ML20237B229

Provides Response to RAI in Support of 980427 & s Re Pressure/Temp Limits

ML20237B229
Person / Time
Site:	Catawba
Issue date:	08/10/1998
From:	Gordon Peterson DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M99629, TAC-M99630, NUDOCS 9808180125
Download: ML20237B229 (11)

Text

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Duks Energy Corpor; tion kf Catawba Nuclear Station 4800 Concord Road York, SC 29745 Gary W Peter (803) 83Id25I *"

n Vice Prnident (803) 831-3426nx

' August 10, 1999 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, D.C.

20555

Subject:

Duke. Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Reply to NRC Request for Additional Information on Amendment Request Regarding Pressure / Temperature Limits TAC Numbers M99629 and M99630

References:

l '.

Letter from Gary R.

Peterson, Duke, to NRC, same subject, dated July 22, 1998.

2.

Letter from Gary R.

Peterson, Duke, to NRC, same subject, dated April 27, 1998.

3.

Letter from Peter S.

Tam, NRC, to Gary R.

Peterson, Duke, same subject, dated March 26, 1998.

In References 1 and 2, Duke Energy Corporation responded to Questions 3 and 4 of the request for additional information contained in Reference 3.

Pursuant to 10 CFR 50.4 and 10 CFR 50.90, Duke Energy Corporation is hereby providing additional information in support of the Reference 1 and 2 responses.

This additional information concerns computer analysis k

performed for the Catawba units with respect to the low

/

temperature overpressure protection (LTOP) heat input

/

transient.

The additional information is contained in to this letter

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i U.S. Nuclear Regulatory Commission J

Pqge 2, August *10, 1998 1

If you have any questions concerning this information, I

please call L.J.

Rudy at (803) 831-3084.

f Very ruly rs,

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-e G ry Peterson LJR/s Attachment xc (with attachment):

L.A.

Reyes U.S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St.,

SW, Suite 23T85 Atlanta, GA 30303 D.J.

Roberts Senior Resident Inspector (CNS)

U.S. Nuclear Regulatory Commission Catawba Nuclear Station P.S. Tam NRC Senior Project Manager (CNS)

U.S. Nuclear Regulatory Commission Mail Stop O-14H25 Washington, D.C.

20555-0001 M. Batavia, Chief Bureau of Radiological Health 2600' Bull St.

Columbia, SC 29207

1 Discus'sion Concerning RETRAN Safety Evaluation Report (SER)

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R'strictions for Catawba Low Temperature Overpressure e

Protection (LTOP) Heat Input Transient

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The following additional information is being provided I

l to further address the SER restrictions associated with l

applying the Catawba RETRAN model to the LTOP heat input l

l transient (Reference 1).

This information is relevant to the proposed Technical Specification changes to the Catawba Nuclear Station pressure-temperature limits.

As described in Reference 1, the RETRAN-02 MOD 005.1 code has been utilized in a Duke Power analysis of the LTOP heat input transient for Catawba Unit 1

with the replacement steam generators.

The acceptability of referencing the MODS.1 version of the RETRAN-02 code is given in Reference 2.

A more detailed discussion of the RETRAN SER restrictions is provided below.

The SER restrictions considered are both those associated with the RETRAN-02 code itself (References 3, 4 and 5,),

and those associated with the Duke Power RETRAN-related topical reports (References 6,

7 and 8).

For the restrictions associated with the Duke Power topical reports, only those relevant to the application of the RETRAN code to the Catawba Nuclear Station are discussed.

For example, since the VIPRE code is not employed in the LTOP heat input transient analysis, restrictions associated with applying the VIPRE code are not applicable.

Additionally, since the LTOP heat input transient is -being applied to the Catawba Nuclear Station (recirculating steam generators),

restrictions associated with applying the RETRAN code to the Oconee Nuclear Station (once-thru steam generators) are also not applicable, and are therefore not discussed.

General RETRAN-02 Restrictions (taken from Reference 3)

1. RETRAN is a large,

complex, sophisticated analysis tool.

The number of volumes and nodes is variable and a

variety of options is available to model different components in a

nuclear power plant.

Because the user has so much freedom, those who wish to submit analyses for review using RETRAN should also submit the corresponding plant or system model (input deck) for review.

Duke Power has met this restriction by submitting j

topical report DPC-NE-3000 (Reference 6).

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2. The subcooled void model is not well qualified.

This i

model must be fully qualified for each condition l

where it is applied, and its use should be restricted to the range of qualification.

l This restriction is not applicable since the subcooled void model is not applied in the LTOP heat input transient analysis.

3. In principal the dynamic s.ip model would be expected to yield more accurate results than the homogeneous equilibrium model when calculating two phase flow.

In the qualification analysis submitted for review l

this was not demonstrated.

In submittals in which the dynamic slip model is

employed, it must bre qualified and sensitivity studies parformed to demonstrate stability and convergence.

It chould be made clear as part of the qualification that the dynamic slip model performs better than the homogeneous equilibrium model and that the application of the model is within the limits of qualification.

This restriction is not applicable since the dynamic

' slip model is not applied in the LTOP heat input transient analysis.

4. The nonequilibrium pressurizer model has not been qualified for those situations where the pressurizer becomes completely empty or full (goes from two phase to single phase).

Appropriate qualification should be submitted with any analysis that places the pressurizer in either of these modes.

This restriction has already been addressed in Reference 1.

5. The bubble rise model has received only limited verification, and its use should be carefully qualified.

The LTOP heat input transient analysis uses no application of the bubble rise model which has not already been justified in References 6 and b.

It should be noted that no boiling heat transfer is modelea for the LTOP heat input transient, and therefore the bubble rise model for phase separation is not significant.

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6. The use of the transport delay option should be aimited to those cases showing a

dominant flow g

direction.

1 This restriction has already been addressed in Reference 1.

1 RETRAN-02 MOD 3 and MOD 4 Restrictions (taken from Reference 4)

1. The RETRAN code is a generically flexible computer code requiring the users to develop their own nodalization and select from optional models in order to represent the plant and transients being examined.

Thus, RETRAN users should include a discussion in their submittals as to why the specific nodalization scheme and optional models chosen are adequate.

This has been ddressed in item 1 in the previous section.

2. Restrictions imposed on the use of RETRAN02 models in the original SER have not been addressed and therefore remain in force for both MOD 003 and MOD 004.

These restrictions have been addressed in the previous section.

3. The countercurrent flow logic was

modified, but continues to use the constitutive equations for bubbly flow; i.e.,

the code does not contain constitutive models for stratified flow.

Therefore, use of the hydrodynamic models for any transient which involves a

flow regime which would not be reasonably expected to be in bubbly flow will require additional justification.

This restriction is not applicable since countercurrent flow logic is not applied to the Lf0P heat input transient.

4. Certain changes were made in the momentum mixing for use in the jet pump model.

These changes are acceptable.

However, those limitations on the use of p

the jet pump momentum mixing model which are stated in the original SER remain in force.

l No jet pumps are modeled for the LTOP heat input transient (PWR).

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5. If licensees choose to use MOD 004 for transient analysis, the conservatism of the heat transfer model i

for metal walls in non-equilibrium volumes should be demonstrated in their plant specific submittals.

Typically, non-equilibrium volume modeling is only applied to the pressurizer.

The LTOP heat input transient models the pressurizer as a single phase,

. water solid volume.

Non-equilibrium volume restrictions are therefore not applicable.

However, i

for other applications, the non-equilibrium pressurizer modeling is justified in Reference 6.

6. The default Courant time step control for the l

implicit numerical solution scheme was modified to 0.3.

No guidance is given to the user in use of default value or any other values.

In the plant l

specific submittals, the licensees should justify the adequacy of the selected value for the Courant parameter.

l The justification for the iterative numerics timestep control is provided in Reference 6.

However, the LTOP heat input transient uses a maximum timestep size that is well below any Courant limit.

RETRAN-02 MOD 5.0 Restrictions (taken from Reference 5)

1. The user must justify, for each transient in which the general transport model is used, the selected degree of mixing with considerations of its coupling to the calculation of other parameters affecting the evaluation of events (e.g.,

reactivity feedback for i

boron transport).

The general transport model is not utilized for the LTOP heat input transient analysis.

2. The user must justify, for each use of the ANS 1979 standard decay heat model, the associated parameter inputs.

Decay heat is not modeled for the LTOP heat input transient analysis.

Decay heat addition is assumed l

to be matched by.the heat removal capability of the l

Residual Heat Removal (RHR)

System.

The transient I

heat input to the primary system is a result of reverse heat transfer from the initially hotter steam generator secondary inventory.

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3. Because of the inexactness of the new reactivity edit feature, use of values in the edit either directly or as constituent factors in calculations of parameters for comparisons to formal performance criteria must be justified.

Reactivity is not edited for the LTOP heat input transient.

Core power is assumed to be zero, since the core decay heat addition is. assumed matched by the heat removal capability of the RHR system.

DPC-NE-3000 RETRAN Restrictions (taken from Reference 9)

1. With respect to analyzing transients which result in a

reduction in steam generator secondary water inventory, use of the RETRAN-02 steam generator modeling is acceptable, only for transients in that category for which the secondary side inventory for the effective steam generator (s) relied upon for heat removal never decreases below an amount which would cover enough tube height to remove decay heat.

The initial steam generator inventory is maximized for the LTOP heat input transient.

This is l

conservative for maximizing the reverse heat l

transfer.

The steam generator tube bundles are never uncovered for the LTOP analysis.

The above l

restriction is therefore satisfied for the LTOP heat input transient analysis.

2. All generic limitations specified in the RETRAN-02 l

SER are applicable.

l These are addressed in the above three sections.

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DPC-NE-3001 RETRAN Restrictions (taken from Reference i

10)

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1.The licensing application of the DPC-NE-3001-P transient analysis methods requires NRC approval of MOD 005 of RETRAN-02 for boron transport calculations.

Reference 5 fulfills the requirement of NRC approval of RETRAN-02 MOD 005.

However, the general transport option is not utilized for the LTOP heat input transient analysis.

2. The licensing application of the DPC-NE-3001-P transient analysis methods requires NRC approval of the thermal-hydraulics topical report DPC-NE-3000.

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Reference 9 fulfills this requirement.

DPC-NE-3 002 RETRAli Concerns /TER Issues (taken from Reference 11)

1. DPC 's Statistical Core Design methodology treats seven variables as key parameters.

Four of these variables were accounted for in DPC-NE-3002.

Of the remaining parameters, the power factors are also input items for systems analysis, which were not presented in the topical report.

Similarly, reactivity feedback was not discussed in this report.

Both of these parameters can significantly influence the course of the transient.

Therefore, when application of the philosophical approach reported in this report is made and submitted for NRC review and approval, review should be made of the modeling of power and reactivity feedback, and to assure that such modeling has no adverse impact on the other modeling described herein.

No power L; reactivity feedback is modeled for the LTOP heat input transient analysis.

As discussed previously, core heat input is zero.

2. Validity of DPC's assumption of 120%

of design pressure as part of the acceptance criteria for Reactor Coolant Pump Locked Rotor should be determined by the NRC staff.

As later determined by the

staff, the acceptance criterion for Reactor Coolant Pump Locked Rotor shall be 110% of design pressure.

This restriction is applicable only to the Reactor Coolant Pump Locked Rotor event, and therefore does not apply to the LTOP heat input transient analysis.

3. DPC documented intent to perform parametric studies in order to select conservative scenarios or assumptions throughout the subject topical report.

Therefore, such parameter studies must be presented when this methodo3ogy is applied.

Sensitivity studies have been performed for the major parameters affecting the results of the LTOP heat input transient.

These include RCS temperature, pressurizer temperature, steam generator

mass, reactor coolant pump start times, and PORV closing l

l-6 1

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times.

The results from the limiting cases are summarized in Reference 1.

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4. No justification was presented for trip and actuation times assumed in the Feedwater System Pipe Break event analysis.

Such justifications must be presented when this methodology is applied.

This restriction applies to the Feedwater System Pipe Break event analysis, and is therefore not applicable to the LTOP heat input transient analysis.

DPC-NE-3002 RETRAN Restrictions (taken from Reference 12)

1. The acceptability of the use of DPC's approach to FSAR analysis is subject to the conditions of SEs on all aspects of transient analysis and methodologies (DPC-NE-3000, DPC-NE-3001, DPC-NE-3002, DPC-NE-2004, and DPC-NE-2005) as well as SEs on the RETRAN and VIPRE'01 computer codes.

The RETRAN-related restrictions have been addressed for DPC-NE-3000, DPC-NE-3001, DPC-NE-3002 and the RETRAN code SEs in the above sections.

The VIPRE l

code restrictions and the Duke Power VIPRE-related topical reports are not applicable since the LTOP heat input transient analysis only employs the RETRAN l

code and methodology.

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2. There are scenarios in which a SGTR event may result in loss of subcooling and the consequent two-phase flow conditions in the primary system.

In such instances, the use of RETRAN is not acceptable without a detailed review of the analysis.

This restriction applies to the SGTR analysis, and is therefore not applicable to the LTOP heat input transient analysis.

It should also be noted that subcooled margin is never lost for the LTOP heat input transient.

3. In the future, if hardware or methodology changes, selection of limiting transients needs to be reconsidered, and DPC is required to perform sensitivity studies to identify the initial conditions in such a way to avoid conflict between transient objectives, such as DNB and worst-case primary pressure.

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As detailed

above, sensitivity studies have been

. performed for the LTOP heat input transient analysis to determine the limiting assumptions.

4. It is emphasized that, when using the SCD methodology to determine DNBR, the range of applicability of the l

selected critical heat flux correlation must not be i

violated.

l DNB is not an acceptance criterion for the LTOP heat input transient, and therefore this restriction does.

not apply.

l S.DPC's assumption of 120% of design pressure as part of the acceptance criteria for Reactor Coolant Pump Locked Rotor is not acceptable; DPC is required to use 110% of design pressure for that limit.

This restriction is applicable only to the Reactor Coolant Pump Locked Rotor event, and therefore does not apply to the LTOP heat input transient analysis.

The RETRAN SER restrictions have therefore been reviewed for applicability to the LTOP heat input transient analysis.

The restrictions listed in the preceding

-sections do' not restrict the application of the RETRAN code to perform the LTOP heat input transient.

References 1.' Letter from G.

Peterson (DPC) to USNRC, " Reply to NRC Request for-Additional Information on Amendment Request Regarding Pressure-Temperature Limits," July 22, 1998.

2.

Letter from Martin J. Virgilio (USNRC) to C. R. Lehmann (PP&L), " Acceptance for Referencing of the RETRAN-02 MOD 005.1 Code," April 12, 1994.

3.

Letter from Cecil O. Thomas (USNRC) to Dr. Thomas W.

Schnatz (UGRA), " Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, 'RETRAN - A Program for One Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,' and EPRI NP-1850-CCM,

'RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,'" September 2, l

1984.

4. Letter from Ashok Thadani (USNRC) to R.

Furia (GPU),

}

" Acceptance for Referencing Topical Report EPRI-NP-1850 1

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1 i

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CCM-A, Revisions 2 and 3 Regarding RETRAN02/ MOD 003 and g

MOD 004," October 19, 1988.

5.

Letter from Ashok Thadani (USNRC) to W.

James Boatwright (TUEC), " Acceptance for Use of RETRAN02 MOD 005.0,"

November 1, 1991.

6. DPC-NE-3000-PA, " Thermal-Hydraulic Transient Analysis Methodology," Duke Power Company, Revision 1, December 1997.

7.

DPC-NE-3001-PA, " Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology,"

Revision 0, December 1997.

8.

DPC-NE-3002-A, "UFSAR Chapter 15 System Transient Analysis Methodology," Revision 2, December 1997.

9.

Letter from Timothy A. Reed (USNRC) to H.

B.

Tucker (DPC), " Safety Evaluation On Topical Report DPC-NE-3000, Thermal-Hydraulic Transient Analysis Methodology (TAC Nos. 73765/73766/73767/73768)," November 15, 1991.

10. Letter from Timothy A. Reed (USNRC) to H.

B.

Tucker (DPC), " Safety Evaluation on Topical Report DPC-NE-3001, Multidimensional Reactor Transients and Safety Analysis Physics Parameters (TAC Nos. 75954/75955/75956/75957),"

November 15, 1991.

11. Letter from Timothy A. Reed (USNRC) to H.

B.

Tucker (DPC), " Safety Evaluation on Topical Report DPC-NE-3002, FSAR Chapter 15 System Transient Analysis Methodology (TAC No. 66850), November 15, 1991.

12. Letter from Robert E. Martin (USNRC) to M.

S.

Tuckman (DPC), " Safety Evaluation for Revision 1 to Topical Report DPC-NE-3002, FSAR Chapter 15 System Transient Analysis Methodology, McGuire Nuclear Station, Units 1 and 2, and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M89944, M89945, and M89946)," December 28, 1995.

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