ML20213C890

Forwards Request for Addl Info Re Flow Distribution,Computer Codes & Pressure Calculations,Required for Completion of Review

ML20213C890
Person / Time
Site:	Columbia
Issue date:	02/22/1979
From:	Phillips L Office of Nuclear Reactor Regulation
To:	Varga S Office of Nuclear Reactor Regulation
References
CON-WNP-0261, CON-WNP-261 NUDOCS 7903260409
Download: ML20213C890 (1)

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IEMORANDUM FOR: Steven A.VVarga. ChMf, Light Water Reactor Branch #4. DPM FROM:

• L.E. Phillips Section Leader, Reactor Analysis Section.

Analysis Branch, DSS l

SUSJECT:

- ROUND ONE QUESTIONS ON WPPSS NUCLEAR PROJECT NO. 2 Plant Name: WPPSS Nuclear Project No. 2

. Licensing Stago: OL Docket No.: 50-397 Milestone No.: 5-22 Responsible Branch and Project Manager: LWR-4. D. Lynch Systems Safety Branch Involved: Analysis Branch Description of Review: Q-l's Requested Completion Date: 2/2/79 Review Status: Aweiting Information Section 4.4 of the WPPSS Nuclear Project No. 2 FSAR has been reviewed by the Analysis Branch. We will require responses to the enclosed questions before we can complete our review. The Systems Analysis Section. Analysis Branch, will have Round One Questions on mass and energy release by March 15,1979.

1 Lauren E. Phillips, Section Leader Reactor Analysis Bection Analysis Eranch

- Division of Systees Safety l

Encleeure:

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R. Mattson R. Tedesco Z. Rosztoczy D. Ly.ach If

Contact:

B.W. Sheron NRR. X27588 Y

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FEB 2 2197g MEMORANDUM FOR: Steven A. Varga, Chief. Light Water Reactor Branch #4, DPM FROM:

L.E. Phillips, Section Leader, Reactor Analysis Section, Analysis Branch, DSS

SUBJECT:

ROUND ONE QUESTIONS ON WPPSS NUCLEAR PROJECT NO. 2 Plant Name: WPPSS Nuclear Project No. 2 Licensing Stage: OL Docket No.: 50-397 Milestone No.: 5-22 Responsible Branch and Project Manager:

LWR-4, D. Lynch Systems Safety Branch Involved: Analysis Branch Description of Review: Q-l's Requested Completion Date: 2/2/79 Review Status: Awaiting Information Section 4.4 of the WPPSS Nuclear Project No. 2 FSAR has been reviewed by the Analysis Branch. We will require responses to the enclosed questions before we can complete our review. The Systems Analysis Section, Analysis Branch, will have Round One Questions on mass and energy release by March 15, 1979.

.v w c Laurence E. Phillips, Section Leader Reactor Analysis Section Analysis Branch Division of Systems Safety

Enclosure:

As stated cc:

R. Mattson R. Tedesco Z. Rosztoczy D. Lynch

Contact:

B.W. Sheron, NRR, X27588 a.h

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m 221-1 i.

• 221.1 Section 4.4 contains no discussion of crud and its effect (None) on CPR and core pressure drop. Provide the ass'mptions u

used for amount of crud in design calculations and the sensitivity of CPR and core pressure drop to variations in the amount of crud present. Also provide data suppc. ting the assumption on crud thickness and discuss how crud build-up in the core would be detected.

• 221;2 The GEXL data base (for the approved correlation) is for (4.4.2.2.1) 7x7 and 8x8 one water rod bundles. No substantial data base has been provided to support the 8x8, two water rod design. The GEXL correlation must be demonstrated to be applicable to the new 8x8 design, by comparison to appli-cable data, prior to issuance of an operating license for WNP No. 2.

Alternatively, the MCPR limit may be increased i

by 0.05 to accomodate GEXL uncertainties.

• 221.3 You state on page 4.4-7. that "There is reasonable assurance, l

(4.4.2.5) therefore, that the calculated flow distribution throughout the core is in close agreement with the actual flow distri-bution of an operating reactor." Does this refer specifically to WNP No. 2 calculations? What operating reactor was used for the data comparison?

• 221. 4 Your flow distribution discussion does not address uncertainties (4.4.2.5, on the flow distribution or the effect of channel flow uncertainty.

Table 4.4-6) coupled with other uncertainties on the MCPR uncertainty. Also -

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Table 4.4-6 does not address flow distribution undirtainties.

Provide this.information.

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• 221. 5 Page 4.4-21 states " Analytical models of the individual flow (4.4.4.5) -

paths were developed as an independent check of the tests.

When using these models for hydraulic design calculations, nominal drawing dimensions are used." Provide the assumptions and equations comprising the model and a comparison of model predictions with data.

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• 221. 6 What fraction of the fuel bundle flow is " water rod flow"?

• 221. 7 Page 4.4-18 of the FSAR states that "the nominal expected (4.4.4.5) bypass flow fraction is approximately 10 percent." What is the calculated bypass flow fraction for WNP No. 2 and what is its uncertainty?

• 221.8 What is the name of the computer program cited in this section?

(4.4.4.5)

Provide references which document the code.

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• 221. 9

.You state that the stability analyses performed in Section 4.4.4.6 l

(4.4.4.6, and for Figure 4.4-6, were perfonned "at the most limiting i

Fig'ure4.4-6) condition that occurs at the end of core life, with power peaking l

• Question asked on one or more other dockets.

221-2 toward the bottom of the core...".

Are the typical values of core stability provided in this section based on end of core life with power peaking toward the bottcm of the core?

If not, provide values of decay ratio for this condition.

Provide the power profile and the void reactivity coefficient used for the analysis.

• 221.10 In discussing the FABLE code on pages 4.4-26, you state that (4.4.4.6)

"As new experimental or reactor operating data are obtained, the model is refined to improve its capability and accuracy."

This means that comparison of old versions of the model with data, as given in Figure 4.4-4, are meaningless for WNP No. 2 if it has been analyzed with an updated version. Are the comparisons of the model with data, as given in Figure 4.4-4, based on the same version of the model as was used for WNP No. 2? If not, provide comparisons using the model.

In addition, provide a description of the code or reference a prior licensing submittal (other than the KAPL reports on STABLE).

• 221.11 On page 4.4-27, the REDY code is referenced as the model used to (4.4.4.6) perform system stability calculations. You also state that the model is periodically refined as new experimental or reactor operating data are obtained.

Is the version of REDY used for WNP No. 2 described in NED0-10802? If not, describe the changes.

• 221.12 We will require that a loose parts monitoring system be installed (4.4.4.6) and operational prior to startup testing. Therefore, a description of the system is needed for our evaluation prior to issuance of an operating license.

Guidance regarding the design requirements for a LPM system can be found in draft Regulatory Guide 1.133

-(Loose-Part Detection Program for the Primary System of Light-Water-Cooled-Reactors). Provide a schedule for submittal of your LPM system description.

• 221.13 Table 4.4-6 describes uncertainties used in the statistical (Table 4.4-6) analysis which is performed to establish the fuel cladding integrity safety MCPR limit. Provide a discussion of and reference where possible the experimental data bases used te derive the uncertainty values listed.

In particular, describe the applicability of these values to the 8x8, two-water rod assembly design.

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