ML20140H422

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Application for Amend to License DR-15,proposed Change 4 for SEFOR
ML20140H422
Person / Time
Site: 05000231
Issue date: 09/03/1970
From:
GENERAL ELECTRIC CO.
To:
Shared Package
ML20140G249 List: ... further results
References
FOIA-97-34 NUDOCS 9705130129
Download: ML20140H422 (7)
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PROPOSED CHANGE NO 4 FOR THE SOUIHWEST EXPERIMENTAL FAST OXIDE REACTOR i

l Re: LICENSE DR-15 DOCKET 50-231

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l GENERAL ELECTRIC COMPANY 4,'gijf ,s j; 310 DeGuigne Drive

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l G-Proposed Change No. 4 (9/3/70)'

for the Southwest Experimental Fast Oxide Reactor

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I. Introduction -l l

Under the authority of License DR-15, General Electric operates the Southwest Experimental Fast Oxide Reactor at a site near Strickler, + ,

1 l Arkansas.

l A revision of the current Technical-Specifications is desired as <

l described herein. The applicable revised pages of the Technical )

. Specifications are also included as Attachment A.

II. Proposed Changes Pursuant to the provisions of 10 CFR 50.59, General Electric requests that the- SEFOR Technical Specifications be changed by substituting i

I pages 3.10-1, 3.10-2, and 3.10-3 in Attachment A of this document for

( corresponding.pages of the current Technical Specifications and by, adding pages 3.13-1-through 3.13-5 in Attachment A of this document to the current -Technical Specifications.' . The proposed changes in Section 3.10 are, indicated by brackets, in the margin. These marks were omitted in Section 3.13, which is new in its entirety.

III. Purpose of the Proposed Change Specification 3.10.E of the present SEFOR Technical Specifications ,

requires that cer'tain additional specifications be approved and incorporated into the Technical Specifications prior to initial operation at 20 MWt. These specifications are submitted herewith, based on experience gained in operating SEFOR up to and including '

the 10 MWt power level, to satisfy the requirements of Specifications 3.10.E and 6.6.B.3.

IV. Discussion A. Limits for unexplained reactor behavior are based on the ability to detect changes in parameters and on the variations which may I occur in these parameters during normal reactor operation.

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1. The cover gas monitor has a range of 1 to 104 mR/h and will read about 20 mR/h,at 20 MWt due to A-41 activity in the cover i gas. If pin hole leaks occur.in one or more fuel rods, the

, cover gas activity will increase rapidly and then stabilize

., at some presently unknown level. Therefore, the limit for ,

this parameter must be expressed as a ratio rather than as an absolute value. A factor of three increase in the cover gas activity was selected as an appropriate initial limit.

2. The reactivity required to rbach a given power level from the zero power condition at 350*F is estimated from the following

. equation:

Ak =

0.692 (T -

350) .0.001785 (T - 350)2 1

260 in - 0.5 P - 0.015 AT 460 + T i

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where:

Ak = reactivity, cents' P. = reactor power, MWt '

T =

average core coolant temperature, *F AT = core coolant temFerature rise, 'F l

. .. l l l To expeditiously accomplish the calculation of Ak, a simple computer program was formulated which uses the reactivity model, the anticipated condi'tions and the reflector calibra-

, tions to predict the reflector positions for the computer con-ditions. For the testing to 10 MWt some 40 comparisons were {

obtained and analyzed. The results indicated a mean differ- i ence between measured and predicted of 1.3 cents with a stan-dard deviation of 4.5 cents. The selection of + 10 cents as j the demarcation between normal and anomalous reactivity results in approximately 95% confidence that disparities greater than this value are truly anomalous. These conclusions are par-

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ticularly germane since the calculated reactor parameters of i the above equation are those which the experimental program is designed to verify. The p'redictive schece is being contin-l ually updated by modifying the above equathn with any perti-L nent experimental results. For example, on the basis of

a previous reactor (RAPSODIE) and fuel testing experience, l

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changes in this equation are inevitable aue to fuel restruc-

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turing and cracking and would not be considered to be anoma- )

lous. Reductions in the power coefficient have already been observed at SEFOR during the initial appi,ach to power as

. would be predicted considering this effect.  ;

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3. The main primary pump flow rate can be predicted by the follow-ing equation:

q ,

100 V (v/400) + 6.01 Q = Main Primary Flow Rate, GPM V =

Main Primary Pump Voltage, Volts This empirical equation resulted from an accumulation of data taken over a wide variety of operating conditions. A least square regression analysis was used to obtain the curve form and coefficients from the data. The limit of +10% was selec-ted to be large,enough to exclude variations d'eu to random errors in data taking and repeatability considerations, but I small enough to detect incipient problems before they become detrimental.

At low flow rates, the applied pump voltage cannot be read j with sufficient repeatability to obtain good correlation l between predicted and measured flow rates. A lower limit of 1500 GPM has therefore been specified for the required com- l

, parisons. The normal range of flow rates called for during reactor operation is 1500 to 5000 GPM. l

4. The upper reactor vessel outlet temperature must be compared to temperatare measured in the main primary coolant system to identif- any unexplained behavior since the planned tests will resu.lt in a wide range of normal values for this tem-perat.ure. Any significant change in core cooling due to an inc case in core bypass flow will cause this temperature' l

comparison to exceed the specified limit and will therefore be detected.

B. Initial operations up to a power level of 15 MWt have been com-pleted without any indication of fission gas leaks from the fuel

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l. by the gross cover gas' monitor.

i The sensitivity of the gross ,

cover,' gas monitor has been demonstrated as required by 3.3.L of I the Technical Specifications; however, there is no operating experience with leaky fuel rods that can be used to establish J firm quantitative limits on acceptable cover gas activity levels, i composition of fission products in the cover gas, or fission l products in the sodium. Experience with operating fast-spectrum l sodium-cooled oxide fueled reactors '

indicats that leaky fuel rods should be expected but that continued operation in- the l presence of " leakers" does not present a safety problem. For .

l example, RAPSOD1E detected its first " leaker" with approximately 10 E d/kg fuel burn-up and has safely operated to fuel burnups in excess of 40 to 50 Wd/kg with no indication of gross fuel failure .

The proposed Specifications 3.13.A.1 and B provide assurance that the reactor will not be operated if the gross cover gas monitor cannot detect the r^elease of fission gas from one addi-tional fuel rod. As discussed in the bases to 3.13.B. fuel failures as defined in 3.3.K are not expected to occur. Nonethe-less, the first positive indication of release of fission gas from a fuel rod will be treated as evidence of a possible fuel failure as defined in 3.3.K of the Technical Specifications. A cover gas sample will be taken for spectral an . lysis. Two or more guinea pig rods under the inner most refueling ports will be examined in the refueling cell to verify that the fuel is not operating near a damage threshold. Sodium samples will be taken and analyzed for fission products. The procedure outlined in 4.9B will be followed to determine if it is safe to resume reactor operations.

It is expected that the results from the various surveillance functions will indicate that the observed increase in cover gas activity was the result of a pin hole leak in a fuel rod and reactor operations would resume.- It is also possible that addi-tional investigation with the reactor operating at reduced power l would be required to obtain sufficient information to allow

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resumption of normal reactor operation. Subsequent reactor opera-tionpillrequirecarefulsurveillanceofgrosscovergasactivity l

,, and composition.

l The safety significance of operating with an undetected fuel rod failure has been discussed previously(4) .

i' In summary, SEFOR has been successfully operated to a power level

! of 15 MWt without indication of loss of clad integrity. Conse-quently, there is no operating experience at SEFOR that can be )

used to establish numerical limits on all of the surveillance functions associated with an investigation of a possible fuel fai'l ure. Experience at EBR-II( .6,D RAPSODIE( ), and BR-5 0) l indicate that system characteristics have an important influence on the behavior of variables used to determine the extent of a loss of fuel clad integrity. Quantitative limits that can be I established based on operating experience at SEFOR and similar l l

  • l reactor facilities are included in the proposed specifications. I Additional limits that will be developed based on operating ex-perience at SEFOR will b' e reported in the quarterly operating reports or in special reports as required by 4.9.B of the Techni- l cal Specifications. l The proposed limits on anomalies other than those associated with the reactor cover gas activity are based on actual operating ex-

, perience at SEFOR. As discussed in the bases for these proposed limits, the values have been selected so that the proper action is taken on variations that are statistically significant. i Actions to be taken in the event any of the proposed limits are 1 reached are covered.in 4.9 of the Technical Specifications. l l

References:

1. G. Kayser, " Problems Due to Fission Products in Circuits of Sodium-Cooled Fast Reactors in the Event of Can Fractures," EURFNR-593 (Original Report No. DRP/SEMIR/ CAD.68.R.575), Nuclear Research Center Cadarache and Saclay l (France), December, 1968. ,

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l 2. A. I. Leipunskii, et al., "Ex'perience Gained from the Operation of the

! BR-5 Reactor,1964-5," Paper 2/3 of Proceedings of the London Conference -

on Fast Breeder Reactors, May, 1966, pp 171-185.

3.Dr.-Stanley J. Stachura (Cadarache Representative, AEC-EURATOM Fast Reac-tor Exchange Program), " Monthly Report No. 18, Fby, June,1969," pp 3,4.

i j 4. SEFOR FDSAR, Supplement 21, p.1-10.-

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5. R.R. Smith & C.B. Doe, " Cladding Failure Simulation Tests in ERB-II,"

-ANL-7067, December, 1966.

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6. R.R. Smith, et al., " Locating and Identifying the Source of the May 24, 1967 Fission Product Release in EBR-II," ANL-7543, April, 1969.
7. R.M. Fryer, et al., " Symptoms and Detection of a Fission Product Release from an EBR-II Fuel Element: Case 1. Defect above Fuel Elevation,"

i ANL-7605, January, 1970.

8. I.A. Efimov, et al., " Isotopic Composition of Fission Products in the Gas of the First Circuit of a BR-5 Reactor with Uranium Monocarbide Fuel," Translation from Atomnaya Energiya, Vol. 26, No. 6, pp. 544-546, J une , 1969.

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