NOC-AE-16003366, First Set of Responses to April 11, 2016 Requests for Additional Information Regarding STP Risk-Informed GSl-191 Licensing Application

First Set of Responses to April 11, 2016 Requests for Additional Information Regarding STP Risk-Informed GSl-191 Licensing Application

ML16154A117
Person / Time
Site:	South Texas
Issue date:	05/11/2016
From:	Connolly J South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16154A127	List: NOC-AE-16003366, First Set of Responses to April 11, 2016 Requests for Additional Information Regarding STP Risk-Informed GSl-191 Licensing Application
References
GSI-191, NOC-AE-16003366, TAC MF2400, TAC MF2401
Download: ML16154A117 (66)

Text

Contents of Attachment 1 should be withheld from public disclosure in accordance with the

-requirements of 10CFR2.390.

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References :

[1] ALION-CAL-STP-8511 -08 , Revision 3 - Risk-Informed GSl-191 Debris Transport Calculation. June 2014.

[2] NUREG/CR-6369 , Volume 2, Drywell Debris Transport Study: Experimental Work.

September 1999.

NOC-AE-16003366 Attachment 2 Page 6 of 8 SSIB-3-3 The NRC staff needs additional information to verify the RoverD computations. For all debris types and sizes , please provide a summary table of the fractions of:

a) all debris transported to the recirculation pool and the strainer, b) the debris retained in structures, and c) the debris settled in the pool.

STP Response:

The limiting case applied to the Rovero analysis was that of the Steam Generator Compartment Case. Debris transport fractions were extracted from the STP transport calculation for that case [1] and are presented in Tables 1 and 2.

Table 1: Transport Fraction Totals: Retained on Structures and To Recirculation Pool Debris Tran;Sport Fraction Debris Tra nsport Fraction Debris Type Debris Size (Retained in Structures) (To Recirculation Pool)

Fines . 0.00% 100.0if.tb LDFG Small Pieces (<6") 36.48%/36.12%

• 0.36% * /5352% ""/63.88% '**"'

(NU KON & TempMat) Large Pieces (>6") 100%/99%

• 1%* /1%***

Intact Pieces 100.00% 0.00%

Microtherm Fines 0.00% 100.00%

Qualified Coatings Fin es Fin es 0.00% 100.00%

U noualified Coatiniu; Fi nes Fin es 0.00% 100.00%

Crud Fines 0.00% 100.00%

Dirt/Dust Fin es 0.00% 100.00%

Latent Fiber Fin es 0.00% 100.00%

• Pre-Er osion/Post-E ro.sion

• As Fin es

• • A5Sma lls

• • " Total Fractio n The two transport fraction values in each row of Table 1 add to unity. Where present the Post-Erosion and Total Fraction values must be taken . This reflects that all of the debris either is held up on structures above the pool or eventually transports to the pool. The small and large piece debris sizes of LDFG that were held up on structures were subject to erosion by containment sprays and produced fines in the process. This percentage is shown as a mass percentage of the total respective source debris size , i.e. 0.36% of the total small piece LDFG debris erodes to fines after being held up on structures, however this debris still reaches the pool and the table is constructed to indicate that.

NOC-AE-16003366 Attachment 2 Page 7 of 8 It is important to note that Rovero computations did not transport any of the small size LOFG despite the "To Strainer" transport fraction based on observations from the July 2008 STP strainer flume test [2]. The sample of small piece LOFG which arrived at the recirculation pool in regions where transport was initially deemed appropriate , valued at 39 .69%, was instead treated as settled out of recirculation pool. In this scenario the debris may be subject to erosion however the computations did not account for these additionally generated fines .

Table 2: T r ansport F raction Totals: To Inactive Cavities, Settled Out of Pool, and To Strainers Debris Transport Fraction Debris Transport Fraction Debris Transport Fraction Debris Type Debris Size

[To Inactive Pool cavities} (Settled Out of Recirculation Pool} (To Strainers}

Fines 1.50% 0.00% 98.50%

LDFG Small Pieces 1<6"1 0.0% 23.83%/22.16%. 2.03% * /39.69% ** /41.72% ***

(NU KON & TempMat) larl!:e Pieces (>6") 0.00% 0.00% 1%.

Intact Pieces 0.00% 0.00% 0.00%

Microtherm Fines 1.50% 0.00% 98.50%

Qualified Coatines Fines Fines 1.50% 0.00% 98.50%

Unqualified Coatings Fines Fines 0.00% 0.00% 100.00%

Crud Fines 1.50% 0.00% 98.50%

Dirt/Dust Fines 5.00% 0 .00% 95.00%

Latent Fiber Fines 5.00% 0.00% 95.00%

• Pre-Erosion/Post-Erosion

• As Fines

• • As Smalls

• • • Total Fraction Marinite debris was also treated as 100% transported in Rovero calculations. This debris type was removed from the plant but since it was used in the 2008 head loss testing it is included as part of this analysis for supportive particulate margin calculations to account for deficiencies in other debris amounts.

The qualified coatings fines consists of all debris types associated with qualified coatings, i.e. epoxy, inorganic zinc, polyam ide primer. While the unqualified coatings fines consisted of those of unqualified systems, i.e. inorganic zinc, alkyds , baked enamel , epoxy. Transport fractions for miscellaneous debris such as equipment labels ,

tags , plastic signs and ty-wraps are not provided as th is debris was assumed to arrive at containment floor but not transport to the strainer. The non-transportability of miscellaneous debris at STP was based on the February 2008 STP strainer testing [3]

where miscellaneous debris was observed not to transport when exposed to STP bound ing flu id flow conditions .

While not specifically requested , Table 2 includes the transport fraction due to inactive pool cavities in addition to fractions that settle out of recirculation and reach th e strainers. The three transport fraction values in each row of Table 2 add to the "To Recirculation Pool" transport fraction . Where present the Post-Erosion and Total Fraction values must be taken . This reflects that all debris that reaches the recirculation pool either ends up in an inactive cavity, settling out of recirculation or at the strainers.

NOC-AE-16003366 Attachment 2 Page 8 of 8

References:

[1] ALION-CAL-STP-8511-08, Revision 3 - Risk-Informed GSl-191 Debris Transport Calculation. June 2014.

[2] AREVA NP Document No. 66-9074541-000 - South Texas Project Test Report for ECCS Strainer Performance Testing, February 2008.

[3] AREVA NP Document No. 66-9088089-000 - South Texas Project Test Report for ECCS Strainer Testing, August 2008.

NOC-AE-16003366 Attachment 3 Attachment 3 Response to DORL-3-1

NOC-AE-16003366 Attachment 3 Page 1 of 2 DORL-3-1 Please provide a specific list of all licensing basis changes, in the application, for which you are requesting NRG review and approval via Section 50. 90 of Title 10 of the Code of Federal Regulations (10 CFR).

STP Response:

Attachment 3 to the August 20, 2015 supplement (ML15246A126 - cover letter, A 129 -

attachment) identifies the changes that require NRG approval per 10CFR50.90. The list below includes those items. The list provides additional detail on the basis for the need for NRG approval.

1. A change in methodology is proposed to allow risk-informed methods to demonstrate satisfactory ECCS sump/strainer performance for certain large breaks instead of the deterministic methods and assumptions described in the current licensing basis. Application of the risk-informed methodology also requires exemption to 10CFR50.46(a)(1) "other properties" and GOG 35, 38 & 41 as described in Attachment 2 to the August 20, 2015 supplement (ML15246A129) and STPNOC letter dated April 13, 2016 (ML16111B204) revising the exemption to 10CFR50.46. NRG approval of this methodology change is required by 10CFR50.59(c)(1)(vii) and (viii). Pc;iragraph (vii) applies because the risk-informed method assumes that breaks with debris not bounded by testing result in core damage; i.e., fission product barrier exceeds a design basis limit, and Paragraph (viii) applies because the risk-informed methodology used to show the risk of core damage and large early release is acceptably small replaces the deterministic method used in the current licensing basis.

This methodology change includes the use of RELAP5-30 as a screening evaluation for in-core effects of debris from the deterministic scope of breaks.

The Rovero risk evaluation is based on all failures being at the strainer (i.e., no in-core risk contribution from smaller breaks). RELAP5-30 is applied to the deterministic scope of breaks that are bounded by the plant-specific testing to confirm there are no smaller (i.e., higher frequency) breaks that cause failure due to in-core effects. The analysis uses the 800°F LTCC PCT based on the WCAP 16793 methodology as its success criteria.

2. The proposed change in methodology for the use of RELAP5-30 is also credited to show satisfactory long-term cooling for the deterministic scope of large HLBs and small CLBs. RELAP5-30 is not required for deterministic scope of large CLBs because the Rovero evaluation shows that there are no debris effects for these breaks. The thermal hydraulic analysis is described in the Rovero discussion of Attachment 1-3 to the August 20, 2015 LAR supplement. The current licensing basis is described in UFSAR Chapter 15.6. The STP current licensing basis thermal-hydraulic analysis is applied only through the reflood

N OC-AE-16003366 Attachment 3 Page 2 of 2 stage to show that the core cooling meets the 10CFR50.46 acceptance criteria such that there are no conditions that would prevent long-term cooling; it does not assess potential effect of in-core debris. The RELAP5-3D method conservatively assumes blockage of certain cooling flow paths by debris during recirculation and long term cooling phases in HLB and small CLB to demonstrate that PCT is acceptable. In addition to flow path blockages, STP conservatively assumes 800°F is the PCT success criterion. NRC approval of this methodology change is required by 10CFR50.59(c)(1)(viii) because the use of RELAP5-3D in this application has not been previously reviewed and approved. However, qpplication of the proposed 800°F PCT has been found acceptable in the NRC review of WCAP 16793. The calculations performed in this methodology change complement deterministic testing by showing that the core damage risk quantification* can be represented by failures from debris effects at the strainer.

There is no contribution from in-core debris effects so only debris quantities that accumulate at the strainer need to be considered.

3. A change to the STP TS for ECCS and C$S incorporating a debris-specific action and required completion time is proposed. All changes to the TS require NRC review and approval per 10CFR50.90.

N OC-AE-16003366 Attachment 4 Attachment 4 Response to ENPB-3-1 through 3

NOC-AE-16003366 Attachment 4 Page 1 of 6 EPNB-3-1 In Attachment 1-2, page 4, of the August 20, 2015, submittal, the licensee stated that the large main steam and feedwater line breaks were not evaluated because recirculation is not required under the plant licensing basis for STP.

a) Please explain why sump recirculation is not required for main steam and feedwater line breaks inside the containment.

b) Please discuss whether any other American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class 2 piping inside the containment, besides the main steam and feedwater lines, are evaluated for debris generation and sump recirculation. If none, explain why ASME Code Class 2 piping inside the containment are not evaluated.

STP Response a) The following will discuss why sump recirculation is not required for main steam line breaks (SLBs) and feedwater line breaks (FLBs) inside the containment. The purpose of sump recirculation is to supply water to the SIS and CSS from the containment sump when the water supply in the Refueling Water Storage Tank (RWST) is depleted. The following discusses the impact of sump recirculation on the SIS and CSS.

Safety Injection System A large main steam or feedwater line break results in a decrease in RCS temperature and subsequent reduction in RCS pressure and water level. The SIS restores the RCS water level and pressurizes the RCS to a pressure no less than the shut-off head of the HHS!, which is approximately 1700 psig. In addition, the SIS provides boron to the RCS which terminates a potential return to power. The restoration of RCS water level, pressurization of the RCS and termination of a potential return to power occurs prior to the depletion of the water in the RWST. In the event safety injection flow were secured, the RCS pressure would be maintained by the pressurizer heaters or at the hot leg saturation pressure if the pressurizer heaters are not available. Since a SLB or FLB does not result in a loss of RCS water inventory, inventory addition from the safety injection system is not required using sump recirculation. Therefore, sump recirculation is not required to support the design function of the safety injection system for a SLB or FLB.

Containment Spray System The purpose of the CSS is to reduce the containment pressure, temperature and minimize the release of radioisotopes (primarily iodines) to the environment.

Containment peak pressure is not a concern for secondary side breaks because they are bounded by the large break LOCA event. Containment peak temperature is bounded by the SLB event. During a SLB event, superheated steam is released into containment which rapidly increases the containment atmospheric temperature. The containment atmospheric temperature quickly decreases to saturation conditions upon the initiation of the CSS. The potential for containment temperature returning to superheated conditions ends when the water inventory in the faulted steam generator is depleted. For SLBs that challenge containment pressure and temperature limits, such as the DEGB SLB, the

NOC-AE-16003366 Attachment 4 Page 2 of 6 water inventory in the faulted steam generator is depleted prior to sump recirculation Therefore, sump recirculation is not required to support the design function of the CSS for the SLB with regard to containment temperature.

During a F_LB, two-phase flow from the break occurs and is therefore bounded by the SLB for the purposes of containment pressure and temperature response.

The containment spray system is not required for radiological effects of secondary side breaks inside containment because no concurrent RCS break is postulated.

For a SLB or FLB, energy is removed from containment by the RCFCs, which is independent of sump recirculation. The CSS does not have a heat exchanger, and does not remove energy from containment. Therefore, the CSS is not required using sump recirculation to ensure the containment pressure and temperature* limits are not exceeded.

Conclusion To summarize, sump recirculation is not required to support the design functions of the SIS or CSS for a SLB or FLB.

For completeness, STPNOC evaluated the risk contribution for SLBs and FLBs for beyond-design-basis failures that would require ECCS recirculation in a "feed and bleed" function. STPNOC assumed these conditions result in core damage. The risk J

contribution does not affect the STPNOC conclusions.

NOC-AE-16003366 Attachment 4 Page 3 of 6 b) All Class 1 welds were evaluated for breaks causing debris generation in the Rovero analysis with the exception of welds beyond a check valve or other isolation valve ir.i the Class 1 piping. In Table 17 starting on Page 57 of Attachment 1-3 of the August 20, 2015 submittal is a list of these welds which are over 600 in number.

The ASME Code Class 2 piping inside containment is not evaluated for breaks that cause debris generation. The Main Steam and Main Feedwater lines are excluded as discussed above. Other Class 2 piping such as Containment Spray and Safety Injection are excluded since the lines are pressurized only for accident mitigation and are not considered as the initiating event of an accident which causes debris that _might challenge the use of the emergency sump in the recirculation mode. Class 2 pipe lines such as charging and letdown of the Chemical and Volume Control System and also Steam Generator Slowdown could be pressurized during normal operation. However, mitigation of breaks in these lines does not require the use of the emergency sump in the recirculation mode. The Residual Heat Removal System has Class 2 lines which are pressurized during the cooldown and startup of the reactor. Again, mitigation-of a break in the Residual Heat Removal System piping does not require the use of the emergency sump in the recirculation mode. Thus for the reasons discussed above, ASME Class 2 lines are not evaluated for: breaks causing debris generation.

NOC-AE-16003366 Attachment 4 Page 4 of 6 EPNB-3-2 In Attachment 1-4, page 22, of the August 20, 2015, submittal, under the heading, Reactor Coolant System Weld Mitigation, the licensee stated that " ... All STP large bore RCS [reactor coolant system] welds susceptible to pressurized water stress corrosion cracking (PWSCC) have been replaced with Alloy 690 material which is not susceptible to PWSCC (SG [steam generator] nozzles) or overlaid with non-susceptible Alloy 52/52M/152 material (pressurizer piping safe ends) with the exception of the reactor vessel nozzle welds ... "

a) Please clarify w~ether "the reactor vessel nozzle welds" discussed in the above statement are the J-groove welds associated with the reactor vessel closure head penetration nozzles to house the control rod drive mechanisms (i.e., control rod drive mechanism (CROM) nozzles), or the full-penetration butt welds associated with the hot-leg nozzles that are attached to the reactor vessel shell.

b) Please discuss of what material the CROM nozzles and the associated J-groove welds in both units are made. If the nozzles are composed of Alloy 600 material and the welds are Alloy 82/182, then discuss why these were not selected as break locations since these materials are susceptible to PWSCC.

c) Please identify: 1) the large bore RCS piping (e.g., hot leg, cold leg, or crossover piping) and other ASME Code Class 1 pipes (e.g., pressurizer surge line, pressurizer spray line, or safety injection piping) that contain either Alloy 690 weld material or are mitigated with Alloy 52/52M/152 material, and that are considered in the GSl-191 evaluation; and (2) all ASME Class 1 piping that is larger than 2 inches that contain Alloy 82/182 weld material, has not been mitigated with Alloy 52/52M/152 material, and are considered in the GSl-191 evaluation.

STP Response:

a. The reactor vessel nozzle welds excepted in the referenced statement are the full penetration butt welds on the hot and cold leg RPV shell. These dissimilar metal welds have not yet been mitigated but have been approved for mitigation by non-welded stress improvement process known as Mechanical Stress Improvement Process (MSIP) - STP Unit 1 spring 2017 (1 RE20) and STP Unit 2 fall 2019 (2RE20). MSIP compresses the pipe adjacent to the weld and creates compressive stresses through -50% of the pipe wall thickness. Eliminating the tensile stresses at the inside surface of the PWSCC susceptible weld removes one of the three conditions required to be present for PWSCC to initiate.

The CROM J-groove weld are addressed as described in the response to part (b) of this RAI.

b. STP replacement RV Closure Head CROM nozzles and associated butt welds (not J-groove welds) are made of Alloy 690/52/152 material and are not considered susceptible to PWSCC. STP RV Closure Heads were replaced in both units (2009 for Unit 1 and 2010 for Unit 2). Due to the relatively small size and location of the CROM nozzles, breaks do not need to be considered since they are bounded by larger breaks that produce much more debris (Ref. NEI 04-07).

NOC-AE-16003366 Attachment 4 Page 5 of 6

c. STP primary loop piping and branch connection pipes are constructed of stainless steel (e.g. SA-351 Gr CF8A for loop piping and SA-312 for branch piping) and is not susceptible to PWSCC.

The RCS piping welds considered for GSl-191 which have been mitigated with material not susceptible to PWSCC are:

• SG nozzles

• Pressurizer nozzles (overlays)

• RPV head CRDM penetrations (as discussed in the response to part (b) of this RAI)

The RCS piping welds considered for GSl-191 which are susceptible to PWSCC which have not been mitigated are: "'

• RPV hot and cold leg nozzle weld as discussed in the response to part (a) of this RAI.

NOC-AE-16003366 Attachment 4 Page 6 of6 EPNB-3-3 Please explain why breaks from the pressurizer heater sleeves and reactor vessel bottom-mounted instrumentation nozzles were not considered as a source of debris generation.

STP Response:

Due to their relatively small size and location, pressurizer heater sleeves and reactor vessel bottom mounted instrumentation nozzles were not considered as a source of debris generation. Any breaks for these items are bounded by larger breaks in the nearest piping which would produce much more debris (Ref. NEI 04-07). Also the only type of insulation debris from a break of one of these items would be from reflective metal insulation which is not considered to transport onto the emergency sump strainers and thus would not be an impact.

)

NOC-AE-16003366 Attachment 5 Attachment 5 Response to ESGB-3-1

NOC-AE-16003366 Attachment 5 Page 1of2 ESGB-3-1 Provide additional details and clarification with respect to the manner in which the total mass of unqualified coatings was calculated. Attachment 1-2, page 67 of 95 of the August 20, 2015, submittal states, in part, that "the weight of applied coatings are determined based on a theoretical coating spread rates (sq. ft per gallon @ 1 mil thick) instead of specific vendor coating spread rates." If a 1 mil (thousandths of an inch) thick coating was assumed for IOZ or epoxy coatings, the analysis may be significantly underestimating the amount of coating debris.

Please describe the thickness used in the analysis for both epoxy and IOZ coatings since the mass of epoxy within the ZOI may be impacted and the mass of unqualified IOZ throughout containment may be impacted.

STP Response:

The theoretical coating spread rate referenced is simply a conversion factor for calculating a coating's area of coverage per gallon (ft2/gal) for a given dry film thickness (DFT). The relationship is defined by the following equation:

Eqn 1 where:

c = Coverage (ft2/gal)

R = Spread Rate (ft2/gal/mil)

DFT = Dry film thickness (mils)

The use of this parameter does not imply that a coatings thickness of 1 mil is assumed.

As an example, consider the following coatings system (these values do not represent a specific STP system, but are typical of what might be used in an unqualified coatings calculation):

Area: 10,000 ft2 DFT: 5 mils Weight/Gallon: 12 lbs/gal

% Solids (by Volume): 50%

% Solids (by Weight): 75%

Spread Rate: 1600 ft2/gal/mil To calculate the dry weight from the given information, the area of coverage per gallon is first calculated using the theoretical spread rate:

NOC-AE-16003366 Attachment 5 Page 2 of 2 2

160oft )

--.~ra_l co.5o)

(

(Spread Rate) (%Solids by Volume) mi ft 2 Eqn 2 Coverage=

. DFT

= 5 mils

= 160- gal The total volume (in gallons) required to cover the area can then be calculated:

Area 10000 ft 2 Gallons Required = Coverage 2 = 62.5 gal f

160 _t_ Eqn 3 gal Dry Weight= (Gallons Required)(Weight per Gallon)(% Soilds by Weight) lbs)

= (62.5 gal) ( 12 gal (0.75) = 562.5 lbs Eqn4 The thicknesses used for unqualified coatings in the STP assessment vary for different substrates and coatings systems. Unqualified epoxy thicknesses ranged from *2 to 22 mils, and the thickness of unqualified IOZ coatings ranged from 2.5 to 6 mils. The dry weight calculations, including the dry weight thicknesses, for every unqualified coatings system considered can all be found in STP's unqualified coatings quantifications (9AC5002#1 and 9AC5002#2).

N OC-AE-16003366 Attachment 6 Attachment 6 Response to SNPB-3-1, -3, -4, -5, -8, -11, -12, -13, -14, -16, -19

NOC-AE-16003366 Attachment 6 Page 1 of 31 Thermal-Hydraulic Review Questions Note 1: the draft SNPB questions sent to STPNOC by e-mail dated October 21, 2015 (ADAMS Accession No. ML16022A177), were subsumed by questions SNPB-3-3 and SNPB-3-20 below.

Note 2: The following SNPB questions are from the criteria set forth in the following two NRG staff guidance documents:

Safety Evaluation for the Westinghouse Topical Report WCAP-16793-NP "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid," Revision 2. This will be identified as "SE for WCAP."

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: L WR Edition," Section 15. 0. 2. The subsection referenced will be provided with the question. This will be identified as "SRP."

SNPB-3-1 Cladding Oxide Please demonstrate that the thickness of the cladding oxide and the deposits of material on the fuel do not exceed 0.050 inches in any fuel region.

Criterion 0.1

Reference:

SE for WCAP STP Response:

Starting on page 74 of 77 of ML083520326, STP RAI #31 and 36. The deposit layer thickness for STP fuel is 13.64 mils.

NOC-AE-16003366 Attachment 6 Page 2 of 31 SNPB-3-3 Clarification on Core Bypass Blockage During the audit, STPNOC considered performing the L TCC analysis with the core bypass open to allow flow in the axial direction. If STPNOC credits the use of the bypass, it should provide test data to demonstrate that the bypass will not block during the scenarios. This test data should bound the flow rates, flow areas, and debris loading expected in the RCS.

Criterion 1.3

Reference:

SRP, lll.3c STP Response:

Flow path analysis basis/background There are six flow paths that need to be taken into account to accurately model the Barrel-Baffle (BB) flow during blockage:

1. Thimble Tube {TT) flow

2. Core former-to-fuel gap flows

3. LOCA hole flows

4. BB flows

. 5. Cold leg to hot leg leakage flow (direct bypass)

6. Upper Head Spray Nozzle (UHSN) flow Of these 6 flows, only the TT and UHSN have been well characterized at STP in specific studies for upper head cooling and removal of thimble. plugs. The other flow paths, (2, 3, 4, and 5) are not characterized except to ensure that the total meets either design (non-conservative for GSl-191), or best estimate.

Assuming minimum flow through the open BB flow path is a conservative approach since it ~educes the amount of flow available to the top of core with the only remaining path being the UHSNs. Since the UHSN path requires cooling water to first go the upper head and then flow down to the top of the core, the water available to the top of the core is delayed and minimized resulting in flow to the top of the core being conservatively~.

estimated.

NOC-AE-16003366 Attachment 6 Page 3 of 31 Phenomenology basis/background STP will assume the largest HLB that is assumed to fail the strainer is 16". While this assumption increases the risk related to the concerns raised in GSl-191, it bounds locations where fine fiber amounts exceed the amounts tested based on an assumed 170 ZOI for unjacketed NUKON.

By adopting the smaller maximum break size, uncertainty associated with complications at larger (up to DEGB) break behavior are subsumed in the assumed failures above 16";

any thermal-hydraulic uncertainties associated with more violent behavior realized at larger break sizes is bounded.

NOC-AE-16003366 Attachment 6 Page 4 of 31 SNPB-3-4 Describe Important Phenomena Please provide a description of the important phenomena being modeled in the L TCC EM for each of the accident scenarios being simulated. These phenomena should include those important to obtaining the correct initial conditions for the long-term phase, and the important phenomena during the long-term phase.

Criterion 1.3

Reference:

SRP, lll.3c STP Response:

HOT LEG BREAK LOCA The important phenomena in a HLB are grouped based on the accident phases:

Slowdown, refill, reflood, long-term core cooling, and core blockage. As break size decreases from large to small, the phases become less distinct, and some phenomena described below become unimportant. For example, in smaller break sizes, the ECCS flow rapidly overcomes the volumetric flow out the break; the RCS and .core are repressurized and flooded.

Slowdown.

The blowdown period is the result of a hypothesized break in the coolant system through which the primary coolant is expelled. Slowdown physical processes include choked flow at the break, fluid flashing and depressurization, and heating of the fuel rods due to degraded heat transfer. During the blowdown phase, some components are affected more than others. In particular, the heat generated in the core may not be adequately removed due to decreased heat transfer resulting from the loss of fluid. The performance of the reactor coolant pumps will degrade as the coolant flashes. The steam generator heat transfer degrades because primary system temperatures are lower than secondary system temperatures during this period. In small break LOCA, heat removal from the steam generators is possible early due to rapid refilling of the RCS.

Important phenomena and processes expected to occur during the blowdown phase are:

• Choked flow at the break. Choked flow is expected to occur due to the pressure ratio between the RCS and RCB being sufficiently high. Both subcooled (initially) and saturated (later) choked flows are expected.

• Flashing and voiding. The volumetric flow through the break greatly exceeds the makeup flow, producing flashing and voiding in all volumes. The primary coolant, initially subcooled, reaches saturation temperature of the RCS water.

• Depressurization. As a result of the volumetric flow difference between break flow and makeup flow, the primary system pressure decreases.

• Release of stored energy and decay heat in the fuel. Unlike the large break in cold leg, the flow through the core during a large hot leg break is expected to accelerate instead of stagnating, removing the stored energy and decay heat efficiently in the initial phase of the accident.

N OC-AE-16003366 Attachment 6 Page 5 of 31

• Boiling on the cladding surface. The heat transfer at the cladding surface is initially dominated by single phase forced convection and rapidly (depending on break size) transitions* to boiling-dominated heat transfer as the system depressurizes.

Refill.

In a hot leg break scenario, refill and reflood are not realized as distinct processes.

Instead, because the combined resistance to flow of the pump and steam generator to the break opposite the vessel side is much higher than the vessel side, safety injection water (from accumulators and SI pumps) preferentially flows through the core to the broken hot leg. Because of the overall pressure gradient developed by the hypothesized break at the core outlet, phenomena and processes such as counter current flow limiting (CCFL) and ECC bypass in the downcomer are not of interest in HLB scenarios.

Reflood Because the pressure gradient favors ECCS flow through the core (in hot leg break), the core is continuously supplied with coolant and flooded almost immediately.

Long-term core cooling.

Decay heat re~oval is dominated by forced convection heat transfer. In the long-term phase, colder water from the RWST is injected into the core until the RWST water is depleted. When the RWST is depleted, the ECCS transfers to sump recirculation mode with a higher injection temperature; the vessel remains full of subcooled water up to the hot loop nozzles_ where the water exits back to the reactor containment pool.

Core blockage.

In recirculation phase, debris may accumulate at the bottom of the core and reduce the ECCS flow through the core. Under the limiting assumption of hypothetical full core blockage with ECCS flow to the core inhibited, the temperature of the water in the core would be expected to increase to saturation, and voiding is expected to occur until water from the UHSNs reaches the top of the core. The following phenomena are expected to occur during this phase:

• Forced to natural convection heat transfer. The heat transfer regime in the core suddenly changes from forced single-phase convection to two-phase natural circulation. As voiding increases, the heat transfer coefficient at the cladding surface continues to reduce. If enough liquid coolant arriving from alternative flow paths replaces the water boil off from the core, the heat transfer regime at the cladding wall is expected to be saturated boiling with the cladding temperature I

slightly above the saturation temperature at the RCS pressure.

• Counter current flow limiting (CCFL). Any injected ECCS water that reaches the core under this scenario will arrive from the top of the core. Due to boiling in the core, steam exiting the core may establish conditions favorable for CCFL.

NOC-AE-16003366 Attachment 6 Page 6 of 31 The most relevant phenomena which are expected to occur during a large break are summarized in the table below. The table is based on a Westinghouse 4-loop plant of CSAU study [1].

Table 1. LB LOCA Important Phenomena Phenomena Phenomena description Phase*

Asymmetries A difference in T-H behavior that can be attributed to the 1, 2 aeometricallv asymmetric arrangement of hardware.

Boiling - film Boiling regime in which vapor blankets all or an appreciable 1,2,3,5 portion of the heating surface.

Boiling - transition A boiling regime that spans the boiling surface between 1,2,3,5 critical heat flux and minimum film boilina Condensation - The process whereby steam is cooled due to contact with a 2 interfacial colder liquid, resulting in a change of phase from vapor to liquid at the interface between the two phases.

Entrainment I The process whereby liquid is captured (entrained) by a 2,3,5 deentrainment high-velocity steam flow. The process whereby liquid departs (deentrained) from a steam flow.

Evaporation - The process whereby a fluid changes from the liquid state 1,2,3,5 interfacial to the vapor state by the addition of enerqy.

Flashing - interfacial The process whereby fluid changes from the liquid state to 1 the vapor state due to a reduction in the fluid pressure, which lowers the saturation temperature.

Flow - countercurrent The process whereby liquid flows opposite (counter) to the 2,5 gas flow direction.

Flow - choked The maximum possible flow through a flow constricting item 1,2 of hardware, usually a nozzle,' orifice, or break in a pipe.

Flow- Flow that has two or more dominant velocity vectors. 2,3,5 multidimensional Examples are multidimensional flows in a PWR core durina refloodina.

Heat conductance - The overall thermal resistance to the flow of heat between 1 fuel-clad qap the fuel pellets and claddina in a nuclear fuel rod.

Heat transfer - forced Process of energy transport by the combined action of heat 2 convection to vapor conduction, energy storage, and mixing motion.

Heat transfer - stored The process by which the energy within a solid structure is 1 energy release released to a lower energy state through one or more heat transfer processes, e.g., conduction and convection.

Applies specifically to the transport of the energy residing in fuel rods operating at full power to the coolant following a reactor trip.

The first peak is associated with the blowdown time period and is caused by the initial stored energy in the fuel rods and degraded fuel rod-to-coolant heat transfer.

lnterfacial shear The friction caused by the velocity difference between two 2,3,5 phases at their interface.

Level The vertical heiaht of a column of sinale- or two-phase fluid. 3 Noncondensable The impact of the presence of noncondensable gases upon 3,5 effects heat transfer or any other phenomenon such as flow, condensation, flashina, and vapor volume expansion.

Oscillations The periodic variation of any given hydraulic characteristic 3,5 between two values.

Power-decay heat Heat produced by the decav of radioactive nuclides. 2,3,4,5 RC Pump- The behavior of a pump under all normal and off-normal 1 performance, including conditions.

degradation a: Phase of the LB LOCA sequence: Slowdown =1, Refill =2, Reflood =3, Long-Term = 4, Core Blockage 5. =

NOC-AE-16003366 Attachment 6 Page 7 of 31 SMALL COLD LEG BREAK The important phenomena in a small CLB are grouped based on the accident phases: Slowdown, natural circulation, loop seal clearance, boil-off, core recovery, long-term core cooling, and core blockage.

Slowdown.

On initiation of the break, there is a rapid depressurization of the primary side of the RCS. Reactor trip is initiated on a low pressurizer pressure set point. RCPs are tripped by the operators when the Emergency Operation Procedures conditions are met (low RCS pressure and ECCS flow). A safety injection signal occurs when the primary pressure decreases to the SI injection set point. The RCS remains liquid solid for most of the blowdown period, with phase separation starting to occur in the upper head, upper plenum and hot legs near the end of this period. During the blowdown period, the break flow is single phase liquid only. Eventually, the rapid depressurization ends and the RCS reaches a pressure just above the steam generator secondary side pressure.

Natural circulation.

At the end of the blowdown period, the RCS reaches a quasi-equilibrium condition which can last for several hundred seconds depending on break size. During this period, the loop seals remain plugged and the system drains from the top down with voids beginning to form at the top of the steam generator tubes and continuing to form in the upper head and top of the upper plenum region. Decay heat is removed by the steam generators during this phase. Vapor generated in the core is trapped within the RCS by liquid plugs in the loop seals, and a low quality flow exits the break. This period is referred to as the natural circulation period.

Loop Seal Clearance.

The third period is the loop seal clearance period. When the liquid level in the downhill side (cold leg tube side) of the steam generator is depressed to the elevation of the loop seal, steam previously trapped in the RCS can be vented to the break. The break flow, previously a low quality mixture, transitions to primarily steam. Prior to loop seal venting, the inner vessel mixture level can drop rapidly, resulting in a deep but short core uncovery. Following loop seal venting, the core level recovers to about the cold leg elevation, as pressure imbalances throughout the RCS are relieved.

Boil-off.

Following loop seal venting, the vessel mixture level will decrease. In this period, the decrease is due to the gradual boil-off of the liquid inventory in the reactor vessel.

The mixture level will reach a minimum, in some cases resulting in a deep core uncovery. The boil-off period ends when the collapsed liquid level in the core reaches a minimum. At this time, the RCS has depressurized to the accumulator set point, and the core boil-off rate matches the delivery of safety injection to the vessel.

NOC-AE-16003366 Attachment 6 Page 8 of 31 Core Recovery and Long-term cooling.

The core recovery period extends from the time at which the inner vessel mixture level reaches a minimum in the boil-off period, until all parts of the core quench and are covered by a low quality mixture. In the long-term cooling period, the entire core is quenched and the safety injection flow equalizes the break flow.

Core blockage.

After the RWST is exhausted, ECCS transfers to recirculation mode. Under a hypothetical full core blockage, due to interruption of the ECCS flow, the temperature of the water in the core is expected to increase to saturation, and voiding is expected to occur until water from alternative flow paths reaches the top of the core. The following phenomena are expected to occur during this phase:

• Forced to natural convection heat transfer. The heat transfer regime in the core suddenly changes from forced single-phase convection to two-phase natural circulation. Liquid coolant from alternative flow paths will replace the water boil off from the core. Heat transfer regime at the cladding wall is expected to be saturated boiling with the cladding temperature slightly above the saturation temperature at the RCS pressure.

• Counter current flow limiting (CCFL). Any injected ECCS water that reaches the core under this scenario will arrive from the top of the core. Due to boiling in the core, steam exiting the core may establish conditions favorable for CCFL. Due to the relatively smaller amount to vapor produced in the core and the larger amount of water inventory in the vessel compared to a large break scenarios, the conditions for CCFL may not be met.

I Due to the longer sump switchover time compared to the large break scenarios; a subsequently lower decay heat to be removed from the core; and a larger water inventory in the RCS (the RCS and vessel are expected to be full of water at this time); these phenomena may have less importance than they do in large break scenarios.

The most relevant phenomena which are expected to occur during a small cold leg break are summarized in the table below. The table is based on Westinghouse 4-loop plant PIRT panel [2].

N OC-AE-16003366 Attachment 6 Page 9 of 31 Table 2. SB LOCA Important Phenomena Phenomena Phenomena description Phase*

Condensation - fluid to surface The process whereby steam is cooled due to contact with a 1, 3 colder surface, resulting in a change of phase from vapor to liquid at the surface.

Condensation - interfacial The process whereby steam is cooled due to contact with a 4,5 colder liquid, resulting in a change of phase from vapor to liquid at the interface between the two chases.

Entrainment I deentrainment The process whereby liquid is captured (entrained) by a high- 3 velocity steam flow. The process whereby liquid departs ldeentrainedl from a steam flow.

Flashing - interfacial The process whereby fluid changes from the liquid state to the 3,4,5 vapor state due to a reduction in the fluid pressure, which lowers the saturation temoerature.

Flow regime - break inlet The characteristics of the flow at the break entrance, e.g., All subcooled liauid, saturated, two-ohase, stratified, vaoor, etc.

Flow - countercurrent The process whereby liquid flows opposite (counter) to the gas 2,3, 6 flow direction.

Flow - choked The maximum possible flow through a flow constricting item of All hardware, usuallv a nozzle, orifice, or break in a nine.

Flow-qap Flow through the hot leg to downcomer gap. 3 Heat transfer:-- post-CHF Heat transfer between the two-phase fluid and the heated surface 4,5 in the liquid-deficient region downstream of the CHF point, i.e.,

the location at which the heat transfer condition of the two-phase flow substantiallv deteriorates.

lnterfacial shear The friction caused by the velocity difference between two 3,6 nhases at their interface.

Level The vertical height of a column of single- or two-phase fluid. 3,4,5,6 Oxidation A chemical reaction that increases the oxidation content of a 4,5 material. Of specific interest is cladding oxidation, which occurs at elevated temperatures, which can occur only under accident conditions.

Power - 3D distribution The axial, radial and azimuthal power variation in a core. 4,5 Power - decav heat Heat produced by the decay of radioactive nuclides. All Power - local peaking (fuel rod) The ratio of power at a location (specific fuel rod) to the core 4,5 averaae oower.

Pressure drop The reduction in pressure with distance. 3 Rewet The post-dryout process in which liquid once again resumes 4, 5 intimate contact with a heated surface.

Stratification - horizontal The variation of physical properties such as temperature or 3 density across the vertical cross section of a fluid body having a primarily horizontal orientation, e.g., the cold leg of a nuclear steam supply svstem.

a:.Blowdown = 1, Natural Circulation = 2, Loop Seal Clearance= 3, Boil-off= 4, and Core Recovery and Lonq-Term= 5, Core Blockar:ie = 6.

References:

[1]. Technical Program Group, EG&G Idaho, Inc., Quantifying Reactor Safety Margins:A Application of CSAU to a LB LOCA, USNRC report NUREG/CR-5249, 1989

[2] S. M. Bajorek, A. Ginsberg, D. J. Shimeck, K. Ohkawa, M. Y. Young, L. E.

Hochreiter, P. Griffith, Y. Hassan, T. Fernandez, and D. Speyer, "Small Break Loss of Coolant Accident Phenomena Identification and Ranking Table (PIRT) for Westinghouse Pressurized Water Reactors," Proceedings of the Ninth International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-9), San Francisco, California (October 3-8, 1999)

N OC-AE-16003366 Attachment 6 Page 10 of 31 SNPB-3-5 Debris at Grid Spacers Please describe how the L TCC EM accounts for potential blockages at the spacer grid in the core above the bottom grid.

Criterion 1.3

Reference:

SRP, lll.3c STP Response:

STP has approximately 2 lbs of CRUD total (ML15901A440, Attachment 3 Pages 8 of 10 and 9 of 10) in the RCS that can be transported to the fuel grids. This small amount (CRUD is primarily oxidized metals and therefore dense) of fine particles is not capable of blocking up fuel channels. Fibrous debris collected at the spacer grids will not cause flow blockage or cause inadequate cooling (ML11292A021, Section 4 ).

N OC-AE-16003366 Attachment 6 Page 11 of 31 SNPB-3-8 How are the Phenomena Modeled Please summarize how the important phenomena are being modeled in the L TCC EM. This discussion should provide the phenomena and a summary of how it is being modeled (e.g.,

through the field equations, by an identified closure relationship).

Criterion 2.2

Reference:

SRP, lll.3a STP Response:

The RELAP5-3D code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code contains models to predict the coupled behavior of the reactor coolant system and the core during loss-of-coolant accident scenarios.

The code is based on a non-homogeneous and non-equilibrium model for two-phase systems that is solved by a partially implicit numerical scheme, and includes important first-order effects necessary for accurate prediction of system transients."

RELAP5-3D includes many component models from which normal operation and transients in PWR systems can be simulated. The component models include pumps, valves, pipes, transient conduction-convection heat transfer systems, reactor kinetics, separators, annuli, pressurizers, accumulators, and control system components. In addition, special process models are included for effects such as form loss, flow at an abrupt area change, fluid-wall friction, branching, and choked flow [1].

The RELAP5-3D code manual is organized in five volumes. The content of each volume is briefly described below:

Volume I: Code Structure, System Models and Solution Methods.

This volume presents modeling theory and associated numerical schemes adopted in RELAP5-3D. The volume is divided into eight sections:

• Section 1: Introduction

• Section 2: Code Architecture

" Section 3: Hydrodynamic Model (including field equations, state relationships, constitutive models, special process models)

• Section 4: Heat structure models

• Section 5: Trip system

• Section 6: Control system

• Section 7: Reactor kinetics model

• Section 8: Special techniques Volume II: User's Guide and Input Requirements Detailed instructions for code application and input data preparation are included in this volume.

The volume is organized in eight sections:

• Section 1: Introduction

• Section 2: Hydrodynamics

• Section 3: Heat structures

• Section 4: Trips and controls

• Section 5: Reactor kinetics

N OC-AE-16003366 Attachment 6 Page 12 of 31

• Section 6: General tables and component tables

• Section 7: Initial and boundary conditions

• Section 8: Problem control An appendix to volume II is included containing a detailed explanation of all input cards.

Volume Ill: Developmental Assessment This volume provides the results of developmental assessment cases that demonstrate and verify the models used in the code. The volume is divided into six sections:

• Section 1: Introduction

• Section 2: Developmental assessment matrix

• Section 3: Phenomenological cases

• Section 4: Separate effects cases

• Section 5: Integral effects cases

• Section 6: Summary and conclusions Volume IV: Models and Correlations This volume discusses in detail RELAP5-3D models and correlations. The volume is organized in eleven sections:

• Section 1: Introduction

• Section 2: Field equations

• Section 3: Flow regime maps

• Section 4: Closure relations for the fluid energy equations

• Section 5: Closure relations required by fluid mass conservation equations

• Section 6: Momentum equation closure relations

• Section 7: Flow process models

• Section 8: Special components models

• Section 9: Heat structures process models

• Section 10: Closure relations required by extra mass conservation fields

• Section 11: Steady-state Volume V: User's Guidelines This volume contains guidelines that have evolved over the past several years through the use of the RELAP5-3D code. The volume is organized in five sections:

• Section 1: Introduction

• Section 2: Fundamental practices

• Section 3: General practices

• Section 4: Specific practices

• Section 5: Pressurized water reactor example applications The important phenomena in LB and SB LOCAs are identified based on the accident phase, and summarized in Table 1 and Table 2 of RAl-SNPB-04 response. The following tables describe how the important phenomena for LB and SB LOCAs are modeled in RELAP5-3D.

N OC-AE-16003366 Attachment 6 Page 13 of 31 Table. I. LB LOCA Phenomena and Modeling Phenomena Phenomena description Model RELAP5-3D Volume/Section Asymmetries A difference in T-H behavior that can Field Equation/Basic 1-3.1.1 3.1.4 & IV-2.1 be attributed to the geometrically Equation: All fluid flow asymmetric arrangement of hardware. equations Boiling - film Boiling regime in which vapor blankets Closure (constitutive) 1-3.3.9 & 1-3.3.10 & IV-4.2 all or an appreciable portion of the Relations (including regime heating surface. maps): Wall-to-fluid energy exchange: film boiling Boiling - transition A boiling regime that spans the boiling Closure (constitutive) 1-3.3.9 & 1-3.3.10 & IV-4.2 surface between critical heat flux and Relations (including regime minimum film boiling maps): Wall-to-phase energy exchange: transition boiling Condensation - The process whereby steam is cooled Closure (constitutive) 1-3.3.11 & IV-4.1 interfacial due to contact with a colder liquid, Relations (including regime resulting in a change of phase from maps): lnterfacial energy vapor to liquid at the interface between exchange: all flow regimes the two phases.

Entrainment I The process whereby liquid is captured Closure (constitutive) 1-3.3.6 & IV-6.1 deentrainment (entrained) by a high-velocity steam Relations (including regime flow. The process whereby liquid maps): lnterfacial departs (deentrained) from a steam momentum exchange: all flow. flow regimes.

Evaporation - The process whereby a fluid changes Closure (constitutive) 1-3.3.11 & IV-4.1 interfacial from the liquid state to the vapor state Relations (including regime by the addition of energy. maps): lnterfacial energy exchange: all flow regimes.

Flashing - interfacial The process whereby fluid changes Closure (constitutive) 1-3.3.11 & IV-4.1 from the liquid state to the vapor state Relations (including regime due to a reduction in the fluid pressure, maps): lnterfacial energy which lowers the saturation exchange: all flow regimes temperature.

Flow- The process whereby liquid flows Closure (constitutive) 1-3.3.6 & IV-6.1 countercurrent opposite (counter) to the gas flow Relations (including regime direction. maps): lnterfacial momentum exchange: all flow regimes. 1-3.4.7 & IV-7.4 Special Process Models:

CCFL Flow - choked The maximum possible flow through a Special Process Models: 1-3.4.1 & IV-7.2 & IV-7.3 flow constricting item of hardware, Critical flow model usually a nozzle, orifice, or break in a pipe.

Flow- Flow that has two or more dominant Field Equation/Basic 1-3.1.11 multidimensional velocity vectors. Examples are Equation: 30 vessel model multidimensional flows in a PWR core during reflooding.

Heat conductance - The overall thermal resistance to the Field Equation/Basic 1-4.12 & IV-9.3 fuel-clad gap flow of heat between the fuel pellets Equation: fuel-clad gap and cladding in a nuclear fuel rod. model

NOC-AE-16003366 Attachment 6 Page 14 of 31 Phenomena Phenomena description Model RELAPS-30 Volume/Section Heat transfer - Process of energy transport by the Closure (constitutive) 1-3.3.9 & 3.3.10 & IV-4.2 forced convection to combined action of heat conduction, Relations (including regime vapor energy storage, and mixing motion. maps): Wall-to-fluid energy exchange: single-phase vapor Heat transfer - The process by which the energy within Field Equation/Basic 1-4 & IV-9.1 & 1-4.12 & IV-stored energy a solid structure is released to a lower Equation: Conduction 9.3 release energy state through one or more heat equation & fuel-clad gap transfer processes, e.g., conduction and convection. Applies specifically to the transport of the energy residing in fuel rods operating at full power to the coolant following a reactor trip.

The first peak is associated with the blowdown time period and is caused by the initial stored energy in the fuel rods and degraded fuel rod-to-coolant heat transfer.

lnterfacial shear The friction caused by the velocity Closure (constitutive) 1-3.3.6 & IV-6.1 difference between two phases at their Relations (including regime interface. maps): lnterfacial momentum exchange: all flow regimes Level The vertical height of a column of Field Equation/Basic 1-3.1.1 3.1.4 & IV-3.2 single- or two-phase fluid. Equation: all: fluid flow equations Noncondensable The impact of the presence of Closure (constitutive) IV-4.1.4 & IV-4.2.1 effects noncondensable gases upon heat Relations (including regime transfer or any other phenomenon such maps): lnterfacial energy as flow, condensation, flashing, and exchange: effect of vapor volume expansion. noncondensables Oscillations The periodic variation of any given Field Equation/Basic 1-3.1.1 3.1.4 & IV-2.1 hydraulic characteristic between two Equation: All fluid flow values. equations 1-3.3 & IV-6 Closure (constitutive)

Relations (including regime maps): Fluid momentum closure: all flow regimes 1-3.3 & IV-4 Closure (constitutive)

Relations (including regime maps): Fluid energy closure:

all flow reQimes Power-decay heat Heat produced by the decay of Field Equation/Basic 1-7 & IV-9.4 radioactive nuclides. Equation: Power generation Pump- The behavior of a pump under all Special Component Models: 1-3.5.4 & IV-8.1 performance, normal and off-normal conditions. Pump component including deQradation

N OC-AE-16003366 Attachment 6 Page 15 of 31 Table 2. SB LOCA Phenomena and Modeling Phenomena Phenomena description Model RELAP5-3D Volume/Section Condensation - fluid The process whereby steam is cooled Closure (constitutive) 1-3.3.9 & 1-3.3.10 & IV-to surface due to contact with a colder surface, Relations (including regime 4.2 resulting in a change of phase from maps): Wall-to-fluid energy vapor to liquid at the surface. exchange: condensation Condensation - The process whereby steam is cooled Closure (constitutive) 1-3.3.11 & IV-4.1 interfacial due to contact with a colder liquid, Relations (including regime resulting in a change of phase from maps): lnterfacial energy vapor to liquid at the interface between exchange: all flow regimes the two phases.

Entrainment I The process whereby liquid is captured Closure (constitutive) 1-3.3.6 & IV-6.1 deentrainment (entrained) by a high-velocity steam Relations (including regime flow. The process whereby liquid maps): lnterfacial momentum departs (deentrained) from a steam exchange: all flow regimes flow.

Flashing - interfacial The process whereby fluid changes Closure (constitutive) 1-3.3.11 & IV-4.1 from the liquid state to the vapor state Relations (including regime due to a reduction in the fluid pressure, maps): lnterfacial energy which lowers the saturation exchange: all flow regimes temperature.

Flow regime - break The characteristics of the flow at the Closure (constitutive) 1-3.1.1- 3.1.4 & IV-3 inlet break entrance, e.g., subcooled liquid, Relations (including regime saturated, two-phase, stratified, vapor, maps): All flow regimes.

etc.

Flow- The process whereby liquid flows Closure (constitutive) 1-3.3.6 & IV-6.1 countercurrent opposite (counter) to the gas flow Relations (including regime direction. maps): All flow regimes.

Special Process Model: CCFL 1-3.4.7 & IV-7.4 Flow - choked The maximum possible flow through a Special Process Model: 1-3.4.1 & IV-7.2 & IV-7.3 flow constricting item of hardware, Choked flow model usually a nozzle, orifice, or break in a pipe.

Flow- gap Flow through the hot leg to downcomer Field Equation/Basic 1-3.1.1 3.1.4 & IV-3.1 gap. Equation: All fluid flow equations Heat transfer - post- Heat transfer between the two-phase Closure (constitutive) 1-3.3.9 & 1-3.3.10 CHF fluid and the heated surface in the Relations (including regime IV-4.1 & IV-4.2 liquid-deficient region downstream of maps): Wall-to-fluid energy the CHF point, i.e., the location at which exchange & lnterfacial energy the heat transfer condition of the two- exchange: transition boiling, phase flow substantially deteriorates. film boilinQ lnterfacial shear The friction caused by the velocity Closure (constitutive) 1-3.3.6 & IV-6.1 difference between two phases at their Relations (including regime interface. maps): lnterfacial momentum exchanqe: all flow reqimes Level The vertical height of a column of Field Equation/Basic 1-3.1.1 3.1.4 & 1-3.3.1 single- or two-phase fluid. Equation: All fluid flow & IV-3.2 equations

NOC-AE-16003366 Attachment 6 Page 16 of 31 Phenomena Phenomena description Model RELAPS-30 Volume/Section Oxidation A chemical reaction that increases the Field Equation/Basic 1-4.15 oxidation content of a material. Of Equation: Heat structure:

specific interest is cladding oxidation, metal-water interaction which occurs at elevated temperatures, which can occur only under accident conditions.

Power-30 The axial, radial and azimuthal power Field Equation/Basic 1-7.2 distribution variation in a core. Equation: Reactor Kinetics Models - 30 kinetics Power - decay heat Heat produced by the decay of Field Equation/Basic 1-7 & IV-9.4 radioactive nuclides. Equation: Power Qeneration Power - local The ratio of power at a location (specific Field Equation/Basic 1-7.2 peaking (fuel rod) fuel rod) to the core average power. Equation: Reactor Kinetics Models - 30 kinetics Pressure drop The reduction in pressure with distance. Field Equation/Basic 1-3.1.1 3.1.4 & IV-3.1 Equation: All fluid flow equations 1-3.3 & IV-6; Closure (constitutive) 1-3.3 & IV-4 Relations (including regime maps): Momentum and energy equations closures Rewet The post-dryout process in which liquid Closure (constitutive) 1-3.3.9 & 1-3.3.10 & IV-once again resumes intimate contact Relations (including regime 4.2 with a heated surface. maps): Wall-to-fluid energy exchange: all flow regimes Closure (constitutive) IV-4.4.5 & IV-4.4.6 Relations (including regime maps): Reflood heat transfer models: Wall-to-fluid heat transfer.

Stratification - The variation of physical properties such Field Equation/Basic 1-3.1.1 3.1.4 & IV-3.1 horizontal as temperature or density across the Equation: Mass, momentum vertical cross section of a fluid body energy equations having a primarily horizontal orientation, 1-3.3.2 & IV-3.1 e.g., the cold leg of a nuclear steam Closure (constitutive) supply system. Relations (including regime maps): Regime maps:

stratified flow

Reference:

[1]. RELAP5-3D Code Manual, Vol. I "Code Structure, System Models and Solution Methods". INEEL-EXT-98-00834, Revision 4.1, September 2013.

NOC-AE-16003366 Attachment 6 Page 17 of 31 SNPB-3-11 Modeling of Important Phenomena Please provide a summary of the important phenomena and discuss how the L TCC EM models these phenomena.

Criterion 3.2

Reference:

SRP, lll.3b STP Response:

The RELAP5 series of codes has been developed at the Idaho National Laboratory (INL) under sponsorship of the U.S. Department of Energy, the U.S. Nuclear Regulatory Commission, members of the International Code Assessment and Applications Program (ICAP), members of the Code Applications and Maintenance Program (CAMP), and members of the International RELAP5 Users Group (IRUG). Specific applications of the code have included simulations of transients in light water reactor (LWR) systems such as loss of coolant, anticipated transients without scram (ATWS), and operational transients such as loss of feedwater, loss of offsite power, station blackout, and turbine trip.

RELAP5-3D is the latest in the series of RELAP5 codes and it includes all the capabilities of the RELAP5 family in simulating the behavior of a reactor coolant system during transients such as hypothesized LOCA scenarios. The mission of the RELAP5-3D development program was to develop a code version suitable for the analysis of all transients and postulated accidents in LWR systems, including both large- and small-break loss-of-coolant accidents (LOCAs) [1].

RELAP5-3D includes many component models from which normal operation and transients in PWR systems can be simulated. The component models include pumps, valves, pipes, transient conduction-convection heat transfer systems, reactor kinetics, separators, annuli, pressurizers, accumulators, and control syst~m components. In addition, special process models are included for effects such as form loss, flow at an abrupt area change, fluid-wall friction, branching, and choked flow [1 ].

The RELAP5 code has been widely used for analysis of LOCA scenarios of different break sizes including large and small breaks. The use of the code has been extended from the initial phases of the accident to the long term core cooling phase [2].

Volume V of the RELAP5-3D code manual [3] describes the modeling techniques developed by INL and researchers of national and foreign institutions for small and large break LOCA analysis. In particular, Section 5 is dedicated to the analysis of steady-state and transients of PWR, and in particular to large break LOCA (Section 5.1.8). The capabilities of the current version of the code and its predecessors have also been assessed against experimental data from integral effect test facilities and referenced in Section 5.1.9 of Volume V.

NOC-AE-16003366 Attachment 6 Page 18 of 31

References:

[1]. RELAP5-3D Code Manual, Vol. I "Code Structure, System Models and Solution Methods". INEEL-EXT-98-00834, Revision 4.1, September 2013.

[2]. NUREG/CR-6770 LA-UR-01-5561, "GSl-191: Thermal-Hydraulic Response of PWR Reactor Coolant System and Containments to Selected Accident Sequences," Los Alamos National Laboratory, August 2002.

[3]. RELAP5-3D Code Manual, Vol. V "User's Guidelines". INEEL-EXT-98-00834, Revision 4.1, September 2013.

N OC-AE-16003366 Attachment 6 Page 19 of 31 SNPB-3-12 Field Equations Please define and provide a summary of the field equations for the L TCC EM. This should include identification of the of the conservation equation (e.g., mass, momentum) and the number of dimensions of the equation. For portions of the RCS model that change in nodalization (e.g., 1-0 to 3-0), a separate description may be necessary. Additionally, demonstrate that these equations are able to model the necessary phenomena.

Criterion 3.3

Reference:

SRP, lll.3b STP Response:

The RELAP5-3D hydrodynamic model is a transient, two-fluid model for two-phase vapor/gas-liquid mixture flow using eight field equations (for eight primary dependent variables). The primary dependent variables are pressure (P), phasic specific internal energies (U 9 , Ut), vapor/gas volume fraction (void fraction) (a9 ), phasic velocities (v9 , Vt),

noncondensable quality (Xn), and boron density (pb). In the one-dimensional equation model, the independent variables are time (t) and distance.

The secondary dependent variables used in the equations are phasic densities (p 9 , pt),

phasic temperatures (T9 , Tr), saturation temperature (Ts), and noncondensable mass fraction in noncondensable gas phase (Xni) for the i-th noncondensable species.

The basic field equations for the two-fluid nonequilibrium model consist of:

• two phasic continuity equations,

• two phasic momentum equations, and

• two phasic energy equations.

The equations are written in differential stream tube form with time and one space dimension (one dimensional model) as independent variables and in terms of time and volume-average dependent variables.

Volume I of the RELAP5-3D code manual [1] provides a detailed description of the:

• field equations: Section 3.1,

• numerical formulation of the basic conservation of mass, momentum, and energy: Sections 3.1.1, and 3.1.2, and

• numerical scheme adopted: Sections 3.1.3-10.

The proposed LTCC EM uses exclusively one dimensional components to simulate the regions of the RCS. Only the one dimensional conservation equations are used 1 .

The use of the two-fluid nonequilibrium model has been largely discussed and proven to be more convenient than previous simplified approaches (for example, the well known homogeneous-equilibrium approximation). Bestion [2] highlighted the need for a two-fluid 1 The RELAP5-3D includes a three-dimensional set of conservation equations described in Section 3.1.11. Although these equations are not used in the EM.

N OC-AE-16003366 Attachment 6 Page 20 of 31 model in the current generation system codes, identifying unacceptable drawbacks of previous models based on improvements of the Homogeneous Equilibrium Model.

The advantage of the two-fluid model is that both mechanical non-equilibrium (phasic slip) and thermal non-equilibrium (different temperatures of the phases) can be modeled.

Mechanical non-equilibrium can be significant in some phases of a loss-of-coolant accident progression. Examples include behavior of the ECC water (when it is injected into the system it does not immediately mix and flow at the same velocity as the steam),

cooling water flowing down the downcomer counter current to escaping steam, countercurrent flow of steam and water occurring in the steam generator tubes during reflux cooling, and stratified flow occurring in horizontal piping components with little interaction between liquid and vapor. Important thermal non-equilibrium phenomena requiring a two-fluid model during LOCA simulation include sub-cooled liquid with direct contact condensation after ECCS injection and superheated vapor during Post-CHF heat transfer in the core [3].

References:

[1 ]. RELAP5-3D Code Manual, Vol. I "Code Structure, System Models and Solution Methods". INEEL-EXT-98-00834, Revision 4.1, September 2013.

[2] Dominique Bestion, "System Code Models and Capabilities", THICKET 2008

- Session Ill - Paper 06, 2013.

[3]. RELAP5/MOD3 Code Manual Volume 6: Validation of Numerical Techniques in RELAP5/MOD3.0; A. S. Shieh V. H. Ransom R. Krishnamurthy, NUREG/CR-5535/Rev 1-Vol VI.

NOC-AE-16003366 Attachment 6 Page 21 of 31 SNPB-3-13 Validation of Closure Relationships For the closure relationships identified, please provide appropriate validation for the use of this relationship over its expected application domain. This validation should include comparisons to separate effects tests and/or integral test data and appropriately address the model's uncertainty. Where appropriate, discuss any similarity criteria, scaling rationale, assumptions, simplifications, and/or compensating errors.

Criterion 3.4, 3.8, 3.9, 4.3, 4.6, 5.2, 5.4, 5.5, 5.6

Reference:

SRP, lll.3b, d, e STP Response:

The RELAP5-3D code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code is able to model the coupled behavior of the reactor coolant system and the core during loss-of-coolant accident scenarios. The code development has benefitted from extensive application and comparison to data from: phenomenological (PT}; separate effects (SET); and integral effects (IET} test cases (including LOFT, PBF, Semiscale, ACRR, NRU, and other experimental programs). A developmental assessment has been performed for the RELAP5-3D computer code. This assessment used a combination of phenomenological, separate effects, and integral effects test cases to investigate how well selected code models perform. A detailed description of the validation matrix, the experimental data used, and the results of the code assessment are included in the RELAP5-3D code manual (Volume Ill).

Tables 1 and 2 summarize the benchmark validation performed on the RELAP5-3D code based on each of the important phenomena for LB and SB LOCA respectively. These phenomena are identified and described in the response to RAl-SNPB-04. Table 3 lists the IET cases included in the code assessment for both LB and SB LOCAs.

N OC-AE-16003366 Attachment 6 Page 22 of 31 Table 1. LB LOCA Phenomena - Developmental Assessment Phenomena Vol IV Sec.# Test Type Test Name Asymmetries 5.6 IET LOFT L2-5 (3-D)

Boiling-film 4.11 SET ORNL THTF Tests 3.07.9B, 3.07.9N, 3.07.9W Boiling - transition 4.10 SET Bennett Heated Tube Tests 5358, 4.11 SET 5294 and 5394 4.12 SET ORNL THTF Tests 3.07.9B, 3.07.9N, 3.07.9W Royal Institute of Technology Tube Test 261 Condensation - interfacial 4.14 SET FLECHT-SEASET Test 31701 4.16 SET UPTF Downcomer Countercurrent Flow Test 6, Run 131 Entrainment I 3.5 PT Bubbling steam through liquid deentrainment 4.14 SET FLECHT-SEASET Test 31701 4.16 SET UPTF Downcomer Countercurrent Flow Test 6, Run 131 Evaporation - interfacial 4.1 SET Edwards' Pipe 4.8 SET GE Level Swell, 1 ft. Test 1004-3

/

4.9 SET GE Level Swell, 4 ft. Test 5801-15 Flashing - interfacial 4.1 SET Edwards' Pipe 4.3 SET Marviken Test 22 4.4 SET Marviken Test 24 Flow - countercurrent 3.7 PT Gravity wave 1-D 3.8 PT Gravity wave 3-D 4.15 SET Dukler air-water flooding 4.14 SET FLECHT-SEASET Test 31701 4.16 SET UPTF Downcomer Countercurrent Flow Test 6, Run 131 Flow - critical 4.1 SET Edwards' Pipe 4.2 SET Marviken Test 21 4.3 SET Marviken Test 22 4.4 SET Marviken Test 24 4.5 SET Marviken Test Jet Impinging Test 4.6 SET 11 Moby Dick air-water Flow - multidimensional 3.3 PT Water over Steam (3-D) 3.12 PT Pure radial symmetric flow (3-D) 3.13 PT Rigid body rotation (3-D) 3.14 PT R-theta symmetric flow (3-D)

Heat conductance - fuel-LB LOCA IET See Table 3 clad gap Heat transfer - forced LB LOCA IET See Table 3 convection to vapor Heat transfer - stored LB LOCA IET See Table 3 energy release lnterfacial shear 4.8 SET GE Level Swell, 1 ft. Test 1004-3

NOC-AE-16003366 Attachment 6 Page 23 of 31 Phenomena Vol IV Sec.# Test Type Test Name 4.9 SET GE Level Swell, 4 ft. Test 5801-4.5 SET 15 Marviken Test JIT 11 Level 3.4 PT Fill-Drain 3.5 PT Bubbling steam through liquid 3.6 PT Manometer 4.13 SET FLECHT-SEASET Test 31504 4.14 SET FLECHT-SEASET Test 31701 4.8 SET GE Level Swell, 1 ft. Test 1004-3 4.9 SET GE Level Swell, 4 ft. Test 5801-15 Noncondensable effects 3.6 PT Manometer 4.16 SET UPTF Downcomer Countercurrent Flow Test 6, Run 131 I Oscillations 3.6 PT Manometer Power-decay heat 3.10 PT Core power Pump - performance, 4.21 SET Full-scale reactor coolant pump including degradation

N OC-AE-16003366

.. Attachment 6 Page 24 of 31 Table 2. SB LOCA Phenomena - Model Validation Phenomena Manual Test Type Test Name Vol/Sec Condensation - 4.17 SET MIT Pressurizer Test ST4 fluid to surface Condensation - 4.13 SET FLECHT-SEASET Test 31504 interfacial 4.14 SET FLECHT-SEASET Test 31701 Entrainment I 3.5 PT Bubbling steam through liquid deentrainment 4.14 SET FLECHT-SEASET Test 31701 4.16 SET UPTF Downcomer Countercurrent Flow Test 6, Run 131 Flashing- 4.1 SET Edwards' Pipe interfacial 4.3 SET Marviken Test 22 4.4 SET Marviken Test 24 Flow regime -

SB LOCA JET See Table 3 break inlet Flow- 3.7 PT Gravity wave 1-D countercurrent 3.8 PT Gravity wave 3-D 4.15 SET Dukler air-water flooding 4.13 SET FLECHT-SEASET Test 31504 4.14 SET FLECHT-SEASET Test 31701 Flow - critical 4.1 SET Edwards' Pipe 4.2 SET Marviken Test 21 4.3 SET Marviken Test 22 4.4 SET Marviken Test 24 4.5 SET Marviken Test JIT 11 4.6 SET Moby Dick air-water Flow-gap SB LOCA JET See Table 3 Heat transfer - 4.12 SET Royal Institute of Technology Tube Test 261 post-CHF [1] VII- 2.3.1.4 IA 2 SET Royal Institute of Technology Tube Test SET lnterfacial shear 4.8 SET

- GE Level Swell, 1 ft. Test 1004-3 4.9 SET GE Level Swell, 4 ft. Test 5801-15 4.5 SET Marviken Test JIT 11

[1] Vll-2.3.1.5 IA-SET Kreisingt test facility

[1] Vll-2.3.2.2 IA: JET BETHSY 6-inch cold leg break

&5 Level 3.5 PT Bubbling steam through liquid 4.13 SET FLECHT-SEASET Test 31504 4.14 SET FLECHT-SEASET Test 31701 4.8 SET GE Level Swell, 1 ft. Test 1004-3 4.9 SET GE Level Swell, 4ft. Test 5801-15 Oxidation See reference [2] and [3]

Power-3D Not currently identified distribution Power - decay heat 3.10 PT Core 12ower Power - local Not currently identified 12eaking (fuel rod) 2 IA: Independent Evaluation

N OC-AE-16003366 Attachment 6 Page 25 of 31 Phenomena Manual Test Type Test Name Vol/Sec Pressure droQ [1] VI 1-2.3.1.2 IA:SET Low Flow and Natural Circulation Ex12eriment at theWSRC Rewet 5.2 IET ROSA-IV Test SB-CL-18 5.4 IET LOBI TestA1-04R 5.5 IET LOFT ExQeriment L2-5 (1-D) 5.6 IET LOFT Ex12eriment L2-5 (3-D)

Stratification - HI IET 2-in LOCA at Krsko Nuclear Power Plant horizontal Table 3. IET Benchmark Test Case Experiment Type Phenomena IET 111-5.4 LOBI Test A1-04R LB LOCA Slowdown phase of a CL LB LOCA in a PWR IET 111-5.5 LOFT L2-5 (1-D) LB LOCA Double-ended CL LB LOCA IET 111-5.6 LOFT L2-5 (3-D) LB LOCA Double-ended CL LB LOCA IET 111-5.1 LOFT L3-7 1-in. SB LOCA 1-in. CLSB LOCA IET 111-5.2 ROSA-IV Test SB-CL-18 6-in. SB LOCA 6-in. (5%) CL SB LOCA IET 111-5.3 Semiscale NC S-NC-1 Loop natural Steady-state single-phase NC circulation IET 111-5.3 Semiscale NC S-NC-2 Loop natural Steady state single-phase, two-phase, and circulation reflux natural circulation IET 111-5.3 Semiscale NC S-NC-3 Loop natural the effective heat transfer area during two-circulation phase loop natural circulation.

IET 111-5.3 Semiscale NC S-NC-10 Loop natural Steady state single-phase, two-phase, and circulation reflux natural circulation

[1] IET-IA (Vll-3.3.2.2) PMK-2 SB LOCA SBLOCA in the cold leg (7.4% cold leg break)

R5M3.1 PMK-2 7.4% Cold Leg Break Experiment (SPE-4)

[1] IET-IA (Vll-3.3.2.3) BETHSY 6.2TC SB LOCA 6 inch cold leg break (5% break area R5M3.1

[1] IET-IA (Vll-3.3.2.4) LOFT L3-6. SB LOCA offtake pipe connected at a right angle to the R5M3.1 cold leg of the active loop. The break orifice had a break area corresponding to a break diameter of 4 inches

[1] IET-IA (Vll-2.3.2.2 BETHSY Test 9.1 b/ISP27 SBLOCA 0.5% Cold Leg Break

&2.3.2.5) 6.2 TC 5.0% (6 inch) cold leg break R5M3.0

[1] IET-IA (Vll-2.3.2.4) BETHSY 4.1a-TC and 5.1a. Natural convection single-phase natural circulation, two-phase R5M3.0 natural circulation, and reflux at various secondary conditions

[1] IET-IA (Vll-2.3.2.3) LSTF SB-CL-18 SB LOCA 5% CL SB LOCA R5M3.0

[1] IET-IA (Vll-2.3.2.6) Semiscale S-NH-1 SB LOCA 0.5% small break loss-of-coolant accident in R5M3.0 the cold leg

NOC-AE-16003366 Attachment 6 Page 26 of 31

References:

[1] RELAP5/Mod3 code manual, Vol. 7: Summaries and reviews of independent code assessment reports, NUREG/CR-5535, INEL-95/0174

[2]. Thomas K.S. Liang, Huan-Jen Hung, Chin-Jang Chang, Lance Wang, "Development of LOCA Calculation Capability with RELAP5-3D in Accordance with the Evaluation Model Methodology", lcone-9, 2001.

[3]. Thomas K. S. Liang, Development of an Appendix K Version of RELAP5-3D and Associated Deterministic-Realistic Hybrid Methodology for LOCA Licensing Analysis, http://www.intechopen.com/

[4]. I. Parzer, B. Mavko, Analysis of RELAP5/MOD3.3 Prediction of 2-lnch Loss-of-Coolant Accident at Krsko Nuclear Power Plant, NUREG/IA-0222.

NOC-AE-16003366 Attachment 6 Page 27 of 31 SNPB-3-14 Simplifying and Averaging Please provide a summary of the key simplifying and averaging assumptions used in the generation of the mathematical models used in the L TCC EM and demonstrate that they are appropriate for the accident scenarios being modeled.

Criterion 3.5 Reference SRP, lll.3b STP Response:

The RELAP5-3D code manual (Volume I) provides a detailed description of the field equations (eight state equations) and the simplification and averaging assumptions used in solving these equations in space and time [1].

The RELAP5-3D hydrodynamic model is a transient, two-fluid model for flow of a two-phase vapor/gas-liquid mixture that can contain noncondensable components in the vapor/gas phase and/or a soluble component (i.e., boron) in the liquid phase. A one-dimensional as well as a multi-dimensional hydrodynamic model is included in the code, although only one-dimensional components are used in this EM. The two-fluid equations of motion that are used as the basis for the RELAP5-3D hydrodynamic model are formulated in terms of volume and time-averaged parameters of the flow. A detailed description of the equations formulation is included in Section 3.1 of Volume 1 [1].

Phenomena that depend upon transverse gradients, such as friction and heat transfer, are formulated in terms of the bulk properties using empirical transfer coefficient formulations. In situations where transverse gradients cann.ot be represented within the framework of empirical transfer coefficients, such as subcooled boiling, special purpose models developed for the particular situation are employed.

The set of basic equations are described in Section 3.1.1 of Volume 1 [1 ]; and differencing (finite differences) to a more convenient set of differential equations upon which to base the numerical scheme are described in Section 3.1.2 [1].

The system model is solved numerically using a semi-implicit finite-difference technique.

The numerical technique is based on replacing the system of differential equations with a system of finite difference equations partially implicit in time. The terms evaluated implicitly are identified as the scheme is developed. In all cases, the implicit terms are formulated to be linear in the dependent variables at new time; a linear time-advancement matrix is solved by direct inversion using the default border-profile lower upper (BPLU). The difference equations are based on the concept of a control volume (or mesh cell, or node) in which mass and energy are conserved by equating accumulation to the rate of mass and energy in through the cell boundaries minus the rate of mass and energy out through the cell boundaries plus the source terms. The velocities at boundaries are most conveniently defined through use of momentum control volumes (cells) centered on the mass and energy cell boundaries. The scalar properties (pressure, specific internal energies, and void fraction) of the flow are defined at cell centers, and vector quantities (velocities) are defined on the cell boundaries.

A detailed description of the semi-implicit numerical method adopted, the guidelines followed in developing the numerical approximations, and the time advancement for the semi-implicit scheme is reported in Sections 3.1.3 and 3.1.4 [1].

NOC-AE-16003366 Attachment 6 Page 28 of 31

Reference:

[1]. RELAP5-3D Code Manual, Vol. I "Code Structure, System Models and Solution Methods". INEEL-EXT-98-00834, Revision 4.1, September 2013.

N OC-AE-16003366 Attachment 6 Page 29 of 31 SNPB-3-16 Single Version of the Evaluation Model Please confirm that a single version of the EM was used during the simulations of the given accident scenarios. This includes confirming that the code version was frozen and the manner for calculating or obtaining inputs did not change.

Criterion 4.1

Reference:

SRP 111.d STP Response:

Yes, a single version of the EM was used. Code version was controlled in accordance with OPGPO?-ZA-0014 "Software Quality Assurance Program" and the manner of controlling inputs did not change.

NOC-AE-16003366 Attachment 6 Page 30 of 31 SNPB-3-19 Initial Test Cases Please provide a summary of the assessment cases performed in order to demonstrate that RELAPS-30 has been installed and is being used appropriately.

Criterion 4.7

Reference:

SRP, lll.3d STP Response:

The RELAP5-3D software has been installed and currently is currently being applied in conformance with the South Texas Project Electric Generating Station Software Quality Assurance Program OPGPO?-ZA-0014 revision 10 [1]. A RELAP5-3D SQA package (per procedure) has been developed using the guidance of paragraph 6.3 of [1]. The SQA documentation [2] has been developed per Addendum 4 of the procedure (Procured Software Developed by Industry Regulators or Industry Organizations). Addendum 4 requires the preparation of a test plan following the guidelines included in Addendum 12, and a test case or test report formatted per Addendum 13. The SQA package [2] includes a summary of the test cases performed to demonstrate that the RELAP-30 software version has been installed and is being used appropriately.

The table below shows the list of run-time environment (RTE) test problems included in the SQA package.

N OC-AE-16003366 Attachment 6 Page 31 of 31 Table 1. SQA Package - RTE Test Problems Case# Input Name Description Basic one dimensional flow test for implementing 3-D 1 3dflow.i capabilities Long term decay heat study with Proposed 2005 ANS 2 ans05.i Standard data (ans05-4)

Long term decay heat study with Proposed 1979 ANS 3 ans79.i Standard data (ans79-3)

Pipe blowdown plus heat structures coupled to the pipe, a 4 edhtrk.i heat structure with a simple analytic solution, reactor kinetics and control components with a few trips Similar to edhtrk.i with nearly implicit advancement used 5 edhtrkn.i instead of semi-implicit advancement Similar to edhtrk.i with time dependent actinide and fission 6 edhtrt.i concentration as a function of power history 7 edrst.i Restart file to check similarity with the results of edhtrk.i Transient file for analyzing water replacing steam in control 8 fldrn2.i volume 9 refbun.i LBLOCA with reflood 10 todcnd.i 5 pipe cells with a hot wall (2 separate heat structures)

Simulation of a four loop pressurized water reactor 11 typ1200n2.i undergoing a small break. Timestep card control word is 15.

References:

[1 ]. OPGP07-ZA-0014 "Software Quality Assurance Program" STI 33856275 rev.10, 04/17/2014.

[2] RELAP5-3D Version 4.1.3 Software Quality Assurance Package, Rev.O, STI 34280651, FSUG File No. 060.

NOC-AE-16003366 Attachment 7 Attachment 7 Affidavit for Withholding for Response to RAl-18

Ci>

AL I 0 N ALION Science & Technology SCltNC!. AHD TECHNOLOGY AFFIDAVIT We, Dominic Mufi.oz, Project Manager and Martin Rozboril, Jr. Assistant Vice President Division Manager (AVPDM) state as follows:

(1) We, Dominic Mufioz, Project Manager, and Martin Rozboril, Jr. AVPDM, Nuclear Services, ALION Science & Technology ("Alion") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in all revisions of ALION Science

& Technology report "Erosion Testing of Small Pieces of Low Density Fiberglass Debris-Test Report," ALION-REP-ALION-1006-04, with the latest revision to date, Rev. 1, dated November 17, 20011. Information from this report was used to support the South Texas Project January 2016 follow-up to 2009 RAI-18. Two versions of the follow-up to 2009 RAI-18 are being provided to the NRC. The first version of follow-up RAI-18 with information from the erosion report, ALION-REP-ALION-1006-04 included forNRC technical review, and the second version ofRAI-18 with proprietary information from ALION-REP-ALION-1006-04 is redacted for public release.

(3) In making this application for withholding of proprietary information of which it is the owner, Alion relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA. 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Alion's competitors without license from Alion constitutes a competitive economic advantage over other companies; Page 1 of3 MAS Affidavit

AL I 0 N

• SCIEHCE AHO TfCHl'tDLDG'f

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future Alion customer-funded development plans and programs, resulting in potential products to Alion;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4) a, and (4) b, above.

(5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by Alion, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Alion, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Alion is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or their delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Alion are limited* to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The document identified in paragraph (2), above, is classified as proprietary because it contains "know-how" and "unique data" developed by Alion within our research and development programs. The development of this document, supporting methods and data constitutes a major Alion asset in this current market.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial Page 2 of3 MAS Affidavit

AL I 0 N

• sc1rncf AHO TECHNOLOGY harm to Alion's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Alion's comprehensive BWR/PWR GSI-191 analysis base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and experimental methodology and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, analytical and experimental costs comprise a substantial investment of time and money by Ali on.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Alion's competitive advantage will be lost if its competitors are able to use the results of the Alion experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Alion would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Alion of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 6th day of May 2016.

Dom:,,;v,/Y/11 ~

Dominic Mufioz ~

Project Manager Assistant Vice President ALION Science & Technology Division Manager, Nuclear Services ALION Science & Technology Page 3 of3 MAS Affidavit

NOC-AE-16003366 Attachment 8 Attachment 8 Definitions and Acronyms

NOC-AE-16003366 Attachment 8 Page 1?f2 ANS American Nuclear Society EOF Emergency Operations ARL Alden Research Laboratory Facility ASME American Society of EOP Emergency Operating Mechanical Engineers Procedure(s)

BA Boric Acid EPRI Electric Power Research BAP Boric Acid Precipitation Institute BC Branch Connection EQ Equipment Qualification BEP Best Efficiency Point ESF Engineered Safety Feature B-F Bimetallic Welds FA Fuel Assembly(s)

B-J Single Metal Welds FHB Fuel Handling Building BWR Boiling Water Reactor GDC General Design Criterion(ia)

CAD Computer Aided Design GL Generic Letter CASA Containment Accident GSI Generic Safety Issue Stochastic Analysis HHSI High Head Safety Injection CCDF Complementary Cumulative (ECCS Subsystem)

Distribution Function or HLB Hot Leg Break Conditional Core Damage HTVL High Temperature Vertical Frequency Loop ccw Component Cooling Water HLSO Hot Leg Switchover CDF Core Damage Frequency ID Inside Diameter CET Core Exit Thermocouple(s) IGSCC lntergranular Stress CHLE Corrosion/Head Loss Corrosion Cracking Experiments ISi In-Service Inspection CHRS Containment Heat Removal LAR License Amendment System Request CLB Cold Leg Break or Current LBB Leak Before Break Licensing Basis LBLOCA Large Break Loss of Coolant CRMP Configuration Risk Accident Management Program LCO Limiting Condition for cs Containment Spray Operability CSHL Clean Strainer Head Loss LDFG Low Density Fiberglass css Containment Spray System LERF Large Early Release (same as CS) Frequency eves Chemical Volume Control LHS Latin Hypercube Sampling System LHSI Low Head Safety Injection OBA Design Basis Accident (ECCS Subsystem)

DBD Design Basis Document LOCA Loss of Coolant Accident D&C Design and Construction LOOP/LOSP Loss of Off Site Power Defects LTCC Long Term Core Cooling DEGB Double Ended Guillotine MAAP Modular Accident Analysis Break Program DID Defense in Depth MAB/MEAS Mechanical Auxiliary Building DM Degradation Mechanism or Mechanical Electrical ECC Emergency Core Cooling Auxiliary Building (same as ECCS) MBLOCA Medium Break Loss of ECCS Emergency Core Cooling Coolant Accident System NIST National Institute of ECWS Essential Cooling Water Standards and Technology System (also ECW) NLHS Non-uniform Latin Hypercube EM Evaluation Model Sampling

NOC-AE-16003366 Attachment 8 Page 2 of 2 Definitions and Acronyms NPSH Net Positive Suction Head, RWST Refueling Water Storage (NPSHA - available, NPSHR Tank

- required) SBLOCA Small Break Loss of Coolant NRC Nuclear Regulatory Accident Commission SC Stress Corrosion NSSS Nuclear Steam Supply SI/SIS Safety Injection, Safety System Injection System (same as OBE Operating Basis Earthquake ECCS)

OD Outer Diameter SIR Safety Injection and PCI Performance Contracting, Recirculation Inc. SR Surveillance Requirement PCT Peak Clad Temperature, SRM Staff Requirements PDF Probability Density Function Memorandum PRA Probabilistic Risk SSE ~afe Shutdown Earthquake Assessment STP South Texas Project PWR Pressurized Water Reactor STPEGS South Texas Project Electric PW ROG Pressurized Water Reactor Generating Station Owner's Group STPNOC STP Nuclear Operating PWSCC Primary Water Stress Company Corrosion Cracking TAMU Texas A&M University QDPS Qualified Display Processing TF Thermal Fatigue System TGSCC Transgranular Stress RAI Request for Additional Corrosion Cracking Information TS Technical Specification(s)

RCB Reactor Containment TSB Technical Specification Building Bases RCFC Reactor Containment Fan TSC Technical Support Center Cooler TSP Trisodium Phosphate RCS Reactor Coolant System UFSAR Updated Final Safety RG Regulatory Guide Analysis- Report RHR Residual Heat Removal UHSN Upper Head Spray Nozzle RI-ISi Risk-Informed In-Service UNM University of New Mexico Inspection USI Unresolved Safety Issue RMI Reflective Metal Insulation UT University of Texas (Austin)

RMTS Risk Managed Technical V&V Verification and Validation Specifications VF Vibration Fatigue Rovero Risk over Deterministic WCAP Westinghouse Commercial Methodology Atomic Power RVWL Reactor Vessel Water Level ZOI Zone of Influence