ML20212N384

Forwards Response to 870106 Request for Addl Info Re 861003 Proposal to Convert Reactor Critical Facility Core to Use Low Enrichment U Fuel

ML20212N384
Person / Time
Site:	Rensselaer Polytechnic Institute
Issue date:	03/03/1987
From:	Harris D RENSSELAER POLYTECHNIC INSTITUTE, TROY, NY
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20212N386	List: ML20212N384 ML20212N392 ML20212N398 ML20212N402
References
NUDOCS 8703130002
Download: ML20212N384 (15)

Text

,3e c+' Department of Nuclear Engineering 9- /"**4 Rensselaer Polytechnic Institute Troy, New York 12180 3590 March 3, 1987 Director of Office of Nuclear Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 i

SUBJECT:

DOCKET NO. 50-225, INFORMATION AS REQUESTED IN YOUR LETTER ON JANUARY 6, 1987 TO DR. D.R. HARRIS FROM RPI.

The enclosure to this letter contains responses to the twenty-six questions which have arisen during your review of our proposal to convert the Reactor Critical Facility core to use low enrichment uranium (LEU) fuel, dated October 3, 1986, under license No. CX-22.

Please contact me at (518) 271-4010 if you need additional information.

Sincerely,

, , S Dr. D.R. Harris Director, RPI Reactor Critical Facility

Enclosure:

As stated.

cc: J.J. Dosa B J 8703130002 070303 PDR ADOCK 0500o225 P PDR

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RENSSELAER POLYTECHNIC INSTITUTE RESPONSES TO QUESTIONS SENT BY JANUARY 6, 1987 LETTER FROM MR. JOHN J. DOSA TO DR. DONALD R. HARRIS CONCERNING HEU/ LEU CORE CONVERSION PROPOSAL DOCKET NO. 50-225 LICENSE NO. CX-22 4

4 Submitted March 3,1987

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2 QUESTION 1. Proposed changes in security exclusion. zones are not considered to be. relevant to core- conversion and will~ not be reviewed by NRC with your proposal. These ' proposed changes should be reviewed and processed according to 10 CFR 50.54(p) or "10 CFR'50.90, whichever is' appropriate.

ANSWER 10 QUESTION 1.- The proposed changes in security exclusion zones are' deal't with in a revised Security Plan transmitted to the NRC under ,'

D.R. Harris and F. Rodriguez-Vera to Mr. Thomas T. Martin, December 23, 1986.

QUESTION 2. Is the multiplication factor referred to in calculations for the fuel storage vault k infinite or k effective?

ANSWER TO QUESTION 2. Criticality calculations for the _ fuel storage rack in_ the Fuel.

Vault were carried out using the DIFXY code, where the x and y axes were perpendicular to the fuel-pins and to their supporting storage tubes. Leakage.in'the z direction, along the fuel pins, was included through a longitudinal buckling of. _

B = 0.0008875 cm! Reflecting boundary conditions were used at the midplane surf aces between storage tubes; thus the array of stored fuel tubes was infinite in extent in the x and y directions. Thus the calculated multiplication included leakage lLn_the z direction only.

QUESTION 3. Can the radial positions of the control rods be altered, if necesary, to assure that they are always adjacent to the outside row of fuel pins?

ANSWER TO QUESTION 3. The_ radial positions of the control rods can not be altered,

. and if necessary to achieve criticality, fuel pins would be accomodated around the control rods (i.e. not always adjacent to the outside- row of fuel pins).

QUESTION _4 . The effective neutron lifetime given in Table 5.1 for the RPI LEU core appears to be very short, especially for a thermally optimized lattice. Please explain this.

ANSWER TO QUESTION 4. The effective neutron lifetime 1* = 12.2 ps as given in Table 5.1 is very short because this is a highly absorbing lattice consequent to the high enrichment. The value of 1* was obtained directly from the LEOPARD output. That this is consistent with the other LEOPARD calculated results is seen -

from 1* = (Wigner - Wilkins 1/v)/(220000.x macroscopic thermal absorbtion cross section) = 13.6 ps for the base case. Here thethermal[a equals 0.1800, almost ten times that for water. The thermal diffusion area, L2 = 1.56 cm2 , is small so

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[ .that jh$ thermal neutroa leakage from the core does not have an-importantfeffect on"l*. ' We note, howeverg r. hat the cited 1*

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does not take.into accounbthose neutrons, leaking into the s

water reflectop and later't'ransporting~bap.c into the core.

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Such a calculat' ion pould require adjoint veli g hting in DIFXY'in ,

an. integral similar to that used here for 0,gg. This was not ,

? done for Cl* because 1* is not an iv.portant quantity in the / '

, transients studied. , . c J '? l ) N :'

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~ QUESTION 52 In the analysis 'of the MHA (vximum hypothetiE2f' accident) for l N your facility it is stated that a. pre-accident power'1evel of- 4, H 200 watts is assumed based upon the Technical Specifications ~

power level li: nit of 100 va' tts and incorporates a factor of two, .

/'

s to account for the cumulative uncertainties associateid with instrument calibration'. Explain whetfchese cuoulative uncertainties are and describe the pescautions tsken,by the ,'

operating staff to mialmize these uncertainties ?furirig noren1 ]

/

operation. ,

ANSWER TO

x. QUESTION. 5., More accurately stated, the cumulative uncertainties associated 7/

with instrument calibration increduce no more thanda factor of '

-? '

two uncertainty in the indic'atof pow'er level. The selection of y -

200 watts actual in-core power.(100 watts indicated on. Linear l

Power Channels 1 and 2) at 1A!tiation of the MHA assumes an ' i even more conservative uncertainty' factor of ab6ut nine. This

j is explained in the response to- Question .No. 6 herein. ,

Regarding the calibration of those channels which indicate 1 reactor power (two linear and two logurithmic), uncercaintieu -

j" are estimated as follL s

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Source of Error pncertainty' '

Level Instrument Res'ponse~ .5%

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CalibraNion' Technique ,'

30% ,

(Foil Activation) ,

TOTAL -

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Implied is that assumptions made in selecting initial power for 3

the MHA are very conservative and consistent with the ,;

philosophy of this analysis. .

Efforts made. by the operating staff to ninimize the specified -

uncertainties are idendified in Technical Specificati(ns 4.1. '

All instrument channels are calibrated . annually utili::ing a

,. foil-activation technique. In addition to this basic maintenanca' requirement, reactor 'startup yrocedures require signature verification of proper instrument fuv. tion prior to any rod withdrawal. This " check" consists of a visual inspection of instrumen't dial setpoints, detecter plateau ,

,- , - - , . . - - - , - , ..;,-4 , , ~ . , - -

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., 4 checks, and a comparison of readings with the neutron source in

- and out of the core. . Finally, operators are instructed to critically evaluate instrument readings in comparison with expected indication and mutual agreement. Any evident or suspected disparity would be cause for immediate shutdown and corrective action.

QUESTION 6.' Linear Power Channels 1 and 2 are assumed to indicate a value.

of 10% on the highest selectable-scale (100 watts indicated with a factor of 2 uncertainty) at the onset of the analyzed MHA. It is then stated that the power level must rise to 1800 watts before the Linear Power Channel Scram is activated. This does not correspond to the basis used for Technical Specification 2.2 which states that 90% of the highest scale for the linear power channels'(the Scram setpoint) corresponds to a power level of 135 watts. Explain and correct this-discrepancy.

ANSWER TO QUESTION 6. The discrepancy arises from a mis-statement of the assumed uncertainty in linear channel readings during the MHA. In fact a 90% reading on the highest scale of each linear power channel

does not correspond with 135 watts. Assumed is an in-core power of 200 watts which is only reflected as 10% of the highest scale on the linear power channels. An accurate measure of such an in-core power level would generate a meter reading of 133% on the highest scale of each linear power channel, provided instrument Scram was not activated. Hence, the assumed uncertainty is on the order of nine rather than two. The factor incorporates any predictable calibration uncertainties and is consistent with the very conservative approach employed throughout evaluation of the MHA.

QUESTION 7. All parameters used to calculate the consequences of the MHA utilize approximately the same level of conservatism, except the value of the proposed reactivity insertion. Reevaluate the MHA with all parameters tc:alved utilizing approximately the same level of conserva isu 'the same percentage of the Technical Specificat ' e ss 'aits).

ANSWER TO QUESTION 7. Most parameters chosen to estimate the consequences of the MHA exceed their respective Technical Specification limits by greater margins than the value of proposed reactivity insertion. Yet this does not imply the value for reactivity insertion is any less conservative. The question posed suggests concern that an experiment of unknown but very large reactivity worth might be conducted, and that the accidental removal of such an apparatus from the core would induce a transient beyond that of the MHA. Yet the administrative limits and operational controls enforced by RPI operators are geared to ensure the possibility of this kind of accident is

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• 5 most ~ remote. The startup tests identified in Question No. 15 herein will include in accurate measurement of Jcore reactivity values, specifically pin reactivity worth and the amount of-

' fuel required for' criticality. This measure, coupled with core loading restrictions listed in Section 7.7 of the Safety Analysis Report, are more -than Ladequate to prevent introducing

] .an experiment 1of.aore than 60d worth (the Technical

~

Specification 1imit). Accordingly, we view the amount am -very conservative ~ for use in the MHA- calculations and appropriate in relation to other parameters chosen. LFinally, the 604 b zf,<

. reactivity insertion chosen is consistent with the value-

utilized in the previous Safety Analysis - Report, submitted to

R. .and approved by the NRC in'1983.

The answer to Question No. 20 herein identifies that the l Technical Specification Safety Limit for the core will be-changed to a temperature limit below that at which UO2 is known to melt. . As a test case the MHA transient -was run with 704 initial reactivity insertion,.the additional 104 included to '

much more than account for any uncertainties in estimation of the reactivity worth of the subject unsecured experiment.

Again reactivity feedback effects of temperature ~and void were neglected. Analysis predicted that at no time during the transient ,would fuel, cladding, or water. temperatures increase more than 0.5 'C. Simply put, even the most severe rectivity.

E transien fails to cause an appreciable rise in material temperatures. The very conservative assumptions of the j' analysis and enormous margin to the specified Safety Limit demonstrate the safe nature of the reactor core.

, f QUESTION 8. What are the units of reactivity used in Table 5.2 of your proposal? You indicate in this table that the values for most of the parameters given for Core B are more restrictive than for Core A.e Does this 'mean the(values are more conservative?

Provide expected values of the, parameters for Core A. Review and correct all' values listed in the " Technical Specification" column against those values given in the proposed Technical Specifications.

ANSWER TO QUESTION 8. A revised Table 5.2 is provided a$ s ADDENDUM TO ANSWER TO QUESTION 8. both the previous (1983) and proposed (1986)

Technical Specifications give reactivity limits in units of dollars, so value in these units are supplied. The calculations for the new cores A and B yield reactivities in absolute units, so values in these units are supplied as well.

Note that Segg is different for the HEU (previous Technical Specifications 1983) and LEU (proposed Technical Specifications 1986) cores. It can be seen from the revised Table 5.2 that Core A is more conservative in some respects, e.g., temperature coefficient of reactivity, but Core B is more conservative in other respects , e.g. , shutdown margin.

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QUESTION 9. Explain why normal operations can be performed safely with Core B when a positive temperature coef ficient of reactivity exists.

ANSWER TO QUESTION 9. Normal operations can be performed safely with Core B when a positive temperature coefficient of reactivity, a T exists 3

because the net positive reactivity insertion from the minimum operating temperature (50 *F = 10 *C) to the temperature at which a becomes negative (91 *F = 32.8 *C, is less than T

B

$0.15 as shown below.

32.8 *C 32.8 f 10 *C a T

dT = 2.1125 x 10 -8 f 10 T2dT - 5.0345 x 10 -6 x B

xf 10 tdt + 1.4233 x 10~ f 10 dT

= (2.1125 x 10~ /3)(32.8 - 1000) -

~0 (5.0345 x 10 /2)(32.8 - 100) + 1.4233 x 10~ (32.8 - 10) 32.8 *C f a T dT = 1.0301 x 10" = $ 0.13466 < $ 0.15 10 *C B

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This value is within the established Technical Specifications. s i

j QUESTION 10. What is the calculated value of the Doppler coef ficient?

ANSWER TO QUESTION 10. At 20*C the calculated Doppler coef ficient of reactivity is equal to -3.404 x 10-5/*C for the solid Core A, and is equal to

-2.979 x 10-5foC for the annular Core B. Calculations of the Doppler coefficient have been carried out up to 920*C, where the values are -1.858 x 10-5 foC for Core A and -1.623 x 10-5/*C

for Core B.

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-QUESTION 11.' -What are the expected individual and total control rod worths with the LEU core?

ANSWER TO QUESTION 11. For Core A the worth of an individual control rod is calculated to be 0.0070216, and the total control rod worth is 0.028086. Similarly for Core B the worth of an individual

- control rod is calculated to be equal to 0.0071388, and the total control rod worth is calculated to be 0.028555.

QUESTION 12. What is the significance of two B10 absorber sections in a control rod, versus one section?

ANSWER TO QUESTION 12. The significance of two B10 absorber sections in a control rod, versus one section comes from the fact that presently available absorber sections have an active length (region actually containing neutron absorbing material) of only 22 inches, so by using an appropriate arrangement of two absorber sections it is possible to obtain an active absorber length of 36 inches-when a control rod is fully inserted in the core (active length, 36 inches).

QUESTION 13. What are the requirements for the BF3 source range counters during startup testing?

ANSWER TO QUESTION 13. The Startup Procedures for the RPI Critical Facility describe requirements for operation of the two BF3 detectors used in startup channels A and B.* These procedures were approved by the RPI Nuclear Safety Review Board (NSRB). Furthermore, Technical Specification F of Technical Specifications Section 3.1 requires that during reactor startup the counting rate range for the BF3 startup channels should be from 2 cps to 104 cps (counts per second).

• see ADDENDUM to Answer to Question 13.

QUESTION 14. Will the water height adjustment capability be utilized during startup testing?

, ANSWER TO I

QUESTION 14. The water height adjustment capability will not be utilized during startup testing. Fuel will be loaded as described in part 5.6 of the Proposed Technical Specifications.

QUESTION 15. Please be aware that the results of the startup tests should be

' kept on file for the life of the facility. These results must be submitted as a written report to the NRC within 60 days of the startup program's completion. The contents of the report i will be reviewed and approved by the NRC before normal operations can be implemented.

ANSWER TO QUESTION 15. We intend to conform with this recommendation.

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. O QUESTION 16. What is the correct date for Reference No. 8 on page 24 of your proposal? Please submit copies of Reference Nos. 2, 4 and 5.

ANSWER TO QUESTION 16. The correct date of reference No. 8 by Kacich is May, 1975.

Copies of Reference Nos. 2, 4, and 5 are given in the ADDENDUM TO ANSWER TO QUESTION 16.

QUESTION'17. Discuss the differences in the types of absorber sections listed in the Technical Specifications definition of control rod assembly and their differert effects on reactivity. The definition is not consistent with Section 5.4.4 of the Technical Specifications nor Section 4.4 of the Safety Analysis .

Report. Please clarify this.

ANSWER TO QUESTION 17. The Technical Specifications definition of Control Rod Assembly is correct and should not be modified. Section 4.4 of the Safety Analysis Report (SAR) has been rewritten, now provides information about the types of absorber sections, as well as a new figure 4.8, showing a control rod basket assembly for the new LEU core.

Section 4.4 of the Technical Specifications has also been corrected.

New Section 4.4 of the SAR and new Section 5.4.4 of the Technical Specifications are given in the enclosed ADDENDUM TO ANSWER TO QUESTION 17, along with correct and unchanged definition of Control Rod Assembly.

QUESTION 18. Definition 1.Q.6 of the Technical Specifications uses a value of $0.070 for the reportability criterion for an unanticipated change in reactivity. Why is this value higher than the value of $0.60 analyzed in the proposal as the MHA?

ANSWER TO QUESTION 18. We concur that the reactivity value of $0.70 identified in Definition 1.Q.6 of the Technical Specifications is inconsistent with the MHA analysis. Accordingly, the value for the reportability criterion for an unanticipated change in reactivity will be listed as $0.60 hereaf ter.

QUESTION 19. All references to fuel elements thrc:ghout the Technical Specifications should be updated to the use of fuel pins.

ANSWER TO QUESTION 19. We concur, and the change will be implemented.

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, 9 QUESTION 20. Provide a more detailed explanation of the use of energy deposition in the fuel as .the basis for the safety limit for your critical facility. ,Show all calcualtions and list the references used.

ANSWER TO QUESTION 20. The energy deposition criterion considered the. amount of energy deposited in the fuel during the MHA transient, neglecting any cooling effects and negative reactivity feedback due to temperature rise. Reference No. 6 of the basic Safety Analysis Report pertains.

Discussion with Mr. John Dosa,- NRC Project Manager, regarding the use of energy deposition as a safety limit criterion.has indicated that the Commission evaluation of'this criterion.

would prolong license issue unnecessarily. This is because other licensees have used temperature limits, and evaluating core-energy deposition poses a new problem for.the NRC.

Accordingly, we propose to alter the Technical Specifications to identify a maximum temperature limit for the UO2 fuel.

Section 2 of the Technical Specifications is given in the ADDENDUM T0' ANSWER TO QUESTION 20.

QUESTION 21. Why should the integrated thermal power not exceed 200 KWh per year for the LEU' core as given in Technica1' Specification 3.1.10?

ANSWER .TO -

QUESTION 21. It is considered prudent to establish an operational limit on integrated core thermal power to preclude the accumulation of significant amounts of long-lived fission products in the core.

The specified value of 200 KWh meets this objective while allowing operation of the reactor in support of research and classroom instruction.

QUESTION 22. Proposed Technical Specifications 3.2.1 and 3.2.2 are less conservative than your current Technical Specifications limits.

Provide an explanation for this with emphasis on safety considerations.

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., 10' ANSWER TO QUESTION 22. Proposed changes in Technical-Specifications 3.2.1 and 3.2.2 are the following:

Core Characteristic HEU Core LEU Core Isothermal temperature coefficient of

<0 for T h>90*F <0-for T th 00 4 reactivity, aT(T).

[ aT(T)dT <$0.11 <$0.15 50*F Void coefficient of reactivity, a y(r) for all r for all r

<0 ~

and Tg>90*F and Tg>100*F Specifically the changes are in a) The threshhold water temperature, Tth, above which aT(T) and a (r) must become negative. It changed from 90*F (HEU core) to 100*F (LEU core).

and in th b) Thevalueoff a (T) dT , which must be smaller than 50*F

$0.15 for the LEU core.

In the case of the Maximum Hypothetical Accident (MHA),

analyzed in Section 5.4 of the SAR, and Answer to Question 7, a fuel temperature rise of less than 0.5'C (0.9'F) above an initial value of 20*C (68'F) is induced. This very small temperature change makes feedback effects negligible, even in the temperature range from 50*F to 100*F for both Core A and Core B pin arrangements. Therefore proposed Technical Specifications 3.2.1 and 3.2.2, do not adversely impact the safe operation of the LEU core.

. QUESTION 23. In the bases for Technical Specifications 3.4.8 and 3.4.9, provide the correct references where 10 CFR 105(1) and 10 CFR 106 are currently listed.

ANSWER TO QUESTION 23. In the Bases for Technical Specifications 3.4.8 and 3.4.9 change: "10 CFR 105(1) and 10 CFR 106" to "10 CFR 20.105 and 10 CFR 20.106."

~

. -D 11 QUESTION.24. Revise the recently proposed Technical Specification 5.6 to be consistent with your recently amended Technical Specification 5.7.

ANSWER TO QUESTION 24. The recently amended

• Technical Specification 5.6 Fuel, will be deleted:after approval of the Proposed Technical Specifications for the new LEU _ core. Then the recently amended

• Technical

' Specification 5.7 will become Technical Specification 5.6, which is given in the ADDENDUM to answer to Question 24.

• Attachment to License Amendment No. 6 Facility Operating License No. CX-22 Docket No. 50-225 QUESTION 25. In Technical Specification 5.6.1 the pin worth value of $0.72 is inconsistent with the maximum single pin worth of $0.20 allowed in Specification 3.1.1. Please explain this.

ANSWER TO QUESTION 25. We concur that the pin worth of $0.72 identified in Technical Specification 5.6 is inconsistent with the value identified in section 3.1.1. Accordingly, the value identified in Technical i Specification 5.6.1 will be listed as $0.20 hereaf ter.

QUESTION 26. Proposed changes to Section 6 of your Technical Specifications are not considered to be relevant to your core conversion and will not be reviewed by NRC with your proposal. The proposed changes should be submitted as an amendment acording to 10 CFR 50.90.

ANSWER TO QUESTION 26. The comment is noted. The proposed changes will be submitted as a license amendment according to 10 CFR 50.90.

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e ADDENDUM TO ANSWER TO i

, QUESTION 8 New Table 5.2: Kinetics Parameters of RPI LEU Core and Technical Specifications 1.

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Tab 1eL5.2. Kinetics Parameters of RPI ' LEU Core and' Technical Specifications.

Technical Specifications LEU Core A. LEU Core B Proposed 1986 Previous 1983 Values Values .(8,gg=0.00765) (8 gg=0.0078)

Excess Reactivity <0.60$ <0.60$

at 68'F (0.00459 <0.00459 (<0.00459) (<0.00468)

.Subcritical Reactivity >0.70$ >0.70$

With One Rod 0.01850 0.01897 (>0.005355) (>0.00543)

Stuck-Shutdown >1.00$ >2.86 Margin 0.02350 0.02397 (>0.00765) (>0.02231)

Core Average <0 for <0 for <0 for <0 for Isothermal T>90*F T>100*F T>100*F T>90*F Temperature Coefficient Core Average Void <-7.65x10-6* <-7.13x10-6* <0 for' <0 for Coefficient of per cc void per cc void T>100*F T>90*F Reactivity Local Void -2.73x10-6 -2.52x10-6 <-0.00043$/cc <-0.00043$/cc Coefficient of per cc void per cc void (<-3.290 (<-3.354 Reactivity.in x10-6/cc void) x10-6/cc void)

Fueled Region-Integrated Reactivity <0.15$ <0.11$

Due to Temperature 0.00006035 0.001030 <0.001148 (<0.000858)

Change, 50*F to Temperature at which aT=0 Reactivity Worth <0.20$ <5.50$

of Standard 0.000528** 0.000333**

Fuel Assembly (<0.00153) (<0.0429)

(HEU Core) or Fuel Pin (LEU Core)

• Values given are at 20*C. Values at higher temperatures are more negative.

• • For removal of fuel pin at periphery of core.

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ADDENDUP.

TO ANSWER TO QUESTION 13 Pages 1,2 and 3 of STARTUP PROCEDURES FOR CRITICAL FACILITY.

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