2 0.5025 0.035 G G 20 j 100 t'O, 2.34 0.432 G

' 100 '. O 1.77 0.482 i

berring.Vil-9 i 97.7 UO.

0.11 tr nat.1 0.4 R2 Pulj 2.31%O:11 14 98.2G t'O:

2.34 0.432

~

1.74 Grt.0 2

VII [ plutonium Zr-2 0.5625 0.03 5 66 20 100t0 2.34 0.482 6

i 100 002 1.77 0.482 beuing. low 9

97.7 00 0.71 st nat. I 0.482

.I gadolinium 2.3 INO.II

  • (VII-Pu-L)J l'

99.13 00:

2.34 0.482 0.67 Gil;O.

~

VII [ plutonium Zr-2 0.5625 0.03 5 66 20 100 00:

2.34 0.482 G

100 002 1.77 0.482 betting, inst ru <

9 97.7 00, 0.71 (U nat.1 0.482 mented (Vil-2.3 PuO Il Pu-l)]

~

PF-10 Zr-4 0.412

' O.025 88 54 99.6500:.

2.0 0.358 9

99.65 00:.

~

1.5 0.358 0.35 Er O.

0.35 Er:0.

SA-l*

Zr-2 0.424 0.022 77 n

100 00

3. 0 -3. 5 0.375 100 00 2.76 6.375, i Poiron rod. pellettred.

8!!sotopic composition of plutonium in atom percent is Pu: Pu?', 0. 41: IN!1' 4 71.3';; Pu'* * < 20.67: Pu!". 6.1i; Pu28: = 1.6 *.

  • 10 rod segnents sere remercJ fron St-1 during the foneth rar...tinz (9u n,

1 l

l l

l 1

l e

4 l

l j.

,~

.

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