ML20196E919
| ML20196E919 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/06/1988 |
| From: | SYSTEM ENERGY RESOURCES, INC. |
| To: | |
| Shared Package | |
| ML19295G794 | List: |
| References | |
| NUDOCS 8812120131 | |
| Download: ML20196E919 (21) | |
Text
{{#Wiki_filter:,2.1 SAFETY LIMITS B.ASES
$1voced. Selt" THERMAL POWER. Low Pressure or Low Flow (Continued) Fuels Co-f *"t'*" ( W)
The ' -- """-- "- --MENG X4 3 I critical power correlation is appli- I cable to the mixed core beginning with cycle 2. The applicable range of the XN-3 correlation is for pressures above 585 psig and bundle mass flux greater than 0.25N1bs/hr-f t8 For low pressure and low flow conditions, a THERMAL POWER safety Itait of 25% of RATED THERMAL POWER for reactor pressure below 785 psig and below 10% RATED CORE FLOW vas justified for Grand Gulf cycle 1 ANF operation based on ATLAS test data. 0"r.ra11, bec ..rse of the design thermal-hydraulic compatibility of the'fMG 8x8 fuel design with the cycle 1 fuel, this ! justification and the associated low pressure and low flow limits remain appli-cable for future cycles of cores contPining these fuel designs. With regard to the low flow range, the core's bypass region will be flooded at any flow rate greater than 10% RATED CORE FLOW. With the bypass region flooded, the associated elevation head is sufficient to assure a bundle mass flux of greater than 0.25 M1bs/hr-f ta for all fuel assemblies which can approach critical heat flux. Therefore, the XN-3 critical power corrsla1.fon is appro-priate for flows greater than 10% RATED CORE FLOW. The low pressure range for cycle I was defined at 785 psig. Since the XN-3 correlation is applicable at any pressure greater than 585-psig, the cycle 1 low pressure boundary of 785 psig remains valid for the XN-3 correlation. GRAND GULF-UNIT 1 8 3-la Amen eent No. l 0121 ~ :1 N 6 i M t:ADOU P at 1 OiWh"fg;". k i
SAFETY LIMITS BASES 2.1.2 THERMAL POWER. Mich Pressure and High Flow The onset of transition boiling results in a decrease in heat transfer from the clad, elevated clad temperature, and the possibility of clad failure. However, the existence of critical power, or boiling transition, is not a di-rectly observable parameter in an operating reactor. Therefore, the margin to boiling transition is calculated free plant operating parameters such as core ' power, core flow, feedwater temperature, and core power distribution. The nar-gin for each fuel assembly is characterized by the critical power ratio (CPR), which is the ratio of the hundle power which would produce onset of transition boiling divided by the actral bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR). The Safety Limit MCPR assures sufficient conservatism such that in the event of a sustained steady state operation at the MCPR safety limit, at least 99.9% of the fuel rods in thei core would be expected to avoid boiling transi-tion. The margin between ca'culated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detailed statistical procedure which considers aNF the uncertainties in monitoring the core operating state. tonee specific uncer- I tainty included in the cafety limit is the uncertainty inherent in the XN-3 b critical power correlat< on.M EM report XN-NF-524(A), Rev.1. "Enon Nuclear l Critical Power Methodology for 8 oiling Water Reactors," Nov. 1983, describes the methodology used in determining the Safety Limit MCPR. The XN-3 critical power correlation is based on a significant body of , practical test data, providing a high degree of assurance that the critical i power as evaluated by the correlation is within a small percentage of the actual critical power being estimated. The assumed reactor conditions used in
' defining the safety limit introduce consorsatism into the limit because bound-ing high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further con-servatism is induced by the tendency of the XN-3 correlation to overpredict the nurter of rods in boiling transition. These conservatisms and the inherent accuracy of the XN-3 correlation provide assurance that during sustained opera-tion at the Safety Limit MCPR there would be essentially no transition boiling in the core, i I
l GRAND GULF-UNIT 1 822 Amendment No. _ i r i
3/4.a power O!STRIEJ)f!ON LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 During two loop operation all AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3. 2.1 b 3.2.1-la, 3.2.1-lb,-+a-3.2.1-Icf as multiplied by the smaller of either the flow-dependent MAPLNGR l factor (MAPFACf ) of Figure 3.2.1-2, or the power-dependent MAPLHGR f actor 3.1.1-14 ov (MAPFAC ) of Figure 3.2.1-3. P 3. 2. . t - t e During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits as determined below: [ l ast AN F f eet b>*G a) f o r41.;l tg; = the limit shown in Figure 3.2.1 1 I as multiplied by the smaller of either MAPFAC ,g MAPFAC p or 0.86; I and _ b) for ped ueltype[thelimitdeterminedin"e";;;.;f;r f.;' tp: l
+Ew. - o r 22.iigasm.mpu y m m y 3 .w menq, m,ne, l
o v O.46. l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THER% POWER.
, 3. 2..t-14 ACTION: oy 3.7..I t e Ouring two loop operation or single loop operation, with an APLHGR exceeding /
the limits of Figures 3. 2.1-1, 3. 2.1 la , 3. 2.1 lb , +< 3. 2.1-Ictas correc tec ey l the appropriate multiplication factor for each type of fuel, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER *ithin the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the required limits:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL PChER, and
- c. Initially and at least once per 12 hours when the reactor is ope ating with a LIMITING CONTROL ROC PATTERN for APLHGR.
The previsions of Specification 4.0.4 are not applicable. d. GRAND GULF UN;T 1 3/4 2-1 Amendment No. l l
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,{ 4VERAGE PLANAR EXPOSURE (mwd /ST)
Q FIGURE 3.2.1-1 MAXIMLM AVERAGEPLANAR LINEAR HEAT GENERATION RA z (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FoR GE F TYPE 8CR2iO 3 .. i
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? 45 50 Average Planar Exposure (GWd/ST)
FIGURE 3.2.1-1 MAPLHGR vs AVERAGE PLANAR EXPOSURE FOR SINGLE LOOP OPERATION. 8XS FUEL
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Average ~ Planar Exposure (GWd/MT)
.O FIGURE 3.2.1-1d MAPLHGR vs AVERAGE PLANAR EXPOSURE FOR ANF FUEL TYPE ANF361E8GZS8
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.o FIGURE 3.2.1 -1 e MAPLHGR vs AVERAGF, PLANAR EXPOSURE FOR ANF 9x9-5 LEAD TEST ASSEMBLY
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. CORE Flcw (% RATED) F FIG'JRE 3.2.1 2 MAPFACf i
Mendment No. 39 l GRAND GULF-UNIT 1 3/4 2-3a
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CORE FLDW (% MATED), F i FIGURE 3.2.3-1 MCPR f GRAND GULT UNIT 1 3/4 2-5 Amenment No.16 I
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RATED MCPR OPERATINC LIMIT = 1.18 i l I 1.00 2 0 20 40 SO 90 100 120 CORE FLOW (Z RATED) l FIGURE 3.2.3-1 MCPR 9
> i i e i i N DENA. PCMet 25% LP140%
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l.0 / 0 20 40 to 80 12 0 CORE THERMAL PO#Dt (% RATED) P FIGURE 3.2.3-2 NPR, GRAND GULF UNIT 1 3/4 2 6 Amenhnt No,16
cs 5
$ 2.40 E
Q* THERMAL POWER 251 5 P $ 401 2.20 -- - CORE I' LOW >501--y -t THERWAL P WEP *%I 3P $ 40I 2.00 -- _ _.. CORE FLOW 3 201 _ , _ . . . . . _ . I 1.80 7
+ + +-
i u A. 2 et CL n 7 $ 1.40 + - - - - - - - THERWAL POWER 40,I < P < 7 0 3 . --- m ALL CORE FLOWh l 1.40 - ~ ~ - - - -
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, I l ALL CdRE FLOWS 1.20 - - - ---
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P ! FIGURE 3.2.3-2 MCPR P
POWER O!5TRIBUT!0N LINITS 3/4.2.4 1,1NEAR NEAT GENERATION RATE . LINITING CONDITION FOR OPERATION 3.2.4 3 ;,:l:O ;;GT 0;;;;MT;;,;, ut; (L.;;;) f., 0; f.,1.. 11. ; ..s,.. The LINEAR HEAT GENERATION RATE (LHGR)f., ::: f.;; shall not exceed the limits shown in Figure 3.2.4-1. APPLICA8ti!Til 0PERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL PCWER. i ACTION: Wi th the '.';L" ef e r,, 0; fw i , ,,, .. ..J f 4 ;t, ;;. 'r4wNt l iai t e , .i t t, it, t LHGR of any 4NG fuel rod exceeding the limit of Figure 3.2.4-1, initiate I corrective action within 15 minuter snd restore the LHGR to within the limit within 2 hours or reduce THERMAL PvWER to less than 2M of RATED THIRMAL POWER within the next 4 hours. S'.'RVE!LLANCE REQUIREMENTS ( 4.2.4 LHGT.'t, f i: r, 0; f.;i O IM f.;l shall be deterstned to be equal l to or less than th'ir allowable limits: ! a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 1 M of RATED TNERMAL POWER,
- c. Initially and at least once per 12 hours when the reactor is operating on a LIMITING CONTROL 200 PATTERN for LHGR, and
, d. The provisions of specification 4.0.4 are not appilcable. 7 - i GRAND GULF-UNIT 1 3/4 2-7 Amendment No, l
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0 5 10 15 20 25 30 35 40 45 50 55 SO 45 % Average Planar Exposure (GWd/MT) 2 O FIGURE 3.2.4-1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE FOR ANF FUEL
~.--__.__: - + -- m-
_9
J/4.2 POWER DISTk!8UTION LIMITS , BASES The specifications of this section assure that the pe'ak cladding temper-ature following the postulated design basis loss of coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46. 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss of coolant accident will not exceed the limit specified in 10 CFR 50.46. The peak cladding temperature (PCT) following a postulated loss of coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondar-ily on the rod to rod power distribution within an assembly. The peak clad ter-perature is calculated assuming a LHGR for the highest powered rod which is - equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steacy ,3. 2. b l e. state gap conductance and rod to roc local peaking f actor. The Technical Speci- or fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR Cf 3.2.kle the highest pe=ered rod divided by its local peaking factor. The Maximum Ave" age Planar Linear Heat Generation Rate (MAPLHGR) limits of Figures 3.2.1 1, 3.2.1-la, 3.2.1 lb, 44 3.2.1-IcNare multipliec by the smaller of eitner the flo. I dependent MAPLHGR f actor (KAPFAC ) or the power dependent MAPLHGR f actor (KAPF AC ) f p corresponding 1: existing core fle= and power state to assure the adherence .c fuel mechanical design bases during the most limiting transient. & ::' -
' ; ; ', ; - ' P/F ' - ) f ; r ;ingi; 1;;; ;;; ;ti:r :: 0. M .
l l Fipve '3 . 2 .1 - 1, For single Icop operation wits ANF 8x8 fuel, a MAPLHGR limit'corresp:Sctng to the product of the ti;M;; :n-i;n;; "C '.:1 MAPLHGR,6and the appropria;e MAPFAC, can be conservatively used. pr:vid:d tMt t': it" :g: ?l r:P " 4 0 ' 2 -^- 4 ,. s u. - 4 . . a... .. tan m e m
.. . e.e.... , ".w.. ,
, , \
MAPFAC f 's are determined using the three dimensional BWR simulator coce to analyzeslowflowrunouttransients.ITr: :u :: fer -e f eel "--de- e e :- j vfd:t f;r .;; b;;;d er the e a;;ing ,;ttinii ;' tM ;;r; '1; 'iciter i- t' j *;;fr: d ;ti; "les Cen;r;l G,et;; in; :.r ; r;;r;;;n;;;;;; ;' ; :::i .' 00 0
#IO. 'i '; ^' Z.7, ;; ;;re T;;;if e;; e G.; ;; the I;ise r PeteT4i;I 'IO. ^.": !
t : :':-t- IInsev t 'n' MAPFAC p
's are generated .:in; th: :=: d;t; h;; ;; tM "CM pto prote:t l the core from plant transients other than core flow increases.
The daily requirement for calculating APLHCR when THERKAL POWER is greater then or equal to 25% of RATED THERHAL POWER is suf ficient since power distribw-tion shif ts are very slow when there have not been significant power or control l L GRAND GULF-Uh!T 1 B 3/4 2 1 Amendment NO.
1 POVER 01STR18t.rt!0N l.IMITS BASES 3/4.2.3 MINIMUMCRITICALPOWERRATJ The required operating Ilmit MCPRs at steedy state operating conditions as specified in Specification 3.2.3 cre derived frne the established fuel clad-ding integrity Safety Limit MCPR, and an analysis of abnorval operational tran-sients. For any abnorsal operating transtant analysis evaluation with the i Initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2. To assure that the fuel cladding integrity Safety 1.ielt is not exceeded during any anticipated abnormal operational transtant, the most liatting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (LPR). The type of transients evaluated were less of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the lar est delta CPR. When added to the Safety Limit MCPR, the required operating lett MCPR of (Reitmee 7) , Specification 3.2.3 is obtained. The power-flow sap of Figure 8 3/4 2.3-1 det mes the analytical basis for generation of the MCPR operating limits 4 /l
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The purpr.,se of the MCPR f and MCPR, is to define operating limits at other than rated core flo.t and power conditions. l 1he MCPR g s are estabitshed to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured. The ref-erence core flow increase event used to establish the MCPRg is a hypothes12ed slow flow runout to maximus, that does not result in a scram (rom neutron flux Iye,t'Cl overshoot esteeding the APRM neutron flux-high level (Table 2.2.1-1 item 2). p N - m' 1- x::t ?b; =h: ': d:;;.nd:.i :n th: ::hth; ;;tt'r; :' thee+e-
' _h . ' _b i t ; r. in ,, t r.; "_ ;; i r; .ht b r. T ,h.- C; .t r;' Cy; t ; .
T.; ' b. ret;; '.;;; 'c;- _ __ a _2 m ___,,_.__2 . ____ . g ;7 g;7_;_;; ;;7; y p_; M&+4. With this basis, the MCPR7 curves are generated from a series of steady state core thermal hydraulic calculations perforced at several core power and flow conditions along the steepest flow control line. In the actual calculations a conservative highly steep generic representatten of the 105% stss>i flow rod-line flow control line has been used. Asserptions used in the original calcula- - tions of this generic flow control line were consistent with a slow flow increase transient duration of several sinutes: (a) the plant heat balance was assumed GRAND GULF UNIT 1 8 3/4 2-4 Amend.sent No.
t POVER DISTRIBUTION LIMITS ' i 8ASES MINIMUM CRITICAL POWER RATIO (Continued) to be in equilibrius, and (b) core menon concentration was assumed to be constant I The generic flow control line is used to define several core power / flow states pnsert'Dl3 at which to perform steady-state core thersal-hydraulic evaluations, j i
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The MCPR, is established to protect the core from plant transients other than core flow increase including the localized rod withdrawal error event. Core power dependent retpoints are incorporated (incremental control rod with-drawat limits) in the Rod Withdrawal Limiter (RWL) System Specification (3.3.6). M M These setpoints allow greater control rod withdrawel at lower core powers where core thereal margins are large. However, the increased rod withdrawa) requires higher initial MCPR's to assure the MCPR safety limit Specification (2.1.2) is not violated. The analyses that establish the power dependent McPR requir'* p.wr.sts ! ments that support the RVL system are presented tic'" ::, .";;nct: 15 ", I (pg For core power below 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine stop valve closure and turbine control valve fast ,9dn closure are bypassed, separate sets of MCPR, limits are provided for high and low core flows to account for the significant sensitivity to initial core flows. For core power above 40% of RATED THERML POWER, bounding power- , dependent MCPR limits were developed. The abnormal operating transients : analyze,d,_f,or
. single toep .. operation are di,scussed in Re,f..er.ence 'h: n
- 5. u.. ---t
.u a . .-
.._.......a ...u..2... < .. ura. ..o.... _.i ...a No change to the MCPR operating limit is required for
':r _n dn'n; M 'r single loop operation. !
At THERMAL POVER levels less than or equal to 25% of RATTD THERMAL POWER, I the reactor will be operating at minimum recirculation pump speed and the modera-l tor void content will be very small. For all designated control rod patterns < l which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable sargin. t 1 1 i GRAND CULF-UNIT 1 8 3/4 2 6 Amendment No. l l l l
-_, . _ _ _ _ _ _ _ _ . , _ _ . . _ . _ . . _ _ _ , _ _ _ _ _ _ . _ _ . ~ . - _ -- - - - - - _ . . - - ,, - - -
DESIGN FEATURES l l 5.3 RE4CTOR CORE i FUEL ASSE4 LIES Esek M 5.3.1 The reactor core shall contain 800 fuel assemblies.Mth ::eofuel - assembTylcontain fuel rods aM 4we. water rods +4a4 with 2ircaloy -M. M* ggt *g ,. Each fuel rod sha! have a design nominal active fuel length of 150 inches. The initial core loading shall.have a design nominal enrichment of 1.708 weight percent U-235. Reload fuel sha1 W Of:it r t:; ps:t::1 d::f r t the initial l l con loading. I.e.ve weche.wlest, the rm t.- hyhtic Q m e @rowie c h.,v uteristics com p :W W t b C0WTROL R00 A55EH8 LIE 0 . h.3.2 The renter core shall contain 193 control rod assemblies, each l coasisting ei a cruciform array of stainless steel tubes contairing a casign i nominal 14's .7 f nches of baron cart >ide, 48 C, puder surrounded by a crue(fore shaped stainless steel sheath. 5.4 REtCTOR COOLANT SYSTEM l DQIGNPRESSUREANDTEMPERATURE 5.4.1 The reactor coolant system is desicaed and shall be saintsined: -
- a. In acccrdance with the code requirements specified in Section 5.2
- of the FSAR, with allowance fer r.orsal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of:
- 1. 1250 psig on the suction side of the ncirculation pump.
- 2. 1650 psig from the recirculation pump discharge to the outlet side of the discharge shutof f valve.
- 3. 1550 psig from the discharge shutoff valve to the jet pueps.
l
- c. For a temperature of 575'F.
4 V0LtME 5.4.2 The total water and stgam vohme of the nactor vessel and recirculation systes is approximately 22,000 ci&ic feet at a nominal T,,, of 533'F. , l l l 't GRMD GULF-UNIT 1 5-5 Amendment No. i
INSERTS FOR PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR CYCLE 4 RELOAD INSERT 'A' The allowable MAPLHGR shown in Figure 3.2.1-1 is a conservative bound during Cycle 4 for all 8x8 fuel types and the Cycle 3 SLO MAPLHGR (Reference 5). The MAPLMGR limit for ANF 9x9 5 fuel is the product of the MAPLtlGR shown in Figure 3.2.1-le and the appropriate MAPFAC. The maximum MAPFAC during single loop operation is 0.86 for all fuel types. i INSERT 'B' Two curves are provided based on the maximum credible flow runout transient for ANF fuel for either Loop Manual or Non Loop Manual operation. The result of a . tingle failure or single operator error during operation in Loop Manual is the runout of only one luop because both recirculation loops are unair independent control. Non loop Manual operational modes allow simultaneous runout of both loops because a single controller regulates core flow. - INSERT 'C' Two flow rates have been considered. The maximum credible flow during a runout transient depends on whether the plant is in loop Manual or Non j Loop Hanual operation. The result of a single failure or single operator erro during Loop Manual operation is the runout of one loop because the two ,ecirculation 1 cops are under independent control. . Runout of both loopt is possible during Non Loop Manual operation because a single 4 contvoller regulates core flow.
- InfiRT 'D' i '
l Loop Manual and Non Loop Manual modes of operation were analyzed.
' Consistent with the single failure / single operator error criterion, one ,
loop runout was postulated for Loop Manual operation whereas two loop runout was postulated for Non loop Manual operation. The maximum core
; flow at loop runout was assumed to be 110% of rated flow. Peaking ;
i factors were selected such that the MCPR for tht bundle with the least margin of safety would not decrease below 1.06. ( I l
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