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

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| issue date = 07/21/2016
| issue date = 07/21/2016
| title = Third Set of Responses to Requests for Additional Information STP Risk-Informed GSI-191 Licensing Application
| title = Third Set of Responses to Requests for Additional Information STP Risk-Informed GSI-191 Licensing Application
| author name = Powell G T
| author name = Powell G
| author affiliation = South Texas Project Nuclear Operating Co
| author affiliation = South Texas Project Nuclear Operating Co
| addressee name =  
| addressee name =  
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=Text=
=Text=
{{#Wiki_filter: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:
{{#Wiki_filter:Soutll li:x.75 Project Electric Gr:11erall11g Statlo11 P.O. Bar 1$9 1%dswortll, T=s 77./$1 July 21, 2016 NOC-AE-16003395 10 CFR 50.12 10 CFR 50.90 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 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)
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:==
==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
: 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 Risk-Informed Approach to Address Generic Safety Issue (GSl)-191 and Respond to Generic Letter (GL) 2004-02", August 20, 2015, NOC-AE-15003241, ML15246A126
: 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.
: 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, ML16082A507 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).
This submittal responds to the third set of RAls from the Nuclear Performance and Code Branch (SNPB). There are no commitments in this submittal.
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
STl34345851
: 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
NOC-AE-16003395 Page 2 of 3 If there are any questions, please contact Mr. Wayne Harrison at 361-972-8774.
* 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.
I declare under penalty of perjury that the foregoing is true and correct.
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.
Executed on:
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.
G. T. Powell Executive Vice President ,
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.
and Chief Nuclear Officer awh Attachments:
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  
: 1. Response to SNPB-3-2,-6, -7, -15, -17, -18, and -20 through -32
-both primary and secondary side, loops, pressurizer, pumps, containment).
: 2. Definitions and Acronyms
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
NOC-AE-16003395 Page 3 of 3 cc:
[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:
(paper copy)                       (electronic copy)
* Period 1: Break Event and Slowdown (300 s. --396 s.)
Regional Administrator, Region IV   Morgan. Lewis & Beckius LLP U. S. Nuclear Regulatory Commission Steven P. Frantz, Esquire 1600 East Lamar Boulevard Arlington, TX 76011-4511           U. S. Nuclear Regulatorv Commission Lisa M. Regner Lisa M. Regner Senior Project Manager             NRG South Texas LP U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (08H04)       Jim von Suskil 11555 Rockville Pike               Skip Zahn Rockville, MD 20852 CPS Energy NRC Resident Inspector             Kevin Pollo U. S. Nuclear Regulatory Commission Cris Eugster P. 0. Box 289, Mail Code: MN116     L. D. Blaylock Wadsworth, TX 77483 Crain Caton & James. P.C.
* Period 2: Refill and Reflood (-314 s. --434 s.)
Peter Nemeth City of Austin Elaina Bail John Wester Texas Dept of State Health Services Helen Watkins
* Robert Free
 
NOC-AE-16003395 Attachment 1 Attachment 1 Response to SNPB-3-2, -6, -7, -15, -17, -18 and -20 through -32
 
NOC-AE-16003395 Attachment 1 Page 1 of65 Preface to SNPB Responses STPNOC has revised its LTCC approaches to the cold leg SBLOCA and the hot leg LBLOCA that reduces the scope of the RELAP5-3D LTCC 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 LTCC 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 -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 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 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.
NOC-AE-16003395 Attachment 1 Page 2 of65 Period 1: Break Event and Slowdown (300 s. - -396 s.)
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 Reactor Trip                                 308 Blowdown 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.
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.
NOC-AE-16003395 Attachment 1 Page 3 of 65
As the accumulators
                                        - Toi lC(S mflawR.nt (e*cluding Accum)     -   TotM AcctsTiul.ttOt mfJowlWtf' -     Corf' Cll 4500                                                                                                                                       16 4000
' 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).
      ,500 12 Vi'    3000 co 2500 to 'i='
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.
u..
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  
w
---*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  
~                                                                                                                                                    z a::  2000                                                                                                                                          0
---*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
0
----SumpSw11chowr --Bloclc.ageTime
-'   1500                                                                                                                                          w>
--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.
u..                                                                                                                                             6   -'
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
Vl w
-CorelnTemperature
Vl c:i:  1000
-coreOut T emperature
~                HPSI injection start 500 LPSI inj*ction start 500 190                                                                  390                                    440                490 SIMULATION TIME [SJ The core collapsed liquid level reaches a first minim um at 396 seconds.
-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.
During this period the primary side of all steam generators is empty.
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.
Period 2: Refill and Reflood (-314 s. - 434 s.)
The PCT follows the satu r ation temperatu r e as shown in the figure below (see description of the core heat transfer regimes).
Due to the location of the break, the refill and reflood phases may not be easily distinguishable compared to typical larg e break scenarios in that cold 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 starts increasing.
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
The injection of the accumulators (accumulators' injection starts at 382 seconds) in the cold legs is preferentially diverted through the core . The collapsed liquid level is partially recovered due to the accumulators' injection. As the accumulators' injection decreases, the core void fraction and liquid mass inventory starts 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 approximately 314 seconds) , provides at total flow over 1500 lbm/s. Most of the SI flow is diverted toward the reactor vessel, and passes through the reactor core, recoveri ng the collapsed liquid level.
-CoreOutTemperiture
The core is completely flooded (collapsed liquid level at the top of the core) at approximately 434 seconds.
-C0teOUtS.tT
During this period the primary sides of all steam generators is empty.
-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:
Other important events observed during this phase are listed in the table below.
 
NOC-AE-16003395 Attachment 1 Page 4 of 65 Phase                                                    Event                                    Time (s)
AFW Activation                                                                       338 Accumulator Activation                                                              383 Refill/Reflood Minimum Core Collapsed Liquid Level                                                 417 Core Full                                                                           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 &deg;F with a slow decrease rate due to the decay power decrease.
The injected ECCS flow 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 appreciable difference is seen between broken loop and intact loops as shown in the figures below.
                    -   Cll SG l , 1' primary s ide    - CLL SGl , J.. pr1mciryside - - -
* Sump Switchover  - - Blockage Time 40 35 3-0 f=' 25 L.L UJ 20 UJ 15 10 1000                            2000                       3000                     4000                   5000         6000 SIMULATION TIME (SJ SG 1 (Intact Loop, Pressurizer Loop) Primary Side Collapsed Liquid Level
 
NOC-AE-16003395 Attachment 1 Page 5 of 65
        -     CLLSG2,1'primarvside    - CLLSG2 , .J,,primaryside - - -
* SumpSwitchovt-r - - SlockageTime 40 35 30 t;:  25
...J w    20 w
...J 15 10 1000                     2000                     3000                     4000                 5000          6000 SIMULATION TIME [SJ SG 2 (Intact Loop) Primary Side Collapsed Liquid Level
 
NOC-AE-16003395 Attachment 1 Page 6 of 65
          - - Cll SG3, 't primary side    - - CLL SG3, ..J, primary side  - - - - Sump Switchover    -  -  Blockage Time 40 35 30
...J LU   20 LU
...J IS 10 0 1000                        2000                          3000                        4000                          5000          6000 SIMULATION TIME (SJ SG 3 (Broken Loop) Primary Side Collapsed Liquid Level
          - - CLL SG4, 1" pnmary side     - - CLL SG4, J, primary side       - - -
* Sump Switchover     -   -   Blockage Time 40 3S 30
...J LU 20 LU
...J IS 10 1000                         2000                           3000                       4000                           5000          6000 SIMULATION TIME [SJ SG 4 (Intact Loop) Primary Side Collapsed Liquid Level
 
NOC-AE-1 6003395 Attachment 1 Page 7 of 65
          - - CLLSGl,secondaryslde    - - - - SumpSw11chowr  - - Bloclc.ageTime    - - AFWSGl      - MFWSGl 60 70 50 60 Vi'
                                                                                                                                      ~
40                                                                                                                        50 CXl UJ 40   !<i:
- ' 30 UJ a::
>                                                                                                                                     5 UJ
-'                                                                                                                             30    9u..
20                                                                                                                               Vl Vl 20   <t
                                                                                                                                      ~
10 10 500             1000              1500              2000                2500            3000          3500      4000 SIMULATION TIME [SJ Intact Loop (Loop 1) Secondary Side Conditions
          - - CLL SG3, secondary side - - -- Sump Switchover       Blockage Time     - - AFWSG3         MFWSG3 60 70 50 60 Vi'
                                                                                                                                      ~
40                                                                                                                        50 CXl i='
u..
UJ 40     !<i:
- ' 30                                                                                                                                a::
UJ UJ s0
-'                                                                                                                            30 u..
20                                                                                                                                Vl Vl 20   <t
                                                                                                                                      ~
10 10 500              1000              1500              2000                2500            3000          3500      4000 SIMULATION TIME [SJ Broken Loop (Loop 3) Secondary Side Conditions
 
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 Temperature -RWST/StJmp2 Temperat Ufe - - -
* SumpSwitdiove r - -   Blockage Time 2ao LI:'  230 UJ a::
\:t ffi  1ao a..
~
UJ f-80 1000              2000                     3000                   4000                       sooo                 6000 SIMULATION TIME [SJ ECCS suction Temperature (Blue Line) and ECCS Injection Temperature (Gray Line)
There is no appreciable effect on the overall primary system behavior due to the increase in the ECCS injection temperature following the SSO time.
Core coolant inlet and outlet temperatures remain subcooled during this phase. After the SSO time, core coolant temperatures increase following 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 similar to the core coolant temperatures as depicted in the figure below.
 
NOC-AE-16003395 Attachment 1 Page 9 of 65
                  -    PCT - - -
* SumpSwitchover -   - BlockageTime    - CorelnTemperature -     coreOut Temperature - coreOUtSatT 500
~      480 u..
0 UJ a::
f-::: 380
<t a::
UJ a..
::::?  280 UJ t--
180 80 290        490            690            890            1090        1290          1490            1690      1890          2090 SIMULATION TIME [SJ Core Temperatures The full core and core bypass blockage is assumed to occur instantaneously at 2100 seconds (360 seconds after the SSO time).
Period 4: Core Blockage and Post-Blockage Long Term Core Cooling (2099 s. and after)
The sudden decrease in the core flow rate due to the instantaneous core blockage at 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 generation in the core. This is also the cause of the increase of the saturation temperature.
The PCT follows the saturation temperature as shown in the figure below (see description of the core heat transfer regimes) .
 
NOC-AE-16003395 Attachment 1 Page 10 of 65
              -  PCT    - - -
* St.ntpSw1tcho\iter - -  Bfc:di:ag,eTime  - C~lnlemper<1tu1e - CoreOutTemperiture -   C0teOUtS.tT - CoreCLl 16 BJ 12 G:'  280
'2....                                                                                                                                             10 ~
UJ                                                                                                                                                    u..
a::
=>                                                                                                                                                    z ti:  230 0
a::
UJ                                                                                                                                                    ~
a..                                                                                                                                                  >
UJ
~                                                                                                                                                6   -I UJ
~ 180 130 80                                                                                                                                         0 1000         1500               1000         1500                         3500                 4500       5000                   6000 SIMULATION TIME (SJ Core Temperatures and Core Collapsed Liquid Level.
The core collapsed liquid level decreases until a stable value of approximately 8 ft is reached . This is an indication that the reactor core inlet/outlet mass flow is 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 rates 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.
* 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
* No appreciable net flow through the SGs' primary tubes is observed an instant before the core blockage time.
'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}  
When the core blockage occurs, a redistribution of the flow within the primary system occurs. The behavior of the SGs immediately before and after the core blockage is described below. The description is complemented with plots of the integral flow rates through each of the loops.
-lnt(CL3->Vessel)  
Broken Loop (Loop 3)
-l nt(CL3->Pump.3)  
The ECCS flow injected in the cold leg 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.
-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.
Most of the ECCS flow through Loop 3 is now forced toward the SG . This flow can be
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  
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{ECCS->Loop2)  
                    -   lnt(HL3*>Vessel} - lnt(CL3->Vessel) - lnt(CL3->Pump.3) - lnt(ECCS->Loop3) - - -
---*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)  
* SumpSwrtchove r - - Slockage l 1me 8000000 6000000
-lnt{CL4->Vessel)  
    'iii'
-lnt(ECCS->loop4)  
_J UJ 4000000
----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  
a::
--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.
3      2000000 0_J u.
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
Vl Vl
-*-* -*-* -*-* . -. -. -. ,., . ,
    <{                                                                                                                          5000                 6000
* 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. 
0     -2000000 UJ
--HTmode6052
    ~
--HT mode 605 16 so Vi' Vl .. !:::: z 46 ::::> a: 44 w co --HT mode 605 4 --MTmode60518
a::
--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
    \,!) -400JOOO UJ 1-z
.... 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
          -6000000
--HTmode6061-04
          -llXXJOOO SIMULATION TIME [SJ Broken Loop (Loop 3) Integral Flow Splits Intact Loops Equipped with SI trains (Loops 2 and 4)
--HTmode6061-06--H T mode606I-OS
The behavior of the intact loops equipped with SI trains appears to be similar during the transient and in particular during the pre- and post-core blockage phases. The ECCS flow injected in the cold leg of intact Loops 2 and 4 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 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 is flow provides 1
--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
The main parameters of the plots are explained here after:
< 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.
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 vessel inlet).
lnt(CLX 7 Pump): Integral flow from CL of Loop X toward the RCP of the same loop (taken 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 slower slope the plot indicates a lower time-average flow rate toward the direction specified. A change in the slope sign is an indication 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     - lnt{ECCS->Loop2)   - - -
* SumpSw1tchover   -   -   BlockageTime 8000000 I
I 500                   1000               1500               1000 I               2500                 3000             3500       4000 I
I I
      -4000000 SIMULATION TIME [SJ Intact Loop (Loop 2) Integral Flow Splits
                        -      lnt(Hl4->Vessel)     - lnt{CL4->Vessel)   -   lnt(ECCS->loop4)   - - - - SUmpSw1td10\<er -   -     Bloc:k*gel1me 8000000 co
....J 6000000 w
~
a::   4000000 3!
I 0
....J u.
Vl     2000000 Vl
<(
:;?
Cl w
~                          500                   1000               1500              2000 I               1500                 3000             3500       4000 a::                                                                                           I
(.!)
I w
I-                                                                                           I z     -2000000                                                                               I I
      *4000000 SIMULATION TIME [SJ Intact Loop (Loop 4) Integral Flow Splits
 
NOC-AE-16003395 Attachment 1 Page 13 of 65 Intact Loop Not Equipped with SI train (Loop 1)
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 UJ
~
~    4000000
;~ I                                                                    I a
UJ                                                                       I
~                    500      1000              1500          2000 I I
2500                      3500      4000      4500 a::
(.!)                                                                     I UJ                                                                       I zf-- . 200000()
    -4000000 SIMULATiON TIME (SJ Intact Loop (Loop 1) Integral Flow Splits Analysis of the case executed also showed that the majority of the flow reaching the top of the core is supplied through the SGs with a small (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 limitation occurs, and how long these conditions are maintained.
 
NOC-AE-16003395 Attachment 1 Page 14 of 65
                                -    * - coreOUtlet                    - - Sump SW1tchover                - - - - Blockage nme 10000 Vi' UJ
  ~
i==
_,   1000
  <i:
c::                                   , . - . - . - . ,., .
t!)
UJ                             ,
* J f--                       .'
z 100 z
0
  ~                      '
f--
u
  <i:    10         '*
:1 u...
u                 I u                 I I
I I
I     I l         I 300                     5300                      10300            15300              20300                    25300          30300 SIMULATION TIME [SJ Core Outlet CCFL Actuation Integral Time (y-axis log scale)
Core Heat Transfer Regimes during Pre- and Post -Blockage Phases This section describes the heat transfer regimes established during the pre- and post-core blockage phases of the transient.
The heat transfer regime fo r the average assembly, hot assembly and hottest rod are shown in the figures below.
 
NOC-AE-16003395 Attachment 1 Page 15 of 65
      - - HTmode6052          - - HT mode 605 4          - - HT mode 605 6    - - HT mode 605 8      - - HT mode 60511        - - HTmode60514
      - - HT mode 605 16      - - MTmode60518            - - HT mode 605 20    - - - - Sump SWitcho~r - - Blockageli me so Vi' Vl
  ~
z    46 a:    44 w
co
  ~    42  - - - --,,-- -+--
z w    40 Cl 0    38                                    Si111lo-Pho1e Liquid
  ~                                            Conv*ction or Subcooled w*ll a:                                           with Void Froction < 0.1) w    36 u..
Vl z    34
  <t:
a:
  ..... 32
  ~
w    30
::r:     1000          1500    2000              2500            3000    3500              4000    4500            5000      5500          6000 SIMULATION TIME [SJ Average Assembly Heat Transfer Regimes 2 2
Heat transfer regimes table available in RELAP5-3D user's manual, volume IV, Section 4.2.1
 
NOC-AE-16003395 Attachment 1 Page 16 of 65
                      - - HT mode 6060 2 - - HT mode 6060 4 - - HT mode 6060 6 - - HT mode 6060 8 - - HT mode 6060 11 - - HT mode 6060 14
                      - - HTmode6060 16 - - HTmode6060 18 - - HTmode6060 20 - - -- Sump Switchover -            -  Blockage Time so Vi' V'l LU 48
.....J t::
z          46 ex:        44 LU co
~          42 z
LU 40                                                                                           Subcooled Nucl*at* Boilin&
0                                                                                               Saturated NudHto Bolllnc 0         38 Sincle-Ph*H Uquld
~                                                Convection 0< Subcooled wall ex:                                             with Vold Froctlo n < 0.11 LU 36 LL V'l z         l4
<(
ex:
I-         l2
~
LU 30
:I:           1000   1500            2000              2500            3000            3500          4000            4500    sooo      5500    6000 SIMULATION TIME [SJ Hot Assembly Heat Tra nsfer Regimes
                      -----*Sump Switchover - - -  Blockage Time  -  -  HTmode6061 - HTmode6061 - HTmode6061-06--HT mode606I-OS
                      - - HTmode6061- ll - - HTmode6061-14                  HT mode 6061-16      HTmode6061-18      HT mode 6061-20 so Vi' V'l LU 48
    .....J t::
z        46 ex:      44 LU co
    ~        42 z
LU 40 0                                                                                             Subcoo*od Nuc!tiato l!lolllnc 0        38 Saturated NuclHto Bollinc
    ~                                            Sincle-PhHo Uquld Conv* ction or Subcooled Wllll ex:
LU 36                                        wit h Void FrllCtlon < 0.11 LL V'l z
    <(
34 ex:
I-      32
    ~
LU 30
:I:
1000  1500            2000              2500            3000            3500          4000            4500    sooo      5500    6000 SIMULATION TIME [SJ Hottest Rod Heat Transfer Regimes As can be seen , the heat transfer in the core before th e SSO time is determined by the large amount of subcooled ECCS flow forced through the core during 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 post-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:
The following plots are included:
* Primary Pressure
* Primary Pressure
Line 98: Line 317:
* Core Inlet/Outlet flow rates
* Core Inlet/Outlet flow rates
* Core Bypass Inlet flow rate
* Core Bypass Inlet flow rate
* Total SG Heat Transfer -PnmPressure  
* Total SG Heat Transfer
----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 '"'&deg; 8300 9300 2.000&#xa3;.02 l.SOOEt-02 1.600&#xa3;+02 1 400&#xa3;+-02 l.200E+02 0:: UJ 1.000&#xa3;+02 0 8.000Et-0 1 a.. 6.000Et-0 1 4.000Et-01 2.000&#xa3;+0 1 0.000&#xa3;.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
                                  -     PnmPressure     - - - - SumpSwitchover - - BlockageTime 320 270
----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.
  ~ 220 Vl a._
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  
UJ a:   170
--
:J Vl Vl UJ a::
Time 1000 2000 3000 4000 SIMULATION TIM E [SJ ECCS and Break Integ r al Flow --CorelnletFlow  
  ~    120 70 20 3300              4:>JO            5300 300 !:>JO     2300                                                       6300
-CoreOotletFlow  
                                                                                                '"'&deg; 8300     9300 SIMULATION TIME [SJ Primary Pressure
----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  
NOC-AE-16003395 Attachment 1 Page 18 of 65
----Su'TipSw1tchover --Bhxkagelime NOC-AE-16003395 Attachment 1 Page 20 of 65 I
                                        - - Tola/CorePa.ver        ---- Sump Switchover      - -    SlockaseTime 2.000&#xa3;.02 l.SOOEt-02 1.600&#xa3;+02 1 400&#xa3;+-02
___ 5_000 ___ 6000--Vl lf H I Vl <t: ::1! -1 000 -1 500 -2000 &#xa3; U:XX-0 1 0 3 '&deg; 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
l.200E+02 0::
---*Sump Switchover --Blockage Time 1000 2500 3000 l 500 4000 4500 SIMULATION TIME [SJ Total SG Primary to Secondary Heat Transfer4 References
UJ       1.000&#xa3;+02
[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.
0         8.000Et-0 1 a..
Provide appropriate references.
6.000Et-0 1 4.000Et-01 2.000&#xa3;+0 1 0 .000&#xa3;.00 1000                2000                      3000                      4000              5000        6000 SIMULATION TIME [SJ Core Decay Power
                                      - - Core Power per Latent Heat    - - -
* SlATlp Switchover  -  -  Blockage Time 200 111) 160 Vi' co
  .=., 140 UJ
  ~    120 0::
u...
u...
0   100
__J 0
co 80 60 40 1000             2000                       3000                       4000                 5000          6000 SIMULATION TIME [SJ Core Boil Off Rate 3 3
The boil off rate _is calculated by dividing the core decay power by the latent heat of evaporation of water at 150 psia.
 
NOC-AE-16003395 Attachment 1 Page 19 of 65
                    -   integral Total Break Row   -   lnt@gralTot.al ECCSFlow     - - - - S1.mp Switchover     - -   Blockag~ Time 8000000 7000000 6000000 V'l 4000000 V'l
      <(
      ~
3000000 2000000 1000000 1000                       2000                         3000                           4000                     5000        6000 SIMULATION TIM E [SJ ECCS and Break Integral Flow
                        - - CorelnletFlow       -   CoreOotletFlow       - - - - SumpSw1tchowr       -  -  BlockageTime 9000 7000 Vi' c__Jo    5000 LU
~
a:        3000
~
0__J u..        1000 V'l V'l
<(
~
          -nxl
          -5000 SIMULATION TIME [SJ Core Inlet and Outlet Flow Rate
 
NOC-AE-16003395 Attachment 1 Page 20 of 65 2000                             -   BypasslnletFJow     - - - - Su'TipSw1tchover -   -   Bhxkagelime 1500
  ~
Vl 1000 co I~,-N:--------
                                                        ,0 0---40 0                          ___5_000___6000- -
Vl Vl
  <t:
lf      H        I
::1! -1000
        - 1500
        -2000 SIMULATION TIME [SJ Core Bypass Inlet Flow Rate
                                          -  Heat to SGs_secondary    - - -
* Sump Switchover  -  -  Blockage Time U:XX-0 1
          &#xa3;0 3
          '&deg; 0 .00h 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 1000 1500    1000        2500             3000             l 500           4000           4500 5000    5500    6000 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.
 
NOC-AE-16003395 Attachment 1 Page 21 of65 SNPB-3-6 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.
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).
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:
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 proposed LTCC 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 1b 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 LTCC EM. Discussion of the confirmation of the initial and boundary conditions is provided below for each of the four input models included in the LTCC EM.
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
Steady-State Input Model The steady-state input model defines the initial conditions for all the LOCA simulations included in the LTCC EM. The steady-state model is prepared under the STP Appendix B preparation of calculations guidelines [1 ], and documented in [2].
[1 ], and documented in [2]. The steady state model defines the plant geometry and the model options selected.
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].
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.
Assumptions, sources of information and margins are also described in the steady-st~te documentation.
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.
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.
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 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.
The important system thermal-hydraulic parameters are set to STP technical specification values, or from other appropriate documentation.
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 LTCC EM is included in the responses to RAI SNPB 3-32, RAI SNPB 3-21, and RAI SNPB 3-22.
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:
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.
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.
LOCA Input Models Three LOCA scenarios are included in the LTCC EM. The scenarios use the same base input file with the sole difference in the break size (16", 6" and 2").
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&deg;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 &deg;F 14.7psia 270&deg;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 model includes all the phases of the accident, in particular:
The simulation results are analyzed to verify the correctness of the predictions and engineering judgment is used to evaluate the results.
Phase 1:   Slowdown Phase 2:   Refill/Reflood Phase 3:   Pre-Core Blockage LTCC Phase 4:   Post-Core Blockage LTCC 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.
References NOC-AE-16003395 Attachment 1 Page24 of65 [1]. South Texas Project Electric Generating Station. Preparation of Calculations.
 
OPGP04-ZA-0307, STI 34177871.
NOC-AE-16003395 Attachment 1 Page23 of65 Parameter          Value                    Parameter         Value 360000 RWST Usable Volume                           HHSI Actuation Delay      6s gal Core Blockage Timing 6min               LHSI Actuation Delay      10 s (from SSO)
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).
Low PZR pressure Set       1871.7             Hot Leg Injection time 5.5 h Point           psi a                 (from break)
[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).
Reactor Trip Signal                           AFW Delay (from 2s                                         30s Processing                                 reactor trie)
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.
Rod droe time         2.8 s                                     130 &deg;F RCP trip: RCS         1444.7           RWST water conditions 14.7psia Pressure Check         psi a Sump Pool MFW isolation Delay       2s                                       270&deg;F Temperature (at SSO)
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.
Accumulator Liquid MFW closure time         10 s                                     1200 ft3 Volume 600 psia Accumulator Water AFW mass flow rate     70.0 Ibis                                   and 120 Conditions            OF 50 psia, 123.4&deg;F-Specific            Containment sprays      4282 AFW conditions Enthalpy             volumetric flow rate    gal/min
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:
                              = 91.5 Btu/bl A discussion of uncertainties identified for the most critical boundary conditions listed above, the importance associated, and the margins included in the LTCC EM is included in the responses to RAI SNPB 3-32, RAI SNPB 3-21, and RAI SNPB 3-22.
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.
Initial conditions for the LTCC (Phase 3 and 4)' are included as part of the simulation of the preceding phases 1 and 2.
All the calculations performed are prepared under the STP Appendix B preparation of calculations guidelines
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.
[1], and properly documented.
 
NOC-AE-16003395 Attachment 1 Page24 of65 References
[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).
 
NOC~AE-16003395 Attachment 1 Page 25 of65 SNPB-3-7 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 LTCC EM are demonstrated to be appropriate for the simulations of LTCC 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.
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.
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:
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 LTCC EM is included in the responses to RAI SNPB 3-32, RAI SNPB 3-21, and RAI SNPB 3-22.
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.
As described in the response to RAl-SNPB-3-06, the model includes all the phases of the accident:
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.
Phase 1: Slowdown Phase 2: Refill/Reflood Phase 3: Pre-Core Blockage LTCC Phase 4: Post-Core Blockage LTCC 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 LTCC phases.
The simulation results are analyzed to verify the correctness of the predictions and engineering judgment is used to evaluate the results. References
Initial conditions for the LTCC (Phase 3 and 4) are included as part of the simulation of the preceding phases 1 and 2.
[1]. South Texas Project Electric Generating Station. Preparation of Calculations.
 
OPGP04-ZA-0307, STI 34177871.
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 LTCC 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.
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:
References
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.
[1]. South Texas Project Electric Generating Station. Preparation of Calculations. OPGP04-ZA-0307, STI 34177871. Rev.7
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:
NOC-AE-16003395 Attachment 1 Page27 of65 SNPB-3-15 Level of Detail 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.
I! Plant operating conditions (initial conditions for the transient)
Criterion 3.6 Reference         SRP, lll.3b STP Response:
The level of detail of the LTCC 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 LTCC; 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 LTCC EM is consistent with the best practice used to simulate LOCA scenarios with RELAPS-30.
The proposed LTCC 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
* Break opening and blowdown phase
* Refill/Reflood Phase
* Refill/Reflood Phase
* Pre-Core Blockage L TCC
* Pre-Core Blockage LTCC
* 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.
* Post-Core Blockage LTCC 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].
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.
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.
Three LOCA scenarios are included in the LTCC 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:
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 main emergency manual operations (EOPs)
* Plant LOCA automatic operations
* Plant LOCA automatic operations
Line 165: Line 451:
* 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.
* 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
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.
[1] RC09989 RELAP5-3D Steady-State Model Rev.O
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:
NOC-AE-16003395 .
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.
Attachment 1 Page29 of65 SNPB-3-17 Validation of the Evaluation Model 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.
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.
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.
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.
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:
These sources are reviewed to confirm they are properly consistent with the purpose of the simulations.
The EM results in reasonable predictions of the PCT for the range of HLB scenarios considered in the LTCC EM. The PCT response to the heat transfer regimes and flow conditions experienced during LTCC in these scenarios is validated by integral and separate effects tests as described in RAI SNPB 3-13.
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 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 thermal~hydraulic response during normal operation (steady-state) and during all the phases of the accident progression.
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
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]. Independent internal and external reviewers (One or more originators, one or more checkers) are identified and adequate verification is conducted.
: 1. Verification of the adequacy of the parameters in use in the EM
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.
: 2. Verification of input models
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.
: 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.
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).
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.
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 review process is conducted in accordance to the STP Appendix B guidelines [1].
The results are analyzed to verify the correctness of the simulation predictions.
Independent internal and external reviewers (One or more originators, one or more checkers) are identified and adequate verification is conducted.
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.
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 thermal-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:
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
* RAl-SNPB-3     Input Verification
-Input Verification
* RAl-SNPB-3     Proper Convergence
* RAl-SNPB-3-25
* RAl-SNPB-3     Non-physical Results
-Proper Convergence
* RAl-SNPB-3     Realistic Results
* RAl-SNPB-3-26
* RAl-SNPB-3     Boundary conditions as prescribed
-Non-physical Results
* RAl-SNPB-3     Thoroughly understood results
* RAl-SNPB-3-27
* RAl-SNPB-3     Independent Peer Review References
-Realistic Results
[1]. South Texas Project Electric Generating Station. Preparation of Calculations. OPGP04-ZA-0307, STI 34177871. Rev.7
* RAl-SNPB-3-28
 
-Boundary conditions as prescribed
NOC-AE-16003395 .
* RAl-SNPB-3-29
Attachment 1 Page 31 of65 SNPB-3-18 Mesh Size Sensitivity Please demonstrate that the L TCC results are independent of mesh size for the accident scenarios under consideration.
-Thoroughly understood results
Criterion       4.7     Reference       SRP, lll.3d STP Response:
* RAl-SNPB-3-31
The sensitivity analysis performed with the LTCC EM on the mesh size and nodalization of selected regions of the primary system demonstrates that the LTCC result (PCT) is not dependent on the mesh size and nodalization adopted.
-Independent Peer Review References
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:
[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 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 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 i~ 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:
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.
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 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.
* 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.
Based on the LTCC important phenomena identified and described in the answer to RAl-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 LTCC 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.
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).
NOC-AE-16003395 Attachment 1 Page 32 of65 During the post-blockage LTCC 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.
Under these conditions, liquid water may preferentially enter the core through colder assemblies and subsequently reach the hot assemblies at lower elevations.
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.
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:
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.
* 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
* 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:
* 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
: 1. The hottest rod heat structure
: 2. The hot assembly heat structure
: 2. The hot assembly heat structure
Line 223: Line 509:
: 2. The average channel, thermally connected to the average heat structure.
: 2. The average channel, thermally connected to the average heat structure.
Both nodalization diagrams use 2.1 axial nodes to allow:
Both nodalization diagrams use 2.1 axial nodes to allow:
* Sharp gradients to be resolved during the L TCC phase in core blockage scenarios
* Sharp gradients to be resolved during the LTCC phase in core blockage scenarios
* Fine axial power shapes be simulated.
* 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.
NOC-AE-16003395 Attachment 1 Page 33 of 65 585 r- 7?     585
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}
513                                      513 2                  590                  2 512        595                          512      696 3
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.
X19 865                                      5 521                                      521 2                                        2 3                                        3 4                                        4 535                                    535 Nodalization Diagrams. One-Channel Core (Base Case, Left); Two-Channel Core (Right)
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:
The cases are executed under the same boundary conditions.
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:
The simulation results in terms of PCT are shown below. No appreciable difference is observed between the two nodalization approaches adopted.
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.
NOC-AE-16003395 Attachment 1 Page 34 of 65
The scope of these sensitivities is to study the var i ability of the PCT against the following thermal-hydraulic parameters
                  - -One-Channel Core {Base Case}     - - -
:
* Sump Switchover* One-Cha me I Core                 - - Blockage Time. One-Channet Core
                  -- Tw~Channel  Core                  - - -
* Sump Si.vrtd*1owr- T\".o-Chann@I Core             - - Blockage Time - Tw<H:hannel Core 800 700 600 L.IJ 500 a::
:J t:(  400                                  I a::                                      I
                  -~- ~
:::::::;::soc
    !** ~200 I
I I
I             I I
100                   I I
I I
I I
1000      1500              2000     2500           3000                 3500                             4500              sooo          5500      6000 SIMULATION TIME [SJ Simulation Results: PCT References
[1]. RELAP5-3D Code Manual - Volume V: User's Guidelines
 
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 for 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)     Appendix K decay heat load with single worst failure and steam generator tube plugging b)      Axial power shape c)      Break sensitivity study with appropriate break size resolution d)      No bypass blockage Criterion      4.7     Reference       SRP, l/l.3d STP Response:
Sensitivity studies are conducted using the proposed LTCC EM to analyze the system response under different conditions. These sensitivities are generated from the 16" break LOCA scenario (base case) which is described in the response to RAl-SNPB-02. The scope of these sensitivities is to study the variability of the PCT against the following thermal-hydraulic parameters:
* Decay power
* Decay power
* SI trains availab i lity
* SI trains availability
* Steam Generators
* Steam Generators' tube plugging
' tube plugging
* Core axial power shape
* 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:
* Core bypass availability at the SSO Three sets of sensitivities are conducted and described below:
SENSITIVITY a) Appendix K decay heat load with single worst 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
* 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.
* Single train failure (1 HPSI + 1LPSI + 1CS) . The failure is assumed to occur at the start of the transient. In order to minimize the total ECCS flow available for cooling ,
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).
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.
* SG tube plugging equal to 10% (maximum allowed by STP design) .
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.
SENSITIVITY b) Axial power shape.
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&deg; F}, rep r esent conservative cases where the RHR HX is not modeled. The 190&deg; 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
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:
(" 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 skewe<I -+-rosin e -+-top skewed Sensitivity b): Power Profiles Results NOC-AE-16003395 Attachment 1 Page 37 of65 The results of the sensitivities performed are summarized in the table below. The results are presented in terms of:
b1) Bottom skewed power shape b2) Top skewed power shape
 
NOC-AE-16003395 Attachment 1 Page 36 of 65 SENSITIVITY d) No bypass blockage.
This sensitivity is performed to show the effectiveness of the core barrel/baffle bypass as alternative flow path during a hypothetical full core blockage at the bottom of the core. This case is executed starting from similar conditions described in (a} , assuming a free core bypass during the post-core blockage phase.
The LTCC EM includes LOCA scenarios with full core and core 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 The main boundary conditions used for the proposed sensitivity study are listed below.
The Sump Pool Temperature is assumed to be the RCS injection temperature. The maximum sump temperature values (270&deg; F}, represent conservative cases where the RHR HX is not modeled . The 190&deg; F cases assume RHR HX is available.
Condition                                Base Case          a)             bl )       b2)           d)
Decay Power(%)                               100            120             100         100         120 SI Trains                                    3               2             3           3           2 SG Tube Plugging(%)                           0             10             0           0           10 Axial Power Shape                      Chopped Cosine Chopped Cosine Bottom Skewed Top Skewed Chopped Cosin e Bypass Blockage(%)                           100            100             100         100           0 RWST Volum e (Ga l)                       360.000       453,000         360.000     360,000     360,000 RWST Temperature (" F)                       130           85             130         130         130 ECCS Injection Temperature at SSO ("F)       270            190            270        270         270 Figure below shows the axial power profiles used to run the sensitivities b1 ) and b2).
0.08 0.07 0.06 c
            ~ 0.05 I!
I&. 0.04 lCl 0.03 0.()2 O.Ql 0
0               .5             10             15           20           25 Axial Node
                              -+- bottom skewe<I       -+- rosin e   -+- top skewed Sensitivity b): Power Profiles
 
NOC-AE-16003395 Attachment 1 Page 37 of65 Results The results of the sensitivities performed are summarized in the table below. The results are presented in terms of:
* The SSO time (s)
* The SSO time (s)
* The Core blockage time (s)
* The Core blockage time (s)
* The Maximum PCT reached after the core blockage Case SSOTime (s) Core Blockage Time (s) Max PCT After Core Blockage (&deg;F) Base Case 1740 2100 367 a) 2654 3014 309 bl) 1738 2098 609 b2) 1739 2099 473 d) 2167 2527 336 Comments In all cases analyzed the PCT is shown to remain under the limit of 800 &deg;F during the post-core blockage L TCC phase. Differences in timing of SSO and core blockage are due to the different injection rate and RWST volumes assumed for each case. Subcooling of ECCS injection (availability of the RHR HX) is the most important parameter for the Case a condition.
* The Maximum PCT reached after the core blockage Case     SSOTime (s)   Core Blockage Time (s)   Max PCT After Core Blockage (&deg;F)
Base Case     1740                 2100                         367 a)         2654                 3014                         309 bl)       1738                 2098                         609 b2)       1739                 2099                         473 d)         2167                 2527                         336 Comments In all cases analyzed the PCT is shown to remain under the limit of 800 &deg;F during the post-core blockage LTCC phase.
Differences in timing of SSO and core blockage are due to the different injection rate and RWST volumes assumed for each case.
Subcooling of ECCS injection (availability of the RHR HX) is the most important parameter for the Case a condition.
References
References
[1]. RC09989 RELAP5-3D Steady-State Model Rev.a SNPB-3-21 Important Sources of Uncertainty Attachment 1 Page 38 of65 Please demonstrate that the important sources of uncertainty are appropriately accounted for in the LTCCEM. Criterion 5.3, 5.2 Reference SRP, lll.3e STP Response:
[1]. RC09989 RELAP5-3D Steady-State Model Rev.a
The important sources of uncertainty are identified and 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 These sources of uncertainty are appropriately accounted for in the EM by defining specific parameters in the steady-state and transient input files. The following hydraulic parameters are identifiable in the proposed L TCC EM. These parameters are classified into two categories based on the on the input file where they are defined. Answer to the RAI SNPB-3-22 addresses the margin to the design specs for each of the uncertainties identified, and the sensitivity analysis proposed.
 
NOC-AE-160033~5 Attachment 1 Page 38 of65 SNPB-3-21 Important Sources of Uncertainty Please demonstrate that the important sources of uncertainty are appropriately accounted for in the LTCCEM.
Criterion       5.3, 5.2         Reference       SRP, lll.3e STP Response:
The important sources of uncertainty are identified and 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 These sources of uncertainty are appropriately accounted for in the EM by defining specific parameters in the steady-state and transient input files. The following thermal-hydraulic parameters are identifiable in the proposed LTCC EM. These parameters are classified into two categories based on the on the input file where they are defined.
Answer to the RAI SNPB-3-22 addresses the margin to the design specs for each of the uncertainties identified, and the sensitivity analysis proposed.
Steady-State Input File Reactor Nominal Power The reactor nominal power is defined in word 2 of card 30000001 of the space independent reactor kinetics input card set [1].
Steady-State Input File Reactor Nominal Power The reactor nominal power is defined in word 2 of card 30000001 of the space independent reactor kinetics input card set [1].
Core Heat Structures Thermal Properties NOC-AE-16003395 Attachment 1 Page 39 of65 The core heat structures are defined in the steady-state input file to simulate the fuel assemblies with a lumped approach described in [1]. The heat structure thermal properties are defined in the input file. In particular:
 
NOC-AE-16003395 Attachment 1 Page 39 of65 Core Heat Structures Thermal Properties The core heat structures are defined in the steady-state input file to simulate the fuel assemblies with a lumped approach described in [1]. The heat structure thermal properties are defined in the input file. In particular:
* Fuel thermal property data (material 001): Thermal conductivity and heat capacity defined as function of temperature in table 100 (cards 20100100)
* Fuel thermal property data (material 001): Thermal conductivity and heat capacity defined as function of temperature in table 100 (cards 20100100)
* Gap thermal property data (material 002): Thermal conductivity and heat capacity defined as function of temperature in table 200 (cards 20100200)
* Gap thermal property data (material 002): Thermal conductivity and heat capacity defined as function of temperature in table 200 (cards 20100200)
* Cladding thermal property data (material 003): Thermal conductivity and heat capacity defined as function of temperature in table 300 (card 20100300)
* Cladding thermal property data (material 003): Thermal conductivity and heat capacity defined as function of temperature in table 300 (card 20100300)
Reactor Vessel Passive Heat Structures Reactor vessel passive heat structure are included in the steady-state input file to account for the unheated structures such as reactor vessel walls, core support plates, and other plena internals.
Reactor Vessel Passive Heat Structures Reactor vessel passive heat structure are included in the steady-state input file to account for the unheated structures such as reactor vessel walls, core support plates, and other plena internals. A detailed description of these structures is included in Section 9.3 of [1].
A detailed description of these structures is included in Section 9.3 of [1]. Reactor Core Axial Power Shape Core heat structures are defined to simulate the core power generation during normal conditions and during accident (decay heat). These heat structures are subdivided into 21 axial nodes to provide an adequate axial nodalization to reproduce the desired axial power shape. Axial power shape is defined by specifying power source multipliers.
Reactor Core Axial Power Shape Core heat structures are defined to simulate the core power generation during normal conditions and during accident (decay heat). These heat structures are subdivided into 21 axial nodes to provide an adequate axial nodalization to reproduce the desired axial power shape. Axial power shape is defined by specifying power source multipliers. These multipliers are defined in cards 16050701-21, 16060701-21, and 16061701-21 respectively for the average assembly, hot assembly, and hottest rod [1].
These multipliers are defined in cards 16050701-21, 16060701-21, and 16061701-21 respectively for the average assembly, hot assembly, and hottest rod [1]. Steam Generator Tube Plugging Four independent coolant loops are included in the EM. Each loop is equipped with a steam generator.
Steam Generator Tube Plugging Four independent coolant loops are included in the EM. Each loop is equipped with a steam generator. The model adopted for the steam generator is described in detail in Section 7.2 of [1]. Tube plugging is modeled in the LTCC EM by reducing the nominal total flow area of the steam generators tubes Ounction X07, steam generator tubes inlet; junction X09, steam generator tubes exit; pipe X08, steam generator tubes), and the nominal total heat transfer area defined in the steam generators u-tubes heat structures (defined in cards 11081000, 12081000, 13081000, and 14081000 [1].
The model adopted for the steam generator is described in detail in Section 7.2 of [1]. Tube plugging is modeled in the L TCC EM by reducing the nominal total flow area of the steam generators tubes Ounction X07, steam generator tubes inlet; junction X09, steam generator tubes exit; pipe X08, steam generator tubes), and the nominal total heat transfer area defined in the steam generators u-tubes heat structures (defined in cards 11081000, 12081000, 13081000, and 14081000 [1].
 
Vessel Flow Bypass Fractions NOC-AE-16003395 Attachment 1 Page 40 of65 Major bypass flow paths are included in the L TCC evaluation model. In particular:
NOC-AE-16003395 Attachment 1 Page 40 of65 Vessel Flow Bypass Fractions Major bypass flow paths are included in the LTCC evaluation model. In particular:
* Flow through the spray nozzles into the upper head is modeled with junction 51301 connecting the downcomer top region (component 51202) with the upper head (component 513).
* Flow through the spray nozzles into the upper head is modeled with junction 51301 connecting the downcomer top region (component 51202) with the upper head (component 513).
* Flow entering into the rod cluster control guide thimbles (included in core bypass modeled with pipe 521 and inlet/outlet junctions).
* Flow entering into the rod cluster control guide thimbles (included in core bypass modeled with pipe 521 and inlet/outlet junctions).
Line 271: Line 599:
* Flow in the gaps between the fuel assemblies on the core periphery and the adjacent baffle wall (included in core bypass modeled with pipe 521 and inlet/outlet junctions).
* Flow in the gaps between the fuel assemblies on the core periphery and the adjacent baffle wall (included in core bypass modeled with pipe 521 and inlet/outlet junctions).
* Leakage flow from the downcomer inlet nozzle directly to the vessel outlet nozzle through the gap between the vessel and the barrel (included in core bypass modeled with pipe 521 and inlet/outlet junctions).
* Leakage flow from the downcomer inlet nozzle directly to the vessel outlet nozzle through the gap between the vessel and the barrel (included in core bypass modeled with pipe 521 and inlet/outlet junctions).
Core Nodalization The proposed L TCC EM include a detailed axial nodalization of the core and heat structures to account for the axial and radial power shapes. In the current model the core is modeled using one-dimensional vertical pipe component (pipe 605) and three independent heat structures (6050, 6060, and 6061). Sensitivity analysis ls conducted to study the effect of the core radial nodalization during the L TCC. Upper Head Nodalization Adequate detail in the nodalization of the upper head region of the reactor vessel is included in the EM to account for the flow through this region. Transient (LOCAl Input Files RWST Usable Volume The RWST is modeled with the use of time-dependent volumes 291, 391, and 491. The volume of the RWST is defined in word 6 of trip card 250. Decay Power Model The ANS79-1 option is specified in the space independent reactor kinetic card set 30000000.
Core Nodalization The proposed LTCC EM include a detailed axial nodalization of the core and heat structures to account for the axial and radial power shapes. In the current model the core is modeled using one-dimensional vertical pipe component (pipe 605) and three independent heat structures (6050, 6060, and 6061). Sensitivity analysis ls conducted to study the effect of the core radial nodalization during the LTCC.
This option uses the 1979 ANSI/ANS (ANS-79) Standard.
Upper Head Nodalization Adequate detail in the nodalization of the upper head region of the reactor vessel is included in the EM to account for the flow through this region.
A power multiplier can be also defined within the same card set.
Transient (LOCAl Input Files RWST Usable Volume The RWST is modeled with the use of time-dependent volumes 291, 391, and 491. The volume of the RWST is defined in word 6 of trip card 250.
Break Size The proposed LTCC EM includes analysis of 16", 6", and 2" breaks. Orientation . NOC-AE-16003395 Attachment 1 Page41 of65 The RELAP5-3D code includes entrainment models when simulating a break in a large horizontal pipe and defines possible azimuthal orientations (bottom, top, side). These options can be selected in the L TCC EM by changing the options of the valve component 013 simulating the break. ECCS Flow Rate The L TCC EM includes a detailed model of the ECCS system for the STP units 1 and 2. Three independent safety injection (SI) trains are modeled. Each train consists of a head injection pump, low-head injection pump, and one accumulator.
Decay Power Model The ANS79-1 option is specified in the space independent reactor kinetic card set 30000000. This option uses the 1979 ANSI/ANS (ANS-79) Standard. A power multiplier can be also defined within the same card set.
The containment spray pumps are not explicitly modeled in the EM but are accounted for in the total ECCS flow rate. Flow rates are defined based on ECCS pump characteristics and line pressure drops as function of the primary system local injection pressure.
 
ECCS Injection Temperature The RWST and sump pool are modeled with the use of the time-dependent volumes 291, 391, and 491. The water temperature is defined as function of the time in the cards of these volumes. Core Barrel/Baffle Bypass Blockage Fraction The blockage fraction is defined in the EM by changing the flow area of the valve component 456 simulating the core barrel/baffle inlet. Core Blockage Fraction The blockage fraction is defined in the EM by changing the flow area of the valve component 457 simulating the core barrel/baffle inlet. Plant Set Points and Delays LOCA plant set points and delays are included in the L TCC EM to account for: 11 Automatic signals and delays due to signal processing and actuation
                                                                        . NOC-AE-16003395 Attachment 1 Page41 of65 Break Size The proposed LTCC EM includes analysis of 16", 6", and 2" breaks.
Orientation The RELAP5-3D code includes entrainment models when simulating a break in a large horizontal pipe and defines possible azimuthal orientations (bottom, top, side). These options can be selected in the LTCC EM by changing the options of the valve component 013 simulating the break.
ECCS Flow Rate The LTCC EM includes a detailed model of the ECCS system for the STP units 1 and 2.
Three independent safety injection (SI) trains are modeled. Each train consists of a high-head injection pump, low-head injection pump, and one accumulator. The containment spray pumps are not explicitly modeled in the EM but are accounted for in the total ECCS flow rate.
Flow rates are defined based on ECCS pump characteristics and line pressure drops as function of the primary system local injection pressure.
ECCS Injection Temperature The RWST and sump pool are modeled with the use of the time-dependent volumes 291, 391, and 491. The water temperature is defined as function of the time in the cards of these volumes.
Core Barrel/Baffle Bypass Blockage Fraction The blockage fraction is defined in the EM by changing the flow area of the valve component 456 simulating the core barrel/baffle inlet.
Core Blockage Fraction The blockage fraction is defined in the EM by changing the flow area of the valve component 457 simulating the core barrel/baffle inlet.
Plant Set Points and Delays LOCA plant set points and delays are included in the LTCC EM to account for:
11 Automatic signals and delays due to signal processing and actuation
* Selected manual actions These are defined with the use of trip logics, control logic, and other control variables.
* Selected manual actions These are defined with the use of trip logics, control logic, and other control variables.
CCFL Parameters NOC-AE-16003395 Attachment 1 Page42 of 65 The CCFL model is selected in the EM at locations specified in [1]. The model is enabled at junctions connecting adjacent vertical components.
 
The EM defines the correlation used for each junction where the model is enabled and the coefficients in use (defined in words 2, 3, and 4 of cards 11 O of the selected junctions).
NOC-AE-16003395 Attachment 1 Page42 of 65 CCFL Parameters The CCFL model is selected in the EM at locations specified in [1]. The model is enabled at junctions connecting adjacent vertical components. The EM defines the correlation used for each junction where the model is enabled and the coefficients in use (defined in words 2, 3, and 4 of cards 11 O of the selected junctions).
References
References
[1]. RC09989 RELAP5-3D Steady-State Model Rev.O SNPB-3-22 Uncertainness and Design Margin NOC-AE-16003395 Attachment 1 Page 43 of65 Please provide a discussion on the impact of the uncertainties considered on the important figures of merit (e.g., peak centerline temperature) for each of the accident scenarios and the margin to the design limit. Criterion 5.7 Reference SRP, lll.3e STP Response:
[1]. RC09989 RELAP5-3D Steady-State Model Rev.O
Identification of the Sources of Uncertainty and Impact The important sources of uncertainty are listed in the response to RAl-SNPB-32.
 
The cladding temperature during the core blockage phase of the L TCC is considered as the most important figure of merit and discussion on the potential impact of each source of uncertainty on the PCT is provided with the response to RAl-SNPB-32.
NOC-AE-16003395 Attachment 1 Page 43 of65 SNPB-3-22 Uncertainness and Design Margin Please provide a discussion on the impact of the uncertainties considered on the important figures of merit (e.g., peak centerline temperature) for each of the accident scenarios and the margin to the design limit.
The importance of each source of uncertainty is classified based on its likelihood to impact the PCT at any time after the core blockage event for the LOCA scenarios included in the EM. The following table summarizes the important sources of uncertainty and list them in order of importance.
Criterion       5.7     Reference           SRP, lll.3e STP Response:
Source of Uncertainty Importance Core Blockage Fraction High Core Barrel/ Baffle Bypass Blockage Fraction High Vessel Flow Bypass Fractions High Decay Power Model High RWST Usable Volume High Break Size and Location High Core Nodalization Medium/High ECCS Injection Temperature Medium/High Axial Power Shape Medium Steam Generators' Tube Plugging Medium Upper Head Nodalization Medium ECCS Flow Rate Medium CCFL Parameters Medium Reactor Nominal Power Low Core Heat Structures Thermal Properties Low Reactor Vessel Passive Structures Low Break Orientation Low Plant Set Points and Delays Low Margin to Design Limits Reactor Nominal Power (Importance  
Identification of the Sources of Uncertainty and Impact The important sources of uncertainty are listed in the response to RAl-SNPB-32. The cladding temperature during the core blockage phase of the LTCC is considered as the most important figure of merit and discussion on the potential impact of each source of uncertainty on the PCT is provided with the response to RAl-SNPB-32.
-Low) NOC-AE-16003395 Attachment 1 Page44 of65 The proposed EM assumes that the reactor core is at its nominal thermal power of 3853 MW when the break occurs. Since the reactor initial power is not expected to play a role in the cladding temperature behavior during the L TCC phase, no margin from the nominal value is considered in the EM. Although conservatism is considered at the reactor shutdown.
The importance of each source of uncertainty is classified based on its likelihood to impact the PCT at any time after the core blockage event for the LOCA scenarios included in the EM.
The EM does not take credit for moderator void negative reactivity coefficient the nominal power is maintained until the reactor trips due to the low pressure signal. Delays in the pressure signal processing and control rod drop time are also included.
The following table summarizes the important sources of uncertainty and list them in order of importance.
This maximize the temperature of the fuel during the blowdown phase of the LOCA and, subsequently, the initial energy stored in the fuel. Core Heat Structure Thermal Properties (Importance  
Source of Uncertainty                       Importance Core Blockage Fraction                           High Core Barrel/ Baffle Bypass Blockage Fraction     High Vessel Flow Bypass Fractions                     High Decay Power Model                                 High RWST Usable Volume                               High Break Size and Location                           High Core Nodalization                             Medium/High ECCS Injection Temperature                     Medium/High Axial Power Shape                               Medium Steam Generators' Tube Plugging                 Medium Upper Head Nodalization                         Medium ECCS Flow Rate                                   Medium CCFL Parameters                                 Medium Reactor Nominal Power                             Low Core Heat Structures Thermal Properties           Low Reactor Vessel Passive Structures                 Low Break Orientation                                 Low Plant Set Points and Delays                       Low
-Low) Core heat structure thermal properties are expected to play a role in the PCT behavior during the early phases of the transients but their impact on the PCT during the time after the core blockage (when decay power reaches lower levels) is considered of low importance.
 
No specific margins are included in the EM besides the one described for reactor nominal power. Reactor Vessel Passive Heat Structures (Importance  
NOC-AE-16003395 Attachment 1 Page44 of65 Margin to Design Limits Reactor Nominal Power (Importance - Low)
-Low) Masses of steel representing the reactor vessel walls and other vessel internal structures are included in the EM. The effect of the heat structures is expected to play a role during the early phases of the accident.
The proposed EM assumes that the reactor core is at its nominal thermal power of 3853 MW when the break occurs. Since the reactor initial power is not expected to play a role in the cladding temperature behavior during the LTCC phase, no margin from the nominal value is considered in the EM. Although conservatism is considered at the reactor shutdown. The EM does not take credit for moderator void negative reactivity coefficient the nominal power is maintained until the reactor trips due to the low pressure signal.
Due to depressurization and cooling, the energy stored in these heat structures is expected to decrease during the accident progression being of low importance during the time subsequent the core blockage event. No additional margin is assumed for this source of uncertainty.
Delays in the pressure signal processing and control rod drop time are also included. This maximize the temperature of the fuel during the blowdown phase of the LOCA and, subsequently, the initial energy stored in the fuel.
Reactor Core Axial Power Shape (Importance  
Core Heat Structure Thermal Properties (Importance - Low)
-Medium) Sensitivity analysis is performed with the proposed EM under different axial power shapes to account for this uncertainty.
Core heat structure thermal properties are expected to play a role in the PCT behavior during the early phases of the transients but their impact on the PCT during the time after the core blockage (when decay power reaches lower levels) is considered of low importance. No specific margins are included in the EM besides the one described for reactor nominal power.
In particular, a cosine shape, a bottom skewed profile, and a top skewed profile are considered (see figure below).
Reactor Vessel Passive Heat Structures (Importance - Low)
0.08 0.07 0..06 "' .!ii o.os 0.()4 .. 0..03 0.02 0.01 0 0 s 10 15 20 Axial Node -.-lx>ttom skewed _.,_cosine skewe<I Steam Generators Tube Plugging (Importance  
Masses of steel representing the reactor vessel walls and other vessel internal structures are included in the EM. The effect of the heat structures is expected to play a role during the early phases of the accident. Due to depressurization and cooling, the energy stored in these heat structures is expected to decrease during the accident progression being of low importance during the time subsequent the core blockage event. No additional margin is assumed for this source of uncertainty.
-Medium) NOC-AE-16003395 Attachment 1 Page 45 of 65 IS The base simulation of the EM assumes 0% steam generators' tube plugging.
Reactor Core Axial Power Shape (Importance - Medium)
A sensitivity analysis is included to evaluate the effect o f higher tube plugging (10%) Vessel Flow Bypass Fractions (Importance  
Sensitivity analysis is performed with the proposed EM under different axial power shapes to account for this uncertainty. In particular, a cosine shape, a bottom skewed profile, and a top skewed profile are considered (see figure below).
-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
NOC-AE-16003395 Attachment 1 Page 45 of 65 0.08 0.07 0..06
-Medium/High)
    .!ii o.os
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 -+---+------<
0.()4
(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 -+--+------<
    ~ 0..03 0.02 0.01 0
(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)
0             s           10           15         20       IS Axial Node
Decay Power Model (Importance
                      - . - lx>ttom skewed _.,_ cosine  ~ top skewe<I Steam Generators Tube Plugging (Importance - Medium)
-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  
The base simulation of the EM assumes 0% steam generators' tube plugging. A sensitivity analysis is included to evaluate the effect of higher tube plugging (10%)
-High) The L TCC EM includes scenarios of different break size and investigates:
Vessel Flow Bypass Fractions (Importance - High)
Due to importance of this source of uncertainty, adequcite margin is included in the proposed LTCC EM. In particular, the EM minimizes the availability of the alternative flow paths that may allow cooling water 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 directly 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 simulated 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 periphery 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.
 
NOC-AE-16003395 Attachment 1 Page 46 of 65 Core Nodalization (Importance - Medium/High)
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 nodalization . 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 transients' simulations (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.
585 585 3
2 512                590                                 2                   590 51 2       595 2                                                        3 i
X19 501                 865 X21
                                              .         X19 501             865 X21 Standard PWR Upper Head Nodalization                   LTCC 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 LTCC EM assumes a RWST usable water volume equal to 360,000 gallons.
Volume 43.1''-+.-==::'---
42 . 1 ''~f---                                  550,000 gal -.--.------1 (22,000 gal)   I Volume above High Alarm I 40.9'_........__ _ HI Alarm -       - - - + - 528,000 gal -+---+------<
(55,000 g.iil)         Working Allowance 37.6''_........__  _   lo Alarm - - - - + - 473,000 gal -+-- - t - - ---<
RWST Total Volume (550,000 gal)
Usable volume (398,000 gal) I   Injection Volumes Ou1let nozzle                                                                      Initiation of switchover Vortex 1 4.4'-~-'--.c-- lo-lo Alarm breaker  13' - -                                                                (43,000 gal)       Transfer Allowance 32,000 gal -+--+------<
                    --+- - - - - --;<-- - ' ' - ' 0 gal _ _.___,,_ _ __,
(32,0GO gal)  I Volume at Empty Alarm I inal Values Including instrurMnt uncertainty Minimum required value For the table below one can estimate the margin to the minim um usable volume included in the EM .
This margin is equal to 81 ,000 gallons.
RWST Volumes                          Volume (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                             441,000 Average Usable Volume                           468,500 Usable Volume (LOCA)                           456,735 Injection Volume 413,735 (LOCA)
 
NOC-AE-16003395 Attachment 1 Page48 of65 Decay Power Model (Importance - High)
The ANS79 decay heat model is used in the proposed LTCC EM. Sensitivity analysis is included which uses the Appendix K decay power requirements of +20%.
Break Size (Importance - High)
The LTCC EM includes scenarios of different break size and investigates:
* A large break of 16-inch diameter
* A large break of 16-inch diameter
* A 6-inch diameter break, representative of medium breaks
* A 6-inch diameter break, representative of medium breaks
* A 2-inch diameter break, representative of small breaks Break Orientation (Importance  
* A 2-inch diameter break, representative of small breaks Break Orientation (Importance - Low)
-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.
While this parameter is not expected to have a direct effect on the PCT behavior during the LTCC and core blockage phases, sensitivity analysis is performed to consider different break orientations.
ECCS Flow Rate (Importance  
ECCS Flow Rate (Importance - Medium)
-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.
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.
This source of uncertainty is expected to impact the cladding temperature after the core blockage time. ECCS Injection Temperature (Importance  
ECCS Injection Temperature (Importance - Medium/High)
-Medium/High)
The ECCS injection temperature is:
The ECCS injection temperature is:
* The temperature of the water in the RWST during safety injection phase
* 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
* The temperature of the sump pool during the recirculation phase The RHR HX cooling effect on the ECCS flow is not modeled in the LTCC EM.
(&deg;F) 130 85 Sump Pool Temperature at SSO (&deg;F) 270 188 (2)
The values implemented in the LTCC evaluation model are compared with the best estimate reference in the table below.
Core Barrel/Baffle Bypass Blockage Fraction (Importance  
Parameter                                 LTCC EM     Best Estimate RWST Water Temperature (&deg;F)                   130             85 Sump Pool Temperature at SSO (&deg;F)             270           188 (2)
-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  
NOC-AE-16003395 Attachment 1 Page49 of65 Core Barrel/Baffle Bypass Blockage Fraction (Importance - High)
-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  
In the proposed LTCC EM the core barrel/baffle bypass is assumed 100% block~d. The blockage is assumed to occur instantaneously 360 seconds after the sump switchover time. This is simulated by closing the inlet valve component 456.
-Low) These parameters are set to their nominal value (based on the STP technical specifications) in the L TCC EM. CCFL Parameters (Importance  
Core Blockage Fraction (Importance - High)
-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
In the proposed LTCC 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.
[1]. STI 34280651 RELAP5-3D Software Quality Assurance.
Plant Set Points and Delays (Importance - Low)
Rev.a. RELAP5-3D User's Manual [2]. NUREG/CR-6770 LA-UR-015561, "GSl-191:
These parameters are set to their nominal value (based on the STP technical specifications) in the LTCC EM.
Thermal-Hydraulic Response of PWR Reactor Coolant System and Containments to Selected Accident Sequences", August 2012.
CCFL Parameters (Importance - Medium)
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)
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)
Long term core cooling, including the impact of debris, using the following two step approach:
References
(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
[1]. STI 34280651 RELAP5-3D Software Quality Assurance. Rev.a. RELAP5-3D User's Manual
(&deg;F) assuming a fully blocked core. This is demonstrated by using deterministic analysis performed with RELAP5-3D.
[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.
(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).
NOC-AE-16003395 Attachment 1 Page 50 of65 SNPB-3-23 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:
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:
(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 (&deg;F) assuming a fully blocked core. This is demonstrated by using deterministic analysis performed with RELAP5-3D.
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:
(2)     The cold-leg large break and cold-leg medium break will rely on a risk-informed approach.
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.
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):
As such, OPGP03-ZA-0307 procedurally requires meeting all pertinent elements of the STP Appendix B program including (but not limited to):
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 10 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"
* 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"
* 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"
* Software Quality Assurance for any software used in analyses; OPGP07-ZA-0014 "Software Quality Assurance Program"
. NOC-AE-16003395 Attachment 1 Page 51 of65
 
                                                                  . NOC-AE-16003395 Attachment 1 Page 51 of65
* Contractor and vendor qualification review; OPGP03-ZT-0138 "Contractor/Staff Augmentation Volunteer Qualification Program"
* 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.
* 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?
NOC-AE-16003395 Attachment 1 Page 52 of 65 SNPB-3-24 Input Verification 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
Criterion
* 6.1 Reference SRP, lll.3f STP Response:
* 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.
The input values are compared against the reference source (see response to SNPB         17) and controlled in the STP engineering calculation process required for all safety-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).
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.
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.
NOC-AE-16003395 Attachment 1 Page 53 of65 SNPB-3-25 Proper Convergence 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.
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:
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 STP QAP does not specifically address non-convergence in a reactor safety code application.
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.
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.
NOC-AE~ 16003395 Attachment 1 Page 54 of65 SNPB-3-26 Non-physical Results 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.
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.
Criterion       6.1   Reference     SRP, lll.3f STP Response:
These kinds of failures are typical in all common reactor safety codes and are well-known to experienced analysts.
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).
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.
NOC-AE-16003395 Attachment 1 Page 55 of65 SNPB-3-27 Realistic Results 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?
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.
Criterion       6.1   Reference       SRP, lll.3f STP Response:
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.
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.
Criterion 6.1 Reference SRP, lll.3f STP Response:
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 non-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.
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.
NOC~AE-16003395 Attachment 1 Page 56 of 65 SNPB-3-28 Boundary Conditions as Prescribed 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.
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).
Criterion       6.1     Reference     SRP, lll.3f STP Response:
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.
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.
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.
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.
NOC-AE-16003395 Attachment 1 Page 57 of65 SNPB-3-29 Thoroughly Understood Results 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.
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.
Criterion       6.1     Reference     SRP, lll.3f STP Response:
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 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.
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.
All the main thermal-hydraulic parameters of the primary system over the entire duration of the transient simulation are included.
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 thermal-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.
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.
NOC~AE-16003395 Attachment 1 Page 58 of65 SNPB-3-30 Quality Assurance Program Documentation 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.
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:
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.
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.
NOC-AE-160033~5 Attachment 1 Page 59 of65 SNPB-3-31 Independent Peer Review 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:
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.
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".
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).
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).
NOC-AE-16003395 Attachment 1 Page 60 of65 SNPB-3-32 Important Sources of Uncertainty Please identify the important sources of uncertainty in the L TCC EM.
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).
Criterion       5.1     Reference     SRP, lll.3e STP Response:
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:
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).
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.
Steady-State Model Uncertainties (1) Reactor Nominal Power (Importance- Low)
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.
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 LTCC, and after the core blockage.
(2) Core Heat Structures Thermal Properties (Importance  
(2) Core Heat Structures Thermal Properties (Importance - Low)
-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
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.
[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)
(3) Reactor Vessel Passive Heat Structures (Importance  
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.
-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.
NOC-AE-16003395 Attachment 1 Page 61 of 65 (4) Reactor Core Axial Power Shape (Importance - Medium)
Although effects on the cladding temperature are not expected to be important.
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.
(4) Reactor Core Axial Power Shape (Importance  
(5) Steam Generators Tube Plugging (Importance - Medium)
-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.
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.
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.
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.
(5) Steam Generators Tube Plugging (Importance  
(6) Vessel Flow Bypass Fractions (Importance - High)
-Medium) The liquid level in the primary side of the SGs' tubes is expected to change during the phases of the accident.
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:
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 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.
* 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.
* 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 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.
* Flow in the gaps between the fuel assemblies on the core periphery and the adjacent baffle wall.
(7) Core Nodalization (Importance  
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.
-Medium/High)
(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.
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].
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  
The nodalization of the core, in particular the radial nodalization (number of channels) is
-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  
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.
-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.
(8) Upper Head Nodalization (Importance - Medium)
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.
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.
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.
Transient (LOCA) Model Uncertainties (9) RWST Usable Volume (Importance - High)
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.
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.
(12) Orientation (Importance
This source of uncertainty is expected to impact the cladding temperature after the core blockage time.
-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.
(10) Decay Power Model (Importance- High)
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.
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.
(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.
(11) Break Size (Importance- High)
This source of uncertainty is expected to impact the cladding temperature after the core blockage time. (14) ECCS Injection Temperature (Importance  
The behavior of the primary system during the LTCC (including the time subsequent to the core blockage) is expected to change with the break size.
-Medium/High)
Cladding temperature is expected to be indirectly impacted by the break size.
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.
 
NOC-AE-16003395 Attachment 1 Page 63 of65 (12) Orientation (Importance - Low)
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 LTCC 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 LTCC 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.
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.
(16) Core Blockage Fraction (Importance- High)
NOC-AE-16003395 Attachment 1 Page 64 of65 (17) Plant Set Points and Delays (Importance  
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.
-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:
 
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
* Reactor trip
* RCP trip
* RCP trip
* ECCS pumps start
* ECCS pumps start
* AFW start and MFW shut off
* 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.
* Signals processing As these parameters are expected to affect the early phase of the transient, their impact on the cladding temperature during the LTCC phase is considered low.
The selection of these parameters may affect the conditions at which the CCFL occurs and, subsequently, have an impact on the core coolability.
(18) CCFL Parameters (Importance- Medium)
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
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.
[1]. RCa9989 RELAP5-3D Steady-State Model Rev.a [2]. STI 3428a651 RELAP5-3D Software Quality Assurance.
These sources and their importance are summarized in the table below.
Rev.a. NOC-AE-16003395 Attachment 1 Page 65 of65 Attachment 2 Definitions and Acronyms NOC-AE-16003395 Attachment 2
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
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)
NOC-AE-16003395 Attachment 1 Page 65 of65 References
Stochastic Analysis, also a FHB Fuel Handling Building short name for the CASA GDC General Design Criterion(ia)
[1]. RCa9989 RELAP5-3D Steady-State Model Rev.a
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)
[2]. STI 3428a651 RELAP5-3D Software Quality Assurance. Rev.a.
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)
NOC-AE-16003395 Attachment 2 Attachment 2 Definitions and Acronyms
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)
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)
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)
ASME American Society of                 ECWS     Essential Cooling Water Mechanical Engineers                           System (also ECW)
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
BA   Boric Acid                         EOF       Emergency Operations BAP Boric Acid Precipitation                       Facility BC   Branch Connection                   EOP       Emergency Operating BEP Best Efficiency Point                         Procedure(s)
* 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}}
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               ~SF      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}}

Latest revision as of 00:52, 5 February 2020

Third Set of Responses to Requests for Additional Information STP Risk-Informed GSI-191 Licensing Application
ML16229A189
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 07/21/2016
From: Gerry Powell
South Texas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
GSI-191, NOC-AE-16003395, TAC MF2400, TAC MF2401
Download: ML16229A189 (72)
v • d • e


Text

Soutll li:x.75 Project Electric Gr:11erall11g Statlo11 P.O. Bar 1$9 1%dswortll, T=s 77./$1 July 21, 2016 NOC-AE-16003395 10 CFR 50.12 10 CFR 50.90 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 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 Risk-Informed Approach to Address Generic Safety Issue (GSl)-191 and Respond to Generic Letter (GL) 2004-02", August 20, 2015, NOC-AE-15003241, ML15246A126
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, ML16082A507 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 361-972-8774.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on:

G. T. Powell Executive Vice President ,

and Chief Nuclear Officer awh Attachments:

1. Response to SNPB-3-2,-6, -7, -15, -17, -18, and -20 through -32
2. Definitions and Acronyms

NOC-AE-16003395 Page 3 of 3 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV Morgan. Lewis & Beckius LLP U. S. Nuclear Regulatory Commission Steven P. Frantz, Esquire 1600 East Lamar Boulevard Arlington, TX 76011-4511 U. S. Nuclear Regulatorv Commission Lisa M. Regner Lisa M. Regner Senior Project Manager NRG South Texas LP U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (08H04) Jim von Suskil 11555 Rockville Pike Skip Zahn Rockville, MD 20852 CPS Energy NRC Resident Inspector Kevin Pollo U. S. Nuclear Regulatory Commission Cris Eugster P. 0. Box 289, Mail Code: MN116 L. D. Blaylock Wadsworth, TX 77483 Crain Caton & James. P.C.

Peter Nemeth City of Austin Elaina Bail John Wester Texas Dept of State Health Services Helen Watkins

  • Robert Free

NOC-AE-16003395 Attachment 1 Attachment 1 Response to SNPB-3-2, -6, -7, -15, -17, -18 and -20 through -32

NOC-AE-16003395 Attachment 1 Page 1 of65 Preface to SNPB Responses STPNOC has revised its LTCC approaches to the cold leg SBLOCA and the hot leg LBLOCA that reduces the scope of the RELAP5-3D LTCC 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 LTCC 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 -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)

NOC-AE-16003395 Attachment 1 Page 2 of65 Period 1: Break Event and Slowdown (300 s. - -396 s.)

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 Reactor Trip 308 Blowdown 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.

NOC-AE-16003395 Attachment 1 Page 3 of 65

- Toi lC(S mflawR.nt (e*cluding Accum) - TotM AcctsTiul.ttOt mfJowlWtf' - Corf' Cll 4500 16 4000

,500 12 Vi' 3000 co 2500 to 'i='

u..

w

~ z a:: 2000 0

$ ~

0

-' 1500 w>

u.. 6 -'

Vl w

Vl c:i: 1000

~ HPSI injection start 500 LPSI inj*ction start 500 190 390 440 490 SIMULATION TIME [SJ The core collapsed liquid level reaches a first minim um at 396 seconds.

During this period the primary side of all steam generators is empty.

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

Due to the location of the break, the refill and reflood phases may not be easily distinguishable compared to typical larg e break scenarios in that cold 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 starts increasing.

The injection of the accumulators (accumulators' injection starts at 382 seconds) in the cold legs is preferentially diverted through the core . The collapsed liquid level is partially recovered due to the accumulators' injection. As the accumulators' injection decreases, the core void fraction and liquid mass inventory starts 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 approximately 314 seconds) , provides at total flow over 1500 lbm/s. Most of the SI flow is diverted toward the reactor vessel, and passes through the reactor core, recoveri 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 generators is empty.

Other important events observed during this phase are listed in the table below.

NOC-AE-16003395 Attachment 1 Page 4 of 65 Phase Event Time (s)

AFW Activation 338 Accumulator Activation 383 Refill/Reflood Minimum Core Collapsed Liquid Level 417 Core Full 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 flow 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 appreciable difference is seen between broken loop and intact loops as shown in the figures below.

- Cll SG l , 1' primary s ide - CLL SGl , J.. pr1mciryside - - -

  • Sump Switchover - - Blockage Time 40 35 3-0 f=' 25 L.L UJ 20 UJ 15 10 1000 2000 3000 4000 5000 6000 SIMULATION TIME (SJ SG 1 (Intact Loop, Pressurizer Loop) Primary Side Collapsed Liquid Level

NOC-AE-16003395 Attachment 1 Page 5 of 65

- CLLSG2,1'primarvside - CLLSG2 , .J,,primaryside - - -

  • SumpSwitchovt-r - - SlockageTime 40 35 30 t;: 25

...J w 20 w

...J 15 10 1000 2000 3000 4000 5000 6000 SIMULATION TIME [SJ SG 2 (Intact Loop) Primary Side Collapsed Liquid Level

NOC-AE-16003395 Attachment 1 Page 6 of 65

- - Cll SG3, 't primary side - - CLL SG3, ..J, primary side - - - - Sump Switchover - - Blockage Time 40 35 30

...J LU 20 LU

...J IS 10 0 1000 2000 3000 4000 5000 6000 SIMULATION TIME (SJ SG 3 (Broken Loop) Primary Side Collapsed Liquid Level

- - CLL SG4, 1" pnmary side - - CLL SG4, J, primary side - - -

  • Sump Switchover - - Blockage Time 40 3S 30

...J LU 20 LU

...J IS 10 1000 2000 3000 4000 5000 6000 SIMULATION TIME [SJ SG 4 (Intact Loop) Primary Side Collapsed Liquid Level

NOC-AE-1 6003395 Attachment 1 Page 7 of 65

- - CLLSGl,secondaryslde - - - - SumpSw11chowr - - Bloclc.ageTime - - AFWSGl - MFWSGl 60 70 50 60 Vi'

~

40 50 CXl UJ 40  !<i:

- ' 30 UJ a::

> 5 UJ

-' 30 9u..

20 Vl Vl 20 <t

~

10 10 500 1000 1500 2000 2500 3000 3500 4000 SIMULATION TIME [SJ Intact Loop (Loop 1) Secondary Side Conditions

- - CLL SG3, secondary side - - -- Sump Switchover Blockage Time - - AFWSG3 MFWSG3 60 70 50 60 Vi'

~

40 50 CXl i='

u..

UJ 40  !<i:

- ' 30 a::

UJ UJ s0

-' 30 u..

20 Vl Vl 20 <t

~

10 10 500 1000 1500 2000 2500 3000 3500 4000 SIMULATION TIME [SJ Broken Loop (Loop 3) Secondary Side Conditions

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 Temperature -RWST/StJmp2 Temperat Ufe - - -

  • SumpSwitdiove r - - Blockage Time 2ao LI:' 230 UJ a::

\:t ffi 1ao a..

~

UJ f-80 1000 2000 3000 4000 sooo 6000 SIMULATION TIME [SJ ECCS suction Temperature (Blue Line) and ECCS Injection Temperature (Gray Line)

There is no appreciable effect on the overall primary system behavior due to the increase in the ECCS injection temperature following the SSO time.

Core coolant inlet and outlet temperatures remain subcooled during this phase. After the SSO time, core coolant temperatures increase following 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 similar to the core coolant temperatures as depicted in the figure below.

NOC-AE-16003395 Attachment 1 Page 9 of 65

- PCT - - -

  • SumpSwitchover - - BlockageTime - CorelnTemperature - coreOut Temperature - coreOUtSatT 500

~ 480 u..

0 UJ a::

f-::: 380

<t a::

UJ a..

? 280 UJ t--

180 80 290 490 690 890 1090 1290 1490 1690 1890 2090 SIMULATION TIME [SJ Core Temperatures The full core and core bypass blockage is assumed to occur instantaneously at 2100 seconds (360 seconds after the SSO time).

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

The sudden decrease in the core flow rate due to the instantaneous core blockage at 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 generation in the core. This is also the cause of the increase of the saturation temperature.

The PCT follows the saturation temperature as shown in the figure below (see description of the core heat transfer regimes) .

NOC-AE-16003395 Attachment 1 Page 10 of 65

- PCT - - -

  • St.ntpSw1tcho\iter - - Bfc:di:ag,eTime - C~lnlemper<1tu1e - CoreOutTemperiture - C0teOUtS.tT - CoreCLl 16 BJ 12 G:' 280

'2.... 10 ~

UJ u..

a::

=> z ti: 230 0

a::

UJ ~

a.. >

UJ

~ 6 -I UJ

~ 180 130 80 0 1000 1500 1000 1500 3500 4500 5000 6000 SIMULATION TIME (SJ Core Temperatures and Core Collapsed Liquid Level.

The core collapsed liquid level decreases until a stable value of approximately 8 ft is reached . This is an indication that the reactor core inlet/outlet mass flow is 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 rates 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' primary tubes is observed an instant before the core blockage time.

When the core blockage occurs, a redistribution of the flow within the primary system occurs. The behavior of the SGs immediately before and after the core blockage is described below. The description is complemented with plots of the integral flow rates through each of the loops.

Broken Loop (Loop 3)

The ECCS flow injected in the cold leg 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

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 .

- lnt(HL3*>Vessel} - lnt(CL3->Vessel) - lnt(CL3->Pump.3) - lnt(ECCS->Loop3) - - -

  • SumpSwrtchove r - - Slockage l 1me 8000000 6000000

'iii'

_J UJ 4000000

~

a::

3 2000000 0_J u.

Vl Vl

<{ 5000 6000

~

0 -2000000 UJ

~

a::

\,!) -400JOOO UJ 1-z

-6000000

-llXXJOOO SIMULATION TIME [SJ Broken Loop (Loop 3) Integral Flow Splits Intact Loops Equipped with SI trains (Loops 2 and 4)

The behavior of the intact loops equipped with SI trains appears to be similar during the transient and in particular during the pre- and post-core blockage phases. The ECCS flow injected in the cold leg of intact Loops 2 and 4 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 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 is 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 vessel inlet).

lnt(CLX 7 Pump): Integral flow from CL of Loop X toward the RCP of the same loop (taken 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 slower slope the plot indicates a lower time-average flow rate toward the direction specified. A change in the slope sign is an indication 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 - lnt{ECCS->Loop2) - - -

  • SumpSw1tchover - - BlockageTime 8000000 I

I 500 1000 1500 1000 I 2500 3000 3500 4000 I

I I

-4000000 SIMULATION TIME [SJ Intact Loop (Loop 2) Integral Flow Splits

- lnt(Hl4->Vessel) - lnt{CL4->Vessel) - lnt(ECCS->loop4) - - - - SUmpSw1td10\<er - - Bloc:k*gel1me 8000000 co

....J 6000000 w

~

a:: 4000000 3!

I 0

....J u.

Vl 2000000 Vl

<(

?

Cl w

~ 500 1000 1500 2000 I 1500 3000 3500 4000 a:: I

(.!)

I w

I- I z -2000000 I I

  • 4000000 SIMULATION TIME [SJ Intact Loop (Loop 4) Integral Flow Splits

NOC-AE-16003395 Attachment 1 Page 13 of 65 Intact Loop Not Equipped with SI train (Loop 1)

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 UJ

~

~ 4000000

~ I I a

UJ I

~ 500 1000 1500 2000 I I

2500 3500 4000 4500 a::

(.!) I UJ I zf-- . 200000()

-4000000 SIMULATiON TIME (SJ Intact Loop (Loop 1) Integral Flow Splits Analysis of the case executed also showed that the majority of the flow reaching the top of the core is supplied through the SGs with a small (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 limitation occurs, and how long these conditions are maintained.

NOC-AE-16003395 Attachment 1 Page 14 of 65

- * - coreOUtlet - - Sump SW1tchover - - - - Blockage nme 10000 Vi' UJ

~

i==

_, 1000

<i:

c:: , . - . - . - . ,., .

t!)

UJ ,

  • J f-- .'

z 100 z

0

~ '

f--

u

<i: 10 '*

1 u...

u I u I I

I I

I I l I 300 5300 10300 15300 20300 25300 30300 SIMULATION TIME [SJ Core Outlet CCFL Actuation Integral Time (y-axis log scale)

Core Heat Transfer Regimes during Pre- and Post -Blockage Phases This section describes the heat transfer regimes established during the pre- and post-core blockage phases of the transient.

The heat transfer regime fo r the average assembly, hot assembly and hottest rod are shown in the figures below.

NOC-AE-16003395 Attachment 1 Page 15 of 65

- - HTmode6052 - - HT mode 605 4 - - HT mode 605 6 - - HT mode 605 8 - - HT mode 60511 - - HTmode60514

- - HT mode 605 16 - - MTmode60518 - - HT mode 605 20 - - - - Sump SWitcho~r - - Blockageli me so Vi' Vl

~

z 46 a: 44 w

co

~ 42 - - - --,,-- -+--

z w 40 Cl 0 38 Si111lo-Pho1e Liquid

~ Conv*ction or Subcooled w*ll a: with Void Froction < 0.1) w 36 u..

Vl z 34

<t:

a:

..... 32

~

w 30

r: 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 SIMULATION TIME [SJ Average Assembly Heat Transfer Regimes 2 2

Heat transfer regimes table available in RELAP5-3D user's manual, volume IV, Section 4.2.1

NOC-AE-16003395 Attachment 1 Page 16 of 65

- - HT mode 6060 2 - - HT mode 6060 4 - - HT mode 6060 6 - - HT mode 6060 8 - - HT mode 6060 11 - - HT mode 6060 14

- - HTmode6060 16 - - HTmode6060 18 - - HTmode6060 20 - - -- Sump Switchover - - Blockage Time so Vi' V'l LU 48

.....J t::

z 46 ex: 44 LU co

~ 42 z

LU 40 Subcooled Nucl*at* Boilin&

0 Saturated NudHto Bolllnc 0 38 Sincle-Ph*H Uquld

~ Convection 0< Subcooled wall ex: with Vold Froctlo n < 0.11 LU 36 LL V'l z l4

<(

ex:

I- l2

~

LU 30

I: 1000 1500 2000 2500 3000 3500 4000 4500 sooo 5500 6000 SIMULATION TIME [SJ Hot Assembly Heat Tra nsfer Regimes

*Sump Switchover - - - Blockage Time - - HTmode6061 - HTmode6061 - HTmode6061-06--HT mode606I-OS

- - HTmode6061- ll - - HTmode6061-14 HT mode 6061-16 HTmode6061-18 HT mode 6061-20 so Vi' V'l LU 48

.....J t::

z 46 ex: 44 LU co

~ 42 z

LU 40 0 Subcoo*od Nuc!tiato l!lolllnc 0 38 Saturated NuclHto Bollinc

~ Sincle-PhHo Uquld Conv* ction or Subcooled Wllll ex:

LU 36 wit h Void FrllCtlon < 0.11 LL V'l z

<(

34 ex:

I- 32

~

LU 30

I:

1000 1500 2000 2500 3000 3500 4000 4500 sooo 5500 6000 SIMULATION TIME [SJ Hottest Rod Heat Transfer Regimes As can be seen , the heat transfer in the core before th e SSO time is determined by the large amount of subcooled ECCS flow forced through the core during 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 post-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

~ 220 Vl a._

UJ a: 170

J Vl Vl UJ a::

~ 120 70 20 3300 4:>JO 5300 300 !:>JO 2300 6300

'"'° 8300 9300 SIMULATION TIME [SJ Primary Pressure

NOC-AE-16003395 Attachment 1 Page 18 of 65

- - Tola/CorePa.ver ---- Sump Switchover - - SlockaseTime 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 1000 2000 3000 4000 5000 6000 SIMULATION TIME [SJ Core Decay Power

- - Core Power per Latent Heat - - -

  • SlATlp Switchover - - Blockage Time 200 111) 160 Vi' co

.=., 140 UJ

~ 120 0::

u...

u...

0 100

__J 0

co 80 60 40 1000 2000 3000 4000 5000 6000 SIMULATION TIME [SJ Core Boil Off Rate 3 3

The boil off rate _is calculated by dividing the core decay power by the latent heat of evaporation of water at 150 psia.

NOC-AE-16003395 Attachment 1 Page 19 of 65

- integral Total Break Row - lnt@gralTot.al ECCSFlow - - - - S1.mp Switchover - - Blockag~ Time 8000000 7000000 6000000 V'l 4000000 V'l

<(

~

3000000 2000000 1000000 1000 2000 3000 4000 5000 6000 SIMULATION TIM E [SJ ECCS and Break Integral Flow

- - CorelnletFlow - CoreOotletFlow - - - - SumpSw1tchowr - - BlockageTime 9000 7000 Vi' c__Jo 5000 LU

~

a: 3000

~

0__J u.. 1000 V'l V'l

<(

~

-nxl

-5000 SIMULATION TIME [SJ Core Inlet and Outlet Flow Rate

NOC-AE-16003395 Attachment 1 Page 20 of 65 2000 - BypasslnletFJow - - - - Su'TipSw1tchover - - Bhxkagelime 1500

~

Vl 1000 co I~,-N:--------

,0 0---40 0 ___5_000___6000- -

Vl Vl

<t:

lf H I

1! -1000

- 1500

-2000 SIMULATION TIME [SJ Core Bypass Inlet Flow Rate

- Heat to SGs_secondary - - -

  • Sump Switchover - - Blockage Time U:XX-0 1

£0 3

'° 0 .00h 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 1000 1500 1000 2500 3000 l 500 4000 4500 5000 5500 6000 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.

NOC-AE-16003395 Attachment 1 Page 21 of65 SNPB-3-6 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 LTCC 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 1b 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 LTCC EM. Discussion of the confirmation of the initial and boundary conditions is provided below for each of the four input models included in the LTCC EM.

Steady-State Input Model The steady-state input model defines the initial conditions for all the LOCA simulations included in the LTCC 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 steady-st~te 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 LTCC 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 LTCC 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 LTCC Phase 4: Post-Core Blockage LTCC 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.

NOC-AE-16003395 Attachment 1 Page23 of65 Parameter Value Parameter Value 360000 RWST Usable Volume HHSI Actuation Delay 6s gal Core Blockage Timing 6min LHSI Actuation Delay 10 s (from SSO)

Low PZR pressure Set 1871.7 Hot Leg Injection time 5.5 h Point psi a (from break)

Reactor Trip Signal AFW Delay (from 2s 30s Processing reactor trie)

Rod droe time 2.8 s 130 °F RCP trip: RCS 1444.7 RWST water conditions 14.7psia Pressure Check psi a Sump Pool MFW isolation Delay 2s 270°F Temperature (at SSO)

Accumulator Liquid MFW closure time 10 s 1200 ft3 Volume 600 psia Accumulator Water AFW mass flow rate 70.0 Ibis and 120 Conditions OF 50 psia, 123.4°F-Specific Containment sprays 4282 AFW conditions Enthalpy volumetric flow rate gal/min

= 91.5 Btu/bl A discussion of uncertainties identified for the most critical boundary conditions listed above, the importance associated, and the margins included in the LTCC EM is included in the responses to RAI SNPB 3-32, RAI SNPB 3-21, and RAI SNPB 3-22.

Initial conditions for the LTCC (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.

NOC-AE-16003395 Attachment 1 Page24 of65 References

[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).

NOC~AE-16003395 Attachment 1 Page 25 of65 SNPB-3-7 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 LTCC EM are demonstrated to be appropriate for the simulations of LTCC 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 LTCC and post-core blockage impact on the PCT. A discussion of uncertainties identified, the importance associated, and the margins included in the LTCC 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 LTCC Phase 4: Post-Core Blockage LTCC 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 LTCC phases.

Initial conditions for the LTCC (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 LTCC 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

NOC-AE-16003395 Attachment 1 Page27 of65 SNPB-3-15 Level of Detail 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 LTCC 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 LTCC; 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 LTCC EM is consistent with the best practice used to simulate LOCA scenarios with RELAPS-30.

The proposed LTCC 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 LTCC
  • Post-Core Blockage LTCC 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 LTCC 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

NOC-AE-16003395 .

Attachment 1 Page29 of65 SNPB-3-17 Validation of the Evaluation Model 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 LTCC EM. The PCT response to the heat transfer regimes and flow conditions experienced during LTCC 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 thermal~hydraulic 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 thermal-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 Input Verification
  • RAl-SNPB-3 Proper Convergence
  • RAl-SNPB-3 Non-physical Results
  • RAl-SNPB-3 Realistic Results
  • RAl-SNPB-3 Boundary conditions as prescribed
  • RAl-SNPB-3 Thoroughly understood results
  • RAl-SNPB-3 Independent Peer Review References

[1]. South Texas Project Electric Generating Station. Preparation of Calculations. OPGP04-ZA-0307, STI 34177871. Rev.7

NOC-AE-16003395 .

Attachment 1 Page 31 of65 SNPB-3-18 Mesh Size Sensitivity 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 LTCC EM on the mesh size and nodalization of selected regions of the primary system demonstrates that the LTCC 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 i~ 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.

Based on the LTCC important phenomena identified and described in the answer to RAl-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 LTCC 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 LTCC 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 LTCC phase in core blockage scenarios
  • Fine axial power shapes be simulated.

NOC-AE-16003395 Attachment 1 Page 33 of 65 585 r- 7? 585

~

513 513 2 590 2 512 595 512 696 3

~

X19 865 5 521 521 2 2 3 3 4 4 535 535 Nodalization Diagrams. One-Channel Core (Base Case, Left); Two-Channel Core (Right)

The cases are executed under the same boundary conditions.

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

NOC-AE-16003395 Attachment 1 Page 34 of 65

- -One-Channel Core {Base Case} - - -

  • Sump Switchover* One-Cha me I Core - - Blockage Time. One-Channet Core

-- Tw~Channel Core - - -

  • Sump Si.vrtd*1owr- T\".o-Chann@I Core - - Blockage Time - Tw<H:hannel Core 800 700 600 L.IJ 500 a::
J t:( 400 I a:: I

-~- ~

soc

!** ~200 I

I I

I I I

100 I I

I I

I I

1000 1500 2000 2500 3000 3500 4500 sooo 5500 6000 SIMULATION TIME [SJ Simulation Results: PCT References

[1]. RELAP5-3D Code Manual - Volume V: User's Guidelines

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

Sensitivity studies are conducted using the proposed LTCC EM to analyze the system response under different conditions. These sensitivities are generated from the 16" break LOCA scenario (base case) which is described in the response to RAl-SNPB-02. The scope of these sensitivities is to study the variability of the PCT against the following thermal-hydraulic parameters:

  • Decay power
  • SI trains availability
  • Core axial power shape
  • Core bypass availability at the SSO Three sets of sensitivities are conducted and described below:

SENSITIVITY a) Appendix K decay heat load with single worst 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 + 1LPSI + 1CS) . The failure is assumed to occur at the start of the transient. In 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

NOC-AE-16003395 Attachment 1 Page 36 of 65 SENSITIVITY d) No bypass blockage.

This sensitivity is performed to show the effectiveness of the core barrel/baffle bypass as alternative flow path during a hypothetical full core blockage at the bottom of the core. This case is executed starting from similar conditions described in (a} , assuming a free core bypass during the post-core blockage phase.

The LTCC EM includes LOCA scenarios with full core and core 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 The main boundary conditions used for the proposed sensitivity study are listed below.

The Sump Pool Temperature is assumed to be the RCS injection temperature. The maximum sump temperature values (270° F}, represent conservative cases where the RHR HX is not modeled . The 190° F cases assume RHR HX is available.

Condition Base Case a) bl ) b2) d)

Decay Power(%) 100 120 100 100 120 SI Trains 3 2 3 3 2 SG Tube Plugging(%) 0 10 0 0 10 Axial Power Shape Chopped Cosine Chopped Cosine Bottom Skewed Top Skewed Chopped Cosin e Bypass Blockage(%) 100 100 100 100 0 RWST Volum e (Ga l) 360.000 453,000 360.000 360,000 360,000 RWST Temperature (" F) 130 85 130 130 130 ECCS Injection Temperature at SSO ("F) 270 190 270 270 270 Figure below shows the axial power profiles used to run the sensitivities b1 ) and b2).

0.08 0.07 0.06 c

~ 0.05 I!

I&. 0.04 lCl 0.03 0.()2 O.Ql 0

0 .5 10 15 20 25 Axial Node

-+- bottom skewettom skewed _.,_ cosine ~ top skewe<I Steam Generators Tube Plugging (Importance - Medium)

The base simulation of the EM assumes 0% steam generators' tube plugging. A sensitivity analysis is included to evaluate the effect of 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 particular, the EM minimizes the availability of the alternative flow paths that may allow cooling water 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 directly 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 simulated 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 periphery 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.

NOC-AE-16003395 Attachment 1 Page 46 of 65 Core Nodalization (Importance - Medium/High)

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 nodalization . 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 transients' simulations (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.

585 585 3

2 512 590 2 590 51 2 595 2 3 i

X19 501 865 X21

. X19 501 865 X21 Standard PWR Upper Head Nodalization LTCC 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 LTCC EM assumes a RWST usable water volume equal to 360,000 gallons.

Volume 43.1-+.-==::'---

42 . 1 ~f--- 550,000 gal -.--.------1 (22,000 gal) I Volume above High Alarm I 40.9'_........__ _ HI Alarm - - - - + - 528,000 gal -+---+------<

(55,000 g.iil) Working Allowance 37.6_........__ _ lo Alarm - - - - + - 473,000 gal -+-- - t - - ---<

RWST Total Volume (550,000 gal)

Usable volume (398,000 gal) I Injection Volumes Ou1let nozzle Initiation of switchover Vortex 1 4.4'-~-'--.c-- lo-lo Alarm breaker 13' - - (43,000 gal) Transfer Allowance 32,000 gal -+--+------<

--+- - - - - --;<-- - ' ' - ' 0 gal _ _.___,,_ _ __,

(32,0GO gal) I Volume at Empty Alarm I inal Values Including instrurMnt uncertainty Minimum required value For the table below one can estimate the margin to the minim um usable volume included in the EM .

This margin is equal to 81 ,000 gallons.

RWST Volumes Volume (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 441,000 Average Usable Volume 468,500 Usable Volume (LOCA) 456,735 Injection Volume 413,735 (LOCA)

NOC-AE-16003395 Attachment 1 Page48 of65 Decay Power Model (Importance - High)

The ANS79 decay heat model is used in the proposed LTCC EM. Sensitivity analysis is included which uses the Appendix K decay power requirements of +20%.

Break Size (Importance - High)

The LTCC 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 LTCC 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 LTCC EM.

The values implemented in the LTCC 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)

NOC-AE-16003395 Attachment 1 Page49 of65 Core Barrel/Baffle Bypass Blockage Fraction (Importance - High)

In the proposed LTCC EM the core barrel/baffle bypass is assumed 100% block~d. 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 LTCC 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 LTCC 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.

NOC-AE-16003395 Attachment 1 Page 50 of65 SNPB-3-23 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 10 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.

NOC-AE-16003395 Attachment 1 Page 52 of 65 SNPB-3-24 Input Verification 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 17) and controlled in the STP engineering calculation process required for all safety-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.

NOC-AE-16003395 Attachment 1 Page 53 of65 SNPB-3-25 Proper Convergence 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.

NOC-AE~ 16003395 Attachment 1 Page 54 of65 SNPB-3-26 Non-physical Results 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).

NOC-AE-16003395 Attachment 1 Page 55 of65 SNPB-3-27 Realistic Results 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 non-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.

NOC~AE-16003395 Attachment 1 Page 56 of 65 SNPB-3-28 Boundary Conditions as Prescribed 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.

NOC-AE-16003395 Attachment 1 Page 57 of65 SNPB-3-29 Thoroughly Understood Results 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 thermal-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.

NOC~AE-16003395 Attachment 1 Page 58 of65 SNPB-3-30 Quality Assurance Program Documentation 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.

NOC-AE-160033~5 Attachment 1 Page 59 of65 SNPB-3-31 Independent Peer Review 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).

NOC-AE-16003395 Attachment 1 Page 60 of65 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:

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 LTCC, 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.

NOC-AE-16003395 Attachment 1 Page 61 of 65 (4) Reactor Core Axial Power Shape (Importance - Medium)

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 LTCC (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.

NOC-AE-16003395 Attachment 1 Page 63 of65 (12) Orientation (Importance - Low)

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 LTCC 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 LTCC 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:

  • Signals processing As these parameters are expected to affect the early phase of the transient, their impact on the cladding temperature during the LTCC 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

NOC-AE-16003395 Attachment 1 Page 65 of65 References

[1]. RCa9989 RELAP5-3D Steady-State Model Rev.a

[2]. STI 3428a651 RELAP5-3D Software Quality Assurance. Rev.a.

NOC-AE-16003395 Attachment 2 Attachment 2 Definitions and Acronyms

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 ~SF 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

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

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