NOC-AE-16003395, Third Set of Responses to Requests for Additional Information STP Risk-Informed GSI-191 Licensing Application

Third Set of Responses to Requests for Additional Information STP Risk-Informed GSI-191 Licensing Application

ML16229A189
Person / Time
Site:	South Texas
Issue date:	07/21/2016
From:	Powell G T South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GSI-191, NOC-AE-16003395, TAC MF2400, TAC MF2401
Download: ML16229A189 (72)

Text

Soutll li:x.75 Project Electric Gr:11erall11g Statlo11 P.O. Bar 1$9 1%dswortll, T=s 77./$1 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 &2 Docket Nos. STN 50-498, STN 50-499 July 21, 2016 NOC-AE-16003395 10 CFR 50.12 10 CFR 50.90 Third Set of Responses to April 11, 2016 Requests for Additional Information Regarding STP Risk-Informed GSl-191 Licensing Application Response to SNPB RAls (TAC NOs MF2400 and MF2401)

References:

1. Letter, G. T. Powell, STPNOC, to NRG Document Control Desk, "Supplement 2 to STP Pilot Submittal and Requests for Exemptions and License Amendment for a Informed Approach to Address Generic Safety Issue (GSl)-191 and Respond to Generic Letter (GL) 2004-02", August 20, 2015, NOC-AE-15003241, ML 15246A126

2. Letter, Lisa Regner, NRG, to Dennis Koehl, STPNOC, "South Texas Project, Units 1 and 2-Request for Additional Information Related to Request for Exemptions and License Amendment for Use of a Risk-Informed Approach to Resolve the Issue of Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized-Water Reactors", April 11, 2016, ML 16082A507 Reference 2 transmitted RAls on STPNOC's application iri Reference 1 and divided the RAls into 3 sets to be responded to in 30-day intervals.

This submittal responds to the third set of RAls from the Nuclear Performance and Code Branch (SNPB). There are no commitments in this submittal.

STl34345851 NOC-AE-16003395 Page 2 of 3 If there are any questions, please contact Mr. Wayne Harrison at 361-972-877

4. I declare under penalty of perjury that the foregoing is true and correct. Executed on: awh Attachments:

G. T. Powell Executive Vice President , and Chief Nuclear Officer 1. Response to SNPB-3-2,-6, -7, -15, -17, -18, and -20 through -32 2. Definitions and Acronyms cc: (paper copy) Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (08H04) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN116 Wadsworth, TX 77483 (electronic copy) NOC-AE-16003395 Page 3 of 3 Morgan. Lewis & Beckius LLP Steven P. Frantz, Esquire U. S. Nuclear Regulatorv Commission Lisa M. Regner NRG South Texas LP Chris O'Hara Jim von Suskil Skip Zahn CPS Energy Kevin Pollo Cris Eugster L. D. Blaylock Crain Caton & James. P.C. Peter Nemeth City of Austin Elaina Bail John Wester Texas Dept of State Health Services Helen Watkins

• Robert Free Attachment 1 NOC-AE-16003395 Attachment 1 Response to SNPB-3-2, -6, -7, -15, -17, -18 and -20 through -32 Preface to SNPB Responses NOC-AE-16003395 Attachment 1 Page 1 of65 STPNOC has revised its L TCC approaches to the cold leg SBLOCA and the hot leg LBLOCA that reduces the scope of the RELAP5-3D L TCC analyses and the need to address these breaks in the SNPB RAI responses.

STPNOC determined that debris effects in the cold leg SBLOCA can be addressed using the same methodology that is described in Reference 1 to the cover letter (Section 3 of Attachment 1-3). The conclusion is that there is not sufficient debris to affect L TCC for the cold leg SBLOCA and there is no need to perform the RELAP5-3D simulation for that event. STPNOC has revised its analysis to consider the maximum deterministically accepted HLB size to be the largest HLB oisman, regardless of the amount of fine fiber generated.

As a consequence, the thermal-hydraulic analysis is limited to 16" and smaller HLB. Therefore, HLB sizes between 16" and DEGB are assessed as risk-informed locations that don't require thermal-hydraulic analysis.

This change adds eight critical weld locations (RPV nozzle welds) to Table 16 in Attachment 1-3 to Reference 1 of the cover letter. There is minimal effect on the risk quantification.

SNPB-3-2 Accident Scenario Progression Please provide a description of the accident progression of the accident scenarios being simulated using the Jong-term core cooling (L TCC) evaluation model (EM). This description should start at the initiation of the break, define each phase, and provide the important phenomena occurring in that phase in the various locations of the reactor coolant system (RCS) (e.g., core, reactor vessel, steam generators

-both primary and secondary side, loops, pressurizer, pumps, containment).

Criterion

1.2 Reference

SRP, llL*3c STP Response:

The description of the accident progression is provided below. The description is limited to the large (16") break in hot leg which is shown to exhibit the same phenomenology but under more severe conditions than smaller breaks. The break is assumed in the hot leg of one of the coolant loops equipped with the SI train (loop 3) and located in the horizontal section of the leg. Each transient simulation is executed as restart from the steady-state simulation

[1], and preceded by a 300-second null transient period. The break is assumed to open instantaneously at the end of the null transient period. The accident scenario progression is divided into four time periods:

• Period 1: Break Event and Slowdown (300 s. --396 s.)

• Period 2: Refill and Reflood (-314 s. --434 s.)

• Period 3: Pre-Blockage Long Term Core Cooling (-434 s. -2099 s.)

• Period 4: Core Blockage and Post-Blockage Long Term Core Cooling (2099 s. and after)

Period 1: Break Event and Slowdown (300 s. --396 s.) NOC-AE-16003395 Attachment 1 Page 2 of65 The break opens instantaneously at the end of the null transient period (300 s.). The primary system rapidly depressurizes due to the mass and energy discharge from the break. At 306 seconds the low pressurizer pressure (1872 psia) signal is reached. This signal trips the SI system (high and low pressure pumps) and the reactor scram. The reactor core is fully scrammed at 311 seconds. Other important events observed during this phase are listed in the table below. Phase Event Time (s) Break Opening 300 Low PZR pressure Signal 306 MFW isolation 308 Blowdown Reactor Trip 308 Reactor Fully Scrammed 311 HHSI Pumps Activation 312 RCP Trip 314 LHSI Pumps Activation 316 As the depressurization continues, voids are created in the core. Since the break is located in the hot side of the primary system (downstream the core exit), core flow stagnation is not observed.

Instead, a large amount of cooling water from the cold side is forced to pass through the reactor core before reaching the break. The cladding temperature steadily decreases during this phase. The core collapsed liquid level also decreases.

The figure below shows the core collapsed liquid level, and the safety injection flow rates during this phase.

Vi' ...... co -' w a:: $ 0 -' u.. Vl Vl c:i: -Toi lC(S mflawR.nt (e*cluding Accum) -To t M AcctsTiul.ttOt mfJowlWtf'

-Corf' Cll 4500 4000 ,500 3000 2500 2000 1500 1000 HPS I i n j ect i on st art 500 L PS I i n j*ct ion start 500 190 390 440 SIMULAT I ON TIME [SJ NOC-AE-16003395 Attachment 1 Page 3 of 65 490 16 12 to 'i=' u.. z 0 > w 6 -' w The core collapsed liquid level reaches a first minim um a t 396 seconds. During t his period the primary s ide of a ll steam generato r s is empty. Period 2: Refill and Reflood (-314 s. -4 34 s.) Due to the location of the break , t h e ref i ll an d ref lo od phases may not be easily distingu i shable compared to typical l a rg e b r eak scenarios in tha t co l d leg. While the SI system injection starts at approximately 314 seconds (HHSI), the refill phase is assumed to start when the core collapsed liquid level s t arts increasing. The injection of the accumulators (accumulators' injecti o n starts at 382 seconds) in the cold legs is preferentially diverted through the core. The co ll apsed liquid level is partially recovered due to the accumulators' injection.

As the accumulators

' injection decreases , the core void fraction and liquid mass inventory s t arts decreasing reaching a second minimum at 417 seconds. As the primary pressure decreases , LHS I pumps are able to start injecting (at approximately 394 seconds).

The LHSI flow combined with the HPSI flow (initiated early in the phase at approx i mately 314 seconds), provides at total flow over 1 500 lbm/s. Most of the SI flow is diverted toward the reactor vessel, and passes through the reactor core , recove ri ng the collapsed liquid level. The core is completely flooded (collapsed liquid level at the top of the core) at approximately 434 seconds. During this period the primary sides of all steam gene r ators i s empty. Other important events observed during this phase are listed in the table below.

40 35 3-0 f=' 25 L.L __, UJ 20 > UJ __, 15 10 Phase Event AFW Activation Refill/Reflood Accumulator Activation Minimum Core Collapsed Liquid Level Core Full NOC-AE-16003395 Attachment 1 Page 4 of 65 Time (s) 338 383 417 434 Period 3: Pre-Blockage Long Term Core Cooling (-434 s. -2099 s.) During this phase the main parameters of the reactor core (core collapsed liquid level , PCT) and the primary system (break flow rates , ECCS flow rate, primary water inventory) do not show appreciable changes over time. The core collapsed liquid level is steadily maintained well above the top of the core by the ECCS cooling water forced to flow through the core. The PCT stabilizes at approximately 300 °F with a slow decrease rate due to the decay power decrease. The injected ECCS f l ow is also partly diverted toward the SGs. The SGs primary side tubes slowly fill up with water from both cold and hot leg sides. All four SGs primary and secondary sides show similar behavior. No appreciab le difference is seen between broken loop and intact loops as shown in the figures below. -Cll SG l , 1' primary s i de -CLL SGl , J.. pr1mciryside

---*Sump Swi t chover --Blockage Time 1 000 2000 3000 4000 5000 6000 SIMULATION TIME (SJ SG 1 (Intact Loop, Pressurizer Loop) Primary Side Collapsed Liquid Level 40 35 30 t;: 25 ...J w 20 > w ...J 1 5 10 -CLLSG2,1'pr i marvside -CLLSG2 ,.J,,primaryside

---*SumpSwitchovt-r --SlockageTime 1000 2000 3000 4000 SIMULATION T I ME [SJ NOC-AE-16003395 Attachment 1 Page 5 of 65 5000 6000 SG 2 (Intact Loop) Primary Side Collapsed Liquid Level 40 35 30 ...J LU 20 > LU ...J IS 10 40 3S 30 ...J LU 20 > LU ...J IS 1 0 0 --Cll SG3, 't primary side --CLL SG3, ..J, primary side ----Sump Switchover --Blockage Time 1000 2000 3000 4000 SIMULATION TIME (SJ NOC-AE-16003395 Attachment 1 Page 6 of 65 5000 6000 SG 3 (Broken Loop) Primary Side Collapsed Liquid Level --CLL SG4, 1" pnmary side --CLL SG4, J, primary side ---*Sump Switchover

--Blockage Time 1 000 2000 3000 4000 5 000 6000 SIMULATION TIME [SJ SG 4 (Intact Loop) Primary Side Collapsed Liquid Level 60 50 40 -' 30 UJ > UJ -' i=' u.. 20 1 0 60 50 40 -' 30 UJ > UJ -' 20 10 500 500 --CLLSGl,secondaryslde

---

SumpSw11chowr --Bloclc.ageTime

--AFWSGl -MFWSGl 1000 1500 2000 2500 3000 SIMULATION TIME [SJ Inta c t Loop (Loop 1) S e c o ndary Sid e Co n d itions --C LL SG3, secondary side ----Sump Switchover Blockage Time --AFWSG3 MFWSG3 1000 1500 2000 2500 3000 SIMULATION TIME [SJ Broken Lo o p (Loop 3) Secondary Side C o nditions NOC-AE-1 6003395 A tt achment 1 Page 7 o f 65 70 60 Vi' ......_ 50 CXl -' UJ 40 !<i: a:: 5 30 9 u.. Vl Vl 20 <t 10 3500 4000 70 60 Vi' ......_ 50 CXl -' UJ 40 !<i: a:: s 0 30 -' u.. Vl Vl 20 <t 10 3500 4000 2ao LI:' 230 UJ a:: ::::> \:t ffi 1 ao a.. UJ f-NOC-AE-16003395 Attachment 1 Page 8 of 65 The SSO time occurs at 1740 seconds when the RWST low-low-level alarm is reached. At this time the ECCS injection switches from the RWST temperature to the sump pool temperature as shown in the figure below. -ECCStnlet T empe r a tur e -RWST/StJmp2 T empera t Ufe ---*SumpSwitdiove r --Blocka g e T i m e 80 1 000 2000 3000 4000 sooo 6000 SIMULATION TIME [S J ECCS suction Temperature (Blue Line) and ECCS Injection Temperature (Gray Line) There is no appreciable effect on t he overall p rim ary sy ste m behavior due to the increase in the ECCS injection tem p era t ure following t he SSO time. Core coolant inlet and outlet temperatures remain subcoo l ed during this phase. After the SSO time, core coolant temperatures increase fo llo wing the ECCS injection temperature change. While the inlet temperature remains subcooled , the outlet temperature reaches the saturation temperature before the core blockage event. Subsequently, void is produced in the core in small amounts as the core liquid inventory slightly decreases. The PCT shows a trend s i milar to the core coolant temperatures as depicted in the figure below.

500 480 u.. 0 UJ a:: :::::> f-::: 380 <t a:: UJ a.. ::::? 280 UJ t--180 80 290 NOC-AE-16003395 Attachment 1 Page 9 of 65 -PCT ---*SumpSwitchover --BlockageTime

-CorelnTemperature

-coreOut T emperature

-coreOUtSatT 490 690 890 1090 1290 1490 1690 1890 2090 SIMULATION TIME [SJ Core Temperatures The f u ll core and core bypass blockage is assumed to occur instantaneously at 2100 seconds (360 seconds aft er the SSO time). Period 4: Core Blockage and Pos t-Blo c kage Long Te r m Core Cooling (2099 s. and after) The sudden decrease in the core flow rate due to the i nstantaneous core blockage a t the bottom of the core produces void in the core, and a subsequent reduction of the core liquid inventory.

A slight increase in the primary pressure is essentially related to the vapor generat i on in the core. This is also the cause of the increase of the saturation temperature.

The PCT follows the satu r ation temperatu r e as shown in the figure below (see description of the core heat transfer regimes).

BJ G:' 280 '2.... UJ a:: => ti: 230 a:: UJ a.. 180 130 NOC-AE-16003395 Attachment 1 Page 10 of 65 -PCT ---*St.ntpSw1tcho\iter --Bfc:di:ag,eTime

-CoreOutTemperiture

-C0teOUtS.tT

-CoreCLl 16 12 10 u.. z 0 > UJ 6 -I UJ 80 1000 1500 1000 1500 3500 4500 5000 0 6000 SIMULATION TIME (SJ Core Temperatures and Core Collapsed Liquid Level. The core collapsed liquid level decreases u ntil a stable value of approximately 8 ft is reached. This is an indication that the reacto r c ore inlet/outlet mass flow i s balanced between the core coolant evaporation rate and the liquid injection from the top of the core. The SGs' primary side liquid inventory and flow ra t es at the core blockage time can be summarized as follows:

• All SG primary tubes are found to be almost full at the time of core blockage from both cold and hot leg sides.

• No appreciable net flow through the SGs' p ri mary tubes is observed an instant before the core blockage time. When the core blockage occurs, a redistribution of the flow within t he primary system occurs. The behavior of the SGs immediatel y before and after the core blockage is described below. The description is complemented with plo t s of the integral flow rates through each of the loops. Broken Loop (Loop 3) The ECCS flow injected in the cold l eg of the broken loop is mostly forced toward the reactor vessel before the core blockage time. At the core blockage time flow from the ECCS is also forced toward the SG. The primary side of the Loop 3 SG is immediately filled and flow is established from the cold side to the hot side of Loop 3, toward the break. Most of the ECCS flow through Loop 3 is now forced toward the SG. This flow can be

'iii' _J UJ a:: 3 0 _J u. Vl Vl <{ NOC-AE-16003395 Attachment 1 Page 11 of 65 assumed to be fully discharged from the break located in the hot leg of Loop 3. The integral flow splits within Loop 3 are shown in the figure below 1. -ln t (H L 3*>Vessel}

-lnt(CL3->Vessel)

-l nt(CL3->Pump.3)

-l n t (ECCS->L oop3) ---*SumpSwrtchove r --Slockage l 1 me 8000000 6000000 4000000 2000000 5000 6000 0 -2000000 UJ a:: \,!) -400JOOO UJ 1-z -6000000 -llXXJOOO SIMULATION TIME [SJ Br ok en Loop (Loop 3) Integral Flow S plits Intact Loops Equ i pped with SI trains (Loop s 2 and 4) The behavior of the intact loops equipped with SI trains appears to be s i milar during the transient and in particular dur i ng the pre-and post-core blockage phases. The ECCS flow injected in the cold leg of intact Loops 2 and 4 i s mostly forced toward the reactor vessel before the core blockage time. At the core blockage time flow from the ECCS is also forced toward the SG. The primary side of the SG of Loops 2 and 4 is also immediately filled and flow is established from the cold side to the hot side of these loops. Th i s flow provides 1 The main parameters of the plots are explained here after: lnt(ECCS 7 Loop X): ECCS Integral flow injected into loop X lnt(CLX 7 Vessel): Integral flow from CL of Loop X toward the reactor vessel (taken at the vesse l inlet). lnt(CLX 7 Pump): Integral flow from CL o f Loop X toward the RCP of the same loop (take n at the pump inlet) lnt(HLX 7 Vessel): Integral flow from HL of Loop X toward the reactor vessel (taken at the vessel outlet). The integral is defined positive in the sense of the arrow. A lower s l ow er slope the plot indicates a lower time-average flow r ate toward the direction specified. A change in the slope sign i s an indicati on of the change in the flow direction.

NOC-AE-16003395 Attachment 1 Page 12 of 65 liquid to the region immediately above the reactor core. The integral flow splits within Loop 2 and Loop 4 are shown in the figures below (see also Note 1). -lnt{HL2*>Ve.ssel}

-1nt(CL2->Vessell

-ln t{ECCS->Loop2)

---*SumpSw1tchover --BlockageT ime 8000000 I I 500 1000 1500 1000 I 2500 3000 3500 4000 I I I -4000000 SIMULATION TIME [SJ Intact Loop (Loop 2) I ntegral Flow Sp l i t s -ln t (Hl4->Vessel)

-lnt{CL4->Vessel)

-lnt(ECCS->loop4)

---

SUmpSw1td10\<er

--Bloc:k*gel1me 8000000 co 6000000 ....J w a:: 4000000 3! 0 I ....J u. Vl 2000000 Vl <( :;? Cl w 500 1000 1 500 2000 I 1500 3000 3500 4000 a:: I (.!) w I I-I z -2000000 I I *4000000 SIMULATION TIME [SJ Intact Loop (Loop 4) Integral Flow Splits UJ Intact Loop Not Equipped with SI train (Loop 1) NOC-AE-16003395 Attachment 1 Page 13 of 65 Immediately before the core blockage time, this loop also appears to be mostly full of water. At the core blockage time, Loop 1 SG fills up to the top, while net flow from the cold side to the hot side is established , contributing to the total liquid flow reaching the top of the core. The integral flow splits within Loop 1 are shown in the figures below (see also Note 1). -lnt(Hl l>>Vessel) --lnt{Cl l*>VesseO ----Sump Switchover

--Bhxkagelime 8000000 4000000 I a UJ a:: (.!) UJ f--500 1 000 I I 1500 2000 I I I 2500 3500 4000 4500 z .200000() -4000000 SIMULATiON TIME (SJ Intact Loop (Loop 1) Int e gral Flow Splits Analysis of the case executed also showed t hat the majority of the flow reaching the top of the core is supplied through the SGs with a sma ll (not relevant) flow passing through the upper plenum sprays. The simulation performed shows that, during the post-core blockage phase, conditions for counter-current flow limitation (CCFL) at the top of the core may occur. During this phase, vapor produced leaves the core by flowing upward through the core outlet, while liquid water (reaching the top of the core through the SGs tubes) moves downward toward the core. Conditions that affect the CCFL at the top of the core include the liquid and vapor velocities , the liquid and vapor properties , and the geometry. The figure below shows the CCFL integral actuation time. This parameter indicates if the conditions for counter-current flow limita tion occurs , and how long these conditions are maintained.

10000 Vi' UJ i== 1000 _, <i: c:: t!) UJ f--z z 100 0 ' ::> ' f--,. , , u 10 '* <i: : 1 _, ,. u... ,. -.. u I u I I I I I I l I 300 -* -coreOUtlet --Sump SW1tchover

-*-* -*-* -*-* . -. -. -. ,., . ,

• J , .' 5300 10300 15300 SIMULATION TIME [SJ NOC-AE-16003395 Attachment 1 Page 14 of 65 ----Blockage nme --*-* -*-*--*-*-* 20300 25300 30300 Core Outlet CCFL Actuation I nt egral Time (y-axis log scale) Core Heat Transfer Regimes during Pre-and Pos t-Bloc k age Phases This section describes the hea t transfer regimes establ i s h e d d u r ing t he pre-and pos t-core blockage phases of the transient.

The heat transfer regime for t he av e rage assem bl y , hot as s emb ly a nd hottest rod are shown in the figures below.

--HTmode6052

--HT mode 605 16 so Vi' Vl .. !:::: z 46 ::::> a: 44 w co --HT mode 605 4 --MTmode60518

--HT mode 605 6 --HT mode 605 8 --HT mode 60511 --HT mode 605 20 ----Sump --Blockageli m e 42 -----,,---+-::::> z w 40 Cl 0 38 a: w 36 u.. Vl z 34 <t: a: ..... 32 w ::r: 30 1000 1500 2000 Si111lo-Pho1e Liqu i d Conv*ction o r Subcooled w*ll with Void F r oction < 0.1) 2500 3000 3500 4000 SIMULATION TIME [SJ 4500 Average Assembly Heat Transfer Regimes 2 5000 2 Heat transfer regimes table available in RELAP5-3D u ser's manual , volume IV , Section 4.2.1 NOC-AE-16003395 Attachment 1 Page 15 of 65 --HTmode60514

.... 5500 6000 so Vi' V'l LU 48 .....J t:: z 46 ::::> ex: 44 LU co 42 ::::> z LU 40 0 0 38 ex: LU 36 LL V'l z <( l4 ex: I-l2 LU :I: 30 1000 so Vi' V'l LU 48 .....J t:: z 46 ::::> ex: LU 44 co 42 ::::> z LU 40 0 0 38 ex: LU 36 LL V'l z 34 <( ex: I-32 LU :I: 30 1000 NOC-AE-1600339 5 Attachmen t 1 Page 1 6 of 65 --HT mode 6060 2 --H T mode 6060 4 --HT mode 6060 6 --H T mode 6060 8 --HT mode 6060 11 --H T mode 6060 14 --HTmode6060 16 --HTmode6060 18 --HTmode6060 20----Sump Switchover --Blockage Time 1500 2000 Sincle-Ph*H U q uld Co nvection 0< Subcooled wa ll with Vold F roctl o n < 0.11 2500 3000 3500 Subcooled Nucl*at* Bo ilin& Saturated NudHto Bolllnc 4000 4500 SIMULATION TIME [SJ Hot Assembly Heat T ra n s f er Regimes sooo -----*Sump Switchove r ---Blockage Time --HTmode6061-02

--HTmode6061-04

--HTmode6061-06--H T mode606I-OS

--HTmode6061-ll --HTmode6061-14 HT mode 6061-16 HTmode6061-18 HT mode 6061-20 1500 2000 Sincle-PhHo U q u l d Con v*ct i on or Subcooled Wllll wi t h Voi d FrllCtlon

< 0.11 2500 3000 3 5 00 Subcoo*od Nuc!tiat o l!lollln c Saturated N u c l Hto B ol li nc 4000 4500 S I MULATION TIME [SJ Hottest Rod Heat Transfe r Regimes sooo 5500 6000 5500 6000 As c an be s een, t he heat t ransfer in the core before th e SSO ti me i s de t e r mine d by the la r ge amoun t of subcooled ECCS flow for c ed through the co r e dur in g this phase. As the temperature NOC-AE-16003395 Attachment 1 Page 17 of 65 of the ECCS injected water increases, the heat transfer regime transitions from single phase to subcooled and saturated nucleate boiling. This regime is also maintained during the blockage phase. Other Thermal-Hydraulic Parameters of Interest The following plots show the behavior of other thermal-hydraulic parameters of interest during the transient.

The following plots are included:

• Primary Pressure

• Core Decay Power and Boil off rate

• Break and ECCS integral flows

• Core Inlet/Outlet flow rates

• Core Bypass Inlet flow rate

• Total SG Heat Transfer -PnmPressure

---

SumpSwitchover --BlockageTime 320 270 Vl a._ UJ a: 170 :J Vl Vl UJ a:: 120 70 20 300 !:>JO 2300 3300 4:>JO 5300 6300 SIMULATION TIME [SJ Primary Pressure '"'° 8300 9300 2.000£.02 l.SOOEt-02 1.600£+02 1 400£+-02 l.200E+02 0:: UJ 1.000£+02 0 8.000Et-0 1 a.. 6.000Et-0 1 4.000Et-01 2.000£+0 1 0.000£.00 1 000 200 111) 160 Vi' -co .=., 140 UJ 1 20 0:: u... u... 0 100 __J 0 co 80 60 40 1000 --To l a/CorePa.ver

---

Sump Swi t chover --SlockaseT i me 2000 3000 4000 SIMULATION TIME [SJ Core Decay Power --Core Power per La t ent Heat ---*SlATlp Switchover --B l ockage T ime 2000 3000 4000 SIMULATION TIME [SJ Core Boil Off Rate 3 5000 5000 NOC-AE-16003395 Attachment 1 Page 18 of 65 6000 6000 3 The boil off rate _i s calcu l ated by dividing the core de cay power by the latent heat of evaporation of water at 150 psia.

8000000 7000000 6000000 V'l 4000000 V'l <( 3000000 2000000 1000000 9000 7000 Vi' -co 5000 __J LU a: 3000 0 __J 1000 u.. V'l V'l <( -nxl -5000 -integral Total Break Row -lnt@gralTot.al ECCSFlow ----S1.mp Switchover

--

Time 1000 2000 3000 4000 SIMULATION TIM E [SJ ECCS and Break Integ r al Flow --CorelnletFlow

-CoreOotletFlow

---

SumpSw1tchowr

--BlockageTime SIMULATION TIME [SJ Core Inlet and Outlet Flow Rate NOC-AE-16003395 Attachment 1 Page 19 of 65 5000 6000 2000 1 500 1 000 Vl -... co _, -BypasslnletFJow

---

Su'TipSw1tchover --Bhxkagelime NOC-AE-16003395 Attachment 1 Page 20 of 65 I

___ 5_000 ___ 6000--Vl lf H I Vl <t: ::1! -1 000 -1 500 -2000 £ U:XX-0 1 0 3 '° 0.00 h OO -1.00E*Ol 3 (,9 -2.00E-01 Vl 0 f--3.00E-01 UJ I -4.00E-01

-5.00E*Dl -6.00E-01 1 000 1 500 SIMULATION TIME [SJ Core Bypass Inlet Flow Rate -Heat to SGs_secondary

---*Sump Switchover --Blockage Time 1000 2500 3000 l 500 4000 4500 SIMULATION TIME [SJ Total SG Primary to Secondary Heat Transfer4 References

[1]. RC09989 RELAP5-3D Steady-State Model Rev.O 4 Positive when transferring from primary side to secondary side. 5000 5500 6000 SNPB-3-6 NOC-AE-16003395 Attachment 1 Page 21 of65 Initial and Boundary Conditions for each Accident Scenario Please demonstrate that the initial and boundary conditions for each accident scenario are appropriate for the given simulation.

This demonstration should focus on the simulations performed under the 10 CFR Part 50 Appendix B Quality Assurance Program. Provide a discussion of the confirmation of the initial and boundary conditions, and describe how these conditions reflect the conditions in the plant. Provide a discussion on the treatment of uncertainties.

Provide appropriate references.

If this demonstration relies on comparisons with results from other computer codes, please provide (1) a description of the code, (2) confirmation that the code has been approved by the NRG, (3) a summary of the simulations the code has been approved to analyze, and (4) an analysis addressing each initial and boundary condition, and how a deviation in that condition would be reflected in the code comparison.

Please confirm that the steady state simulation is consistent with plant operation (e.g., pressure drop around the loop). Confirm that important system parameters are being applied with their TS values or values assumed in the UFSAR as appropriate (e.g., flow rates, temperatures).

Criterion

1.4 Reference

SRP, lll.3c STP Response:

The proposed L TCC EM consists of four RELAP5-3D input models (the steady-state input model, and the input models for the 16", 6" and 2" HLB LOCA scenarios) which are used to perform the simulations.

The simulations are performed under the STP 1 b CFR Part 50 Appendix 8 quality assurance program. The initial and boundary conditions are demonstrated to be appropriate for the simulations of the LOCA scenarios included in the L TCC EM. Discussion of the confirmation of the initial and boundary conditions is provided below for each of the four input models included in the L TCC EM. Steady-State Input Model The steady-state input model defines the initial conditions for all the LOCA simulations included in the L TCC EM. The steady-state model is prepared under the STP Appendix B preparation of calculations guidelines

[1 ], and documented in [2]. The steady state model defines the plant geometry and the model options selected.

The initial conditions are set for the 100% power operating condition at normal operating temperature and normal operating pressure; and applied to all LOCA calculations.

All the plant geometrical and operating conditions included in the model are documented in [2]. Assumptions, sources of information and margins are also described in the documentation.

NOC-AE-16003395 Attachment 1 Page 22 of65 The validation of the steady-state model is described detail in [2]. In general, the validation is based on direct comparison with plant data (when available) to show conformance with actual plant conditions.

In other cases, (where plant data are unavailable) the steady state output from the approved STP RETRAN model [3,4] are used to ensure fidelity to the design. The validity of initial conditions definitions with applicable references is confirmed with any deviations described in the STPNOC Engineering Calculation RC09989 Rev. 0 [2]. All deviations noted are deemed to be insignificant to the calculation results. Model uncertainties are identified and ranked based on their importance in respect to the LTCC and post-core blockage impact on the PCT. A discussion of uncertainties identified, the importance associated, and the margins included in the L TCC EM is included in the responses to RAI SNPB 3-32, RAI SNPB 3-21, and RAI SNPB 3-22. The steady state simulation is consistent with the STP plant operating conditions.

The important system thermal-hydraulic parameters are set to STP technical specification values, or from other appropriate documentation.

LOCA Input Models Three LOCA scenarios are included in the L TCC EM. The scenarios use the same base input file with the sole difference in the break size (16", 6" and 2"). The model includes all the phases of the accident, in particular:

Phase 1: Slowdown Phase 2: Refill/Reflood Phase 3: Pre-Core Blockage L TCC Phase 4: Post-Core Blockage L TCC The model includes adequate details of the plant characteristics, and plarit LOCA emergency operations (manual and automatic) to simulate the phases of the accident listed above. These characteristics are simulated with the use of a set of boundary conditions summarized below.

Parameter Value RWST Usable Volume 360000 gal --Core Blockage Timing 6min (from SSO) Low PZR pressure Set 1871.7 Point psi a Reactor Trip Signal 2s Processing Rod droe time 2.8 s RCP trip: RCS 1444.7 Pressure Check psi a MFW isolation Delay 2s MFW closure time 10 s AFW mass flow rate 70.0 Ibis 50 psia, 123.4°F-AFW conditions Specific Enthalpy = 91.5 Btu/bl Parameter HHSI Actuation Delay LHSI Actuation Delay Hot Leg Injection time (from break) AFW Delay (from reactor trie) RWST water conditions Sump Pool Temperature (at SSO) Accumulator Liquid Volume Accumulator Water Conditions Containment sprays volumetric flow rate NOC-AE-16003395 Attachment 1 Page23 of65 Value 6s 10 s 5.5 h 30s 130 °F 14.7psia 270°F 1200 ft3 600 psia and 120 OF --4282 gal/min A discussion of uncertainties identified for the most critical boundary conditions listed above, the importance associated, and the margins included in the L TCC EM is included in the responses to RAI SNPB 3-32, RAI SNPB 3-21, and RAI SNPB 3-22. Initial conditions for the L TCC (Phase 3 and 4)' are included as part of the simulation of the preceding phases 1 and 2. Since no simulations or plant data are available for comparison with the proposed LOCA scenarios included in the EM, a direct comparison of the simulation results for the accident scenario (Phase 1 and 2) cannot be performed.

The simulation results are analyzed to verify the correctness of the predictions and engineering judgment is used to evaluate the results.

References NOC-AE-16003395 Attachment 1 Page24 of65 [1]. South Texas Project Electric Generating Station. Preparation of Calculations.

OPGP04-ZA-0307, STI 34177871.

Rev.7 [2] RC09989 RELAP5-3D Steady-State Model Rev.O [3]. Calculation NC-07087, Rev. 0, "Mass and Energy Release for Main Steamline Break Inside Containment. (STI: 32464076).

[4]. "RETRAN-30" -A program for transient thermal-hydraulic analysis of complex fluid flow systems. Vol.1. NP-7450,, Version 3, Electric Power Research Institute (1998).

SNPB-3-7 Attachment 1 Page 25 of65 Initial and Boundary Conditions for the Long-Term Phase Please demonstrate that the initial and boundary conditions for each accident scenario at the beginning of the long-tenn phase are consistent with those conditions which are expected.

This demonstration should analyze the RELAP5-3D calculations for the conditions at the beginning of the reflood stage, and show that those calculations are reasonable compared with known behavior.

This analysis should include a comparison between the conditions calculated by RELAP5-3D and the current large and small break LOCA safety analyses . Criterion

1.4 Reference

SRP, lll.3c STP Response:

The initial and boundary conditions applied to the L TCC EM are demonstrated to be appropriate for the simulations of L TCC phases of the LOCA scenarios included in the EM, and are consistent with the conditions expected during this phase. Discussion on the adequacy of the initial and boundary conditions for each accident scenario adopted in the EM is included in the response to RAl-SNBP-3-06.

All the calculations performed are prepared under the STP Appendix B preparation of calculations guidelines

[1], and properly documented.

All the plant geometrical and operating conditions included in the model are documented.

Assumptions, sources of information and margins are also described in the documentation included for each LOCA scenario.

Model uncertainties are identified and ranked based on their importance in respect to the L TCC and post-core blockage impact on the PCT. A discussion of uncertainties identified, the importance associated, and the margins included in the L TCC EM is included in the responses to RAI SNPB 3-32, RAI SNPB 3-21, and RAI SNPB 3-22. As described in the response to RAl-SNPB-3-06, the model includes all the phases of the accident:

Phase 1: Slowdown Phase 2: Refill/Reflood Phase 3: Pre-Core Blockage L TCC Phase 4: Post-Core Blockage L TCC The model includes adequate details of the plant characteristics, and plant LOCA emergency operations (manual and automatic) to simulate all the phases of the accident listed above, including the pre-core blockage and post-core blockage L TCC phases. Initial conditions for the L TCC (Phase 3 and 4) are included as part of the simulation of the preceding phases 1 and 2.

NOC-AE-16003395 Attachment 1 Page 26 of65 Since no simulations or plant data are available for comparison with the proposed LOCA scenarios included in the EM, a direct comparison of the simulation results defining the initial conditions for the L TCC phases cannot be performed.

The simulation results are analyzed to verify the correctness of the predictions and engineering judgment is used to evaluate the results. References

[1]. South Texas Project Electric Generating Station. Preparation of Calculations.

OPGP04-ZA-0307, STI 34177871.

Rev.7 SNPB-3-15 Level of Detail NOC-AE-16003395 Attachment 1 Page27 of65 Please confirm that the level of detail (e.g., phenomena modeled, initial and boundary conditions, overall assumptions) is consistent between STP's LOCA licensing basis analysis and the simulations performed in the L TCC EM. Criterion

3.6 Reference

SRP, lll.3b STP Response:

The level of detail of the L TCC EM, including the phenomena modeled, the initial and boundary conditions, and the overall assumptions, are consistent with STP licensing basis analyses that use conservative boundary and initial conditions where they have been shown to be important for the L TCC; or by performing sensitivity analyses (see for example, SNPB RAI 3-22) to ensure that adopted values are bounded. In addition, the level of detail included in the L TCC EM is consistent with the best practice used to simulate LOCA scenarios with RELAPS-30.

The proposed L TCC EM consists of one steady-state model and three models to simulate LOCA scenarios of different break size. The steady state model defines the plant geometry and the model options selected.

An adequate level of detail is included in the steady-state model to represents all the regions of the primary system, including reactor vessel and internals, reactor core, coolant loops with RCPs and SGs. Models are included in the EM to simulate the important phenomena of the plant during all phases of the accident progression:

I! Plant operating conditions (initial conditions for the transient)

• Break opening and blowdown phase

• Refill/Reflood Phase

• Pre-Core Blockage L TCC

• Post-Core Blockage L TCC The initial conditions are set for the 100% power operating condition at normal operating temperature and normal operating pressure; and applied to all LOCA calculations.

All the plant geometrical and operating conditions included in the model are documented in [1]. Assumptions, sources of information and margins are also described in the steady-state documentation.

Three LOCA scenarios are included in the L TCC EM. The scenarios use the same base input file with the sole difference in the break size (16", 6" and 2"). The model includes all the phases of the accident listed above.

NOC-AE-16003395 Attachment 1 Page 28 of65 The model includes adequate details of the plant characteristics to simulate all the phases of the accident progression.

This includes:

• Plant LOCA main emergency manual operations (EOPs)

• Plant LOCA automatic operations

• Plant set points

• Delays The level of detail included in the EM is consistent with the STP licensing basis analyses and with the RELAP5-3D LOCA analysis best practices.

References

[1] RC09989 RELAP5-3D Steady-State Model Rev.O SNPB-3-17 Validation of the Evaluation Model NOC-AE-16003395 . Attachment 1 Page29 of65 Please provide appropriate validation demonstrating that the L TCC EM will result in a reasonable prediction of the important figures of merit for the accident scenarios considered.

Demonstrate that the validation covers the range of the accident scenarios used in the L TCC EM. This validation should include comparisons to 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 4.2, 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 EM results in reasonable predictions of the PCT for the range of HLB scenarios considered in the L TCC EM. The PCT response to the heat transfer regimes and flow conditions experienced during L TCC in these scenarios is validated by integral and separate effects tests as described in RAI SNPB 3-13. The validation of the EM is conducted in order to verify the adequacy of the parameters defined in the input models (plant geometry, materials properties, initial conditions, boundary conditions, assumptions, margins, set points, delays), and to confirm that the model provide reasonable predictions of the plant response during normal operation (steady-state) and during all the phases of the accident progression.

No plant data or experimental results are accessible to compare with the predicted accident progression, and, in particular, for the specific LOCA scenarios (hot leg breaks) analyzed.

Subsequently, the validation of the EM is based on: 1. Verification of the adequacy of the parameters in use in the EM 2. Verification of input models 3. Judgement of the simulation results Verification of the Adequacy of the Parameters in use in the EM Thermal-hydraulic parameters used to prepare the RELAP5-3D input files are retrieved from approved STP sources. All sources used are referenced, listed and properly documented, as per STP Appendix B guidelines.

These sources are reviewed to confirm they are properly consistent with the purpose of the simulations.

Important sources of uncertainty are identified, ranked based on the potential impact on important parameters (PCT), and, when possible, adequate conservatism is included in the EM, or dedicated sensitivity studies are conducted.

The parameters verified through this process are selected and implemented in the RELAP5-3D input files. Verification of the Input Models The RELAP5-3D input models are subject to review to verify that the selected plant parameters and other conditions/setting are correctly implemented in the input models. The review process is conducted in accordance to the STP Appendix B guidelines

[1]. Independent internal and external reviewers (One or more originators, one or more checkers) are identified and adequate verification is conducted.

NOC-AE-16003395 Attachment 1 Page 30 of65 The simulation results are reviewed in detail during the process to verify adequacy of the parameters implemented in the EM. Judgement of the Simulation Results The simulation results are analyzed in detail for all the phases of the accident progression.

Plant operating conditions, used as initial conditions of the LOCA scenarios, are simulated with the steady-state input model. As stated in the answer to RAl-SNPB-3-06, the validation of the steady-state results is based on direct comparison with plant data (when available) to show conformance with actual plant conditions.

When plant data are unavailable, the steady state output from the approved STP RETRAN model is used to ensure fidelity to the design. The simulation results of the accident procession cover all the phases of break opening (LOCA initiation), blowdown, refill, reflood, and LTCC (pre-and post-core blockage).

As stated in the response to RAl-SNPB-29, the simulation results are accurately verified by the use time tables, plots, and other information available in the output files. These include all the main thermal-hydraulic parameters of the primary system over the entire duration of the transient simulation.

The results are analyzed to verify the correctness of the simulation predictions.

Different parameters are generally plotted and combined iil the same figures to confirm that the predictions are in reasonable agreement with the expected behavior.

The analysis of the results is supported by the use of additional control variables defined in the input files, and other specialized cards to allow the monitoring of other hydraulic parameters of interest not directly available in the output files. Engineering judgment is performed.

Additional details on the input models verification and other techniques adopted to verify the adequacy of the EM and the produced simulation results are included in the responses to:

• RAl-SNPB-3-24

-Input Verification

• RAl-SNPB-3-25

-Proper Convergence

• RAl-SNPB-3-26

-Non-physical Results

• RAl-SNPB-3-27

-Realistic Results

• RAl-SNPB-3-28

-Boundary conditions as prescribed

• RAl-SNPB-3-29

-Thoroughly understood results

• RAl-SNPB-3-31

-Independent Peer Review References

[1]. South Texas Project Electric Generating Station. Preparation of Calculations.

OPGP04-ZA-0307, STI 34177871.

Rev.7 SNPB-3-18 Mesh Size Sensitivity NOC-AE-16003395 . Attachment 1 Page 31 of65 Please demonstrate that the L TCC results are independent of mesh size for the accident scenarios under consideration.

Criterion

4.7 Reference

SRP, lll.3d STP Response:

The sensitivity analysis performed with the L TCC EM on the mesh size and nodalization of selected regions of the primary system demonstrates that the L TCC result (PCT) is not dependent on the mesh size and nodalization adopted. The nodalization diagram adopted for the RELAP5-3D model is prepared in compliance to the general LOCA analysis guidelines described in Volume V of the RELAP5-3D user's manual [1]. In particular:

• The size of the nodes is defined such that the ratio of the length and hydraulic diameter is approximately equal to unity or greater than one, to allow spatial convergence and maintain the applicability of constitutive models

• The total number of nodes in the system is also optimized against the run time 11 When possible, the regions are subdivided in approximately equally sized nodes, with a relatively finer nodalization adopted only selected regions of the system The reactor system is subdivided into regions reflecting the RELAP5-3D specific guidelines for LOCA simulations of typical Westinghouse 4-loop PWR. The proposed nodalization is based on the described in Volume V Section 5.1 [1]. The model includes four independent coolant loops with hot leg, cold leg, intermediate leg, and steam generators.

The nodalization adopted for these regions also follows the guidelines provided by the RELAP5-3D user's manual. In particular, for: a The size of the nodes adopted to simulate the legs, and specifically the number of nodes to simulate the intermediate leg.

• The size of the nodes and total number of nodes of the primary side steam generator's u-tube.

• The nodalization of the secondary side of the steam generators.

Based on the L TCC important phenomena identified and described in the answer to SNPB-3-04, and the accident scenario progression described in the answer to RAl-SNPB-3-02, node size sensitivity study is exclusively performed for the reactor core to demonstrate that the pre-and post-blockage L TCC results are independent of the mesh size adopted for this region. The 16" break in hot leg LOCA scenario is selected for this sensitivity.

This case is shown to exhibit the same phenomenology but under more severe conditions than smaller breaks.

NOC-AE-16003395 Attachment 1 Page 32 of65 During the post-blockage L TCC phase, liquid water reached to the top of the core through the hot legs. Vapor produced in the core may reach sufficient axial upward velocity to establish conditions for CCFL which may temporary prevent liquid from entering the core. It may be also expected that the vapor generation and, subsequently, its axial upward velocity is larger in location of the core with higher power sharing (hot assembly).

Under these conditions, liquid water may preferentially enter the core through colder assemblies and subsequently reach the hot assemblies at lower elevations.

Recirculation patterns may be expected between cold and hot assemblies and, if sufficient amount of liquid water enters the core, the core coolability is maintained.

The proposed sensitivity study is designed to enhance the counter-'current flow limiting conditions at the hot channel by:

• Simulating the core with two independent flow channels representing the average channel and the hot channel respectively, to compare with the base nodalization where all fuel channels are lumped together into one single vertical pipe component.

• Associating the average heat structure to the average flow channel, and the hot structures to the hot channel, to compare with the base case where all heat structures are connected to the single pipe component

• Removing the cross flow between average and hot channels to enhance the CCFL at the exit of each channel, and, in particular, at the hot channel. The nodalizations used for this sensitivity are shown in the figure below. The single core channel model (base case, left) is thermally connected to three different heat structures:

1. The hottest rod heat structure

2. The hot assembly heat structure

3. The average assembly heat structure.

The two-channel core nodalization (right) includes:

1. The hot channel, thermally connected to the hottest rod and to the hot assembly heat structures.

2. The average channel, thermally connected to the average heat structure.

Both nodalization diagrams use 2.1 axial nodes to allow:

• Sharp gradients to be resolved during the L TCC phase in core blockage scenarios

• Fine axial power shapes be simulated.

X 19 2 512 521 2 3 4 585 513 590 595 3 865 53 5 r-2 512 521 2 3 4 513 585 696 NOC-AE-16003395 Attachment 1 Page 33 of 65 7? 5 5 35 Nodalization Diagrams.

One-C hannel Core (Base Case, Left); Two-Channel Core (Right) The cases are executed under the same boun d ary conditions.

The simulation results in terms of PCT are shown below. No appreciable difference is observed between the two nodalization approaches adop t e d.

800 700 600 L.IJ 500 a:: :J --One-Channel Core {Base Case}

Core t:( 400 I ---*Sump Switchover*

One-Cha me I Core ---*Sump Si.vrtd*1owr-T\".o-Chann@I Core a:: I !** 200 : :::::::;::soc I I I I I I 100 I 1000 1500 References I I I I I 2000 2500 3000 3500 SIMULATION TIME [SJ Simulation Results: PCT [1]. RELAP5-3D Code Manual -V olume V: User's G ui delines NOC-AE-16003395 Attachment 1 Page 34 of 65 --Blockage Time. One-Channet Core --Blockage Time -Tw<H:hannel Core 4500 sooo 5500 6000 NOC-AE-16003395 Attachment 1 Page 35 of 65 SNPB-3-20 Specific Sensitivity Studies During the audit , the NRG staff identified a number of sensitivity studies that would be important f or the NRG staff review of the proposed L TCC evaluation methodology.

STP is requested to perform the following sensitivity studies and submit plots of the relevant figures of merit and important timings for L TCC analysis:

a) b) c) d) Criterion Appendix K decay heat load with single worst failure and steam generator tube plugging Axial power shape Break sensitivity study with appropriate break size resolution No bypass blockage 4.7 Reference SRP , l/l.3d STP Response:

Sensitivity studies are conducted using the proposed L TCC EM to analyze the system response under different conditions.

These sensitivities are generated from the 16" break LOCA scenario (base case) which is desc r ibed in the response to RAl-SNPB-02.

The scope of these sensitivities is to study the var i ability of the PCT against the following thermal-hydraulic parameters

• Decay power

• SI trains availab i lity

• Steam Generators

' tube plugging

• Core axial power shape

• Core bypass availability at the SSO Three sets of sensitivities are conducted and desc rib ed below: SENSITIVITY a) Appendix K decay heat l o ad with single w or st failure and steam generator tube plugging. This case is generated from the base case and combines the following:

• Augmented decay power (+20%) from the ANS-79 model used in the base case

• Single train failure (1 HPSI + 1 LPSI + 1 CS). The failure is assumed to occur at the start of the transient.

I n order to minimize the total ECCS flow available for cooling , the unavailable SI train is assumed to be in one of the intact loops (loop 2).

• SG tube plugging equal to 10% (maximum allowed by STP design). SENSITIVITY b) Axial power shape. Two simulations are included in this sensitivity study. These simulations are originated from the base case which uses an axial cosine power shape [1], and modified as follow: b1) Bottom skewed power shape b2) Top skewed power shape SENSITIVITY d) No bypass blockage.

NOC-AE-16003395 Attachment 1 Page 36 of 65 This sensitivity is performed to show the effec t iveness of the core barrel/baffle bypass as alternative flow path during a hypothetical full core blockage at the bottom of the core. Th i s case is executed starting from similar conditions described in (a}, assuming a free core bypass during t he post-core blockage phase. The L TCC EM includes LOCA scenarios with full core and co r e bypass blockage of different break sizes (16", 6" and 2" break). The analysis of the scenarios has found similar behavior of the primary system. Similar phenomenology and accident progression described in the response to RAl-SNPB-02 is observed for other break sizes. For this reason , additional sensitivities on the break size resolution are not performed.

Boundary Conditions T h e main boundary conditions used for the proposed sensitivity study are listed below. The Sump Pool Temperatur e is assumed to be the RCS injection t emperatu r e. The maximum sump temperature val u es (270° F}, rep r esent conservative cases where the RHR HX is not modeled. The 190° F cases assume RHR HX is avai l able. C ond iti o n Bas e C ase a) b l) b2) d) Decay Powe r(%) 1 00 120 100 100 120 S I Tra in s 3 2 3 3 2 SG Tub e P l ugg in g(%) 0 10 0 0 10 A xi a l Powe r S hap e C h opped Cos in e C h opped Cos in e Bottom Skewed Top Skewed Ch o pped Cos in e Bypass B l oc kage(%) 10 0 100 100 100 0 RWST Vo lum e (Ga l) 360.000 453 , 000 360.000 360 , 000 360 , 000 RWST Temperature

(" F) 130 85 130 130 130 ECCS Inj ect i o n T e m pe r atu r e at SS O ("F) 2 7 0 1 90 270 2 70 270 Figure below shows t he a xi al pow er p r of i le s u se d to r un t he sensit i vi t ies b1) and b2). 0.08 0.07 0.06 c 0.05 I! I&. 0.04 .. Cl l 0.03 0.()2 O.Ql 0 0 .5 10 15 20 25 Axial Node -+-bottom skewettom skewed _.,_cosine skewe<I Steam Generators Tube Plugging (Importance

-Medium) NOC-AE-16003395 Attachment 1 Page 45 of 65 IS The base simulation of the EM assumes 0% steam generators' tube plugging.

A sensitivity analysis is included to evaluate the effect o f higher tube plugging (10%) Vessel Flow Bypass Fractions (Importance

-High) Due to importance of this source of uncertainty, adequcite margin is included in the proposed LTCC EM. In particu l ar, the EM minimizes the availability of the alternative flow paths that may allow cooling wa t er to reach the core in the event of a core blockage at the bottom of the core. 1) Leakage flow from the down comer inlet nozzle directly to the vessel outlet nozzle through the gap between the vessel and the barrel are not accounted in the EM. Possible cooling water leaking from the downcomer inlet nozzle di r ectly to the vessel outlet region is not accounted in the EM. 2) Flow entering into the rod cluster control guide thimbles is not independently modeled but instead lumped into the core bypass channel s i mulated with the pipe component 551, which is assumed to be blocked after the sump switchover time. 3) Flow in the gaps between the fuel assemblies on the core per i phery and the adjacent baffle wall is not independently modeled but instead lumped into the core bypass channel simulated with the pipe component 551 , which is assumed to be blocked after the sump switchover time. 4) Flow introduced between the baffle and the barrel (modeled with the pipe component 551) is assumed to be blocked after the sump switchover time.

X19 Core Nodalization (Importance

-Medium/High)

NOC-AE-16003395 Attachment 1 Page 46 of 65 The proposed EM simulates the reactor core with a one-dimensional vertical pipe component and three heat structures representing the average assembly , the hot assembly , and the hottest rod. Sensitivity in the core nodalization is performed to account for the uncertainty in the nodal i zation. This considers a nodalization of the core with two vertical pipes representing the average channel and hot channel. Upper Head Nodalization (Importance

-Medium) The upper plenum sprays and the flow paths between the upper head and the top of the core represent alternative flow paths through which the ECCS cooling water , injected into the cold legs, may reach the top of the core in the event of a core blockage. The nodalization of the reactor vessel upper head region is conceived to include more details of the flow paths between the upper plenum sprays and the top of the core, compared to the nodalization generally adopted for LOCA trans i ents' s i mulations (1 , 2). Guide tubes and the volumes surrounding guide tubes are also included in the EM. The nodalization adopted prevents water from flowing downward the guide tubes until the surrounding volumes are full. 585 58 5 3 2 5 1 2 590 2 590 2 51 2 595 3 i X2 1 501 X19 X21 865 .. 501 865 Standard PWR Upper Head Nodalization L TCC EM Adopted Nodalization NOC-AE-16003395 Attachment 1 Page 47 of65 RWST Usable Volume (Importance

-High)The RWST water volumes are identified in the figure below. The proposed L TCC EM assumes a RWST usable water volume equal to 360 , 000 gallons. Volu me 4 3.1-+.-==:: '---

550,000 gal -.--.------1 (22 , 00 0 ga l) I V olume ab ov e H i gh A larm I 4 0.9'_........_ __ HI Alarm ----+-528 , 000 gal -+---+------<

(55 , 000 g.iil) 37.6_........_ __ lo Alarm ----+-473,000 gal -+---t-----< Work i ng A llowance RWST T o t al V ol ume (5 50 , 000 gal) Usable volume (398,0 0 0 g a l) I Inj ec tion Vo lume s Ou1l et n ozz le Initiation of switchover Vortex lo-lo Alarm br ea k e r 1 3' --(43 , 000 ga l) T r a nsfer A llo w ance 32 , 000 gal -+--+------<

(3 2 , 0 GO g a l) I Volume a t Empty A larm I --+-------;<----' 0 ga l _ _.___,, ___ _, inal Values Including instrurMnt uncerta i nty Minimum requ i red value For the tab l e bel o w one c an estimate the mar gi n t o t h e m inim um usable vo l ume included in the EM. This marg i n i s equal t o 81 , 000 gallons. RWST Volum e s Vo l ume (gal) Hi Alarm (Nominal) 528 , 000 Lo Alarm (Nominal) 473,000 lo-lo Alarm (Nominal) 75,000 Max Usable volume 496 , 000 Min Usable volume 44 1 ,000 Average Usable Volume 468,500 Usable Volume (LOCA) 456,735 Injection Volume 413 , 735 (LOCA)

Decay Power Model (Importance

-High) NOC-AE-16003395 Attachment 1 Page48 of65 The ANS79 decay heat model is used in the proposed L TCC EM. Sensitivity analysis is included which uses the Appendix K decay power requirements of +20%. Break Size (Importance

-High) The L TCC EM includes scenarios of different break size and investigates:

• A large break of 16-inch diameter

• A 6-inch diameter break, representative of medium breaks

• A 2-inch diameter break, representative of small breaks Break Orientation (Importance

-Low) While this parameter is not expected to have a direct effect on the PCT behavior during the L TCC and core blockage phases, sensitivity analysis is performed to consider different break orientations.

ECCS Flow Rate (Importance

-Medium) The ECCS flow rate has a similar impact to what described in (9). The sump switchover time and, subsequently, the time at which the core is assumed to be blocked in the EM, is strictly related to the injected flow rate. A larger ECCS flow rate results in an earlier core blockage time and, subsequently, to a higher decay heat generated in the core at the time of core blockage.

This source of uncertainty is expected to impact the cladding temperature after the core blockage time. ECCS Injection Temperature (Importance

-Medium/High)

The ECCS injection temperature is:

• The temperature of the water in the RWST during safety injection phase

• The temperature of the sump pool during the recirculation phase The RHR HX cooling effect on the ECCS flow is not modeled in the L TCC EM. The values implemented in the L TCC evaluation model are compared with the best estimate reference in the table below. Parameter LTCC EM Best Estimate RWST Water Temperature

(°F) 130 85 Sump Pool Temperature at SSO (°F) 270 188 (2)

Core Barrel/Baffle Bypass Blockage Fraction (Importance

-High) NOC-AE-16003395 Attachment 1 Page49 of65 In the proposed LTCC EM the core barrel/baffle bypass is assumed 100%

The blockage is assumed to occur instantaneously 360 seconds after the sump switchover time. This is simulated by closing the inlet valve component 456. Core Blockage Fraction (Importance

-High) In the proposed L TCC EM the core is assumed 100% blocked. The blockage is assumed to occur instantaneously 360 seconds after the sump switchover time. This is simulated by closing the inlet valve component 457. Plant Set Points and Delays (Importance

-Low) These parameters are set to their nominal value (based on the STP technical specifications) in the L TCC EM. CCFL Parameters (Importance

-Medium) Sensitivity study is included in the EM to account for any effect of the CCFL parameters on the PCT by minimizing the accessibility area. These parameters are identified by the vapor/gas intercept (word 3 of card 8451110) and slope (word 4 of card 8451110) References

[1]. STI 34280651 RELAP5-3D Software Quality Assurance.

Rev.a. RELAP5-3D User's Manual [2]. NUREG/CR-6770 LA-UR-015561, "GSl-191:

Thermal-Hydraulic Response of PWR Reactor Coolant System and Containments to Selected Accident Sequences", August 2012.

SNPB-3-23 NOC-AE-16003395 Attachment 1 Page 50 of65 Evaluation Model in an Appendix B Quality Assurance (QA) Program To address Generic Letter (GL) 2004-02, STP demonstrates its compliance with 10 CFR 50.46(b)(5)

Long term core cooling, including the impact of debris, using the following two step approach:

(1) The hot-leg large break, hot-leg medium break, hot-leg small break, and cold-leg small break will be demonstrated to be in compliance with 10 CFR 50.46(b)(5) by ensuring that the long-term core temperature does not exceed 800 degrees Fahrenheit

(°F) assuming a fully blocked core. This is demonstrated by using deterministic analysis performed with RELAP5-3D.

(2) The cold-leg large break and cold-leg medium break will rely on a risk-informed approach.

The hot-leg large break, hot-leg medium break, hot-leg small break, and cold-leg small break analyses are used to demonstrate compliance with 10 CFR 50.46(b)(5).

Therefore, certain design control measures are required, as specified in 10 CFR 50, Appendix B (Ill): Design control measures shall be applied to items such as the following:

reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of materials; accessibility for inservice inspection, maintenance, and repair; and delineation of acceptance criteria for inspections and tests. However, it is not apparent that the RELAP5-3D analysiS' was performed under a QA program satisfying the requirements of Appendix B. Please demonstrate that the RELAP5-3D analysis was performed under a QA program which satisfies the requirements of 1 O CFR 50, Appendix B, or provide a similar analysis that was performed under such a program. Criterion

6.1 Reference

SRP, lll.3f STP Response:

The RELAP5-3D analysis is governed by the STP procedure, OPGP03-ZA-0307 "Engineering Calculations" which is a quality procedure in compliance with the STP Appendix B quality assurance program (Operations Quality Assurance Plan, "OQAP"). OPGP03-ZA-0307 is required under the OQAP for STP engineering analyses supporting quality products and equipment.

As such, OPGP03-ZA-0307 procedurally requires meeting all pertinent elements of the STP Appendix B program including (but not limited to):

• Error reporting; OPGP03-ZX-0002 "Condition Reporting Process"

• Qualification (training) for applicable procedures; OPGP03-ZT-0136 "Engineering Support Personnel Training Program: OPGP04-ZA-0010; "Engineering Support Personnel Qualification"

• Software Quality Assurance for any software used in analyses; OPGP07-ZA-0014 "Software Quality Assurance Program"

. NOC-AE-16003395 Attachment 1 Page 51 of65

• Contractor and vendor qualification review; OPGP03-ZT-0138 "Contractor/Staff Augmentation Volunteer Qualification Program"

• Records Management and Control of Quality Records: OPGPO?-ZA-0001 "Records Management" and OPGP04-ZA-0328 "Engineering and Vendor Document Processing" The RELAP5-3D analysis is performed by qualified personnel in accordance with STPNOC procedure for Engineering Calculations, OPGP03-ZA-0307.

SNPB-3-24 Input Verification NOC-AE-16003395 Attachment 1 Page 52 of 65 Please provide details of how STP's QA program controls over the input deck for the L TCC EM. How are the input values verified?

What inputs are users given permission to change and how are such changes controlled?

Criterion

• 6.1 Reference SRP, lll.3f STP Response:

The input values are compared against the reference source (see response to SNPB-3-17) and controlled in the STP engineering calculation process required for all related work (Q). Each analysis must follow the STP Engineering Calculations procedure OPGP04-ZA-0307.

From time to time changes are required to engineering calculations due to new information, due to new regulatory constraints, and so forth. Any changes required to be made to a safety-related engineering calculation, require that the person follow the procedure (which details requirements for revision).

As part of the calculation procedure requirements, calculations that involve changes to the method of analysis described in the UFSAR must be evaluated to determine if prior NRC approval is required.

SNPB-3-25 Proper Convergence NOC-AE-16003395 Attachment 1 Page 53 of65 Please explain how the QA program ensures the code converged properly.

Such indicators commonly include nonphysical state properties and excessive mass error. Demonstrate that if the code did not converge numerically, the analysts would be alerted to the error messages and act appropriately.

Criterion

6.1 Reference

SRP, lll.3f STP Response:

The STP QAP does not specifically address non-convergence in a reactor safety code application.

All calculations performed using OPGP04-ZA-0307 require that both the performer and reviewer are qualified to perform safety-related calculations.

The analysts are able, based on experience, to recognize non-convergence.

When a RELAP5-3D solution is failing to converge, analysts will observe that the time step size decreases to a minimum and the solution progresses too slowly; if the minimum time step size is specified too large, a machine failure will be produced.

These kinds of failures are typical in all common reactor safety codes and are well-known to experienced analysts.

The most relevant check for proper step siz,e (primarily temporal) is the "mass error" tracking performed automatically in RELAP5-3D.

The mass error is a global check on the solution accuracy and is commonly addressed by making adjustments to limits on the adaptive time step routine through input. The maximum time step specified in the input file is generally reduced to follow the real time step required by the code. This is performed by running time step sensitivities to the scope of minimizing the mass error of the simulation.

Convergence is assured if the code obtains a solution; that is, referring to Lax's well-known theorem for convergence of finite difference formulations, the finite difference scheme converges when a solution is obtained, otherwise the scheme will fail. When RELAP5-3D fails to converge, the machine will stop with an error that is due to an impossible calculation (such as property look-up table out of data). Convergence and accuracy of the solution are different aspects that must be addressed by the analysts.

SNPB-3-26 Non-physical Results 16003395 Attachment 1 Page 54 of65 Please explain how the QA program ensures identification of non-realistic results such as liquid over vapor, unphysical oscillations that could be numerically induced, or any other nonphysical results that may lead to effoneous conclusions concerning the code's calculated thermal-hydraulic behavior.

Criterion

6.1 Reference

SRP, lll.3f STP Response:

The STP QA program does not specifically address each possible non-physical result that may be produced in a reactor safety code application.

All calculations performed using OPGP04-ZA-0307 require that both the performer and reviewer are qualified to perform safety-related calculations.

The analysts are able, based on experience with common reactor safety codes such as RELAP5-3D and knowledge of STP plant response, to understand when the code is producing non-physical results. If any non-physical results are found, the performer and checker will assure non-physical results are properly addressed (typically by making required corrections and re-running the code).

SNPB-3-27 Realistic Results NOC-AE-16003395 Attachment 1 Page 55 of65 Please explain how the QA program ensures the physical results are realistic.

Where the calculated flow regimes and heat transfer modes should be studied to ensure that the code is not assuming unrealistic conditions?

Criterion

6.1 Reference

SRP, lll.3f STP Response:

In general, the STP QA program ensures that the software used in safety-related engineering calculations complies with STP procedure OPGP04-ZA-0307 which also requires any software used to be qualified under the STP software quality assurance program, OPGP07-ZA-0014.

The SQA program itself ensures that the software is being used for the purpose intended and is validated in the domain of use. Finally, as mentioned in RAI SNPB-3-26 response, calculations performed under OPGP04-ZA-0307 are required to be performed by qualified personnel.

Any clearly physical results (assuming conservative assumptions) produced by a computer code used in a safety-related calculation are screened by the performer and reviewer (SNPB-3-2 response discusses a review and screening process).

If any non-physical results are found, the code is rerun with appropriate inputs to remove these artifacts.

SNPB-3-28 Boundary Conditions as Prescribed Attachment 1 Page 56 of 65 Please explain how the QA program ensures that the boundary conditions are occurring as prescribed.

Boundary conditions and others that control the direction of the transient (e.g., valves opening, pumps beginning to coast down, or heater rod power turning off) should be checked by the user to ensure expected performance.

Criterion

6.1 Reference

SRP, lll.3f STP Response:

The STP QA program requires that any software used in a safety-related application must meet the requirements of the STP SQA program. Under this program, the software is checked to be capable of accepting the inputs used for boundary conditions and applying them properly.

Additionally, simulations are performed using the STP engineering procedure, OPGP04-ZA-0307 by qualified personnel.

The simulation is described in the documentation such that an independent qualified reviewer can understand and evaluate the method and results. Thus, the boundary condition definitions are reviewed by at least two qualified personnel.

SNPB-3-29 Thoroughly Understood Results NOC-AE-16003395 Attachment 1 Page 57 of65 Please explain how the QA program ensures that every aspect of the calculation is thoroughly understood.

The depressurization rate, various indications of core heatup, drain rate of the system at various locations, liquid holdup, indications of condensation or evaporation, transition from subcooled to two-phase break flow, and other conditions should all be explainable.

Also, the results of the user's calculation should be understood from the perspective of previous calculations done on the same or similar facilities.

Criterion

6.1 Reference

SRP, lll.3f STP Response:

The STP QA program requires that the person who prepares the simulation develop a package for review in sufficient detail such that the reviewer is able to independently understand and reproduce the calculation.

The simulation output includes time tables, plots, and other information.

All the main thermal-hydraulic parameters of the primary system over the entire duration of the transient simulation are included.

These parameters are analyzed to verify the correctness of the simulation results. Different parameters are generally plotted and combined in the same figures to confirm that the predictions are in reasonable agreement with expectations.

The analysis of the results is supported by the use of additional control variables defined in the input files, and other specialized cards to allow the monitoring of other hydraulic parameters of interest not directly available in the output files. When possible, simulation results are compared with available plant data or other simulations, and engineering judgment is performed.

SNPB-3-30 Quality Assurance Program Documentation Attachment 1 Page 58 of65 Please demonstrate that the documentation for the QA program includes procedures to address all relevant areas including, but not limited to, design control, document control, software configuration control and testing, and co"ective actions. Criterion

6.2 Reference

SRP, lll.3f STP Response:

As explained in the response to SNPB-3-23, material evidence of compliance with each required element of the STP Appendix B program is procedurally required to be verified and documented by a second review check (a qualified reviewer) prior to completing the procedure.

SNPB-3-31 Independent Peer Review Attachment 1 Page 59 of65 Please demonstrate that the QA program used independent peer review in the key steps appropriately.

This should include a description of the steps where independent peer review was applied and how independence was defined and obtained.

Criterion

6.2 Reference

SRP, lll.3f STP Response:

OPGP03-ZA-0307, "Engineering Calculations" defines the roles of the preparer and checker in clearly defined procedural steps. Additional review and approvals by supervision (Step 3.2.4), primarily for compliance and completeness, are also included.

Step 3.2 "Calculation Review and Approval" requires the calculation package to be assembled and forwarded to a qualified individual (fully qualified to use the procedure under the STP training program) for review of "completeness, clarity and accuracy".

The procedure requires the calculation to be prepared prior to sending it to the qualified reviewer ("Checker);

the calculation must "Present a description of the analysis used such that the Checker can understand and reconstruct the method used to perform the calculation" (Step 3.1.6.3).

The procedure further requires the reviewer to "develop a comprehensive understanding of the calculation methodology and content and be able to respond to any questions about the calculation." The procedure is not complete until all review comments are resolved (Step 3.2.3).

SNPB-3-32 Important Sources of Uncertainty Please identify the important sources of uncertainty in the L TCC EM. Criterion

5.1 Reference

SRP, lll.3e STP Response:

NOC-AE-16003395 Attachment 1 Page 60 of65 The major sources of uncertainty are identified based on their potential impact on the cladding temperature immediately after the core blockage time, and the long-term core cooling period subsequent to the core blockage.

These sources of uncertainty are classified based on the expected relative importance (Low, Medium, Medium/High, and High). Steady-State Model Uncertainties (1) Reactor Nominal Power (Importance-Low) In the proposed EM, the reactor core is assumed to be at its nominal power when the break occurs [1]. The reactor initial power determines the decay heat initial value at the reactor shut down. The uncertainty of the reactor nominal power is expected to have a minim'al impact on the cladding temperature during the L TCC, and after the core blockage.

(2) Core Heat Structures Thermal Properties (Importance

-Low) Heat capacity and thermal conductivity of fuel, claciding, and gap are specified in the EM as part of the thermal properties required for the core heat structures

[1]. These parameters are known to play a role in the predicted behavior of the cladding temperatures during initial phase of a LOCA. Due to the expected lower temperatures in the core heat structures at the time of core blockage and subsequent lower energy stored in the fuel, these parameters are not considered to have an important impact on the prediction of the cladding temperature after core blockage.

(3) Reactor Vessel Passive Heat Structures (Importance

-Low) Passive heat structures are defined in the EV to simulate the mass of steel of the reactor vessel and other internals

[1]. The presence of these heat structures may affect the simulated behavior of certain regions of the reactor vessel such as upper plenum, downcomer, and lower plenum. The importance of the impact may depend on the break size and location.

Although effects on the cladding temperature are not expected to be important.

(4) Reactor Core Axial Power Shape (Importance

-Medium) NOC-AE-16003395 Attachment 1 Page 61 of 65 The core axial power shape is expected to change during the fuel cycle due to the presence of burnable poisons. During a hypothetical core blockage scenario, the average void in the core is expected to increase.

Liquid water is expected to reach the core from the top. The axial location of the core power peak and the overall power shape may have a direct impact on the cladding temperature.

(5) Steam Generators Tube Plugging (Importance

-Medium) The liquid level in the primary side of the SGs' tubes is expected to change during the phases of the accident.

In particular, during the period immediately after the core blockage event, redistribution of the flow through the primary system is expected to occur, forcing the ECCS flow through the steam generators.

Flow established through the u-tubes may affect the heat transfer between the primary and secondary sides of the steam generators.

The SGs' tube plugging is expected to affect the flow behavior through the steam generators and the subsequent heat transfer from/to the primary coolant during the core blockage scenario.

This may have a direct impact on the cladding temperature since it affects the conditions of the alternative flow paths included in the EM. (6) Vessel Flow Bypass Fractions (Importance

-High) The majority of the primary system coolant flow passes though the core during normal operation.

A certain fraction of the total primary coolant flow is diverted to vessel flow paths. The following flow paths or core bypass flow are considered in the EM:

• Flow through the spray nozzles into the upper head for head cooling purposes.

• Flow entering into the rod cluster control guide thimbles to cool the control rods.

• Leakage flow from the downcomer inlet nozzle directly to the vessel outlet nozzle through the gap between the vessel and the barrel.

• Flow introduced between the baffle and the barrel for the purpose of cooling these components and which is not considered available for core cooling.

• Flow in the gaps between the fuel assemblies on the core periphery and the adjacent baffle wall. During a hypothetical full core blockage at the bottom of the core, these flow paths may represent alternative paths through which the cooling water reaches the core, and, subsequently, have an important impact on the core coolability and cladding temperature.

(7) Core Nodalization (Importance

-Medium/High)

The nodalization adopted to simulate the primary system is carefully selected in the EM based on the LOCA simulation guidelines included in the RELAP5-3D users' manual [2]. The nodalization of the core, in particular the radial nodalization (number of channels) is NOC-AE-16003395 Attachment 1 Page 62 of65 an important factor for the proposed EM due to the expected flow behavior in the core during the core blockage phase: colder coolant reaching the top of the core through alternative flow paths may enter the core preferentially through cold channels where the vapor core exit velocities are expected to be lower than the ones at the exit of hot channels.

Liquid cross flow may help to cool down the core. The prediction of these phenomena may be affected by the number of channels used to simulate the core. (8) Upper Head Nodalization (Importance

-Medium) As mentioned above (6), the upper plenum sprays are considered one of the main alternative flow paths through which ECCS water injected into the cold legs may reach the core during a hypothetical core blockage scenario.

The nodalization of the upper head region (including upper plenum sprays and upper guide tubes) is important for this EM. Transient (LOCA) Model Uncertainties (9) RWST Usable Volume (Importance

-High) The RWST usable volume is defined as the actual volume that is available for injection during the safety injection phase of a LOCA. When the usable water is depleted, the sump switchover procedure is initiated.

The sump switchover time and, subsequently, the time at which the core is assumed to be blocked in the EM, is strictly related to the RWST usable volume. A larger volume results in a delayed core blockage time and, subsequently, to a lower decay heat generated in the core at the time of core blockage.

This source of uncertainty is expected to impact the cladding temperature after the core blockage time. (10) Decay Power Model (Importance-High) Different models of decay power are available in RELAP5-3D.

The model adopted may have an impact on the prediction of the cladding temperature since it affects the core heat flux. (11) Break Size (Importance-High) The behavior of the primary system during the L TCC (including the time subsequent to the core blockage) is expected to change with the break size. Cladding temperature is expected to be indirectly impacted by the break size.

(12) Orientation (Importance

-Low) NOC-AE-16003395 Attachment 1 Page 63 of65 The location of the break is assumed to be in one of the largest pipes of the primary system (cold or hot legs). The azimuthal location of the break (break orientation) is normally expected to affect the early phase of the accident progression.

The orientation of the break may impact the quality of the mixture discharged through the break also during the L TCC and core blockage phases but his direct impact on the cladding temperature may not be of particular importance.

(13) ECCS Flow Rate (Importance-Medium) The ECCS flow rate have a similar impact to what described in (9). The sump switchover time and, subsequently, the time at which the core is assumed to be blocked in the EM, is strictly related to the injected flow rate. A larger ECCS flow rate results in an earlier core blockage time and, subsequently, to a higher decay heat generated in the core at the time of core blockage.

This source of uncertainty is expected to impact the cladding temperature after the core blockage time. (14) ECCS Injection Temperature (Importance

-Medium/High)

The ECCS injection temperature affects the rate of heat removal of the decay heat during the L TCC with subsequent impact on the conditions of the core (void fraction, core temperatures) at the core blockage time. (15) Core Barrel/Baffle Bypass Blockage Fraction (Importance-High) The core barrel/baffle bypass is one of the most important alternative flow paths in the reactor vessel, allowing flow to reach the top of the core during an event of core blockage.

The fraction of the core barrel/baffle blockage at the time of sump switchover affects the amount of coolant passing through this alternative flow path and, subsequently the core coolability.

(16) Core Blockage Fraction (Importance-High) The assumed core blockage fraction (blocked flow area/ total core flow area) determines the amount of coolant reaching the core after the blockage, and the subsequent cladding temperature.

NOC-AE-16003395 Attachment 1 Page 64 of65 (17) Plant Set Points and Delays (Importance

-Low) Set points and delays are defined in the EM as part of the control logic. The control logic simulates automatic and manual operations such as:

• Reactor trip

• RCP trip

• ECCS pumps start

• AFW start and MFW shut off

• Signals processing As these parameters are expected to affect the early phase of the transient, their impact on the cladding temperature during the L TCC phase is considered low. (18) CCFL Parameters (Importance-Medium) Counter-current flow limited conditions may occur in certain location so the primary system. The core exit is one of the most likely location where this condition may occur, in particular during the core blockage phase, where liquid water moves downward into the core while the vapor proceeds upward. Parameters of the CCFL model may are specified in the EM to simulate this phenomenon.

The selection of these parameters may affect the conditions at which the CCFL occurs and, subsequently, have an impact on the core coolability.

These sources and their importance are summarized in the table below. Number Description Origin Importance 1 Reactor Nominal Power Steady-State Model Low 2 Core Heat Structures Thermal Properties Steady-State Model Low 3 Reactor Vessel Passive Structures Steady-State Model Low 4 Axial Power Shape Steady-State Model Medium 5 Steam Generators' Tube Plugging Steady-State Model Medium 6 Vessel Flow Bypass Fractions Steady-State Model High 7 Core Nodalization Steady-State Model Medium/High 8 Upper Head Nodalization Steady-State Model Medium 9 RWST Usable Volume Transient Model High 10 Decay Power Model Transient Model High 11 Break Size and Location Transient Model High 12 Break Orientation Transient Model Low 13 ECCS Flow Rate Transient Model Medium 14 ECCS Injection Temperature Transient Model Medium/High 15 Core Barrel/ Baffle Bypass Blockage Fraction Transient Model High 16 Core Blockage Fraction Transient Model High 17 Plant Set Points and Delays Transient Model Low 18 CCFL Parameters Transient Model Medium References

[1]. RCa9989 RELAP5-3D Steady-State Model Rev.a [2]. STI 3428a651 RELAP5-3D Software Quality Assurance.

Rev.a. NOC-AE-16003395 Attachment 1 Page 65 of65 Attachment 2 Definitions and Acronyms NOC-AE-16003395 Attachment 2

NOC-AE-16003395 Attachment 2 Page 1 of2 Definitions and Acronyms ANS American Nuclear Society ECCS Emergency Core Cooling ARL Alden Research Laboratory System (also ECG) ASME American Society of ECWS Essential Cooling Water Mechanical Engineers System (also ECW) BA Boric Acid EOF Emergency Operations BAP Boric Acid Precipitation Facility BC Branch Connection EOP Emergency Operating BEP Best Efficiency Point Procedure(s)

B-F Bimetallic Welds EPRI Electric Power Research B-J Single Metal Welds Institute BWR Boiling Water Reactor EQ Equipment Qualification CAD Computer Aided Design Engineered Safety Feature CASA Containment Accident FA Fuel Assembly(s)

Stochastic Analysis, also a FHB Fuel Handling Building short name for the CASA GDC General Design Criterion(ia)

Grande computer program GL Generic Letter that uses the analysis GSI Generic Safety Issue methodology 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 HVAC Heating, Ventilation

& Air CET Core Exit Thermocouple(s)

Conditioning CHLE Corrosion/Head Loss ID Inside Diameter Experiments IGSCC lntergranular Stress CHRS Containment Heat Removal Corrosion Cracking System ISi In-Service Inspection CLB Cold Leg Break or Current IOZ Inorganic Zinc Licensing Basis LAR License Amendment CRMP Configuration Risk Request Management Program LBB Leak Before Break cs Containment Spray LBLOCA Large Break Loss of Coolant CSHL Clean Strainer Head Loss Accident (also LLOCA) css Containment Spray System LCO Limiting Condition for (same as CS) Operation eves Chemical Volume Control LDFG Low Density Fiberglass System LERF Large Early Release OBA Design Basis Accident Frequency DBD Design Basis Document LHS Latin Hypercube Sampling D&C Design and Construction LHSI Low Head Safety Injection Defects (ECCS Subsystem)

DEGB Double Ended Guillotine LOCA Loss of Coolant Accident Break LOOP/LOSP Loss of Off Site Power DID Defense in Depth MAAP Modular Accident Analysis DM Degradation Mechanism Program NOC-AE-16003395 Attachment 2 Page 2 of 2 Definitions and Acronyms MAB/MEAB Mechanical Auxiliary RMS Records Management Building or Mechanical System Electrical Auxiliary Building RMTS Risk Managed Technical MBLOCA Medium Break Loss of Specifications Coolant Accident (also RPV Reactor Pressure Vessel MLOCA) RVWL(S) Reactor Vessel Water Level NIST National Institute of (System) Standards and Technology RWST Refueling Water Storage NLHS Non-uniform Latin Tank Hypercube Sampling SBLOCA Small Break Loss of Coolant NPSH Net Positive Suction Head, Accident (also SLOCA) (NPSHA -available, SC Stress Corrosion NPSHR -required)

SI/SIS Safety Injection, Safety NRC Nuclear Regulatory Injection System (same as Commission ECCS) NSSS Nuclear Steam Supply SIR Safety Injection and System Recirculation OBE Operating Basis Earthquake SR Surveillance Requirement OD Outer Diameter SRM Staff Requirements OQAP Operations Quality Memorandum Assurance Plan SSE Safe Shutdown Earthquake PCI Performance Contracting, STP South Texas Project Inc. STPEGS South Texas Project Electric PCT Peak Clad Temperature Generating Station PDF Probability Density Function STPNOC STP Nuclear Operating PRA Probabilistic Risk Company Assessment TAMU Texas A&M University PWR Pressurized Water Reactor TF Thermal Fatigue PWROG Pressurized Water Reactor TGSCC Transgranular Stress Owner's Group Corrosion Cracking PWSCC Primary Water Stress TS Technical Specification(s)

Corrosion Cracking TSB Technical Specification QA Quality Assurance Bases QDPS Qualified Display Processing TSC Technical Support Center or System Technical Specification RAI Request for Additional Change Information TSP Trisodium Phosphate RCB Reactor Containment UFSAR Updated Final Safety Building Analysis Report RCFC Reactor Containment Fan UNM University of New Mexico Cooler USI Unresolved Safety Issue RCS

• Reactor Coolant System UT University of Texas (Austin) RG Regulatory Guide V&V Verification and Validation RHR Residual Heat Removal VF Vibration Fatigue RI-ISi Risk-Informed In-Service WCAP Westinghouse Commercial Inspection Atomic Power RMI Reflective Metal Insulation ZOI Zone of Influence