ML19206A560
| ML19206A560 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/07/1978 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19206A540 | List: |
| References | |
| BAW-1497, NUDOCS 7904200142 | |
| Download: ML19206A560 (20) | |
Text
.
EAW-1497 I
Jime 197R 1
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JUSTIFICATIO:1 FOR RD' OVAL OF ORIFICE ROD ASSD13 LIES Ili THREI MILE ISLA2m UNIT 2, CYCLE 1 1
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BAECCCK & '4ICLOX 1
Poaer Generation Group Iluelear Pcaer Generation Divisio1 P. O. Box 1260 Lynchburg, Virginia 24505 9, s. ? cock & Wilcox 4B h I:
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1-1 1.
INTRCDUCTION.
2.
TEE?liAL-HYDRAULIC DESIGN.
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COT.E LOADING PLAN
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u IF !ATICNS 5-1 5.
PROPOSED MODIFICA~I:U.
A-1 REFIRENCES.
List of Figures Figure 3-3 3 _.
Feedwater Te=perature Decrease 4-2 4-1.
niI-2 Initial Core Loading Plan 2.1-1.
D1I Unit 2 Reactor Core Safety L1=it 5-2 5-3 2.1-2.
Reactor Core Safety Limits.
2.2-1.
Trip Setpoint for Nuclear Overpower Based on RCS Flow and 5-4 k:ial Pcuer I= balance 2.2-2.
A11oaabic Value for Nuclear Overpcuer Eased on RCS Flow and 5-5 kcial Power Imbalance 2.1.
n!I Unit 2 Pressure /Ter.perature Limits at Maxi =u= Allowable 5-6 Power for Mini =ua DNER.
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1.
INTRODUCTION This report provides justification for continued operation of the first cycle of Three Mile Island Unit 2 (TMI-2) at the rated core power of 2772 MWt follow-ing the re= oval of orifice rod assemblies (ORAs) fro = the core.
The ORAs are used to li=it bypass flow through fuel asse=blies with empty guide tubes.
A systen flow of 102." of design flow has been used in these analyses which offsets the increased core bypass flow due to re= oval of ORAs.
An evaluation of thernal-hydraulic perfor:snce has been =ade based en the l
increase in system fler and renoval of ORAs and has been compared to the anal-I and Fuel Densification Report.2 This evalua-yses presented in the TMI-2 FSAR tion shows that the effects of the recaval of forty ORAs and the increase in reactor coolant flow rate provide i= proved safety =argins relative to those j
reported in the TMI-2 FSAR1 and Fuel Densification Report.2 1
The use of retainers 3 to provide positive.olddown of burnable poison rod as-l s emb lies (SPRAs) in the retainder of cycle 1 has also been considered.
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2.
T9ERMAL-h7DRAULIC DESIGN j
The ther=al-hydraulic design evaluation supporting continued cycle 1 operation l
used the =ethods and =odels described in reference 2 with the following excep-l tions:
i 1.
An increase in core bypass flow due to CRA renoval.
2.
An increase in syste= flow, l
3.
The inclusion of retainers to provide positive holddown of EPRAs.
j During the initial porti:n of cycle 1 operation, fuel asse=blies which did not I
contain control rods, II..is, or neutron sources had ORAs installed in the guide tubes to =inimize ::re bypass flow.
The =axi=un core bypass flow, with ORAs installed in forty fuel asse=bly locations, was 6.04% of syste= flow.
Thirty-eight ORAs will be recoved for the re=ainder of cycle 1.
Two fuel as-t j
se=blies will contain primary neutron sources and codified ORAs.
The thernal-hydraulic analysic assumed a total of forty vacant fuel asse=blies and resulted in a taxi =u= core bypass flow of 7.6%.
As previously noted, a systen flow of 102% of design flow was used in the anal-l ysis (see Table 2-1) which offsets the affect of the increased bypass flow.
This syste= flow rate is conservatively based on a predicted four-pu=p flow rate of 105% of design flow as verified during startup testing.
Retainers will be installed on all fuel asse=blies containing BPRAs and pri-t
=ary neutron sources with codified ORAs. This retainer design is described in reference 2.
The additional for= loss due to retainer installation has been included in the calculation of core flow distribution.
The limiting fuel asse7bly does not contain a IPRA during cycle 1 operation.
Maximum design conditions and significant parameters are shown in Table 2-1 for cycle 1 operation with and without the CRAs.
The potential affect of fuel rod bow on DSSR was considered by incorporating suitable targins into DN3 limited core safety limits and RFS setpoints (pres-sure te=perature limits and flux / flow setpoint).
The ma:.i=un rod bow penalty was cal:ulated fro = the equation:
2-1 Babcock a \\Vilcox
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.c i2 - 0.065 + 0.001449 /_3U_
Co whe' e' AC = rad bow =agnitude, =ils, C, = initial gap (138 mils),
BU = =axi=uc asse:bly burnup, C'd/=tU.
The pressure-te=perature limit cur"cs shcwn in Figure 2-1 (section 5 of this report) provide the basis for the variable low-pressure trip setpoint.
These curves have been changed fr:= those of bases Figure 2-1 of the TMI-2 Technical Specifications.4 The revised pressure-temperature 11=1ts cover an 11.2% fuel rod bou penalty, based on an assumed taxi =ua assecbly burnup of 33,000 mwd /ctU, while incorporating the core flow changes discussed chose.
The flux / flow trip setprint was deter =ined by analyzing an assumed one-pucp coastdown starting frc= an initial indicated power level cf 102%.
The Tech-nical Specification flux / flow setpoint of 1.05 was re-evaluated based on the initial conditions deter =ined with the ORAs out.
The 1.05 setpoint provides coverage of a 9.1% rod bow penalty in the analysis.
The =axi=uc cycle 1 burn-up is 19,422 mwd /=tU.
Using this burnup, a tod bow penalty of 9.1% is calcu-lated.
A thermal cargin credit equivalent to 1% D GR to of fset the red bow penalty has been used as a result of the flow area (pitch) reduction factor included in all ther:21 hydraulic analysis.
Applying the 1% credit against the 9.1% calculated penalty results in an 8.17. penalty to be applied to the analysis.
Therefore, the present flux / flow setpoint provides core than ade-quate rod bow penalty coverage for cycle 1 operation.
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Table 2-1.
Ther:al-Hydraulic Desien Conditions Densif'n Revised TMI-2 FSAR Reoort Cycle 1 Design power level, Mh't 2772 2772 2772 System pressure, psia 2200 2200 2200
~
RC flow, gpa 369,600 369,600 377,000 Vessel inlet coolant te=perature, 100% power, F 557 557 557.2 Reference design radial-local power peaking facter 1.783 1.783 1.783 Reference design axial flux shape 1.5 cos 1.5 cos 1.5 cos Hot channel factors Enthalpy rise 1.011 1.011 1.011 Heat flux 1.014 1.014 1.014 Flow area 0.98 0.98 0.98 Active fuel length, in.
144.0 141.7 141.7 Average heat flux, 100% power, Btu /h-ft2 a)
)
b) 185,000 188,000 188,000 CHF correlation
'4-3 EAW-2 BAW-2 Mininua DNBR, 1127; power 1.39 1.62 1.65 (a) Based on the active fuel length arid cold fuel pin diaceter.
(b) Based on the densified active fuel length and het fuel pin dia=eter.
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These analysis results deconstrate that the renoval of forty ORAs froa TMI-2,
': hen combined '. tith the increased reactor coolant systen fic.i rate, result in improved core safety targins relative to those defined in references 1 and 2.
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3.
TRANSIRST ANALYSIS The DNER related transients presented in reference 2 have been reviewed for applicability to operatien with the ORAs re=oved.
The four pump acastdown is the loss-of-coolant-ficw (LCCF) transient analyzed in the Densif cation Report.
The minitu: DN3R during this transient was 1.65 (BAF-2).
The initial condi-tions for these transients are at 102% power.
Re-analysis at 102% power with CRAs recoved shows an increase of 1% in the initial DU3R.
The higher initial minitu DN3R =akes the res its of the transients analyzed for the Densification Report applicable and c::servative for the revised cycle 1.
All loss-of-coolant flow ::2nsients, with the exception of the loss of one p ur; from four pu=p operation, will result in a reactor trip initiated by the pu p etnitors.
The cos t limiting LCCF transient for which the pump =cnitors provide DNER protection is the four pu=p coastdown which has been shown to be acceptable.
The one pump coastdown fro four purp operation is the most limiting flow transient by virtue of its use in determining the flux / flow trip setpoint.
The flux / flew rip is based on preventing the minicu= DNBR f roc going below the design value plus the rod bow penalty.
Therefore, a one pucp coastdown with the resulting flux / flow reactor trip will result in the most limiting DNBR during nor:21 operation.
The TMI-2 FSARI has been reviewed for the = cst limiting DN3R transients of acderate frequency since the one pu=p coastdcwn does nor appear directly as an accident.
The cost limiting FSAR transient is the e::cessive heat removal accident (feedwater terperature decrease).
This transient has been re-analyzed for revised cycle 1 operation with the same input as uscd in the FSAR.
The results of the re-analysis are shown on Figure 3-1.
The minicu DNER is 1.5S (EAa-2) versus a 1.43 (W-3) reported in the FSAR.
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Figure 4-1.
TMI-2 Initial Core Leading Plan TUEL TRANSFER CANAL l
LT UX UC RL U3 FS UA U1 U4 l 3
Cl39 3153 C124 3153 C132 3M 3 Lca.;
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L C137 3162 A022 3192 C150 3156 C142 3150 C149 3199 A019 3165 0134 LOSI LO52 L353 Lc:a
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Fuel !.a se-bly ID (PS-ED: Cat ch 1; US-T6 5 5-r;:3n ch 2; T7-UY ;3a.ch 3)
I Cen::ci C apanent ID (CXXX - CFA; A:CG - APS 21., E:GX - 3Pc; OW.UO.0G;
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LCXX -RE!A!NER) g MTL NJ00 precedes all f uct asse=51y ID's.
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4.
CORE LOADING PLAN Figure 4-1 shows the revised core loading plan for the re=ainder of cycle 1.
All fuel asse=blies are re=aining in their original core locations, i.e.,
no fuel shuffle will take place.
The changes occurring are:
1.
Retainers vill be installed on all BPRAs.
2.
Thirty-eight CFJ_s will be re=cved.
3.
Two ORAs will be ecdified and installed in the pricary neutron source locations (3-12 and ?-4).
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9 5.
PROPOSED MCDIFICATIONS TO TEC'dNICAI. SPECIFICATI0:iS The Technical Specifications have been revised for the retainder of cycle 1 operation.
Changes were the result of the following:
1.
The pressure-tenperature limits have been revised to incorporate the affects of ORA renoval, retainer installation, and rod bow penalty.
2.
Syste= flow of 102% of design flow was used.
3.
The low pressure setpcint has been raised to account for the LCCA s=all break analysis (bacic; function only).
4.
Instrument drift nuriers have been included for calibration drif t in accordance with iten 2.C. (3)f. of the opera ting license.
Figures 2.1-1, 2.1-2, 2. 2-1, 2. 2-2, and 2.1 (Tech Spec nu= bering) illus trate the revisions to previous Technical Specification limits.
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Figure 2.1-1.
TMI Unit 2 Reactor Core safety Li=it 2'00 4
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RCS Pressure - High Trip P = 2355 Psig.
T=619 F 2300 O
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Temperature-High Trip o
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2200 2-ACCEPTABLE P = 2l00 Psig T
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1500j RCS Pres:;ure - Low Trip 8
P = 1900 Psig f
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Figure 2.1-2.
Reactor Core Safety Li=its THER'lAL P;4ER,s 123 CN3R LlWIT
(-33.6,112) "
- (24.5,112)
-- 113 gy/pi Lluti A;;EPTAELE
(-49. 2,1 C3) 4 pi;3p
-- 113 (43.3,103)
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(-33.S,24.5)
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(43.3,73.1) 3 g 4 pggp CPERATICN
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63 Atlal Perer le:atance, 5 C'JRVE REACTOR C0:LANT FLCf (0?M) 1 377,C;3 2
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Figure 2.2-1.
Trip Se: point for Nuclear Overpower Based on RCS Flew and Axial Power Imbalance 5 0F RATED THERMAL POWER 110 (105.0)(15 8 105)
(-23.7,105) 1 100 l
l l ACCEPTABLE
(-37.1, S S. 5) l (29.8,S3.5) l4 PUMP 90 4
l0PERATION l
1 80 (73.1) 1 I
e I ACCEPTABLE 70
( - 37.1, S S.'S )
1 3 & 4 peup (29.5,66.6)
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I ACCEPTABLE 40 (29.8,33.4) f
(-37.1,39.4) i l 2,3, & 4 l PUMP 30 IGPERATION 3
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Allowable Value for Eclear overpower Based on
. t RCS Flow and Axial Power It'calance 5 0F RATED THER.',tAL P0ifER
- 110 (105.125) l I
i l ACCEPTABLE 100 g
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'(93.54)
IOPERATICN S0 l
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ACCEPiASLE l
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- Figure 2.1.
D1I Unit 2 Pressure /Tenperature Li=its at Maximum Allowable Power for Minicun D:33R 2400
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1600 580 600 620 6'40 660 Reacter Outlet Tem;araturo, F
_ CURVE
_ G PM POWER
_ PUMP 3 0?ERATING (TYPE OF LIMIT) l 377,000(100$)
112%
FOURPUMPS(DN8R) 2 280,400 (7'.4%)
84.6%
THPEEPUMPS(DN5R) 4 3
182,800 ('48.5%)
57.4%
.ONE PUMP PER LOOP (QUALITY)
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- a - m REFERENCES I
Three Mile Island Nuclear Station, Unit 2 -- Final Saf ety Analysis Report.
Docket No. 50-320.
2 Three Mile Island, Unit 2 Fuel Densification Report, BAW-1455, Eabcock &
Wilcox, Lynchburg, Virginia, July 1977.
3 EPRA Retainer Desi n Report, 3AW-1496, Babcock & Wilcox, Lynchburg, Virginia, F
- ay, 1978.
NUF.IG-0432, ihree Mile Island Nuclear Station Unit 2 Technical Specification, Ap,endix A to License No. DPR-73, February 3, 1978.
A-1 Babcock & Wilcox t; rr q 1 c.i
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