ML19329A880
| ML19329A880 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/08/1977 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19329A876 | List: |
| References | |
| NUDOCS 8001150772 | |
| Download: ML19329A880 (51) | |
Text
{{#Wiki_filter:q@ __ hq rLEL PMD EDU KFFECT ON DA'!!S-EhS9C iTIT J Tr.Ci:N : CAL SP 41JICAT;0F', The Labcoch 6 Wilcox Cenpany has evaluatcd thc thern,1-: d7 raulie Ce a;a of t:m !v.vic-Besse Nuclear Power Station Unit No. 1 and has identified thernal margins which can be used to offset the DNBR reduction uhich results from fucl rod bcuing. The rod bow versus burnup curve used for this evaluation is consistent with that given in Reference (1) and in Section 3.2 of the " Interim Safety Evaluation Report on Ef fects of Fuel Rod Bowing on Thernal :fargin Calculations for Liaht Water Reacters". The red bou versus burnup curve has been used in conjunctica uith the Uestinghouse curve of DSER reductien, versus rod bcu to deternine the reduction in DMER as a function of burnsp. In daterrd n-ing a rod bcu penalty, the following thermal cargins have been identified to help of f. e t the DNER reduction:
- 1. The flou area (pitch) reduction factor, which is used to acccunt for the reducticn in DNSR due to pitch reduction fren fabrication tolerances and initial rod bcw.
This factor has been determined to be cquivalent to a reduction of 1.3% in EM3R.
- 2. Renoval of the densificaticn pcwer spike factor from DMBR analyses. This factor has been determined to be equivalent to a reducticn of 1.1" in DN3R.
Ecced on this evaluation and t .e therral cargins prescated in (1) and (2) hove, a reduction in DNBR is required for Davis-Besse Unit Cne. The required reduction for Cycle 1 of Davis-Besse 1 has been accennodated by adjusting the Pressure-Tenperature Limits and the corresponding variable low pressure trip set point as given in Bases Figure 2.1 and Figure 2.1-1, respectively, of the Davis-Besse Unit 1 Technical Specifica-tions, and in the axial power inbalance and overpeteer limits as given in Fi ures 3 2.1_. 2.2-1 and 2.2-2 of these Technical Specificaticns. The axial peuer-imbalance envelepes , Figures 3.2-1, 3.2-2 and 3.2-3 have also been revised to cetpensate for predicted red bou pouer cpikes. Changes have also been nade in the RPS trip setpoints and allevable values listed in Table 2.2-1 of the Technical Specifications to account for instru=ent drif t in accordance with the meeting betueen NRC and TECO ef December 10, 1976.
References:
(1) Letter, JF Stols (NRC) to LE Roe (TECO), "Ef fects of Fuel Rod Bowing on Departure from Nucleate Boiling", dated 1/21/77 I
--- . - . 8001150772 I
PROG, &-REVIEW DOPY
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3.2.1 AXIAL PC'JER IM3AL# ICE shall be maintained within the limits sh5;n
- 3. 2. - 3, on Figures 3.2-1 arwi 3.2-2) ct.nd
. APPLIC;3! LIT /: .MOCE 1 above 40% of PATED THEFSAL PC'WER.*
ACTIC.l: With AXIAL FC'JER IMEALRICE exceeding the limits specified above, either:
- a. Restcre the AXI'L PC'JER I:3.2LC'E c within its limits within 15 minutes, er .
- b. Be in at least HOT STAN05Y within 2 hcurs.
3 V SURVEILLA' ICE 'ECUITEMENTS
. 4.2.1 The AXIAL FC'WER IMSALifiCE shall be :ietam,insd ta .e within limits at least once every 12 hcurs wnen abcle 'C" of FATED T9ERFAL FC'42R except wnen the AXI.AL PC'AER I"3ALAiCE alarm is inc::erable, t.ien calculate the AXIAL FC'WER IMIALANCE at least cnce per hour.
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= See Special Test Exceptien 3.10.1 a Ww1 $
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g p . . p . c p) 6dfl! 3/4.2 P0'.ER Of;. .! BUTTON LIMITS - -
- t. cr;
.:=-
- - - - - ~
0 Tb :rc! fin a_ct of U n 12 ,Mn ,cavide c : :.ur. = c f f. e
*: e. c j during Conditica i ( cc.nl 0;sa;icnj and I: (ir.cic;. s ci .". 4 .a frequency) events by: (a) maintaining the minimum CNER in the core
> 1,32 during normal operation and during short tena transients, (b) maintaining the peak linear power density < 18.4 kw/f t during r.ormal operation, and (c) naintaining the peak ccUer density < 20.4 kw/f t during short-tern transients. In addition, the above criteria mu:t be met in order to meet the assumptions used for tne loss-of-coolant accidents.
The pcuer-imbalance envelope defined in Figures 3.2-1 and 3.2-2, and the insertion limit curves, Ficures 3.1-1 and 3.1-3 are based on LOCA analyses wnich have defined the maxicum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200 F folicwing a LOCA. C;sraticn.cutside of tSe 1 power-inbalance envelc;e alene cces no: constitute a sitection that would . cause the Final Acceptance Criteria to be exceeded should a LOCA occur. 1 The power-imbalance envelcoe represents the boundary of cceratien limited by the Final Acceptanca Criteria only if the control rods are at the inser-tion limits, as defined by Figuras 3.1-1 and 3.1-3 and if a 4.92 percent QUADRANT PO'lER TILT exists. Addi icnal ccnserva; ism is intrccuctad (') by application of: s
- a. Nuclear uncertainty factors.
- b. Thermal calibration uncertainty.
. c. Fuel densification effects.
- d. Hot rod manufacturing tolerance factors.
G. ?, t ed st Fue I ro d b,ua effee . s l The ACTION statements which permit limited variaticns frem the basic requirements are acccmpanied by additional restrictions which ensures that the original criteria are met. The definitions of the design limit nuclear pcwer peaking factors as used in these specifications are as follcws: F Nuclear Heat Flux Hot Channel Factor, is defined i the maximum t 0 local fuel rod linear ccwer density divided by the averace fuel rod linear pcwer density, assuming ncminal fuel pellet and red dimensicns.
! oom . D DJ ' ' J a o JL '
- m T
' ~
- 0 '
. '0 3 DAVIS-BESSE, UNIT 1 B 3/4 2-1 ._ _(f)_,./L. _a i ..
JAll 121977 TO 4 -
-~ ~ -- -
,; 3y 7 5 'cu),v us , i :.' "B
'PCPER t
CIST.T EUTIIW LI!ITS
/
u . 1
' ; ,a. 5..r- ..
jr l I" riuclu.c Enthai, j "H a S ; C: ' ~ :1 S-;;: , is defi: .' s ' h-t.!! ratic cf the in:cgrEl ci lin;ar ;c.or aicrg the rc:. ca taich minimum Di3R occurs to the average red pcwer. It has bcen determined by exter.sive analysis of possible ocerating powar shapes that the design limits on nuclear pcwcr peaking and cn minitun C::ER at full pcwer are met, prcvid:d: N
- FQ < 2.94; F4H < l.71 Power Peaking is not a directly cbservable cuantity and therefore limits have been established en the bases of the AXIAL PC'iER IMSALA;CE prcduced by the pcwer peaking. I: has been determined that tn2 abcVe hct
- channei fac cr limits will be tot previded th3 fcilcwing conditions are maintained. ,
- 1. Control rods in a single group mcve together with no individus.i rod insertion differinc. b.v mera than .+ 5.55 (indicated position) frca the grcup average height. .
- 2. Regulating rod grcups are secuenced wifh cverlapping groups as O. required in Specifica:icn 3.1.3.5.
- 3. The regulating red insertion limits cf Specificction 3.1.3.6 are maintainet. .
- 4. AXIAL POUER IMEALA;tCE limits are taintained. The AXIAL PC'lER IMBALA!;CE is a measure of the diffsrenca in power between the top and bottcm halves of the ccre. Calculations of core avcrage axial peaking facters for tany piants and ceasurements frca operating plants uncer a variety of oceratinc ccnditions have been correlated with AXIAL POWER IMSAL' ;CE. 'The correlatica shows that the design ocwer shape is not exceeded if the AX*AL POWER IMBALAUCE is maintained between -W percent and .Sh4 l
' percent at iATED THEFJ'AL POWER. + Wo - i e.o l 1 4 The design limi power peaking factors are the most restrictive calculated at full power for the range frca all centrol reds fully withdrawn , i to minimum allowable control red inser:icn and are the ccre D.';3R cesign l 1 basis. Therefore, fer operation at a fraction of RATED THEPF.1.L POWER, tha i design limits are met. When usl.ng incere detectors to make pcuer cis:ribu- l tien caps to determine F Qand F"aH:
- a. Thecensurementoftotalpeakingfacter,F$eas , shall be increased by 1.4 percent to acccunt fcr manufacturing tolcrances and fur:acr increascd by 7.5 percent tc acccun Q
for measurement errer. DAVIS-BESSE, Uti1T 1 d Mg2
. D OdA JAri 12 E/7 n' ..
D Td _T "' c w - T s .J . _ L ai
-- .s,-- . = -2
- n ..
.13s.,c l.
i .
" 0
' ~ p
,i.
". G.'.'rR DITTII,IIT10' LIMITS j .
.i . PASES 1 .i i ti ,,
- b. The censurement of enthcipy rise Fct channel factor, F7q, shali be increased by 5 percent to ac.ccunt for measurement er'cr. r For Condition II events, the core is protected from exceeding 20.4 kw/ft locally, and frem goinc belou a minimua C?iER of 1.32, by
- automatic protection on power, AXIAL PC'.-lER DISALA:;CE, pressure and i temperature. Only conditions 1 through 3, above, are candatory since the AX1AL PC'.-lER D*.3ALANCE is an explicit input to the Reactor Protection System.
j The QUADRAli PC'.!ER TILT limit assures that the radial ;ctier distribu-tion satisfies the desica values used in the ocwer cacabili:V analysis. Radial pcuer distributi5n measure:.an:s are mac'e during star;hp teiing i and pericdically during pcwer operation. The QUADRANT PO',!ER TILT limit at which corrective action is red'uired provides D:13 and linear heat generation rate protecticn with x-y piane power tilts. In the event tne til; is not corrected, the margin for uncertainty en Fg is reinstated by reducing the power by 2 'cercent for
- each percent of ilt in excess of the limit.
~
- 0 3/4.2.5 C:'S FARA:'ETERS .
The limits cn the D:13 related parameters assure that each of the parameters are maintained within the normal steady state enveloce of-operation assumed in the transient and acciden analyses. The limits are consistent with the F5AR initial assumations and have been analy:ically demonstrated adequate to maintain a minimum CNBR of 1.30 throughout each analyzed transient. The 12 hour periedic surveillance of these parameters through instru-ment readout is sufficient to ensure that the parameters are restored within their limits following load changes and other excected transient 4 operation. The 18 month pericdic measurement of the RCS total ficu rate using delta P instrumentation is adequa:e to detect ficw degradatica and ensure correlatien of the ficw indication channels with measured ficw such that the indicated percent ficw will provide sufficient verification j of flow rate on a 12 hour basis. O o-D D wol ~ V " ]~T L~ - g# O . : 4 DAVIS-BESSE, UNIT 1 B 3/4 2-3 , .. 4 . i JAN 12 i377 i
m< v PROOE &-REVIEW COFY,- I, ,, 4 . I '
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8 1 SECTIC:' 2.0 j t 3 SAFETY LIMITS N4D LIMITIriG SAFETY SYSTEM SETTI'GS f 1, - t s g . 2 . .. I 1'
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P.EACTOP. CCRE 2.1.1 The cc=bination of the reactor coolant core outlet pressure and cutlet temperature shall not excaed the safety limit shown in Figure 2.1-1. APPLICASILIT/: M0CES 1 and 2. L ACTICd: Whenever the point defined by the c mbination of reactor cecian core cutlet ressure and cutle: :Em erat:.re has exceeded tne safety limit, 3
- i be in HOT STA:iC5Y wi:nin One hcur. .
REACTOR CCRE 2.1.2 The ccmbinatica of reactor THER"AL PCWER and AXIAL PCWER I:iEALA::CE shall r.ct exceed :ne safety limit shc.4n in Figure 2.1-2 fer the various ccabinations of :.vo, three and fcur reactor coolant pump cperation.
'{ .- APPLICA3IL!TY: M00E 1. '
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J ACTIOt: 1 5 Whenever' the ;cint defined by the c :bination of Reactor Ccciant Systam i ficw, AXIAL PCWER IMSALA :CE ana THE??'AL PC'.iER has exceeded the appr:priata v safety limit, be in HOT STA GY witnin one hcur. REACTCR CCClatti SYSTE'4 PRESSUP.E 1
'j 2.1.3 The Reactor Ccolant System pressure shall not exceed 2750 psig. -
9 a APPLICA3ILIT/: MOCES 1, 2, 3, 4 and S. ACTIC't : . S MODES 1 and 2 - Whenever the Reactor Ccolant System pressure has ex-
',3 ceeded 2750 psig, be in H0T STA:iCSY with the React:r Coolant System pressure within its limit within ene
^} hour.
M0CES 3, 4 - Whenever the Reac:Or Ccolant System pressure has and S exceeded 2750 psig, reduce the P.eac:Or Ccolant Sys:2.j pressure to within its limit wi:nin 5 minutes. 3 oo D - 1 D . 1 4'
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' . . . ,r-.,.. . . ._
declare the channel incoerible and a20'? th* a??,*lCa3!". 'sres.~[T!',"' requiremen: cf Specification 3.3.1.i ur.til the en.annet ? ' ' ' ' 2 . - .- OPER/,BLE s:stus with its trip se:poin adjusted consistent wi:n the Trip Se:poin value. i 1 H 4 i . 1 . a
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,b 1. t4anual Reactor Trip liot Applicable riot Applicabic E 2. liigh Flux < 105# ?. of RATED TilERl1AL POWER < 105.54'of RATED THEF. MAL POWER
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i l PROCT.&RB/IEW COPY s . . 3
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V The su nary statt.7.cnt: contained in this sectic1 provide the bases for che
,: specifications of Sectica 2.0 and are not considered ar art vf a
'..'.cee ..r.-.s....2't specification as proviced in 10 C. R 50.35.
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The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation whicn would result in the - . Overheating cr 1cn products to t.,e reactor cco t ant. release ct- 11 u ne fuel cladding is prevented by restricting fuel cperaticn to within the nucicate boilin"a regir.c unere the heat transfer coef ficient is large and
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o add 4..-3 sc "m r3. m' m- a.' n...g . a r. - " r n. d,, e s l i c .Y.'./' c'.- ' v v c. C. w . a. ce. ol a r. . . .e .= w'.r. = 'u i co-r, . ~ [c.e.6p . $ .2 t..wr e.n. . Opera t. ion >". ova. . + L... n. .rc c r ka"...d m .a r..".-' ' . '. .b a. r. "m r ..'. e .=. h.a .'. s.i l '. . .c":i...' . wo.. u l d - o. ..c u l ' '. .". =. .v..- a. c .. c '. ". a. c.i. '.'in . r.e..~*..~.."..".=-s *..~..=.v.==. . e. '. ' . . " . =
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ci i . '.1 t.n. +..h.o. ] e. r. ., ] .L. o. .= *. 3. 1. o. .v. , {3 j e.d. 4. C .' *. i. ". A. A. .# .. *".O. ". ..= " m i r. '.,. n.13 . s., . .
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'l 3 o. n. I va. . .a .s.r .s c,ea u,n c- nry- r. ny.e,+e.
o r ,.Il 't l n o
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C o n f:. C a n v-n. . . 4. a- C .h~s r u, n margin to C"3 for all cperatir.g conditions. The curve presented in Figure 2.1-1 represents the ccnditions at which a minimum DNBR of 1.32 is predicted for the maximun pessible thermai pc.ier 1125 when the reacter ccolar.: ficw is 121.3 x 10i its/hr, wnich is the design flew rate for four cperating reacter coolant pumps. This curve is based on the folicwing hot channel facters with potential fuel densificationa effects: p.,paar a.o e win c, F = 2.55; N F'A H
= 1.71; F'1Z = 1. 50 Q
The design limit ecwer peaking factors are the most restrictive calculated at full pcwer for the range frca all control reds fully withdrawn to minimum allcwable control red withdrawal, and form the core C"SR design basis. CN N b D es es S .,
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hr ~ :: .' _ _ ti - - - lit i T. 2 : x E r '.c r e t- 1 :. 9 mum . t.pn:: ' N- e: it ty 1 i.Y n:- closely tr.sa it actually d:es bectu;e the reactur : rip pres;ures era measured at a iccatica wnere the indicated pressure is abou: 30 psi less than core outlet pressure, providing a more conservative margin to the safety limit.
.The curves of Figure 2.1-2 are based on the core restrictive of two thercal limits and i"' t :-he effects of pctential fuel densification; e.o
.4ceum w p rem an s.t. e,o :
- 1. The 1.32 D:;3R limit prcduced by a nuclear pcwer peaking factor of F = 2.55 or the ccabination of the radial peak, q
axial peak and pcsi ica of the axial peak that yields no less thcn c 1.22 0::ER.
- 2. The cctbinatica of radiai and axial ceak tha: causes central fuel melting at tne hot spot. The limit is 20.4 kw/ft.
Power peaking is net a directly observable'cuantity and therefore limits have been estabiished on the basis of the reactor pcwer imbalance A produced by the pcwer peaking. v The specified ficu rates for curves 1, 2, and 3 cf Figure 2.1-2 correspond to the expe:ted mininua ficw rates with four pumps, nrce pumpt, and one pucp in each lo:p, respectively. The curve of Figure 2.1-1 is the cost restrictive of all possible reactor ccolant purp-caximca thernal cwer cctbina:icas shown in BASES Figure 2.1. The curves of SASES Figure 2.i represen the conditiens at which a minimun 0:iSR of 1.32 is precicted at the taxicum possible thermal pcwer for the number of reactor coolant pumps in cpera icn or the local quality at the poin: of minicua 0:;3R is equal to +225, whichever conditica is more restrictive. wer e v 4.ves u- o w m- +re-r,m_ emrs on e,n .e., a-s m am pen;ci~ ria. Using a local quality limit of +225 at the point of miniaun C::BR as a basis for curve 3 of SASES Figure 2.1 is a conservative criterion even though the cuality at the exit is higher than the quality at the point of minimum D:;SR. The D:'BR as calculated by the EL'.i-2 D iS correlatica centinually increases frca poin: of cininum D::SR, so that the exit D:;3R is alt.ays higher. Extrapolatien of the correlati0n beyond its published range of +22% is justified en the basis of expericental da p
- quality D , D OaS ;
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J r ! For each curve of BASES Figure 2.1, a pressure-te:cerature point 3 I above and to the left of the curve aculd resuit in a C33?. greater than 1.32 or a Iccal cuality at the coin of minicum D EP. iess than -225 . . ..
. for that particular reac:or coolant pump si:uaticn. ir.e 1. :., o....-nen l .
curve for four pump cperation is =0re restric;ive than any c:her reactor coolant pump situation because any pres:ure/tecterature point abcVe and to the lef t of tne cur pump curve w111 be above and :: ne lef:
- of the other curve:.
4 <>.
- a. 1.3. 3.. : . ,, . n .o. e. r m.. . r.- e. v. e , _ ". p e.r. .- . ". r.:
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r#. "...a. Reactor Coolant Systra frca everaressuri:2:icn and thereby Orsvents One release of radicr.uclides centair.ed in tne reac::r cociant frc= reaching I the containment atmasphere. { J **
'h i e . c. .a. c *. .. yn r a. c. .,"w
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a -. '" i c. o 3 .e d '. . .. ., , N. .n ". s.". .- .c. *. . . '. . s .= m. . . . #. ".. . t. .i ~. . w. . . s '. =. .r. o
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therefore c:nsistent wi a the design critaria and asscciatec c:ce requirements. The entire Reactor Coolant Systam is hydrotested at 3125 psig,1255 , of design pressure, to descnstrate integrity prict t: initiai :peraticn. ' L. .
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i < t . . , . ,. et:.:- L..._._...____ _ _.__.._____ . _ . . _. ._ l 2.2.1 REACTOR PROTECTIC ! SYSTEM INSTRU:E:ITATIC1 SETFOINTS The Reactor Protec:icn Systcm Instrumentaticr. Trio Set:cint s;ecified in Tcble 2.2-1 are the values at wnich the Reac:ce Tri;; ar2 set f:r saca I parameter. The Trip Se: points have been selected to cnsure that the reactor core and reactor cc lant system cre prevented frca exceedi ; l their safety limits. Oper::ica with .a trip set:cint less conserva-ive I than its Trip Setpoint but within itt, specified Ailcwable Value is a:cect-able on the basis that each Alicuable 'lelue is equal to or le:s inan he j drift allcwance assumed for each : rip in the safety analyses. a 1-
+i The Shutdcwn Bjpass pr0vides for byp2: sing certain functicn3 Of the
- - Reactor Protection Systt'n in order to permit con rol r:d drive tcs::,
aero pcuer FRYSICS TESTS and cer ain startup and shutd:wn precadure:. The purpose of the Shutdown Eypass High Pressure trip is to revan: normal cotration wi .h Shu:dcwn Sycass activa ed. This high pressure rip setpoint is Icwer t.ian the ncmai Icw pressure rio se- cin so na-
- m. the reacter ~ust be trisced before the bvo. ass is initia;;d. The Mic.h .
3 Flux Trip Se::oint of < 5.G5 c.c ven s any significan reac :r pcwer
.j -
< frca being pr:cuced. -Sufficien: naturai circulation wculd be available .
.n .s ir none cr the reactor c: clan-i r.:.ra..,l - 0....,
.c o re:Ove . .,. Or vu: .__D pumps were cperadng. .
Manual React:r Trio The Manual Reactor Trip is a redur. dant channel to the aut: atic Reactor Protecticn System ins rumentatica channels and provides car.ual reactor trip capability. P Hich Flux A High Flux tric at high power level (neutron flux) provides reactor core -rctection agains reactivity excursions which are tco rapid to be protected by temperature and pressure prctective circuitry. During normal station operation, reactor trio is initiated when the d reactor pcwer level reacnes 105.55 ci rated power. Cue to calibra:ica - and instrument errors, the maximum actual power at whicn a trip wculd be actuated could be 1125, which was used in the safety analysis. ~ s
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l RC Hich Tcecerature .
, The RC High Tc crature trip 1 619'? prcvent: the reactor cutlet j temperature frca exceeding the design limits and acts as a tackup tr:p for all pc.-;cr excur:icn tran:ients.
Flur - a Flux-Flc s
- l The power level trip setpoint : reduced by the reactor ccolant i Lystcm flew is bascd on a #iux-to-f'. w r::ic ..hich has cecn as ablisned
}
to ace = date ficw dccreasing transien: frc: nigh p wer where pro-1 taction is not provided by the Hign Fiux/h::cr of Reactor C: clan: ? umps On Trips. The pc,.er level tric set;oint creduced by' the pcwer-t:-ficw ratio
; provides both high power level and icw fica rotecticn in the even: the s reactor pcwer icvel increases cr the reac::r coolan: ficw rate cecreises.
1 i p. ,,,g ..-ic. , m../. ;.. e ..--g--
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.,. <... s. .,......
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.,,v...e I overpcuer D;E prc:cc:icn for all :: des of ;uma c; era:ica. Fcr every d ficw rate thcre is a maximun :ermissibia ;0e.er levei, ar.d for every
) pc;,cr level :here is a minimum pernissi le icw ficw rate. Typical ;;wer
, level and icw ficw rate c::binations it. :he pump situaticns of Table 2.2-1 are as follows: -
e i 1. Trip would cccur when four reacter coolant pumps are operating if power is 109.00% and reac:ce fi:w rate is 1005, or ficw O rate is 92 @and pcwer level is 100"., j 2. Trip wculd cecur when three reac:cr c:olant pu=cs are c: era:ing if power is 60.105 and react:r flew rate is 74.7%, or ficw a rateis69.5%andpoweris75~.
?- 9 3. Trip would oc:ur when one reactor coolant pumo is operating in
-i -
each leap (total of wo pumps cperating) if the pcwer is 53.00L s .i and reacter ficw rate is 49.0% or ficw rate is 45.F' and the ,
.'l power level is 49.0%.
. For safety calculatiens the maximum calibration and instrumentation
~
o o errors fer the pcwer level were used. i m o ,. s D 0 , ts . o a< . i '- n-
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u.:.:.L.::w . BASES 1 1 i The AXIAL POER D'EALAt:CE beundaries are established in order to prt. vent reactor thermal limits frca being exce:ded. The:e tner .21 l limits are either pcwcr peaking kw/f t limi s or C:iBR limits.
.I-f .
1
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P.C Pressure f.:u. Mi-h and Frn- .ure Te tearature
. The High and !.:w trips are provided to limit the pressure range in
].
which reactor Opera:icn is permitted. Curing a slex reactivity inserti:n s::r:co a:cid:nt frcn icw ;;wer or a slew reactivity in:ertien fr:m high ;cwer. the F.C High Pres:ura
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-l j point f r ?.C Hign Pressure, 2255 psig, has been established to cair:ain
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transient. The :C. Higa Pressure trip is t:ckec u; by the cres:uri:cr f' cede safety v:ives f r RCS cver pressure ar ery.:n, and is tneref:re i set icwer taaa tne set ::ressure fcr these vaP es, 2 35 psig. The RC High Pressure trip also backs up the High Flux trip. 1 The RC 1.cw Pressure,1985 psig, and RC Prcssure-Tem::eriture (13.01 Tout *F-5773; osig, Trip Setpoints have been established to maintain the C :S ratic grea:er than er equal to 1.32 for these cesign accicen:s :nat
< result in a Orcssure reducticn. It also prevents reac:Or coeraticn a:
pressures belcw the valid range of C 43 correlatien limits, protecting ,
; against C:13 .
.a d -
Hich Flux / ! umber of Reacter Ccclant Punos On
'j In conjunction with the Flux - a Flux-Ficw trip the High Flux /: umber
- 1 of Reacter C:olant Pumcs On tric prevents the minimum core Cli3R fr:m decreasing beicw 1.32 by tri;cing -he react:r due to the less of reac: Or
' ' { , coolantpump(s). The pumo meni :rs also restric: the pcwer level for the number of pumps in cperation.
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- DAVIS-BESSE,tJilIT I B 2-6 S t.. u s .IS 1
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j 1 Containment Hich Prassure + The Contair. ment High Pressure Trip 5etpoint < 4 psig, provides ~~ l positive assurance that i reactor trip will cccur in :ne unlikely
; cycnt of a :te:a line failure in the contair. ment vessel cr a ioss-of-1 ccolant accidant., even in the absence of a RC Lcw Pressure trip. -
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l DAVIS-BESSE-1 LOCA AilALYSIS - DESIGN PRESSURE DRCP EVALUATION
- 1. Introduction ,
The Davis-Besse-1 (DB-1) LOCA Analysis, reported in B&W topical report BAW-10105, was per formed in the winter of 1974-1975, and was based on the system pressure distribution available at that time. In comparing that evaluation to one for a nearly identical plant which utilized pressure distributions developed in 1976, an erroneous input value affecting the reverse ficw pressure drop for the ! inlet nozzle was discovered. Correction of the DB-1 model only in the area of the inlet nozzle (upgrading the nozzle model to the 1976 values) would have made it necessary to reduce the peak linear heat rate. Such a reduction is, however, inapprcpriate, in light of the availability of improved design pressure distribution models based en actual operating plant experience. Data from the i 177 FA operating plants, unavailable when the 08-1 analysis on file with the NRC was performed, has been used to alter and verify system pressure distributions. These improved distributions are the basis upon which. 'to construct the LOCA j evaluation. This paper presents the results of our revised arid improved l evaluation of the loss of coolant accident for the DB-1 plant. l The subject of conservatisms and the appropriateness of the margins in present and past calculations has been carefully considcred in this analysis. System l l pressure distribution has heretofore been based on a best estimate evaluation i using the data available. Nearly endless parameter variations would be necessary i i if conservatisms were to be placed within the pressure drop model itself. Therefore the industry and the NRC has chosen to use a best estimate approach for these inputs. B&W concurrs with this approach and has used it consistently l
i in past analyses. In order to ensure that the transient flow utilized in determining the cladding temperatures is conservative relative to the expected case, Appendix K requires that a higher than realistic discharge model (break flow model) be used with a spectrum of break sizes. All analyses performed to demonstrate compliance with the criteria of 10CFR50.46 are then based on the worst spectrum case. In this way, flow conservatism is positively assured. For B&W plants, the worst case result has always been the largest cold leg pipe break considered. Conservatism of this approach is demonstrated by comparing the worst case spectrum result (CD = 1.0) to the expected case (CD = 0.6). I i !
- For the DB-1 plant type, a double-ended cold leg break with a discharge j coefficient of 1.0 yields a peak cladding temperature 3500 F higher than the 1
l expected case, the same break with a discharge coefficient of 0.6. This l
\
difference reflects the conservatism on only transient ficw and ignores other ) known conservatisms in Appendix K. As the revised pressure drop evaluaticn does not alter the spectrum approach, the results contained here remain i suitably conservative for demonstrating compliance to 10CFR50.46. Using the revised system pressure distribution has resulted in a slight drop in the peak cladding temperature of about 300 F from the original DB-1 evaluation. The analysis also includes certain model adjustments, detailed l i later, which have been imposed by the NRC since the 08-1 original submittal ) and were therefore not considered in the original work. I - l l-2 1 i
, ,--,--m e< -- e-,+, -m +c - - . - - - - - -=m-e - n - - -- - ,- , - v ---
- 2. Sumr.ary and Conclusions 1
The peak cladding temperatures obtained for the re-evaluation of the 08-1 6-foot LOCA Limit Analysis, based on the corrected inlet nozzle model and the revised system pressure distribution, are 2133 F for the unruptured node and 1999*F for the ruptured node. These results are 33 FU and 21 F lower, respectively, than those obtained in BAW-10105 due to the enhanced reflooding of the core. Comparisons of other dependent variable evolutions, which are presented in Section 4, show that the overall transient is very similar. Thus, the trends of the spectrum, sensitivity, and LOCA limits studies reported
<n BAW-10105 -emain valid. Since the results are less severe than t'1ose reported in BAW-10105, the resulting LOCA limits curve presented in that docunent is conservative and remains a viable base for assuring compliance to 10CFR50.46.
,J 4
. - . . - , . - s.
- 3. Basis for System Pressure Distribution Alteration The determination of system pressure drop and fica distribution is cade through use of the SAVER computer code, described in B&W topical report BAW-10072. Although detailed information can be obtained from the topical, SAVER can reasonably be described as an automated table of standard industry pressure drop models as used in the design of process flow equipment. In developing the actual distributions for use in design, a normalization to existing experimental data obtained from operating plants or from vessel model flow tests is performed. Thus, best estimate values for pressure distribution are created.
For the original DB-1 LOCA submittal, as reported in B&W tcpical report BAW-10105, operational data was not yet available and the resulting pressure distributions were therefore based exclusively on calculations. The basis for this submittal is an altered pressure distribution model which contains the influence of operating and vessel model ficw data. Figure 3-1 shows the lccations where pressure drop measurements were made during Oconee 1 hot functional testing. Using this data the OTSG and reactor vessel unrecoverable pressure drops were obtained by removing momentum, elevation, and calculated reactor coolant piping friction effects. The pressure distributions within the reactor vessel and the OTSG were then adjusted to match this data. Figure 3-2 is a gross comparison of the original distribution, BAW-10105 bases, and the present best estimate. Note that the reactor vessel pressure drop when based en hot functional test data is slightly lower than the original and that the vessel model flow test results substantiate the decrease. The difference in the OTSG pressure drop is addressed f
later in Figure 3-4. For the piping there is essentially no change. Figure 3-3 compares the reactor vessel incremental pressure drops. The original calculation, based on an older design, included a long leg U-baffle wi h a prr.ssure drop of 3.45 psi. The new calculation is based on the existing desiga which does not have a long leg U-baffle. Tie upper internals pressure drop, 17.36 psi, was adjusted to match the measured total vessel value of 53.4 psi. The value thus obtained matches that measured in the vessel model flow test within 5%. Figure 3-4 shows the incremental 0TSG pressure drops. The new values are listed beside the drawing and the original model values are in parenthesis. In the upper head, a pressure drop of 2.22 psi, was used in order to match the hot functional data, 2*.67 psi total drop. A comparisen of the old and new inlet drops of 2.22 + 0.74a and 5.9 shows a difference of almost 100%. The original drop considered the inlet as an abrupt expansion which is a much larger loss than really exists. Figure 3-5 gives a synopsis comparison between the original and present pressure drop models along with the data which percipita ed the change. The col ur.3.i labeled new forms the basis for the LOCA calctlations presented here in Section 4.2. Figure 3-6 provides a summary of the information presented ; in this section. l 1 I 3-2 l
i: . ; l 0 ', . OTSG oP (3
' ~ ~ - - -
CD
=24.6NttaEc est -
_O_ I I- . ( -/ >9
\ :
19
<4 (-)
i m .. P .. p i ,. h' ( b
~h -
REACTO'R VESSEL AP uf1 REC
= 53.4. PSI PUMP AP = P1-P2 REACTOR VESSEL AP = P1-P3 STEAM GENERATOR oP = P3-P2 .
t . RV AND OTSG AP BASED ON OCONEE HFT DATA . FIG 3-1
V ORIGINAL NEW VMFT CALCULATED RVo P USING SAVER & VMFT 55,90 53.4 53,1* (*BAW-10037, REV 1) CALCULATED SG oP USING HFT DATA 8 SAVER 30.07 24,67 CALCULATED PIPINGoP 16.00 16.23 TOTAL 101.97 94.30 COMPARISON OF ORIGINAL AND NEW , RCS PRESSURE DROPS FIG 3 - O a e k e 4
~
COMPARIS0il 0F CALCULATED REACTOR VESSEL PRESSURE DROPS FIG 3-3 FLOW ORIGINAL NEW REAS0!1 FOR PATH CALCULATI0il
' CALCULATIOil , DIFFEREMCE INLET 12.75 8.62 l u-BAFFLE =[45 DOWilCOMER .20 .21 --
LOWER I!!TERilALS 9.63 12.71 CORE 12.02 14.50 g_gDESIGN VERSlJS CURRE?iT MARK-B
~
UPPER liiTERilALS 21.'22 17.36' -
*It1PUT TO MATCH TOTALAP OF 53.4 TOTAL 55.9 53.4
, NOTE:
THIS oP IS WITHIN 5% OF THE UPPER INTERBIALS 6P AS MEASURED ON THE VMFT. REF. BAW 10012 4 9~
, , ww., w , .-=emmea...--.
OTSG PRESSURE DROP FIG 3-4 AP = .748 p(0RIGIllAL P = 5.9 PSI) 1
. D = 2.22 USED TO MATCH HFT DATA .
(NOTE: ABRUPT EXPAi1SI0i! aP WOULD BE = 5.7 PSI) 6P = 21.07 (ORIGlilAL = 23.7 PSI) AP = .53 (ORIGIt!AL = .52) 6P = .085 EXTRAPOLATED OC0!!EE DATA INDICATES OTSGSPunaec " 24.67 PSI (0RIGillAL APOTSG = 30.1 PSI) E e F**whW g g. W 4 qw w, gg M g 3- ge e N* p+g M' @ S*M > @ W mm @e M% atW h-wMmS h &
COMPARIS0il 0F ORIGINAL AliD l'EW 4 PRESSURE DROP O!! TECO DB-1 IN CRAFT CODE
, 1 FIG 3-5 f ~
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.E A1 1 PATH ORIGINAL NEW REASON liFFERF'k;CE VEW - G Ib
'1-2 13.95 16,87 2.92 CORE DESIGil 2-3 20.97 16,91 - 4.06 VMFT DATA 3-4 8.41 8.30 SAME
~~~4-5 31.77 26.74 - 5.03 MEASURED DA':
5-6 5.86 E.60 SAME 6-1 21.95 19.72 - 2.23 U-EAFFLE
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SUMMARY
- 1. THE ORIGINAL PRESSURE DROPS ARE BASED STRICTLY Oil ,
l CALCULATIONS.
- 2. NEW PRESSURE DROPS ARE BASED ON:
- 1. MEASURED RCS PRESSURE DROPS
' 2 '. APPROVED STAi!DARD CALCULATIOil METHODS (SAVER) (BSW 10072) 3.~~ RESULTS OF THE 177 FA VESSEL MODEL FLOW TEST (B&W 10037)
- 3. THE ORIGINAL AND NEW DATA HAVE BEEN COMPARED AND THE REASON FOR THE CHANGES HAS BEEN EXPLAINED.
1 l '
- I .
,I ~ l . 1 l
1 1 j 4. Revised Loco Pressure Distribution Analysis i Section 3 discussed the revised loop pressure drops. As noted in that section, the new pressure drops are based on test data received from operating B&W plants, or from prototype tests such as the vessel model flow tests. These i revised pressure drops resulted in a decrease of 8.8 psi in the total system pressure drop relative to that assummed in BAW-10105. From the core outlet
~
j to the pump discharge, the overall pressure drop decreased by 14%. The 6-foot LOCA limit case reported in BAW-10105 was reanalyzed using the revised loop , pressure drops, which includes the inlet nozzle correction. The results of the l analysis are present on Table 4-1. During the first half of the transient, leak flow from the break on the vessel side, Figure 4-1, increased relative to BAN-10105. Due to the removal of the U-baffle from the LP calculation for the inlet nozzle, the resistance to flow out the break decreased and thereby resulted in the increased flow out the break. However, flow through the break on the pump side, Ftqure 4-2, also shows increased leak flow during the early stages of the transient. The break on the pump side exhibited the increased flow due to the reduction in the overall resistance from the vessel to the break on the pump side, i.e. from the top of the core through the hot leg, steam generator and across the pump j in the broken loop. While the reactor coolant pump in the broken leg is a resistance during the blowdown, at one second in the transient, it only represents i approximately 70% of the resistance from the vessel to the break on the pump side. The reduction in the pressure drop over this path reduced the overall resistance and resulted in 2% more flow out the break on the pump side. During the later stages of blowdown, leak rates for both paths were lower than that l
~mg ,7-q . . . , - - . . '.._y .w , _.,,,y ,_.. __._ *' ~ ' * ' - - - - --C r-yy- ,-&+ gn,t<--g -+.w_7.__#
i in BAW-10105 due to the decreased system pressure, Figure 4-3.for the revised pressure drop case. The end of blowdown occur-ed at 22.4 seconds in comparison to 24.0 seconds for the 6-foot LOCA limits case in BAW-10105. Figure 4-4 presents the hot spot core flow comparison. The effect of the i i increased flow out the break on the vessel side is compensated for by the
=
increased flow out the break on the pump side. In fact, core flow was increased for the first half of the transient. As shown in Figure 4-6, the increased 4 core ficw resulted in lower cladding temperatures for the positive core flow phase of the accident. During the negative core flow phase of the accicent, the revised loop pressure model did not cool as well as the analysis in i BAW-10105. However, the effect of the decreased cooling during the negative i core flow period was compensated by the increased cooling during the positive J core flow period and the earlier end of blowdown. At the end of blowdown, both cases predicted approximately the same cladding temperature. The REFLOOD model was modified to include the revised system pressure drops. Table 4-2 provides a comparison of the flow path resistances used in REFLOOD for this case and the BAW-10105 case. In this analysis, the additional 0.25 psi pressure drop to account for the effect of steam-water interaction due to high pressure injection was included in the cold legs. Figure 4-5 presents a ccmparison of the reflooding rates. The revised pressure distribution c- se yielded higher i flooding rates than those in BAW-10105 due to the decreased resistance from the downcomer to the break, which resulted in increased vent valve ficw, and the decreased loop resistance, which resulted in more loop flow. A THETA analysis was performed to determine the cladding temperature response. The version used included the post-CHF heat transfer logic described in the l i letter K.E. Suhrke to S.A. Varga dated 1/24/77. A peak linear heat rate of 4-2
18.4 kw/f t wcs used which is the 6-foot LOCA linit reported in BAW-10105. The results of the analysis are presented in Figures 4-6 and 4-7 for the unruptured and ruptured nodes. As shown, the peak unruptured node temperature was 21330F and the peak ruptured node temperature was 19990F. This is a reduction of 330F and 21 F, respectively, relative to BAW-10105. The majority of this difference is caused by the enhanced reflooding rates in the revised pressure distribution evaluation. As shown above, the effect of the inlet nozzle correction is compensated for by the revised system pressure drops. Since trends, i.e. core f'.ow, break flow, etc., are similar to the results presented in BAW-10105, the sensitivity studies and spectrum analyses are performed at the 6-foot elevation, the relative effects of the revised pressure distribution remains the same. For the LOCA limits, the changes in pressure distributien at any elevaticn in the core is small relative to the overall chang throughout the systam. Henca, the effects of the revised system pressure distribution model would be essentially unaffected by the core elevation studied. Further, with the reduction in loop resistances core reflooding rates would be increased. Th refore, the analysis presented in BAW-10105 is conservative and remains a valid licensing document for assuring confonaance of CB-1 to 10CFR50.46. k 4-3 l
l -- l . Table 4-1 Summary of Results ) i ) Original Revised Study System Pressure BAW-10105 Distribution CRAFT Run Name T0415Q3 INCll1V i CRAFT / THETA Linear Heat 18.6/18.4 18.6/18.4 Rate, kw/ft CFT Actuation Time,s 16.61 15.96 Rupture Time /% Blockage,s/% 16.04/74.0 15.76/64.2 l End of Blowdown,s 24.0 22.4 l l End of Eypass,s 24.0 22.4 Mass Remaining in Vessel 2214 1632 at End of Bicwdown, iba End of Adiabatic Heatup,s 33.2 31.4
' Peak Unrupturea Node Cladding 2167/66 2133/59.5 Temperature / Time,F/s Peak Ruptured Node Cladding 2021/43.5 1999/41 Temperature / Time,F/s Local Metal-Water Reaction,% 3.25 4.62 j
4 l 1 l l 4 9
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n,,. ,, ,, ,.,- ,. -y-,.g,_ , - - .
s TABLE 4.2 Comparison of Flow Path Resistances for REFLOOD Revised System flode 2 BAW-10105 Pressure Distribution From To Area, Ft (a) Resistance, K Resistance, K 1 5 7.574 2.003 1.665 5 6 17.95 5.343 3.866 6 7 25.52 9.0375 8.5397 7 8 6.3495 26.01(b) 25.76(b) 7 9 6.359 26.05(b) 26. 53(b&d) 8 4 6.257 2.8897 4.8755(d) 1 SP 7.574 2.003 1.665 SP 6P 17.95 5.343 3.866 6P 7P 25.52 9.0375 8.5397 7P 9P 12.699 26.01 (b) 25.76(b) 9P 4 12.514 2.8897 4.4(dl 3 2 45.89 6.233 7.581 2 1 67.301 12.068 14.72 1 4 .4.276 4.2(cl 4.2(c) CFT 3 .7213 6.423 6.423 a - average areas over length of flow path h - includes locked-rotor pump resistance c - vent valves d - includes .25 psi penalty for steam-water mixing
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---------- Revised System Pressure Distribution ,
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