ML20052A693

Forwards Request for Addl Info Re PSAR Nuclear Design

ML20052A693
Person / Time
Site:	Clinch River
Issue date:	03/23/1982
From:	Check P Office of Nuclear Reactor Regulation
To:	Longenecker J ENERGY, DEPT. OF
References
NUDOCS 8204290038
Download: ML20052A693 (6)

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l Distribution Docket Filej NRC PDR March 23, 1982 Local PDR

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R. Stark /.C Docket No.: 50-537 C. Thomas'y s.

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Mr. John R. Longenecker T. Wehneri.

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Licensing and Environmental Coordination W.Fostery

/0 "er Clinch River Breeder Reactor Plant P. Shuttleworth xg{g U. S. Department of Energy, NE-561 E. Tomlinson" Washington, D.C.

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SUBJECT:

CLINCil RIVER BREEDER REACTOR PLANT, REQUEST FOR ADDITIONAL INFORMATION As a result of our review of your application for a construction permit for the Clinch River Breeder Reactor Plant, we find that we need addi-tional information in the area of Core Performance. Please provide your final responses by May 15, 1982.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.

If you desire any discussion or clarification of the information requested, please contact R. M. Stark, Project Manager (301) 492-9732.

i Sincerely.

Original Signed by Paul S. Check Paul S. Check, Director CRBR Program Office j

Office of Nuclear Reactor Regulation i

Enclosure:

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Dr. Cadet H. Hand, Jr., Director Barbara A. Finamore Bodega Marine Laboratory S. Jacob Scherr University of California Ellyn R. Weiss P. O. Box 247 Dr. Thomas B. Cochran Bodega Bay, California 94923 Natural Resources Defense Council, Inc.

Daniel Swanson 1725 I Street, N.W.

Office of the Executive Suite 600 Legal Director Washington, D.C.

20006 U. S. Nuclear Regulatory Commission Eldon V. C. Greenberg Washington, D.C.

20555 Tuttle & Taylor 1901 L Street, N.W.

William B. Hubbard, Esq.

Suite 805 Assistant Attorney General Washington, D.C.

20036 State of Tennessee Office of the Attorney General L. Ribb 450 James Robertson Parkway LNR Associates Nashville, TN 37219 Nuclear Power Safety Consultants 8605 Grimsby Court William E. Lantrip, Esq.

Potomac, MD 20854 City Attorney Municipal Building P. O. Box 1 Oak Ridge, TN 37830 George L. Edgar, Esq.

Morgan, Lewis & Bockius 1800 M Street, N.W.

Washington, D.C.

20036 Herbert S. Sanger, Jr., Esq.

General Counsel Tennessee Valley Authority Knoxville, TN 37902 Chase Stephens, Chief Docketing and Service Section l

Office of the Secretary U. S. Nuclear Regulatory l

Commission Washington, D.C.

20555 Raymond L. Copeland l

Project Management Corp.

l P. O. Box U l

Oak Ridge, Tennessee 37830

MAE 22 EG2 QUESTIONS TO THE APPLICANT ON PSAR NUCLEAR DESIGN CS 491.1 In Section 3.1, "Conformance with General Design Criteria," there (3.1) is no design criterion comparable to 10 CFR 50, Part A, Criterion 28, " Reactivity Limits." Why is this general design criterion not a part of Section 3.17 Are appropriate limits on the potential amount and rate of reactivity increase discussed in this criterion going to be quantitatively specified?

CS 491.2 Please explain why Criterion 29, " Protection against Anticipated (3.1)

Operational Occurrences," of 10 CFR 50, Part A is not a criterion in Section3.1 of the PSAR, CS 491.3 The specific speed of response requirements does not seem to be (4.2.2.1) presented in Section 4.2.2.1.3.

Where is it presented?

CS 491.4 Why aren't there sufficient calculational uncertainities listed (4.2.3.3) to enable one to judge the fragility of the PCRS and SCRS scram conclusions?

CS 491.5 In one case you claim one dollar of reactivity is inserted within (4.2.3.3)

.3 seconds after rod motion begins. Why doesn't this agree with the presentation in Figure 4.2-122?

Why aren't the treatments of PCRS and SCRS scrams parallel?

CS 491.6 Are the PCA position indicators and dampers also being tested?

(4.2.3.4)

Which of the tests mentioned in Section 4.2.3.4.1 have been completed and documented?

CS 491.7 The Source Range Flux Monitoring (SRFM) System is described in (4.3.2.1)

Section 4.3.2.1.5.

The applicant is requested to provide the minimum acceptable neutron flux at the detector, the maximum acceptable gamma-ray dose rate at the detector, and the calculated neutron flux and gamma-ray dose rate at the detector.

Al so provide a description (method, models, etc) of how these calcul-ations were made.

In this same section it is indicated that SRFft detector operating characteristics will be experimentally verified in ZPPR critical experiments which will mock up the actual CRBR installations as close as practicable.

In addition transport calculations will be l

employed to account for neutron streaming effects in the cavity I

which cannot be mocked up in the ZPPR.

Please respond to the following:

(1)

Describe the transport calculations (method, models, etc) to be perfonned.

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(2)

What plans are being made to verify that the planned ZPPR experiment is an accurate test of the SRFM detectors for the actual CRBR?

(3)

Are transport calculations being planned to determine the radiation environment at the SRFH detectors for the actual CRBR configuration as well as the ZPPR configurations?

If so, describe these planned calculations, e.g., method, model, etc.

CS 491.8 The applicant indicates that power distribution limits are (4.3.2.2) derived from maximum allowable peak heat generation rates for nominal and anticipated operational conditions, which combined with the rod mechanical and thennal parameters, assure that incipient fuel melting does not occur in the fuel pellet with peak power. What are these specific, quantitative, power distribution limits? What are these specific, quantitative, power distribution limits? What are the maximum (quantitative) allowable peak heat generation rates (linear power) for nominal and anticipated operational conditions? What clad and coolant temperatures correspond to these maximum peak heat generation rates?

CS 491.9 What type of instrumentation is planned to allow detection of (4.3.2.2) flux (power) tilts in the core at operational levels?

491.10 In Section 4.3.2.2.9c clarify the expression "3s equivalent (4.3.2.2) uncertainties".

CS 491.11 The Doppler reactivity constant has been computed for temperatures (4.3.2.3) above 2100 degrees K (e.g., 3000, 4000 and 5000 degrees K). What assumptions were made, or how uncertain are these high temperature coefficients when your basic 30-group library probably only gives temperature dependent cross sections to 2100 degrees K? Can you argue that any safety considerations only very weakly depend on accurate high temperature Doppler coefficients?

CS 491.12 Uncertainty in the Doppler constant has been based principally (4.3.2.3) on the analysis of the SEFOR Core I and II experiments perfonned by GE.

Is the extrapolation from SEFOR to CRBR simply from one reactor to another or between reactors and methods?

If the PSAR l

l method for calculating the Doppler constant is different from that used by GE to calculate the SEFOR Doppler constant, please provide a comparison of the two methods.

Also provide justification as to why the uncertainty of the GE method should be accepted as that for the CRBR method.

The PSAR also indicates that Ref. 9 provides l

data for extrapolation from SEFOR to LMFBR power reactors accounting for differences in core composition, core spectrum, etc.

Provide justification that the SEFOR to LMFBR power reactors (1973 l

type) extrapolation should be identical to the SEFOR to CRBR extrapolation.

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CS 491.13 It appears that Expansion and Bowing Reactivity Coefficients are (4.3.2.3) caaputed by integrating core movenents over axial and radial worth curves.

Has the reactivity of the small core reconfiguration been accounted for?

As the core heats up the structural and fuel naterials expand increasing the size of the core.

Does the mass of sodium necessarily increase in the expanded core? That is, is it possible that sodium expands enough with higher temperature that its mass in the core stays the same or even decreases?

CS 491.14 There appears to be insufficient control rod neutronics.

Provide (4.3.2.4) descriptions of calculations and data that characterize control rod burnup, nanagement, and flux and power distributions.

CS 491.15 In order to establish the criticality of the hot-full power CRBRP, (4.3.2.7) why not perform a direct K calculation at hot-full power conditions?

CS 491.16 Why were the minimum critical configuration calculations performed (4.3.2.7) with only P cross sections?

g CS 491.17 Section 4.3.2.9 and Table 4.3-35 report neutron flux and fluence (4.3.2.9) data at locations within the core and in structural components outside the core, i.e., core barrel and reactor vessel.

Please respond to the following question regarding the ex-core calculations:

1.

What was the calculational nethod used?

2.

What was the geometrical model used and what modeling approximations were made?

3.

What was the neutron source used in the calculations? What was the basis for this source and what approximations were made in incorporating it into the ex-core calculations?

4.

What procedure was used to insure that the ex-core neutron fluences, over plant life, are conservative and represent-ative of the worst points on the core barrel and pressure vessel?

5.

What is the accuracy of the calculated fluxes?

6.

What are the limiting flux (fluence) values for the core barrel and pressure vessel?

CS 491.18 Cite references where your procedures (methods, codes, models, (4.3.3) and data) have been clearly compared with some other laboratory's procedures for the calculation of Doppler coefficients, sodiun void coefficients, control rod worths, power and flux distri-butions, material worths, burnup, bowing reactivity coefficients, power coefficients, temperature defects, startup coefficients, etc.

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Some of the fundamental neutronics parameters of the CRBR are Doppler coefficients, sodium void coefficients, control rod worhts, power and flux distribution, burnups, and bowing reactivity coefficients.

Identify the particular safety consider-ation that you feel is the nost impacted, limited, or made uncertain by each of the above parameters.

Then, identify the most uncertain link in the calculational chain for each of the parameters.

CS 491.19 The applicant's method of calculating the CRBR and applying (4.3.3) biases derived from critical assembly investigations is similar to that perfonned for the FFTF.

Since startup measurements on the FFTF have been completed, what investigations have been performed to discover which methods and calculations did not stand up well for the FFTF and hence may be suspect for the CRBR?

Does the 30-group neutron e.ross section library (your basic starting point) have a reference? Can this exact library be obtained in order to reproduce any of your calculations?

CS 491.20 In this section you state that ZPPR-4 control rod bank worth (4.3.3.7)

C/E values range from 0.95 to 1.04.

Why do these numbers differ from those presented in Table 4.3-40?

CS 491.21 Section 15.1.2 describes qualitative core limits for normal (15.1.2) operations, transients, and accidents. Are specific, quantitative, design limits going to be specified?

If not, please justify why qualitative limits are preferrable to quantitative limits.

CS 491.22 No sodium boiling is used as a limit for extremely unlikely (15.1.2) faul ts in Table 15.1.2-2.

This limit does not appear to have a specific value (temperature) as it depends on the coolant pressure.

If this criterion results in a variable quantitative temperature limit for the various events considered why is the corresponding design limiting coolant temperature (and its basis) not specified for each event?

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