ML20236P779

Forwards Request for Addl Info Re Rev 0 to NSPNAD-8608, Reload Safety Evaluation Methods for Application... & Rev 0 to NSPNAD-8609, Qualification of Reactor Physics Methods. Response Requested by 870914

ML20236P779
Person / Time
Site:	Prairie Island
Issue date:	08/05/1987
From:	Scaletti D Office of Nuclear Reactor Regulation
To:	Musolf D NORTHERN STATES POWER CO.
References
NUDOCS 8708130042
Download: ML20236P779 (8)

Text

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AUG 0 51987 ,

E

ket Nos. 50-282 DISTRIBUTION

and 50-306 *Ipocket Files,@ 0GC-Bethesda NRC PDR 'DScaletti Local PDR Edordan PDIII-3 r/f JPartlow l GHolahan ACRS(10)

Mr. D. M. Musolf, Manager DWigginton PKreutzer Nucle u Support Services Northern States Power Company 414 Nicollet Mall Midland Square, 4th Floor Minneapolis, Minnesota 55401

Dear Mr. Musolf:

L

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - NSPNAD-8608 AND NSPNAD-8609 During)the course of the staff's review of your topical reports NSPNAD-8608,(Rev.0 ,

Nuclear Generating Plant," and NSPNAD-8609, (Rev. 0), " Qualification of Reactor '

Physics Methods," the need for additional information has been identified.

To allow the staff to complete its review in a timely manner, please provide

! the additional information requested in the enclosure to this letter by September 14, 1987.

Should you have any questions regarding this request, please contact me at (301)492-8146.

The information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under Pub. L.96-511.

l Sincerely, Dino C. Scaletti, Project Manager Project Directorate III-3 4 Division of Reactor Projects l

Enclosure:

As stated cc: See next page l- '

\

l- Office: LA/PpIII-3 PM P 3 PD PDIII-3 i Surname: ' P6h4tzer D sti/tg DWig inton i l

Date: 08/5 /87 08/3/87 08/ /87  !

I 970013o042 070005 l' PDR ADOCK 05000202 P PDR

- ______J

I4 l - .

'Mr. D. M. Musolf Northern States Power Company Monticello Nuclear Generating Plant h

cc:

Gerald Charnoff, Esquire Commissioner of Health l- Shaw, Pittman, Potts and Minnesota Department of Health Trowbridge 717 Delaware Street, S. E.

2300 N Street, NW Minneapolis, Minnesota 55440 Washington, D. C. 20037

~0. J. Arlien, Auditor U. S. Nuclear Regulatory Comission Wright County Board of Resident Inspector's Office Commissioners Box 1200 10 NW Second Street P Monticello, Minnesota 55362 Buffalo, Minnesota 55313 Plant Manager i Monticello Nuclear Generating Plant

( Northern States Power Company Monticello, Minnesota 55362 Russell J. Hatling Minnesota Environmental Control Citizens Association (MECCA)

Energy Task Force 144 Melbourne Avenue, S. E.

Jiinneapolis, Minnesota 55113 g Dr. John W. Fennan Minnesota Pollution Control Agency 520 Lafayette Road St'. Paul, Minnesota 55155-3898 Regional Administrator, Region III U. 5. Nuclear Regulatory Comission 799 Roosevelt Road Glen Ellyn, Illinois 60137

p .

REQUEST FOR ADDITIONAL. INFORMATION Reload Safety Evaluation Methods for Application Topical Report

Title:

to the Monticello Nuclear Generating Plant Tupical Report Number: NSPNAD-8608 (Revision 0)

Topical Report Date: September 1986

1. Does the DYN0DE-B fuel rod gap heat transfer coefficient account for expo -

sure and fuel temperature dependence and, if not, what error does this siin-plification introduce?

?.

What direct moderator heating fraction is used and is this value conserva-tive for the transients to be analyzed (Table 4.1-1, MSIV closure, etc.)?

3 Comparisons have been presented for the DYN0DE-B a How do DYN0DE-B and NDH compare with respect to Doppler reactivity?

4. In the DYN0DE-B/REDY comparisons, what REDY input was unknown and how was Was this input data adjusted to improve the DYN0DE-B/REDY it determined?

comparisons?

" 5.

The reduction in voids in the top of the core is expected to affect the axial albedo for the upper reflector. Has this effect been accounted for 3 ,

and, if not, what is the effect of this simplification on the DYN0DE-B pre-dictions? l

6. Are any codes that have not been approved by the NRC being used to provide input to DYN00E-B7

[. The recirculation loop modeling for both REDY Hasand ODYN similar has been verified by qualification comparison to recirculation pump trip tests.

been performed for DYNODE-87 6.

What is the direction of conservatism for each input parameter, for which a conservative uncertainty allowance wiil be included, for the transients to be analyzed (Table 4.1-1,, MSIV closure, etc.)?

Are the values for the

9. Are the void model in DYN0DE-B and NDH identical? g void concentration parameter, Co, and drift velocity, If not,Vwhat j, used in the is the effect NDH calculations the same as used in 2DYN0DE-B7 calculated by DYN0DE-B and on t of this inconsistency on the k and M DYN0DE-B results? .

10.

t_ist all significant code and modeling differences between DYNODE-B, and REDY and ODYN and provide estimates of the effect of these differences on the DYN0DE-B predictions when it cannot be demonstrated that the differ- -

ences provide improved modeling or more conservative results.

e

11. Reference-7 recommends the mechanistic rather than the Profile-Fit void model for transient applications. Since DYN0DE-B allows both the mechan-istic and Profile-Fit void model, what is the basis for the selection of the Profile-Fit model?

12. The DYNOCE-B definition of the volumetric flow fraction, 6, the concentra-tion parameter, Co, and the drift velocity, Vg j, involve arbitrary con-stants (viz., C00, C01, by , 6 1, V )3, g Ygj2). How are these constants de-termined and what uncertainty is introduced into the DYH0DE-B calculations by the selection of these constants? Also, the definition of 6 in DYN0DE-B appears to be in error.

13. Describe in detail the core thermal-hydraulic model used to determine the axial pressure, vuid, flow and enthalpy distributions. Have the resulting f: equations been tested for numerical stability?

14 In the calculation of steam dome pressure, what Uncertainty is introduced by the use of the " steam-dome pressure model" Yher than the "non-equilibrium steam-dome pressure model?"

15. In the static flow distribution calculation, how is the bypass flow frac-tion determined and does it vary during the transient?

16. How are the feedwater flow, recirculation flow, power level, turbine bypass y

and stop valve controller lead-lag, lag and controller constants determined o

- and do they change for each cycle?

17. The flux, 4, rather than the source, S=v{fD, satisfies the standard time-dependent diffusion equation. Has the additional term 08g (v{f) been ac-

- counted for in the DYN0CE-B source equations and, if not, what errc,r is in-troduced by this approximation?

18. What is the mechanism responsible for the underprediction of the scram curves (Figure 3.1-4) and can this result in a non-conservative overpredic-tion for other static and transient states?

19. How do the DYN0DE-B and ODYN peak powers in the load rejection, feedwater controller failure and MSIV closure transients of Figures 3.2-93, 3.2-100 and 3.2-107, respectively, compare? Are these differences due to DYNODE-B and ODYN modeling differences and, if so, why should they not be considered as a measure of the uncertainty in performing transient analyses of Monti-cello reloads? ,

20. Describe how DYN0DE-B is used in the calculation of the fuel misloading er-ror and how the reactivity input is determined. How are radial redistribu-tion effects accounted for?

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21, in the application of DYN0DE-B to the control rud withdrawal event, what error is introduced by not including the radial flux distribution changes explicitly in the calculation? Does the non-equili,brium model include the time dependent mass and energy balance for the (1) riser and dome if steam (2) riser liquid (3) dome liquid and (4) the entrapped steam, not, what error is introduced by this approximation?

22. Explain any dif ferences between the Table 4.2-1 initial conditions and in-put parameters and the corresponding values and conditions assumed in the ODYN analyses. What effect do these differences have on the DYN0DE-B pre-dictions of ACPR, peak pressure and dect.y ratio for the transients to be analyzed?

23. How are the uncertainties in the bundle power and relative inlet flow due to differences in the static and transient radial power and flow distribu-tions accounted for in the determination of ACPR7 24 What range of operating state variables, including power level, flow, inlet 1 l

subcooling, control rod pattern and exposure, were used to detennine the collapsing factor ( AF)? Demonstrate that this set of states is sufficient in view of the wide range of intended applications (Table 4.1-l', MSIV closure, etc.). What is the Doppler reactivity co?la.nsing factor (AF) and ,

how is this uncertainty accounted for?

) Describe in detail the method used to determine the DYN0DE-B equivalent 25.

one-dimensional k . M2, gy and albedos from the three-dimensional NDH solutions. Describe the perturbed states used in this determination in terms of core power, flow, inlet subcooling, pressure and exposure. Demon-strate that these selected perturbed states provide an adequate representa-tion of the transient states encountered in the events to be analyzed (Ta-ble 4.1-1, MSIV closure, etc.) .How are the k , M 2, gy and albedos de-termined for the control rod insertion / withdrawal events?

26. The ODYN model has had difficulty in predicting core inlet flow oscilla-tions above 5 Hz. If DYN0DE-B will be required to analyze oscillations above this frequency, demoactrate that DYN0DE-B does not have the same dif-ficulty.

27. The qualification data base provided to demonstrate the accuracy of the DYN0DE-B code (e.g., Tables 4.4-1, 4.4-2 and 4.4-3) is insufficient in the number and quality of the comparisons to allow a reliable estimate of the code uncertainty. For example, the Peach Bottom turbine trip calculations were nonnalized to insure that DYN0DE-B reproduced the measured peak and integrated power, and the comparison for the Monticello turbine trip start-up test includes a large (-300%) DYNODE-B/ measurement transient power discrepancy. A detailed code uncertainty analysis is therefore required to insure there is suf ficient margin to the thermal-hydrulic design basis and -

the reactor coolant pressure boundary limit.

I

4 Provide a listing of the importar.t sources of uncertainty in the DYN00E-B predictions required for the intended reload analyses. Consideration should be given to factors such as: void coef ficient , . controller set-points, jet pump loss coefficients, scram reactivity, void model, separator

- model, steam line model, neutronics collapsing, etc. Estimate the 951 probability limits for these uncertainties, and determine the corresponding ACPR/ICPR for each uncertainty for the turbine trip without bypass tran-sient. Determine the corresponding a-pressure (%) for each of these uncer-tainties for the MSIV closure event with position switch scram failure.

Also, provide an estimate of the corresponding uncertainty in the calcu-lated ' decay ratio.

28. What mesh is used in the MDC representation of the steam line and does this satisfy the stability criteria?: The . steam line flow in Figure 3.2-96 does not' exhibit the same behavior as the ODYN prediction. What is causing this difference?

29. How does the DYN0DE-B decay heat precursors model compare with more recent revisions of this standard (e.g., the ANS standard of September,1978)?

A e

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+ l REQUEST FOR ADDITIONA'L INFORMATION QUESTIONS ON THE NSP REACTOR PHYSICS METHODS QUALIFICATION NSPNAD-8609 i

1. In order to eliminate ' selected TIP readings from the statistit,al analysis, it should be demonstrated that the eliminated TIP signals are erroneous and are not, in fact, a result of differences between the design model and the as-built core. What is the. increase in the reliability factors for MCPR, LHGR and APLHGR when no TIP signals are eliminated?

2. Describe in detail how the value %Ak/%AV = .0077 is determined from the data in Table 3.3.1.

3. Provide quantitative justification (using results from references 5-8 if appropriate) that the 95/95 upper tolerance limit on the Doppler coeffi-cient is RFoop = .10.

P 4 The NDH model has been nonnalized to the Monticello cycles 7 through 10 measurement data and, consequently, the reliability factors determined from the cycle 7-10 calculation / measurement differences are smaller than ,

those for a cycle to which NDH has not been normalized. What increase in- '

the reliability factors is expected for future cycles and how is this ac-  !

counted for?

5. In view of the differences between the PWR and BWR measurement systems and-the source of the measurement system errors, demonstrate that the factor of three reduction in the number of measurements is adequate to account

, for the lack of independence of the Monticello measurement errors.

6.~

In the calcul lon of both the void coefficient and control rod worth reliability factors, the error in the void and control rod reactivity defects, 6Ak, is assumed to b0 the same as the error in the statepoint keff, 6keff. In fact, in the determination of the reactivity defect, ak = k ff 2 - k' eff 1 the statepoint error, 6keff, to a good approximation"subfrac,tsout"andthereactivitydefecterror,6ak,is independent of the statepoint error, 6keff. Therefore, provide a calculation of the void coefficient and control rod worth reliability factors based on the error in predicting the void and control rod reactivity differences.

7. Based on the comparisons of Table 3.6.3 and Figure 3.6.44 it is concluded that all y-scan measurement data was not included in the power distribu-tion comparisons. On what basis was the measured data discarded and what effect does this data selection have on the reliability factors?

8. Are the generic normalization factors based on data from cycles 7 through-10? Are these factors intended for use in all future cycles of Monti-cello? What core parameters affect these normalization factors? How are these factors affected by operating history?

9. Describe the method used to generate the radial albedos and leakage fac-tors. How is void dependence in the radial albedos determined? Are these albedos updated for each cycle? How sensitive are the albedos to expo-sure, rod pattern, temperature and core loading?

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10. Describe the procedure used to derive the correction factor for a hundle moved from a peripheral to a central location. How sensitive is this cor-rection facter to exposure, rod pattern, temperature and core loading?

11. Describe the spectrum correction factor used to correct for the extra-polated flux.

12. How are incore detector signals calculated? Specifically, indicate how the contributions from each of four dissimilar uncontrolled / controlled assemblies are derived, indicating the, parameter dependence.

13. Do any of the few-rod criticals listed in Table 3.1.1 include the wiWi-drawal of the highest worth (strongest) rod at the time the critical was measured? If not, how would this withdrawal effect the results of the measurement / calculation comparisons?

14 How are uncertainties in the fuel pin temperature associated with power changes accounted for in the Doppler reliability factor? Similarly, how are uncertainties resulting from differences between the as-built and as-sumed dimensions and/or materials and fuel densification treated?

15. What effect do the differences (e.g., cross sections) between the initial version of CASMO used in the Kritz benchmarking and the more recent ver-sion used by NSP (CASMO-II) have on the reliability factors? l b 16. Are control rod history effects accounted for in the NDH calculations and, if not, how are the uncertainties introduced by this simplification ac-counted for?

17. Has the effect of excluding from consideration 8 of the 48 axial values of

- the instrument signals been evaluated? What is the increase in the uncer-tainty and, correspondingly, what is the additional allowance by which the power distribution reliability factor must be increased when this data is not excluded?

18. If NSP selects the option to provioe its own support for the process com-puter and generates its own data for this system, how will the change in uncertainty be accounted for in the safety limit?

19. Describe the fuel loadings for the cycles 7 through 10 cores which are in-cluded in the verification process of the NDH code. Provide infonnation on fuel types, U235 - enrichment, gadolinia, water rods, etc. Are the fuel loadings of cycles,7 through 10 representative of cycle 14 and future cycles?

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