ML19210A731
| ML19210A731 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/23/1978 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19210A727 | List: |
| References | |
| NUDOCS 7910310546 | |
| Download: ML19210A731 (20) | |
Text
{{#Wiki_filter:. FETROPOLITXI EDISON CCMPA:iY JERSEY CE:iTRAL PCWFR & LIGHT COMPMIY AND PENNSYLVA'i!A ELECTRIC CCMPANY THREE MILE ISLAND NUCLEAR STATICN UNIT 1 Operating License :io. DPR-50 Dceket No. 50-289 Tecnnical Snecification Chance Recuest No. 8h This Technical Specification Change Request is subnitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Statien Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included. METROPOLITA'I EDISON CCMPANY f gy - rM -
/ < ice President Sworn and subscribed to me this d3 day of , 1978.
V
& , / 75 Notary 7tiblic [jg GEORGE J. TROFF2R Notary PubUc, Readr:g, Sarks Co.
f% C:mmI:si:n Ex;::re: .'an. 25, W^2 1493 174 E91031u Y 9f
**hree Mile Island Hucle=.r Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Dcchet No. 50-289 T g} Technical Snecification Change Reauest No. 8h BBD Bd \b The Licensee requests that the following changes be made to the TMI-l License No. DPR-50, Appendix A, Technical Specifications: ,
- 1) The attached pages 2 k, 2-9 and 3-3ha replace the corresponding pages ,
of the existing ca-k '-C Specifications.
- 2) The attached fisres 2.1-2, 2.3-1 and 2.3-2 replace corresponding figures of the existing = ' '-=' Specifications.
- 3) he attached fig;.res 2.5-23, 3.5-2D and 3.5-2F be inserted where there e.re pages reserved r:r :le figures in the existing Technical Specifications.
? ear :: f:r Change S~-'-=~ Specificatic: :htnge 2equest No. 70 (January 9,1978) was submitted basei :n a Cycle 3 bu nup :f 2 0 1 10 EFFD. Amend =ent A to Technical Specifica-ti:n Int ge Request 30. -- ( A;ril 03,1978) was submitted as a conservative set o' a-- ' cal Specificati::s basei en a 315 EFFD Cycle 3 and only specified operating para =eters to 125 1 5 E ?D of Cycle 4 operation. This Change Request in:lude: .he changes resi ing frem the actual 28T.1 EFFD Cycle 3 and refine-nent of the conservative Technical Specification subsitted as Amend =ent A to Technical S1 ecification Change Request No. 70, to extend Cycle 4 operation to 265 1 15 EFPP, as well as those changes to operating limits necessary to ac-ccanodate the discrepancies between measured and predicted radial and total peaking noted during Cycle h startup.
In Metropolitan Edison Cc=pany letter (GQL 07h3) of April 20, 1978, a co..mitment was made to reanalyze the reactor coolant system peak pressure folleving a feedwater line break and to submit the results with the refined cycle h Technical Specifications. Therefore, this submittal also includes the revised reactor coolant system high pressure trip set-peint based on the reanalysis performed by Babcock & Wilcox. The revised setroint is acceptable for cycle h and subsequent cycle cperation. Safety Evaluation JustiReing Change Tne Proposed Technical Specifications and the Cycle h Reload Report (Revised June 1973) are structured relative to the current Technical Specifications and c?arating history to date. The revisions made to the Cycle b Reload Report (Perision 1, January 1978) include the changes based en, 1) Arend ent No. 39, April 27, 1978, (in response to Tech Spec Change Request No. 70,Atendment A, April 3,1973); and 2) A=endment No. LO, May 19,1978 (in response to Tech Spec 1493 175
Change Request Hos. 79 and 80); 3) the ret valuation of Cycle h based on the actual EOC-3 burnup of 287.1 EFPD to determine the key parameters for a Cycle h burnup from 0 to 280 EFPDs; h) the generic study over the small break LOCA spectrum; and 5) the high pressure trip setpoint renalysis. Centerline fuel melt margins and steady-state DNB margins to limiting linear heat rates were reviewed for impact on the Core Protection Safety Limits and Setpoints for Cycle h. Centerline fuel melt margins versus reacter power is-balance are more restrictive than the current TMI-1 Technical Specifications due to the power distribution anc=oly observed in the Cycle h Startup Tests, and as a result, the calculational nuclear uncertainty factors were increased . to 115 for the radial peak and 13.50% for the total peak. Figure 2.1-2, " Core Protection Safety L its, CC-1" indicates a negative imbalance scre restrictive than the current negative imbalance safety limits. Further refinement in the analysis to take inte acccunt the calculation / measurement adjustment, resulted in changes to Figre 2.3-2. Figure 2.3-2 and Figure 2.1-2, which provides the basis data for Figure 2.3-2, have been refined to account for the extension of Cycle 3 to 2871 EFFDs and to incorporate the increased uncertainty factors. 3cth Figures are valid fec Cycle h (0 to 280 EFPDs). The DN3 dependent points cf the pressure temperature limit curves (variable 1cv pressure trip) and the flux / flow trip setpoints are based upon. design peak-ihg racher than calculated peaking. The Cycle h cargin between calculated nuclea pin peaks and the reference design peaks used for ther=al-hydraulic analysis was found to increase due to the Cycle 3 extension. Therefore, all Technical Specification limits based en design peaking remain conservative for Cycle h. The the=al-hydraulic rod bow penalty and fuel temperature / pin pressure analyses were reviewed for applicability during Cycle h operation. Both the
=aximum assembly burnup (31094 K4D/>CU) and the hot assembly burnup (16,000 M4D/5EU) at EOC h (280 EFPDs) were below the assembly burnup assumed for the rod bov penalty (33000 E4D/MTU). Also, the calculated maximum pin burnup (35hhT E4D/>CU) and the pin power history were bounded by the values used for the fuel temperature / pin pressure analysis.
The =echanical. performance of the fuel was found to be acceptable relative to cladding stress, strain and creep collapse for Cycle h. Therefore, all Technical Specification limits based en fuel integrity are valid for Cycle k. The two-hour thyroid dose vis reanalyced using the actual EOC-3 burnup of 287.1 EFFDs and a Cycle h of 280 EFPDs. The Cycle h deses remain only a very small fraction of 10 CFR 100 limits, and are acceptable for Cycle h operation. The accident analyses were reviewed based en the refined Cycle h values, and the Cycle h parameters were compared to the FSAR/densification report analysis values (See Table 1). Table 1, in conjunction with Table 2, illustrates that for all accidents considered in the FSAR, the initial conditions defined by Cycle h prra=eters produce less severe transients than the initial conditions assumed 1a the FSAR analysis. Because the Cycle h transients are bounded by previously accepted analyses, no reanalyses of these events have been performed. 1493 176
D**D 7~% j wwS & h$ = Se:ti:n 3.5.2, " Control Rod Group and Power Distribution Linits, was reviewed f:r changes in shutdown margin, ejected rod vorth and peaking =argin to LOCA kv/ft limits for Cycle k. The current Technical Specification Figures 3.5-2A, 3.5-2C and 3 5-2E are more restrictive than the like figures developed from the =ost recent analysis. Therefore, in order to expedite the licensing re-view process, the current, more restrictive, limits for Figures 3.5-2A, 3 5-2" and 3.5-22 have been retained for the first 130 EFPDs of Cycle h. The remaining Cycle 4 period (130 to 280 EF?Ds) vith respect to rod position , limits and the LOCA dependent peser imbalance envelope, vill be governed by the attached Figures-3 5-23, 3.5-2D and 3.5-2F. In general, the rod position , licits have a slightly Ore restrictive shutdown margin limit curve and per-rissitie Operating regien than the figures previously submitted. The high p:ver perating regi:n :f Firres 3.5-23 and 3.5-2D has been retained by im-pesing tre restri::1 e p:ver i= balance envelope (Figure 3.5-2F), and AFSR p si-" -4t curve lFirre 3.5-2H, A=endment No. h0, May 19, 1978). Re-visei calculations inii:st a that Cycle h's mini =um shutdown margin at 280 EFF:s was 2.19% Ak/k, v c- < ve the 1% &/k requirements. The -e4"" efe::ei rod verth was 0.3?" l'-:/k at Hot Zero Power and 0.25% Ak/k at Hot Full ?:ver, ve'.1 below the l' ':s Of 1.0% Ak/k and 0.65%Ak/k, respectively. Fi: re 3.5-23 for 2 and 3 purp cperation was developed by power scaling the li:1 ing b:: dary for the "?e issible Region" of Figure 3.5-2B in the recog-niti n :f the ordinate cf Firre 3.5-2D. The shutdown margin limit of Figure 2.5-23 vas not scaled, aiiing aliitional conservatism to that limit in Figure 3.5-22 . . 3ased en the above safet- evaluation review, it can be concluded that the revisei Technical Specification changes to those previously submitted for Z -l's Cycle h support a full power Cycle h operation for 0 to 280 EFFDs without endangering the' health and safety of the public. Pages 2 4 and 2-9 and Figure 2.3-1, have been changed to reduce the RC System high pressure trip setpoint to 2390 psig so that in the event of a feedvater line break, the peak RCS pressure vill not exceed 2750 psig. Attached to this safety evaluation are the results of the reanalysis which justify the 2390 psig set-point for Cycle h and subsequent cycles. TMI-1 reduced the setpoint to 2190 psig prior to obtaining greater than h0% FP during Cycle h startup. This issue d0es not involve a safety concern and is resolved with this Technical Specifica-tion sutrittal. Face 3-3ha is being submitted to correct a typographical error in the quadrant tilt value as deternined by the out-of-core detector system. The value should be +22.96 rather than the current Tech. Spec. value of +22 92%. Tr.ese Technical Specification changes discussed above and the Cycle b Reload Report (Revised June 1978) support a full power 28C EFFD Cycle k, and do not create a threat to the health and safety of the public. Cisssification of A ndment (10 CFR 170.22) Tnis a.endment to authorize operation in Cycle k to 265 + 15 EFFD is the result 1493 177
h-of refinements to the original Cycle h Technical Specification Submittal (Technical Specification Change Request No. 70, January 9, 1978, as amended April 3, 1978) and do not involve a significant hazards consideration. There-fore, the Licensee has determined that this is a Class III amendment. The reactor coolant system high pressure trip setpoint submittal included herewith is a refinement of the analysis submitted on April 17, 1978 (GQL 0669), and is in response to our eccmitment of April 20,1978 (GQL OTh3). The License
- has deter =ined that the resulting changes to the Technical Specifications are not subject to the scheduled amendment fees. ,
Therefore, the proper renittance for this Change Request is $h,000.00. 1493 178
TABIE I COMPARISON OF KEY ACCIDENT AND TilANSIEN1' PAllAMETERS F0ll CYCLE h# FSAR and deasif'n Parameter report valre Predicted Value l Doppler coeff (BOC), A k/k/ F -1.17 x 15- -1.48 x 10 -5 Doppler coeff (EOC), A k/k/ F -1.33 x 10- -1,60 x 10-E Moderator coeff (B00), A k/k/ F +0.5 x 10 -h -0.71 x 10-Moderator coeff (EOC), A k/k/ F -3.0 x 10- -2 53 x 10-All rod group worth, % A k/k 10.0 8.62 Initial boron conc. (IIFP) ppm 1200 1045 Boron reactivity worth (70 F), ppm /1% A k/k 75 73 Max ejected rod worth (liFP), % A k/k 0.65 0.25 Dropped rod worth (IIFP), % A k/k 0.h6 0.20 -A> ,gy
- Based on a 287 EPPD Cycle 3 and 280 EFPD Cycle h.
L.e3 4 e
TABLE 2 COMPAF4 SON OF CYCLE h* PAllAMIS'ERS TO Tile FSAH Cycle 4 margin Safety Margin greater than Trarsient Key Parameters (y) Increases with FSAR margin? Moderator dilution accident Initial boron lower concentration yes concentration (BOC). Boron reactivity worth larger value no(2) Moderator coefficient more nigative value yes (BOC) Cold kater Accident ModeraLor coefficient less r.egative yes (EOC) Doppler coefficient more negative yes (EOC) Loss of Coolant Flow Doppler coefficient more negative yes (BOC) Moderator coefficient more negative yes (DOC) Steam Line Failure Moderator coefficient less negative yes 9 (EOC) Rod Ejection Accident Doppler toefficient more negative ye's 45* (BOC) m Moderator coefficient more negative yes __. (BOC) co C3 Ejected rod worth smaller worth yes
- Based on a 287 EFPD Cycle 3 and 280 EFPD Cycle h
- Cycle h Margin Surety Margin greate- than Transient Key Parametern(y) Juereanen with PSAR Margin't Dropped Control Rod Dropped Rod Worth Smaller worth Yes Mclerator Coefficient Less Negative Yes (EOC)
Doppler Coefficient Less Negat.lve No (EOC) Rod Withdrawal Accident Doppler Coefficient More Negative Yes (BOC) Moderator Coefficisnt More Negative Yes (BOC) NOTES: (1) Certain key parameters do not normally vary from cycle to cycle. These have been excluded in this table: RC pump flow; RC pump flow characteristics, and design radial-local and axial peaking factors. (2) The quotient of initial boron concentration and boron worth yields a lower reactivity addition (higher safety margin) than the FSAR reference analysis. g (3) Although a more negative EOC Doppler is not conservative for the dropped rod accident, the transient results are still conservative with respect to the FSAR analysis because of the smaller Cycle h dropped rod worth and less negative moderator cuei11cient.
~
W U
May 1, 1978 PEAK RC PRESSURE FOLLOWIUG FEEDWATER LINE BREAK TMI-l Cycle h The Peak Reactor Coolant System Pressure follovirg a Feedvater Line Break has been recalculated for Cycle 4 and subsequent cycles based on the following as-su=ptions.
- 1. Safety valve relief rate - 156#/s @ 2500 psig (combined relief rate)
- 2. Rosemont pressure transmitter error of 45 psi in a degraded environment ,
- 3. Trip string delay time of h50 =s .
L. High pressure trip setpoint 2kO5 (curve 1); 2395 (curve 2); and 2390 (cume 3). Tne results in the attachei figure show that unacceptable peak system pressures result with the use of the 2LOS psig trip setpoint for Cycle 4 and subsequent cycles. This conclusion supercedes that submitted to the NRC April 20, 1978, G;L 0%3, where too great a safety valve relief flovrate was used. To pr: duce acceptable peak system pressures following a feedvater line break for ?/cle h, it vill be necessary to lower the high RC pressure trip setpoint tc 2335 psig, keeping the safety valve setpoint at 2500 psig. This change vill resn't in acceptable peak pressures for a reasonable range of moderator reactivity coefficients. The attached figure (curve 3) indicates that the peak system p- a " " ~ is less than 2750 psig for moderator coefficients more negative than -0.35 X 10 d AK/K/F. This, acceptable range includes the present Cycle h moderator coefficient of -0.63 X 10 " AK/K/F. The analysis was performed also for a 2390 psig setpoint for reactor trip. Feat syste pressure re=ains below the 2750 psig limit for all negative moderator coefficients. It can be concluded that a high pressure trip setpoint of 2390 psig would suffice for Cycle h and all subsequent cycles since a Tech Spec exists to assure that the moderator coefficient will always be negative. The analysis performed is insensitive to Doppler coefficient since the fuel terperature changes (increases) by only a few degrees prior to developnent of the peak sy This analysis is based on a Doppler coefficient of -1.L9 X 10 gtem pressure. AK/K/F vhich should be conservative with respect to subsequent cycles. The results of this analysis indicate that a setpoint of 2395 psig is necessary to give acceptable results for Cycle 4. A setpoint of 2390 psi- will produce acceptable results for Cycle h and subsequent cycles. *fhe 239. psig high RC pressure trip setpoint has been incorporated in the TMI-1 cyc'.e h refined Tech Spe:s in order to have a fixed setpoint for subsequent cycles. M 1493 182
i
.i.......
i' I l l ; ..;.. t
! t- : l I
l l ,! ! - l l l t -
' i j- -
,i -. l l.
- l .
l l i . ( ,r- ..'.l .
.r--
I ! i - - '
-- ----{ -- -
. t i i
, e i r
l J.9(,o- , ,
. ; : . 1 .. j . I i i L _ _. .
.._._ __.t _ _ _ _ _ . . . . ... .__i .. . . ;
[ l 1. ,. i e 4 , i O .
. .l :!- +
/ . g .1 - *
.~ !
-!!: .l + -
-l ,,,,,,,-
# i- lL : i- i .
- i. : l.
d.M8
":l Q
.....s.
lm m [ - fi
.,p. .:.. _ . ._. .. :.
- l l- t-
.. n I: -
f _.
.i ...
p .
. _ . . . . . ...g
..t. * - -
- .. .i.
.l. - . . . . .
.i: -
- 1
.i . . i:r-- \ -
.1 2. !
/ -
i:
,a .
~4 ., (./NAcc.cw4c:.s 4 # i I "'. - i' l
c 'q_e.2So ., y n 9 S 1 - t
.e ( p, ..
- 1 :..t:_
l.
- pmspyg
--l.
i
- l
=: ..... = ij.n .;. g { .t. ,
-l. ir . . g. j . ,. j: ;l: ,
.--j : Q :- - '%:)' , \.i.AJ ';i. . g . , . : l ! . .. !i.. --gl:.i..
]-"'
~
;i 21-
.... ..,_, . -i 4r ]: t- f :li. _'i. - t. .1 J- F ~p :l: -
. 'trli.l: : i. :
,:: t t,.3
.- = = }.:2: = V , -n,, '. .. . . . . .
- .nt . r= .
.:..in-5:'. :rr. ....
i .
.i .!
r:' . . 4 . , .]
._. y
_. . _ . . .p..r. .: p
. ;. .. 1.: -.. :. 7. .. . c. . ...
. . . . . . }. .. t. : .
. . 7.
i: 7- . . . . , . .. 1- T-~- l'"
~ ~~ =. +i -T. :-i;izi: -- ..-- -
2-li-- r- ----r=-- - --- -- _.. .._. ._...g.._-__ . = - - - - - =;= ,_: : ...
= =- lv. -_.:=-- =
. , . , , , . . _ _ . . _ _ . . _ . . . . . = = - = = = -:- = . .. . , : -- c -- - - - - - - -
u-
':!.pr3r ' iri :=1-H - := . :. :M =n=
-- - iEy;;; .gj,i- j;_.j;l . .. gi.~..=1_,
_ . h:-- . . _. . .. _... .r- - . . . _ . . _ _ . . . _ :...... _ _ _ _ . ._..__. _ _ . . . . _ _. .. _.. I : =- =_-! n1 : t -- -
- -t---
=n= : == =; = : ; == = l _~. ut. .: =_.
__ _ . s == : r= {, =:= =.= .. u.-l=:: _ === =:==.__...... : -:= mi--- n: - l- .a- - --
- .2.
. f _=__..._....l_..i_. ___. _ , . . .
= p.- .-t. . , . . . .._ . . . . . t. ..mr :
..7.....
..; _ _ } _ _ _ _. = r- .: _ ---- -- -- -
..t---
- s. .g r___. i L; =__[_; =u .~l . . . . . _
s . . .0, .. lg,,. 1._ _ , _. . ..o.y
--t_..... . j . .. - _ ,. .} =. . . . . . j . . t .r-o_3 :=I ,4.to :l. , ,r r= t - .
p : = ;- _ .p.... _ , p:= = .r...- it .____.. a W W6=. W . _. . r N --- : t_ Moor etho r2 ~~do; A-Wic/E.U"- .t :j=Vri yo - :- - - =f---- - t-N:: ---$1-3~5I ~-T '.!$- -
--iEli-i}- 5-* : Im---{~=IE-rIN- =I htii: lMm'
-+ --
-5l#- IIir p .. _ . . l---. - +p-.-. . .. . . . . . .. p .;.. .. r .b 7t:- : :::: f n :b=j=: - l unj u . ...:!=:-
- : .l.
__ = f=. . r - . . g . _ - ;.:l=j..jn: _. .I . .. m : _. 'nn((1..
. .:p.
. . . ---+ .--'f .--4i.:..
3 =pr . cup r, . .= p_..
- .: .;}: .l:.
.y. .;$. . :p g . _ . _ , . ....... . . . ... .
- 7. .
g.. . ....g. .. .p : .;; q. .p ,
.-..g .j,...l. .g .l. .
..f . ....,. , g.g.. .
, gg [ g g .g g. j i nnJ-- -iji. . : :. :jli ..:]i r - -[i::. g :
[ gg ' _g i
=t . .
--"t:
..j... --
t-
.f.
._pPr-
' " r ---
---" .i. :!: ....j pM fe .43 r fp Go! AA l
rd /h7- :
?
- l! :I ii:l: ij: : it. i I: 1. !!: i: :h ' .
l t - l- !- !- .(
.l. 1:
_S.v_.RGMEE_& tire : Afo, % e S w psia
? t .i. :b t -
. .: ( i. l t i
.l;-
ERESS usel . r/Mns'...Eencrz;_.%T [. t I i N1 I
'i '; 5.. ';-_..
,Ii .
t i l l l' _j-^r
,.!1']G.]} P A?i3 SQ_^lt 1", 7tClP! ! ;
[
.W - l8 40 .l 3 L .2 Q . .
.... ,. h, o ....'
f - l..
.oeJ.>, v a ns.
.._i -
i
- i..l...r -
!: .i .
i l .
. i l
I l t 1493 183
2.2 SAFETY LIMITS - RFACTOR SYSTEM PRESSUP.E b'Jhh Anolicability , UU[jk ,i C g YL
' Applies to the 14-4t en reacter coolant syste= pressure.
Objective To =aintair. the integrity of the reactor coolant syste= and to prevent the . release of significant a: cunts of fission product activity. i Seecificati_on . 2.2.1 Tre reacter cecitnt syste= pressure shall not exceed 2750 psis when there are fuel asse=blies in the reactor vessel. Eases s The reactor coolz_. syste: (1) serves as a barrier to prevent radiocuelides in the reactor coolant frc= reaching the atc= sphere. In the event of a fuel cladding failure, the reae cr coole=t syste= is a barrier against the release of fission products. Establishing a syste= pressure li=it helps to assure the integrity of the reactor ecclant syste=. The naxi=== transient pressure allovabic in the reacter coolant syste= pressure vessel under the ASMI Code, Section III, ic 230% of design pressure. Thus, the safety linit ,of 2750 psig (110% of the 2500 psig design pressure) has been established. (2) The =axi=u settings for the reacter high pressure trip (2390 psig) and the pressurizer code safety I valven (2500 psig) (3) have been established in accordance with ! ASME Boiler and Pressure Vessel Cede, Section III, ArtTele 9, Winter,196o to assure that the reactor coolant syste= pressure safety l'i=it is not exceeded. The initici hydrcstatic test was ceniucted at 3125 psig (125% of design pressure) to verify the integrity of the reactor ecolant syste=. Additicnal assurance that the reactor coolant systa= pressure does not exceed the safety li=it is provided by setting the pressuricer electrc=atic relief valve at 2255 psig. (h) Seferences (1) FSAR, Section h (2) FSA3, Section b.3.10.1
!493 184 (3) FSAR, Section h.2.h.
(L) FSA3, Tabic L1 2-h
TABLE 2.3-1 REACTOR PROTECfION SYSTEM TRIP SETIING LIMITS Four Reactor Coolant Three Reactor Coolent One Reactor Coolant Pumps Operating Pumps Cperating Picnp Operating in (Nortinal Operating (Hominal Operating Each Loop (Nominal Shutdown Power - 100%) Power - 75%) Operating Power ':9%) Dypnoo
- 1. Nuclear power, Etx. 105 5 105 5 105 5 5 0 (3)
% of rated power .
- 2. Nuclea Power based on 1.08 tirno flow minus 1.00 times flow atinus 100 times flow minue Bypassed flov (2 and imbalance reduction due to reduction due to reduction due to max. of rated power imbalance (s) imbalance (s) imbalnnce(s) 3 lent power based NA NA 91% Bypnosed
{ugonpumpmonitors, 5 max. % of rated power I4. liigh reactor coolant 2390 2390 2390 1T20 b} y system prescure, psig, w mnx. 5 Low reactor coo 3 ant 1800 1800 - 1800 , Dypassed nyotem pressure, poig min.
- 6. Varinble low reactor coolant nyotem pressure (11.T5 Tout-5103) (1) (11.T5 Tout'-5103) (1) 11 75 Tout-5103) (1) Bypasaca g
psig, mia.
~
O i. T. Reactor coolant temp. 619 619 619 619 F. , Itax. U i __ 8. Iligh Reactor Duilding is I4 I4 18 co pressure, poig, max. (1) Tout la in degreco Fahrenheit (F) (2) Henctor coolant pystem flov, % *D ' (3) Ad:ninistratively controlled reduction act only during reactor shutdown (1 ) Automatically set when other segments of the RFS (no specified) are bypassed 1 (5) The pump monitors also produco a trip on: (a) 1000 of two reactor coolant pumps in one reactor coolont loop, and (b) loca of one or two reactor coolant pumps during t'vo-pump op ation. (6) Trip uett.ings limits are setting limits on the setpoint side of the protection sys@m bist
- 2. The control rod group withdrawal limits (Figures 3.5-2A, 3 5-2B, 3 5-2c, 3 5-2D, and 3.5-2H, shall be reduced 2 Percent in power for each 1 percent tilt in excess of the tilt limit.
- 3. The operational imbalance limits (Figure 3 5-2E, and 3 5-2F) shall be reduced 2 percent in power for each 1 percent
' ~
tilt in excess of the tilt limit.
- f. Except for physics or diagnost'ic testing, if quadrant tilt is in excess of +26.75% determined using the full incere detector system (FIT), or +15.21% detemined using.the minimum incere detector system (MIT) if the FIT is not available, or +22 96% l determined using the out of core detector system (OCT) when neither the FIT nor MIT are available, the reactor vill be placed in the hot shutdown condition. Diagnosite testing during power operation vith a quadrant tilt is permitted provided that the thermal power allevable is restricted as stated in 3 5 2.h.a above.
- g. Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated pover.
l493 i86 - O e 3-3ha
Thermal Power Level, f.
-- 120 (112)
I (44,112) DNBR Limit (-30. l l 2) -
- 110 ACCEPTABLE 4 PUMP OPERATION -
- 100 Kw/Ft Kw/Ft .
Limit ^~ Limit (-30,87.I) -- 90 (87.1) 2 (44,87.i) (-53,80) ACCEvTABLE (s o , s o )
- 80 3&4 .' UMP OPERATIOS 70
(-30,59.6) -- 60 (ss 6) 3 (44,59.6) (- 53,55. I ) (58.4,50.4) ACCEPTABLE (-42. 4,42,4) 2,3 & 4 PUMP -- 50 (L9.2,49.2) OPERATION
-- 40 .
. 30
-- 20 10 1 I I I 1 t l .
l t 50 -40 20 -10 0 10 20 30 40 50 60 , Reactor Power Imbalance, f. Curve ReactorCoolantFlow(Ib/hr) I 139.8 x 10 6 2 104.5 x 10 6 3 68.8 y 10 6 CORE PROTECTION SAFETY LIMITS T M l - 1, Cycle h Figure 2.1-2 1493 187
2500 P = 2390 psig 2300
^
E ACCEPTABLE T = '619 F p OPERATION I 2100
/
8
~r 5 1930 @
I
- UNACCEPTABLE
$ g '
- P = 1800 psig OPERATION 1700 1500 B00 620 640 540 560 580 Reactor Outlet Temperature, F TMl-1 PROTECTION SYSTEM MAXIMUM ALLDEABLE SET POINTS Figure 2.3-1 1493 188
Inermal Power level, $
- 120
(-f7,108) - 110 (108) (17,108) l l
- 100 81 = 1.28 M2 * -1 8 f ACCEPTABLE 4 PUMP
- -9 '
(35,90) ' OPERATION (-35,85) I (80.7) i _ gg
/ I ACCEPTABLE
- 70 3 & 4 PUMP OPERATION (35,62.7)
-- 60
( -3 5,57. 7 ) l - I (53.1)[ - l
-- 50 1 I ACCEPTABLE - 40 2,3 & 4 PUMP OPERATION (35.35.1)
(-35,30.I) -_ 30 , l , I l -- 20 ' I = e o 9 lC ' 10 ra .
-l II
~
ln , n ,
= l
' ' ' i iI i , , , , , , ,
60 -50 -40 -30 20 -10 0 ' 10 20 30 40 50 60 70 Reactor Power Imualance, 5 PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS FOR REACTOR POWFR IMBALANCE TMI -1, cycle h Figure 2.3-2 1493 189
POWER 234,102 274.1,102 LEVEL 100 - NOT ALLOWED bul0FF
= 925 g ,
274.1.9 -- RESTRICTED 80 -
, 248.2,80 70 -
SHUTOOWN MARGIN LIMIT 200,70 g [ 60 - E PERMISSIBLE N 50 - 173.50 180,50 OPERATING E REGION 40 - d a 30 - 20 - 122.15 140,15
.a _
0 0.2.3 i i I a e t e e s . h 0 25 50 75 100 125 150 175 200 225 250 275 300 _ Rod Inder, % Withdrawn 4 m
- O 25 I
50 I 75 l
. 100 Q . e Grcup 7
_ 0 25 50 75 100 , I I 'l t up 6 ROD POSITION LlHITS FOR 14 PUNP OPERATION O 25 50 75 100
' ' ' 8 FROM 125 1 5 EFPD TO 265 i 15 EFPD Grcup 5
. THi-l, Cycle 14 Figure 3.5-2B
~ $
Q *,
o a N
** E
" n m
n.
. 2 -
o
- n. .. ;
e m e4 N* o - o N o 0 o o- % M og O o e %O w N +
-o -
* ;0 - Nm
.. - N .
N. 2 - - m - 2 g a o z a2' _>- o o w _ m N o N m o w m z a
-x 2 O >. x 65, -o o .*
- a<- o e o
O se - o *- N 2 w w - N e-cc a m -o a. < o w N N
- a. o g
y
. o y o a.
m a o N o o N e 2 m h. m
-c m
N 3
- o m-N o e 3 o
x e. . o u o n,, o o-ch a o > y - e e m - - e
< c o.
i 2 >- - c m 3 z- . 2 o o ..a , e ce m o o N N o
=>
z V e
=
- 8 o -0 w CL 2
o3 m a. m o-N m o co N
~
=c m o >- x a w us < m
- s w O o o o e m u a = N o
.J k.
o x - e m-N N m. N. o I I I f 1 I e f f f o o-o o o o o o o o o o o o o e N w e a n N - 3[Cemo((y *2tH jo % 'J Wod 1493 191
Power, 7. of 2535 HWt
~
RESTRICTED REGION
-22.25,102 -
100 ' ..
-23.0s,92 >
o Il.26,92 g f
-27.72,80 --
80 o iI.26,80 .
-- 70
-- 60 PERMISSIBLE OPERATING --
50 REGION . f-
-- 40 30 20 10 I f f f I I l .I I -
-40 -30 -20 -10 0 10 20 30 40 50
, Axial Power Imbalance, 7 POWER IMBALANCE ENVELOPE FOR OPERATION FROM 125 i 5 TO 265 15 EFPD TH i - 1, cycle h Figure 3.5-2F 1493 192}}