ML19276H108
| ML19276H108 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/30/1978 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| References | |
| BAW-1473, NUDOCS 7910100504 | |
| Download: ML19276H108 (39) | |
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BAW-1473' November 1977 Revised (June 1973)
T.4REE MILE ISIXiD UNIT 1 CYCLE 4 RELOAD REPORT 9
1415 121 Babcock & Wilcox
BAW-1473 November 1978 Revised (June 1978)
THREE MILE ISLAND UNIT 1 CYCLE 4 RELOAD REPORT BABCOCK & WILCOX Power Generation Group jfjr j99 Nuclear Power Generation Division.
I7 13 iLL P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox
o
- l CONTENTS Page
~..
1.
INTRODUCTION AND
SUMMARY
1-1 2.
C?ERATING HISTORY 2-1 3.
GENERAL DESCRIPTION 3-1 4.
FUEL SYSTEM DESIGN.
4-1 4.1.
Fuel Assenbly Mechanical Design 4-1 4.2.
Fuel Rod Design 4-1 4.2.1.
Cladding Collapse 4-1 4.2.2.
Cladding Stress 4-1 4.2.3.
Cladding Strain 4-1 4.3.
Thernal Design...
4-2 4.4.
Ma:erial Design 4-2 4.5.
Operating Experience.
4-2 5.
NUCLEAR DESIGN........
5-1 5.1.
Physics Characteristics 5-1 5.2.
Analytical Input.
5-2 5.3.
Changes in Nuclear Design 5-2 6.
THERMAL-HYDRAULIC DESIGN.
6-1 7.
ACCIDENT AND TRANSIENT' ANALYSIS 7-1 7.1.
General Safety Analysis 7-1 7.2.
Accident Evaluation 7-1 8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS.
8-1 9.
STARTUP PROGRAM -- PHYSICS TESTING 9-1 10.
REFERENCES.
10-1 1415 123 Babcock & Wilcox
- 111 -
Revised June 1978 List of Tables Table Page h-l.
Fuel Design Pare =eters and Dimensions h-3 h-2.
Fuel The: al Analysis Para =eters.
hh 5-1 DE-1, Cycle 4 Physics Pars =eters 5h 5-2.
Shutdown Margin Calculation for CE-1, Cycle h........
5-6
~'
6-1.
Ther-al-Hydraulic Design Conditiens 6-2 7-1.
Cc=parisen of ey Fars=eters for Accident Analysis.
7k 7-2.
Sounding Values for A11cvable LCCA Peak Linear Heat Rates 7h List of Fi.rres 13 71i;re 3-1.
Ccre Leading Diagrs= f:: OE-1, Cycle a............
3-3 3-2.
5 '"- ent and Burnu; Distribution for DC-1, Cye'.e k..... 3h 3-3 Cencrol Rod Locatiens 'er OE-1, Cycle h 3-5 5-1.
3CC (k ara), Cycle i.sc-Ci=ensicnal Relative Fever.
Distribution - Full ?cver, Iquilibrius Xenen, APSRs Inserted 5-7 8-1.
Protectica Syste= Maxi =u= Allevable Setpoints..
8-2' 3-2.
Cere P otectica Safety Limits.
8-3 8-3 Protection Syste= Maximus Allevable Setpoints for Reactor
?cver I= balance.
8h 8 k.
Red Positica Limits for Four-Punp Operation Frc= 12515 EFFD to 265 1 15 EFPD - EU-1, Cycle b............... 8-5 8-5 Red Position Limits for Two-and Three-Pu=p Operation Frc= 125 1 5 te 265 1 15 IFFD - SE-1, Cycle h............. 8-6 8-6.
- Power I= balance Envelope for Operatica Frc= 125 + 5 to 265 +
15 EFFD.
8-7 8-7 A?SR Fosition Limits for Operation Frc= 0 to 265115 IFPD 8-8 1415 124
- iv -
Sabecck i 'Jilecx
1.
INTRODUCTION AND
SUMMARY
This report justifies the operation of the Three Mile Island nuclear Station Unit 1'(TMI-1) (cycle 4) at a rated core pcver of 2535 !Gt. Included are the required analyses, as cutlined in the USNRC docu=ent, " Guidance for Propcsed.
License A=end=ents Relating to Refueling," June 1975 To supper cycle h cperation of the DC-1, this report e= ploys analytical techniq2es and design bases established in reperts that have been submitted and received technical approval by the USNRC (see references).
The design for cycle h is based on operation in the feed-and-bleed er rods-cut
=ede.
All nuclear para =eters pertinent to accident analyses have been calculated consistent vith this =cde of operatien. Section 5.3 descrites the change to feed-and-bleed operation.
Cycle h reactor para =eters that are related to pcver capability are su==arized in this report and referenced to cycle 3.
All the accidents analyzed in the FSAR have been revieved for cycle h operation, and in casec where cycle h characteristics proved to be censervative with respect to those analyzed previously, norav analysis was performed.
The Technical Specifications have been reviewed, and the =edifications required for cycle h cperation are justified in this report. Based on the analyses per-fomed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for I=ergency Core Cccling Systems (ECCS), it has been concluded that TMI-1, cycle h can be safely cperated at the rated core pcver I415 125 level of 2535 :Ut.
1-1 3abecek & Wilcox
Revised June 1978 2.
OPERATING HISTORY
..~
The reference cycle for the nuclear and ther=al-hydraulic analyses of the Three Mile Island, Unit 1 is the operating cycle 3 Cycle 3 achieved criti-cality en May 13, 1977, e ' a'ter cero power testing attained loof, power on May 20, 1977 No centrol r:d interchange was nade during cycle 3, which was ec pleted en March 18, 1973, after 287 IFPD. No operating anc=alies occurred during the fourth cycle. 2e operation of cycle ' began April 29, 1978.
The design cycle length is 265115 :tro, and no control red interchanges are plar.ned.
1415 126 2-1 3abecek & Wilecx
3.
GENERAL DESCRIPTION The TMI-l leactor core is described in detail in Section 3 of the Final Safety Analysis Report for that Unit.1 The cycle k core consists of 177 fuel asse:-
blies (FAs), each of which is a 15-by-15 array containing 208 fuel reds,16 centrol red guide tubes, ard one incere instru=ent guide tube. The undensified i
nc=inal active lengths of the fael rods are lhh inches for batches le and 2b, lh2.6 inches for batch h, and ih2.25 inches for batches 5 and 6.
All fael as-senblies in cycle h =Mntain a constant nc=inal fael loading of h63.6 kg of ure=iu=.
The cladding is cold-verked, Zircalcy h with an CD of 0.h30 inch and a vall thickness of 0.0265 inch. The Pael consists of dished-end, cylindrical pel' ets of uraniu= dicxide (see Table h-2 for data).
Figare 3-1 is the core leading diagra= for TM1-1, cycle h.
The initial enrich-
=ents of batches ic, 2b, ard h vere 2.06, 2.75, and 2.6h vt 5 uraniu=-235, re-spectively. Batches 5 and 6 have a 2.85 vt 5 uraniu=-235 enrich =ent.
All the batch 3 asse=blies vill te discharged at the end of cycle 3, and the batch h and 5 assemblies vill be shuffled to new locatiens. The batch 6 asse=blies vill occupy *he periphery of the cere. The 13 batch le and the eight batch 2b asse=blies vill occupy interier core locatiens. Note that the designations le and 2b are used to identify asse=blies frc= the original batch 1 and batch 2 fuel. The le end 2b asse=blies vere re=oved frc= the cere at the end of cycle 1 and cycle 2, respectively (see Table h-1). They are being reinserted into the cycle h core to lever feed batch size require =ents and spent fuel storage, thereby producing a more efficient fuel cycle. It should be noted that the asse=blies referred to as batch la in the cycle 3 Relcad Report (reference 2) are new designated as 1b; batch la is now the re=ainder of batch 1 asse=blies which have not been scheduled for reinsertion. Figare 3-2 is an eighth-core =ap sheving each asse=bly's burnup c5sL2Zx 3-1 3
at the beginning of cycle h and its initial enrichnent.
Cycle h vill be operated in a reds-out, feed-and-bleed nede. The core reac-tivity centrol vill be supplied nainly by soluble beren and supplenented by 61 full-length, As-In-Cd control rod assenblies (CRAs). In addition to the full-length centrol rods, eight axial power shaping rods ( ApSRs) are provided for additional control of axial power distribution. The cycle h locatiens of the 69 centrol reds and the group designations are indicated in Figure 3-3 The core locaciens of the 69 control reds for cycle h are identical to these cf the reference cycle 3 Zevever, the group designations differ between cycle h and the reference cycle c nininice power peaking. No centrol rod interchanges er burnable poison rods are necessary for cycle h.
The nc=inal systes pressure is 2200 psia, and the ecre average densified nc=inal linear heat rate is 5.72 kW/ f: at the rated core ;cver of 2535 rit. The heat rate is slightly higher than in the reference cycle 3 (5.71 kW/ft) due to the shorter stack height of batch 6 relative to the discharged batch 3.
}k
)
3-2 3abecek & Wilecx
Revised June 1978 Figure 3-1.
Core Loading Diagram for TMI-l Cycle 4 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 i
A 6
6 6
6 6
B C7 Ell C9
{
6 6
6 4
4 4
6 6
6
^^
y DI 37 (yly) 39 G12 B4 B12 C
( 2)
(Cy1) y 4
6 4
23 5
5 1C 2B 4
6 4
Gl.
35 A6 D6 A8 D10 A10 Bil D9 0
5 6
L 6
5 4
5 4
5 5
4 6
6 G2 II D5 B6 310 D11 l,cyl) g,1{)
gik)
E14 G14 g
6 4
5 1C 4
5 1C 5
4 1C 5
4 6
l y32) lI4 II C3 CS A9 E8 Cl3 E12 FIS (ff j) l 6
6 23 5
4 5
4 5
4 5
4 5
23 6
6 G3 D2 Ti l 72 H5 A7 G8 G15 H13 F14 F12 D14 G13 6
4 5
l5 4
5 4
5 4
5 4
5 4
6 E3 L"
13
'l H7 F4 J
H9 Kl' fu3 HIS L12 31
$C b[ '
~
7 }
ICl}
5
[C 6
4 5
5 4
1C 4
5 4
6
~
K3 ' N2 La H3 K1 K8 R9 Hll Ll4 L12
' N14 K13 l12 6
4 5
4 5
4 5
4 5
4 5
4 5
4 6
3 L1 M4 03 MS R7 08 013 M12 L15 p'f 2) gy j) ll g
6 6
28 5
4 5
4 5
4 5
4 5
2B 6
6 K2 M2 N5 P6 05 P10 Nil flr M14 K14 (Cy 1)
(Cy1)
'Cy 1) 6 4
5 4
5 5
4 5
4 6
Ic IC IC N
N7 P5 R6 N6 R8 N10 R10 Pil K12 6
6 4
5 5
4 5
4 5
5 4
6 6
0 K4
(
}
4 6
4 5
4 6
4 C
B p
/
07 115 c9
\\
6 6
6 4
4 4
6 6
6 R
6 6
6 6
6 I
xxx Pr..u. Cyc1. txxue (exc.Pe.. ex.c J
pgg $$
i413 129 3-3 Babcock & Wilcox
Revised June 1978 Figure 3-2.
Enrichment and Eurnup Distribution for TMI-1, Cycle 4 8
9 10 11 12 13 14 15 2.06 2.64 2.85 2.06 2.85 2.06 2.64 2.85 14,263 16,658 8,177 13,600 7,612 14,263 16,681 0
2.85 2.64 2.85 2.64 2.85 2.64 2.85 K
~~ -
3,167 18,S81 11,377 15,454 5,557 20,258 0
2.35 2.64 2.85 2.75 2.85 2.85 h
5,198 15,360 7,123 23,C 9 0
0 2.06 2.85 2.64 2.85 M
11,718 8,839 17,606 0
2.64 2.85 2.85 16,285 0
0 2.64 0
16,288 P
R x.xx Initial Enrichment xx,xxx BOC Surnup, >St 14i5 130 3-4 Babccck & W'ilcox
Figure 3-3.
Control Red Locations for DtI-1, Cycle 4 X
l A
8 3
7 3
C 1
6 6
1 D
7 8
5 8
7 E
1 5
,2 2
5 1
F 3
8 7
5 7
8 3
G 6
2 4
4 2
6
~
J 7
5 5
3 5
5 7
X 5
2 4
4 2
6 L
3 S'
7 5
7 8
3 1
1 5
2 2
5 1
8 5
8 7
N 7
0 1
6 6
1 P
3 7
3 R
I Z
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 Grouc No. of rods Function 1
8 Safety 2
8 Safety X
Group Numoer 3
9 Safety 4
4 Safety 5
12 Contr.ol 6
8 Control 7
12 Control 8
8 APSRs total # 69 Babcock & Wilcox 3-3 1415 131
4.
FUEL SYSTEM DESIGN 4.1.
Fuel Asse=bly Mechanical Desi2n
~ ~ -
The types of fuel asse=blies af.d pertinent fuel design parameters for TM1-1, cycle 4 are listed in Table 4-1.
All fuel asse=blies are identical in concept and are =echanically interchangeable. All results, references, and identified conservatis=s presented in section 4.1 of reference 2 are applicable to the cycle 4 reload core.
4.2.
Fuel Rod Desien 1:'e =echanical evaluation of the fuel red is discussed below.
4.2.1.
Claddine Collapse Creep collapse analyses were perfor=ed for three cycle assembly power histories.
te batch 2b reinserted fuel is = ore limiting than the other batche's due to lower prepressurization, lower pellet density, e;Ler lenger previous incore exposure f
time.
The batch 2ni asse=bly power histories were analyzed, and the =ost limiting assembly was determined.
The pcwer history for the cost limiting assembly was used to calculate the fast neutron flux level for the energy range above 1 MeV. The collapse time for the most limiting assembly was conservatively determined to be more than 30,000 EFFH (effective full power hours), which is greater than the maximum projected resi-dence ti=e- (Table 4-1).
The creep collapse analysis was perfor=ed based on the conditions set forth in references 2 and'3.
4.2.2.
Cladding Stress The batch Ic and 2b reinserted fuel is the most limiting for cladding stress.
The results presented in reference 4 are applicable.
4.2.3.
Cladding Strain The fuel design criteria specify a limit of 1.07. on cladding plastic circumfer-stial strain. The pellet design is established for a plastic cladding strain 1415 132 Babcock & Wilcox 4-1
of less than 1,5 at values of =exi=u= design local pellet burnup a.nd heat gen-eration rate, which are censiderably higher than the values that the TMI-l fuel is expected to see.
his vill result in an even greater =argin than the analysis de=enstrated. The strain analysis is also based en the maxi =u= allow-able value for the fuel pellet dia=eter and density and the lovest per=itted tolerance for the cladding ID.
h.3.
Ther=al resien The incoming batch 6 mel is ther-Dy and gec=etrically si=ilar to the batch 5 fuel of cycle 3.
The TAFY5 rael pin analysis perfer=ed for tatch 5 Pael also applies to batch 6.
An analysis was also perfor=ed for batch 6 using the fuel 6
perfor=ance code, TA00. Where differences cce= red between correspending calculated values of each ecde, the = ore conservative values vere chosen for batch 6 design.
"~ne=nal analysis of the fael reds assu=ed in-reactor densificatica to 96.55 TD.
The average fuel te=pe stres (Table 4-2) for batches 1 through 5 are taken frc= the TAFY analyses -/..ich define the linear heat rate (LE) capability for each batch. The average te=;eratre shown fer batch 6 vas taken frc= an average pin analysis using the TACO cede. The value shown represents the 30L (100 mwd /
=tU) average fuel te=peratr e at 5.80 kW/ft. The average te=peratre decreases with burnup to a value of 1120F at 38,000 mwd /=tU.
Linear heat rate capabilities are based on centerline fuel =elt.
Batch 6 linear heat rate capability was deter =ined based en the lover tolerance li=it (LTL) of the fuel density specification. The design LER capability used for batch 6 was 20.15 kW/ft, which was calculated by the TAFY code and is the same as that for batches k and 5 The TACO analysis for batch 6 fael gives a higher L'IR capability.
Therefore, the = ore conservative TAFY LE capability was used in the 2 sign of batch 6.
h.k.
Material Desien The chemical compatibility of all possible fuel-cladding-coolant asse=bly in-teracticu for the batch 6 fuel asse=blies is identical to that of the present Pael.
h.5 Oceratin Exterience 3&W's operating experience with Mark 3,15-by-15 fuel asse='.;17 design has veri-fled the adequacy of the fuel asse=bly design. As of August 31, 1977, the fol-loving operating experience has been ecliected for the seven 3&*41TT-FA plants using the Mark 3 Pael asse=bly:
1415 133 h-2 3abecek & 'Jilecx
Revised June 1978 lowing operating experience has been collected for the seven B&W 177-FA plants using the Mark 3 fuel assembly:
Current cycle Cumdlative Current max asse=bly net electrical Reactor cycle burnuo, Wd/etU output, Wh Oconee 1 3
26,300 18,134,699 Oconee 2 3
23,400 13,475,779 Oconee 3 2
25,700 13,907,914 TMI Unit 1 3
26,700 15,259,750 ANO Unit 1 2,
24,294 12,044,505 Rancho Seco 1
19,664 8,328,383 Crystal River 3 1
4,400 1,881,750 Table 4-1.
Fuel Desien Para eters and Dimensions res ce Twice-burned assys Once-burned assys Batch 2b Batch 4 Batch ic Batch 5 Batch 6 Fuel assembly type Mark 33 h rk 34' hrk 32 hrk 34 h rk 34 No. of assemblies 8
56 13 48 52 Fuel red CD, in.
0.430 0.430 0.430 0.430 0.430 Fuel rod ID, in.
0.377 0.377 0.377 0.377 0.377 Flexible spacers, type Corr'd Spring Corr'd Spring Spring Rigid spacers, type Zr0 Zr-4 Zr0
=#~4 U#-'
2 2
Undensified active fuel length, in.
144.0 142.6 144.0 142.25 142.25 Fuel pellet OD (mean specified), in.
0.3700 0.3700 0.3700 0.?695 0.3695 Fuel pellet initial
- density, TD 92.5 93.5 92.5 94.0 94.0 Initial fuel enrich =ent vt : 235U 2.75 2.64 2.06 2.35 2.85 Burnup (BOC) Wd/stU 23,049 17,216 13,276 7,309 0
~
l Cladding collapse time, EFPH
>30,000
>30,000
>30,000
>30,000
>30,000 Residence Time, EFFH 23,976 19,682 17,832 20,258 20,088 l
Babc]c,< ; $" ilcaxT34 4-3 6.
4
Table 4-2.
Fuel Thermal Analysis Parameters Densified fuel carameters(*} Batch ic Batch 25 Batch 4 Batch 5 Batch 6 Tellus diadeter, tu.
0.3640 0.3640 0.3645 0.3646 0.3646 Fuel stack height, in.
141.12 141.12 140.46 140.47 140.47 Nominal LHR at 2568 EJt, kW/ft 5.77 5.77 5.80 5.80 5.80 -
Avg fuel temp at nominal LHR,
- F (BOL) 1335 1335 1320 1315 1400 LERtog,fuci=elt, kW/ft 19.6 19.6 20.15 20.15 20.15 i.
(*)Densification to 96.57. TD assumed.
1415 135
~
Babcock & 'Niicox 4_4
5 NUCLEAR DESIGN 5.1.
Physics Characteristics Table 5-1 cc= pares the core physics para =eters of cycles 3 and h; these values vere generated using pDQ07 for both cycles. Since the core has not yet reached an equilibriu cycle, differences in core physics para eters are to be expected between the eveles.
f The lenger design life cf cycle h vill produce a corresponding larger cycle differential burnup than designed for the reference cycle 3.
The icver accu =ulated average core burnap at the end of cycle k is mainly due to the discharge at the end of cycle 3 of batch 3 fuel which had a high burnup history. Figure 5-1 illustrates a representative relative pcuer distributica for the beginning of the cycle h at
^211 pover with equilibriu: xenen and group 8 inserted.
The critical boren concentrations are approxi=ately the sa=e as those of reference cycle 3 The hot, full-pever control red worths are similar in both cycles except for grcup 7, which is significantly higher in cycle k, being ec=pesed of 12 control red asse=blies rather than eight as in cycle 3.
Control red verths are sufficient to =aintain the required shutdown =argin as indicated in Table 5-2.
The differences in the parsneters between cycles 3 and h are due to changes in radial flux distribution, isotopics, and the difference in cycle lengths. The ejected rod verth in Table 5-1 are the maxi =us calculated values. It is difficult to ec= pare values between cycles or between red patterns since neither the rod patterns frc= which the CRA is ejected nor the isotopic distributions are identical.
Calculated ejected rod vorths and their adherence to criteria are censidered at all tines in life and at all pcver lev-1s in the develep=ent of the red position li=its presented in section B.
The =axi=== stuck red verth at the end of cycle h is si=ilar to that for the reference cycle 3 but is lever at t.
beginning of the cycle.
1415 136 5-1 3abcock & Wilecx
s The following conservatic=s were applied for the shutdevn calculations:
1.
Poison =aterial depletion allowance.
2.
105 uncert'ainty en net red vorth.
3 Flux redistributien penalty.
Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional =cdel.
The shutdown ce.lculation was analyzed at 277 EFPD. The =aximum allovable inserted rod verth is s= aller in cycle h than in cycle 3 due to the operatien of this cycle in a feed-and-bleed =ede in which control rod group 7 is only partially inserted in the core during the entire cycle.
The cycle h power deficits frc= hot zero power to hot full power are lover than those for cycle 3 due to the less negative =oderator coefficients in cycle k.
The differential beren and xenon vorths are similar fer cycles 3 and 4 The effective delayed neutron fractions for cycle 4 show a decrease with burnup (also repe ed in the reference cycle 3).
5.2.
Anal.-tical Input The cycle L incere =easu e=ent calculation constants used for ce=puting core power distributions were prepared in the se=e =anner as for the reference cycle.
5.3 Chances in !iuclear Desien Cycle h is designed to operate in a feed-and-bleed = ode in contrast to the redded cperatien of cycles 1, 2, and 3 The =ajor difference in operational
= odes during equilibrius, steady-state conditions is that no full-length cen-trol rods are inserted into the core.
(A srm, bite insertica, approxi=ately 105, of cne regulating bank is =aintained to allow discrete changes in soluble boron and to acc0==cdate s=all te=perature and lead de=and changes. ) During lead follow operation the regulating bank is inserted into the core only to offset power Doppler reactivity changes. Transient xenen reactivity effects are ec=pensated by char.ging the soluble boren concentration.
The se=e calculational =2thods end design infor=atien used in reference cycle 3 vere used to obtain the i=portant :1uclear design para =eters in cycle k.
Additicnal calculations were perfor=ed for soluble boren control, shutdevn, reactivity control, and radiation analyses due to the =cdification in the =cde of operation. As in cycle 3, both A? SEA and C3A positien li=its, as well as power i= balance li=its, p
1415 137 5-2 3abecek & Wilcon
vill be specified based on LOCA analyses. These operational 14-4ts and the RPS limits (Technical Specification changes) for cycle k are presented in section 8.
i 4
1415 138 33 3atecek & '411ccx
Revised June 1973 Table 5-1.
TMI-1. Cvele 4 Physics Parameters Cvele 3(
Cvele 4 5)
Cycle length, EFPD 270 277 Cycle burnup, Mk'd/mtU 8341 8557 Average core burnup - EOC,INd/=tU 18,352 18,165 Initial core loading, stU 82.1 82.1 Critical boron - 30C, ppm (no Xe)
HZP(c), group 8 (37.5% ud) 1317 1226 HZP, groups 7 and 3 inserted 1155 1093 HFP, group 3 inserted 998 1045 Critical boro'n - EZ, ppe (eq Xe)
RZP 380 285 group 8 u...s.:.. we, eq Xe) p.g 84 16 Control rod worths - :_P, 300, % ak/k Group 6 1.18 1.12 Group 7 0.84 1.48 Croup 8 (37.5% vd) 0.54 0.46 Con:rol rod worths - ?_ P, ECC,(d)
- ak/k Group 7 1.11 1.57 Group 8 (37.5% vd) 0.50 0.50 Max ej ected rod worth
'd2?, % Ak/k(*
30C 0.34 0.81 EOC 0.77 0.81 Max stuck rod worth - HZP, % ak/k 30C 2.42 2.01 EOC 2.06 2.06 Power deficit, HZP to HFP, % ak/k 30C
-1.58
-1.29 EOC
-2.15
-2.05 Doppler coeff - 30C, 10-5 (ak/k/*F) 100% power (0 Xe)
-1.47
-1.43 Doppler coeff - EOC, 10-5 (ak/k/*F) 100% power (eq Xe)
-1.51
-1.60 Moderator coeff - HFP, 10 (ak/k/*F) 30C (0 Xe, 1045 ppm, group 8 ins)
-0.91
-0.71 EOC (eq Xe, 17 ppm, group 8 ins)
-2.54
-2.53 Boron worth - HFP, ppe:/% ak/k SOC (1150 ppm) 107 104 EOC (17 ppm) 97 94 Xenon worth - HFP, % ak/k 30C (4 EFFD) 2.59 9 65 E0C (equilibrium) 2.641415 139 5 -!.
Babcock & Wilcox
Revised June 1978
~
Table 5-1.
(Cont'd)
Cycle 3(*
Cvele 4( )
t Effective delayed neutron
~
f raction -- HFP BOC
'O.00584 0.00583 EOC 0.00524 0.00520
(* Cycle 4 data are for the conditions stated in this report.
The cycle 3 core conditions are identified in reference 2.
f (b) Based on 253 IF?D at 2535 MWe, cycle 2.
(c)E2P denotes het :ero power (532F T""8); HFP dunotes hotfull power (579F I""5).
(d)246 EFFD in cycle 3; 277 EFPD in cycle 4.
(e) Ejected rod worth for groups 5 through 8 inserted.
(f) Based on 287 EF?D at 2535 MWe, cycle 3.
s
}k h
Babcock & Yliicox 5-5
Revised June 1973 Table 5-2.
Shutdown Margin Calculation for TMI-1. Cvele 4 BOC, % Ak/k EOC.("}% Sk/k Available Rod Worth C)
Total rod worth, HZP 8.62 8.78 Worth reduction due to burnup o r poison material
-0.37
-0.46 Maxi =u::: stuck rod, HZ?
-2.02
-2.06 Ne: vorth 6.23 6.26 less 10% uncertain:y
-0.62
-0.63 Total available v:::h 5.61 5.63 Reeuired-Red Worth
?:ver deficit, HF? :: EZ?
1.29 2.05 Max allovable inser:ed rod worth 0.34 0.41 Flut redistributic_
0.53 0.98
!c:a1 required vor:'n 2.16 3.44 Shu:de i Marzin To:al available minus cctal required 3.45 2.19 Note: Required shutdown margin is 1.00% ak/k.
(*}277 EFFD.
(b)HZP denotes hot zero power (532F T**E); HFP denotes hot full power (579F Tavg).
1415 141 Babcock & Wilecx 5-6
Revised June 1973 Figure 5-1.
BOC (4 EFPD), Cycle 4 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, APSRs Inserted
- 8 9
10 11 12 13 la '
15 H
0.94 1.05 1.23 0.97 1.21 0.91 0.83 0.79
~..
K 1.24 1.cc 1.16 1.05 1.18 0.85 0.7.9 f,
8l L
1.32 1.05 1.03 0.93 1.18 0.69 M
0.99 1.16 0.95 0.99 N
1.03 1.13 0.70 0
0.55 P
R s
X Inserted Rod Group Number X.XX Relative Power Density
- Calculated results from 2D pin-by-pin PDQ07 1415 l42 5-7 Babcock & Wilcox
Revised June 1978 6.
THERMAI.-HYDRAULIC DESIGN
~~
The inconing batch 6 fuel is hydraulically and geometrically similar to batch 5 fuel. The only difference between cycles 3 and 4 is the core configuration.
f' 2
The cycle 2 DN3R analysis was used for reference cycle 3 ; this analysis is also valid for cycle i as discussed below. The core configuration used for cycle 2 analysis consis:ed of 60 Mark 33 assemblies and 117 Mark 34 assemblies with the cost limiting (h:t) assembly being a 33 assembly. The cycle 4 con--
figuration consists of 13 :: ark 32, 8 Mark B3, and 156 Mark B4 assemblies with the ::sc li=1 ting assn:b'/ being a 34.
Both the Mark 32 and B3 assemblies have a hi her resi;tance to flew than the Mark B4 assembly.
3 The =initu: DN3R calculated fr:: the cycle 2 analysis was compared to the min-inus DN3R obtained fro: ananalysisofanall-$4 core. The cycle 2 analysis p:ovided the more restrictive =ini=um DN3R. For cycle 4 the addition of the higher resistance Mark 32 and 33 assemblies provides additional DNBR margin.
The higher resistance hrk 32 and 33 assemblies will tend to increase flow through the limiting Mark 34 assembly. Therefore, the cycle 2 DN3R analysis bounds that for cycle 4.
This conservatism is amplified when peaking factors are considered. The mini-mum DN3R calculated from the cycle 2 analysis is based on a radial-local peak of 1.733.
The maximum radial-local peak calculated for cycle 4 operation, including 8% nuclear uncertainty, is 1.547 at 30L.* This decreases to 1.403 at the end of cycle 4.
This provides a 13.2% margin to the design peak at EOC and a 21.3% margin at E0C. This margin ensures that the design conditions shown in Table 6-1 will not be exceeded during cycle 4 operation.
All ochar thermal hydraulic analyses that uere applicable to cycle 3 remain applicable to cycle 4.
1415 143
- Calculated results from 2D pin cy-pin PDq07.
l 6-1 Sabcock & Wilcox
Table 6-1.
Thermal-Hydraulic Desis:n Conditions Cycle 1 Cycles 2, 3, and 4 Power level, W e 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 100.0
.106.5 Vessel inlet coolant te=perature at 554.0 555.6 100% power, F Vessel outlet coolant te=perature at 100% power, F 603.8 602.4 Ref design radial-local power peaking factor 1.78 1.78 Ref design axial flux shape 1.5 cosine 1.5 cosine Active fuel length, is.
Table 4-2 Table 4-2 Avg heat flux (100.5 power), Btu /h-f t2 171,470 174,870 2
Maxi =u= heat flux (100% pover), Beu/h-ft (for DNER calculation) 457,825
'466,903 ~
CHF correlation W-3 B&W-2 Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flov area 0.98 0.98 l'ini=us DNER (densified fuel) 1.46 1.877 (114% power)
(112% power) 1415 144 Babcock 3. V/ilecx 6-2
Revised June 1978 P00RORBINAL 7.
ACCIDENT AND TRANSIENT ANALYSIS 7.1.
General Safety Analvsis si Each FSARI accident analysis has been examined with respect to changes in cycle 4 paraceters to deter ine the effect of the cycle 4 reload and to ensure that therral perfor=ance during hypothetical transients is not degraded.
The effetts of fuel densification on the FSAR accident results have been evalu-ated and are reported in reference 4.
Since batch 6 reload fuel assemblies centain fuel rods with a theoretical density higher than those considered in reference 4, the conclusiens in that reference are still valid.
The dose evaluations in the FSA2 were based on conservative values for fuel burnup and power peaking; hcwever, i= proved fuel utilization and improved cal-culational methods have led to a higher plutonium-to-uranium fiss.on ratio.
Since plutonium has a higher iodine fission yield than uranium, more iodine activity will be produced and thus the thyroid doses will be slightly higher than reported in the FSAR.
A comparison has been made between the 2-hour
- thyroid doses associated with the I
major accidents in Chapter 14 of the FSAR with the 2-hour thyroid doses that would result from the cycle 4 iodine activity inventory. The results show that although the thyroid doses for cycle 4 increase by 6 to 19% over the l.
FSAR, the cycle 4 doses are still only a very s=all fraction of 10 CFR 100 limits.
7.2.
Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas:
core thermal parameters, thermal-hydraulic parameters, and kinetics parameters including the reactivity feedback coefficients and control rod worths.
Core thermal properties used in the FSAR accident analysis were design oper-ating values based on calculational values plus uncertainties.
Comparisons of s
e Badh & Wilcox 7_1
Revised June 1978 first con values (FSAR values) of core ther=al parameters and subsequent fuel batches to paraseters used in cycle h analyses are given in Table h-2.
A co=-
parisen of the cycle h ther=al-hydraulic mvf=us design conditiens to the pre-vious cycle values is presented in Table 6-1.
These para =eters are ec==on to all the accidents considered in this report. A ec=parison of the key kinetics paraseters frem the FSAR and cycle k is provided in Table 7-1.
The feedvater line break has been analyzed with respect to pri=ary system pres-f sure respense censidering the high-pressure reactor trip and pressurizer code safety valve setpoints, 2390 and 2500 psig, respectively. For cycle h pars =eters, the peak RC systen pressure at the pu=p discharge is below the 2750 psig criteriou. The analysis was perfor=ed using a conservative Doppler coefficient a=d using various =cderater coefficients. Acceptable RC peak syste= pressures resu'ted for all nonpositive =cderator ccefficients.
A generic LCCA analysis has been perfor:e:1 for the 3W 177-FA lowered icop !SS using the Final Acceptance Criteria ECCS evaluation model reported in reference 7 This analysis is generic in nature since the li=iting values of the key para =eters for all plants in this category were used. Further= ore, the ec=bination of the average fuel te=perature as a function of linear heat rate and the lifettre pin pressure data used in the LOCA limits analysis to l
conser"ative ec= pared to those calculated for this reload. Thus, the analysis and the LCCA limits reported in references 7 and10 provide conservative results l
for the operation of T:C-1, c'/cle h fuel. A tabulation sho. ring the beunding values fer allevable LCCA peak linear heat rates fer T:C-1, cycle h fuel are 1415 146 P00R OR M<
provided in Table 7-2.
It is concluded by ex -Matica of cycle h core ther=al properties and kiretics properties with respect to acceptable previcus cycle values that this core re-10c.d vill not adversely affect the ability to safely operate the CH-1 plant during cycle k.
Considering the previcusly accepted design basis used in the 7-2 3abecek a !Tilecx
Revised June 1978 FSAR and subsequent cycles, the transient evaluation of cycle h is considered to be bounded by previously accepted analyses. 'The initial conditions of the transients in cycle h are bounded by the FSAR and/or the fuel densification report (reference h) and/or subsequent cycle analyses.
4 1415 147 7-3 3abecch 5 'Jilecx
Revised June 1978 Table 7-1.
Comoar.' son of Kev Paraceters for Accident Analysis' FSAR and densif's Parameter report value
~ Predicted value
~
~
Doppler coeff (BOC), ak/k/*?
-1.17 x 10
-1.48 x 10
-3
-5 Doppler coeff (EOC), ak/k/*F
-1.33 x 10
-1.60 x 10 Moderator coeff (BOC), Ak/k/*F
+0.5 x 10~
-0.71 x 10
~
Moderator coeff (ECC), ak/k/*F
-3.0 x 10
-2.53 x 10 "
~
All rod group worth, (H2P) 7. Ak/k 10.0 8.62 Initial boron conc. (HTP), pps 1200 1045 3oren rasetivity worth (70*F),
75 73 pp:/1': ak/k Max ejected rod worth, (EF?), P. Ak/k 0.65 0.25 Dropped rod worth (HFP), " ak/k 0.46 0.20 Table 7-2.
Sou.cing Values for Allowable LOCA Feak Linear Heat Rates _
Allowable peak linear Core elevation, ft heat rate, kW/ft 2
15.5 4
16.6 6
18.0 8
17.0 10 16.0 1415 148 7a Babcock & Wilcox
Revised June 1978 8.
PROPOSED MCDIECATIC::S TO TECH:TICAI, SPECIFICATICRS The Technical Specifications have been revised for cycle k operation. The changes vere =ade as a result of the folleving:
1.
TMI-l has been changed frc= a rodded to a feed-bleed = ode of operation for l
cycle k.
This change is not regarded as a =ajor change in the operating mode since T!E-1 vas operated in essentially a rods-cut configuratica dur-ing the latter part of the previous cycles. hbalance linits, centrol rod position li=1ts, and A?S'. positica li=its are utiliced to control power peak h g and linea-heat rates. A power level cucoff of 925 full power is used to control pcVer peaking due to transient xenon effects.
2.
Se Technical Specifica:icn limits based on DNBR and LER criteria include l
appropriate allevances fer projected fuel red bov penalties, i.e., poten-tial reduction in 2:3R and an increase in pcver ;es2s. A statistical ec=-
bination of the nuclear ancertainty factor, engineering hot channel factor, and red bow peaking penalty was used in evaluating LER criteria, as app' roved in reference 8.
3.
Per reference 9, the power spike penalty due to fuel densificatien was not l
used in setting the DN3R-and ECCS-dependent Technical Specification li=its.
h.
3ased upon a reanalysis, it beca=e necessary to reduce the RCS High Pressu e Trip Setpoint so that in the event of a feedvater line break, the Peak RCS pressure vill not exceed 2750 psig.
I 1415 149 S-1 3abecek & 'Jilecx
Figure 8-1 hotection Systen Itd== A11cvable Setpoints 2500 P = 2290 psig 2300 ACCE? TABLE e
=-
g' l
OPERATION T = S19 F 0
l 2 00 S
E Y
N d
A8 s%
t 1530 I
c' I
s UNACCEPTABLE g
P = 1500 ::s i g OPERATION 1700 1500 540 550 580 600
'620 640 Reactor Outlet Temperature. F 1415 150 7
Babcock & Wilcox 8-2
Revised June 1978 Figure 8-2.
Core Protection Safety Limits Thermal Power Level, f.
-- 120 I
ONBR Limit (112)
(su.112)
(-M,ll2)
ACCEPTABLE
-- 110 4 PL'HP OPERATION
-- 100 Kw/Ft Kw/Ft Limit L.imet
(-30.87.1) 7 90 (sr.1) 2 (44.s7.1)
(-53,80)
ACCEPTA2L E 3 M FUMP.
-- 20 (60. s o )
CPERATION
, 70
(~
'$')
(- 53. 55.1 )
__ 60 (ss.s) 3 (45.59.s)
ACCEP iA!L E (5s.4.58.4)
(-4 2. '4. '42. 4) 2.3 & 4 PUMP
-- EO OPERATf03 (0.2.49.2) 40 30 20 10 I
i f
f I
f g
g
, 50
-40 20
-10 0
10 20 30
'0 50 60 4
Reactor Power imbalance, f, Curve Reactor Coolant FJow (Ib/hr)
I 139.8 x 106 6
2 10'4.5 x 10 S
3 68.8 x IC 1415 151 8-3 Babcock & Wilcox
Revised June 1978 Figure 3-3.
Prote:: tion System Maximum Allowable Setpoints for Keactor Power Imbalance Thermal Power Level, i
_120
( 17.108) -
110 (108) (37 ggg)
I l
l
-.100 l
M2 = -1.0 MI = 1.28 ACCEPTABLE A PUMP 0? ERAT i ct.
.- 30 (35.90)
(-35,85)
I (80.7) 6 ou l
ACCEPTABLE 70 l
3 & 4 PUMP CPEP.ATION (35,52.7)
( -3 5,5 7. 7 )
l i
(53.1)l l
-- 50 l
I l
ACCEPTABLE 40 2,3 & 4 PUMP (35*35*1)
OPEP.ATlCN i
(-35,3 0.1 )
l 30 l
1 1
-- 20 l
lo l
o o
l2
~
l 10 I ff E
\\;
E I
t I
f f
t !
I t
f f
f f
f f
-10 60
-50
-40 20
-10 0
10 20 30 40 50 50 70 -
Reactor Power Imaalance. i 1415 152
- 3. h Babcock 2. Wilcox
Revised June 1973 Figure 8 h.
Red Position Limits for Four Pamp Operation Frc= 125 1 5 IFPD To 265 115 IFFD - TMI-1, Cycle k paser
~
234.102 274.1.102 LEVEL NOT ALLOWE0 CUiOFF 100
= 92%
2'4.1.92 g
^'
RESTRICTED 80 248.2,'90 SH'JTOC#N MARGIN LIMIT 70 200.70 3
50 3
PEF.MISSISLE 50 173.50 180.50 OPERATING REGION
=
40 a'
30 f
20 122.15 140.15 10 0.2.3 0
i t
i I
t t
t t
i e
0 25 50 75 100 125 150 175 200 225 250 275 380 Ro2 Index. i U Inarawn 0
25 50 75 100 1
f f
f I
Group 7 0
25 50 75 100 I
f I
t t
0 25 50 75 100 f
f f
f f
Grouc 5 1415 153 Pabcock & Wilcox 3-5
Revised June 1978 Figure 8-5 Rod Position Limits for Two-and Three-Pu=p Operation From 125 : 5 to 265 2 15 EFPD - TMI-1, Cycle 4 234.102 248.2.102 100 NOT ALLOWED
~
80 SHUTDOWN HARGIN i
.2 70 LIMIT 4
A 60
=
2 50 173.50 a-40 OPERATING y
REGION 2
30 RESTRICTED FOR 20 2 AND 3 PUMP 122.15 140.15 10 0.2.3 0
f f
f f
f 0
25 50 75 100 125 150 175 200 225 250 275 300 Rod Index 7. Withdrawn 0
25 50 75 100 l
f f
f f
Group 7 0
25 50 75 100 I
I t
f f
Group 6 O
25 50 75-100 t
t i
l i
Group 5 1415 154 8-6 Babcock & Wilcox
Revised June 1978 Figure 8-o.
Power Imbalance Envelope for Operation From 125 5 to 265 : 15 EFFD Power, X of 2535 MWt
- - 110 RESTRICTED REG 10M
-22.25,102 100
-23".01,92 Il.26,92
~~
o 90 II.26,80 80 o
-27.72,80' 70 60 PERHISSIBLE OPERATING 50 REGION 40 30 20 10 I
I I
I I
I i 1
1 I
I
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power imbalance, 7 kk\\b g.7 Babcock & Wilcox
Revised June 1978 Figure 8-7 APSR Fosition Limits for Operation From 0 to 265 15 EFPD 57.9,102 100 o 6.1,102 o,
RESTRICTED 90
> 6.1,92 REGION 80
< 0,80 57.9,80 70 100,70 W
N 50 o
50 C
,E PERMISSIBLE 40 OPERATING REGION 30 20 10 I
I f
f I
I g
l 0
10 20 30 40 50 60 70 80 90 100 APSR, ". Withdrawn 1%T5'.<'4b5SS
~
14\\S
\\S6 83 Babcock ! Wilcox
F.evised June 1978 9
SILO. NIP ??.0 Gar: - 1.fts:Os TESTI:ro Tne Startup Test Progras associated with ecre perfor=ance as cutlined belov, was cenpleted on June 11, 1973. These tests verified that core perfor=ance is within the assu=ptions of the safety analysis and provide the necessary data for
~
continued safe operation. S.e results of these tests vill be sitb=itted to the li?.C by July 31, 1973.
i oreeritical Tests 1.
Control rod trip test Zero Pever Physics Tests 1.
Critical baron concentra:ica j
2.
Te=perature reactivi y : efficient All reds out; grcup 3 in, a.
b.
Groups 5 through 3 inserted; groups 1 through h cut 3
Centrol rod group resetivity vorth h.
Ejected control rod reactivity verth
'?ever Tests 1.
Core power distribution verificatien at approximately k0, 75 and 100% full power with nor=al control red group configuration.
l 2.
In-core versus out-of-core detector i= balance ec: relation verificatien.
3 Power Doppler reactivity coefficient at approxi=ately 1005 FP.
k.
Te=perature reactivity coefficient at approxi=2tely 1005 FP.
5 Dropped control rod core pcVer distributien verificatien at approximately k0% FP.
k 1415 157 9-1 set =cca s a11= x
Revised June 1978 10.
REFERE' ICES
- Three Mile Island !!uclear Station, Unit 1, Final Safety Analysic Report, US:TRC Docket No. 50-289.
Three Mile Island Unit 1 Cycle 3 Reload hport, 3AW-lkh2, 3abcoch & Wilcox, Lynchburg, Virginia (1976).
3 A. J. Echert, et al., Progra: to Deter.ine In-Reactor Perfor.ance of 3 W Fuels - Cladding Creep Collapse, 3AW-1C08h, Rev.1, Babcock & Wilcox, Lynch-brg, Virginia, ?tovember 1976.
"C ' ?uel Densifi:ati:n Ee; rt, 3AW-1389, 3abcock & Wilcox, Lyn:hburg, Virginia, June 1973.
e' T. ':. Merge.n and H. 5. T.ao, TATI - Fuel Pin Te=perature and Gas Pressure M Cysis, 3AW-1Cohh, 2at:::% i Wilcox, Lynchburg, Virginia, May 1972.
P. F.. 3:cuit, et al., !A : - Fuel Perfor=ance Analysis, 3AW-100STP, Rev 1, 31:::2 & *a*ilcox, Lyn:Y:n 5, Virginia, May 1976.
7 E. C. Jcnes, et al., I :5 h Cysis of 3&W's 177-FA Levered Loop USS, BAW-10103. Rev 2, Eabecek & Wil cx, Lynchburg, Virginia, April 1976.
S. A. Varga to J. H. Taylor, Letter, "Co==ents on 3&W's Sut tt ttal on Com-bination of Peaking Factors," May 13, 1977 9 ' S. A. Var;::a to J. E. Taylor, Letter, "U.:date of 3AW-10355, ' Fuel Densification Repor:'," December 5,1977 10 J. H. Taylor to R. L. Baer, Letter Report, " Analysis of S=all 3reaks in the Reactor Coolant Punp Discharge Piping for the 31W Lovered Loop 177FA Plants",
Babcock & Wile 3., Lynchburg, Virginia, May 1, 1978.
P00RORRU1
\\h\\5
\\SS 10-1 3abcock i Tilcox