RNP-RA/06-0027, ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis.

ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis.

ML061080524
Person / Time
Site:	Robinson
Issue date:	03/31/2006
From:	Progress Energy Carolinas
To:	Office of Nuclear Reactor Regulation
References
RNP-RA/06-0027 ANP-2512(NP)
Download: ML061080524 (29)

Text

United States Nuclear Regulatory Commission Attachment V to Serial: RNP-RA/06-0027 26 pages including cover page H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING REVISION TO CORE OPERATING LIMITS REPORT (COLR) REFERENCES ANP-2512(NP)

Loss of Forced Reactor Coolant Flow Analysis for Robinson March 2006 Non-Proprietary Version

ANP-2,r 12(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson March 2006

Customer Disclaimer Important Notice Regarding the Contents and Use of This Document Please Read Carefully AREVA NP, Inc.'s warranties and representations concerning the subject matter of this document are those set forth in the agreement between AREVA NP, Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither AREVA NP, Inc. nor any person acting on its behalf:

a. makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or

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The information contained herein is for the sole use of the Customer.

Inorder to avoid impairment of rights of AREVA NP, Inc. in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (inthe patent use of the term) of such information until so authorized in writing by AREVA NP, Inc. or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document.

sA AFt EVA NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loc s of Forced Reactor Coolant Flow Analysis for Robinson Page i Nature of Changes Itern Paqe Description and Justification

1. All Initial release.

ARE/A NP, Inc.

Aft EVA NON-PROPRIETARY ANP-251 2(NP)

Revision 0 Los s of Forced Reactor Coolant Flow Analysis for Robinson Page ii Table of Contents 1.0 Introduction ............... 1-1 2.0 Conclusion .2-1 3.0 Analytical Methodology .. 3-1 3.1 Nodalization .3-1 3.2 Chosen Parameters .3-2 3.3 Sensitivity Studies .3-2 3.4 Definition of Event Analyzed and Bounding Input .3-2 4.0 Loss of Forced Reactor Coolant Flow (FSAR 15.3.1) . .4-1 4.1 Identification of Causes and Event Description .4-1 4.2 Acceptance Criteria .4-1 4.3 Analysis Results .4-2 5.0 References .5-1 List of Tables Table 3.1 Key Assumptions .3-3 Table 3.2 Key Input Parameters Biases. 3-4 Table 4.1 Sequence of Events .4-3 List of Figures Figure 3.1 S-RELAP5 Reactor Vessel Nodalization .3-5 Figure 3.2 S-RELAP5 Reactor Coolant System Nodalization (Loop 1).3-6 Figure 3.3 S-RELAP5 Steam Generator Secondary System and Steam Line Nodalization (Loop 1).3-7 Fig Lre 4.1 Core Power .4-4 FigL're 4.2 Reactor Coolant Tempeatures .4-5 Figure 4.3 RCS Flow Rate .4-6 Figure 4.4 Pressurizer Pressure .4-7 FigLre 4.5 Pressurizer Level .4-8 FigL re 4.6 Reactivity .4-9 Figure 4.7 DNBR Trend .4-10 ARE'/A NP, Inc.

A AFt EVA NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page iii Nomenclature BOr Beginning-of-Cycle CE Combustion Engineering CHF Critical Heat Flux DNB(R) Departure from Nucleate Boiling (Ratio)

ESI: Engineered Safety Features HTP High Thermal Performance HF1' Hot Full Power LOCF Loss Of Forced Reactor Coolant Flow MDNBR Minimum Departure from Nucleate Boiling Ratio PORV Power Operated Relief Valve RCP Reactor Coolant Pump RC:S Reactor Coolant System RP. Reactor Protection System RNI' Robinson Nuclear Plant SAFODL Specified Acceptable Fuel Design Limit TS Technical Specification W Westinghouse AREVA NP, Inc.

A AR EVA NON-PROPRIETARY ANP-251 2(NP)

Revision 0 Los; of Forced Reactor Coolant Flow Analysis for Robinson Page 1-1 1.0 Introduction The analysis documented herein describes a LOCF analysis for RNP using the S-RELAP5 computer code. This analysis demonstrates the application of the Reference 1 methodology to the RNP.

AREVA NP, Inc

A ;' EVA% NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 2-1 2.0 Conclusion Based on the results of this analysis, margin exists to the DNB SAFDL. Because the core power does not increase appreciably during this event, the challenge to the fuel centerline melt SAI:DL is not limiting. The pressurization transient does not present a severe challenge to the maximum pressure criterion since system temperatures and pressure increase less significantly for a loss of flow event compared to complete loss of load type events. Therefore, the event acceptance criteria are met.

ARE/A NP, Inc.

A.

AR EVA NON-PROPRIETARY ANP-251 2(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 3-1 3.(1 Analytical Methodology The analysis is performed using the approved Reference 1 methodology. The S-RELAP5 code is used to model the primary and secondary side systems of the RNP and to calculate reactor power, total reactivity and fluid conditions (such as coolant flow rates, core inlet temperatures, pressurizer pressure and level). The MDNBR for the event is calculated using the thermal-hydraulic conditions from the S-RELAP5 calculation as input to the XCOBRA-IIIC code (Reference 2) along with the HTP CHF correlation (Reference 3). The RNP core is composed solely of AREVA NP HTP fuel assemblies, thus mixed core considerations do not apply.

3.1 Nodalization The plant configuration is represented by an S-RELAP5 model. The S-RELAP5 model nodalizes the primary and secondary sides into control volumes representing reasonable homogenous regions, interconnected by flow paths, or "junctions". The reactor vessel, RCS piping and steam generator nodalization diagrams are shown in Figures 3.1 to 3.3. The current analysis is based on a RNP specific model.

In general, the plant nodalization is defined to be consistent wherever possible for different plant types. Most of the differences existing between the model used for the current RNP analysis (W 3-loop plant) and that for the sample problem in Reference 1 (CE 2 x 4 plant) are attributed to plant specific differences. The RCS loop nodalization (including the steam generator primary sid e components) is different between W and CE plants to accommodate the different loop configurations. Namely, the loop configuration for the RNP application described herein consists of three individual loops each with one hot leg, a U-tube steam generator, a cold leg and a RCP. The Reference I sample problem, on the other hand, is based on a CE plant design which consists of two coolant loops each with one hot leg, a U-tube steam generator, two cold legs and two RCPs.

The vessel nodalization is very similar, differing only in the details of the downcomer and the lower downcomer and lower head flow paths to accommodate the flow skirt in the CE vessel (not present in the W design).

The steam generator secondary and steam line models are nodalized slightly differently between the current model for RNP and the Reference I sample problem model, namely, the ARE-VA NP, Inc

A AR EVA NON-PROPRIETARY ANP-251 2(NP)

Revision 0 Lo:3s of Forced Reactor Coolant Flow Analysis for Robinson Page 3-2 steam generator downcomer and boiler regions in the current model each contain one fewer noJes. Although the number of nodes decreased by one in each of these regions, the characteristics of the steam generator, specifically the volume distribution in the downcomer and the heat transfer to the boiler region, are more accurately captured. The overall effect of these changes on the analysis is negligible.

Other plant specific differences include the number and location of the main steam safety valves, the geometry of the pressurizer surgeline and the pressurizer PORV design.

3.2 Chosen Parameters The parameters and equipment states are chosen to provide a conservative estimate of the challenge to DNB. The biasing and assumptions for key input parameters are consistent with the approved Reference 1 methodology. The key assumptions are given in Table 3.1 and the biasing of key parameters is provided in Table 3.2. The process of defining the biasing and assumptiDns for key input parameters is consistent with the Reference 1 sample problem.

3.3 Sensitivity Studies This event is controlled primarily by the primary system flow coast down. The S-RELAP5 code assessments in Reference I validate the model relative to this controlling parameter. Thus, no additional model sensitivity studies are needed for this application.

The biasing of input parameters is chosen to produce a conservative estimate of the challenge to DNB for this application. Thus, no additional input parameter sensitivity studies are needed.

3.4 Definition of Event Analyzed and Bounding Input The event is analyzed from full power initial conditions since the margin to the DNB limit is minimized at the beginning of the event. BOC kinetics parameters are used resulting in positive moderator reactivity feedback and an increase in core power. The input parameter biasing and assumptions for this event, shown in Tables 3.1 and 3.2, are consistent with the approved meihodology.

AREVA NP, Inc

AFIEVAY NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Fcrced Reactor Coolant Flow Analysis for Robinson Page 3-3 Table 3.1 Key Assumptions Parameter Assumption Time of loss-of-offsite power Offsite power is available Mitigating systems

• Low primary flow RPS trip Available

• Pressurizer PORVs Available

• Pressurizer spray Available Operator actions No operator actions are credited Single failure No single failure in the ESF affects this event.

All loops in operation consistent with HFP operation.

Number of operating loops ARE/A NP, Inc.

A AS'EVA. NON-PROPRIETARY ANP-2512(NP)

Revision 0 Los; of Forced Reactor Coolant Flow Analysis for Robinson Page 3-4 Table 3.2 Key Input Parameters Biases Parameter Bias Initial reactor core power (MMt) Rated thermal power plus calorimetric uncertainty Maximum TS value [ I plus Initial RCS vessel average temperature (F) measurement and control deadband uncertainties [ I Nominal value [ 3minus Initial RCS pressure (psia) measurement and control deadband uncertainties [

Initial RCS flow rate (Mlbm/hr)

Scram reactivity (pcm)

Moderator temperature coefficient (pcm/°F) Most positive TS value Doppler reactivity coefficient (pcm/IF)

Peliet-to-clad gap conductance and fuel rod thermal properties (Btu/hr-ft2 -OF),

Low RCS flow RPS trip setpoint (% of initial Nominal minus uncertainty flow)Noiamiuunetny Low RCS flow RPS trip time delay (sec) Maximum AREVA NP, Inc.

AFl EVAS NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 3-5 I

Figure 3.1 S-RELAP5 Reactor Vessel Nodalization ARENA NP, Inc.

AS' EVAt. NON-PROPRIETARY

. ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 3-6

[

I Figure 3.2 S-RELAP5 Reactor Coolant System Nodalization (Loop 1)

ARENA NP, Inc.

A A R EVA NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Paae 3-7

[

I Figure 3.3 S-RELAP5 Steam Generator Secondary System and Steam Line Nodalization (Loop 1)

ARE.VA NP, Inc

A AR EVA NON-PROPRIETARY ANP-251 2(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 4-2 No single failure of the ESF will affect the analysis.

4.3 Analysis Results The transient response is shown in Figures 4.1 to 4.7. An event summary is given in Table 4.1.

The transient is initiated by tripping all three primary coolant pumps. As the pumps coast down, the flow reduces causing a reactor trip on a low primary coolant flow signal with rod insertion beginning at 3.0 seconds.

As the flow coasts down, primary temperatures increase. The vessel average coolant temperature increases about 30 F before being turned around by the power decrease following reactor scram. This increase in temperature causes a subsequent core power rise resulting from moderator reactivity feedback due to the positive coefficient being modeled. The core povier peaks at about 105% of rated power.

The temperature increase also causes an insurge into the pressurizer and resultant pressurization of the RCS. The pressurizer PORVs open to minimize pressure which in turn maximizes the DNB challenge. The compensated pressurizer PORV opens at 3.0 seconds and preients the pressure from exceeding 2315 psia at 5.0 seconds. The pressure decreases as the core power level continues to drop.

The calculated MDNBR is 1.192 relative to the 95/95 limit for the HTP CHF correlation of 1.141.

AREVA NP, Inc.

A AF. EVAt NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson -Page 4-3 Table 4.1 Sequence of Events Event Time (sec) Value Three-pump coastdown initiates (see Figure 4.3) 0 2.0 87%

RCSofflow initial RCS flow reaches low flow trip RPS setpoint Scram occurs (see Figure 4.6) 3.0 Core power peaks (see Figure 4.1) 3.0 105% of

_____________________ ___ ____ ___ ___r ate d pow er Compensated pressurizer PORV begins to open 3.0 MDNBR occurs 4.1 1.192 Uncompensated pressurizer PORV begins to open 5.0 Pressurizer pressure peaks (see Figure 4.4) 5.0 2315 psla Uncompensated pressurizer PORV fully closed 6.3 Average RCS coolant temperature peaks (see Figure 4.2) 8.5 579 *F Compensated pressurizer PORV fully closed 10.1 AREVA NP, Inc.

AVA:e A ' NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 4-4 120 100 80 I-60 0

0 c1 40 U

20 0

0 20 Time (s)

Figure 4.1 Core Power AREVA NP, Inc

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Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 4-5 620 600 RCS Avg. Thot

.RCS Avg. Tcold a 580 _ _ _ __-- - Core Average 0

C-E 560 I-540 P 520 . . . . . . . I . . . . ..

0 5o 10 15 20 Time (s)

Figure 4.2 Reactor Coolant Temperatures AREVA NP, Inc.

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Revision 0 Lois of Forced Reactor Coolant Flow Analysis for Robinson Page 4-6 40000 35000 [

30000 0

25000 co 20000 0

U-a)at 15000 10000 5000 0-0 5 10 15 20 Time (s)

Figure 4.3 RCS Flow Rate ARE/A NP, Inc,

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Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 4-7 2400 2300 2200

.0M A-2 2100 0

2000 1900 1800 0 5 10 15 20 Time (s)

Figure 4.4 Pressurizer Pressure AREMA NP, Inc.

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Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 4-8 100 80

-c 60 CL

-1

-6 9 40 20 0

0 5 10 15 20 Time (s)

Figure 4.5 Pressurizer Level AREVA NP, Inc.

A AFZ EVA NON-PROPRIETARY ANP-251 2(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Paqe 4-9 2.0 1.0 0.0

-1.0 Z;t

-2.0 .

.5.:

4)

-3.0

-4.0

-5.0

-6.0 .

-7.0 .

o 5 10 15 20 Time (s)

Figure 4.6 Reactivity AREVA NP, Inc.

AS'1EVA NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 4-1 0 5.0 4.0 z

a:

E 3.0 C

al N

,u 2.0 0

z 1.0 0.0 Time (s)

I Figure 4.7 DNBR Trend' a "he Tong DNBR Is a RELAP calculated core-wide DNBR used to estimate the approximate time of rninimum DNBR to establish core parameters for input into the XCOBRA-IIIC code to evaluate 1ADNBR. The illustrated Tong DNBRs are normalized to the minimum value.

ARENA NP, Inc.

A AIq EVA NON-PROPRIETARY ANP-2512(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Robinson Page 5-1 5.(I References

1. EMF-2310(P)(A) Revision 1, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, Framatome ANP, May 2004.

2. XN-NF-82-21 (P)(A) Revision 1, Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, Exxon Nuclear Company, September 1983.

3. EMF-92-153(P)(A) Revision 1, HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, Siemens Power Corporation, January 2005.

AREVA NP, Inc.

United States Nuclear Regulatory Commission Attachment VII to Serial: RNP-RA/06-0027 4 pages including cover page H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING REVISION TO CORE OPERATING LIMITS REPORT (COLR) REFERENCES Framatome ANP Affidavit Regarding Proprietary Attachment VI

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

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CITY C)F LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing in Regulatory Affairs, for Framatome ANP ("FANP"), and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by FANP to determine whether certain FANP information is proprietary. I am familiarwith the policies established by FANP to ensure the proper application of these criteria.

3. 1am familiar with report ANP-2512(P), Revision 0, entitled "Loss of Forced Reactor Coolant Flow Analysis for Robinson," dated March 2006 and referred to herein as "Document." Information contained in this Document has been classified by FANP as proprietary in accordance with the policies established by FANP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by FANP and not made available to the public.

Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

6. The following criteria are customarily applied by FANP to determine whether information should be classified as proprietary:

(a) The information reveals details of FANP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

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(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for FANP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by FANP, would be helpful to competitors to FANP, and would likely cause substantial harm to the competitive position of FANP.

7. In accordance with FANP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside FANP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

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9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this 144L day of . M L , 2006.

t !I Ella F. Carr-Payne NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/09 E:LLA F.CARR-PAYNE - I NOTARY PUBLIC Commonwealth of Virginia My (commission Expires 8-31-09