ML19220C144
| ML19220C144 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/15/1977 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904280292 | |
| Download: ML19220C144 (5) | |
Text
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N AUG 15 Cocket No. 5.1-2:-
MEMORANDUM FOR:
D. B. Vassallo, Assistanc Director for LWR's, OPM FRCM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, CSS SU5 JECT:
THREE MILE ISLAND UNIT NO. 2 - SER SUPPLEtiENT Plant Name: Three Mile Island Unit rio. 2 Docket No.: 50-320 Licensing Stage: 01 M'lestone No.: 27-21 Resconsible 3rancn 1 Project danager:
L'aR-4, H. Silver Systems Safety Branch Involved: Reactor Systems 3rancn Cescription of Review: SER Supplement Review Status: Awaiting Infomation Enclosed is a draft SER supplement for Three Mile Island Unit tio. 2.
The supplement addresses two of the outstanding issues--sump tests and ECCS small braak spectrum. The spectrum of breaks issue has been satis-factorily resolved. The vortex control issue is closed while verification of adequate NPSH is still open.
Open issues as presented in Section 1.7 of the SER are discussed belcw using SER identifying numbers.
(3) Page 1-7 Revised Steam Line Sreak Analvsis Tne applicant has cor.itted to modify the system at first refueling in con-formance with staff positions. The current issue involves obtaining a more definitive understanding of the potential cons <;uences of a steam line break during the first cycle of operation. Applicant e.vpects to provide analysis by mid-August using beginning of life reactivity control values wnich are sxpected to show a much more favorable subcritical behavior following scram.
The staff has excressed concern that the return to power indicated in earlier analysis, could lead to fuel damage. At present, B&W does not analyze this nortion of the transient for po:sible fuel damage. This topic is being persued on a generic basis with Babcock & Wilcox. A staff position will be fomulated following a meeting scheduled for August 9,1977.
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Contact:
Jim Watt. NRP 88 008 7 90428 07 ? U
Dis +ribution
- 2. 2. Vassalle Aug 15 197~(Cocker F1h toe xeacing NRR Reading Watt Chron (2) Page 1-e Reactor Euildinc Suno Test Results This is addressed in the sucolement.
Vortex control nas been denenstrated.
Adequate NPSH is yet to ce presented based on composite of experimentai results.
a (3) Page 1-8 Installation of flow measurecent devices to assure adeouate flow to limit baron concentratior The applicant must provide assurance that this scheme for flow measuremen-will not exoose the operator to excessive radiation. Exposure should oe censidered in the paths followed in getting to or from the sensors, the installation thereof, ano wnile ficw readings are being taken.
The staff also rquires assurance that appropriate measures will be taken to store and maintain this equipment; also that adequate instructions anc procedures are develooed for its use.
(3) Page 1-3 Review of Feed Line Break Analysis The staff expects the steam line break analysis to bound the feedwater line break consequences. As the same systers and coaconents are utilized in mitigating the consequences of botn secondary system breaks, tne staff believes it prudent to withhold judgment od the feedwater line breaks until a piping system satisfactory for steam line breaks has been defined.
(3) Page 1-9 Startua Pressure Protection The applicant intends to respond in mid-August 1977.
(4) Page 1-9 Emergency Core Coolino Analysis Hodifications This issue is closed in the enclosed supplement e ginal sic )51 Y p.1.tosa.
D. F. Ross, ir., Assistant Director for Reactor Safety Divisienodf Systems Safety
Enclosure:
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DRAFT SAFETY EVALUATION REPORT SUPPLEMENT Three Mile Island Unit No. 2 1.
Introduction The Safety Evaluation Report (SER) for Three Mile Isiand Unit 2, dated September 1976 stated that a supplemental report would be issued to update the SER in those areas where ou evaluation had not been completed. This supplement includes an uodate relative to the results of Reactor building sump tests and Emergency core cooling analysis modifications.
2.
Reactor Building Sump Tests 6.2.2, 6.3.4 In lieu of preoperational tests to demonstrate vortex control and adequate NPSH in the recirculation mode, the applicant com.itted to a model test program. A series of developmental tests were performed in a 1/3 scale model of the containment and sump.
The model included one half of the containment building floor with all obstructions.
Provisions were made to introduce flow from various locations to simulate possible break locations. Test variables also included water level, water temperature and flow rates from the two sump outlet pipes.
From visual observations it was concluded that the general layout was conducive to a strong circulation pattern in the rocm containing the sump.
Swirl patterns were observed to form which SS C'i G
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2 would attach to the outlet pipes forming an air entraining vortex.
Vortices shedding from obstructions were observed,to drift over the outlet pipes.
These would occassionally attach to an outlet pipe forming an air entraining vortex. A development program involving twelve configurations was conducted bercre vortex control was demonstrated. The staff witnessed tests with and without various vortex suppression devices in place in October 1976.
Tests of up to 50% blockage of screens and trash racks were performed. A report U) submitted "" the applicant concluded that the results may be safely projected to the full scale design.
The staff has re-viewed this report and agrees with its conclusion. On this basis, the staff concluded that adequate vortex control has been demonstrated.
Final assurance of adequate available NPSH for the low pressure injec-tion and containment spray pumps will be provided later.
Final pressure loss calculations will be based on loss coefficients determined during the scaled test program and system losses measured during other preoperational 'ests.
The staff will report on the NPSH results in a later suppleme-to the SER.
3.
Emergency Core Cooling Analysis Modifications 6.3.3 The applicant submitted the requested small and transition break analyses. The results of the calculations confimed that the worst case break had been previously identified in SAW-10103 and is properly Q O\\ '
3-4 presented in section 6.3.3 of the Safety Evaluation Report. No clad temperatures higher than 660 F were indicated in the additional analyses. The st;ff has completed their review of the methods and results concluding that the 3pectrum of breaks study has been satisfactorily completed.
Reference E.
References 1.
Hydrodynamics of Vortex Suppression in the Reactor Building Sump Decay Heat Removal System, Three Mile Island Unit 2.
Alden Research Laboratories, February 1977 ARL 46-77/M202FF.
2.
Letter from Steven A. Varga to Xenneth E. Suhrke, February 18, 1977.
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