ML20235C982

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Div of Reactor Licensing Rept to ACRS in Matter of Comm Ed Application for CP for Quad Cities Units 1 & 2, Supplemental Rept
ML20235C982
Person / Time
Site: Quad Cities, 05000000
Issue date: 12/07/1966
From:
US ATOMIC ENERGY COMMISSION (AEC)
To:
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ML20235B311 List: ... further results
References
FOIA-87-111 NUDOCS 8709250076
Download: ML20235C982 (6)
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y OFMCFAL4dSEtNL(7 DECy 79gg U. S. ' ATOMIC ENERGY COMMISSION DIVISION OF REACTOR LICENSING L

REPORT TO ADVISORY COMMITTEE SAFEGUARDS ON REACTO IN_THE MATTER OF_

COMMONWEALTH EDISON COMPANY APPLICATION FOR CONSTRUCTION PERMIT FOR TH DOCKETS E OUAD-CITIES UNITS 1 AND 2 _

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. . 4 Note by the Director of the Divisioneactor of R Licensing The attached report has been prepared by th consideration by the ACRS .at its meeting. December 1966e Division of React t yv 1

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I. Introduction The regulatory staff report regarding Commonwealth Edison's Quad-Cities .

application for a construction permit was sent to the Committee on November 21, 1966. Since writing this report, the staff has received Amendment No. 4, dated November 28, 1966. This present report summarizes the results of orr continued review of the Quad-Cities application.

II. Amendment No. 4 Amendment No. 4 was prepared principally to supply more detailed information on the engineered safeguards in response to the ACRS request in its letter on Dresden Unit 3, dated August 16, 1966. This additional system and equipment design data, functional performance evaluation, and information on physical arrangement for each of the engineered safeguards systems is under review by the staff. Our review of the details of these systems will continue as construction of the plant progresses. This information, although supplied for the Dresden Unit 3 engineered safeguards, is applicable to the Quad-Cities reactors except for che redundancy indicated in the core spray systems discussed in our previous report. Analysis to date of the new information supplied on the safeguards systems is presented in section IV below.

In addition to information on engineered safeguards systems, Amendment No. 4 contains further information submitted in response to questions raised at the ACRS subcommittee meeting of November 17, 1966. The subcommittee requested information on (1) corrosion protection of,the suppression pool, (2) tornado destruction of the f acility stack, (3) turbine rotor f ailure, (4) fabrication, inspection and test requirements for conventional systems versus General Electric BWR primary system components, and (5) criteria for detection of primary system leakage within

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_OFMGAL USE-ONET the drywell. No new information was supplied which would change the conclusion in our previous report that the plant can be built and operated safely.

III. Missile Protection Capability of Drywell Shield Plug On the basis of the staf f's analysis and conversations with Mr. James Proctor of the Naval Ordnance Laboratory, we believe that the drywell shield plug would withstand any credible missile caused by turbine overspeed f ailure or tornado-carried sections of the facility stack without loss of function. Mr. Proctor had available to him the analyses supplied in Amendments 3 and 4 to the application t

and indicated that the raethod used to calculate miscile penetration in the concrete was acceptable and that the ultimate capability of the shield plug would be about 16 million ft.lb. per square foot of impact area. The highest energy missile originating from the turbine or stack would have signi*icantly less energy than 16 million f t.lb. per square foot on impact with the concrete.

IV. New Information on Safeguards Systens A. Water Source Four screened headers till be previded in the suppression poc1, three cf which are conservatively calculated to be able to supply the water for the core spray, LPCI and HPCI systems simultaneously. We believe that this arrangement is accept-able since multiple paths from the cooling water source are provided. In addition, these systems could be supplied from the condensate storage tank if the suppression pool water source was for some reason not available.

B. LPCI As recommended in our previous report, the applicant has now provided isolation capability on the line which cross-ties the two halves of the LPCI system. The staff believes that this provides additional assurance that at least part of the system will be operable in case of a single component failure.

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@EHGHAL4JSErONLlt V.- Desien' Basis Nuclear Excursion The following' discussion presents a summary of the consequences of energy re' leases'from~ rod ejection or dropout accident.

To insure the integrity of the reactor . vessel and vital vessel internal structures, the severity of potential nuclear excursions is limited by engineered safeguards. The table below illustrates the importance of'the control rod velocity limiter and the rod thimble supports to limit excursion energy release.

Peak energy. releases for other rod worths may be obtained from Figure 68 of the' Plant Design and Analysis Report.

Peak Calculated Rod Velocity and Energy Release with Control Rod Worth of 2.5%

With Velocity Limiter Without Velocity Limiter Rod Velocity tEnergy Release Rod Velocity Energy Release

1. Rod dropout 5 ft/see 200 cal /gm 20 ft/sec 375 cal /gm
2. Rod ejection 10 ft/sec ---

15 ft/sec ---

with thimble supports

3. Rod ejection 7 30 ft/sec 7400 cal /gm >30 ft/sec 7400 cal /gm without i thimble supports i It should be noted that case '2' in the above table (the rod ejection with J thimble supports) does not result in a significant amount of reactivity insertion since the rod is restrained f rom moving more than a few inches by the thirble j supports.

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The design basis nuclear transient has been taken to be the dropout of a rod worth 2.5% at 5 ft/second. The maximum rod worth is limited to this value by pro- k cedural controls backed up by a rod-worth-minimizer computer which limits the rod

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_4 patterns available to the operator such that no rod, if dropped, could insert in excess of 2.5% in reactivity. The nuclear transient from this rod-dropout accident results in a calculated peak fuel enthalpy of 200 cal /gm (i 25 cal /ge).

Fuel melting takes place between 220 and 280 cal /gm. The applicant has stated that no mechanical damage would result f rom this accident because the major fraction of the energy would be stored in fuel which maintains its clad integrity.

Excursiont which result in peak fuel enthalples between 300 and 400 cal /gm are calculated to result in pressure gradients on the order of 10 to 100 psi.

This energy is transferred to the water by fuel which fragments or melts and is rapidly expelled into the water. The table below summarizes the estimated damage from these excursions.

Resultant Damage to Vessel Internals from Nuclear Excursions Peak Fuel Enthalpy, cal /gm __ Component Damage 375 Channel Box Deformation of the channel box sufficient to interfere with control rod motion.

3?5 Opper Plenum Hold-down bolts yield Head

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400 Channel Box Yields and splits 400 Upper Plenum Hold-down bolts yield and Head plenum head moves 3-5 feet.

For nuclear excursions with large amounts of fuel above 425 cal /gm (vaporiza-tion threshold), the integrity of the primary system itself would be threatened.

As stated above, excursions of this magnitude are avoided by procedural controls on rod worths and by engineered safeguards.

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The staff is continuing its review of the excursion model and damage threshold assumptions used by General Electric. A seminar will be held December 19, 1966, to discuss the company-private document, Nuclear Excursion Analysis, APED-5797, which was recently made available to the staf f. We believe that the engineered safeguards provided will insure that potential n'uclear excursions will be limited to peak enthalpies well below the calculated threshold for major mechanical effects.

VI. Conclusion i

Our conclusion remains unchanged from that stated in our previous report on these reactors; namely, that these reactors can be built and operated without undue risk to the health and safety of the public.

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