ML20058N303

Forwards Response to NRC 900717 Request for Addl Info

ML20058N303
Person / Time
Site:	University of Missouri-Rolla
Issue date:	08/09/1990
From:	Freeman D MISSOURI, UNIV. OF, ROLLA, MO
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9008140039
Download: ML20058N303 (18)

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/, j Nuclear Reactor Facility f II UNIVERSITY OF MISSOURI-ROLLA Nuclear Reactor Rolla. M0 Cb4010249 Telephone (314) 341 4236 f

j August 9, 1990 l Document Control Desk i U.S. Nuclear Regulatory Commission  !

Mail Stop 10-D-2),

Washington, D.C. 20555 Docket'No. 50-123 Dear Sirs!

Please find enclosed our response to your " Request For Additional l Information" dated July 17, 1990.

We hope this satisfies your information needs. Please contact me ,

at (314) 341-4236 if you require additional information.  ;

Sincerely, ltY David W. Freeman Reactor Manager DWF/lp.

Enclosure 1

xc: Alexander Adams, Jr., U.S. NRC Dr. A.E. Bolon, University of Missouri-Rolla l- A. Burt Davis, U.S. NRC, Region III l' Don L. Warner, University of Missouri-Rolla ,

Signed before-'me this 9th day of August, 1990.

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"~ 1TO REQUEST FOR' J.

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ADDITIONAL INFORMATION- .

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1. Question it Our question No. 4 from our-February 8, 1990  !

request for additional information has been misinterpreted. .

We believe that the 1.5% delta k/k insertion for the fuel  ;

handling accident is credible. Your introduction to the i accident in your 1984 Safety Analysis Report (SAR) along with the analysis in your HEU to LEU conversion application result in a complete analysis. Please resubmit the accident analysis using information contained on pages 9-13 and 9-14 through the middle of the second paragraph from your 1984 SAR i

and the information on pages 9-14 through 9-16 of your conversion submittal. This analysis should include table XV from your May 8, 1990 submittal.

Response to Ouastion 1.

The fuel handling accident analysis has been revised per your request. Information contained on pages 9-13 and 9-14 of the 1984 SAR has been combined with information presented on .

pages 9-14 through 9-16 of our conversion submittal. Revised Section 9.6 is presented in Attachment A. *

2. Question 2. Our question No. 5 from our February 8, 1990 request for additional information discusses the applicability of Part 100 to research reactors. .Part 100 is not applicable to research reactors. Your license renewal analysis was not deemed appropriate by the NRC staff. In ,

our Safety Evaluation Report (NUREG-1086) for the renewal of your_ operating license, the staff performed an independent analysis of the maximum hypothetical accident based on Part

20. _This was done because your analysis was based on Part 100. However, we agree with your assertion that this accident is not impacted by the conversion to LEU fuel. We requested the changes to make your SAR consistent with the regulations.

Response to Question 2.

Per your request, the analysis presented in Section 9.7,

" Failure of a Fueled Experiment" has been revised. The references to 10CFR Part 100 have been removed and calculated  ;

doses to offsite personnel are now compared with the requirements presented in 10CFR Part 20. The revised Section ,

9.7 is presented in Attachment B. -

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' Attachment A. Revised SAR Section 9.6, i

"He'.imum Reactivity Insertion" o

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f 9.6 Maximum Reactivity Insertion In this section mechanisms which could give rise to a large reactivity increase (such as p > 6 ) are analyzed -in order to I

identify the maximum reactivity insertion for which a safety analysis needs to be performed. The Technical Specifications I specif y the excess reactivity f or the UMRR as f ollows:  ;

The f uel loading shall be such that the excess reactivity above the reference core condition will be no more than 1.5% delta k/k, except that the excess reactivity may be ,

increased up to a maximum of 3.5% delta k/k for the ,

purposes of control rod calibraton only. This increase in excess reactivity above 1.5% delta k/k will be permitted no i more than twice a year and for no more than five '

consecutive working days each time. The reactor shall be operated only by a licensed Senior Operator when the excess reactivity is greater than 1.5% delta k/k.

In spite of extensive staff discussions and literature research no credible accident scenario has been f ound which could possibly lead to a sudden release of excess reactivity larger than 1.5% delta k/k. Therefore. an instantaneous insertion of the excess reactivity larger than 1.5% delta k/k has been excluded f esm f urther analysis.

Experiments at the Curtiss-Wright Research Reactor (11) have shown that the worth of a fuel element at the core periphery is less than 1.5% delta k/k. This is consistent with l

experience at the UMRR gained with different core configurations. Depending on its position at the core periphery a standard fuel element can be worth between 0.5% and 1.5% dslta k/k. (The reactivity worth of a fuel element within the reactor l

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core is not well known since the Technical Specifications

' preclude reuctor operation with an empty internal lattice 1 position.)

A hypothetical accident can be postulated assuming that a fuel' element has been placed next to a barely subcritical core, thus resulting in a positive step reactivity insertion of 1.5%-

delta k/k. A sudden reactivity insertion of such a magnitude j would cause the reactor to become prompt critical with a subsequent exponential power increase. The reactor period at the.

beginning of the prompt critical power excursion can be J approximately calculated (14) from the expression  ;

T4 - E--

Po -8 where'fp = prompt neutron lifetime (5x10-5 sec) 1 Rev. 4 l 8 = delayed neutron fraction (0.0065) i i

The reactor period corresponding to the above postulated  ;

I reactivity insertion is 6 msec.

In the analysis of short power excursions the total energy release and the resulting maximum fuel plate temperature are two of the most important physical parameters. A transient code Rev. 4 1 PARET described in the previous section has Leen used to analyze this accident. Its results can be summarized as follows:

Initially, the reactor is assumed to be at some low power (1W).

After the postulated reactivity insertion the power increases I

steeply until a peak power of 588 MW is reached at about 0.144

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9 sec. At this time energy released during the transient amounts Rev. 4<

to 5.4 MWs. A strong negative feedback caused by moderator solding (about -O.70% delta k/k) and the Doppler effect (about

-0.16% delta k/k) reverses the sign of the period and brings the

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reactor to delayed supercritical. From this time on, therefore, the reactor power rapidly decreases and the transient quickly i

dies away.

The maximum fuel centerline temperature is not i reached until about 0.16 sec. At this time, it amounts to 4350 C ,

which is still distinctl:t below the melting temperature of cladding. In the reactsr average channel a fuel temperature of  ;

2450 C is indicated. Subsequently, both temperatures start to '

decrease with the reactor being now subcritical with keff = i 0.9981.

The results of the theoretical analysis are supported by a i large collection of data from the excursion experiments performed i at the BORAX-and SPERT facilities (17). Especially, some of the SPERT-1 experiments using the DU-12/25 core are applicable to the UMR Reactor since the fuel geometry and composition are very  !

similar (18). A detailed comparison is given in Table XV. '

In the series of'SPERT-experiments, there was an experiment in which the induced reactor period was 6 ms. The total energy released in the excursion was 13.2 MWs. Onset of the self-limiting mechanism occurred when about 7.2 MWs of the l

L thermal energy was generated. No damage to the fuel cladding was

' observed (19). From the experiments it was concluded that the

= mechanism responsible *.'or self-limiting the power excursion l

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consists of fuel and moderator thermal expansion and boiling.

Rev. 4 (The latter being the dominant shutdown mechanism.) This finding is consistent with the results of our theoretical analysis.

• Table XV. Comparison of Important Fuel Data UMRR UMRR SPERT-1  !

Geometry. Plate , Plate Plate Rev 4 l HEU LEU i

Length (cm) 61 61 61 -

Width (cm) 7.6 7.6 7.6

-Thickness (cm) 0.15 0.127 0.15 Water gap (cm) 0.49 0.31 0.45 Fuel Material U 38 0 -Al U 3 Si 2-Al U-Al Enrichment (%) ~90 19.8 100 Weight fraction of U 0.36 0.48 0.24

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Thickness (mm) 0.51 0.51 0.51 Cladding Material A1 6061A1 Al Thickness (mm) 0.51 0.38 0.51 Both the theoretical and experimental analyses have shown pey, 4 that the above-postulated accident can safely be terminated by a self-limiting shutdown mechanism. This is a rather surprising result. However, the short time constant of the thin fuel plates allows a large amount of energy to be transferred into the water

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.'* 9 17 channels even during very short reactor periods. Consequently, boiling (together with the Doppler feedback) becomes the rapid Rev. 4 and dominating shutdown factc'. Such an accident can, therefore, be terminated even if the safety instrumentation, e.g. both power safety channels, were ir. operable.

In spite of this UNRR inherent safety feature, the following administrative steps have been established in the Standard Operating Procedures which are designed to prevent a fuel handling accident:

(1) All fuel handling is done in accordance with written procedures.

(2) Loadings are planned to include the sequence of loading and positions of individual elements. Also a loading schedule is prepared prior to commencement of loading, j l

i (3) Loading operations are done by qualified personnel under the

-direct supervision of a licensed Senior Operator.

-(4) Fuel handling tools are kept locked with the keys secured to prevent unauthorized movement of fuel.

i (5) Loading of the core is done from the inside to the outside.

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4 9-28 Finally, it should be pointed out that the assumptions leading to this accident are very unlikely, and therefore it is not believed that such an accident would ever happen. The analysis, however, has been quite useful in showing.the inherent safety potential of the UMRR. Therefore, no effects on the 4

health and safety of the public nor on the reactor staff are to be expected from this type of accident.

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J Attachment B. Revised SAR Section 9.7, ,

" Failure of a Fueled Experiment" i

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S. 7 Failure of a Fueled Experiment l J

1 In this section an analysis is performed to assess the )

l hazard ' associated with the failure of an experiment in which fissile material has been irradiated in the reactor. In the scenario of this accident it is assumed that a capsule containing irradiated fissile material breaks and a portion of -

t; f i ssi on product inventory becomes airborne. The i consequences of the release are analyzed for both the reactor .

f staff and general public. Since the potential impact of this postulated accident is greater than in any other accident analyzed, the failure of a fueled experiment is designated as the maximum hypothetical accident for the UMRR.

The limiting . criterion used in the analysis of a fueled ex peri ment is the power generated within the irradiated fissile materi al . In this anal ysi s the consequences of a failed experiment generating 1.W and 100 W, respec ti vel y, were studied.

The, fission products expected to become airborne, in=the case where the experiment capsule wrire to lose its integrity, are noble gases and elemental iodine. -Other fission products and actinides are not vol ati l e at the temperature (which is l

esuentially at room temperature) at which the f ueled experiment would be performed. The amount of noble gases and radiciodine is -assumed to be that specified in (13), i.e.: 100% of the noble gaser and 50% of the iodine inventory. If the experiment

(' were to be run in the reactor pool a credit for the absorption of iodine in water can be taken (14). This par ti ti on factor

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9-20 i amounts to 10, i.e. only about 5% of the total iodine inventory i

would reach'the Reactor Building atmosphere in an accident. '

i A conservative assumption was made in the analysis in that the irradiation time was considered to be 1.1 finite. Theref ore, t c the fission inventory used in the analysis represents for some long-lived radionuclides, e.g. Kr-85, most likely an overly ,

-conservative value. Fur ther more , it was _ assumed that the fission products are instantaneously released and uniferaly distributed in the Reactor Building air. The free volume of the Reactor Building is approximately 1.7 x 10* m.

The external dose rate (in mrem /hr) due to Y- and B-radiation was calculated using the relationships given in (15) 6 Y

= 9.43 x 10'*x XxE Y where X = radionuclide concentration (Ci/cm')

• E = average y-energy per disintegration (MeV)

T and bg = 8.24 x 10*

• x X x -E g where~E = average B-energy per disintegration (MeV). The dose 3

rate = to the thyroid ssn rem /hr) due to the inhalation of 1 radiciodines is given by 0 T = DCF M BxX l where DCF = dose-conversion f actor f or the thyroid (rem /Ci) I l

B = breathing rate (cm*/hr) l k X = radiciodine concentration (Ci/cm')

The standard breathing rate recommended (14) is 1.25 x 10' ' cm'

/hr. The thyroid dose-conversion factors are given in Table XVI. Rev. 4 f

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i Table T XVI. Iodine Dose-Conversion Factors f or the Thyroid (14) Rev. 4

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Isotope DCF (rem /C1) 1-131 1.0 x 10' {

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1-132 6.6 x 10 I-133 1.8 x 10' I-134 1.1 x 10 8 I-135 4.4 x 10' l

The calculated saturation activity for e,ach respective--

radioisotope and its concentration in the Reactor Building af ter experiment failure is chown in Table XVil f or the experiment Rev. 4 ;

power of.1 W. Also shewn it his table are the associated T-and 6- radiation energies together with calculated dose rates for the.whole-body, skin, and the thyroid. With a T -dose rate in the reactor building as high as 250 mrem /hr any one of

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radiation area monitors would cause an automatic reactor shutdown, audible and visual alarms in the control room, and in addition the reactor bridge monitor would activate the building evacuation alarm system. From the past experience, i t is known that the reactor building can be evacuated within 3 minutes.

For the purpose of this analysis it is assumed that the time '

elapsed between the release of radioactivity and the end of evacuation is 5 minutes. Therefore, it is assumed in the

. f ollowing calculation that the exposure time to the members of the reactor staff is 5 minutes. The resulting radiation doses are: whol e-body dose 20. 6 mrem, skin dose 11.21 mrem, and 'the thyroid dose 0.93 rem.

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Table XVII. Rev. 4 '

Dose Rates in the Reactor Building from a Failed Fuel Experiment (Power = 1 W) 3 3g m ISOTOPE A sa (Cl) E (MeV) 5,(MeV) X,(Ci g) 3 (m ) ) )

I-131 2.45 E-2 3.71 E-1 1.97 E-1 7.20 E-12 2.52 E O 1.17 E 0. 9.00 E 0 I-132 3.71 E-2 2.40 E 0 4.48 E-1 1.09 E-11 2.47 E 1 4.03 E O 9.00 E-2 1-133 5.48 E-2 4.77 E-1 4.23 E-1 1.61 E-11 7.25 E 0 5.60 E O 3.62 E O I-i34 6.06 E-2 1.94 E O 4.55 E-1 1.78 E-11 3.26 E 1 6.65 E O 2.45 E-2 1-135 5.06 E-2 1.78 E 0 3.08 E-1 1.49 E-11 2.50 E 1 3.78 E O 8.16 E-1 Kr-83m 5.90 E-3 2.60 E-3 1.01 -2 3.47 E-12 8.51 E-3 2.95 E-2 Kr-85m 1.27 E-2 1.51 E-1 2.7 _-1 7.47 E-12 1.06 E O 1.37 E O Kr-85 2.53 E-3 2.11 E-3  ? E-1 1.49 E-12 2.12 E-1 2.74 E-1 Kr-87 2.00 E-2 1.37 E O ss E O 1.18 E-11' 1.52 E 1 1.02 E 1 Kr-88 3.12 E-2 1.74 E O 3.41 E-1 1.84 E-11 3.02 E 1 5.17 E O  ?

Kr-89 3.96 E-2 1.60 E 0 1.33 E O 2.33 E-11 3.52 E 1 2.56 E 1 y Xe-131m 2.53 E-4 2.00 E-2 1.40 E-1 1.49 E-13 2.81 E-3 1.72 E-2 Xe-133m 1.35 E-3 3.26 E-1 1.55 E-1 7.94 E-13 2.44 E-1 1.01 E-1 Xe-133 5.48 E-2 3.00 E-2 1.46 E-1 3.22 E-11 9.11 E-1 3.87 E O Xe-135m 1.77 E-2 4.22 E-1 9.74 E-2 1.04 E-11 4.14 E O 8.35 E-1 Xe-135 5.23 E-2 2.46 E-1 3.22 E-1 3.08 E-11 2.90 E 1 8.17 E O Xe-137 5.31 E-2 1.50 E-1 1.37 E O 3.12 E-11 4.41 E 0 3.52 E 1 Xe-138 5.57 E-2 1.10 E 0 8.00 E-1 3.28 E-11 3.40 E 1 2.16 E 1 Total 2.47 E 2 1.34 E 2 1.36 E 1 J

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In Table XVIII dose rates in the reactor building from a Rev 4 I i

failed fuel experiment generating 100 W are shown. It was )

assumed.that this experiment would be run at the reactor core i

'within the water pool. Therefore, as discussed previously, in ,

the calculatinn of the iodine concentration in the reactor building air a retention factor of 10 was assumed for the reactor-pool. Aeuumb.g the same evacuation time as above, the respective j rad \ation doses to the staff members are calculated: whole-body l ;

1.37 rem, skin 0.96 rem, and to the thyroid 11.3 rem. l i

For the radiation calculations outside of the reactor building it was assumed that all fission products released in the reactor building would leak out within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Since the reactor building does not have any windows and has only a'few -

openings such as the ones for fans, air conditioners, etc., which could be readily sealed from outside in the case of an emergency, the assumption about the leak rate is considered to be conservative. Another conservative assumption was made in that _

no radioactive decay and hence no decrease in the source strength  ;

. was taken into account while calculating the dose rates outside the reactor building. The radionuclide concentration just Rev. 4' L outside of the reactor building was calculated using the building wake dispersion factor of 2.0x10-2 sec/m3 (discussed in Section

! 7.6).  ;

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Dose Rates in the Reactor Building '

From a Failed Fuel Experiment (Power = 100 W)

ISOTOPE X 8 (C1/cm )

3 3 (mrem) n, g-(mre) g nr bT (r my I-131- 7.20 E-11 2.52 E 1 1.17 E 1 9.00 E 1 ,

I-132 1.09 E-10 2.47 E 2 4.03 E 1 9.00 E-1 j I-133 1.61 E-10 7.25 E 1 5.60 E 1 3.62 E 1  ;

I-134' 1.78 E-10' 3.26 E 2 6.65 E 1 2.45 E-1 I-135 1.49 E-10 2.50 E 2 3.78 E 1 8.16 E O ,

Kr-83m 3.47 E-10 8.51 C-1 2.95 E O Kr-85M 7.47 E-10 1.06 E 2 1.37 E 2  !

l Kr-85 1.49 E-10 2.12 E 1 2.74 E 1 Kr-87 1.18 E-9 1.52 E 3 1.02 E 3  ;

g , Kr-88 1.84 E-9 3.02 E 3 5.17 E 2 Kr-89 ,2,33 E-9 3.52 E 3 2.56 E 3 Xe-131m 1.49 E-11 2.81 E-1 1.72 E O Xe-133m 7.94 E-11' 2.44 E 1 1.01 E 1 Xe-133 3.22 E-9 9.11 E 1 3.87 E.2 Xe-135m 1.04 E-9 4.14 E 2 8.35 E 1 Xe-135 3.08 E-9 2.90 E 3 8.17 E 2 Xe-137- 3.12 E-9' 4.41 E 2 3.52 E 3 Xe-138 3.28 E-9 3,40 E 3 2.16 E 3 Total 1.64 E 4 1.15 1 4 1.36 E 2

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9-25 The whole body dose:to un individual located just outside Rev. 4 the reactor building was calculated using the methodology presented above for the same failed fuelen experiment scenarios.

The resulting whole-body dose associated with the 100 watt fueled experiment was 242 mrem. (The whole body dose sssociated with the 1 watt fueled experiment was lower by about twv orders of magnitude.) This dose is a factor of two lower than the 10CFR Part 20 annual limit of 500 mrem exposure to individuals in unrestricted areas.

It is concluded that experiments using fissile material can be irradiated at the UMRR within the power limits analyzed in this section.

. hare is no undue hazard to the general public nor to the reactor staff in the very hypothetical case of a' failed experiment as postulated'and analyzed above.

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