ML20058H797

Forwards Request for Addl Info Re 900430 Application for Amend to License R-84

ML20058H797
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	11/14/1990
From:	Mendonca M Office of Nuclear Reactor Regulation
To:	Irving G DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY
References
NUDOCS 9011260046
Download: ML20058H797 (4)

Text

. . . . . . . . , , , , , , , , , , , , , , , ,

, . November 14, 1990 Docket No. 50-170 George W. Irving, Ill, Colonel, BSC, USAF Director Armed forces Radiobiology Research Institute Bethesda, Maryland 20814-5145

SUBJECT:

REQUEST FOR ADDITIONAL INF0PNATION We are continuing our review of your application for amendment of Facility Operating License No. R-84 submitted on April 30, 1990. During our review of your application for amendment, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed Request for Additional Information within 60 days of the date of this letter, so that we may continue our evaluation of your application. If you have any question on this review please contact me at (301) 492-1128.

The reporting.and/or record' keeping requirements contained in this letter affect-fewer than ten- respondents; therefore, Office of Management and Budget clearance is not required under Public Law 96-511.

Sincerely, Original signed by:

Marvin M. Mendonca, Senior Project Manager Non-Power Reactor, Decommissioning and Environmental Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated DISTRIBUTION RDoctet< File'd, NRC & Local PDRs PDNP r/f D. Crutchfield W. Travers S. Weiss E. Hylton M. Mendonca OGC E. Jordan ACRS(10) 0FC :PDNP/gA :PDNP/PM g :PDNP/ ),):~  :  :  :

,-----:--- p-.-----:------------:-....- .4/-:-PDNP:PM 3---:------------:------------:..

TNAME :EQton- :MMendonca:st:SWeis  : AAd 8 :

.../--:-----------:-----------:-------  :

DATE :11/ /90;

-..--:---[-..-----:.......-....:--.....-----:--A11/

11/[p/90 :11/7/90  :

6 /90 :  :  :

OFFICIAL RECORD COPY Document Name: AFRRI FUEL FOLLOWER QUESTIONS

\

..D /g Nl 9011260046 901114 k PDR ADOCK 05000170 p PNV

f 'o,, UNITED si ATEs y p, NUCLEAR REGULATORY COMMISSION

.c j WA$mNGTON, D. C. 20555 k..../ November 14, 1990 Docket No. 50-170 George W. Irving, III, Colonel, BSC, USAF Director-Arnied Forces Radiobiology Research Institute Bethesd6, Maryland 20814-5145

Dear ;olonel:

Irving:

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION We are untinuing our review of your application for amendment of Facility

-Operating License No. R-84 submitted on April 30, 1990. During our review of your application for amendment, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed Request for Additional Information within 60 days of the date of this letter, so that we may continue our evaluation of your application. If you have any question on this review please contact me at (301) 492-1128.

The reporting and/or record keeping requirements contained in this letter affect fewer than ten respondents; therefore, Office of Management and Budget cle6rance is not required under Public Law 96-511.

Sincerely,

.. %dh A<~ -

Marvin M. Mendonca, Senior F>oject Manager Non-Power Reactor, Deconsnissioning and-

-Environmental Project Directorate Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation Endlosure:

As stated i

. . \

l Reouest for Additional Infornation

1. Provide your testing program for meesurement of significant core character-istics with Fuel Follower Control Rods (FFCR). Include consideration of reactivity temperature coefficients, shutdown margin, excess reactivity and rod worth. Also, _ include acceptance criteria for test results which assure that core characteristics are within the limits assumed in the safety analyses for the cotire spectrum of core operations and burnup conditions. We are especially interested in your assessment of shutdown margin, since your submittal did not seem to address shutdown margin in a detailed manner.

2. Provide a copy of reference 1 to your safety analysis, i.e., General Atomics letter to Mark Moore dated October 28, 1988.

3. For Page 3 of the safety analysis, " Power Density in FFCR Fuel Element:"

a. A power level of 1.0 Megawatt (MW) was assumed. Provide safety analyses to support the requested steady state operation power of level of 1.1 MW.

l -b. Provide analyses that verify i. hat the most limiting power peaking condition was assumed in your analyses for both steady state and pulse power operations.

1. Possibly the most limiting condition is where a rod of 12%

uranium by weight (wtt) is surrounded by 8 wt% fuel, ii. Include in your analyses consideration dur59 pulsing of the i condition before significant heat has flowed from the rod to

the coolant.

'4. In the analysis for " Maximum Temperature in FFCR fuel Element," the free convective heat transfer coefficient, h, was experimentally determined as e

described in Appendix A of the safety analysis. Appendix A's determination was based on B-ring conditions. Provide analyses that the use of the B-ring's heat transfer coefficient is acceptable for use in the D-ring locations. Include evaluation of all pertinent pt.rameters, e.g., tempera-ture differentials, power profile, etc.

l 5. Provide additional description of the FFCR. Include H/Zr ratio and l burnable poison content.

6. Attachment I, page 1-1, item 6 refers to a 1961 General Atomics report.

Provide appropriate reference or safety analysis to demonstrate applicability to the current stainless steel fuel rod configurations.

7. Attachment I, page 1-4, item 12 refers to ANSI 15.4 I a. Is this the appropriate reference? Provide any needed corrections.

L

i

-2

b. Provide wording to assure that it is clear that the one month interv61 is to be maintained as an average over the long term in accordance with ANSI 15.1.

8. Provide verification that the proposed definition of fuel element and the current definition of standard control rod acceptably establish the FFCR configuration. Include considerations of Attachment I, page 1 4, item 13.

9. Provide a sumniary of the results of FFCR operations in similar reactors.

If possible, include th9 fuel element temperature measurements and/or calculatiens in both st(edy state and pulse operations to provide addition 61 assurance. Include comparisons of wt1 and loading patterns for FFCR and adjacent fuel Letween ycur facility and the other similar reactors you' re mentioned in ,vour submittal.

10. Provide appro)riate safety analysei to verify that the impact of increased stesdy state power level is limited by your previous safety analyses, or provide revised safet) analyses specifically for the power increase.

I

• MR p.

i:

I 1

o l

1 l