ML19210B476
| ML19210B476 | |
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|---|---|
| Site: | Crane  |
| Issue date: | 01/31/1978 |
| From: | BABCOCK & WILCOX CO. |
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| BAW-1473, NUDOCS 7911080702 | |
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Text
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BAW-1473 November 1977 Fevisien 1 (January 1978)
THREE MILE ISLAND UNIT 1 CYCLE 4 RELOAD REPORT 1565 304 Babcock & Wilcox 791108079 Q
BAW-1473 November 1977 Fevisien 1 (Janua / 1978)
THREE MILE ISLAND UNIT 1 CYCLE 4 RELOAD REPORT 1565 305 BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox
CONTENTS Page 1.
INTRODUCTION AND
SUMMARY
1-1 2.
OPERATING HISTORY 2-1 3.
GENERAL DESCRIPTION 3-1 4.
FUEL SYSTEM DESIGN.
4-1 4.1.
Fuel Assembly 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.
Thermal Design.
4-2 4.4.
Material 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 1565 306 Babcock & \\Vilcox
- lii -
List of Tables Page Table 4-1.
Fuel Design Parameters and Dimensions...
4-3 4-4 4-2.
Fuel Thermal Analysis Parameters 5-4 3
5-1.
TMX-1, Cycle 4 Physics Parameters..........
5-2.
Shutdown Margin Calculation for TMI-1, Cycle 4 5-6 6-2 6-1.
Thermal-Hydraulic Design Conditions..
.......- 7-3 7-1.
Comparison of Key Parameters for Accident Analysis T-2.
Bounding values for Allowable LOCA Peak Linear Heat Rates T-3 List of Figures Figure 3-3 3-1.
Core Loading Diagram for TMI-1, Cycle 4 3-2.
Enrichment and Burnup Distribution for TMI-1, Cycle 4 3-4 3-5 3-3.
Control Rod Locations for TMI-1, Cycle 4....
5-1.
BOC (4 EFFD), Cycle 4 Two-Dimensional Relative Power Distribution -- Full Power, Equilibrium Xenon, APSRs Inserted..
5-7 8-2 8-1.
Core Protection Safety Limits 8-2.
Protection System Maximum Allowable Setpoints for Reactor 8-3 Power Imbalance 8-3.
Rod Position Limits for Four-Pump Operation From 0 to 125 8-4 1 5 EFPD - TMI-1, Cycle 4 8-4.
Rod Position Limits for Four-Pump Operation From 125 ! 5 EFPD 8-5 to 265 15 EFPD -- TMI-1, Cycle 4 8-5.
Rod Position Limits for Two-and Three-Pump Operation From 0 to 125 5 EFPD -- TMI-1, Cycle 4 8-6 8-6.
Rod Position Limits for Two-and Three-Pump Operation From 125
! S to 265 15 EFPD -- TMI-1, Cycle 4 8-7 8-7.
Power Imbalance Envelope for Operation From 0 to 125 ! 5 EFPD 8-8 8-8.
Power Imbalance Envelope for Operation From 125 5 to 265 8-9 15 EFPD 8-9.
APSR Position Limits for Operation From 0 to 265 i 15 EFPD.
8-10 1565 307 Babcock & Wilcox
- iv -
4 1.
INTRODUCTICN AND SLM!ARY This report justifies the operation of the Three Mile Islar.d Nuclear Station Unit 1 (TMI-1) (cycle k) at a rated core pcVer of 2535 MWt.
Included are the required analyses, as outlined in the USNRC document, " Guidance for Proposed License Amend =ents Relating to Refueling," June 1975 To support cycle h operatien of the TMI-1, this repcrt employs analytical techniques and design bases established in repcrts 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 or rods-cut mode. All nuclear parameters pertinent to accident analyses have been calculated consistent with this mode of operation. Section 5.3 describes the change to 1
feed-and-bleed operation.
Cycle h reactor parameters that are related to power capability are su==arized in this report and referenced to cycle 3.
All the accidents analyzed in the FSAR have been reviewed for cycle h operation, and in cases where cycle h characteristics proved to be conservative with respect to those analyzed previously, no nev analysis was performed.
The Technical Specifications have been reviewed, and the modifications required for cycle h operation are justified in this report. Based on the analyses per-formed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Energency Core Ccoling Systems (ECOS), it has been concluded that TMI-1, cycle h can be safely operated at the rated core power level of 2535 MWt.
1565 308 1-1 Babcock & Wilcox
C O
2.
OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of the Three Mile Island, Unit 1 is the operating cycle 3.
Cycle 3 achieved criti-cality on May 13, 1977, and af ter zero power testing attained 100% power on May 20, 1977. No control rod interchange was planned for cycle 3, which is scheduled for completion in ea.rly Maren af ter 270 10 EFPD. No operating anom-alies occurred during the first three cycles that would adversely affect fuel performance during the fourth cycle. The operation of cycle 4 is scheduled to begin in Apri] 1978.
The design cycle length is 265 1 15 EFPD, and no control rod interchanges are planned.
1565 309 Babcock t.Wilcox 2-1
3.
GEERAL DESCRIP'" ION The TMI-l reactor core is described in detail in Section 3 of tne Final Safety Analysis Report for that Unit.1 The cycle h core censists of 177 fuel asse=-
blies (FAs), each of which is a 15-by-15 array containing 208 fuel rods,16 control rod guide tubes, and one incere instrument guide tube. The undensified ncminal active lengths of the fael rods are lhh inches for batches le and 2b, lh2.6 inches for batch h, and 1h2.25 inches f or batches 5 and 6.
All fuel as-semblies in cycle h maintain a constant nemiral fuel leading of h63.6 kg of uranium. The cladding is cold-worked, Zircalcy h with an CD of 0.h30 inch and a vall thickness of 0.0265 inch. The fuel consists of dished-end, cylindrical pellets of uranium dioxide (see Table h-2 for data).
Figure 3-1 is the core loading diagram for TMI-1, cycle k.
The initial enrich-sents of batches le, 2b, and h were 2.06, 2.75, and 2.6h vt % uranium-235, re-spectively. Batches 5 and 6 have a 2.85 vt % uranium-235 enrich =ent.
All the batch 3 assemblies vill be discharged at the end of cycle 3, and the batch h and 5 assemblies vill be shuffled to new locations. The batch 6 assemblies vill cecupy te periphery of the core. The 13 batch le and the eight batch 2b asse=blies vill occupy interior core locations. Note that the designations le and 2b are used to identify asse=blies from the original batch 1 and batch 2 fuel. The le and Eb assemblies were removed frem the core at the end of cycle 1 and cycle 2, respectively (see Table h-1).
They are being reinserted into the cycle h core to lover feed batch size require =ents and spent fuel stcrage, thereby producing a more efficient fuel cycle. It should be noted that the assemblies referred to as batch la in the cycle 3 Reload Report (reference 2) are ncv designated as lb; batch la is new the remainder of batch 1 assemblies which have not been scheduled for reinsertion. Figure 3-2 is an eighth-core map showing each assembly's burnup 3-1 Babecek 's '411cox 1565 310
at the beginning of cycle h and its initial enrichment.
Cycle h will be operated in a rods-out, feed-and-bleed ucde. The core reac-tivity control vill be supplied =ainly by soluble boren and supple =ented by 61 full-length, Ag-In-Cd control rod assemblies (CRAs).
In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of axial power distribution. The cycle h locations of the 69 control rods and the group designations are indicated in Figure 3-3 The core locatiens of the 69 control rods for cycle h are identical to those of the reference cycle 3 However, the group designations diffe7 between cycle h and the reference cycle to minimize power peaking.
No control rod interchanges or burnable poisen rods are necessary for cycle 4.
The nominal system pressure is 2200 psia, and the core average densified nominal linear heat rate is 5 72 kW/ft at the rated core power of 2535 MWt.
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 1565 311 3-2 Babcock & Wilcox
Figure 3-1.
Core Loading Diagram for THI-1, Cycle 4 I
A B
El Ms
- 7 s
e e
e e
i s
a e
89 PS Elo Pl2 slo P4 El P7 NT
!s fc s
5 2
4 s
4 e
s a
0 (12 Pil slo att is as 8s Ps 89 e
a e
s s
e s
e 5
s e
a e
Ele Wit alt All Plc ti l Ps as 89 M2 E2 g
c' s
s C
s 4
C s
9 8
e a
s Ll3 Les uit A13
.as RF us 43 ut Li L3 I
(cV2)
(cT2) e s
to s
e s
e s
i s
e s
as e
a 113 E89 Ll2 Lil 411 89 ts El 43 L2 L4 A2 13 G
s e
s e
5 e
s a
s e
s e
s e
s Mal Fi2 Nis u13 sis al 74 57 si (3
NI Le is I-(CYl)
(CTI)
(Cfl)
(CYl)
(Cff)
-f H
s a
se s
le s
e ic e
s le s
Ic e
e sIl Die F12 Fil N13 als s4 a7 Ns F2 F4 02 e3 4
4 s
e s
e s
e s
e s
e 5
4 s
FIS Fls (32 cl3 ts as Cs c3 El FI F3 g
(cia)
(ci2) e s
!s 5
s 5
e 5
s 5
e s
te e
a slo tie alt oli sto es se os on 12 s2 (fi')
!!")
I!" 's s
s e
sit sie aio oso as os as es so N
e e
s s
a e
s a
s s
e s
s 09 s3 cl0 s12 Os et Ce s7 of 0
gg,33 gg,ij gg,33 e
e i
28 s
ic 5
2e e
a e
ct til c7 p
s e
a e
e e
a e
a R
s s
e a
s I
Z l
2 3
4 5
6 7
8 9
10 11 12 13 14 15 III PFevious Cycle Location (escept 2s netto) 1 Baten 1565 312 3-3 Babcock & Wilcox
Figure 3-2.
Enrichment and Burnup 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 H
14263 16182 7647 13600 7056 14263 16071 0
2.85 2.64 2.85 2.64 2.85 2.64 2.85 K
7636 18217 10723 14918 5176 19704 0
2.95 2.64 2.85 2.75 2.85 2.85 L
4792 14740 6690 23049 0
0 2.06 2.85 2.64 2.85 N
11718 8304 16979 0
2.64 2.85 2.85 N
15685 0
0 2.64 0
15689 P
R 1565 313 X.XX Initial Enrienment XXXXX BOC Burnup Babcock & Wilcox 3-4
Figure 3-3.
Control Rod Locations for TMI-1, Cycle 4 X
I A
8 3
7 3
C 1
6 6
1 0
7 8
5 8
7 E
1 5
2 2
5 1
F 3
8 7
5 7
8 3
G S
2 4
4 2
6 H
7 5
5 3
5 5
7 X
S 2
4 4
2 6
L 3
8' 7
5 7
8 3
M 1
5 2
2 5
1 N
7 8
5 8
7 0
1 6
6 1
P 3
7 3
R 1
Z 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 Group No. of rods Function 1
8 Safety 2
8 Safety X
Group NumDer 3
9 Safety 4
4 Safety 5
12 Control 6
8 Control 7
12 Control 8
8 APSRs total # 69 Babcock & Wilcox 3-3 1565 314
4.
FUEL SYSTEM DESIGN 4.1.
Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for TMI-1, cycle 4 are listed in Table 4-1.
All fuel assemblies are identical in concept and are mechanically interchangeable. All results, references, and identified conservatisms presented in section 4.1 of reference 2 are applicable to the cycle 4 reload core.
4.2.
Fuel Rod Design The mechanical evaluation of the fuel rod is discussed below.
4.2.1.
Cladding Collapse Creep collapse analyses were performed for three cycle assembly power histories.
The batch a3 reinserted fuel is more limiting than the other batches due to lower prepressurization, lower pellet density, end/cr longer previous incore exposure time.
The batch 2b assembly power histories were analyzed, and the most limiting assembly was determir.ed.
The power history for the most 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 EFPH (ef fective full power hours), which is greater than the maximum projected resi-dence time (Table 4-1).
The creep collapse analysis was performed based on the conditions set forth in references 2 and' 3.
4.2.2.
Cladding Stress The batch 1c 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 1.0% on cladding plastic circumfer-ential strain. The pellet design is established for a plastic cladding strain Babcock & Wilcox 4-1 1565 315
of less than 1", at values of maximum design local pellet burnup and heat gen-eration rate, which are considerably higher than the values that the 041-1 fuel is expected to see.
This vill result in an even greater =argin than the analysis de=cnstrated. The strain analysis is also based on the maximum allow-able value for the fuel pellet dia=eter and density and the lovest permitted tolerance for the cladding ID.
h.3.
Thermal Desien The incoming batch 6 fuel is ther= ally and gecmetricefly similar to the batch 5 fuel of cycle 3.
The TAFY5 fuel pin analysis perfor=ed for batch 5 fuel also applies to batch 6.
An analysis was also perfor=ed for batch 6 using the fuel 6
perfor=ance code, TAC 0. Where differences cecurred between corresponding calculated values of each code, the = ore conservative values were chosen for batch 6 design.
Ther=al analysis of the fuel rods assu=ed in-reactor densification to 96.5" TD.
The average fuel te=peratures (Table h-2) for batches 1 through 5 are taken from the TAFY analyses which define the linear heat rate (L*43) capability for each batch. The average te=perature shown for batch 6 was taken frem an average pin analysis using the TACO code. The value shown represents the 30L (100 !Gd/
mtU) average fuel te=perature at 5.80 kW/ft.
The average te=perature decreases with burnup to a value of ll20F at 38,000 !Gd/=tU.
Linear heat rate capabilities are based en centerline fuel melt. Batch 6 linear heat rate capability was determined based en the lover tolerance limit (LTL) of the fuel density specification. The design LFR 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 L and 5 The TACO analysis for batch 6 fuel gives a higher LHR capability.
Therefore, the =cre conservative TAFY LHR capability was used in the design of batch 6.
h.h.
Materia? Desien The che=ical ec=patibility of all possible fuel-cladding-coolant assembly in-teractions for the batch 6 fuel assemblies is identical to that of the present fuel.
h.5 Orerating F.xterience B&W's operating experience with Mark 3, 15-by-15 fuel asse=bly design has veri-fled the adequacy of the fuel assembly design. As of August 31, 1977, the fol-loving operating experience has been collected for the seven 3&W 177-FA plants using the Mark B fuel assembly:
h-2 Babecek & Wilcox 1565 316
Current cycle Cumulative Current max assembly net electrical Reactor cycle burnuo, Wd/mtU 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 Design Parameters and Dimensions res ue1 Twice-burned assys Once-burned assys Batch 2b Batch 4 Batch ic Batch 5 Batch 6 Fuel assembly type Mark B3 Mark B4 Mark B2 Mark B4 Mark B4 No. of assemblies 8
56 13 48 52 Fuel rod OD, 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 Zr02 Zr-4 Zr02 Zr-4 Zr-4 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 0.3695 0.3695 Fuel pellet initial Density, % TD 92.5 93.5 92.5 94.0 94.0 Initial fuel enrichment ut % 235U 2.75 2.64 2.06 2.85 2.85 Burnup (BOC) Wd/mtU 23,049 16,625 13,276 7,410 0
Cladding collapse Time, EFFH
>30,000
>30,000
>30,000
>30,000
>30,000 Residence Time, EFFH 23,976 19,272 17,736 19,200 19,080 1565 317 4-3 Babcock & Wilcox
Table 4-2.
Fuel Thermal Analysis Parameters Densified fuel parameters (*
Batch le Batch 2b Batch 4 Batch 5 Batch 6 Pellet diameter, in.
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 MWt, 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 LHR to ( fuel melt, kW/ft 19.6 19.6 20.15 20.15 20.15
(*}Densification to 96.5% TD assumed.
1565 318 Babcock & Wilcox 4-4
a 5.
NUCLEAR DESIGN 5.1.
Physics Characteristics Table 5-1 compares the ccre physics parameters of cycles 3 and h; these values were generated using PDQ07 for both cycles. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles.
The longer design life of cyale h vill produce a corresponding larger cycle differential burnup than designed for the reference cycle 3.
The lever accumulated average core burnup at the end of cycle h is =ainly 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 pcver distribution for the beginning of the cycle h at full power with equilibrium xenen and group 8 inserted.
The critical bcron concentrations are approximately the same as those of reference cycle 3.
The hot, full-power control rod verths are similar in both cycles except fcr group 7, which is significantly higher in cycle h, being ec= posed of 12 control rod assenblies rather than eight as in cycle 3.
Control rod verths are sufficient to =aintain the required shutdevn margin as indicated in Table 5-2.
The differences in the parameters between cycles 3 and h are due to changes in radial flux distributien, isotopics, and the difference in cycle lengths. The ejected rod worth in Table 5-1 are the maxi =um calculated values.
It is difficult to ec= pare values between cycles or between rod patterns since neither the rod patterns from which the CRA is ejected nor the isotopic distributions are identical.
Calculated ejected rod vorths and their adherence to criteria are considered at all times in life and at all pcver levels in the develop =ent of the red positien limits presented in section 8.
The maximum stuck rod vorth at the end of cycle h is similar to that for the reference cycle 3 but is icver at the beginning of the 1565 319 7 *-
5-1 3abcock & '411cox
The' following conservatisms were applied for the shutdevn calculations:
1.
Poison =aterial depletien allevance.
2.
10% uncertainty en net red worth.
3 Flux redistributien penalty.
Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional =odel.
3e shutdown calculation was analyzed at 277 EFFD. The maximum allovable inserted rod worth is smaller in cycle h than in cycle 3 due to the operation of this cycle in a feed-and-bleed mode in which control rod group 7 is cnly 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 moderator coefficients in cycle k.
The differential boren and xenon verths are similar for cycles 3 and h.
Ths effective delayed neutron fractions for cycle h show a decrease with burneg (also reported in the reference cycle 3).
5.2.
Analytical Input The cycle h incore measurement calculation constants used for computing core power distributions were prepared in the same manner as for the reference cycle.
5.3 Changes in Nuclear Design Cycle h is designed to operate in a feed-and-bleed mode in contrast to the rodded operation of cycles 1, 2, and 3 The =ajor difference in operational modes during equilibrium, steady-state conditions is that no full-length con-trol rods are inserted into the core.
( A s=all, bite insertion, approximately 10%, of cne regulating bank is maintained to allev discrete changes in soluble boren and to accommodate small temperature and load de=and changes. )
During load follev operation the regulating bank is inserted into the core only to offset power Doppler reactivity changes. Transient xenen reactivity effects are compensated by changing the soluble boren concentration.
The same calculational methods and design information used in reference cycle 3 vere used to obtain the i=portant nuclear design parameters in cycle h.
Additional calculations vere performed for soluble beren control, shutdevn, reactivity control, and radiatien analyses due to the modificatien in the mode of operation. As in cycle 3, both APSRA and CRA position lL=its, as well as power bdbalance limits, 1565 320 5-2 3abcock & Wileen
will be specified based en LOCA analyses.
'"hese operational limits and the RPS limits (Technical Specification changes) for cycle h are presented in section 8.
1565 321 5-3 3abecek & Wilcox
Table 5-1.
TMI-1, Cycle 4 Physics Paracieters Cycle 3(b)
Cycle 4 270 277 Cycle length, EFPD 8341 8557 Cycle burnup, mwd /mcU Average core burnup - EOC, mwd /mcU 18,352 17,844 Initial core loading, atU 82.1 82.1 Critical boron - BOC, ppm (no Xe)
HZP(c), group 8 (37.5% vd) 1317 1250 HZP, groups 7 and 8 inserted 1155 1115 HFP, group 8 inserted 998 1084 Critical boron - EOC, ppm (eq Xe) 08 group 8 (37.5% wd, eq Xe) 4 Control rod worths - HFP, BOC, % ak/k 1.18 1.10 Group 6 0.84 1.48 Group 7 Group 8 (37.5% wd) 0.54 0.46 Control rod worths - HFP, EOC,(d) % ak/k Group 7 1.11 1.57 Group 8 (37.5% wd) 0.50 0.50 Max ejected rod worth - HZP, % ak/k
- BOC 0.34 0.81 EOC 0.77 0.81 Max stuck rod worth - HZP, % ak/k BOC 2.42 1.95 EOC 2.06 2.03 Power deficit, HZP to HFP, % ak/k BOC
-1.58
-1.28
-2,15
-2.05 EOC Doppler coeff - BOC, 10-5 (ak/k/*F) 100% power (O Xe)
-1.47
-1.49 Doppler coeff - E0C, 10 5 (ak/k/*F) 100% power (eq Xe)
-1.51
-1.59 Moderator coeff - HFP, 10 4 (ak/k/*F)
BOC (0 Xe, 1084 ppm, group 8 ins)
-0.91
-0.63 EOC (eq Xe, 17 ppm, group 8 ins)
-2.54
-2.52 Boron worth - HFP, ppm /% ak/k BOC (1000 ppm) 107 105 EOC (17 ppm) 97 95 Xenon worth - HFP, % ak/k BOC (4 EFPD) 2.59 2.63 EOC (equilibrium) 2.64 2.73 1565 322 Babcock & Wilcox 5-4
Table 5-1.
(Cont'd)
Cycle 3(*
Cycle 4 Effective delayed neutron fraction - HFP BOC 0.00584 0.000586 EOC 0.00524 0.00522
(* Cycle 4 data are for the conditions stated in this report.
The cycle 3 core conditions are identified in reference 2.
(
Based on 253 EFPD at 2535 MWe, cycle 2.
(" HZP denotes hot zero power (532F T
)
full
- 8
- HFP denotes hot power (579F T""E).
(d)246 EFPD in cycle 3; 277 EFPD in cycle 4.
- Ejected rod worth for groups 5 through 8 inserted.
1565 323 Babcock & Wilcox 5-5
Table 5-2.
Shutdown Margin Calculation for TMI-1, Cycle 4 BOC, % Ak/k EOC,(*}% Ak/k Available Rod Vorth Total rod worth, HZP( )
8.71 8.81 Worth reduction due to burnup of poison material
-0.37
-0.46 Maxi:su:s stuck rod, HZP
-1.95
-2.03, Net worth 6.39 6.32 Less 10% uncertainty
-0.64
-0.63 Total available worth 5.75 5.69 Required Rod Worth Power deficit, HFP to HZP 1.28 2.05 Max allowable inserted rod worth 0.40 0.42 Flux redistribution 0.40 0.90 Total required worth 2.08 3.37 Shutdown Margin Total available - total required 3.67 2.32 Note: Required shutdown margin is 1.00% Ak/k.
(*)277 EFPD.
( )HZP denotes hot zero power (532F T##8); HFP denotes hot full power (579F T
).
y 1565 324 Babcock & Wilcox 5-6
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 14 15 H
.99 1.14 1.32 1.01 1.25
.91
.81
.76 K
1.33 1.13 1.21 1.08 1.17
.83
.76 N
L 1.37 J.09 1.07
.90 1.13
.65 M
1.01 1.15
.93
.96 N
1.00 1.10
.67 0
.52 P
1565 325 R
INSERTED R00 GROUP NUMBER X.XX RELATIVE POWER DENSITY 5-7 Babcock & Wilcox
6.
TE!AL-HYDPJWLIC DESIGN The inecming batch 6 fuel is hydraulically and geometrically similar to batch 5 fuel. The only difference between cycles 3 and h is the core configuratien.
2 The cycle 2 DN3R analysis was used for reference cycle 3 ; this analysis is also valid for cycle h as discussed below. The core configuration used for cycle 2 analysis consisted of 60 Mark 33 asse=blies and 117 Mark 3h assemblies with the most limiting (hot) assembly being a 33 assembly. The cycle h configuration consists of 13 Mark 32, 8 March B3, and 156 Mark Sh assemblies with the most limiting assembly being a Bh.
Both the Mark 32 and 33 assemblies have a higher resistance to flew than the Mark Bk assembly.
The minimum DIGR calculated from the cycle 2 analysis was ec= pared to the miniram DNER obtained from an analysis of an all Bk core. The cycle 2 analysis provided the more restrictive minimum DN3R. For cycle h the addition of the higher resistance Mark B2 and 33 assemblies provides additional DNBR =argin.
The higher resistance Mark 32 and 33 assemblies vill tend to increase flow through the limiting Mark Sh assemb'.y.
Therefore, the cycle 2 DNER analysis bounds that for cycle k.
This conservatism is a=plified when peaking factors are censidered. The minimus DNER calculated from the cycle 2 analysis is based on a radial-local peak of 1.783.
The maximam radial-lecal peak calculated for cycle k operatien, including 8%
nuclear uncertainity, is 1.637 at BOL. This decreases to 1.h21 at the end of cycle h.
This provides an 8.2% margin to the design peak at ECC and a 20.3%
margin at ECC. This margin ensures that the design conditions shown in Table 6-1 vill not be exceeded during cycle h operation.
All other thermal hydraulic analyses which vere applicable to cycle 3 remain applicable to cycle k.
1565 326 6-1 3abcock & Wilcox
e Table 6-1.
Thermal-Hydraulic Design Conditions Cycle 1 Cycles 2, 3 and 4 Power level, HWe 2568 2568 System pressure, psia 2200 2'l00 Reactor coolant flow, % design 100.0 106.5 Vessel inlet coolant temperature at 100% power, F 554.0 555.6 Vessel outlet coolant temperature 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, in.
Table 4-2 Table 4-2 2
171,470 174,870 Avg heat flux (100% power), Btu /h-ft 2
Maximum heat flux (100% power), Btu /h-ft (for DNBR calculation) 457,825
~466,903 CHF correlation V-3 B&W-2 Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 Minimum DNBR (densified fuel) 1.46 1.877 (114% power)
(112% power) 1565 327 Babcock z.Wilcox 6-2
7.
ACCIDENT AND TRANSIENT ANALYSIS 7.1.
General Safety Analysis Each FSARI accident analysis has been examined with respect to changes in cycle 4 parameters to determine the effect of the cycle 4 reload and to ensure that thermal performance during hypothetical transients is not degraded.
The effects of fuel densification on the FSAR accident results have been evalu-ated and are reported in reference 4.
Since batch 6 reload fuel assemblies contain fuel rods with a theoretical density higher than those considered in reference 4, the conclusions in that reference are still valid.
The dose evaluations in the FSAR were based on conservative values for fuel burnup and power peaking; however, improved fuel utilization and improved cal-culational methods have led to a higher plutonium-to-uranium fission 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 dnses associated with the I
major accidents in Chapter 14 of the FSAR with the 2-hnur 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 8 to 15% over the FSAR, the cycle 4 doses are still only a very small 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 an 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 Babcock & Wilcox 7_1 1565 328
first core values (FSAR values) of core thermal parameters and subsequent fuel batches to paracaters used in cycle 4 analyses are given in Table 4-2.
A com-parison of the cycle 4 thermal-hydraulic maximum design conditions to the pre-vious cycle values is presented in Table 6-1.
These parameters are common to all the accidents considered in this report. A comparison of the key kinet-ics parameters from the FSAR and cycle h is provided in Table 7-1.
A generic LOCA analysis has been perfor=ed for the BW 177-FA lowered loop NSS using the Final Acceptance Criteria ECCS evaluation model reported in reference 7.
This analysis is generic in nature since the limiting values of the key parameters for all plants in this category were used.
Further= ore, the com-bination of the average fuel temperature as a function of linear heat rate and the lifetime pin pressure data used in the LOCA limits analysis (reference 7) is conservative compared to those calculated for this reload.
Thus, the analy-sis and the LOCA limits reported in reference 7 provide conservative results for the operation of TMI-1, cycle 4 fuel. A tabulation showing the bounding values for allowable LOCA peak linear heat rates for TMI-1, cycle 4 fuel are provided in Table 7-2.
It is concluded by examination of cycle 4 core thermal properties and kinetics properties with respect to acceptable previous cycle values that this core re-load will not adversely affect the ability to safely operate the TMI-1 plant during cycle 4.
Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 4 is considered to be bounded by previously accepted analyses. The initial conditions of the transients in cycle 4 are bounded by the FSAR and/or the fuel densification report" and/or subsequent cycle analyses.
1565 329 Babcock a.Wilcox 7-2
Table 7-1.
Comparison of Key Parameters for Accident Analysis FSAR and densif'n Parameter report value Predicted value
~
~
Doppler coeff (BOC), ak/k/*F
-1.17 x 10
-1.49 x 10 Doppler coeff (EOC), ak/k/*F
-1.33 x 10
-1.59 x 10 '
~
~
~
~
Moderator coeff (BOC), ak/k/*F
+0.5 x 10
-0.63 x 10
~
~
Moderator coeff (EOC), ak/k/*F
-3.0 x 10
-2.52 x 10 All rod group worth (HZP) 4 Sk/k 10.0 8.71 Initial boron cone. (H2P) ppm 1200 1084 Boron reactivity worth (70*F),
75 74 ppm /1% ak/k Max ejected rod worth (IIFP), ". Ak/k 0.65 0.28 Dropped rod worth (HFP), % ak/k 0.46 0.20 Table 7-2.
Bouncing Values for Allowable LOCA Peak Linear Heat Rates Allowable peak linear Core elevatic., ft heat rate, kW/ft 2
15.5 4
16.6 6
18.0 8
17.0 10 16.0 1565 330 Babcock s.Wilcox 7-3
8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications hsva been revised for cycle 4 operation.
The changes were made as a result of the following:
1.
TMI-l will be changed from a rodded to a feed-bleed mode of operation for cycle 4.
This change is not regarded as a major change in the operating mode since TMI-1 was operated in essentially a rods-out configuration dur-ing the latter part of the previous cycles.
Imbalance limits, control rod position limits, and APSR position limits are utilized to control power peaking and linear heat rates. A power level cutoff of 92% full power is used to control power peaking due to transient xenon effects.
2.
The quadrant tilt limit will be changed from a maximum actual core tilt of 3.41% to a maximum actual core tilt limit c' 4.91%.
A maximum actual core tilt limit of 4.92% was used in cycles 1 and 2, where.ss cycle 3 used a maxi-mum actual core tilt limit of 3.41%.
The larger tilt limit was used for cycle 4 due to the larger operating windows for feed-bleed operation.
3.
The Technical Specification limits based on DNBR and LHR criteria include appropriate allowances for projected fuel rod bow penalties, i.e.,
poten-tial reduction in DNBR and an increase in power peaks. A statistical com-bination of the nuclear uncertainty factor, engineering hot channel factor, and rod bow peaking penalty was used in evaluating LHR criteria, as approved in reference 8.
4.
Per reference 9, the power spike penalty due to fuel densification was not used in setting the DNBR-and ECCS-dependent Technical Specification limits.
1565 331 Babcock & Wilcox 8-1
Figure 8 Core Protection Safety Limits THERIAL POWER LEVEL, 5 120 DNBR tIulf (112) 1 (44,112)
(.44.112)
ACCEPTABLE 4 PuuP OPERATION
- 100 KI/FT lIEII LIEli
( 44.17.1) 80 (17.1) 2 (44,87.1)
WW
( 80.80)
ACCEPTABLE 80 3 & 4 PUMP OPERATION 70 (59.8) 3 (44.59.8)
( 44,59.6) gg 5W8. 0
( 49.2,49.2) 1.CCEPTA9t.E
" 50 (43.2.49.2) 2,3, & 4 PUMP OPERATION
,, 4g 30 20 10 i
i i
60 50 40 30 20 10 0
10 20 30 40 50 60 Reactor Poser lanalance, S CURVE REACTOR COOLANT FLOW (In/nt)
I 139.8 x 106 2
104.5 x 106 3
88.8 a 106 1565 332 Babcock & Wilcox 8-2
Figure 8-2.
Protection System Maximum Allowable Setpoints for Reactor Power Imbalance THERMAL POWER LEVEL, 5 120
(.17,108)
~~'
('
(17.108) 100 l
y,j, ACCEPTABLE M2 = -1.0 4 PUMP
(-35,90)
OPERATION
- 96 (35,90)
I (80.7) i dU l
CCEPTABLE 70 3 f. 4 PUMP
(-35,62.7)
OPERATION (35,62.7) l
-- 60 l
1 (53.1)
I 50 I
ACCEPTABLE
- 40
(-35,35.1)
OPIkA10N l
__ 30 I
l
-- 20 o
o I
4 M
{!
,, ' l
-- 10 u
I J
, i i
i e
i i
-70 60 50 40 30 20
-10 0
10 20 30 40 50 60 70 Reactor Power imoalance, 5 1565 333 8-3 Babcock 8. Wilcox
t Figure 8-3.
Rod Position Limits for Four-Pump Operation From 0 to 125 2 5 EFPD - TMI-1, Cycle 4 P0wfA 274.1.102 LEVEL
~
CU70FF 138,102 100
= 925 274.1.92 N07 ALLOWE0 90 RESTRICTED
~
$HUTDOWN BARGIN Lluli 200,70 E
70 mm 2
20,60 50 85.50 5
40 PERMl 331BL'E 30 20 3,8.5 10 40,10 0,0 t
I I
t 9
t f
f f
f f
0 25 50 75 100 125 150 175 200 225 250 275 300 Rod index, 5 witnarann 25 50 75, 10,0 0,
Group 7 0
25 50 75 100 t
t t
t 1
0 25 50 75 100 Group 6 1
f f
i Group !
1565 334 Babcock 8. Wilcox g_4
Figure 8-4.
Rod Position Limits for Four-Pump Operation From 125 : 5 EFPD to 265 15 EFPD - TMI-1, Cycle 4 POWER 230.102 274.1.102
' LEVEL 100 PCT ALLCWED 25 274.1.92 g
RESTRICTED 80 243.2,30 SHUTDOWN MARGIN Llulf 70 A
N 60 PERMl!31BLE 188.50 50 E'
40 30 20 118.15 120,10 10 0, 2. 7 i
i t
e e
f 3
25 50 75 100 125 150 175 200 225 250 275 300 Roa inats, $ setnarann 0
25 50 75 100 t
i i
i i
0 25 50 75 100 n
f i
i 0
25 50 50 75 100 Group 8 i
f 1
I i
Group 5 1565 335 Babcock & Wilcox 8-5
Figure 8-5.
Rod Position Limits for Two-and Three-Pump Operation From 0 to 125 5 EFPD - TMI-1, Cycle 4 138.102 150.102 248.2.I02 igg W T AL O ED RESTRIOTED FOR 3 Pt,'uP to 00,89 80 mf 70 LIElf 120,75 IMUTDQfN BARGIN j
80 PERul $st BL E S
SO 85,50 8
~
ii 40 8
30 A
20 4 'I 0,I.5 60,15 10 t
e f
I 1
't t
f 1
I f
0 25 50 75 100 125 150 175 200 225 250 275 300 Rao India, 5 ot tnataan 0
25 50 75 100 I
f I
f f
Group 7 q
2,5 Sq 7{
1q0 0
25 50 75 I00 Group 6 i
i Grour 5 1565 336 8-6 Babcock s Wilcox
Figure 8-6.
Rod Position Limits for Two-and Three-Pump Operation From 125 : 5 to 265 : 15 EFPD - TMI-1, Cycle 4 230.102 248.2.102 100.
90 j
80 NOT ALLCEED 5
SHUT 00fN MARGIN
\\
3 70 LIMIT s
60 p
s 168,50 50 f
8 PERulSSIBLE 30 20 118.15 130,15 30
~0,2.7 i
t 0
25 50 75 100 125 150 175 200 225 250 275 300 Roo inces, 5 tithdrawn 0
25 50 75 100 t
I f
f f
50 75 iOO 0
25, Gr0Up 6 0
25 50 75 100 t
t t
f I
Group 5 1565 337 Babcock & Wilcox 8-7
Figure 8-7.
Power Im'oalance Envelope for Operation From 0 to 125 : 5 EFPD POWER, 5 0F 2535 H t RESTRICTED REGION
- 110
-30.80,102' I~
- 100
-30.21,92 80 13.45,80
-- BG o
-38.25,80 i
70 PERMISSIBl.E OPERATING HEGION 60
- 50
-- 40
- 30
-- 20 10 i
i i
i i
i i
i t
-50 30
-20
-10 0
10 20 30 40 50 Axial Power imoalance, "
1565 338 Babcock & Wilcox 8-8
Figure 8-8.
Power Imbalance Envelope for Operation From 125 2 5 to 265 : 15 EFPD POWER, 5 0F 2535 MWt RESTRICTED REGION
- - 110
-30.07,102
--100 12.87,102 1
-29.14,92 i
- 90 13.45,92
-36.81,80
-- 80 13.45,80
-- 70 PERMI SSI BL E OPERATING
-- 60 REGION
-- 50
- 40 30
-- 20
- 10 i
i i
i i
i i
i i
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power imoalance, 5 1565 339 8-9 Babcock & \\Vilcox
Figure 8-9.
APSR Position Limits for Operation From 0 to 265 t 15 EFPD 57.9,102 100 RESTRICTED REGION 90 57.9,80 100,70 7
g e
5 60 PERMISSIBLE 7
I"U 50 REGION f.
E 40 30 20 10 1
I I
I f
f f
I f
0 10 20 30 40 50 60 70 80 90 100 APSR 5 witndrawn 1565 340 Babcock & Wilcox 8-10
9.
STARTUP PROGRAM - PHYSICS TESTING The planned Startup Test Program associated with core performance is outlined below.
These tests verify that core performance is within the assumptions of the safety analysis and provide the necessary data for continued safe operation.
Precritical Tests 1.
Control rod trip test.
Zero Power Physics Tests 1.
Critical boron concentration.
2.
Temperature reactivity coefficient.
a.
All rods out; group 8 in.
b.
Groups 5 through 8 inserted; groups 1 through 4 out.
3.
Control rod group reactivity worth.
4.
Ejected control rod reactivity worth.
Power Tests 1.
Core power distribution verification at approximately 40, 75, and 100% full power with normal control rod group configuration.
2.
In-core versus out-of-core detector imbalance correlation verification at approximately 75% full power.
3.
Power Doppler reactivity coefficient at approximately 100% FP.
4.
Temperature reactivity coefficient at approximately 100% FP.
5.
Dropped control rod core power distribution verification at approximately 40% FP.
1565 341 Babcock & VVilcox 9-1
10.
REFERENCES 1
Three Mile Island Nuclear Station, Unit 1, Final Safety Analysis Report, USNRC Docket No. 50-289.
Three Mile Island Unit 1, Cycle 3 Reload Report, BAW-lkh2, Babcock & Wilcox, Lynchburg, Virginia (1976).
A. J. Echert, et al., Program to Determine In-Reactor Perfor=ance of B&W Fuels - Cladding Creep Collapse, BAW-1008h, Rev. 1, Babcock & Wilcox, Lynch-burg, Virginia, Nove=ber 1976.
h TMI-l Fuel Densification Report, BAW-1389, Babcock & Wilcox, Lynchburg, Virginia, June 1973 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-100hh, Babecek & Wilcox, Lynchburg, Virginia, May 1972.
6 R. H. Stoudt, et al., TACO - Fuel Performance Analysis, BAW-10087P, Rev 1, Babcock & Wilcox, Lynchburg, Virginia, May 1976.
I R. C. Jones, et al., ECCS Analysis of B&W's 177-FA Lovered Loop NSS, BAW-10103, Rev 2, Babcock & Wilcox, Lynchburg, Virginia, April 1976.
S. A. Varga to J. H. Taylor, Letter, "Cc==ents en B&W's Submittal on Cem-bination of Peaking Factors," May 13, 1977 9
S. A. Varga to J. H. Taylor, Letter. "Undate of BAW-10055, ' Fuel Densification Report'," December 5, 1977 1565 342 10-1 Babecek & Wilcox