ML20064A562
| ML20064A562 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 09/04/1990 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20064A560 | List: |
| References | |
| NUDOCS 9009140104 | |
| Download: ML20064A562 (18) | |
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' -ATTACHMENT I TECIINICAL SPECIFICATION CIIANGES t
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- si c REAcios TRIP SYSTEM IN51NSENTAT10N TRIP SEIP0lNTS:
AtteltetE VALUES 7 FUNCTIONAL UNIT TRIP SETPOINT i c N. A. - ; 5 1. Manual Reactor Trip N.A. A 2. Power Range, Neutron Flux Low Setpoint 1 25% of RATED Low Setpoint 125 of RATED.
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g THEmmt. POWER 10Emmt. POKR . m E High Setpoint 1 1995 of RATED High Setpoint 1 11 5 of RAlts m THEment. POWER THEWEL POWER
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- 3. Power Range, Heutron Flum, 1 5% of RATED TIEN EL POWER with i 5.5% of RATED THENRIL POWER 1 High Positive Rate a time constant i 2 seconds with a time constant 1 2 seconds
- 4. Power Range, flestron Flux, i 5% of RATED THEWEL POWER with i 5.SE of: RATED THEMIAL POWER ~ '
High Negative Rate a time. constant 1 2 seconds with a time constant > 2 seconds-
"' 5. Intermediate Range, fleutron < 25% of RATED INEN14L P0bER- < 30E of RATED TIENERL POWER
.i Flux
- 6. Source Range, IIeutron Flux = ~ 1 105 counts per second i 1.3 x 105 counts per second:
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- 7. Overtemperature AT See Note 1 See Note 2 See Note d l g
- 8. Overpower ai
> 1935 psig - 1
- 9. Pressurizer 9ressure--Low > 1945 psig gg -
" " 10. Pressurizer Pressure--High 1 2385 psig~ 1 2395 psig _ ,
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-- 11. Pressurizer Water Level--Migh .1 92% of instrument span- i 93K'of lastrument' span-i
- 12. Low Reactor Coolant Flow > 90E of i: 2- flew per loop *' > 88.8K of. des 4gn flew per loop" l-M4m messwed Mn,%~ - mess red ,
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__ 2~ . 1' TABLE 2.2-1 (Continued). M E REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS -
'", -NOTATION (Continued)- -
h (Continued) NOTE 1: ts
= Time constant utilized in the measured T,,g lag compensator, is 1 2 see 1 at RATED THERMAL POWER, T' = $ 588.2*F Reference T,,
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3 _ P = Pressurizer pressure, psig - P' = 2235 psig (Nominal RCS. operating pressure), , 5 = Laplace transform operator, sec 8
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and f (al) is a function of the indicated difference between top and botton, detectdrs of the power-range nuclear ion chambers; with gains to be selected based on measured ins; ~Joent response during plant startup tests such that:
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(i) for.qt ~9b between -29% and M ; f (AI) = 0, where q andqgarepercent. RATED .,. g g THERMAL POWER in the top and bottom halves of the-core respectively, and tq '* 9b is total THERMAL POWER in percent of. RATED THEIMAL POWER; lii) for.each percent that the magnitude of qg qb exc
- .. the AT Trip 5etpoint -- ,[$ .
5S shaII be automatically reduced by 3.151% of its value at RATED THERNAL POWER; .and - xx
+10%
PP for each percent that the magnitude of qt - qb exceeds @dE, _ the AT. Trip Setpoint : (iii) . shall be automatically reduced by 1.50% of its _value at RATED THERMAL POWE 22 _.
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FLOW PER- LOOP. = 96250 gpm 333 UNACCEPTABLE : OPERATION ,! 2400 880- Ds 1,
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FIGURE 2.1-1 UNITS 1 AND 2 i . P.EACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION L McGUIRE - UNITS 1 and 2 2-2a Amendment No. (Unit 1) Amendment No (Unit 2)
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bE ' " .1 ' ATTACHMENT II JUSTIFICATION. Over the course of time, degraded steam generator tubes have been plugged or sleeved. resulting in a reduction of reactor coolant system flow. LAs a result of steam' generator tube plugging during the refueling outage of-McGuire Unit 2,-reactor coolant system (RCS) flow may decrease further.
~!' - This-reduction in flow may make it difficult to meet the (Table 2.2-1 Item 12 'as annotated) flow required by technical specifications to reach 100%
- power 1To alleviate this concern, analyses have been performed to justify a 1% reduction in RCS flow. These analyses show that the reduction in flow-would not have a'significant impact on any of the accident analyses presented in Chapter 15 of, the Final Safety Analysis Report (FSAR). . In addition, none of the analyses presented in Chapters 3, 4; or.6 are significantly'affec+ad by this change. Therefore, the proposed change in.
minimum measure" flow will not result in a reduction of safety margin. Another sinor, administrative change is being made to delete obsolete refer. aces to the RTD bypass manifold, which has been removed'from each unit. . Removal of the manifolds was approved by the NRC staff by Facility _ Operating License Amendment Numbers 84 (Unit 1) and 65'(Unit 2). Another minor change, of an editorial nature, involves the replacement of the term " design flow".with'" minimum measured flow." This will correct-the parameter of. interest. The minimum measured flow differs.from thermal design flow by an amount equal to instrument uncertainty. Therefore, while the design flow may be' inferred from the measured flow, the technical specification more correctly applies to the measured flow. s (
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i COMPARISON OF EFFECT OF REDUCED FLOW ON . FSAR ANALYSES-A . 5 4 o, i i s 3 i n-i l= a ,, t ;' ! a' l >. l-- i t,- l I i p. i U 5 l l i.
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m jij % ,. . l Blowdown Reactor Vess_el 'And loop' Forces (FSAR' Sections 3.6.4.1 and 3.9.1.5) ; With a reduction in Thermal Design Flow (TDF), minor changes could be expected j in LOCA reactor vessel forces, primarily as a resul_t of a slight decrease- in ~ L cold leg temperature. For the reduction In TDF from 95500 to 94500 gpm and
, the less than one degree cold leg terrperature change, the vessel "id core L barrel horizontal forces has been estimated to increase less than 0.4%. This change is~ covered by. margin in the structural analysis and the reduction in ,
thermal design flow rate as indicated above has no adverse' effect on vessel '
, internals, core components, and coolant. loop piping structural adequacy. -
l Thermal Hydraulic __Desian (FSAR Section 4.4) ,
; The thermal hydraulic design for McGuire Units 1 and 2 was analyzed for a 1%
reduction in minimum measured flow from 388880 gpm to 385000 gpm and a reduction in thermal design flow from 382000 gpm to 378000 gpm._ The reduced , flow rate resulted in a slight reduction in core DNB limits, and therefore a - replacement Technical Specification Figure 2.1-1, Reactor Core Safety Limit - Four Loops in Operation, was generated to reflect the lower flow. The Axial Flux Difference limits, Technical Specification Section 3.2.1 are unchanged,- t and all of the current thermal hydraulic design criteria bre satisfied at the- i
-reduced flow conditions.
As previously noted above, revised core thermal limits were generated re-flecting the reduced RCS flow of 385000 gpm (minimum measured flow). Based upon _these new protection limits, it was determined that the current Over-temperature and Overpower AT (OTAT/0Pt.1) setpoint equation constants (see Note 1 of Table 2.2-1) for McGuire are conservative and provide the necessary protection. However, the dead band of the f( AI) function had to be revisec: ' from its current range of -29% > q t -q b to -2 W > qt - qb > 7%.
, Containment functional Deslan (FSAR Section 6.2.1) 4 The pertinent ef fect of a reduction in RCS flow is the resultant impact on RCS !
temperature. A review was performed by Westinghouse relative to the , assumptions used for the following McGuire FSAR containment analyses: s 6.2.1.3 Mass and Energy Release Analysis for Postulated Loss of Coolant Accidents 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary u System Pipe Ruptures Inside Containment 1
,4 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies of ECCS This review concluded -that the proposed reduction in flow will have a negligible effect on .the RCS temperature used in the containment analyses.
Therefore, the results of those analyses remain valid.
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% Acci6nt' Analysis '(FSAR Chapter: 15)
>Thf following'section spec;fically addresses _the impact of a reduction inLthe.
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1RCS ' flow on; the transients presented i in - Chapter - 15 of . '.he- McGuire FSAR.' .
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m basis ' for thel evaluations was' that' a.1%- reduction' in tha RCS flow would not~- significantly affect D the: system transient responselfor any event.- This is_ i ~s . , 5
-valid- since the systec. transient response . is primarily governed' byf the initiatirg ; event i .e. , an -increase or decrease -in the c secondary . side ' heat
~ , ; > removal,f a sudden.decreasc in the RCS flow rate, a reactivity anomaly, or by.
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a- change in the _ reactor coolantEinventory. The system transient: response is 3 not' affected bycthe initial RCS flow assumption, unless the reduction in the ; 4@ 3e initiali RCSf flow 'is suf ficiently large' to impiir the " steady-state. core heat-
> transfer or' steam 'gener'ator heat transfer capability. This ' is. not. theicase r
i)M forg a 1% ' reduction in RCS flow. 'In addition, for the events in which 'a; }f decrease in RCS 1 flowdoccurs, the flow co stdown is ; determined- by the pump Therefore,; the g'jy characteristics andonot; the Aitial RCS flow assumption. . f ' system transient; response for aach of the non-LOCA events discussed below : g remains _ unchanged.
, Feedwater System Malfunction that Results in ; fecrasse in' Feedwater j ', mTemperature (FSAR Section 15.1.1)
( %^ ?This ANS Cnndition' 11 event is bounded by "Feedwater Sys em Malfunction'.that [ ; Result' in anf Increase in . Feedwater ' flow" (15.1.2). The a fety. analysis 'ONB T y, design basis is met and the conclusions of the FSAR remain 'alid.
$. Feedwater System Malfunctions that Result in an Increase in Feedwater Flow (FSAR Section 15.1.2)
For this ANS Condition II event, cases are analyzed for both full power and g;b ' c zero power conditions. The zero power case, as discussed in the FSAR, is 3 bounded .by)" Uncontrolled RCCA Bank Withdrawal from a Sabcritical or Low Power i Startup Condition" (15.4.1).
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1 For the full power case, the transient is g effectively terminated by a turbine trip and feedwater isolation on high-high 14 steam .gmerator..lovel. This analysis assumed a minimum measured flow of- , j' f 393600 gpm. If a ' MMF of 385000 gpm were assumed, the system transient f respor ,e would not be signif t.antly affected. Since the revised core: limits j' reR _ ting the reduced RCS flow are not exceeaed at the most limiting point in
- g. t> transient, the~ safety analysis DNBR limit 'is met. Therefore, the f (. m iusions'of the FSAR remain valid.
$ 1 L .y E_guipment Malfunction or' Oyerating Failure that Results in Increasino-pr Steam Flow (FSAR Section 15.1.3) m:.y S
' For' this ANS Condition 11 event, cases are analyzed at beginning and end of life conditions both with and without automatic rod control. In all cases,
!;f the' transient approaches an equilibrium condition and a reactor trip does not result. This analysis assumed a minimum measured flow of 393600 gpm. If a % MMF of 385000- gpm .were assumed, the system transient response would not be significantly affected. Since the revised core limits reflecting the reduced RCS flow are not exceer.ed at the most limiting point in the transient,' the safety analysis - DNBR 'imit is met. Therefore, the conclusions of the FSAR remain valid. L l h{ Np
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Inadvertent" 0)ening of: aT Steam Generator Relief or - Safety Valve'(FSAR . U [u Section-15.1.4)= ,' o This ANS' Condition II. event is bounded by " Spectrum of Steam 1 Piping Failures 4 :Inside 'and Outside Containment" (15.1.5), since the Condition II DNB criterion ,
. is applied to that Condition TIV event, and since both are uncontrolled steam line depressurizations. The safety ' analysis DNB design basis is met 'and the
= conclusions of the FSAR remain valid.
- 3 Spectrum of Steam' Pipino Failures Inside and' 0utside Containment (FSAR ' ;
Section 15.1.5)- 'i For this ANS Condition IV event,.the ANS Condition 11 criterion of meeting the. ; a DNB limit is applied. ' The analyses for both with the without' offsite power - are performed ;at.zero power conditions and assume peaking factors consistent . with the' most reacthe kCCA stuck oct of the core. The current licensing :r ' basis' analysis w6s performed for the UHI Elimination effort and assumed a L thermal dwsign, flow of 377000 gpm which~1s more conservative than'the reduced F TDF or 378000 gpm. Therefore, the safety analyses DNBR limit is met and the ' conclusions of-the FSAR remain valid, For the steamli_ne~ breaks occurring outside . containment, the analysis were-
- performed ri zero-power conditions, consistent with the analysis presented in
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g E- 15.1.'1. . A > hetion in the RCS flow reduces -the primary to secondary heat transfer and thus minimizes the return to power. The ef fect of a postul?ted ~
. secondary system pipe rupture. outside- containment is that an adverse environment is created. Decause of the adverse environment, various ,
< components, .such m the steam generator power-operated reHef valves, .are (
assumed to ' fail . or not operate as designed. Therefore, the ef fect that the. s various consequential . component failures would have on. the steamline' break B ' core response event was examined (see Reference 6 which-presented the results- , 19 ~o f. analyses performed in-1985). It was determined in a review of Reference 6' , that the 'steamline break currently presented in FSAR Section 15.1.5 remains bounding. ,
- Loss of External Electrical Load (FSAR Section 15.2.2) p .
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-This ANS Condition 11 event is bounded by " Turbine Trip'? (15.2.3). The safety.
analysis DNB design basis is met and the conclusions of the FSAR remain valid. 9 Turbine Trip (FSAR Section 15.2.3)
-f For this ANS Condition II event, cases aia analyzed at beginning and end of L ' life conditions both with and without pressurizer control. This analysis assumed a minimum measured flow of 388886 gpm. If a MMF of 385000 gpm were
- assumed, the system transient response would not be significantly affected.
n Since the revised core limits reflecting the reduced faS flow are not exceeded 6 at the most limiting point in the transient, the safety analysis DNBR limit is i' not. Therefore, the conclusions of the FSAR remain valid. c p
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In' advertent Closure of MSIVs (FSAR' Section' 15.2.4)' Wf cThis ANS Condition 11 event is bounded by " Turbine Trip" (15.2.3). _ThM safety _
'#' : analysis DNB design l basis is met,and the conclusions of the FSAR remain valid.
un ~ . . 1 ,_ Loss of Non-Emergency A-C Power toLthe Station Auxiliaries (FSAR;Section. W ,
- 15. 2. 6 ) --
Y This ANS Condition II event is: analyzed _ to show -that adequate heat removal'
?' ' capability 1 exists _ to remove n core decay heat and ' stored energy ' following -
,n reactor trip. The analysis presented 'in the FSAR. assumed a thermal' design; ,
, flow of 377000 gpm which. is more conservative than the reduced TDF of '378000
$p gpm 'Therefore, all the safety criteria are satisfied- and the conclusions- of'
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g the FSAR remain valid. g loss of Normal Feedwater (FSAR Section 15.2.7)
'Thist ANS Condition ~ II ' event is' analyzed to show that adequate- heat removal-capability ; exists -to remove core decay heat and stored. energy following The analysis presented in the FSAR' assumed a thermal design-
' reactor trip.
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' flow of 377000 gpmtwhich is more conservative than the reduced TDF of'378000 9
gpm. Therefore, all 'the safety criteria' are satisfied and the conciusions of the FSAR remain valid. Feedwater System Pipe Break (FSAR Section 15.2.8)- This ANS Condition lV event is analyzed to show that the core remains in place and ' geometrically intact with n loss of core cooling capability. This is ensured by applying;the: criterion that no culk boiling occut :in-the hot leg. 11he analysis presented in the FSAR assumed a' thermal design flow of-377000 -gpm y which is=more conservative than the reduced TDF of 378000 gpm. Therefore, all the r sa fety criteria are . satisfied and the conclusions of the FSAR remain y! valid. Partial Loss of- Forced Reactor Coolant Flow (FSAR _ Section 15.3.1)
'For this ANS Condition 11 event, the transient is terminated by a low RCS-loop flow reactor trip. This analysis assumed a minimum measured flow 'of 388880 gpm. 'If a MMF of 385000 gpm were assun.ed, the system transient response would j'
not be significantly; af fected. Therefore, the li:niting statepoint for this event remains- unchanged and was evaluated for the . lower flow. The safety j analysis DNBR limit is met'and the conclusions of the FSAR remain valid,
' Complete loss of Forced Reactor Coolant Flow (FSAR Section 15.3.2) ;
1 For -this ANS Condition II event, the transient i s. te rmi.nated by an undervoltage or underfrequency reactor trip. This analysis assumed a minimum .i measured flow of 388880 gom. If a MMF of 385000 gpm were assumed, the system transient response would not be significantly affected. Therefore, the limiting statepoint for this event remains unchanged and was evaluated for the
- lower flow. The safety analysis DNBR limit is met and the conclusions of the
'FSAR remain valid.
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f :jW ja __a Reactor Coolant Pung Shaft Seizure (FSAR Section 15.3.3) 1Th'is . ANS Condition IV event - is ^ analyzed -to demonstrate . that Lthe peak. clad temparature=doeslnot: exceed 2700 F and the RCS preseure remains below'110s of
- the. designs (110 x - 2485 psig .= 2733.5 - psig). The ' analysis' presented in . the .
FSAR'. assumed a thermal 1 design flow of 37_7000 gpm which is' more conservative
'than the: reduced TDF of.378000 gpm. ~ Therefore, all the safety criteria are-satisfied and the conclusions of the FSAR remain valid..
Reactor Coolant Pump Shaf t Brea_b (FSAR Section 15.3.4)-
. This ANS Condition IV event is bounded by " Reactor Coolant' Pump Shaft Seizure"
( 15.3.3).- Therefore,.the conclusions of the FSAR remain valid. Uncontrolled'RCCA Bank Withdrawal from a_ Subcritical- or Low Power Startup Condition (FSAR Section 15.4.1) For this - ANS Condition II - event, the analysis is performed at zero power h condi ti on s .-- The analysis presented in the FSAR assumed a flow of 171672 which is the two pump' equivalent. flow of 373200 gpm, which is more conservative than the reduced TDF of. 378000 gpm. Therefore, the conclusions of the FSAR remain
. valid.. ,
Uncontrolled-RCCA Bank Withdrawal at Power (FSAR Section 15.4.2) For, this ANS Condition II event, various powe. levels and reactivity insertion 1 rates ' for both . minimum and maximum reactivity feedback are analyzed. The transients are terminated by an Overtemperature AT or High. Neutron Flux reactor trip.. This analysis assumed a minimum measured flow of. 388880 gpm. If-a MMF of 385000 gpm were assumed, the system transient response would not
.be: significantly affected. Since the revised core -limits reflecting the reduced RCS flow are not exceeded at the most limiting point'in the transient,
- b< the safety analysis DNBR limit is met. Therefore, the conclusions of the FSAR remain valid.
RCCA Hisoperation (FSAR Section 15.4.3) For the events presented in this section of the FSAR, the effect of the reduced RCS flow has been evaluated by reanalysis. Therefore, the safety analyses DNBR limit is met and the conclusions of the FSAR' remain valid. Startup of an Inactive Reactor Coolant Pump (RCP) with Low Hot Lea lemperature (FSAR Section 15.4.4) For this ANS Condition II event, the analyses presented in the FSAR assumed' an N-1 loop flow consistent with a minimum measured flow of 386000 gpm. If an-N-1 loop flow consistent with a MMF of 385000 gpm were assumed, the system transient response would not be significantly a f f ec t.ed. It has been determined that the reduction in MMF to 385000 gpm would ha"e negligible impact on the DNBR at the limiting statepoint of the transient. Therefore, the conclusions of the FSAR remain valid. 1
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- CVCS Malfunction-that' Results in a Decrease in the Boron Concentration in 4 <; the Reactor Coolant (FSAR Section 15'.4.6).
- This ANS' Condition'Il eventfis> analyzed toLshow that adequate time exirts! for S operator' action : to terminatet a .dilutioni event prior to a ' loss of shutdown h margin.- For; the_ Mode- 1. 'and ; 2. analyses, "the assumption of RCS flow is not .
L explicitly modeled. - Thus,J atreduction in the RCS- flow- willL noti impact L the L
", calculation of time available to loss.of; shutdown margin. -Withl respect to the t
-DNBR criterion,. theat ' power cases, Modes 1 and 2, are bounded 'by.
P. " Uncontrolled RCCA Bank Withdrawal: at Power" (15.4.2). -Therefore, - all the gn safety criteria are. satisfied:and the conclusions of the FSAR remain valid. c V" Inadvertent Loading and_ Operation of a Fuel Assembly in an
+ Improper Position (FSAR.Section 15.4.7).
'The analysis of the Inadvertent Loading and Operation of a Fuel Assembly in an-Improper Position event does not axplicitly model the RCS flow. Therefore, it-is not impacted by-the RCS flow reduction. ,
Spectrum of RCCA Ejection Accidents (FSAR Section 15.4.8)
~ This ANS Condition IV - event is : analyzed to demonstrate that the peak clad.
temperature does not exceed 2700 F and that-the percentage of fuel me*; the hot spot in the core Jremains below 10L The analysis presented in i FSARL assumed a thermal. designa flow of 377000 gpm which is more conservative than--
-the reduced TDF 'of 378000 gpm. Therefore, all the safety criteria. are satisfied and the-conclusions of the FSAR remain valid.
A review of the McGuire FSAR Chapter 15.4.8 indicates that - conservative assumptions were used in the . analysis of the release of radioactivity to the environmentH in the eventL of- a postulated ~ rod ejection accident. These assumptions are listed in Section 15.4,8.3 of the McGuire FSAR. The reduction
, ' of the thermal design flow will not -influence. the assumptionsn made in the analysis, and therefore will not af fect the consequences of the rod ejection accident as reported;in the McGuire Final Safety Analysis Report.
Inadvertent-Operation of Emergency Core Cooling System (ESCS.). Durina Power Operation (FSAR Section 15.5.1) For this ANS Condition II event, the transient is initiated by a spurious safety-injection signal. .The. injection of borated water drives nuclear power and 'RCS . temperature down. Honce, the nost limiting thermal - hydraulic conditions occur at the initiation of the transient. The analysis presented
- in the FSAR assumed a minimum measured flow of 393600 gpm. If a MMF of 385000 gpm were assumed, the system transient response would not be significantly affected. Since the revised core limits reflecting the reduced RCS flow are
.not-exceeded at the most limiting point in the transient, the safety analysis DNBR limit is met. Therefore, the conclusions of the FSAR remain valid.
Accidental Depressurization of the Reactor Coolant System (FSAR Section 15.6.1) For this ANS Condition II event, the transient is terminated by an Overtemperature AT reactor trip. The analysis presented in the FiAR assumed a l
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' system : transient response would 'not be --significantly affected. Since the:
h % 4 revised core 1_imits _ reflecting the reduced RCS flow are -not exceeded at the
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' %", ' most 11.miting pointLin the transient, the safety analysis DNBR limit.1s met.
p. g Therefore, the conclusions of-the FSAR remain. valid. L i - Steam Generator Tube Rupture (FSAR Section 15.6.2): y e. l ' For. the Steam Generator Tube. Rupture (SGTR) event, 'the FSAR -SGTR ' analysis was 37
,. performed to evaluate the radiological consequences offan SGTR accident. The
@- a; - major factors that af fect the radiological doses ;of an : SGTR event are- the 3, d amount of ; primary coolant transferred to the = secondary side of E the failed "i steam. qenerator through the -ruptured tube and the steam-' released from the f( , n faded generator to the atmosphere. The'effect on.these major factors of a 1%- t t *
. reduction in thermal design flow has been determined.
D . - A sensitivity ' analysis was completed to assest %ct of the 1% thermal-
. design : flow reduction. .~The results of - the / analysis for; this
~ decrease in thermal design flow show a less the .e to.the primary to secondary breakflow.and a slight decrease to thw etmospheric steam release:via theifailed steam generator. The . increase in : breakflow i s much smaller tW the margin afforded by- the breakflow assumed in- the r ,diological analys, g presented in the? McGuire FSAR. A radiologi. cal evaluatian incorporating the "results ;of the flow - reduction was performed basad on the dose methodology' utilized in 'Section 15.6.2 of the McGuire FSAR. The results of the
. radiological consequence evaluation indiu ted that .he resultant offsite doses for the sensitivity'. case are bounded by the McGuire FSAR SGTR analysis offsite
' doses. .Therefre, the . McGuire FS AR SGTR analysis conclusion that the doses are within-tF .imits of 10CFR100 remains-valid for a 1% reduction in thermal design flow.
Loss' of Coolant ~ Accident (FSAR Section 15.6.4) 1 - 7The Westinghouse Small .-Break LOCA analysis consists of a thermal hydraulic Reactor ' Coolant System -(RCS) and a hot. rod analysis. For McGuire these J: analyses.' were performed with the Westinghouse small break ECCS evaluation model using the NOTRUMP code (Ref. 5). For the reduction in loop TDF from 95500 to 94500 gom the temperature in the cold. leg was calculated to be 0.3 degrees lower than the case analyzed at the original TOF. Thus'the density of the cold leg with the reduced TDF would be slightly greater than the case analyzed. Because the di1ferences in densities-of water in the cold leg between the two cases i s 'very small (less than-0.005%), there would be no significant difference in RCS depressurization rate or reactor trip time. l After reactor trip there would be no significant difference in the thermal I hydraulic response between the original analysis and the reduced TOF case. Addi tionally, for a small break LOCA, a minor perturbation in the initial
-operation ~ conditions should not have an impact on calculated peak clad temperature because of the event sequence. The peak clad temperature of a small break LOCA occurs af ter loop seal clearing. The coolant inventory and core mixture level af ter loop seal clearing are strongly dependent on steam t-
~
", w mo. m ~ ,o , .4 c. . ,,. ? . , ( , u c , generator pperation conditions andl loop seal clearing oscillstions during .'the transient.- -These factors are not affected by.the initial loop flow rate..
. In! addition 'to the local _ hydraulic conditions,, the peak rod cladding _ and_ fuel ,
. temperatures'are dependent upon<the rod internal parameters as initializedrat- 1
'steadyfstate fullv poweri conditions. Calculations have L showni that J these L
parameters are' not sensitive to changes -in4 the coolant temperatu.'e, pressure and flowrate. : Consequently,H the ef fect - of - the , small l change
- s ini the in-core q conditions accompanying.the flow reduction will,be negligible.
Based on the abbve' discussions, the red"4 tion in TDF. ' rom 95500,gpn to 94500 > gpm' willLhave no significant impact on tne small bream LOCA analysis 7results. Q~ TThus, the reductica in TDF will have no significant impacti on the :McGuire-S
- small break margin to the_ DCT limit of 2200 F.
The largelbreak LOCA; analysiscwas performed. with the 1981c version of the Westinghouse ECCS Evaluation Model using_the BASH code'(Ref. 4), and'asthermal~ design f;10w n rate i of 93500 : gpm. Sensitivity = studies have - shown- that small reductions-'in .thermala design ; flow result; in the calculation . of higher 'DCTs with ' , the evaluation model. -The .present -analysis therefore ' remains -
, V conservative for the proposed reduction in' thermal design flow, a '
The reduction inNThermalf Design- Flow._(TDF) has _no adverse ef fect on long . term , cc re : cooling :f allowing L am large ; break. LOCA: since this 1_s controlled by :the: safety > systems," the reactivity of . the - core, and ' the total ' mass of1 primary . coolant and J soluble boron- that are collected in the- cont ininent7 sump. since the temperatures'of.the primary loop have been changed by less than one degree due to the reduction in_TOF, the change in primary coolant massLis' negligible. N: er 2 3 l r
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' -'01 < J A Tdecreaset from13888801 gpm. to':3850001 gpm in:. ' the 'McGuire Nuclear S'tation .
gA M;n " : Technical:= Spect fication minimum measured: flow will not- adversely' af.fect the ; ' 4 3M Esteddy-state ori transient analyses' documented int Chapters 3, 4, 6"and' 15' of f , . W,,, ,
-the'McGuire Nuclear Statio'n final Safety Analysis, Report l,. ; !
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- 1. -McGuire Nuclear Station Final Safety Analysis Report;; Revised 12/8h= , q i ; i 2. WCAP-9220-P-A. .(Proprietary),. WCNP-9221: (Non-Proprietary),.
Y m -Eicheldinger, C., Westinghouse: ECCS EvaluationiModel -r 1981 Version,"- . Revi si on :- 1,-l 1981. ., p., , . 3 S[ ,
;3.-.
WCAP'-8339.1(Non-Proprietary),- Bordelon, F.M. , et' al . , J "Westinghousei ECCS - Evaluation Model - Summary," June 1974. -
.a 6 4. 'WCAP-10266-P-h'(Proprietarf.Besspiata,'J.'J1etal""The1981'Versionof .
3, the. Westinghouse ECCS. Evaluation.Model Using the BASH-Code," March'1987; . 4 : T
.'5. ' WCAP-1005' 4 -P-A (Proprietary)', Lee, -.N.: et al . , . " Westinghouse ? Small: Break t ECCS Evaluation Model Using the NOTRUMP Code," August 1985.
L{ t .q
, 7 6. DAP-85-0~/9, '"McGuire. Steamline ' Break Core Response Analysis -
with j'; Consequential. Failures'due to.Superheated Steam," March. 25, 1985'. i c
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.t. y ATTACHMENT III
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, The1 fol10 wing' analysis, _ performed pursuant ^ tot :10CFR50.91; ^ shows; thst; the-prop 6 sed 1 amendment will' not create a<significant- hazards- consideration, : as determined by.the criteria of 10CFR50.92, s
iThis amendmentL will _not significantlyc increase the
.1. probabilityLor
- consequences'of any accident previously evaluated;
-N, No: componentL modification, system realignment, or. change ;inioperat on; will occur ' which could_ affeet. the probability ofi .any Laccident . or 3D m ac transient. The reduction in flow will , not change _- the probability. of
* : actuation- of ; any' Engineered Safetya Feature - (ESF); or 'other . device. The..
consequences of f previously analyzed accidents- have been found to ' be; , 4 finsignificantly . dif ferent' if ' a 1% : lower flow . rate is assumed' in':the H k . analyses. ' The system transient - response. is ;not af fected by : the~ initial RCS flow assumption, unless tb. initial assumption is so 'iow.as tolimpair-
. the? steady-state core cooling capability or the: steam: generator: heat a vansfer capability. This is clearly-not the case wi,th at 1% reduction:in?
M,A-g ; JCS flow. 11 '2j This.' amendment, will not create the possibility _of: any new or' different--
,# -accident not previously evaluated. @ e No . component modification, system realignment,_ or change :in . operating ;
procedure will occur which could create the probability of aLnew- event. y, t 'not;previously considered. The reduction in flow will ' not_ initiate any - new ' events. :All credible accident scenarios have beer considered. 4
- a. 3. .This amendment will not - involve a - significant decrease in a margin; of.
,f . safety.
,N '
"As< described in Attachment II.,the-decrease in RCS ficw has been analyzed)
' ' and : foun'd' tot h ave an insignificant 1 ef fect 'on the applicable transient
@ analyses in the FSAR. The reduced flow rate resulted in slightly reduced
% 10NB limits. Figure 2.1-1 provides' revised core safety; limits for T-avg.
3 .as a function; of power at the xreduced flow rate. These limits will W .~ provide equivalent assurance that. operating parameters will . remain
' , ". . acceptable, n-F 'The A'xial Flux Difference limits given:in T.S. 3/4,2.1 are unchanged, 'and g' all ofJthe current the'rmal hydraulic design criteria are satisfied.at the reduced flow conditions. The current overtemperature AT and overpower AT setpoints are conservative and provide the necessary- protection. However,,
6 -the dead band of the f(AI) function (see Note 1 of Ta' ale 2.2al) was revised from -29% > q q > +9% to -29% > q q +7%. The effect of
- this' change is to askure protection in the evenh >of ' a power imbalance M< between the top and bottom of the core. No margins of sa4ety are reduced
% 'by these changes.
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