ML16054A428

Revision 33 to the Updated Final Safety Analysis Report, Section 14, Plant Safety Analysis

ML16054A428
Person / Time
Site:	Monticello
Issue date:	01/26/2016
From:	Northern States Power Co, Xcel Energy
To:	Office of Nuclear Reactor Regulation
Shared Package
ML16054A376	List: L-MT-16-004, Revision 33 to Updated Final Safety Analysis Report, Drawings NF-36058 Through NH-36243-1, Rev. 82 (Public) ML16054A388 ML16054A413 ML16054A414 ML16054A415 ML16054A416 ML16054A417 ML16054A418 ML16054A419 ML16054A420 ML16054A421 ML16054A422 ML16054A423 ML16054A424 ML16054A425 ML16054A426 ML16054A427 ML16054A428 ML16054A429 ML16054A432 ML16054A434 ML16054A435 ML16054A436 ML16054A438 ML16054A439 ML16054A440 ML16054A441 ML16054A442 ML16054A443 ML16054A444 ML16054A445 ML16054A446 ML16054A447 ML16054A448 ML16155A258 ML16155A259
References
L-MT-16-004
Download: ML16054A428 (177)

Text

SECTION 14

SECTION 1414.1

14.1.1 14.1.2

14.1.3

14.1.4

14.1.5 SECTION 1414.2

SECTION 1414.314.3.1

14.3.2

14.3.3

14.3.4

SECTION 1414.4

14.4.1

14.4.2

14.4.3

14.4.4

14.4.5

SECTION 1414.5

14.5.1

14.5.2 14.5.3

SECTION 1414.6

14.6.1 14.6.2 14.6.314.6.414.6.5

SECTION 1414.7

14.7.1

14.7.2

14.7.3

14.7.4

14.7.5

14.7.6

14.7.7

14.7.8

SECTION 1414.814.8.1

14.8.2

SECTION 1414.1014.10.1

SECTION 1414.11

SECTION 14

SECTION 14AUPDATED SAFETY ANALYSIS REPORT

?l 0 Xcel Energy* 2015 Xcol Enorgy Nuclear Analysi s and Desi gn Page 1 of 19 July 2015 Section Re v is ion: 0 2 19

Revis ion: 0 3 19

R ev i sion: 0 4

se cti ons. A RT S Average Po'vV&r R an ge Mo ni t or , Rod Bl ock M onitor, an d Techni cal Specifi cat ion #

G ard-e I OffiCi al Core Mo nit or ing S ystem fo*r M onti ce ll o MCPR M in imu m Crit i cal Po\'VG r Ratio OPRM Os cill ation Powe r Range Mon itor Revi sion: 0 5 Option A Sc r am Time rep r esen tat ive of the Tech nical Specification re qu i reme n ts

Revi sion: 0 0 I w lo Revi sion: 0 7 19 The fi nal core lo a ding pa ttern an alyzed in th is re port \VaS tr ansmitted to GNF in co rr esponden ce 3). described in Cycle GE14*P1 < < 13 3 781 26 2-17GZ 1 41 n {41 781 27 1 4" {4 338 1 26 ,_ JYS 00 1 04 ,.,.JYY 54 1 Number Of Bundles 0 44 56 24 24 6 Initial Avg.

13 11 l1 3.91 3. 3. 3, 3.89 3.89 Rev i sion: 0 T he min i mum sh ut dovvn margin (SD M) r epo rted bek>w is b ased on mode r ator tem pe r ature of expos u re of11.849 GWD J M'TU corre s ponds to tlile min i mum pr evious cycle expos u re Operations Manua l 8.03.04 .. 05 req u ir es tha t t he m inimum torus water tempe r at u re is greater ca l cul ated SDM resu lt s be l\we n 68°F an d 6S"F is insign ificant as co m pa r ed to t he u ncerta inti es was f ou nd to be suff i cie n t. A co n servative deple ti on s tr ategy was u tilized in t he evalua t ion of st an d by l iquid co ntr ol shutdown margin. The r esuHs p resented are cornsorvatively based on an en d of t ho prevtous reactor, f rom a full po\A/er and m inimum con trol r od invento ry to a sutrc ri tica l cond rtion at any t ime in th e cycle u nder t he most reactive fr ee st ate by the i nject i on of 660 pp m boron.

Rev i sion: 0 9 19 This soction ide ntif ies the t ransient and accident an al yses performed as part of the cu rren t cycle the limitations and condi tions. Th ese c ycle-sp ecific H im ita t ions and are me t for Mo nti cello an other i ssue were eva l ua t ed. Th ese eve nt s are listed be l ow.

1) 2) 3) 4) 7)

Event Pr i mary System Pressure Increase Generator loa d Reiectio n wit h Bypass Fail ure T urt:line T ri p with Bypass Fail ure Mai n Steam I sola t ion Valve CloStxe (One I All Vatves) Tu rtine T ri p with Bypass Fail ure w/o Positio n Scram 1 Main Steam I solation Valve Closure Sc ra m Loss of Condenser Vacuum Reactor Vesse l Water Temperature Decrease I nadvertent HPC I Actuat ron wrth L8 T urbine Tnp Posittve Re activity Insertion Rod Wlthdrawa J Erro r Reactor Vesse l Coolant I nventory Dec r ease Core Coolant Flow Decrease T rip ot One Recirc ula tion Pu mp or Two Rearcu!at ion Pumps Rec i rct.Jation Pump Se i zure. Core Coolant Flow I ncrease-Slow Re ci'ctJa t!on Con t rol Failure -In crease (MCP RF) Slow Reci'CIAa tio n Control Fail ure -I nc rease (MAPlHGR,)" Fas t Recircula No n Contro l Failure -I ncrease Stanup of an I dle Recuculat i on LOop Fuel Loading Errors M isp laced Boodte-Acaclen t 1 Performed ror ASME Vessel Overpressure Com pli ance. Revis ion: 0 Current

./ ./ ./ ./ ./ (il SLO) ./ ./

Revis ion: 0 Reactor fu ll powe r i nitia l cond itions that app ly to th e lim iting transient an al ysis are summarized transio nt an alyses are fisted in Tab le 4.3. Paramete r Value Va l ue Increased Core Low COfe Fklw Flow Ra t ed Thermal Powe r 2004 46.1 Analysis Powe r I Core Flow Ana lysis Reacto r Pressure Time in Cycle Number of*SIRVs for Analysis 5 i n-serv i ce in-service two

Pslg low Vesse l Water L evel Scram 1 %open L ow low Vesse l Water 1 Hig h Pressure Pslg High Reac to r Vesse l Wa ter L eve 1 1 Ps ig Revi sion: 0 1 25 10915 85 85 *55 11 62 11 70 ror be t imes.

Op t ion 1 be be Nollimit in g be 2

to 3

Revi sion: 0 OLMCPR fo r m oasured scram in sertion ti mes. 3 5.2 1 77 1.57 1.62 core monitori ng system use s more de tailed li rrits for e ach f ue l bu nd le latt ice.

Rev i sion: 0 15 111 2.1.2 specifica t ion sta t es that the press ur e measu r ed in the r eactor steam dome sh all no t exceed 1 332 Psig. The p ressure safety limit of 1332 Psig as measured in tho ves sel st eam space was derived fr om t he design pr ess u res of t he r eactor pressure vesse l, st ea m space pip ing, and water spac e piping. The pressure safety lim it was ch ose n as the lowo r of th e pr ess u re 1 332 i nitial co n ditions:

• 100% (2004 1 105%

(60.5 x10 61010.0 1 170.0 1320 T ec h nical SpeciflCat i on li mi t of 1332 Ps i g. Th e calc ulated maximum st eam line pr essure of 1314 1 332 134 4 1 375 Monticello uses TRACG for the MSIV cmure-Wi too ut Position Sc r am analysis.

Thi s ana!y sis is rt11 at 1 00%

1 (1010 psig) and uncerta i nty are applied as a pressure adde r to t he res ul t of the event to obtain the fina l r eported

Revi sion: 0 rod th ermal*mechanica l des i gn and l icens i ng basis , wit h resped to both steady state operations, and transien t and accident eve nts. Th ermal overpower lim its arc de fin ed to eva l uate t he pote nti al for fuel ce n te rlin e metting. Mec h anical overpower lim its are deftned to eva l uate the pot entia l for f ue l cladding ove rs train. Fe edwater Cont r oUe r Fa aure e vent mecha ni cal design and li oens i ng basis crite r ia for th e plant.

Revision:

0 To pr ovide Monticello Nu cl ear Generating Pl ant -with opera t in g im pr ovemonts, expanded operating do main an:a l yses were perfor med in Refere n ce 1 fo r maximum extended l oad line lim it r eload an alyses in Reference 1.

Revi sion: 0 A reload DSS.CO eva lu ation h as beo n performxt in accorda n ce Volith t ho l icensing rnGthodolog y 0) tr ip l in ear segment.

povve r int ercop t WasP.. TRIP Was.p.sREAK RDF-Rec i rc u la ti on Drive F'low

Revi sion: 0 SECTION 14

SECTION 1414.1

14.1.1 14.1.2

14.1.3

14.1.4

14.1.5 SECTION 1414.2

SECTION 1414.314.3.1

14.3.2

14.3.3

14.3.4

SECTION 1414.4

14.4.1

14.4.2

14.4.3

14.4.4

14.4.5

SECTION 1414.5

14.5.1

14.5.2 14.5.3

SECTION 1414.6

14.6.1 14.6.2 14.6.314.6.414.6.5

SECTION 1414.7

14.7.1

14.7.2

14.7.3

14.7.4

14.7.5

14.7.6

14.7.7

14.7.8

SECTION 1414.814.8.1

14.8.2

SECTION 1414.1014.10.1

SECTION 1414.11

SECTION 14

SECTION 14AUPDATED SAFETY ANALYSIS REPORT

?l 0 Xcel Energy* 2015 Xcol Enorgy Nuclear Analysi s and Desi gn Page 1 of 19 July 2015 Section Re v is ion: 0 2 19

Revis ion: 0 3 19

R ev i sion: 0 4

se cti ons. A RT S Average Po'vV&r R an ge Mo ni t or , Rod Bl ock M onitor, an d Techni cal Specifi cat ion #

G ard-e I OffiCi al Core Mo nit or ing S ystem fo*r M onti ce ll o MCPR M in imu m Crit i cal Po\'VG r Ratio OPRM Os cill ation Powe r Range Mon itor Revi sion: 0 5 Option A Sc r am Time rep r esen tat ive of the Tech nical Specification re qu i reme n ts

Revi sion: 0 0 I w lo Revi sion: 0 7 19 The fi nal core lo a ding pa ttern an alyzed in th is re port \VaS tr ansmitted to GNF in co rr esponden ce 3). described in Cycle GE14*P1 < < 13 3 781 26 2-17GZ 1 41 n {41 781 27 1 4" {4 338 1 26 ,_ JYS 00 1 04 ,.,.JYY 54 1 Number Of Bundles 0 44 56 24 24 6 Initial Avg.

13 11 l1 3.91 3. 3. 3, 3.89 3.89 Rev i sion: 0 T he min i mum sh ut dovvn margin (SD M) r epo rted bek>w is b ased on mode r ator tem pe r ature of expos u re of11.849 GWD J M'TU corre s ponds to tlile min i mum pr evious cycle expos u re Operations Manua l 8.03.04 .. 05 req u ir es tha t t he m inimum torus water tempe r at u re is greater ca l cul ated SDM resu lt s be l\we n 68°F an d 6S"F is insign ificant as co m pa r ed to t he u ncerta inti es was f ou nd to be suff i cie n t. A co n servative deple ti on s tr ategy was u tilized in t he evalua t ion of st an d by l iquid co ntr ol shutdown margin. The r esuHs p resented are cornsorvatively based on an en d of t ho prevtous reactor, f rom a full po\A/er and m inimum con trol r od invento ry to a sutrc ri tica l cond rtion at any t ime in th e cycle u nder t he most reactive fr ee st ate by the i nject i on of 660 pp m boron.

Rev i sion: 0 9 19 This soction ide ntif ies the t ransient and accident an al yses performed as part of the cu rren t cycle the limitations and condi tions. Th ese c ycle-sp ecific H im ita t ions and are me t for Mo nti cello an other i ssue were eva l ua t ed. Th ese eve nt s are listed be l ow.

1) 2) 3) 4) 7)

Event Pr i mary System Pressure Increase Generator loa d Reiectio n wit h Bypass Fail ure T urt:line T ri p with Bypass Fail ure Mai n Steam I sola t ion Valve CloStxe (One I All Vatves) Tu rtine T ri p with Bypass Fail ure w/o Positio n Scram 1 Main Steam I solation Valve Closure Sc ra m Loss of Condenser Vacuum Reactor Vesse l Water Temperature Decrease I nadvertent HPC I Actuat ron wrth L8 T urbine Tnp Posittve Re activity Insertion Rod Wlthdrawa J Erro r Reactor Vesse l Coolant I nventory Dec r ease Core Coolant Flow Decrease T rip ot One Recirc ula tion Pu mp or Two Rearcu!at ion Pumps Rec i rct.Jation Pump Se i zure. Core Coolant Flow I ncrease-Slow Re ci'ctJa t!on Con t rol Failure -In crease (MCP RF) Slow Reci'CIAa tio n Control Fail ure -I nc rease (MAPlHGR,)" Fas t Recircula No n Contro l Failure -I ncrease Stanup of an I dle Recuculat i on LOop Fuel Loading Errors M isp laced Boodte-Acaclen t 1 Performed ror ASME Vessel Overpressure Com pli ance. Revis ion: 0 Current

./ ./ ./ ./ ./ (il SLO) ./ ./

Revis ion: 0 Reactor fu ll powe r i nitia l cond itions that app ly to th e lim iting transient an al ysis are summarized transio nt an alyses are fisted in Tab le 4.3. Paramete r Value Va l ue Increased Core Low COfe Fklw Flow Ra t ed Thermal Powe r 2004 46.1 Analysis Powe r I Core Flow Ana lysis Reacto r Pressure Time in Cycle Number of*SIRVs for Analysis 5 i n-serv i ce in-service two

Pslg low Vesse l Water L evel Scram 1 %open L ow low Vesse l Water 1 Hig h Pressure Pslg High Reac to r Vesse l Wa ter L eve 1 1 Ps ig Revi sion: 0 1 25 10915 85 85 *55 11 62 11 70 ror be t imes.

Op t ion 1 be be Nollimit in g be 2

to 3

Revi sion: 0 OLMCPR fo r m oasured scram in sertion ti mes. 3 5.2 1 77 1.57 1.62 core monitori ng system use s more de tailed li rrits for e ach f ue l bu nd le latt ice.

Rev i sion: 0 15 111 2.1.2 specifica t ion sta t es that the press ur e measu r ed in the r eactor steam dome sh all no t exceed 1 332 Psig. The p ressure safety limit of 1332 Psig as measured in tho ves sel st eam space was derived fr om t he design pr ess u res of t he r eactor pressure vesse l, st ea m space pip ing, and water spac e piping. The pressure safety lim it was ch ose n as the lowo r of th e pr ess u re 1 332 i nitial co n ditions:

• 100% (2004 1 105%

(60.5 x10 61010.0 1 170.0 1320 T ec h nical SpeciflCat i on li mi t of 1332 Ps i g. Th e calc ulated maximum st eam line pr essure of 1314 1 332 134 4 1 375 Monticello uses TRACG for the MSIV cmure-Wi too ut Position Sc r am analysis.

Thi s ana!y sis is rt11 at 1 00%

1 (1010 psig) and uncerta i nty are applied as a pressure adde r to t he res ul t of the event to obtain the fina l r eported

Revi sion: 0 rod th ermal*mechanica l des i gn and l icens i ng basis , wit h resped to both steady state operations, and transien t and accident eve nts. Th ermal overpower lim its arc de fin ed to eva l uate t he pote nti al for fuel ce n te rlin e metting. Mec h anical overpower lim its are deftned to eva l uate the pot entia l for f ue l cladding ove rs train. Fe edwater Cont r oUe r Fa aure e vent mecha ni cal design and li oens i ng basis crite r ia for th e plant.

Revision:

0 To pr ovide Monticello Nu cl ear Generating Pl ant -with opera t in g im pr ovemonts, expanded operating do main an:a l yses were perfor med in Refere n ce 1 fo r maximum extended l oad line lim it r eload an alyses in Reference 1.

Revi sion: 0 A reload DSS.CO eva lu ation h as beo n performxt in accorda n ce Volith t ho l icensing rnGthodolog y 0) tr ip l in ear segment.

povve r int ercop t WasP.. TRIP Was.p.sREAK RDF-Rec i rc u la ti on Drive F'low

Revi sion: 0