ML20137C502

Forwards Responses to NRC Questions Re 850419 Submittal of Rev 3 to Topical Rept NSPNAD-8102

ML20137C502
Person / Time
Site:	Prairie Island
Issue date:	11/19/1985
From:	Musolf D NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
References
TAC-57644, TAC-57645, NUDOCS 8511260518
Download: ML20137C502 (6)

Text

o ~~o Northern States Power Company 414 Nicollet Mall Minneapons. Minnesota 55401 Telephone (612) 330 5500 November 19, 1985

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Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PL\NT Docket Nos. 50 282 License Nos. DPR 42 50-306 DPR 60 Responses of NRC Questions on Revision 3 of Topical Report NSPNAD 8102 Attached are responses to Staff questions concerning our April 19, 1985 submittal of Revision 3 to Topical Report NSPNAD 8102.

Please contact us if you further questions.

0' W avid Musolf Manager Nuclear Sup ,et Services DKM/TMP c: Regional Administrator III, NRC NRR Project Manager. NRC Resident Inspector, NRC C Charnoff Attachment 0\

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Attachment Paga 1 of 5 l PRAIRIE ISLAND NUCLEAR GENERATING PLANT l

l Responses of NRC Questions on Revision 3 of Topical Report NSPNAD-8102 Question 1 The one pass core modeling will be used in the Prairie Island licensing hot channel thermal hydraulic calculations: (a) Provide details of the radial subchannel layouts which model the reactor core and hot channel for both a homogenous core of the same fuel design and a transitional mixed core of more than one fuel design. (b) How do you ensure that these layouts are detailed enough to provide correct crossflow and

. turbulent mixing of the hot channel with the remainder of the core that results in the most conservative DNBR?

Response

A typical radial subchannel layout is shown in Figures C-1 and C-2 of Reference 1.

The guidelines used to generate the radial subchannel layout are described on pages 273-278 of Reference 1. Briefly, the guidelines involve modelling the hot quarter assembly on a subchannel basis with the surrounding quarter assemblies modeled as lumped channels. The remaining full assemblies are modeled as lumped channels. The hot fuel rod is arbitrarily raised to the Tech Spec FAH value.

A conservative axial power shape is applied to all rods to raise the hot fuel node to the Tech Spec FQ limit. The same guidelines are used to generate subchannel layouts for mixed and homogeneous core designs. For mixed core designs, two separate

'VIPRE models are used to evaluate the response of the hot Exxon and hot Westinghouse quarter assemblies. These guidelines were determined to be sufficient to correctly model crossflow and turbulent mixing the COB % IIIC/MIT code (Reference 2).

Therefore, the same level of radial detail previously used in COBRA IIIC/MIT, is used for the VIPRE application.

Question 2 The hot channel is defined as the one which has the lowest DNBR. The channel surrounded by the fuel pins having the highest peaking factors is not necessarily always the hot channel. How are the hot channels determined for the one pass modeling of the mixed core and homogeneous core fuel designs, respectively?

Response

The hot channel is assumed to occur within the quarter assembly containing the fuel rod with the highest peaking factor. This quarter assembly is modeled on a subchannel basis (Reference 1, Figure C-1). The relative pin powers from the quarter assembly with the lowest peak pin to average pin factor are superimposed on this location. These relative pin powers are then raised up such that the hot pin relative radial power is equal to the Tech Spec FAH limit. The DNBR is calculated for all subchannels within this quarter assembly. The hot channel is therefore calculated to be the one with the MDNBR regardless of the location of the peak rod. For mixed core applications, two separate models are used, one to calculate the response of the hot Exxon TOPROD channel and one to calculate the response of the hot Westinghouse channel.

Attachment Page 2 of 5 Question 3 Table F.5 of NSPNAD-8102, Rev.3, provides the values of the turbulent momentum factor, transverse momentum factor and crossflow resistance factor for the Westinghouse Improved Optimized fuel. What are the bases for these values? What are the values of the same parameters for the Exxon standard and Exxon TOPROD fuel designs? What are the values used in a mixed core of various fuel designs.

Response

The values for the turbulent momentum factor, transverse momentum factor, and crossflow resistance factor are the same for Westinghouse Improved Optionized Fuel Assemblies (IOFA) and Exxon TOPR00 and standard designs.

The turbulent momentum factor is the recommended value from Reference 3. The values for the transverse momentum factor and the crossflow resistance factor were determined from sensitivity studies performed with the COBRA IIIC/MIT code (Reference 1 pages 275-276). Since the two codes use these values similarly, the sensitivity study results can be applied to the VIPRE code.

Question 4 What are the heat transfer correlations used in the Prairie Island application with regard to single phase forced convection and subcooled and saturated nucleate boiling?

Provide your justification for using these correlations.

Response

Single phase forced convectien - Dittus-Boelter.

Subcooled boiling - Jens-Lottes.

Saturated nucleate boiling - Thom.

These correlations were selected to maintain consistencey with the previously approved (Reference 2) COBRA IIIC/MIT methods. This combination of correlations is also justifiable, for our application,' based on the benchmarks NSPNAD performed to the Westinghouse test data (Reference 1, Table F.2).

Question 5 In Prairie Island licensing application, is the fuel pellet-cladding gap conductance calculated by the VIPRE gap conductance model or an user input constant value? If the gap conductance is an input constant, what is the value (and justification) used?

Response

In the Prairie Island licensing application the fuel rod conductance is not modeled in VIPRE but rather in DYN00E-P, the systems code (Reference 1, Appendix B). In DYN00E-P, a constant gap conductance, of approxiu tely 550 Btu /hr ft 8 *F, is conservatively used. The heat flux from DYN00E-P fs then forced on the VIPRE model.

This application of the fuel rod conductance model has been previously approved (Reference 2) by the staff.

Attachment i Page 3 of 5 l Question 6

( NSPNAD-8102, Rev.3 refers to the new fuel to be loaded in Prairie Island cores as the 14x14 Westinghouse Improved Optimized fuel. What are the differences between the l- Westinghouse Improved 0FA and Westinghouse OFA fuel designs?

Response

The Westinghouse Improved Optimized Fuel assembly has a six inch natural uranium l axial blanket at the top and bottom of the fuel rods.

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l Question 7 Appendix F.4 of NSPNAD-8102, Rev.3 indicates that the MONBR will now be evaluated

using the WRB-1 correlation instead of W-3. However, WRB-1 is not qualified for use l with the Exxon standard and TOPROD fuel designs. Is W-3 still used for the Exxon l fuel?

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Response

The WRB-1 correlation will only be used for the Westinghouse 10rA. The W-3 l correlation will continue to be used for the Exxon Standard and TOPROD fuel.

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Question 8 ,,

Describe the quality assurance program to be used in the application of VIPRE-01 in safety analysis. This is, explain what steps are taken to. maintain consistency in safety analysis applications.

l (a) What steps will be taken by Northern States Power to assure that only the l

approved versions of VIPRE will be used?

l (b) How will future modifications to .VIPRE be implemented for Prairie Island?

Response

The VIPRE code will be controlled according to NSPNAD Policies and Procedure NAP 5.001A Rev.4, " Computer Program Control" (attached). This is the same procedure that is used to control all computer program within the Northern States Power Nuclear Analysis Department (NSPNAD). " bis procedure covers the use of codes and making modifications to codes. This procedure was audited by Mr R H Brichley of the

! NRC in January 1983 (References 4 and 5). A brief description of how VIPRE is

! controlled using this procedure follows.

The VIPRE-01 code package includes a series of standard test cases developed by the authors of the code, Battelle Northwest Laboratories. After we install the code on our computer system, we run these standard test cases and compare the output of our cases to the output from the cases Battelle ran to ensure that we can reproduce their results. In order to test any coding we have added to VIPRE-01, we also develop our own test cases and run these at this time. As an example, VIPRE-01 has a subroutine specifically designed for utilities to add a CHF correlation, which NSP used to add l the WRB-1 correlation.

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Attachment Page 4 of 5 Once all the test cases have been run and the results and comparisons have been documented, the code is approved for use by the Manager, Nuclear Analysis. At this point, the code version is archived and passworded and no changes to the source are allowed. If the code has not been reviewed and approved for use by the NRC, it is labeled as a " Developmental" code and is archived to the developmental storage area.

Currently, VIPRE is labeled a " Developmental Code". Developmental codes may be used for safety related calculations. If the code has been reviewed and approved for use by the NRC, it is labeled as a " Production" code and is archived to the production storage area. Only production codes may be used for safety related calculations.

Future modifications are handled in a similar manner. All previous test cases are rerun to ensure that the original coding is still correct. New test cases are developed in addition to the original ones to test all new coding added to VIPRE-01.

Once tested and approved, the new version is archived and replaces the previous version, so that there is only one approved version available for "ee at any one time.

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Attachm:nt Page 5 of 5 References

1. NSPNAD-8102P Rev.3, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to P.I. Units," March 1985.

2. " Safety Evaluation by the Office of Nuclear Reactor Regulation of the Reactor Physics and Reload Safety Evaluation Methods Technical Reports NSPNAD-8101P and NSPNAD-8102P for the Northern States Power Company Prairie Island Nuclear Generating Plant, Units 1 and 2," February 17, 1983.

3. EPRI NP-2511-LLM, VI-V3.

4. Letter; J. F. Streeter, NRC to C. E. Larson, NSP, "Special Safety Inspection,"

March 9, 1983.

5. Letter; C. E. Larson, NSP to J. G. Keppler, NRC, " Response to Special Safety Inspection Report Number 50-282/83-01 (DPRP); 50-306/83-01 (DPRP),"

April 4, 1983.

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