ML19270H157

Forwards Requests for Addl Info Re Reactor & Sys Analysis & Containment Subcompartment Analysis.Also Forwards Marviken Rept to Be Used Re Justification for C-E FLASH-4A Methods. W/O Rept

ML19270H157
Person / Time
Site:	Waterford
Issue date:	05/18/1979
From:	Baer R Office of Nuclear Reactor Regulation
To:	Aswell D LOUISIANA POWER & LIGHT CO.
References
NUDOCS 7906230084
Download: ML19270H157 (9)

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UfaTED STATES j

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Docket No. 50-382 Pr. D. L. Aswell Vice President, Power Production Louisiana Power and Light Company 142 Delarande Street

aw Crleans, Louisiana 70174

SUBJECT:

WATERFORD UNIT 3 - REQUEST FOR ADDITIONAL INFORMATION Enclosed are requests for additional information concerning reactor and systems analysis and an additional request concerning containnent sub-compartment analysis.

In order to maintain our review schedule, you should provide complete responses to the 221 and 222 series of questicns by August 24, 1979 and to question 022.23 by July 20, 1979.

Questicn 222.1 refers to a Marviken report, one copy of which is enclosed b

(Enclosure 2).

The Marviken report is referenced because the data in the report demonstrates that, for two-phase flow, the mass flow rate per unit area for orifices is higher than for pipes.

We are forwarding the Marviken report so that it may be used to respond to question 222.1 regarding justification for the CE FLASH-4A methods for prediction of flow through orifices. The Marviken report is being for-warded under the following conditions:

1.

The report will not directly, indirectly, or otherwise be used, or duplicated except as may be necessary to acccmplish the task set forth above; 2.

The report will only be transferred, disseminated, disclosed or otherwise comunicated, in whole or in part, to people or organi-zations involved in the task; 3.

The report will not be transferred, disseminated, disclosec or otherwise comunicated, in uncle or in part, :: ar.y cerson or crganization out'sice of the Lnited St ates,

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'r. D. L. Aswell YIf ; 5 37; 5.

The report will not be published, either in w.bcle or in part, prior to publicaticn in the open literature by the Marviken Project or until authorization is cbtained from the Farviken.

Project Board to so publish; and 6.

All of the above conditions shall be made a part of any transfer permitted under 2 above.

Sincerely,

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debert L. 'Baer, Ch ief Light Water Reactors Eranch No. 2 Division of Project Manacement

Enclosures:

1.

Request for Additional Information 022.23 221.3-221.8 222.1-222.8 2.

Marviken MXC-207, Interim Report - Results f rom Test 7 (MX3-49, August 1978) ccs: See next page 2247 313 e

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e. :. L. Aswell ces w/ enclosure 1:

I Malcolm Stevenson, Esq.

Mcnroe & Lemar.n 1424 Whitney Building New Orleans, Louisiana 70130 Mr. E. Bl ake Shaw, Pittman, Potts and Trowbridge 1800 M Street, N. W.

Washington, D. C.

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Mr. D. B. Lester Production Engineer Lcuisiana Power & Light Company 142 Celarende Street New Orleans, Louisiana 70174 Lyman L. Jones, Jr., Esq.

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Gillespie & Jones 910 Security Homestead Building 4900 Veterans Memorial Boulevard Metairie, Louisiana 70002 Luk e Fontana, Esq.

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)1A Gillespie & Jones LL

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I4 824 Esplanade Avenue New Orleans, Louisiana 70116 Stephen M. Irving, Esq.

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Cre American Place, Suite 1601 Eaton Rcute, Louisiana 70825 w/ o enclosures:

Mr. Lars Carlbom Marviken Board Chairman Studsvik Energiteknik AB Marviken Experiment 61024 Vikbolandet Swecen

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RECUEST FOR ADDITIONAL INFORMATI0t!

WATERFORD UNIT 3 022.0 CONTAINMENT SYSTEMS 022.23 Your response to Questien 022.2 is inccmolete.

Provide (6.2.1) a response that considers pipe breaks not only in the reactor coolant systems but also in the feedwater and steam lines.

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222-1 379 220.0 Systems Analysis Section. Analysis Branch 222.1 Since pipe restraints are provided for large primary system (6.2.1) lines that penetrate the reactor cavity and steam generator compartment, limited off set type breaks were analyzed.

For breaks of this type the break geometry may resemble an orifice in the broken pipe.

Data by a number of investigators has demonstrated that for two-phase flow the mass flow rate per unit area for orifices is higher than for pipes. Justify that the CE FLASH-4A methods are conservative for prediction of flow through orifices. Ori-fice and short nozzle flow data is found in (1) NEDO-13418,

" Critical Flow of Saturated and Subcooled Water at High Pressure,"

by Sozzi and Surtherland, July 1975, (2) " Blowdown Flow Rates of Initially Saturated Water", by V. Simon, Topical Meeting on Water-Reactor Safety, Salt Lake City, Utah, March,1973, (3) Choked Expansion of Subcooled Water and the I.H.E. Flow Model", by R.L. Collins, Journal of Heat Transfer, May 1978, and (4) "The Marviken Full-Scale Critical Flow Tests Interim Report; Results from Test 7".

222.2 Provide justification for not perfoming analyses of postulated (6.2.1) steam and feedwater ruptures within the steam generator compartment.

See also question 022.23.

.2 Provide an analysis of the effect of nitrocen release from the (6.2.1) safety injection tanks on the containment or:::ar: follcwing a postulated double ended pump suction break LOCA.

Include a table of nitrogen flow and enthalpy as a function of time.

222.4 Figures 6.2-13c and 6.2-13d indicate that feedwater flow is (6.2.1) terminated following a postulated main steam line break in approximately 7 seconds. Table 15.1-13 states that the main feedwater isolation valves close at 22.1 seconds. Justify this apparent inconsistency.

222.5 The analysis method for the inadvertent opening of a steam (15.I'.l.4) generator dump valve has not been presented by the reactor vendor nor approved by NRC. Therefore, the following infor-mation is required concerning analysis of this event.

a.) A detailed nadalization diagram used to analyze the transient.

b.)

A complete discussion cf the calculatic al method and models used in the evaluation of the transient.

c.)

A detailec #1:w :iag-am f;r tne tri ary and sacer.da y systans icantif iri; ill c0mCCne.~.5 C;r.5icered 4-

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Calculati0nal model use0.

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222-2 WY.1 e ryg 222.6 The analysis method for the steam line break has not been pre-(15.1.3.1) sented by reactor vendor nor approved by NRC. Therefore, the following information is required concerning analysis of this event.

a.) A detailed discussion of the calculational method, including the system noding diagram, used in the evaluation of the accident including all the codes used.

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b.) A detailed flow diagram for the primary and secondary system used in the calculational model and identify all the com;onents considered in the analysis.

c.) Describe in detail hcw the core thermal-hydraulic effects were evaluated including the calculation of DNSR correlations used, and assumptions regarding mixing of loop flows.

d.) Provide the calculational method used to generate transient

..f.-A axial and radial power distributions.

Describe how these D"

peaking factors were considered in the therral-hydraulic

. calculations.

e.) Provide calculational model used to determine the nuclear and thermal-hydraulic effects for the first 20 seconds for

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both BOL and EOL conditions from full power.

Include the changes in moderator density due to rapid depressurization of the primary system.

f.) Describe in detail how the time dependent pressure drop in the average and hot fuel channels was calculated.

g.) For the high pressure safety injection system, describe the flow path into the core, the calculational model used for g

evaluating the time for the fluid to reach the center of

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the core, and method used for determining the resultant reactivity feedback.

Provide the resultant reactivity versus time for the cases analyzed.

h.) Provide experimental verification for the analytical models used to evaluate the steam generator performance (such as froth level, heat transfer coefficients, water and steam flow rates).

i.)

Describe in detail the calculaticnal model used to calculate steam discharge rate from tne broken steam line and justify tne initial licuid mass inventory assumed in the steam generat:r.

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222.7 The analysis rethod for the feedwater line break has not been (15.2.3.1) presented by reactor vendor ner approved by NRC. Therefore, the following information is required concerning analyses of this event.

a.) A discussion of the calculaticnal methods including the system noding diagram, ased in the evaluation including all the codes used.

b.) A detailed flow diagram for primary and secondary systems used in the calculational model and identify all the compo-nents.

c.) The calculational model used to obtain the discharge rates out of the feedwater line including the mass inventory in the steam generators during the feedwater line break accident.

222.8 The analysis method for the steam generator tube rupture has not (15.6.3.2) been presented by reactor vendor nor approved by NRC. Therefore, the following information is required concerning analysis of this event.

a.) The details of the nadalization model used in the analysis.

b.) The model used to calculate the steam generator tube discharge rates.

c.) A description of the cceputer program used to calculate the primary and secondary system pressures and the results for a typical calculation.

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MAY ! 3 7;g 221.0 Reactor Analysis Section, Analysis Branch 221.3 Provide an explanation of hcw the effects of possible (4.4.4.2) crud deposits on the core ficw and pressure drop are included in the thermal-hydraulic design.

Provide the magnitude of the increase in the core pressure drop i

due to core crudding. Provide the assumptions used to estimate the crud buildup.

Provide a description of the instrumentation available and the surveillance requirements and procedures which would alert the reactor operator to an abnormal core ficw or core pressure drop during steady-state opera-tion.

221.4 The. staff has developed interim criteria for evaluating (4.4.2) the effect of rod bow on DMB for application to the Ccmbustion Engineering 16 x 16 fuel assemblies. Use of the staff report " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calcu-lations for Light Water Reactors", dated February 16, 1977 presents an acceptably conservative treatment of rod bowing.

The applicant should present,.in the Technical Specifications, its proposed thermal margin reduction.

221. 5 Provide the results of fuel assembly tests applicable to tha (4.4.4.2) Waterford 3 design to verify the values of the less coefficients for the upper and icwer end fittings and spacer grids.

221.6 The staff is performing a generic study of the hydraulic stability (t.4.4.5.3)of light water reactors, including the evaluaticn methods used for Waterford 3.

The results of the staff study will be applied to the acceptability of the stability methods new in use by reactor vendors. Piovide a commitment to take 'any actions requi red by the results of the s taff's~ stucy.

221.7 It is the intent of the staff to risiew the Loose Parts Monitoring

( 4. a. 6.1 ) System for Waterford Unit 3 against the guidelines of Regulatory Guide 1.133 (draft).

In this regard, the following information, in addition to that supplied in the response to Question 221.1, shculd 53 incluced:

(1) A descriotion of the program ca:: ability for distinguishing between a lcose cart and nor al background noise.

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4 (2) Quantification of the online sensitivity of the system in terms of mass and kinetic energy, and

( 3') Discussion of the training program for plant personnel.

221.8 With regard to the Core Protection Calculator System, we require that the following information be provided:

1.

Identification of the Software Specifications upon which the Waterford 3 CPCs are based, 2.

the data base constants used in the CPC algorithr.s,

3. ' the test report for verification of the Waterford 3 CPC i

software, and

4. modifications to the proposed technical specifications.

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