ML103620211

Draft Response to Us EPR Design Certification Application RAI No. 413, FSAR Ch. 7, Supplement 3

ML103620211
Person / Time
Site:	Framatome
Issue date:	12/22/2010
From:	Canova M Office of New Reactors
To:	AREVA NP
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Download: ML103620211 (62)

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Canova, Michael From: Canova, Michael Sent: Wednesday, December 22, 2010 9:58 AM To: ArevaEPRDCPEm Resource

Subject:

FW: Non-PROPRIETARY DRAFT Response to U.S. EPR Design Certification Application RAI No. 413, FSAR Ch. 7, <AE>

Attachments: RAI 413 Supplement 3 Response US EPR DC (Public) - DRAFT.pdf From: RYAN Tom (AREVA) [1]

Sent: Wednesday, December 22, 2010 9:56 AM To: Carneal, Jason Cc: PANNELL George (AREVA); BUDZIK Dennis (AREVA); Canova, Michael; BRYAN Martin (EXTERNAL AREVA); Tesfaye, Getachew; SLOAN Sandra (AREVA); BROWNSON Doug (AREVA)

Subject:

RE: Non-PROPRIETARY DRAFT Response to U.S. EPR Design Certification Application RAI No. 413, FSAR Ch. 7, Jason - due to this RAI 413 draft being on the agenda for the 1/4/10 Public Meeting, I am sending you the Non-Proprietary (Public) version.

Please let me know if you have any questions, Thank you, Tom Ryan Project Engineer Regulatory Affairs, New Plants AREVA NP An AREVA and Siemens company 7207 IBM Drive - CLT2B Charlotte, NC 28262 Phone: 704-805-2643, Cell : 704-292-5627 Fax: 434-382-6657 From: BRYAN Martin (External RS/NB)

Sent: Friday, December 17, 2010 4:55 PM To: Tesfaye, Getachew Cc: DELANO Karen (RS/NB); ROMINE Judy (RS/NB); BENNETT Kathy (RS/NB); RYAN Tom (RS/NB); PANNELL George (CORP/QP); HALLINGER Pat (EXT); WILLIFORD Dennis (RS/NB); BUDZIK Dennis (EP/PE); Carneal, Jason; Canova, Michael

Subject:

PROPRIETARY DRAFT Response to U.S. EPR Design Certification Application RAI No. 413, FSAR Ch. 7,

Getachew, To support the final response dates for nine questions in RAI 413, a draft response to RAI 413 questions 07.08-21, 07.08-24, 07.08-27, 07.08-28, 07.08-30, 07.08-31, 07.08-35, 07.08-37, and 07.08-40 is attached.

Because AREVA NP believes some of this material in the response is proprietary, an affidavit is attached to request withhold from public disclosure in accordance with 10 CFR 2.390. Let me know if the staff has questions of if the response can be submitted as final.

1

Tlhanks, Martin (Marty) C. Bryan Uf§. EPR Design Certification Licensing Manager AREVA NP Inc.

Ted: (434) 832-3016 7012561-3528 cell Martin. Bryan .ext(5areva.com From: BRYAN Martin (External RS/NB)

Sent: Monday, December 13, 2010 8:40 PM To: 'Tesfaye, Getachew' Cc":' DELANO Karen (RS/NB); ROMINE Judy (RS/NB); BENNE1T Kathy (RS/NB); RYAN Tom (RS/NB); PANNELL George (CORP/QP)

Subject:

Response to U.S. EPR Design Certification Application RAI No. 413, FSAR Ch. 7, Supplement 2

Getachew, AREVA NP provided a schedule for technically complete and correct responses to the questions in RAI 413 on September 08, 2010. Supplement 1 response to RAI No. 413 was sent on November 19, 2010, to provide a revised schedule.

To provide additional time to interact with the NRC a revised schedule is provided below (bolded dates have changed).

Question # Response Date RAI 413 07.08-10 March 15, 2011 RAI 413 07.08-11 March 15, 2011 RAI 413 07.08-12 March 15, 2011 RAI 413 07.08-13 March 15, 2011 RAI 413 07.08-14 March 15, 2011 RAI 413 07.08-15 January 28, 2011 RAI 413 07.08-16 March 15, 2011 RAI 413 07.08-17 March 15, 2011 RAI 413 07.08-18 January 28, 2011 RAI 413 07.08-19 February 22, 2011 RAI 413 07.08-20 January 28, 2011 RAI 413 07.08-21 January 28,2011 RAI 413 07.08-22 January 28, 2011 RAI 413 07.08-23 January 28, 2011 RAI 413 07.08-24 January 28, 2011 RAI 413 07.08-25 January 28, 2011 RAI 413 07.08-26 February 22, 2011 RAI 413 07.08-27 January 28, 2011 RAI 413 07.08-28 January 28, 2011 RAI 413 07.08-29 February 22, 2011 RAI 413 07.08-30 January 28, 2011 RAI 413 07.08-31 January 28, 2011 RAI 413 07.08-32 February 22, 2011 RAI 413 07.08-33 January 28, 2011 RAI 413 07.08-34 January 28, 2011 2

RAI 413 07.08-35 January 28, 2011 RAI 413 07.08-36 January 28, 2011 RAI 413 07.08-37 January 28, 2011 RAI 413 07.08-38 January 28, 2011 RAI 413 07.08-39 January 28, 2011 RAI 413 07.08-40 January 28, 2011 RAI 413 07.08-41 January 28, 2011 RAI 413 07.08-42 March 15, 2011 Sincerely, Martin (Marty) C. Bryan U.S. EPR Design Certification Licensing Manager AREVA NP Inc.

Tel: (434) 832-3016 702 561-3528 cell Martin. Bryan.ext(aareva.com From: BRYAN Martin (External RS/NB)

Sent: Friday, November 19, 2010 4:51 PM To: Tesfaye, Getachew' Cc: DELANO Karen (RS/NB); ROMINE Judy (RS/NB); BENNETT Kathy (RS/NB); PANNELL George (CORP/QP)

Subject:

Response to U.S. EPR Design Certification Application RAI No. 413, FSAR Ch. 7

Getachew, AREVA NP provided a schedule for technically complete and correct responses to the questions in RAI 413 on September 08, 2010. To provide additional time to interact with the NRC a revised schedule is provided below for questions 07.08-36, 07.08-39, and 07.08-41.

Question # Response Date RAI 413 07.08-10 March 15, 2011 RAI 413 07.08-11 March 15, 2011 RAI 413 07.08-12 March 15, 2011 RAI 413 07.08-13 March 15, 2011 RAI 413 07.08-14 March 15, 2011 RAI 413 07.08-15 December 17, 2010 RAI, 413 07.08-16 March 15, 2011 RAI 413 07.08-17 March 15, 2011 RAI 413 07.08-18 December 17, 2010 RAI 413 07.08-19 January 28, 2011 RAI 413 07.08-20 December 17, 2010 RAI 413 07.08-21 January 28, 2011 RAI 413 07.08-22 December 17, 2010 RAI 413 07.08-23 December 17, 2010 RAI 413 07.08-24 January 28, 2011 3

RAI 413 07.08-25 December 17, 2010 FAl 413 07.08-26 December 17, 2010 RAI 413 07.08-27 December 17, 2010

,RAI 413 07.08-28 December 17, 2010 RAI 413 07:08-29 January 28, 2011 RAI 413 07.08-30 January 28, 2011 RAI 413 07.08-31 January 28, 2011 RAI 413 07.08-32 January 28, 2011 RAI 413 07.08-33 December 17, 2010 RAI 413 07.08-34 December 17, 2010 RAI 413 07.08-35 January 28, 2011 RAI 413 07.08-36 December 15, 2010 RAI 413 07.08-37 January 28, 2011 RAI 413 07.08-38 December 17, 2010 RAI 413 07.08-39 December 15, 2010 RAI 413 07.08-40 January 28, 2011 RAI 413 07.08-41 December 15, 2010 RAI 413 07.08-42 March 15, 2011 Sincerely, Martin (Marty) C. Bryan U.S. EPR Design Certification Licensing Manager AREVA NP Inc.

Tel: (434) 832-3016 702 561-3528 cell Martin. Bryan.ext(areva.com From: BRYAN Martin (External RS/NB)

Sent: Wednesday, September 08, 2010 4:33 PM To: Tesfaye, Getachew Cc: DELANO Karen (RS/NB); ROMINE Judy (RS/NB); BENNETT Kathy (RS/NB); PANNELL George (CORP/QP)

Subject:

Response to U.S. EPR Design Certification Application RAI No. 413, FSAR Ch. 7

Getachew, Attached please find AREVA NP Inc.'s response to the subject request for additional information RAI 413.

The following table indicates the respective pages in the response document, "RAI 413 Response US EPR DC.pdf," that contain AREVA NP's response to the subject questions.

Question # Start Page End Page RAI 413 07.08-10 2 2 RAI 413 07.08-11 3 3 RAI 413 07.08-12 4 4 RAI 413 07.08-13 5 5 RAI 413 07.08-14 6 6 RAI 413 07.08-15 7 7 RAI 413 07.08-16 8 8 RAI 413 07.08-17 9 1 _.,9 RAI 413 07.08-18 10 10 4

P,ý1A 413 07.08-19 11 11

'F-%I 413 07.08-20 11 12

',AI413 07.08-21 13 13 FAI 413 07.08-22 14 14 Rr.qI 413 07.08-23 15 15 RAI 413 07.08-24 16 16 PAl 413 07.08-25 17 18 RAI 413 07.08-26 19 19 RAI 413 07.08-27 20 20 RAI 413 07.08-28 21 21 RAI 413 07.08-29 22 22 RAI 413 07.08-30 23 23 RAI 413 07.08-31 24 24 RAI 413 07.08-32 25 25 RAI 413 07.08-33 26 26 RAI 413 07.08-34 27 27 RAI 413 07.08-35 28 28 RAI 413 07.08-36 29 29 RAI 413 07.08-37 30 30 RAI 413 07.08-38 31 31 RAI 413 07.08-39 32 32 RAI 413 07.08-40 33 33 RAI 413 07.08-41 34 34 RAI 413 07.08-42 35 35 A complete answer is not provided for 33 of the 33 questions. The schedule for a technically correct and complete response to these questions is provided below.

Question # Response Date RAI 413 07.08-10 March 15, 2011 RAI 413 07.08-11 March 15, 2011 RAI 413 07.08-12 March 15, 2011 RAI 413 07.08-13 March 15, 2011 RAI 413 07.08-14 March 15, 2011 RAI 413 07.08-15 December 17, 2010 RAI 413 07.08-16 March 15, 2011 RAI 413 07.08-17 March 15, 2011 RAI 413 07.08-18 December 17, 2010 RAI 413 07.08-19 January 28, 2011 RAI 413 07.08-20 December 17, 2010 RAI 413 07.08-21 January 28, 2011 RAI 413 07.08-22 December 17, 2010 RAI 413 07.08-23 December 17, 2010 RAI 413 07.08-24 January 28, 2011 RAI 413 07.08-25 December 17, 2010 RAI 413 07.08-26 December 17, 2010 RAI 413 07.08-27 December 17, 2010 RAI 413 07.08-28 December 17, 2010 RAI 413 07.08-29 January 28, 2011 RAI 413 07.08-30 January 28, 2011 RAI 413 07.08-31 January 28, 2011 5

RAI 413 07.08-32 JanuarV 28, 2011 JRAI 413 07.08-33 December 17, 2010 RAI 413 07.08-34 December 17, 2010 RAI 413 07.08-35 January 28, 2011 RAI 413 07.08-36 November 19, 2010 RAI 413 07.08-37 January 28, 2011 RAI 413 07.08-38 December 17, 2010 RAI 413 07.08-39 November 19, 2010 RAI 413 07.08-40 January 28, 2011 RAI 413 07.08-41 November 19, 2010 RAI 413 07.08-42 March 15, 2011 Sincerely, Martin (Marty) C. Bryan U.S. EPR Design Certification Licensing Manager AREVA NP Inc.

Tel: (434) 832-3016 702 561-3528 cell Martin. Bryan.ext(@areva.com From: Tesfaye, Getachew [2]

Sent: Monday, August 09, 2010 3:46 PM To: ZZ-DL-A-USEPR-DL Cc: Mott, Kenneth; Spaulding, Deirdre; Jackson, Terry; Canova, Michael; Colaccino, Joseph; ArevaEPRDCPEm Resource

Subject:

U.S. EPR Design Certification Application RAI No. 413(4772), FSAR Ch. 7 Attached please find the subject requests for additional information (RAI). A draft of the RAI was provided to you on June 4, 2010, and discussed with your staff on July 22, 2010. Draft RAI Questions 07.08-19, 07.08-21, 07.08-23, and 07.08-41 ,were modified as a result of that discussion. The schedule we have established for review of your application assumes technically correct and complete responses within 30 days of receipt of RAIs. For any RAIs that cannot be answered within 30 days, it is expected that a date for receipt of this information will be provided to the staff within the 30 day period so that the staff can assess how this information will impact the published schedule.

Thanks, Getachew Tesfaye Sr. Project Manager NRO/DNRL/NARP (301) 415-3361 6

Request for Additional Information No. 413(4772), Revision 1, Supplement 3 8/9/2010 U. S. EPR Standard Design Certification AREVA NP Inc.

Docket No.52-020 SRP Section: 07.08 - Diverse Instrumentation and Control Systems Application Section: ANP-10304 QUESTIONS for Instrumentation, Controls and Electrical Engineering I (AP1000/EPR Projects) (ICEIt)>

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 2 of 56 Question 07.08-21 Provide additional explanation of the changes made to S-RELAP5 as utilized for the D3 analysis presented in ANP-10304 Rev 1, including the following:

a. A description of Heat Transfer modifications (i.e., fluid temperature, Inayatov multiplier, LIQHTC) made to the S-RELAP5 code and the purpose of the change, and

b. The validation basis for Tavg change made to the S-RELAP5 code.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protective function. The Staff Requirements Memorandum to SECY 93-087, Item II.Q, states that the vendoroi'°applicant shall analyze each postulated common-mode failure for each event and shall dem 0onstrate adequate diversity within the design for each of these events.

Section A.2.5 of Appendix A to ANP-1 0304 Rev 1 state/s that minor cha ges to the S-RELAP5 code were made to reflect improved heat transfer in the ste generator secondary system.

Response to 07.08-21 Purpose of S-RELAP5 heat transfer model modification The purpose of the S-RELAP5 heatetransfer modification is to provide a more realistic computation of steam generator hea transfer.

Previous S-RELAP5 computations havet-use -ariety of input choices to increase steam generator heat transfer to achieve target secondary side pressure during steady-state initialization. Typical input'choices have included an increase in steam generator tube area and the use of a small hydraulic diameter toiJncrease the steam generator heat transfer. It is desired to use more realisticinput options to achieve the target secondary side pressure.

Description of the chan S-RELAP5 includes two new input options that improve steam generator heat transfer. They are:

1. An option for "Tare"; and,

2. An option to use the "lnayatov" heat transfer multiplier.

Both options improve the accuracy of heat transfer computation in steam generators and aid in achieving the target secondary side pressure. While an adjustment in heat transfer is still needed to get the target secondary side pressure, it is done by applying the LIQHTC heat transfer coefficient multiplier available in S-RELAP5 input.

Option to Use Average Fluid Temperature for Wall Heat Transfer The original S-RELAP5 heat transfer computation from the wall to the fluid is based on the fluid nodal temperature adjacent to the surface of the heat structure. Figure 07.08-21-1 shows two scalar nodes that are the fluid control volumes for mass and energy. Temperatures are shown

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 3 of 56 at scalar cell centers, with, TK on the left side and TL on the right. The heat transfer from a heat structure adjacent to Node L uses TL, but, that temperature is actually assigned to the nodal boundary (exit), so, if the flow is from the left-to-right, TK is assigned to j and TL is assigned to j+1.

A more accurate temperature for computing wall heat transfer is to use an average of the assigned temperatures for the fluid node facing the heat structure. That change is implemented as an option in S-RELAP5 for best estimate non-LOCA analysis.

Option to Include Inayatov Multiplier Single-phase, turbulent, forced convection heat transfer coefficient in S-RELAP5 is based on the Dittus-Boelter correlation as follows:

h. = max[O.023 Re'.8 pr*4,7.86]kf Dh To include the Inayatov multiplier in the Dittus-Bo4 insfer correlation the above correlation is modified as follows:

h = max/0.023 D 2 Ref 7Pr8*.,

The Inayatov multiplier (P 1 P2/1 5nJfrmELAP5/MOD3.3. P1 and P 2 are the tube pitches and D is the tube oute eter..J-he multiplier range is from 1.1 to1.6. The U.S.

EPR steam generator has-a-u gular tube array, so P1 = P 2 = P. The lnayatov multiplier therefore becomes -F .AP5 has an input option for P/D.

The Inayatov multiplier is appl the same way to two-phase forced convection using the Chen correlation (macro"-part).

LIQHTC Multiplier LIQHTC is a liquid heat transfer coefficient multiplier that is part of the S-RELAP5 Code Uncertainty Analysis input. LIQHTC multiplies the single phase heat transfer coefficient and also the nucleate boiling heat transfer coefficient macro term. LIQHTC is selected manually for use in the steady-state initialization to achieve the desired secondary side pressure. LIQHTC continues to be applied during the transient calculation.

Validation approach The validation computations demonstrate the improved accuracy of the Tave option. S-RELAP5 can produce results consistent with the theoretical solution and independent nodal solutions using Texit and Tave.

AREVA NP Inc.

Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 4 of 56 All validation computations use a simple steam generator consisting of a single tube and a tube wall. By defining boundary conditions on the tube wall and at the inlet and exit of the channel, it is possible to consider three types of validation computations.

1. Simple Steam Generator The steam generator is modeled with a constant shell side temperature to represent Tsat and a constant tube side heat transfer coefficient. This steam generator has a theoretical solution from which it is possible to assess the accuracy of the Texit and Tave fluid temperature definitions. The steam generator also has nodal (finite difference) solutions for the Texit and Tave fluid temperature definitions that have a direct relationship to the S-RELAP5 solutions. This simple modeling is also used to define input files for S-RELAP5 for a variety of situations including the Texit and Tave options and with forward and reverse flow.

The fluid temperatures computed by S-RELAP5 are compared to the theoretical and nodal solutions in Excel.

2. Single-Phase Forced Convection The steam generator is modeled as in Item 1, except tha>the single-phase heat transfer coefficient is computed by using the S-RELAP5,,eat transfer correlations. The shell (right) side wall temperature is constant and the tube (left) side water temperature decreases by heat transfer across the tube wall. The,,ieattransfercoefficient computation also includes the Inayatov multiplier selected by inputioption. The single-phase heat transfer coefficient computed by S-RELAP5 is compared to its ndependnt computation in Excel.

3. Two-Phase Forced Convection The steam generator is meled as in-Item 1 except roles of the shell and tube are switched. The shell (right) side.boils and-the two-phase heat transfer coefficient is

\'e computed on the shellisTd-e by using the S-RELAP5 heat transfer correlation for two-phase forced convectiqn The heat-tran'sfircoefficient computations also include the Inayatov multiplier. The tLbe (left) side wall temperature is a constant and greater than Tsat on the shell (right) side, thus, the shell-side water boils. The two-phase heat transfer coefficients computed by S-REAPh5 are compared to its independent computation in Excel.

Mathematical Solutions fora Simple Steam Generator First consider the theoretical solution for a simple steam generator. Then consider the nodal solutions using the Texit and Tave definitions.

Theoretical Solution The theoretical solution for heat transfer in a simple steam generator is based on the following:

" Hot water, at a constant flow rate, drives the heat transfer process across a tube wall where the outer surface is constant at Tsat. There is no subcooled region on the Tsat side of the wall.

• The circular tube wall is arbitrarily thin to eliminate the thermal resistance of the wall.

• The tube-side heat transfer coefficient is constant.

AREVA NP Inc.

Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 5 of 56 The heat transfer process for this simple steam generator has a theoretical solution that can be defined from first principals. The solution is expressed as:

• (T - Ta,, (Tý - T.,

where a= hJ MC Nodal Solution Usino T-... - -1 A finite difference solution for the liquid temperature in the sara'e'simple steam generator, using Texit, produces a recursive relationship that can be e xpresseds:\

Tj-Ts,,,- (l+aAx) u Given the inlet temperature, the nodal temperature Tj canbe advanced along the channel.

Nodal Solution Usina T.-

A finite diffeYehcbe solution for the in the same simple steam generator, using Tave, produces a recursive relatjol T, - TS. ,)

Given the inlet temperatur'e, the nodal temperature Tj can be advanced along the channel.

S-RELAP5 Model of Simple Steam Generator S-RELAP5 is used to model the simple steam generator as described above. It is directly applicable to the theoretical and nodal solutions considered in Excel. The modeling consists of a flow channel (pipe component) with volumes placed at the each end for application of pressure and temperature boundary conditions. Those volumes are connected to the flow channel by junctions at each end. The inlet junction defines the flow boundary condition.

Heat is transferred between the fluid channel and the tube wall heat structures.' The wall is cylindrical, thin and has high thermal conductivity to eliminate its resistance to heat transfer.

The heat transfer coefficient is held constant. The outer surface of the wall is also held at a constant temperature. Figure 07.08-21-1 shows the nodal arrangement.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 6 of 56 Validation summary results The following plots present representative results from the three types of validation computations.

Simple Heat Exchanger Figure 07.08-21-3 shows the theoretical solution and the Texit solutions from S-RELAP5 and the corresponding Texit nodal solution from Excel. The solutions using Texit are in agreement, but, they are both above the temperatures from the theoretical solution.

Figure 07.08-21-4 shows the results of the same computation but using Tave in S-RELAP5 and in the corresponding nodal solution from Excel. The overlay of the three solutions shows the improvement when using Tave. The plots also show the excellent agreement among all three computations. The same excellent agreement is found for reverse flow and when switching the tube and shell sides.

The agreement between the independent computations done in Excel for the temperature in a simple steam generator and the results of the S-RELAP5 computations validate the implementation of the Tave option in S-RELAP5.

Single-Phase Forced Convection Validation?

This case considers the following:

• The single-phase forced convection heat transfer on the tube (left) side is computed by using the S-RELAP5 heattransfer correlations presented above.

• The shell (right) side tube wall temperature is a constant.

• An lnayatov multiplier of RID 1.5 is selected for this validation computation.

The shell (right) side wall temperature is a constant. The heat transfer across the tube wall from the tube side to the shell side decreases the temperature of the liquid on the tube (left) side.

Excel performs an independent calculation of the heat transfer coefficient including an lnayatov multiplier of 1.5. The heat transfer coefficient is uniform because of uniform flow rate and fluid properties.

Figure 07.08-21-5 shows the heat transfer coefficient from S-RELAP5 and from Excel. The heat transfer from S-RELAP5 is essentially a constant along the length as expected. The Excel value is very close to the result computed in S-RELAP5.

The independent computations in Excel validate the implementation of the lnayatov multiplier applied to the single-phase heat transfer coefficient computed in S-RELAP5.

Two-Phase Forced Convection Validation This case considers the following:

• S-RELAP5 computes the two-phase forced convection heat transfer coefficients on the shell (right) side.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 7 of 56

" Set shell (right) side temperature to Tsat.

• Set tube (left) side temperature greater than Tsat to transfer heat to the shell side.

" Set the Inayatov multiplier to 1.5 on the shell side. The value of 1.5 is arbitrary.

The two-phase heat transfer coefficient is defined by the Chen correlation. The Chen correlation for boiling heat transfer consists of "macro" and "micro" heat transfer coefficients terms as follows:

q"=hma,(Tw - Tf )+ hmic(Tw - at,)

The forced convection or "macro" part of the Chen correlation is bWsed on the Dittus-Boelter correlation, and, with the Inayatov multiplier, it is written as:

05 hm,, 0.023- 0 - .53 Re0 8 Prf0 4 -- F The F-factor and the "micro" part of the Chen correlation are not modified by the Inayatov multiplier. The F-factor is defined as:

l 1.0 - 0 .1 *1" ' ,

2.3 5 >1\

For homogeneous two-phase fow thelMaIrtinelliLparameter is defined as:

0,5, 0.1 Note that the inverse of the Martinelli parameter is just notation and that an inverse is not actually done. The parameterVis defined in terms of steam quality and the fluid properties as shown.

The two-phase forced convection heat transfer coefficient is independently computed in Excel.

The Excel computation mimics the S-RELAP5 computation by computing the two-phase mixture enthalpy rise and steam quality along the heated channel. The components of the Chen correlation are computed as a function of quality. The combined macro and micro heat transfer coefficients are reported along with the S-RELAP5 coefficients in Figure 07.08-21-6. The heat transfer coefficient increases with distance from the inlet as the vapor quality increases. This agreement is very good given the complexities of the two-phase heat transfer computation and the changing fluid properties with pressure along the flow path in the S-RELAP5 computation.

The Excel computation uses uniform fluid properties representative of those from taken from the corresponding S-RELAP5 case.

The independent computations in Excel validate the implementation of the Inayatov multiplier in the two-phase heat transfer coefficient computed in S-RELAP5.

AREVA NP Inc.

Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 8 of 56 Nomenclature Subscripts C Specific heat capacity, BTU/(Ibm- 0 F) Ave Average D Diameter,'ft Exit Exit Dh Hydraulic diameter, ft F fluid, liquid F Chen correlation F-Factor In Inlet h Heat transfer coefficient, BTU/(sec-ft 2- OF) J Nodal index k Thermal conductivity, BTU/(sec-ft-°F) Mac Macro m Mass flow rate, Ibm/sec Mic Micro P Tube pitch, ft Sat saturation Ph Heated perimeter, ft W Wall Pr Prandtl number q,, Heat flux, BTU/(sec-ft 2)

Re Reynolds number T Temperature, OF x Axial distance, ft X Steam quality, mass fraction P Viscosity, Ibm/(ft-sec) 3 p Density, Ibm/ft Xtt-1 Martinelli Parameter

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 9 of 56 Figure 07.08-21-1 -S-RELAP5 One-Dimensional Nodalization Scheme Mass and energy control volume or cell Vector node or junction __

Vg, Vf I I Nodal Model for Simple Steam Generator IHIS 124, 10 nodes!

JUN 111 JUN 131 Comp 101 I-* -T Comp 201 FcCm p 12,4, 10 noe~s:]

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 10 of 56 Figure 07.08-21 Steam Generator Solutions using Teit Tube Side Temperature, Using Texit

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 11 of 56 Figure 07.08-21-5 - Single-Phase Heat Transfer Coefficients LIQUID HEAT TRANSFER COEFFICIENT 6.0 5.0 ur 4.0 i-i-

3.0

.I-2.0

.. i 1.0 0.0-0 Figure 07.

6.0 4 5.0 4.0 I-3.0 I-2.0 M 1.0 0.0 10 Distance from Inlet, ft

AREVA NP Inc.

Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 12 of 56 FSAR Impact:

The U.S. EPR FSAR will not be changed ais a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 13 of 56 Question 07.08-24 Provide additional information to support the reliability of the predicted DAS trip actuation on Low SG Level, including the following:

a. A description of the S-RELAP5 SG level model and the SG narrow range level instrumentation model, including validation basis, and

b. An evaluation of the decalibration effects of the MSIV closure on the narrow range instrumentation and how decalibration is treated in the simulation model.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protective function. The Staff Requirements Memorandum to SECY 93-087, Item II.Q, states that the vendor,,or applicant shall analyze each postulated common-mode failure for each event and shall dermonstrate adequate diversity within the design for each of these events.

The engineering analysis of the Inadvertent MSIV CIosure event presented in Section A.3.3.2 of ANP-10304 Rev 1 shows that a DAS reactor trip actuation occurs on Low SG Level, terminating the secondary system pressure excursion at about 164'0 psia, or 100 psia below the D3 analysis criterion of 120 percent of secondary system design pressu re (1.20 x 1450 psia = 1740 psia).

The Inadvertent Closure of MSIV event analy.ed-in the FSAR results in PS reactor trip actuation (at approximately 6 seconds) on High SG Pressure/,,which>is not available through DAS.

Therefore, an explicit D3 engineeringanalysis~\s pr'ovided in ANP-1 0304 Rev 1 Section A.3.3.2 for the Inadvertent Closure of MSIV event. The D3 analysis shows that at approximately 130 seconds DAS provides a reactor trip actuation on Low SG Level. The pressure in the affected SG reaches approximately 1640 psiaerri1A.3%iOf the secondary system design pressure (design pressure is 1450.psia, per,,FSAR Table 10.3-1). Considering the rate of change of the SG pressure excursion and its calculated peak value relative to the D3 analysis criterion (1640 psia peak SG pressure versus 1\740 psiaýcriterion), additional information on the SG level model and the DAS Low SG\Ievel trip funiction is required in order for staff to complete its review of the Inadvertent Closure of'MSIV D3 analysis.

Response to 07.08-24 Item 07.08-24(a):

Description of S-RELAP5 Narrow Range (NR) and Wide Range (WR) steam generator (SG)

Level Model The S-RELAP5 model for the best estimate non loss-of-coolant accident (non-LOCA) analyses implements a mechanistic tap-to-tap differential pressure measurement of the SG narrow range and wide range level signals. This scheme is used only in the best estimate non-LOCA model, where steam generator (SG) pressure can rise rapidly following main steam isolation valve (MSIV) closure or turbine trip. This can potentially cause SG level shrink that may not be tracked with complete accuracy by the more conventional collapsed liquid level scheme used in the U.S. EPR FSAR Tier 2, Chapter 15 analyses. Since nominal instrument setpoints are used in the best estimate analyses, capturing the SG level shrink/swell phenomena, if it occurs, is an important consideration. The best estimate small-break LOCA (SBLOCA) S-RELAP5 model

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 14 of 56 retains the collapsed liquid level scheme used in the U.S. EPR FSAR Tier 2, Chapter 15 design basis analyses because rapid SG secondary pressure and level changes are not expected for SBLOCA scenarios.

The tap-to-tap differential pressure is comprised of the constituent liquid and vapor static heads, as follows:

Plower tap - Pppertap : pfgAzf + Pg (1) where the effective heights of the liquid and vapor constituents sum to the overall tap-to-tap elevation difference:

Azf + AZg = Zuppertap -- Z lower tap (2)

The steam generator liquid level is indicated as the effective heightof the liquid, which is expressed as a percentage of the tap-to-tap span, is atfllows:

Liquid level = Azf X100% (3) zupper tap Zlower tap The effective height of the liquid, in terms of the tapse-to, on in~~~~~ thisp

~~~ rltomabeotiefrmeaio / pderential pressure, for substitution in this relation may be obtained from eq*uation\(),),;\gy first solving equation (2) for the effective height of the vapor and substitutingitinto equaoi6n (1),as follows:

rtý - Azf Plower tap -Pupper tap , zo tap-.( er tap f Pggta+( lupper tap which may be rearranrged to obtain:

Az Plower tap "l\ uppe taPp- g (Zupper tap - Zlo... tap

\/(Pj - P )g By incorporating this result into equation (3) and simplifying, the liquid level may be expressed as:

Liquid level = rPlower ta t P - Pupper ap Zuppertap - Ziowertap

_ Pgg"x X 100%

(4)

When implementing this into the S-RELAP5 model, it was assumed based on preliminary I&C design information that the plant's narrow range and wide range liquid level indication algorithms would both be calibrated for saturated conditions at 1200 psia, and would not be corrected for transient-related deviations from those conditions. The densities in equation (4) are therefore treated as constant values and are evaluated at the 1200 psia saturated state regardless of the dynamic conditions in the SG. The constants are defined in the non-LOCA

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 15 of 56 model supplied in the Response to RAI 413, Question 07.08-41, as control variables (CVs) 40 (for pf) and 41 (for pg).

It is also assumed that the S-RELAP5-calculated pressures for the model volumes containing the taps as listed in Table 07.08-24-1, with appropriate adjustments to account for the dynamic heads and the static head differences between the volume-centered elevations and the tap elevations, are suitable to use as the tap pressures in equation (4).

For example, using the S-RELAP5 model volume numbers shown in Figure 07.08-24-1 for SG 1:

Table 07.08-24-1 -SG Level P5 Volume Elevations L

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application , Page 16 of 56 Applying the elevations in Table 07.08-24-1 and equations (4), (5), and (7) above, the narrow range liquid level for SG 1 becomes:

- -K Similarly, applying the elevations'in/w/ ale 0e'.081-24-1 and equations (4), (5), and (7) above, the

_ i l wide range liquid level for SG 1becomes-~

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 17 of 56 The original collapsed liquid level control functions are retained in the non-LOCA model for comparison purposes, but all trips and controls that reference SG level use the new differential pressure signals.[

]

Item 07.08-24(b):

Steam Generator Level Transient Response The models of SG level are compared for the main steam isolation valve (MSIV) event because it presents the more severe drop of SG level.

Figure 07.08-24-2 shows the NR level response for the4MSIV closur&e*vent for end of cycle (EOC) conditions. The reactor trip on diverse actuatiOn system (DAS) low SG level at 15 percent NR level occurs at 133 seconds after MSllosure./The plots show that ifthe collapsed liquid level scheme had been used for DASqinput instead of the differential pressure algorithm the reactor trip (see Figure 07.08424--3.) on lowlevel would have occurred 5 to 10 seconds earlier.

Figure 07.08-24-4 shows the WR level responsefor the MSIV closure event. In this case the collapsed liquid level signal leads tlhe differential pressure WR signal, demonstrating that the shrink/swell phenomenon has*a larger influence\on the full-range level transmitters than for the narrow range transmitters.

Figure 07.08-24-5 shows the butlet pressure for the affected SG. The timing of the pressure peak at -140 seconds coincides with the approximate time of minimum SG WR level (differential pressure.,signal) shown in Figure 07.08-24-4.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 18 of 56 Figure 07.08-24-1 - Diversity and defense-in-depth (D3) Secondary System Nodalization

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 19 of 56 Figure 07.08-24 SG-4 NR Level - D3 MSIV Closure E)SG-L NR <cv8l02>

--- El Collapsed Liq Level <cv8l 01>

100 90 0

80 70 CL) 0~ 60

_0

("

50

-j U(o 40 c° z(D 0

30 0) 20 10 0

0 100 200 300 400 500 Time (s)

ID:21007 3Nov2009 08:34:03 D3-MSIVC eoc.dmx

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 20 of 56 Figure 07.08-24 Reactor Power - D3 MSIV Closure 150.0 100.0 3,-

0 0

t5 n'-

50.0 0.0 0 100 200 300 400 500 Time (s)

ID:21007 3Nov2009 08:34:03 D3-MSIVC eoc.dmx

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 21 of 56 Figure 07.08-24 SG-4 WR Level - D3 MSIV Closure

- SG-L NR <cv8l54>

--- E Collapsed Lig Level <cv8153>

100 90 80 70 0~

C,)

60 C-)

50 (D

CO 40

_0 30 20 10 0 L 0 100 200 300 400 500 Time (s)

ID:21007 3Nov2009 08:34:03 D3-MSlVCeoc.dmx

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 22 of 56 Figure 07.08-24 SG-4 Outlet Pressure - D3 MSIV Closure 1800.0

=3 0-0)

C/)

500.0 FSAR Impact:

The U.S. EPR FSAR will hAchanged as a result of this question.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 23 of 56 Question 07.08-27 For the RCP Rotor Seizure event described in ANP-1 0304 Rev 1 Section A.3.4.3, provide the following additional information:

a. A comparison of the RCS flow coastdown rate assumed for D3 evaluation versus the flow coastdown rate shown FSAR Figure 15.3-9;

b. The calculated initial (t=O seconds) and minimum DNBR for the RCP Rotor Seizure event as analyzed in both Section 15.3.3 of the FSAR and the D3 analysis;

c. A table that lists the sequence of events for the RCP Rotor Seizure event, showing times of DAS reactor trip actuation, beginning of control rod insertion, minimum DNBR, and turbine trip.

In addition, explain the differences in initial DNBR margin between the FSAR Chapter 15 analysis and the D3 analysis, i.e., identify the best estimare assumptions and assess their beneficial effects on IDNBR.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in component design and.phrinciples of operation, shall be used to the extent practical to prevent loss of the protective function. The Staff Requirements Memorandum to SECY 93-087, Item II.Q, S/atie"s-that the vendor or applicant shall analyze each postulated common-mode failure for each e aent and shall d'e'monstrate adequate diversity within the design for each of these events."

ANP-1 0304 Rev 1 Section A.3.4.. rovides an engineering argument for the RCP Rotor Seizure event (PA), stating that DAS wIll~actuate/a-reactor'trip on Low-Low RCS Flow (one loop), and therefore provide protection comparable to tlYe-PS as shown in the FSAR Section 15.3.3 analysis. The FSAR -apa!*i..shst that the DNBR SAFDL is exceeded, resulting in fuel damage. The staff is/not able to identifydesign descriptions that would permit sufficient understanding in order to complete the safety evaluation.

Response to 07.08-27 The reactor coolant pump (RCP) rotor seizure event was not specifically analyzed for the diversity and defense-in-depth (D3) assessment, but was evaluated by a quantitative comparison with U.S. EPR FSAR Tier 2, Chapter 15 analysis using best estimate assumptions.

The overall conclusion from this comparison was that the U.S. EPR FSAR Tier 2, Chapter 15 analysis is bounding and that fuel failure fractions and offsite consequences would be less for D3 as a result of the use of best estimate assumptions.

In the D3 assessment the potential detrimental effect of the lower diverse actuation system (DAS) low-low reactor coolant system (RCS) flow trip setpoint and longer time delay was offset by the use of best estimate assumptions. The best estimate assumptions were all beneficial in the comparison demonstrating, in overwhelming fashion, that the U.S. EPR FSAR Tier 2, Chapter 15 analysis is bounding. This is illustrated below by discussing the U.S. EPR FSAR Tier 2, Chapter 15 analysis assumptions with a comparison to the corresponding best estimate assumption.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 24 of 56 The U.S. EPR FSAR Tier 2, Chapter 15 system and core response for the rotor seizure event is evaluated with the S-RELAP5 and LYNXT computer codes. It was conservatively predicted that eight percent of the core experiences departure from nucleate boiling (DNB)-induced cladding failure. Radiological analysis of this event assumed a bounding fuel failure fraction of 9.5 percent. As presented in U.S. EPR FSAR Tier 2, Table 15.3-1-Decrease in Reactor Coolant System Flow Rate Events - Key Input Parameters, the initial conditions are biased to achieve conservative results. Specifically, the parameters power, pressure, and RCS flow rate, are chosen to penalize the departure from nucleate boiling ratio (DNBR) evaluation. Table 07.08.27-1 presents the comparison between parameters used in U.S. EPR FSAR Tier 2, Section 15.3 and those assumed in the D3 assessment.

Although the low-low RCS flow trip would be delayed in the D3 analysis (DAS setpoint) as compared to the U.S. EPR FSAR Tier 2, Chapter 15 analysis, this delay is only a fraction of a second as a result of the steep flow decrease from the seized rotor shown in U.S. EPR ESAR Tier 2, Figure 15.3-9. The flow to the inlet of the core in the';D3 analysis would not be impacted as much because the RCPs in the unaffected loops wo~uld continue

• run. In the U.S. EPR FSAR Tier 2, ChapterI 5 analysis, loss of offsite power/(LOOP) was assumed upon reactor trip at which time the RCPs in the unaffected loops coast down./The initial flow in the RCS is higher under best estimate conditions. The initial flow assuedan1the n..th U.S.

U.S. EPR EP FSAR Tier 2, Chapter SRTer2/hpe 15 analysis is based on thermal design flow with a design core bypass fraction. Overall these flow effects are conservatively judged to be a-slight penalty~on DNB performance.

The initial power distributions in the U.S. EPR FAR'Ter 2, Chapter 15 analysis as compared to best estimate conditions result in alsgnificant reduuction in Fq (50 percent) and FAH (15 percent) as illustrated in Table 07.08-27-1/ he)Fesponse to RAI 413, Question 07.08-19 provides an overview of the differences between the power distributions used in the D3 analyses as compared to the power distributionsit iitithei U.S. EPR FSAR Tier 2, Chapter 15. Overall, this results in a significant improvement in

1.  % initial _DNB performance as illustrated in Table 07.08.27-The Response to R\113, Question 07.08-23 provides an overview of the differences in the SCRAM reactivity curves which illstrate that, under the best estimate assumptions, with respect to the U.S. EPR PSAR Tier 2, Chapter 15 analysis, SCRAM reactivity is greater at any given time following a reactor'trip which results in a lower power-to-flow ratio throughout the event which is beneficial with respect to DNBR performance.

This discussion demonstrates that the credited best estimate items for D3 more than offset the effects of a delayed reactor trip on DAS, and that the DNB performance is equal to or better than in the U.S. EPR FSAR Tier 2, Chapter 15 analysis. Therefore, the U.S. EPR design is adequate in addressing a software common cause failure in the protection system (PS) during RCP rotor seizure and shaft break events.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 25 of 56 Table 07.08-27-1 - D3 Rotor Seizure Event Parameters- Comparison FSAR Tier 2, Chapter 15 versus D3 FSAR Tier 2, Parameters D3, DNBR Section 15.3.3 Evaluation DNBR evaluation Initial reactor power (MWt) 4590 4612 Average RCS temperature (OF) 594 594+/-4 Initial PZR pressure (psia) 2250 2250+/-50 Initial RCS loop flow rate (gpm) 124,741 119,692 Low-low flow trip setpoint (time delay) 44% (1.30,4ec) 50% (1.05 sec)

Flow to core Inlet RCPs'irn3 Impacted by unaffectedl oops LOOP (no RCPs o*,ctinue, I available)

Fq 2.6 FAH 1.70 Scram (pcm) BOC--HFP 6161($10.35)

EOC', iFý ' 7353 ($14.28)

Core bypass fraction (%) \ n 5.5 Initial DNBR (normalized to/§SAFDL) \ 2.41 1.30 FSAR Impact:

The U.S. EPR FSAR,, will not be changed as a result of this question.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 26 of 56 Question 07.08-28 For the Uncontrolled RCCA Withdrawal at Power event described in ANP-10304 Rev 1 Section A.3.5.2, provide the following additional information:

a. The calculated initial (t=0 seconds) and minimum DNBR for the Uncontrolled RCCA Withdrawal at Power event, and

b. A table that lists the sequence of events for the Uncontrolled RCCA Withdrawal at Power event, showing times of DAS reactor trip actuation, beginning of control rod insertion, minimum DNBR, and positioning of the RCCA bank being withdrawn.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in component design and principl s of operation, shall be used to the extent practical to prevent loss of the protective function/,The Staff Requirements Memorandum to SECY 93-087, Item lI.Q, states that the .e dor,ýrrapplicant shall analyze each postulated common-mode failure for each event and shall, demonstrate adequate diversity within the design for each of these events.

The Uncontrolled RCCA Withdrawal at Power event is\anayzed assuming full power initial conditions at both BOC and EOC conditions, and with the RCCAs inserted to the Technical Specification Power Dependent Insertion Lirmiit-(PDIL). The time in cycle life-time (BOC vs.

EOC) affects moderator temperature coeffc\ient-an -control,rd bank worth. The FSAR Section 15.4.2 analysis states that the reactor system is protected.by PS Low DNBR, High LPD, Excore High Rate of Change, High Core PoWer-Levewl\d High Pressurizer Level reactor trip functions, none of which are provided by DAS'.The, D3 engineering analysis provided in ANP-10304 Rev 1 Section A.3.5.2, however, shows, that 2 ~s actuates a reactor trip on Low SG Level and that reactor power peaks at approximt"ly1008 erce The staff is not able to identify design descriptions that would permit sufficient understanding in order to complete the safety evaluation.

Response to 07.08-2*8 Item 07.07-28 (a):

The departure from nucleate boiling ratio (DNBR) normalized to the specified acceptable fuel design limit (SAFDL) values are given rather than the calculated values to facilitate direct comparison to the results presented for the Uncontrolled Rod Cluster Control Assembly (RCCA)

Withdrawal at Power transient in U.S. EPR FSAR Tier 2, Figure 15.4-54.

For the diversity and defense-in-depth (D3) Uncontrolled RCCA Withdrawal at Power transient analysis, the lowest DNBR is reached during the beginning of cycle (BOC) case. The transient is simulated by an insertion of reactivity corresponding to the withdrawal of the Bank D from the PDIL to the all rods out position at the maximum RCCA extraction speed. The initial DNBR normalized to the specified acceptable fuel design limit (SAFDL) is 2.41 and the minimum DNBR normalized to the SAFDL value reached during the transient is 2.00 at 287.4 seconds.

Item 07.07-28 (b):

Table 07.08-28-1 presents the sequence of events for the Uncontrolled RCCA Withdrawal at Power transient.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 27 of 56 Table 07.08-28 RCCA Withdrawal at Power - Sequence of Events Event Time (sec)

Bank D Withdrawal from PDIL beginning 0.0 Bank D withdrawal from PDIL end 72.0 DAS low SG level delay (RT signal) 290.1 DAS RT with delay (rod release for scram) 290.5 Minimum DNBR 288.0 DAS turbine trip (TT) with delay 291.1 The Uncontrolled RCCA Withdrawal at Power transient is shown to be more challenging in the U.S. EPR FSAR Tier 2, Section 15.4.2 evaluation than in the D3'*nalysis. This is because of differences between the U.S. EPR FSAR Tier 2, Chapter 15/.an&D3 analyses that impact the

/ / \1 N initial DNBR and linear power density (LPD) margins, the use of best estimate core physics parameters, and other modeling differences that contribute to the overall event progression and severity.

Initial DNBR and LPD Margin The Response to RAI 413, Question 07.08-,l9-will-proovidean overview of the differences between the best estimate assumptions used in 'the D3 analysis as compared to the parameters within the U.S. EPR FSAR Tier 2, C*hapter 15ig,\Tlie key it6ems affecting the initial margin in DNBR and the LPD for the uncontrlf'edlRCCA\wvlthdrawal at power, as compared to the U.S.

EPR FSAR Tier 2, Section 15 nalysis, include the following:

Power Distributions The power distributio applied for ,,, L.3Uncontrolled RCCA Withdrawal at Power DNB and LPD analysis is a representative'nominal power distribution for beginning of cycle (BOC) conditions. These best estimate power distributions will be described in the Response to RAI 413, Question 07.08-19-Part b Numerous power distributions are evaluated for actuation of the low DNBR reactor trip for the U.S. EPR FSAR safety analysis, as described in the Incore Trip Setpoint and Transient Methodology for U.S. EPR Topical Report, ANP-10287P-000.

To initiate the Uncontrolled RCCA Withdrawal at Power transient for the U.S. EPR FSAR Tier 2, Section 15.4.2 analysis, these power distributions are scaled to the low DNBR limiting conditions for operation (LCO), the high LPD LCO, or the FAH LCO, whichever occurs first. For the representative case presented in U.S. EPR FSAR Tier 2, Figure 15.1-17, the power distribution was scaled from a nominal FDH of 1.42 to 1.47 (3.5 percent increase) to initiate the transient.

Application of Uncertainties D3 DNBR analyses use best estimate S-RELAP5 thermal-hydraulic boundary conditions within the subchannel code LYNXT to confirm the DNBR is above the technical specification safety limit of 1.0 with a 95 percent probability and 95 percent confidence (95/95) level. Since best

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 28 of 56 estimate conditions are applied for D3 DNBR analyses, the 95/95 level is the critical heat flux correlation limit with no additional uncertainties.

D3 LPD analyses use the best estimate power distribution and the maximum reactor power, without additional uncertainties, to determine the maximum linear heat rate during the transient.

This linear heat rate is compared to the limiting linear heat rate corresponding to fuel centerline melt or one percent clad strain.

For the Uncontrolled RCCA Withdrawal at Power analysis in the U.S. EPR FSAR Tier 2, Section 15.4.2, the uncertainties applicable to the Low DNBR and High LPD reactor trips are applied as described in the Incore Trip Setpoint and Transient Methodology for U.S. EPR Topical Report, ANP-10287P-000. To provide reasonable assurance that the DNBR is above the safety limit of 1.0 with a 95 percent probability and 95 percent confidence level;>uncertainties described in Section 5 of ANP-10287P-000 are applied. For LPD, the uncertainties described in Section 6 of ANP-10287P-000 are applied to provide reasonable assuraacetfNat the LPD is less than the V/ e*\1 \

LPD corresponding to fuel centerline melt or one perceptclad strain with 95 percent probability and a 95 percent confidence level.

Additional Key Differences Relevant to the Uncon*roll/edBank Withdrawal Event In the U.S. EPR FSAR Tier 2, Section 15.4;2:analysis the moderator is at its maximum value (zero, as shown U.S. EPR FSAR Tier 2, Figure 15.4-7) to minimize the moderator feedback.

The moderator and Doppler coefficients for UJS. EPR-SFAR Tier 2, Section 15.4.2 and D3 analyses are compared in the Respoffse to RAI 1, Question 07.08-23. The Uncontrolled RCCA Withdrawal at Power causes an RCS temperature increase. Because the best estimate modeirator coefficient in the D3aanalysis is negative, the response to the RCS temperature increase is an insertion of negative reactivity-wjjich compensates the reactivity increase due to the RCCA withdrawal as shown in Figure 07.08-28-1.

Another key difference in the mobOdelin f this event is the fact that in the U.S. EPR FSAR Tier 2, Section 15.4.2 an~alysis the reactivity addition due to the RCCA withdrawal was assumed to continue until the reactor trip was initiated while the D3 analysis maintained the reactivity addition magnitude So that the overall reactivity addition from the withdrawal was constrained to the worth of the length of thec6ontrol rods withdrawn from the PDIL.

The negative reactivity contribution of the moderator and the termination of positive reactivity insertion upon full withdrawal of the RCCA contribute to the transient in a way that limits the peak power achieved and to turn power downward prior to the reactor trip, as shown Figure A.3.5-1 of ANP-10304, Revision 1. This reduces the DNBR degradation during the event compared to that of the U.S. EPR FSAR Tier 2, Section 15.4.2 analysis. As a result, the U.S.

EPR FSAR uncontrolled RCCA withdrawal analysis reached a peak power of about 120 percent with a minimum DNBR value normalized to the SAFDL of 1.11 while the D3 analysis reached a peak power of approximately 108 percent with a minimum DNBR value normalized to the SAFDL of 2.0. Both analyses demonstrate adequate margin to the safety limit.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 29 of 56 Figure 07.08-28-1 - RCCA Withdrawal at Power Event - Reactivity 0.40 0.30

/

0.20

-0.10

-:0.200

-E

. D-e~o"*r eff~ect \1

-0.30 *-.4 Mo~tffrator Effect J\

T otal Rea t - A,-

0 100 150 200 250 300 Time (s)

ID:06188 3Nov2009 13:20:03 UCBfWhfp~boc.dmx:2 FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 30 of 56 Question 07.08-30 Explain the assumption that RCSL will respond to the Boron Dilution event as described in Section A.3.5.5 of ANP-10304 Rev 1.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protective function. The Staff Requirements Memorandum to SECY 93-087, Item II.Q, states that the vendor or applicant-shall analyze each postulated common-mode failure for each event and shall demonstrate adequate diversity within the design for each of these events.

Section A.3.5.5 of ANP-10304 Rev 1 describes a Boron Dilution/'e/*vent where RCSL responds to the reactivity transient by automatic insertion of RCCAs, thereby alerting the main control room operator of the dilution event. Sections 4.1,4.12, and A.22/stýte.tat RCSL is assumed not to be available in the D3 analysis. The U.S. EPR, Tier 2, Section 7.7;2.3.10, state that "...RCS boron concentration is calculated in the PS..." and "our redundant limitation signals from the PS are transferred to RCSL." Therefore, upon postulated failure of the\PS, the "transfer" of PS limitation signals to the RCSL would not occur. The\staff/*could not identify sufficient design descriptions that would clearly describe how the RCSL would respond to the Boron Dilution event as described in Section A.3.5*5 of AN-.0304 Rev 1\

Response to 07.08-30 There are two possible plant responsesfor aboron dilution event at power. One possible response is that the reactor control, surveillance nd limitation system (RCSL) is functioning properly and the other is that the RCSL.-ise _itn manual control or is not functioning. Section A.3.5.5 of ANP-1 0304 Revision 1\discussed both situations. Both cases were evaluated to cover the possible responses.,, a software common cause failure (SWCCF) is present in the protection system (,PS,) resulting\ina complete failure of the PS, the transfer of limitation signals to the RCSL will mostlikely not o'ccur and the RCSL will not function. This is the same as when the RCSL is in manual ontrol. Conversely, when the SWCCF in the PS resulted in a partial failure, it is possible for tleRC-SLi/to function normally. The diversity and defense-in-depth evaluation discussed in Sectipn A.3.5.5 of ANP-10304 Revision 1 presents both possibilities to demonstrate that acceptance criteria are met in the event of either scenario.

For the case where RCSL is functioning properly (partial failure of PS), RCSL inserts rod cluster control assemblies (RCCAs) to maintain the reactor coolant system (RCS) average temperature (Tavg) and core power. As the position of the RCCAs approaches the power-dependent insertion limit (PDIL), an alarm will alert the operator that a possible dilution is in progress. Rod movement is blocked so that the PDIL is not exceeded. As in the case for when the RCSL is not functioning or is in manual operation, sufficient time (several hours) is available for the operator to detect and terminate a dilution of the RCS before the shutdown margin is lost.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 31 of 56 Question 07.08-31 Provide an explanation of difference in the DNBR transient between the no-rupture and with-rupture RCCA Ejection cases described in Section A.3.5.6, including:

a. Identification of any differences in analysis assumptions (e.g., initial conditions, reactivity parameters) that affect the transient,

b. A comparison of the key reactor parameters affecting DNBR, e.g., power level, peaking factors, reactor pressure, coolant temperatures, core flow.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical

• to prevent loss of the protective function. TheStaff

/ / Requirements Memorandum to SECY 93-087, Item II.Q, states that the vendobor applicant shall analyze each postulated common-mode failure for each event and shall dc montrate adequate diversity within the design for each of these events.

The applicant reports that the RCCA Ejection event ,`assurming no vessel rupture, does not exceed the DNBR SAFDL, whereas for the rupturecasesthe DNBR decreases below the SAFDL. The staff is not able to identify design descriptioos that would permit sufficient understanding in order to complete the safety' evaluation.

Response to 07.08-31 Item 07.08-31 (a):

The rod ejection event for divety an .defenis.. -depth (03) considered three different scenarios for coolant leakage frozmn "break" sizes identified in Section A.3.5.6 of ANP-10304, ranging from no break,,area-to

,/' Z *\..maximum X \* break area, sized from the control drive flange housing inner diameter. The'threesizesprovide.,a spectrum of possible coolant leakage path sizes if the control drive were t*)Je ejected fro.mn the reactor by the pressure driving head from a flange break. This also allows consideration of the impact of the depressurization of the reactor coolant system (RCS) on'\the departure from nucleate boiling ratio (DNBR) performance.

Each of the events is initialized to the same operating conditions and each "ejects" a control rod with a worth of 65 pcm by adding the equivalent point kinetic worth in dollars (1$ = 1 beta, or delayed neutron fraction) over a timeframe of 0.05 second to simulate the withdrawal from the hot full power (HFP) dependent insertion limit (-50 percent inserted)

The three-dimensional (3D) transient power shapes were determined for the fuel assembly of interest from a rod ejection calculation with the three-dimensional nodal kinetics code NEMO-K using constant inlet thermal hydraulic conditions. This captured the initial power shape redistribution in the assembly of interest, following the methods of U.S. EPR Control Rod Ejection Accident Methodology Topical Report, ANP-10286P. The total core power histories were determined from the point kinetics S-RELAP5 model. These accounted for the reactivity feedback effects from the depressurization and heatup of the RCS occurring after the addition of reactivity from the ejected control rod. The inputs to the LYNXT DNBR calculation were a combination of the transient 3D power shapes in the form of peaking factors and the total core

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 32 of 56 power, mass flux, RCS pressure, and inlet temperature in the form of histories normalized to the initial conditions.

Item 07.08-31 (b):

The sequence of events for each break size case is provided in Tables 07.08-31-1,' 07.08-31-2, and 07.08-31-3. The axial power shapes along with their associated radial peaking factors for the peak fuel rod are provided in Figure 07.08-31-1. The transient shapes are shown for key time points of the transient and are a subset of the full data set from the 3D kinetics calculation.

A comparison of the responses for the total core power, the core exit pressure, the core inlet temperature, the minimum departure from nucleate boiling ratio (MDNBR)/specified acceptable fuel design limit (SAFDL), the peak fuel and cladding temperatures, and the peak average enthalpy rise in the fuel are provided in Figures 07.08-31-2 through 07.08-31-6. The information is shown for each case out beyond the time of the MDNBR poinf. The information shown for each case is beyond the time of the minimum DNBR point/,Thh-time ranges presented are from 200 seconds for the no break case to a few seconds aft reactor trip\for the break cases. The information in the figures allows the following observations to be made":,

• The depressurization rate is proportional to the break size (see Figure 07.08-31-3).

• The moderator density reduction is proportional to the\/magnitude of the depressurization and, therefore, so is the magnitude of negZativ.e reactivity. addition.

" Core power reduction is proportional to the magnitudeof negative reactivity addition (see Figure 07.08-31-2). The initial heatup of the cdre inle't temperature is about the same for each of the cases due to the iritial-energy deposition into the coolant during the initial 6 seconds of the event; beyord this point, lowercore powers lead to lower increases in the inlet temperature (see Figurle,\07.087-3173)-

• MDNBR degradation-in-creasesas the rate of depressurization increases. For each case with break, the SA/ "/ Lis ovilated,(see

\ ,*.. Figure 07.08-31-4). The results of fuel failure census, performed in accordance with the methods presented in ANP-10286P, indicate less than 0.3 percent thfe rods enter DNB (one criterion for fuel failure) which is far below the limit of 30 percent fbelbfailures!

" Once the MDNBR has exceeded the SAFDL the peak cladding temperatures rapidly increase but are limited by the reduction in the fuel rod heat fluxes due to the decreasing core power. At no time do the clad temperatures exceed the limit established in Section 2.2 of ANP-10286P (see Figure 07.08-31-5).

" The peak fuel temperatures are influenced primarily by the core power history because the peak linear heat generation rate is occurring at elevations below the location at which MDNBR exceeds the SAFDL limit. This is reflected in the peak average enthalpy rise response provided in Figure 07.08-31-6. At no time do the fuel temperatures exceed the melting point limit established in Section 7.3 of ANP-10286P (see Figure 07.08-31-5).

" The peak fuel enthalpy rise is impacted by the local heat transfer capability as well as the internal heat generation rate. Violation of the MDNBR SAFDL is followed by an enthalpy rise excursion as a result of the degradation of heat removal capability. The reactor trip eventually terminates the local temperature excursions. This occurs for these cases before the MDNBR SAFDL is violated at the elevation where the peak LHGR is located. At no time

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 33 of 56 does the enthalpy rise exceed the limit of 150 cal/gm established in Section 2.2 of ANP-10286P (see Figure 07.08-31-6).

Table 07.08-31 Sequence of Events for Rod Ejection for Case with No Break Table 07.08-31-2 - Sequence of Events for Rod Ejection for Case with 0.025 ft2 Break Event \> Parameter Time (sec)

Peak core power reached 110.7% 0.066 High corepower level delay (PSRTinot acUtive) Trip 455 7.2 MDNBR/SAFDL limit reaced , 1.000 29.0 Low hot leg saturation marginlaeay (PSRT-/not active) Trip 460 29.5 Low PZR pressuredeta4(psýRT\rqt active)' Trip 15 55.8 Low hot leg pressure delay (PlS RT'nIot active) Trip 5 57.8 Minimum MDNBR/SAFDL reached 0.862 69.0 DAS low hot leg pressure delay Trip 88 69.1 DAS RT with delay -/ Trip 900 69.5 DAS turbine trip (TT) with delay Trip 899 70.1 Transient terminated 1281.6

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 34 of 56 Table 07.08-31 Sequence of Events for Rod Ejection for Case with 0.048 ft2 Break Event Parameter Time (sec)

Peak core power reached 110.7% 0.072 High core power level delay (PS RT not active) Trip 455 8.8 MDNBR/SAFDL limit reached 1.000 14.5 Low hot leg saturation margin delay (PS RT not active) Trip 460 15.1 Low PZR pressure delay (PS RT not active) Trip 15 28.2 Low hot leg pressure delay (PS RT not active) Trip 59 28.8 DAS low hot leg pressure delay /,<Trip 88 34.8 Minimum MDNBRISAFDL reached. // 0.864 35.0 DAS RT with delay / ,\\Trip 900 35.2 DAS TT with delay \irip 899

\T,/ 35.8 Safety injection system (SIS) by PS or by DAS /7 Trip"1,058 58.1 Transient terminated /7 ^ _ ,_ 195.1

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 35 of 56 Figure 07.08-31 Peak Fuel Rod Axial Shapes from NEMO-K and Best Estimate Models U.S. EPR D3 Rod Ejection BOC HFP Fuel Rod Power Shapes 170 160 150 140 130 120 110 Static Shape

£100 FdH=1.7594

• O CO 90 - Shape t=0.Os FdH=1.5963 Shape t=6.9s E80 FdH=1.8523

-- BE BOC FdH=1.476 70 60 50 40 30 20 10 0

0 0.5 1 1.5 2 2.5 FQ=FdH*Fz

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 36 of 56 Figure 07.08-31 Fraction of Core Power Response to Rod Ejection from BOC HFP I---mNEMO-K Constant BCs -RKTPOW No break - - RKTPOW_0.025ft^2 break - - - RKTPOW_0.048ft^2 break 1.15 1 1.10

- 1.05 0

4- 1.00 0

Lu 0.95 0.90 0.85 200

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 37 of 56 Figure 07.08-31 Core Exit Pressure and Inlet Temperature Responses to Rod Ejection from BOC HFP

-P No Break P 0.025ft^2 Break - - - P 0.048ft^2 Break

-T--Tin No Break - e-.Tin 0.025ft^2 Break -- -o-- Tin 0.048ft^2 Break 580 578 576 574 JZ.

572 w

570 S L) 568 "

566 O 564 562

-4 560 200

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Response.to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 38 of 56 Figure 07.08-31 MDNBR/SAFDL Response to Rod Ejection from BOC HFP I- No Break - - 0.025ftA2 Break - - - 0.048ftA2 Break - - Limit I 2.50 2.25 2.00 1.75

, 1.50 Cl) 1.25 z

0 2 1.00 0.75 0.50 0.25

when at -2130 psia and FOP=1.018 0.00 200

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 39 of 56 Figure 07.08-31 Peak Fuel and Cladding Temperature Responses to Rod Ejection from BOC HFP

- TFuelMax-NoBrk - - - TFuelMax-025brk - -TFuelMax-048Brk

-TCMadMax-NoBrk - - - TCadMax-025brk - -TCladMax-048Brk 3500 3000 2500 a' 2000 E

~01500 a,

C-

, 1000 (O

500 0

200

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 40 of 56 Figure 07.08-31-6 - Peak Average Enthalpy Rise Responses to Rod Ejection from BOC HFP US EPR D3 Rod Ejection BOC HFP Peak Radial Average Enthalpy Rise

- DeltaPeakAvgEnth-NoBrk - -DeltaPeakAvgEnth_025brk - - - DeltaPeakAvgEnth_048brk 40 35 30 25 20 a)

• 20 10

• 15 S10-0) 200 FSAR Impact:

The U.S. EPR FSAR will not be as a result of this question.

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Response to Request for Additional Inf6rmation No. 413, Supplement 3 U.S. EPR Design Certification Application Page 41 of 56 Question 07.08-35 Identify the credited diverse means to address the loss of the PS initiated safety functions of RT on a high pressurizer level and CVCS isolation on high pressurizer level, as discussed in ANP-10304, Revision 1, Section A.3.6.2," CVCS Malfunction that Increases RCS Inventory," in accordance with the D3 policy stated in SRM to SECY-93-087, Point 3. If the credited diverse means are manual actuations, provide the detailed design descriptions that would address the guidance of Standard Review Plan (SRP) Appendix 18-A, "Crediting Manual Operator Actions in Diversity and Defense-In-Depth (D3) Analyses." Applying the credited diverse means, provide the following additional information:

a. A description of the operator action sequence, starting with recognition of the increasing RCS inventory event and ending with isolation of CVCS*

b. The time within which the operator can accomplish the'rerquired actions to isolate CVCS and terminate the event, and

c. Identification of the procedure or procedure type (e.g., EPGs)that will prescribe the steps to accomplish the required operator action and whether a,\special D3 coping procedure is required.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in componentsdesign and principles of operation, shall be used to the extent practical to prevent loss of the protectiviefun'ction. The Staff Requirements Memorandum to SECY 93-087, ItemJl.Q, statesthat th&Vendor or applicant shall analyze each postulated common-mode failure for-each event and shall demonstrate adequate diversity within the design for each of these events.

For the Increase in RCS Inventory eventwitWSWCCF, as described in Section A.3.6.2 of ANP-10304 Rev 1, the transietyfiddes notterminate on Pressurizer High Level as is the case with the Thr Chiper and iolain Fonro ste anlysis, V ) the pressurizer high level causes a reactor strp a RCS) innto entth Frel analysis, neither the reactor trip nor the CVCS isolation sp3 occurs, and the pres'urizer fills soliud. Although Section A.3.6.2 of ANP-1 0304 Rev 1 states that pressurizer PSRVs are cgabn plpofrelieving water, thus ensuring that the RCS pressure boundary is maintained, sepffreleuir indication and procedures should be available to ensure that the plant operations personnel recognize and terminate the event in a timely manner. The staff is not able to identify design descriptions that would permit sufficient understanding in order to complete the safety evaluation.

Response to 07.08-35 The chemical and volume control system (CVCS) malfunction that increases reactor coolant system (RCS) inventory event results from a spurious actuation, either by a control system or operator action, of the CVCS that adds fluid to the RCS without letdown. At steady-state full power operation, one charging pump of the CVCS is operating and the CVCS control system maintains the pressurizer level through control of the letdown valve position. The CVCS malfunction is assumed to isolate letdown and start the second charging pump. Two charging pumps inject fluid to the RCS from the volume control tank (VCT) without letdown.

In the U.S. EPR FSAR Tier 2, Chapter 15 analysis of this event the pressurizer high level causes a reactor trip and CVCS isolation. Both these features are part of the protection system

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 42 of 56 (PS). For the diversity and defense-in-depth (D3) analysis, with a software common cause failure (SWCCF) in the PS, neither the reactor trip nor the CVCS isolation occurs. The discussion provided in Section A.3.6.2 of ANP-10304, Revision 1 conservatively estimates that it will take 24 minutes for the pressurizer to fill solid ifthe CVCS malfunction continues unabated. 24 minutes is sufficient time for the operator to recognize and terminate the charging flow before the pressurizer overfills. Under this scenario there are several indications/opportunities that would alert the operator that the pressurizer is filling uncontrollably. The operator will first see a deviation alarm when pressurizer level exceeds the normal pressurizer level control band. The VCT tank level will decrease resulting in the automatic start of the boron and reactor water makeup pumps.

This assessment is conservative because systems that would be available for event mitigation are not credited. In the U.S. EPR design, there exists a pressurizer level limitation function, separate from the PS process automation system. This function'is intended to improve plant availability by avoiding reactor trip and other safety functionattu*tions for events that lead to increasing level in the pressurizer. As illustrated in Figuie 07.08-35-1,,the limitation function will isolate charging and pressurizer spray when the pressurizer level reaches 70 percent.

Therefore, the pressurizer level limitation function will termin'ate this event well before filling the pressurizer. This automatic feature will actuate approximately 8.5 minutes after event initiation.

Figure 07.08-35-1 ',-'Pessurize L I Setpoints PZR LEVEL (% range) CONTROL LIMITION FUNCTION PROTECTION 100%

.Normal Spray 95%

Valves Valvesll 80 ISOLATION OF THE cvcs 80% Max2p .. . ............................ CHARGING LINE 75% Maxlp . RT N fPZR spray ISOLATION M

70% iee funcn ..........

.. CVCS charging flow ISOLATION PZR normal & auxiliary spray ISOLATION PZRL + 7% Aia~xfunction / / ADJUST CVCS HP red station to MAX flowrate

<<54.3% ... . ,/IO ng/ Closing of CVCS 34% < PZRL..%............... HP reducing station Amin function PZRL -7% A............................................................. ADJUST CVCS HP red station to MIN flowrate 2nd CVCS pumps actuation 15% Min2 level function CVCS letdown ISOLATION 12% Min3 level function ........... ............. ..........................CVCS letdown ISOLA TION 0% PZR heaters OFF FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 43 of 56 Question 07.08-37 Provide information to justify that manual isolation of th6 main control room as described in Section A.3.9 of ANP-10304 Rev 1, will occur in a timely manner, including:

a. A description of how the need for manual isolation of themain control room is recognized,

b. The time line for manual control room isolation, including recognition of need for isolation and achievement of isolation,

c. Identification of the procedure or procedure type (e.g., EPGs) that will prescribe the steps to accomplish the required operator action and whether a special D3 coping procedure is required, and

• d. A discussion of whether the operator actions required4 <anually isolate the main control room represent a diverse means of protectiyv*;*action to ensure the D3

• radiological analysis acceptance criteria are met.

10 CFR Part 50, Appendix A, GDC 22, requires, inpart that-design techniques, such as functional diversity or diversity in component design~and.priciples of operation, shall be used to the extent practical to prevent loss of the protective function. The Staff Requirements Memorandum to SECY 93-087, Item II.Q, states that thie endor or applicant shall analyze each postulated common-mode failure for each eVent-and-shalŽ'demonstrate adequate diversity within the design for each of these events.

For the Radiological Consecuence s'of'accidents with SWCCF in the PS, Section A.3.9 of ANP-10304 Rev 1 states that DAS 17"',4* dbes

\ not. provide

• ', . automatic V. control room isolation. Analyses performed by the applicant, howeve~r~indicate-!tat manual isolation of the main control room should take place within 30-minutes of an event initiation. The staff is concerned that manual isolation of the main control-room mayrepresent a required D3 coping action and is a considered a vulnerability which\shoulldaddress the guidance contained in the BTP-7-19 acceptance criteria which states, part:

The applicant/licensej~lsbouIld(1) demonstrate that sufficient diversity exists to achieve these goals, (2) identify the vulnerabilitiesdiscovered and the corrective actions taken, or (3) identify the vulnerabilitiesdiscovered andprovide a documented basis thatjustifies taking no action.

Therefore, the staff request additional information to address this vulnerability.

Response to 07.08-37 Isolation of the main control room (MCR) is required following events with radiological consequences. The events of interest include loss-of-coolant accident (LOCA), steam generator tube rupture, main steam line break, reactor coolant pump locked rotor, rod ejection, and fuel handling accident. Under normal conditions, the protection system (PS) would automatically actuate the MCR emergency filtration system upon receipt of a high radiation

. signal in the MCR air intakes or a primary containment isolation signal. In the event of a software common cause failure (SWCCF), automatic isolation of the MCR is assumed not available, but the radiation signal and alarms are still present. In each of these scenarios with a

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 44 of 56 SWCCF in the PS, the MCR high radiation air intake alarms would alert the operator that MCR isolation is required. Since the air intakes to the MCR are close to the release point in each scenario, the alarm is expected to occur relatively early in the event (within approximately 5 minutes). The diversity and defense-in-depth (03) analysis described in Section A.3.9 of ANP-10304, Revision 1 relies on this alarm function to alert the operator so that the MCR is isolated within 30 minutes.

Emergency response procedures for the U.S. EPR are not developed as a part of Design Certification. It is, however, anticipated that either abnormal operating procedures or emergency operating procedures will include instructions in response to high radiation at the MCR intakes. These instructions will either direct the operator to confirm MCR filtration system actuation or provide for manual isolation. A special D3 coping procedure is not anticipated.

FSAR Impact:

The U.S. EPR FSAR will not be changed

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 45 of 56 Question 07.08-40 For Section A.2.5 of ANP-10304 Rev 1, describe how the models for pressurizer pressure and level control are validated to assure that they accurately describe the plant response.

10 CFR Part 50, Appendix A, GDC 22, requires, in part, that design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protective function. The Staff Requirements Memorandum to SECY 93-087, Item II.Q, states that the vendor or applicant shall analyze each postulated common-mode failure for each event and shall demonstrate adequate diversity within the design for each of these events.

Section A.2.5 of ANP-10304 Rev 1 states that the pressurizer pressure and level control systems are included in the S-RELAP5 best estimate non-LOCA model. The staff is not able to identify design descriptions that would permit sufficient unders anding in order to complete the safety evaluation.

Response to 07.08-40 The S-RELAP5 models for the best estimate analyses feature a 10-node pressurizer component that utilizes a standard collapsed liquid levelScheme foirlevel measurement identical to that used in the U.S. EPR FSAR Tier 2, Chapte-,5eigna'ýimodels. The reactor coolant system (RCS) primary nodalization for non loss-of-ceolaibaccident (non-LOCA) analyses is shown in Figure 07.08-40-1, and theRGS no iali/ation for small-break LOCA (SBLOCA) analyses is shown in Figure 07.*0840-2, )The pressurizer (PZR) components are almost the same as the non-LOCA model'except that the normal pressurizer spray and the pressurizer safety relief valves (PSRVs) arehot in'cludedjthe SBLOCA model because those subsystems are not usually challenged-okrequired for SBLOCA events. The PSRVs in the best estimate non-LOCA model are idýnticateto thbs\from the U.S. EPR FSAR Tier 2, Chapter 15 design basis models except'that nominalopening and closing setpoints are used instead of nominal plus instrument uncer'ainty.

Pressurizer Level Indication 'a Control Table 07.08-24-1 in the Response to RAI 413, Question 07.08-24-1 shows the PZR level tap locations relative to the bottom of the PZR vessel.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 46 of 56 PZR level control is accomplished by the chemical and volume control system (CVCS). The S-RELAP5 CVCS model is described in the Response to RAI 413, Question 07.08-39. The nominal control setpoint at 100 percent rated power is 54.3 percent. Table 07.08-40-3 shows the level and pressure control setpoints that are implemented,-

Figure 07.08-40-3 and Figure 07.08-40-4 show the CVCS logic thaltýi\implemented in the S-RELAP5 models to control PZR level during transient~sFigure 07.08-40-5 shows the pressurizer level response to three representative transients:

1. Loss of all four reactor coolant pumps (RCPs).

2. 1006 pcm rod drop at end of cycle (EOQ).

3. Rod ejection accident with pressure boundaF" bre6k" Figure rate of 07.08-40-6 shows the change of pressure corresponding fort. / :I PZR\pressure

• response for the same events. The rothse typical events is not extreme enough for concern about the decalibration effects associat~ld '" -"wi ...

dc ithe-daffieroeenttal pressure level taps in the as-built pressurizer. There is no transient primary pressure profile for a diversity and defense-in-depth (3) event that would, inalidat ,theý'ecllapsed liquid level scheme used in the S-RELAP5 model.

Pressurizer Pressur ndication and Control The parameters presented in Table 07.08-40-3 apply to the operation of the PID controller used in the RCS pressure control function.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 47 of 56 Table 07.08-40-4 and Table 07.08-40-5 correspond to the function blocks in Figure 07.08-40-7 that are used to determine spray valve position. A linear interpolation is used to find values intermediate to those provided.

The desired flow (Ib/s) is the flow through the normal spray valves, in addition to the constant bypass spray flow. Flow is based on constant volumetric flowrate at cold leg temperature =

562.5'F and pressurizer pressure = 2250 psia.

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 48 of 56 Table 07.08-40-1 - Pressurizer Level Tap Elevations Location Elevation Comments Lower level tap position 27.7559 in. Measured relative to bottom of PZR Upper level tap position 461.024 in. Measured relative to bottom of PZR Table 07.08404- Pressurizer Control Trips S-RELAP5 Trip(s)

Signal Action(s) Setpoint Specifying

___ Setpoint Terminate PZR auxiliary A " \T High-high PZR level 'v 70%o Trip 379 spray and RCS charging 7*

• 7%'/ above*.

High PZR level Maximize RCS letdown .crtrol setpoint Trip 474 If pressurizer normal spr~ay'is demanded but is unavailable High PZR pressure dm e bt i ,, a 2349.7 psia Trip 383 or insufficient, actuate-PZZR auxiliary spray \.V Low-low-low PZR Deenergize all PZR heafe12---3 rods \i3 1 level Low-low PZR level Termirnatie letdown \ \ 15% Trip 476 Low PZR level Miniimiize 1etdo'w(n and'start 7% below. Trip 472 standby:?ch!arging pump control setpoint Table 07.08-40"3 - Pressurizer Pressure PID Controller Constants Parameter Value Gain, KP [ ]

Integral action time, Tr [ ]

Derivative action time, Td [ ]

Lag time of derivative action, T, [ ]

C-OUTP upper limit, UL [ ]

C-OUTP lower limit, LL [ ]

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 49 of 56 Table 07.08-40-4 - 102F or 103F = f (C - OUTP)

C-OUTP Value [%] Desired Spray Flow [Ibis]

Table 07.08-40 102Por*1*03P f(F)

Desired Spray Flow [Ibis] ý Desired Valve Position [% open]

-10 \.\ 0 0 ~ .0

.3.3 // ) 30 6.6 40 11.9 50 21 .4K ",/ 60 66 .1/ 100 88:2\ 100

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 50 of 56

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 51 of 56 Figure 07.08-40-2 - SBLOCA RCS and Pressurizer Nodalization

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 52 of 56 Figure 07.08-40-3 - CVCS Control Logic

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Response to Request for Additional Information No. 413, Supplement 3 U.S, EPR Design Certification Application Page 53 of 56 Figure 07.08-40-4 - CVCS Flow Control

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 54 of 56 Figure 07.08-40-5 - Pressurizer Level Response for Various Events 60 0

0 10 20 30 40 50 Time (s)

ID:02279 3Nov2009 12:20:11 clocfboc.dmx:l

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 55 of 56 Figure 07.08-40-6 - Pressurizer Pressure Response for Various Events 2500 , ,

2400 2300 co r- 2200 0ZE2100 "

C 2000.*

a.

gn1800 1700<-

1600 Loss of 4 RCP

---

RD eoc 1006 pcm

- REA with full Break 1500 I I 1 1 0 10 20 30 40 50 Time (s)

ID:02279 3Nov2009 12:20:11 clocf boc.dmx:l

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Response to Request for Additional Information No. 413, Supplement 3 U.S. EPR Design Certification Application Page 56 of 56 Figure 07.08-40-7 - Mode I RCS Pressure Control FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.