ML20151E623

Forwards Response to NRC 880610 Request for Addl Info Re 860912 Application for Amend to License R-103,supplemented on 870911 & 880311,authorizing Use of Extended Life Aluminide Fuel Element Contining Higher U Densities

ML20151E623
Person / Time
Site:	University of Missouri-Columbia
Issue date:	07/22/1988
From:	Brugger R, Mckibben J MISSOURI, UNIV. OF, COLUMBIA, MO
To:	Alexander Adams Office of Nuclear Reactor Regulation
References
NUDOCS 8807260146
Download: ML20151E623 (8)

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Research Reactor Facility UNIVERSITY OF VISSOURI July 22, 1988 Research Park Columtna Missoun 65211 Telephone (314) 882-4211 I

Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Station P1-137 Washington, D. C. 20555 l l

ATTENTION: Mr. Alexander Adams, Jr., Project Manager '

Standardization and Non-Power Reactor i Project Directorate l

REFERENCE:

Docket 50-186, License R-103 I University of Missouri l SUBJ ECT: Response to NRC Reque.st for Additional Information l

Dear Sir:

We have completed our response to questions 1 and 2 in your letter dated l June 10, 1988 concerning our application dated September 12, 1986 as supple- l mented in letters dated September 11, 1987 and March 11, 1988 for an amendment l to our R-103 operating license. This anendment would authorize the use in our i reactor of a newly developed extended life aluminide fuel (LEAF) elemdnt con- l taining higher densities of uranium and a burnable poison.

The two questions and our answers are attached. We had completed the answers to questions 3 and 4 usirl our original a,deling techni- ,; but in double checking our computer out ,ut, we noted the code as modeluo computed different pressure drops across the two channels in the model. In an attempt ,

to check the sensitive of the results to how the flow and pressure drop rela- l tionship is modeled, we changed the input for this relationship. The code would not converge on a flow rate to match the pressure drops across the two channels in our model. We are going to work with W. Woodruff of the RERTR group at Argonne National Laboratory (ANL), who wrote the PARET/ANL version of the code, to correct this problem in the PARET code. He was on vacation when the problem was discovered and will not return till the end of July. Until he returns from vacation, we will not be able to estimate how long it will take to resolve the problem. When we resolve the problem with ANL, we will inform you of what we have found and give you an estimate of how long it will take to complete the answers to questions 3 and 4, and any additional reanalysis that may be required due to the change.

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'I Director of Nuclear Reactor Regulation.

July 22,,1988 Page 2 r

-If you have any questions concerning this, please contact me'at (314)-882-5204.

Si ncerely, W

/ }-- k J . $r Kibben P.eactor Manager Reviewed and Approved: l /

[O F W gr Robert M. Brugger Director xc: Mr. A. Adams

. Subscribed and sworn to before me this 22nd day of July,1980.

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! ADDITIONAL.INFORMATION REQUESTED IN NRC LETTER, 6/10/88 Page 1 I

QUESTION NUMBER.1

.In Table 4.1 on enthalpy. rise, safety 11imits basis, an overall' product uof 8.81

'is _ reported. . This appears to be a typographical error. Please address.

ANSWER The overal'1 product for flow-related factors given in Table 4.1 in our answer to NRC questions dated March 11, 1988 is a. typographical error. The value.given as 8.81 should read as 0.81. The next page'is a revised Table 4.1.

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Corrected page 15 of answer.to questions 4 and 5 in  !

the MURR March 11, 1988 submittal.

1 Table.4.1 Summary of MURR Hot Channel Factors On . Enthal py Ri se. . . . . . . . . . . . . . . Safety New limits Fuel Basis Design

.Powe -related Factors Nuclear Peaking Factors Radia1........................................ 2.220 -

1.416 No n-u ni f o rm Bu rnu p . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.112 -

Local (Circumferential)....................... 1.040 1.330 Axia1......................................... 1.000 1.000 Engineering Hot Channel Factors on Enthalpy Rise Fuel Content Variation........................ 1.030 1.030 Fuel Thickness / Width Variation. . . . . . . . . . . . . . . . 1.030 1.040 Overall Product................................. 2.72 2.02 Flow-Related Factors Core / Loop Fl ow Fracti on. . . . . . . . . . . . . . . . . . . . . . . 1.000 1.000 Assembly Minimum / Average Flow Fraction. ..... .. 1.000 1.000 Channel Minimum / Average Flow Fraction I nl et Va ri at i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.000 1.000 Width Variation............................. 1.000 1.000 Thi ckness Va ri ati on. . . . . . . . . . . . . . . . . . . . . . . . 1./1.080 -

Within Channel Minimum / Average Flow Fraction 1/1.25 Thi ckness Va ri ati on. . . . . . . . . . . . . . . . . . . . . . . . 1./1.050-Ef fect i ve Fl ow Area. . . . . . . . . . . . . . . . . . 0.3231/0.3505 0.3231/0.3505 Overall Product................................. 0.81 0.74 ]

On Heat Flux...................

Power-Related Factors Nuclear Peaking Factors l

Radial........................................ 2.220 -

1.416 I Non-uniform Burnup............................ 1.112 - )

Local (Ci rcumf erential ) . . . . . . . . . . . . . . . . . . . . . . 1.040 1.330 Axial......................................... 1.432 1.347 Engineering Hot Channel Factors on Flux ,

Fuel Content Vari.ati on. . . . . . . . . . . . . . . . . . . . . . . . 1.030 1.030  !

Fuel Thi ckness/Wi dth Vari ati on. . . . . . . . . . . . . . . . 1.150 1.250 l Overall Product.................................... 4.35 3.27 Energy Fraction Generated in Fuel Plate......... 0.930 0.930

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Rev. JUL 2 2 1333 l

l ADDITIONAL INFORMATION REQUESTED IN NRC LETTER, 6/10/88 Page 2 QUESTION NUMBER 2

2. In the last sentence of page 16, the sentence reads, "... by changing the maximum

-acceptable-DNBR:. This appears to be a' typographical error. Please address. It is suggested that the corresponding COBRA DNBR be given for the new fuel design.

Table 4.2 indicates that COBRA is predicting 'approximately a'6 percent greater margin than the BOLERO safety limits. Will the COBRA minimum DNBR reflect this difference?

ANSWER:

In the MURR letter dated March 11, 1988, the last sentence on page 16 of the answer to questions 4 and 5 should read as follows:

For COBRA, the local hot spot engineering factor for fuel thickness / width variation can be taken into consideration by changing the minimum acceptable DNBR:

C Minimum BOLERO DNBR =

B Ho p

= 2.0 ,

l COBRA DNBR = Critical Heat Flux , Critical Heat Flux <

COBRA Hot Spot 1.03 BOLERO Hot Spot 1.15 therefore for a 1.15 local hot spot engineering factor:

1 Minimum COBRA DNBR = BOLERO DNBR = 2.23 1

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A revised page 16 is attached to this answer. .

l The minimum COBRA DNBR is 2.23 for tht. MURR safety limit basis. The minimum i COBRA DNBR corresponding to a BOLERO DNBR of 2.0 can be calculated for other local hot spot engineering factors for fuel thickness / width variation (HSEF) by the following equation:

Minimum COBRA DNBR = BOLERO DNBR = x 2.0

ADDITIONAL ~INFORMATION REQUESTED IN NRC LETTER, 6/10/88' Page 3 ANSWER TO QUESTION NUMBER 2 (cont'd)

As stated in the March 11, 1988 submittal, this minimum DNBR is based on using the ,

Bernath critical heat flux correlation (l) and corresponds to I true ratio to DNB of ~

1.2 as experimental measured.(2,3)

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The following method is.used to vecify a core is within the MURR safety limits:

1. The nuclear peaking factors are determined for the worst case cnndition. ~

2. The Engineering peaking factors and flow related factors are evaluated.

3. COBRA-3C/MURR will be run to determine what the maximum power level is for the COBRA DNBR corresponding to a BOLER0 ONBR = 2.,00 for the factors obtained in steps 1 and 2. Runs are made for the six sets of pressure /

flow / temperature conditions given in Table 4.2 of the March 11, 1988 submittal.

4. The core is within the MURR safety lunits if each of the maximum allowable core power levels determined in step 3 is equal to or greater than the l corresponding power levels given in Table 4.2 under COBRA-Safety Limit Basis. j i

This insures the safety margins established by the MURR Limiting Safety System l Settings are not reduced by the new core configuration.

In answer to questions 4 and 5 in our March 11, 1988 submittal, we assumed a HSEF of 1.25 to be conservative since we have not yet fabricated the fuel plates. Using this HSEF, the following COBRA DNBR was used for the 1.27 kg fuel design in Table 4.2: l Minimua COBRA DNBR = x 2.0 = 1.2 = 2.43 3 l03

l ADDITIONAL INFORMATION REQUESTED IN NRC LETTER, 6/10/88 Page 4-ANSWER T0 QUESTION NUMBER 2 (cont'd)

As noted in your Question 2*in the NRC June 10, 1938_ letter, the COBRA calculates-slightly, greater power levels than BOLERO for the same curresponding DNBR. This11s probably due to COBRA being a newer more modern t$ermohydraulics code allowing for -

more accurate modeling. The above method of checking a core is within' safety limits does not attempt to take advantage of this improved accuracy but maintains any un-necessary over conservatism built into the safety limits as a result of the use of BOLER0 in 1973.

REFERENCES

- 1. Bernath, L., "A Theory of Local-Boiling Burnout and Its Application to Existing Data," Chem. Eng. Progm. Symp. Ser. , 56, No. 30,95-116 (1960) .

2. Croft, M. W., "\dvanced Test Reactor Burnout Heat Treasfer Tests," UASEC Report 100-24475, Babcock and Wilcox, Co., (January 1964).

3. Waters, E. D., "Heat Transfer Experiments for the Advanced Test Reactor," USAEC Report BNWL-216, Battelle-Northwest, (May 1966). ,

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Corrected page 16 of answer to questions 4 and 5 in the 14JRR March 11, 1988 sub;ai ttal . i QUESTIONS 4 AND 5 (cont'd) the power level corresponding to a DNBR = 2.0 = Bernath critical heat flux divided by the "BOLER0" hot spot heat flux. For COBRA the hot channel factor of 2.72 and the axial power distribution can be given in the input, so the heat flux then has the fuel thickness / width variation engineering factor of 1.03 for the hot channel.

The local hot spot engineering factor of 1.15 cannot be input to COBRA. There-fore, the "BOLER0" hot spot heat flux = (1.15/1.03) "COBRA" hot spot heat flux.

For COBRA, the local hot spot engineering factor for fuel thickness / width variation can be taken into consideration by changing the minimur: acceptable ]

DNBR:

B0LERO DNBR = _ Critical Heat Flux , i,. 0 BOLERO liot Spot COBRA DNBR = Critical Heat Flux , Critical Heat Flux COBRA Hot Spot 1.03 BOLER0 Hot Spot 1.15 therefore for a 1.15 local hot spot engineering factor: ]

1 COBRA DNBR = BOLERO DNBR = 2.23 l

I Rev. JUL 9a 1E3 l

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