ML20028E965

Forwards Addl Info on Sodium Spill Vols for Inerted Cells, Per Request.Info Will Be Incorporated Into PSAR Amend

ML20028E965
Person / Time
Site:	Clinch River
Issue date:	01/27/1983
From:	Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:	Check P Office of Nuclear Reactor Regulation
References
HQ:S:83:199, NUDOCS 8301280304
Download: ML20028E965 (5)

Text

O Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:199 JAN 271983 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Check:

ADDITIONAL INFORMATION ON S0DIUM SPILL VOLUMES - CLINCH RIVER BREEDER REACTOR PLANT Enclosed is additional information on sodium spill volumes for inerted cells as requested by the staff. It will be incorporated into the Preliminary Safety Analysis Report in a future amendment.

Any questions regarding the information provided can be addressed to Mr. W. Pasko (FTS 626-6096) or Mr. D. Florek (FTS 626-6188) of the Project Office Oak Ridge staff. -

Sincerely, J n R. Longene er Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy Enclosure cc: Service List Standard Distribution Licensing Distribution go o !

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t c) pipa embedded in cencrete and weld:d directly to cell liner

. . with full pInetration welds.

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Penetrations between inerted cells having a comon atmosphere may also consist of an open pipe sleeve which is welded to the cell liner 59 at each face with full penetration welds.

3A.8.3 Design Evaluation The piping integrity investigation analysis of crack growth due to all design duty cycle events indicated negligible crack growth. Based upon this evaluation, it is concluded that no leaks will occur under operation in accordance with the piping design specifications, i

The sodium and NaK components and piping in the CRBRP nuclear steam supply system and auxiliary systems are all designed to prevent leakage. The liners in cell which contain sodium or NaK should therefore not be exposed to any conditions more severe than those corresponding to normal plant operation.

However, accidental sodium leaks or spills cannot be precluded and therefore 45 must be considered in the design of the liners.  ;

To accommodate the effects of accidental sodium leaks or spills, the cell liners will be designed for a design basis sodium spill. Based on the operating experience of existing sodium facilities and previous assessments of sodium spills, the amount and/or leakage rate of an accidental sodium spill into a cell are minor. Leaks which could develop from flaws, fatigue, creep, etc. are expected to be much less than the Design Basis Leak as discussed in Reference 2 of Section 1.6.

3A.8.3.1 Sodium Spill Evaluation The evalfa o f the consequence of sodium spills is provided in PSAR Section 15.6.7 The method and criteria for evaluation of the cell liners are discussed in Section 3.8-B.

3A.S.3.2 Brittle Failure Potential of the Liner in Irradiated Areas The increase in ductile-brittle transition temperature due to neutron damage is estimated to be less than 100F for the reactor cavity liner. This is based on damage function analysis, which indicates that 37 the damage level for the neutron spectrum in the reactor cavity will be approximately 100 times lower than that for LWR reactor vessels.

For the neutron embrittlement evaluation of the cell liner plate, the methods and limits established by USNRC Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials" will be used. The only area of the plant exposed to neutron fluence is the reactor cavity. By considering the worst case exposure condition 18 2 for (E >0.1the reactor M V) and cavity 6.1 xce]n/cmlingr lo where the maximu fluence is 7.8 x 10 n/cm ,

(E <f.0M V), the maximum adjustment of Nil-ductiTity temperature (NDT) is 10 F a$d does not require trace element 59 control. This indicates that the liner steel is not effected by neutron embrittlement nor does gama radiation result in steel degradation.

Amend. 59 Dec. 1980 3A.8-3

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. INSERT TO 3A.8.3.1 Cells other than those analyzed in PSAR Section 15.6 are analyzed in a similiar fashion. The cells will be designed to accommodate the peak pressure from these spills. For the cells that are specified as a 15 psig cell, the worst case maximum peak pressure is calculated to be 13 psig. For the cells with a 12 psig design pressure the peak cell pressures are in the range of 2-6 psig.

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ATTACHENT C DESIGN BASIS LEAKS To accommodate the effects of accidental sodium leaks or spills, the cell liners shall be designed for a design basis sodium spill. Based on the operating experience of existing sodium facilities and previous assessments of sodium spills, the amount and/or leakage rate of an accidental sodium spill into a cell are minor. To insure that cell structures are conservatively designed, leak rates, spill volumes and spray Impingements used to establish structural loadings are based on leakage from a sharp edged circular orf fIce whose area is equal to one quarter of the pipe wall thickness multiplied by the pipe inside diameter. This leakage criterton is consistent with the Intent of the moderate energy fluid system leak defined for fluid systems with low stored energy In NRC Branch Technical Position MEB3-1, " Postulated Break and Leakage Locations in Fluid System Piping Outside Containment."

Leak rate time histories for the Primary Heat Transport and Reactor Cavity Cells are defined in Table 15.6.1.4-1. Leakage flow-rates for postulated leaks in the Auxiliary System piping in Reactor Containment and Reactor Service Building inerted cells range from approximately 100 GPM to 6 GPM.

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TABLE 3.8-B.C-1 SPILL VOLUMES IN INERTED CELLS Cell Spill Volume (Gallons) Spill Temperature OF 101A See 15.6.1.4 See 15.6.1.4 101C 101D 101E 102A 25,400 830 102B 25,400 830 103 25,400 830 104 25,400 830 107A 25,400 830 107 B 25,400 830 107C 25,400 830 121 See 15.6.1.4 See 15.6.1.4 122 "

123 131 700 200 (NaK) 132 400 830 141 360 830 157A 17,500 830 157B 17,500 830 157D 17,500 830 157E 17,500 830 331 4,650 , 600 331A 4,650 600 337 15,250 600 351A 700 400 351B 15,250 600 351C 15,250 600 351D 15,250 600 357 4,800 600 357A 4,800 600 357B 4,800 600 360 15,250 600 360A 15,250 600 361 15,250 600 386 840 600 387 630 600

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