ML20011D611

Forwards Responses to Items 1,2 & 3 of Re High Enriched U/Low Enriched U Conversion at College.Anticipates Shipping Spent Fuel as Scrap to ORNL Using 6M Shipping Containers

ML20011D611
Person / Time
Site:	05000199
Issue date:	12/12/1989
From:	Berlin R MANHATTAN COLLEGE, RIVERDALE, NY
To:	Michaels T Office of Nuclear Reactor Regulation
References
NUDOCS 8912280095
Download: ML20011D611 (20)

Text

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t anheNan WANHATTAN COLLEGE PA4 CAY WEcHANICAL EwelNEERING DEPARTWENT aivEmpatE.NEw vonx ipn

,hhh crie) er>osas 9-M December 12,1989 Mr. Theodore Michaels, Project Manager Non Power Reactor, Decommissioning and Eisvironmental Project Directorate Division of Reactor Projects - II,IV, Y and Special Projects Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Cornmission l

Washington, DC 20$55

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SUBJECT:

Response to Questions Regarding HEU/ LEU Conversion at Manhattan College.

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(Your letter of October 30, 1989) i

Dear Mr. Michaels:

t Enclosed are responses to items 1. 2,and 3 of the enclosure te your letter of October 30,1989 relative to the conversion from HEU to LEU fuel in the MCZPR.

As I indicated in our recent phone conversations, we are anticipating shipping the spent fuel j

as scrap to ORNL using the 6M shipping containers in which the new LEU fuel elements l

will be shipped from Babcock and Wilcox. B&W has told the DOE that the current projected

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date for availability of the LEU fuel is approximately the end of December. I will keep you informed as to updates in the schedule.

)

t Since cly, f

Robert E. Berlin Reactor Administrator ec:

ROC Members i

Dr. J. Lestingi

()

f 8912200095 89)212 PDR ADDCK 05000199 P

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desconse"toItame 1. 2. and 3 of Enclomore to NRC Letter of October 30;'1989 4-Item 1.

Fuel Londing Plan-Before insertion of the new LEU fuel into the core area, the reactor tank and. all :its ~ inside components will be cleaned and-e:camined, and approximstly.

4500 gallons deminerali :ed water will then be injected into.the tank. A one-curie Pu-Be neutron source will be inserted into the source holder et the core for control' console instrut+ntation= checking. -Three calibrated portable detectors used for neutron, bets-part.ielo and ganna rays estimation uP j be p:epared for backcround manitnring and fuel loading. In order to avoid eai abrupt chance in core reactivity as we]) as power level, th+ fuel incerU $

> order has been propcsed es r.h:wi in Ficurel.

Control console meter readin2s win ba made during core conve:w$on, and e.11 outcomes wi31 serve &c.the tesis for. initial. ' eriticality. estitstion (critical mass estimation).

The detailpi procedure for fuel 3cading and criticality test are as follows:

1 J hat. h rnose To lo:d new LEU feel int:.he core to the mininum eritica] mstr in L raf-and efficient menner, and increase critical lodding te, the' u:minut2 al kehl- -

e:: cess reactivity-of. 0.44% ek/k.

L. Summary 2f T3At Fuel elements will be'loahd incrementally, as shownfin Figure 1.

Counts will be taken at each individual fuel element loading for all rods in the full-

~in position. and at-the 50% and 100% out positions.. The reciprocal of these counts will be plotted e gair.st the number:-of fuel elements:in the core end; a

extroplated to predict the expected critical-loading.. This [.rocess will be continued : until criticality is achieved.

Upon _ reaching _ criticality,the regulating Jrod -will be calibrated using the positive period method.

Ur.ing--

this calibrated rod the loading will, then be-adjusted ito the maximum e.@ese reactivity of 0.44% Ak/k.

.t 3,, Prereauicite Onerations A. General

'The following operations and > conditions will" be fulfillsd beibre loading the initial. fuel elements.

No deviations from.-these. conditions wil]-be' accepted.'

1. Tests to determine the instrument response to neutrons, gama-rays, brid-the ' blackness of control rods will have been completed and the re.wPs

'found'to be satisfactory.

2. All-conditions as list % in Table 1,

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1. All required _ check-out procedures will-have bee 6 completed, withithe W

reactor and involved instrumentation determined to be in proper operating con'dition. These check-outs shall include:

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a. Physical check procedures.

b. Console check-out procedures.

2. Verify that the startup channel instrumentation is adjusted properly, with respect to the minor change of core geometry (replace HEU full fuel elems, at

• 46 in core by the LEU full fuel element at $814).

C. Source A one curie Pu-Be neutron source will be in the source holder in' core posil: ion S-1 and positioned vertically so as to give a count rate of greater than 2 eps.

D. Health Physics Equipment The minimum portable health physics equipment available-for initial start-up and power operation is listed in Table 2.

4. Prentions l

A. General The test for initial criticality is the most important phase of-the operational test procedure because most of the variables effecting the outcome of the test have been predicted analytically (by the ANL) rather than

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experimentally.

In view of a possible discrepancy between measured and i

calculated critical loading, and uncertainty in the theoretical results for the core, the critical test shall be undertaken presuming that the resultant critical loading is relatively unknown. The initial loading will be made with caution.

The reactof behavior after each fuel loading step will be analyzed for nearness to criticality before proceeding with the subsequent step.

B. Fuel Elements The new fuel elements to be inserted to achieve reactor operation are to j

be inspected for damage to the channels or defects on the surface that may have occurred during storage and shipment to the site.

Of particular importance is a check for evidence of corrosion or blistering on the cladding surface.

This pre-loading inspection is important, since it will help preclude the possibility of fuel element' failure during subsequent operation.

Each fuel element should always be handled with clean lint-free gloves and with extreme care..

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Where the loading procedure is interrupted for' any'significant period of

time, the readings of the previous element loading will be retaken prior to proceeding with additional loading. This will check for any failure or drift in the start-up instrumentation.

D. By-pass of Scram Circuits The following scram circuits must be bypassed in order to obtain control power:

1 1.

A bypass to eliminate a reverse as a consequence of the gamma recorder being down scale. This bypass will be utilized during startup until the gamma recorder reads on scale.

2.

A bypass to eliminate a reverse as a consequence of the linear recorder being down scale. This will be, utilized in the initial fuel loading until the linear recorder reads on scale.

E. Safety of Personnel During the initial approach to criticality any precautions deemed necessary by the reactor supervisor will be observed to insure the safety of personnel and equipment.

All personnel associated with the experiment will have been briefed as to procedures to be followed in the event any emergency shall occur.

4

5. Initial Reseter Conditions l

l

1. Reactor water at normal operating level (7 feet high).

2. Reactor water at normal operating temperature range (60-80*F).

3. Reactor water purification system operable.

4. Reactor core empty of any dummy and live elements.

5. Reactor room ventilating system operable.

6. Neutron source in position S-1.

7. Two control rods in full down position.

l B. Procedures l

i

1. Recheck positions of source (S-1), UIC (I-2), and BF3 (I-1).

2. Carefully load the first fuel element into position as shown by Figure 1. ;

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4. Take th'ree, one-minute' counts on each' start-up' counting' circuit.~Re~co~rdic.V f

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5. Withdraw both' cont'rol rods full-out (100%),' and wait one-mh, ute.

6. Repeat step 4.

7. Insert both control rods full-in and repeat step 4.

8. Load a new fuel element into the next scheduled pc-ition (Figure 1).

9. Repeat steps 3, 4, 5, 6, and 7.

10. Plot the reciprocal of the count rate readings versus the number of fuel elements in the core. Draw lines through each set of points plotted.

The intersection with the X-axis gives an approximation of the critical loading.

11. Continue loading fuel elements in the sequence listed in Figure 1, repeating steps 8 through 10 after each fuel element is added, until criticality is approached. The Reactor Supervisor will then determine from a study of the reciprocal count plots whether to proceed with the i

loading of an additional full fuel element on the next loading. In case 1/M does not behave as analytical predicted, or it is anticipated a supercritical condition may occur due to a full fuel element insertion, a partial fuel element with all movable fuel plates will be used in lieu i

of a full fuel element on the next-loading.

12. After criticality is indicated, level off the neutron flux by insertion of the control rods.

13. Check exact criticality by removal of the source after the flux level has been stabilized by the control rods.

14. When exact criticality has been obtained, record the control rod positions abcurately (fine position-indicator), also record readings of the Log Count Rate Meter and the Linear Channel.

15. On reaching initial criticality the MCZPR regulating rod will be calibrated using the period measurement technique.

16. Using the calibration of the stainless steel regulating rod the excess reactivity of the initial critical loading will be determined.

17. Adjust the core loading by using partial fuel element to li2 nit the excess reactivity to a value of 0.44% Ak/k.

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1. Core loading records lCih.oluding fuel element' serial" numbers 1andecor_e-_

positions;l detector' serial numbers,-- type and-locations

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2. Initial start-up records, ' including run numbers, instrument ri$ lings,

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reciprocal count rates, incremental fuel mass, total fuel mass'for each loading, control rod positions, loading time, reciprocal count curves, and other such data as may be required at the time of initial start-up, i

8. Personnel during Start-up

1. Health Physicist

2. Radiation Safety Officer

3. Chief Reactor Supervisor

4. Reactor Operators

5. Technical Support Engineer /Technicien (Note: Our Health Physicist is the only person who will be present who participated in the initial HEU core loading and start-up test on March 24, 1964, However, the Chief Reactor Supervisor has participated in other loadings of the HEU core after periodic maintenance. The MCZPR Technical Support Engineer / Technician was previously employed in refueling of reactors in the Naval Nuclear program) l l

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1. Eater purification and recirculating pump OPERABLE.

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2. UIC in position I-2 as indicated in the core diagram (Figure 1).

3. S. tart-up source in source holder and positioned as-indicatedLin the core '

position S-1 (Figure'1).

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4. BFg chamber in core position I-1.(Figure 1).

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Table 2 Item Ha

1. Slow and fast neutron rurvey meter and counter.

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2. Cutie-Pie detector

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ofusinfasu6 critical'schemeto' determine-thereactor~ power s,

In Method I level, the source should$always stay "in" core to provide the necessary reduce,since the removal of,the neutron source neutron flux l'or power estimation, the power level and' thereafter-'an-(sourceless) will -immediately unpredictable transient power would complicate the suboritical experiment.

To assure that the maximum licensed power level of 0.1 watt will not be exceeded during core conversion, a well-established sourceless critical experiment will be employed (see attachment I).

In this experiment, the last step of reaching criticality is to remove the source "out" of the core (sourceless),

and the power level will be determine from the indication on the picoammeter based on the ratio of obtained power scale to the licensed power setting of 0.1 watt at 5.0x10~' scale on picoammeter (this setting was labeled by the NRC in 1964 ).

In hethod II, all the indicated typographical errors have been corrected, and all the equations with specific constants are rewritten and explained as follows:

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a q As CsR o.0 23

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is the average thermal flux seen by the gold (Au) foil.

where A (t ) is the absolute activity per gram of bare gold foil at time tg 0g with unit of disintegration /see-gram, f"d is the gold (Au) thermal neutron activation cross section at 2200 meters /seo, with value of 85 barns.

Cg R is the Cadmium Ratio and is equal to A /Aa, and here Ag is g

the absolute activity per gram of bare gold foil, and Aes is the absolute activity per gram of Cadmium-covered gold foil.

At.Wt. is the atomic weight of gold (Au) with the value of 107.

0.6023 is the Avogadros number. The reason we use 0.6023 instead of using 0.6023x10*"

theconsiderationthatwecancancel10'gsnumberisbasedon to represent Avogadr with the value of cross sectio *n (barns) of 6-d" in equation 1, where 1 barn is equal to 10 cm*,

act h

is the decay constant of gold ( Au), and is calculated by the equation of A =1n 2/Ty,, =0.693/Ty,. Here T g is the half-life of gold (Au) with value of 64.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (233,280 seconds),

t is the delay time between exposure and counting of gold (Au) g foil, with unit of second, t*

is the duration time of exposure of gold (Au) foil, with unit of second. 4

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is the total mass of U-235 in core at time gold (Au) foils are irradiated, with unit of grams. For the HEU core, the total mass of U-235 in core is.about 3,024 grams.

p is the average thermal neutro'n flux in core. For the HEU 8

core, 6,rt is about 1,000,000 neutrons /seo-cm. The relation-cert between the average lux in core and the flux of gold (Au) foil is fe,re : 4 x0.895, where 0.695 is the correction factor for the neutr n flux estimation in the MCZPR core.

P /P is the ratio of total power (i.e. maximum powep = 0.125 watt) r d to the power result'ing only from thermal fissions, and the value of this ratio is 1.20.

3.54x10'" in equation 2 is actually the proportional constant between the power and the product of parameters in the right-hand-side of equation 2; that is, we can obtain this constant based on the following calculation:

0.125 (watt) = 3.54x10 " x 3,024 x 1,000,000 x 1.20

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(C R), all the other parameters such as power Except for.the Cadmium Ratio d

peaking factors and power distribution have been estimated by the ANL, using the LEU reference core (see attachment II) with both Monte Carlo and Diffusion

- Theory codes. The Cadmium Ratio, however, will be reevaluated in terms of neutron flux at the specific location in the LEU core, since the new LEU core configuration will be different from that of the original HEU core..The core configuration-change was suggested by the ANL based on the consideration of obtaining a more symmetric core geometry; due to the replacement of fuel #46 in HEU core by. fuel '414 in LEU core, more effective fission in the fuel elements and hence a more uniform thermal neutron flux distribution in the core can be expected.

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d TTACH MEAIT I

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i APPROACH TO CRJTICALITY i

1 INTRODUCTION j

Since the use of a nuclear reactor as an experimental machine requires a thorough knowledge of its physical makeup, the first part of this experiment will be devoted to a description of the reactor components and equipment and a visual inspection _ of the facility. A daily reactor checkout will then be performed to 1

introduc'e the operation and safety aspects of the reactor. The reactor checkout includes a' step by step check of the reactor conditions, the neutron monitoring instruments and the scram circuits.

The concepts involved in an approach to critical with the fine control rod or regulating rod will be brought out in the latter part of the experiment. The approach to critical will be performed with the' intention of showing the participant j

what is occurring in the reactor core during startup. The experiment will actually simulate the last steps of a criticality experiment, which is the basic experiment performed to determine the critical mass of a new system. The final result of the experiment will yield the position of the fine control rod at which the reactor -

becomes a self sustaining neutron chain reacting system.

II.

THEORY When a neutron source is placed near a suberitical reactor, the fissions induced in the fuel multiply the source neutrons entering the core. The multi-plication is defined as the ratio of the total thermal neutron flux,- due to source.

neutrons and fission neutrons to the. flux due to the source alone. The degree of multiplication is dependent upon the magnitude of the reproduction factor of the assembly. The equ& tion relating k,ff and multiplication is readily derived from the sketch below where Q neutrons were introduced each second.

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The multiplication, M, of the source neutrons as previously defined equals i

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i Since this expression has been derived for a suberitical reactor, k,ff c1, the expression for the multiplication may'be rewritten after cancelling out the Q l

1 M=

1 - K,ft It can be noted from this equation that as k,ff approaches one, the multiplication approaches infinity and 1/M would approach zero.

The experimental arrangement for the approach to etitical on.the fine control rod can be seen in the diagram below, Fine Control Rod WBe BF.3 Neutron Detector Neutron Source The measurement of the neutron population at the detector location can be assumed

' to be proportional to the thermal neutron flux in the reactor core. With the fine control rod at the edge of the core, the reproduction factor equals (k,ff g.

Let C

equal the measurement of the neutron population by the detector. If the rod o

is moved into the core to the desired position, the reproduction factor becomes (k,gg)1 and the new neutron population measurement C.

With all neutron measure-i ments normalleed to C, the multiplication at any rod position is given by o

A linear plot of the inverse of the multiplication versus rod M = Cg/Co.

i position can be extrapolated to a 1/M value of hero where the fine control rod is in its critical position. The linear extrapolation is based on the assumption that the change in the reproduction factor per centimeter movement of the fine control rod is constant. This assumption has been found to be true only after the rod has been moved by approximately 15% of its total travel.

When a reactor is made supercritical, K,ff pl, the neutron population will begin to rise to higher levels. This rise in neutron level can be described as a summation of exponentials.

a

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$(t) =

Aeg 1=o Where the A 's are constants determined by the initial conditions of the g

reactor and the c)g's are the m + 1 roots of the fundamental inhour equation.

The rate at which the flux rises is dependent upon the magnitude of k,ff, the amount that k,ff is above unity.

All the terms in the expression above except the first, have negative exponentials when positive reactivity is added, and their contribution to the flux decreases rapidly to zero after a short time has elapsed.

This summation then reduces to, gg f 0 '='- fo 0

[, = flux at t = 0 where u.), = the inverse of the reactor period The positive asymptotic period T = 1/v.)o, can easily be determined from a plot of neutron population versus time. The period is defined as the time it takes the neutron level to increase by a factor of o, after the transient terms have died out. Another important parameter is the doubling time.

This is defined as the time it takes the neutron level to double. It is related to the period by the following expression:

d J2 where t2 is the doubling time T=

2n2 It is possible to determine the reactivity (k ff-1)/K,ff of the assembly which corresponds to a certain measured positive period through the use of the basis in'-hour equation. This equation, derived from the time dependent diffusion equation, involves the delayed neutron parameters and the thermal neutron lifetime of the reactor in question.

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When loading fuel into a reactor, one should expect to obtain results such as tbn ones chown in the figure below.

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Weight of Fiscionable Material Added A - Non-conservative approach. ' Detector too close to source.

May cause a dangerous overestimate for next loading.

B - Over-conservative approach. Detector too'far from source -

shadowed by fuel elements and other materials.

I-Ideal approach 5

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, Proceddre:

I In the actual experiment, approach to criticality is attained by gradual removal of a control rod rather than by adding fuel to the reactor' core.

Remove the source and take a five minute background count on the scaler. Replace the source.

Remove the enim rod to some position which has been pre-determine'd by the instructor. Take a five minute count on the scaler. The number of counts per minute (minus background) will be N.

o Remove the reg rod 10% and take a one minute count (N). N /N o

will be the reciprocal of the multiplication factor. On standard cartesian coordinate graph paper,. plot N /N vs. percentage of reg rod withdrawal.

o Using a straight edge make a projection to the zero axis of No/N. Where does this line strike the zero axis ?

Remove the reg rod 20% and take a one minute count (N). Plot N /N on the graph paper. Set the best fitting railroad curve to pass through o

the three points on the graph paper and extrapolate to the zero axis. Where does the extrapolated curve strike the zero axis ?

Continue the procedure described in the last paragraph, gradually withdrawing the reg rod and using the best fitting railroad curve on the last three points plotted.

Always wait for a steady state condition before taking a reading.

Why ?

Be sure to take one reading with the reg rod withdrawn 50%. Why?.

When the ' reactor is almost critical, all the students will try to guess the critical position of the reg rod. The instructor will then bring the reactor to criticality and remove the so'urce.

Dis cus sion A graph is enclosed for the r' elative reg rod worth. How do you explain the shape of this etTve ?

How close was your estimate for critical position of the reg rod to the one actually obtained?

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RELATIVE REG ROD WORTH.

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0. 0 1.000

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i A' T TA C HME N T

.1 LEU CORE Power / Element, milllWatta 5.4 6.1 4,9 Peak Powe. In Element /

l 32 'l l 43 l l 54 l Average Power in Core Fuel 1.94 1.99 1.go l

l 6.7 8.7 8.0 5.1 l 22 l l 33 l l 44 l M

2.24 2,58 2.52 1.73 5.4 8.5 9.1 6.9

{ 12 l l 23 l l 34 l l 45 l 1.80 2.56 2.59 2.21 l i j

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5.8 7.1 6.3 l 13 l

,l 24 l l 35 l l

2.02 2.20 2.16 4.2 1.0 146 1 1251

~~

1.61 0.26 1

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11tm 3. Emergejter Shutdguut While the emergency shutdown rod is currently located on the wall of the reactor facility within arms reach of an operator positioned on the reactor platform to permit removal and insertion in one rapid motion, we will reposition the rod during the initial core loading to r

be suspended in the reactor tank outside the core. This will permit the rod to be inserted even more rapidly, should this be required, i

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