ML19345E714

Notifies That Licensee Inadvertently Failed to Include Pictures in Analyses Part of Addl Info on Proposed Change 19.Pictures Encl

ML19345E714
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	01/28/1970
From:	Walke G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Morris P, Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML19345E712	List: ML19345E711 ML19345E714
References
NUDOCS 8102050691
Download: ML19345E714 (12)

Text

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- CORSumBIS PoWCr y

,,x company General Offices: 212 West Machigan Avenue. Jackson, Michigan 49201. Area Code S17 788-0550 January 28, 1970 1

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l Dr. P. A. Morris, Director WM

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Division of Reactor Licensing (yf/e'r' US Atomic Energy Commission Re: Docket 50-155 - Proposed Change Washington, DC 20545 No. 19 to Technical Specifications

Dear Dr. Morris:

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D. J. Skovholt On Wednesday, January 28, 19r70, we transmittee to you forty (h0) copies of "Additiona3" Information on Proposed Change No.19 to Technical Specifications.

We, inadvertently, fr.i:ed to include the picturec in the analyses part of this transmittal. Therefore, we are transmitt'ng these under cover of this letter and would appreciate it if you w d

see that they al.

cluded in your forty (h0) copies.

Yours very truly, lil Jke M

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WGF/fs Gerald J. Walke Nuclear Fuel Management Administrator

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ADDITIONAL INFORMATION Change %.

19 Consumers Power Company -

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Docket.No. 50-1551 geguisted y@

Request for Change to the Technical Specifications-License ' No. DPR-6

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FUEL TRANSIENT BEHAVIOR

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p BLENDED PuO -UO2 2

& PARTICLE SIZE CHARACTER 1ZATION -

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A.

Introduction General Electric has proposed the irradiation of three Pu0 -UO2 e ntaining 2

fuel bundles in Consumers Power Company's Big Rock Point reactor as part of the EEI program. ' At the beginning of cycle, the plutonium in these bundles would make up 27 percent of the plutonium in tt.e core. At the end of the cycle, the plutonium bundles

- woeld contain 20 percent of the plutonium in the co

. The totc1 plutonium in the core is expected to remain essentially constant at 65 kg throughout the cycle.

These Pu0 -UO fu is were fabricated using physically mixed powders and a 2

2 cold press and sinter process. The size and distribution of the clutonium oxide particles within the fuel have important consequences relating to transient fuel per-formance. Recognition of these consequences has led to transient testing and analysis and to careful characterization and control of particle size during the EEI production

. campaign. Transient test work in SPERT began in 1968. Some four test pins have been

-irradiated to date and seven more tests are planned in 1970.

B.

SPERT Transient Test Results The SPERT tests form the basis for our statement that Pu0 -UO fuels behave 2

2 similar to UO fuels. As stated above, four tests of physically mixed, Pu0 -UO f""1 2

2 2

have been completed in the SPERT Capsule Driver Core. The test fuel pins were manu-factured by General Electric in the Plutonium Laboratory at Vallecitos, as was the EEI fuel-in question. Two fuel pins contained cold pressed and sintered pellets and two contained annular, cold mressed and sintered pellets with the center hole scaled to the center hole of the EEI fuel for BRP. The characteristics of these fuels are given in Table : and Table 11 lists the test results. The major impression of the test results can be der lv.vl by observing Figures 1 to 6, which are post-transient photo-graphs of the Pu0 -UO test pins and UO test pins of similar design. Observe that 2

2 2

none of these fuel materials were dispersed even though the cladding has melted and broken in some cases. Figures 2, 3 and 6 illustrate clearly that the fuel pellets remained intact af ter 275 cal / gram energy insertions.

. These test observations form the primary basis for considering 265 cal / gram

' es a conservative failure threshald for physically blended Pu0 -UO fu 1.

Neither 2

2 SPERT nor General Electric observers have detected any significant difference between the transient behavior of these mixed Pu0 -UO test pins and standard UO L st pins.

2 2

2

- These tests lend credence to calculations which indicate no significant consequence in using this f uel form. These calculations are discussed further in Section D.

C.

Plutonium Particle Size Distribution The Pu0 -UO fuel rods in question were produced by mixing oxalate-derived 2

2 Pu02. powder with Eldorado UO2 p wder. Both the Pu0 and UO produced by the above 2

2 processes have inherently small crystallite sizes so a particle in the sense discussed hereaf ter is an agglomerate of either Pu0 r UO crystallites or a mixture of both.

2 2

Track Etch counting techniques and neutron radiography have been found to give adequate resolution to define the distribution of particles in Pu0 -UO f " l*

• 2 2

Alpha radiography is used to determine the particle size distribution and neutron radiography is used to detect large particles.

1 Alpha autoradiography is used to determine the particle distribution by statistically sampling several cross sectional areas of pelletized fuel from each blended powder batch. The Track Etch method used is capable of detecting-particles as small as 20 microns in diameter. A 1000 micron wide band on each Track Etch film.

is examined on a transmitted light microscope. The size and number of particles is determined for 10 micron size increments and a mass distribution plot is constructed from the data. The particle population follows a log normal distribution, which is characteristic of dispersions in powders. A linear regression is performed on the distribution curves from five samples per powder blend which yields the most probable

' distribution.

l The average distribution of the EEI fuel was linearized to log normal type distribution and is shown as Figure 7.

This distribution was derived using the assumption that all particles are 100 percent Pu0. A comparison of the alpha track 2

• General Electric Service Mark f

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,. density of the particle with that of a 100 percent p uton a stan ard indicated that l

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the pa;:ticles were only about 20 to 40 percent plutonia, which makes the above dis-tribution calculation conservative. Furthermore, only part of the plutonium could be accounted for in particles above 20 microns in most cases, inferring that the remainder is in particles below the detection limit. This conservatism is not taken into account in establishing homogeneity at this time because of uncertainties in relating Track Etch density to plutonia concentration.

Neutron radiography is used as a non-destructive method of examin,e 100 percent of the fuel for large particles. Neutron radiography can resolve particles as small as 200 microns and a statistical chance of a few particles above 100 microns exists, as predicted by the distribution curve determined by alpha-autoradiography.

Observations of the neutron radiographs of all the fuel in question show the individual pellets to contain 0 to 4 particles above 200 microns. The worst case observed showed that about 0.05 percent of the plutonia exists in particles larger than 200 microns (assuming that these particles are 100 percent plutonia). Comparison of these particles on neutrographs with those obtained fram 100 percent plutonia microsphere standards indicates that most of the particles are not pure plutonia. The overall impression of the neutrographs is that the fuel contains a sparse population of particles (10 to 130 per rod) in the range 200 to 500 microns. No particle greater than 500 microns exists in the fuel.

D.

Impact of Size Distribution on Predicted _ Transient Behavi,or An evaluation has been made of the effects of plutonium particle size dis-tribution on the transient behavior of EEI mixed oxide fuel for BRP. Both calcula-tional results and experimental data are available to help make this evaluation.

The bulk of the energy would be generated in plutonium particles. It is obvious that in order for the particles to transfer their heat to the surrounding fuel matrix they must be at some temperature higher than the average matrix temperature.

A simple, conservative spherical heat transfer model has been formulated and calcula-tions with that model have produced the following results, f

. The pa rticles in lowest enriched fuel will experience the highest tempera-tures for a given average energy deposition. The lowest enrichment used in these

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bundles war 1.6241 Since primary concern is for fuel below 265 cal /gm energy deposi-tion (fuel above this energy is assumed to be dispersed in any event), all calculation were done for this energy deposition. Any increase in enrichment or decreasa in energy deposition would cause a reduction in consequences discussed below. Also, any decrease in Pu concentration in the particles would reduce the consequences.

The calculations show that particles up to 50-60 microns in diameter behave essentially like homogeneous fuel. The particlcs may be completely melted but they transfer 'their heat 'quickly enough to keep them from generating. any significant vapor pres su re. As can be seen from Figure 7, some 15% of plutonium may be in particles greater than 50 microns in diameter. These-particles can experience local peak enthal-pies that are extremely high. However, the frcction of the particles large enough to reach very high peak entha?py is very small. The mass weighted average peak enthalpy in the particles greater than 50 p in diameter would be approximately 750 cal /gm. Much of this energy is required to heat the particles so that only the energy that goes into creating vapor is available for conversion into mechanical energy. These considerations lead to an estimated maximum thermal energy in the plutonium particles available for

' conversion to mechanical energy of 0.3% of the total energy.

The-peak particle enthalpics occur very shortly after peak power and as tbu power decreases the particle enthalpies decrease rapidly. Measurements of clad temper-a tures during an excursion were made by SPERT, and the data indicate that at the time of peak power and for a short period af ter the cladding temperature does not rise more t han a few hundred degrees. At the end of the transient, the clad temp rature is rising rapidly and eventually reaches very high values up to and including meltinh. The con-clusion is unat at the time of peak particle enthalpy the cladding is relat: vely cool and possesses considerable strength. Calculations indicate that the fuel and the cladding can easily absorb the potential mechanical energy available without gross loss of integrity.

The potential exists, however, for very localized failures being generated by the few particles in the 200-500 micron range. These failures might occur if one of these large particles should happen to be located near the cladding. These large i

particles make up a very small fraction of the fuel. The fuel dispersal that might i

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1

> result from such isolated perforations is considered negligible compared to the fuel above 265 cal / gram, which is assumed to be dispersed in any event. Also, the tests in SPERT summarized above had a few of these large particles within the test pins and no effect is immediately apparent. Cerf anly they caused no gross effects.

One final factor to be considered is the effect of burnup. Experience has shc=m that even af ter. very short periods of irradiation the particles become smeared and most of the probleus resulting from the existence of particles disappear. This strengthens the arguments made above in that 'the potencial effects of the particles exist only when the cladding is very ductile and capable of absorbing considerable energy.

E.

Impact of Size Distribution on Doppler Feedback The hold-up of heat in the plutonium pr.rticles might indicate that Doppler feedback would be affected. This is not the case. As was mentioned earlier, 85% of the particles behave as though they are homogeneously mixed with the UO. The calcula-2 tions indicate that the remaining 15% of the particles give about 90% or more of their heat promptly, although some 98% of the heat is ~promptly transferred to the UO. Also 2

the contribution to Doppler of the Pu-241 in the particles actually can increase Doppler feedback. The ef fect the plutonium fuel has on the Doppler is certainly -less than 1% which is within the uncertainties of Doppler calculations in the past.

F.

Ef fect of Beta Changes __on Reag_tivity Accident The effects of beta changes on the consequences of a reactivity accident were evaluated. The local ef fects resulting from adding the plutonium bundles to core were considered. The results of the investigation indicated that the reduction in beta resulting from loading the EEI bundles could result in a 1-2% increase in the core peak enthalpy. This effect was considered small in comparison to other conserv-ative features of the analysis so adjustment was not made to take this effect into account.

_ i G.

The bone dose is shown in the Big Rock Point Final Hazards Summary Report, Section 13, page 29, to be always less than 0.1 rem on all ccnditions of exposure time, location and diffusion regime. A conservative approach would indicate a site boundary dose increase in the same proportion as the plutonium inventory increase. As shown in A, above, the additional experimental plutonium will constitute a maximum of 27 percent of the core plutonium inventory, which added to 4 percent inventory of the 32 rods being irradiated under Change 18 to the Technical Specifications, constitutes a maximum of 31 percent of core inventory plutonium in experimental rods and bundles.

As plutonium accounts for about 25 percent of the dose due to non-volatila solids at the site boundary the total dose due to non-volatile solids would be conservatively increased a maximum of 8 percent.

The dose at the site boundary is still then calculated to be less than 0.1 rem.

H.

<a the reactivity of these bundles is less than those of the "E-G" Reload Fuel (see Table 1 of Proposed Tech. Spec. Change 19), the calculated severity of the hypothetical bundle drop accident described in the Final Hazards Summary Report, Section 12.9, page 21, is reduced. For the same reason, there are no additice.al criticality problems introduced when these Pu0 -UO bundles are placed 2

2 in the new fuel storage vault.

I.

The maximum pellet radial peak power generation in the highest power UO -Pu0 2

2 pellet is a nroximately 10 percent higher than average at beginning-of-life and decreases thereafter. This additional peaking on the pellet surface is mitigated by several facts.

1.

During a transient the temperature near the pellet surface is relatively cool due to the rapid heat transfer to the clad. Thus the extra energy generated on the surface is not reflected in temperature.

2.

Transient test data on like pins as described in Section B, above, use pellet average energy density as the measurement parameter. Thus, in making comparisons to the hypothetical excursions, the average radial enthalpy is used.

3.

The small increase in temperature is well within the uncertainty band of the excursion calculations.

e TABLE I: - SPERT TEST FUEL DESCRIPTION t-Fuel. Type GEXPR GEXPF.-CH Cladding l Material

---

Zr-2(10% Cold Worked)-------

Cladding ODJ(in.)

.3125

.3125 Cladding Thickness (in.)

.020

.020 Fuel Material Pu02 physically blended with natural UO2

' Fuel-Density (g/cm )

10.3 10.3 Pellet Center-Hole ID (in.)

0.

.114

- Fuel Enrichemnt 5.5%'Pu-239,.3% Pu-241 5.4% Pu-239,.3% Pu-241 Active Fuel 1 Length (in.)

5.2 5.2

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Fuel Pellet Diameter (in.)

.268

.268 TABLE II:

SPERT TEST RESULT,S i

Physical

' Energy Deposition Pressure Condition i

Test #

Fuel Type (cal /gm)

(Psi)

After Test 562 GEXPR 27.4 0

See Fig. 2,3 566 GEXPR 223 0

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707-G EXPR-CH 223 0

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'731 GEXPR-CH 277-0 See Fig. 6' i

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Report of the Nuclear Safety Program Division of Idaho Nuclear Corporation f

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2 Figure 2: Post Test Photograph of Test # 562. GEXPR Fuel Rod, 275 cal /gm Test. Photograph Courtesy of the SPERT Project, Idaho Nuclear Corporation i

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si Figure 3: Closeup Post-Test Photograph of Test #562.GEXPR Fuel Rod, 275 cal /gm Test. Photograph Courtesy of the SPERT Project, Ids.ho Nuclear Corooration

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l Figure 5: Post Test Photograph of. Test #707. GEXPR-CH Fuel Rod, 223 cal /gm Test. Photograph Courtesy of the SPERT Project, Idaho Nuclear Corporation.

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Figure 6: Post Test Photograph of Test #731. GEXPR-CH Fuel Rod, 277 cal /gm Test. Photograph Cour tesy of the SPERT Project, Idaho Nuclear Corporation.

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Figure 7 I

Plutonia Particle Distribution I

i EEI Fuel 100 1.00 Linearized Plot of Data 80 0.80 --

Mean Particle Size 38 Microns

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DATE RECEIVED

' DATE OF DOCUMENT Consumers Power Co.

Jan. 23, '70 2-2-70

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OTH ER.

REeORr, M EMO:

JacksoD, Michigan (Gerald W lke) x (Myr MyMRIZED) l OTHER:

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1 & 59 conformed cys. rec'd TO:

y CTioN NECESSARY O

CONCURRENCE O

DATE ANSWERED l

hO ACTION HELI 23ARY C COMMENT O

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POST OFFICE FILE CODES CLAint F RECEIVED BY DATE U

REG.NO:

DATE REFERRED TO (Must B. UnCtassified)

DESCillPTION:

Ltr. trans. tM foll&ving in support l

tiemann-2-2 of their Proposed Change No. 19 to w/9 cys. - F)R ACTION Tech Specs:

(i0 copies rec c) i 4

aNetosuREs:

Distribution-*

Additional Information re blended

-rec. f1Le cy.

1-D. ' thompson i

fuel transient behavior and 1-FDR Cop;/

1-DTIE j

Puo2-UO2 1-USIC particle sir.e characterization......

2-cW h ' e including Figures 1 thlu 7 1-00C (P A) 1-royd 1-Saltzman l-Skovholt 1-DuceMdine 1-H. Pri_ce_&_Stnff e

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