ML19317E143

Addendum to Rept 2 to ACRS Re Duke Power Co CP Application

ML19317E143
Person / Time
Site:	Oconee
Issue date:	07/06/1967
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
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ML19317E142	List: ML19317E140 ML19317E143
References
NUDOCS 7912160013
Download: ML19317E143 (12)

Text

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W 3% 6 U. S. ATOMIC ENERGY COMMISSION DIVISION OF REACTOR LICENSING REPORT TO ADVISORY COMMITTEE ON REACTOR SAFEGUARDS IN THE MATTER OF DUKE POWER COMPANY CONSTRUCTION PERMIT APPLICATION FOR OCONEE UNITS 1, 2 AND 3

_ ADDENDUM TO REPORT NO. 2

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Note by the Director, Division of Reactor Licensing The attached report has been prepared by the Division of Reactor Licensing for consideration by the ACRS at its July, 1967 meeting.

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@K.3C0AL U$2 @NLT 1.0 Introduction During the June, 1967 ACRS meeting, the applicant, Duke Power Company, was requested to submit in writing answers to qiistions on various aspects of the' plant design.

In our Report No. 2 to the Committee dated June 16, 1967, we stated that we would comment on these answers in a supplemental re-port and this addendum is submitted to this end.

2.0 Amendment No. 5 Amendment No. 5 to the Duke Power Company's application for three units at its Oconee Nuclear Station includes 11 answers to questions asked by the Committee and additional information to confirm comeitments made orally to the staff.

Our comments on the information submitted in Amendment No. 5 follow:

2.1 Cold Water Iniection The results of an analysis of the thermal transient seen by the reactor vessel during safety injection after a blowdown were presented.

Ductile yield-ing, brittle f racture and fatigue failure were considered.

The results of the analysis indicate no loss of vessel integrity due to the thermal transient during core injection.

The details of the calculational pLocedure were not pre sented, however, and we cannot conclude that the problem is finally resolved.

We are engaged in reviewing the Westinghouse approach to the maalysis of this problem and expect to meet with B&W in the future for a similar detailed review.

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@K3C0AL U$2 @ME 2.2 Solutions to the " Steam Bubble" Problem The applicant has acknowledged the possibility of core floeding being prevented by formation of a vapor lock or " steam bubb:c" between the core and a water leg in the steam generator after a cold leg pipe break.

Two methods of relieving hot leg pressure to the cold leg have been proposed as solutions f

to this problem:

(1) check valves located on the Core Support Shield which

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would be held closed by higher pressure in the outer annulus during, ump oper-ation or natural circulation, or (2) a rupture disk which would be designed 4

to blow out under internal steam pressure but which would withstand the ex-ternal operating pressure differential. We understand that the applicant prefers the check valves at present but that alternate means will continue to be studied as the design progresses.

The check valves would be designed and supplied by a valve manufacturer with experience in the use of check valves requiring similar specifications.

The present conceptual design would use several check valves with a maximum flow area of 10 f t These would open on a pressure difference of less than 1 psi.

At 3.5 psi (at which point the core would be 1/2 covered) about 1500 lbs of opening force would be applied to a 24-inch valve.

We believe that the check valves proposed would provide an acceptable solution to the steam bubble problem. Other potential problems arise, however, because of the use of these valves which must be considered in the design:

(1)

The Core Support Shield must be locally strengthened to compensate for the removal of material.

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. (2) The forceful opening of the valves against the reactor vessel during blowdown must be considered.

(3)

The consequences of loss of a valve must be evaluated or the design must provide assurance against such loss.

(4)

The ef fect on normal operation must be considered, particularly any possibility of bypassing or short-circuiting the core during pump operation or natural circulation.

(5)

The valves must be capable of test and inspection.

We believe that the above problems are all capable of solution and that the present proposal could resolve the steam bubble problem.

2.3 Blowdown Forces on Reactor Internals The applicant has proposed that the stress levels to be met in the anal-ysis of blowdown forces on reactor internals correspond to the minimum speci-fication yield strength value specified in Section III of the ASME Code.

For stainless steel the basic allowable code stress intensity (Sm) is 90% of yield stress at temperature (for carbon steel S,would be 2/3 yield strength at temperature).

The applicant has stated that the design stress levels are conservative but it should be pointed out that the conservatism lies in the stress limits set by the code rather than in the applicant's use of them.

While some defor-mation is expected at the stress level corresponding to the yield stress, gross deformation would not be expected since the deformation at the yield stress corresponds to a strain of only 0.002.

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. Based on our analysis of the applicant's proposal, we conclude that the proposed stress limits are acceptable since (1) they are being used for the design of vessel internals which will not have continuous high stress levels rather than on the pressure barrier itself, (2) conservative margins are inherent in the stress levels specified by the code, and (3) the loadings are not expected to be applied more than once, in the service life of the vessel.

I 2.4 Maior Pipe Break Within Primary Cavity

'Ihe applicant has stated that the primary shield pit could withstand the transient pressure resulting from an 8 ft break in the primary piping and has indicated that physical limitations, incloding pipe stops, would prevent a pipe break with area greater than about 4 ft. We believe that with design attention to physical restrictions within the primary cavity a pipe break larger than 8 ft can be ruled out.

2.5 Design of Submerged Weir in Intake Canal, In Supplement 5 the applicant submitted the dimensions of the under-water weir to be placed in the intake canal as well as the results of a stability analysis.

Subsequently, at the request of the subcommittee we have asked Dr. Newmark to review the dam design from the standpoint of foun-dations, materials and earthquake resistance. We have obtained further information from the applicant informally for this purpose.

Dr. A. J. Hendron of Dr. Newmark's staff will submit a report on the design of the underwater weir for the July ACRS meeting on this application.

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3.0 Core Cooling Analvsis for the Spectrum of Breaks At the request of the subcomnittee, we have held further discussions with the applicant on the ability of the emergency core cooling systems to The cope with the full spectrum of primary coolant system pipe break si~

applicant had previously submitted, in the PSAR, a full analysis of the break) and had double ended rupture of the largest coolant pipe (14.1 ft also presented information on a spectrum of hot leg ruptures with respect to system pressure and mass release during the blowdown.

In our previous reports to the Committee we have stated that the anal-ysis was not complete but that there was enough information available to provide assurance that the spectrum of breaks would be covered and that a construction permit could be issued. The applicant's continuing efforts in this area have resulted in more detailed information, particularly in the area of cold leg breaks and peak clad temperature.

In addition to the new data, as presented in this report, we have compared various calculations made by Westinghouse with those of B&W.

Table I summarizes the more detailed information now available on the Oconee design for the spectrum of hot and cold leg coolant line breaks.

Information on the time and extent of core uncovery and the pressure decay for the spectrum of hot leg breaks was presented in Figures 14-41 and 14-42 of the PSAR. A mass release of 310,000 lbs on Figure 14-41 indicates uncovery of the top of the core and the bottom of the core is uncovered at a mass re-lease of about 400,000 lbs.

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- The information in Table I assumes that two accumulators are initiated at 600 psi and that one high pressure pump functions.

The flow rate is varied as a function of the pressure driving force.

The pressure transient in the vessel is calculated without the effects of the accumulators and the injection water flow superimposed on this calculation.

The applicant states that this is conservative since injection flow would condense steam in the annulus and lead to f aster pressure decay and more residual water inventory.

The " mini-mum quiet level" indicated in the table is an equivalent static core water level.

t Although the applicant has not completed the analysis and we have not completed our review, the information in Table I provides further assurance that the proposed design is practicable.

We have also compared the Babcock and Wilcox calculations for Oconee to calculations made by Westinghouse for the Indian Point II and Carolina Power and Light Reactors.

These comparisons are presented in Figures 1 through 3 of this report.

Figure 1 shows the pressure decay for the three systems for a large cold leg break.

The Indian Point 11 reactor is of similar size but of slightly lower average coolant enthalpy (575 Btu /lb vs 588 for Oconee).

The Carolinas reactor is smaller in coolant volume (9800 ft vs 11,800 for Oconee).

The results of a comparison of the 3 ft break size was similar to that presented in Figure 1 for the 8.5 ft break.

The time to uncover the top of the core is presented in Figure 2 for the three reactors for break sizes ranging from 14 ft to 3 ft The core uncovery time is shorter for the Oconee case than for the Westinghouse reactors for the cold leg break.

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. Figure 3 illustrates the major difference between the B&W and Westinghouse calculations.

For the small break (about 0.5 ft ) the pressure in the Oconee reactor is calculated to " hand up" at a higher pressure and to retain a larger fraction of the water inventory. Although both manufacturers use the FLASH code to calculate the blowdown transient, the Westinghouse calculation employs a constant steam-separation factor while B&W varies the amount of steam sepa-ration with pressure.

This assumption results, in the B&W calculation, in retention of a larger fraction of the water within the vessel and a higher vessel pressure.

The greater mass calculated to remain af ter blowdown is qualitatively substantiated by the results of LOFT semi-scale blowdown tests which resulted in substantially greater qucntities of water remaining in the vessel than were predicted on the basis of the FLASH code with a constant steam separa -

tion factor.

We urderstand from B&W and from Phillips monthly report that a modification is being made to incorporate a variable steam separation factor in the Phillips calculational model.

Although our review of the spectrum of breaks is not yet complete we believe that the results to date indicate a diligent effort by B&W and the applicant and that there is assurance that the analysis will be satisf actorily completed.

4.0 Conclusion The further information provided in Amendment 5 which verified oral commitments and amplified the previous submittals and the information obtained M-fi

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. orally on the spectrum of breaks analysis do not change our conclusions as forth in Report No. 2 to the ACRS dated June 16, 1967.

In summary, we set believe that the Oconee units can be built and operated without undue risk to the health and safety of the public.

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TABLE I

_ SPECTRUM OF COOLANT LINE BREAKS Minimum Positive Quiet Hot Hot Peak Moderator Break 1

Level Spot Spot Temp.

(Full-Sizg Coolt.nt (ft. above Uncovered Re-Covered Hot Spot Power -

(ft')

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'_ Core Bottom)_

_(Seconds)

_(Seconds)

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_ Seconds) 14 hot

-6.8 4.0 17.5 1950 2.1 S.5 hot

-5.4 6.2 20.6 1700 3.5 cold

-7.5 5.0 22.3 1740 3

hot

-2.3 17.6 33.5 965 1.6U cold

-5.6 13.4 32 1320 1

hot

+4.7 790 1.42U cold

+4.0 790 0.4 hot

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1.4 cold

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730 O.05 cold

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1/ Hot spot at about +3 ft.

2/ Scram assumed.

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