ML20151C327
| ML20151C327 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 03/31/1988 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20151C300 | List: |
| References | |
| BAW-2015, NUDOCS 8804120329 | |
| Download: ML20151C327 (57) | |
Text
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BAW-2015 March 1988 i
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THREE MILE ISLAND UNIT 1 CYCLE 7 RELOAD REPORT l
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Mart:21 1988 i
'DIREE MILE ISIAND LNIT 1 CYCIE 7 RELOAD RERRP i
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B & W FUEL CMPANY P. O. Box 10935 Lyrriturg, Virginia 24506-0935 B&W Fuel Comparty
N Page 1.
INIBODUC. TION AND SUt9%RY....................
1-1 2-1 2.
OPERATING HIS'ICRY 3-1 3.
GDUAL DESCRIPTION 4-1 4.
FUEL SYSID4 DESIGi.......
4.1.
Ibel Asserably Mechanical Design 4-1 4.2.
Ibel Rcd Designs.....................
4-1 4-1 4.2.1.
Cla&iing Collapse 4-2 4.2.2.
Cladiing Stress 4-2 4.2.3.
Cla& ling Strain 4.3.
Therinal Design......................
4-2 4-3 4.4.
Material Design 4.5.
OperatirrJ Experience...................
4-3 t
5-1 5.
NUCEAR DESIGi.
5-1 5.1.
H1ysics Characteristics 5-2 5.2.
Analytical Input..
5-2 5.3.
Changes in Nuclear Design 6-1 6.
THERMAIeHYIRAULIC DESIGi....................
7-1 7.
ACCIDD(I AND TPRGIDTP MRLYSIS 7-1 7.1.
General Safety Analysis 7-1 7.2.
Accident Evaluation...................
8.
HOICSED PODIFICATIQE TO TEQNICAL SPECIFICATIONS.
8-1 6
9-1 9.
STARIUP HOGRAM - IHYSICS 7ESTING...............
9-1 9.1.
Precritical 7tsts....................
9-1 9.1.1.
Control Rod Trip 74st 9-1 9.1.2.
RC Flcw.....................
9.2.
Zero Power Physics Tests.................
9-2 9.2.1.
Critical Boron Correntration...........
9-2 9.2.2.
Terperature Reactivity Coefficient........
9-2 9.2.3.
Control Rod Group / Boron Reactivity Worth.....
9-2 9.3.
Power Escalation Tests..................
9-3 9-3 9.3.1.
Core Syurnetry W.................
r' 9.3.2.
Core Power Distribution Verification at Intar=ilate Power level (IPL) and 100% FP L
With Ncr:Linal Control Rod Position 9-3 11 I
B&W Fuel Comparty 4
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,..,..,. - ~,... _. _,..,
.,,,,n
-c..-
CCtTrDTIS (Cbnt'd)
Page 9.3.3.
Inoc,re Vs. Excore Detector Imbalance Correlation Verification at the IPL 9-4 9.3.6
'I4mperature Reactivity Coefficient at 100% FP.
9-5 9.3.5.
Power Doppler Reactivity Coefficient at 100% F.'
9-5 9.4.
Procedure for Use if Acceptance Criteria Not Met.....
9-5 10-1
- 10. REFERINCES.
List of Tables Table 4-1.
Fuel Design Parameters ard Dimensions.............
4-4 5-3 5-1.
M -1 Cycle 7 Ehysics Parameters 5-2.
Shutdown Margin Calculation for 'IMI-1 Cycle 7.........
5-5 6-1.
Maximum Design Conditions, Cycles 6 ard 7...........
6-3 7-1.
Conparison of Key Parameters for Accident Analysis 7-3 7-2.
Bounding Values for Allcuable IOCA I%ak Linear 7-3 Heat Rates for 'IMI-1 List of Floures Figure 3-3 3-1.
Core Icadirg Diagram for 'IMI-1 Cycle 7 3-2.
Enrichment and Birnup Distribution for 'IMI-1 Cycle 7 3-4 3-3.
Control Rod Iocations and Croup Designations for 3-5
'IMI-1 Cycle 7.
~
3-4.
IEP 2nrichnent ard Distribution for 'IMI-1 Cycle 7.......
3-6 5-1.
BDC (4 EFPD), Cycle 7 TWcr-Dimensional Relative Power Distribution - n111 Power, Equilibrium Xenon, APSRs 5-6 Inserted................,..........
8-3 8-1.
Core Protection Safety Limit 8-4 8-2.
Core Protection Safety Limits 'IMI-1 Cycle 7.
8-5 8-3.
Core Protection Safety Bases 8-4.
Protection System Maximum Allowable Setpoints.........
8-6 8-5.
Protection System Maxinum Allowable Setpoints 8-7 for Axial Power Imbalanos, 'IMI-1 Cycle 7 8-6.
Rod Position Setpoints for Four-Purp Operation Frtan 0 8-8 to 40 +10/-0 EFPD, 'IMI-1 Cycle 7 8-7.
Rod Position Setpoints for Four-Pu p Operation Frcan 40 +10/-O to 100 +10/-0 EFPD, 'IMI-1 Cycle 7..........
8-9 8-8.
Rod Position Setpoints for Four-Purp Operation After 8-10 100 +10/-0 EFPD, 'IMI-1 Cycle 7..........
8-9.
Rcd Position Setpoints for 'Ihree-Purp Operation Frcan 8-11 O to 40 +10/-0 EFPD, 'IMI-1 Cycle 7...............
iii B&W Fuel Company
T3st of Fiaures (Cbnt'd)
Figurs Page 8-10. Rod Position Setpoints for 'Ihrws-Pung Operation Fran 8-12 40 +10/-0 to 100 +10/-0 DTD, 'IMI-1 Cycle 7 8-11. Rod Position Setpoints for 'Ihree-Punp Operation After 8-13 100 +10/-O EFPD, 'IMI-1 Cycle 7........
8-12. Rod Position Setpoints for 'Iko-Punp Operation Frun 8-14 0 to 40 +10/-O EFPD, 'IMI-1 Cycle 7.........,.....
8-13. Rod Positico Setpoints for 'I%o-Punp Operation Fran 8-15 40 +10/-0 to 100 +10/-0 DTD, 'IMI-1 Cycle 7 8-14. Rod Position Setpoints for 'I%o-Punp Operation After 8-16 100 +10/-0 EFPD, 'IMI-1 Cycle 7............
8-15. Axial Power Imbalance Envelope for Operation 8 17 Fran O to 40 +10/-0 IFPD, 'IMI-1 Cycle 7 8-16. Axial Power Dnbalance Envelope for Operation Fran 40 +10/-0 to 100 +10/-0 DTD, 'IMI-1 Cycle 7.....
8-18 8-17. Axial Power Imbalance Envelope for Operation 8-19 After 100 +10/-0 EFPD, 'IMI-1 Cycle 7.
8-18. IDCA Limited Maxirun Allowable Linear Heat Rate, 8-20
'IMI-1 Cycle 7.
iv B&W Fuel Company
. }
1.
INIRODUCI' ION AND SLM4ARY t
'Ihis report justifies the operation of the Three Mile Islard Nuclear Station Unit 1 (cycle 7) at a rated core power of 2568 PHt.
Included are the required analyses as outlined in the USNRC hnnant, "Guidance for PrW_
License Amendments Relating to Refueling," June 1375.
'Ib support cycle 7 operation of the TMI-1 nuclear station, this report ertploys analytical techniques and design bases established in reports that have been subnitted aM received technical approval by the USNRC (see references),
t
'Ibe design for cycle 7 raised the rated thermal power frcan 2535 to 2568 PHt, which was the ultimate core power level identified in the 'Ihree Mile Island Unit 1 Final Safety Analysis Report (ISAR).1 Cycle 7 reactor parameters that are related to power capability are summarized in this report. All the accidents analyzed in the IBAR have been reviewed for cycle 7 operation.
'Ibe cycle 7 characteristics proved to be conservative with respect to all applicable safety limits and criteria.'
'Ibe Technical Specifications have been reviewed, and the mMifications requir.d for cycle 7 operation are justified in this report.
Rml on the analyses performed, which take into account the postulated effects of fuel densificatico ard the Final Acceptance Criteria for emergency core cooling systems (ECCS), it has been concluded that 'IMI-1, cycle 7 can be safely operated at the rated core power level of 2568 PWt.
1-1 B&W Fuel Company
3 2.
OPERATDC HISICRt r
'n.e reference cycle for the ruclear and thermal-hydraulic analyses of the
'Ihree Mile Islard, Unit 1 plant is the operating cycle 6.
Cycle 6 achieved criticiality on March 23, 1987, ard after zero power testing began powcr op-eration on March 26, 1937. Cycle 6 is scheduled for empletion in June 1988 after 425 i 15 EFPD.
No operatirg anomalies have occurred during the sid.
cycle.
'Ibe operation of cycle 7 is scheduled to beain in August 1988. 'Ihe design cycle length is 445 15 EFFD.
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r 2-1 B&W Fuel Company
i 3.
GDEPAL DESCRIFFION The 7MI-l reactor core is described in detail in section 3 of the Firal Safety Aralysis Report for the Unit.1 The cycle 7 core consists of 177 fuel assemblies (FAs), each of which is a 15 by 15 array containing 208 fuel rods,16 ALwl rod guide tubes, and one incore instrument guide tube. The fuel is ccuprised of dished-erd, cylindrical pellets of uranium dioxide clad in col.1-worked Zircalcy-4.
All fuel aSalies ir. cycle 7 have a constant ncninal fuel loadire of 463.6 kg of uranium. Tne undensified raninal active fuel legths, theoretical densities, fuel and fuel red dimensions, and other related fuel parameters are given in Table 4-1 for all fuel asser:blies.
Figure 3-1 is the fuel shuffle diagram for 1MI-1, cycle 7.
The initial enrictment of batches 7B, 8B, ard 9A is 2.85 wt% U-235; batches BA ard 9B have an enrictment of 2.95 wt% U-235, ard the batch 9C enrichment is 3.63 wt% U-235.
All batch 6 and 27 batch 7 assablies will be disc arged at the end of cycle 6.
The remainire 25 batch 7B, and all the batch 8 medlies will be shuffled to new locations. The 76 fresh batch 9 assemblies will be loaded in a syur:etric checkerboard pattern thren.ghout the core.
One batch 9A FA, which contains 9 rods of 2.99 st% U-235 will be in core location K-12 (see Section 8 for criteria).
Five other batch 9A FAs ard two batch 9B FAs, containirq two rods each of 2.83 wt% U-235, will be in core locations D-5, D-7, E-4, H-7, K-8, M-6 and N-9.
Figure 3-2 is an eighth-core map
.howing the acely burnup ard enrichment distribution at the beginning of f
cycle 7.
Cycle 7 will be operated in a feed-and-bleed mode. Reactivity is controlled by 61 full-length Ag-In-ci ocotrol rods, 68 burnable poison red ase.emblies (BPPAs), ard soluble boron shim.
In addition to the full-lergth control rods, eight Inoonel axial power shapirq rods (gray APSRs) nre prWided for adiitional control of the axial pcuer distrib2 tion.
The cycle 7 locations of the 69 control rods ard the grcup designations are the same as in cycle 6, ard are indicated in Figure 3-3.
The cycle 7 locations ard enrichments of the BWAs are shown in Figure 3-4.
3-1 B&W Fuel Company
'1he ncninal systen pressure is 2200 psia and the oore average ncainal linear heat rata is 5.73 kW/ft at the rated oore power of 2568 )Mt.
l l
3-2 B&W Puel Coenpany
Figure 3-1.
Core Loading Diagram for TMI-1 Cycle 7 FVtt itAa5Ftt C M r--
i 1
M M
79 M
38 A
gg gg gg ggg ggg M
M 88 SC 78 M
as M
as I
L3 F
m3 F
C13 F
n13 F
L13 M
M fl K
70 M
70 M
78 K
4A til F
M F
A4 F
A10 F
812 F
M6 as M
3A M
as H
d M
If, M
M M
IB 0
C10 F
M9 F
me F
4 F
mit F
La F
C6 K
75 M
as k
M M
N H
M M
78 SC g
F 02 F
Fl F
M2 F
M14 F
K2 F
014 F
88 IS M
Is M
sA M
M M
M M
88 9C 88 M
F 09 C12 P
011 F
M10 F
F7 F
Ll F
D$
F C4 07 88 K
78 M
88 H
M M
N N
se M
Tl M
88 6
pg pg y
ggg
,g3 og gg g
pgg a
py 78 78 M
88 M
M M
78 M
M M
M M
79 78 El C3 F
M11 F
K14 F
tl F
$2 F
ml F
013 til as M
Tl N
88 H
84
'A te la 88 M
70 l:
88 Lt F
L1 F
Fil F
C4 F
M3 F
Fl F
L18 F
L7 E
se as M
M M
M la et la SA lA te K
at at k
49 012 F
411 F
Fil F
89 F
(6 F
nl F
04 47 M
Tl M
68 H
84 la 88 lA 64 M
75 M
F m2 F
G14 F
E!
F (14 F
87 F
R14 F
M M
M 94 84 SA 44 M
'84 94 6A K
GB Clu F
G4 F
(4 F
tl F
E!!
F nF F
04 8A K
78 9C 78 M
78 M
18 M
4A 1
0 gge p
pg og agg pgg a
pg 88 M
88 SC 71 M
M M
to F3 F
03 F
03 F
013 F
Fil 88 84 78 84 C4 i
i t
g4 gg af gtg gig L
I J
1 3
3 4
5 6
7 8
9 10 II II l3 11 laten a
Fr..iw i Cicie tocati a l
3-3 B&W Fuel Company
figure 3-2.
Enrichment and Burnup Distribution for TMI-1 Cycle 7 8
9 10 11 12 13 14 15 2.85 2.95 2.85 2.85 2.85 3.63 2.85 2.85 17,558 0
13,331 0
16,507 0
15,735 17,554
)
2.85 2.85 2.85 2.85 2.85 3.63 2.85 X
16,000 0
11,292 O
16.347 0
16,344 2.95 2.85 2.85 3.63 2.85 2.85 L
16,817 0
16,236 0
12.711 16.695 2.85 2.85 2.85 3.63 M
13,329 0
13,746 0
2.95 3.63 2.85 16,451 0
16.035 2.95 16,788 P
1 A
X.XX Initial enrichment XX,XXX BOC burnup, mwd /mtU 34 B&W Fuel Company
Figure 3-3.
Control Rod Locations and Group Designations for THI-1 Cycle 7 X
Fuel Transfer Canal
)
A 1
6 1
B 3
5 5
3 C
D 7
8 7
8 7
E 3
5 4
4 5
3 F
1 8
6 2
6 8
1 G
5 4
2 2
4 5
H W-6 7
2 4
2 7
6 Y
5 4
2 2
4 5
K L
1 3
6 2
6 8
1 M
3 5
4 4
5 3
i N
l 7
8 7
8 7
3 5
5l 3
0 l
P l
l 1
6 1
R I
i Z
1 2
3 4
5 6
7 8
9 10 li 12 13 14 15 X
Group Number Group No. of Rods Function 1
8 Safety 2
8 Safety 3
8 Safety 4
9 Safety 5
12 Control 6
8 Control 7
8 Control 8
8 APSRs 3-5 54W Fuel Comparty
Figure 3-4.
LBP Enrichment and Oistribution for THI-1 Cycle 7 8
9 10 11 12 13 14 15 t
H 1.000 1.000 1.312 K
1.000 0.800 0.500 0.200 L
I L
0.800 0.800 1.312 M
1.000 0.800 0.200 4
I N
0.500 0.200 0.200 0
1.312 1.312 0.200 P
e p
0.200 I
i R
No. of Concentration.
BPRAs wt% B C
[
4 12 1.312 8
1.000
[
X.XXX LBP Concentration 16 0.800 (wtt B C in Al 0 )
8 0.500 4
23 i
24 0.200 I
l Total 68 I
i 3-6 B&W Fuel Company
]
1
- 4. FUDr. SYSIIM DESIGN 4.1.
Fuel Assembiv Mechanical Damian The types of fuel assenblies aM pertinent fuel design parameters for M-1, cycle 7 are listed in Table 4-1.
All fuel assemblies are identical in w.pt aM are mechanically interchargeable.
Retainer medlies will be used en the two fuel====mblies containing the regenerative neutron sources (RNS) and the 68 assenblies containing BPPAs.
The justification for the design aM use of the retainers described in references 2 and 3 is applicable to the RNS retainers in cycle 7 of M-1.
All results, references and identified conservatisms presented in section 4.1 of the 4
cycle 6 reload report are applicable to the cycle 7 reload core.
The batch 9C fuel uses Zircaloy rather than Imenel as the material for the intermediate spacer grids as reported in reference 5.
The NRC safety evaluation 6 of that report requires that a licensee who is irrorporating that design suknit a plant-specific analysis of ocmbined seismic and IICA loads accordirq to /fpeniix A to Standard Review Plan 4.2.
The analysis that was presented in reference 5 envelopes the M -1 plant de. sign requirenants. Therefore, the margin of safety reported for the Mark BZ fuel
===ambly is applicable to M -1.
4.2.
7 bel Rod Desians The pin pre-pressure in 38 of the batch 9 fuel assemblies has been reduced 50 psi to inprove fuel performanon.
This ra+M pre-pressure has been considerad in all mechanical aralysee.
The mechanical evaluation of the fuel rods is d h e W below.
4.2.1.
Claddim Co11 arse Creep collapse anslyses wre perferned for the fcur different fuel batch I
power histories.
Because of its longer previous incore exposure time, the batch 7B fuel is more limitirq than the other batches.
The batch 7B 4-1 l
B&W Fuel Company l
I
assenbly p:ver histories were analyzed and the nest limiting assembly was determined.
'Ihe power history for the nest liraiting assenbly was used to capare with a conservative generic creep collapse analysis.
'Ihe collapse time for the nest limiting assembly was conservatively detemined to be more than 35000 EFHi (effective full power hours), which is greater than the maximum projected residence time (Table 4-1).
'Ihe cruep collapse analysis was perfomed based on the conditions set forth in reference 7.,
4.2.2.
Claddim Stress
'Ihe 'IMI-l stress parameters are enveloped by a conservative fuel rod stress analysis.
For design evaluation, the primary menbrane stress nust be less than two-thirds of the mininum specified unirradiated yield strength, ard all stresses (primary and seoordary) nust be less than the mir inn specified unirradiated yield sLvrgth.
In all cases, the ratvin is in excess of 30%.
'Ihe folicwing ocmtisms with respect to 'IMI-l fuel were used in the analysis:
1.
Izu post-densification internal pressure.
2.
High system pressure.
3.
High themal gradient across the cladding.
I
'Ibe stresses reported in reference 8 for core 1 fuel represent conservative values with respect to the cycle 7 core.
4.2.3.
Claddim Strain
'Ibe fuel design criteria specify a limit of 1.0% on cladding circumferential plastic strain.
'Ibe pellet design is established for plastic claddirq strain of less than 1% at values of maxinum design loca'. pellet burnup and heat generation rate, which are considerably higher than the values the 'IMI-1 fuel is expected to see. 'Ihe strain analysis is also based on the mv4==
specification value for the fuel pellet diameter ard density ard the lowest permitted tolerance for the cladding ID.
4.3.
'Ihernal Desiern All fuel aedlies in the cycle 7 core are therrally similar. 'Ibe design of the batd 9 Mark B2 assenblies is such that the themal performance of this fuel is equivalent to the fuel design uu:d in the remairder of the 4-2 B&W Fuel Company
core.
The analysis for all fuel was performed with the ECO2 code as described in reference 9.
Ncaninal undensified irput parameters used in the analysis are presenteC in Table 4-1.
Densification offacts were accounted for in 2 002.
The results of the thermal design evaluation of the cycle 7 oore are sumarized in Table 4-1.
Cycle 7 oore protection limits are MM on a c
linear heat rate (IER) to centerline fuel melt limit of 20.5 kW/ft as deter-mined by the TACO 2 cede.
The maximm fuel assenbly burnup at IDC 7 is predicted to be less than I
34,000 !H:!/mtU (batch 8A).
The fuel rod internal pressures have been evaluated with TA002 for the highest burnup fuel rods and are predicted to be less than the nominal reactor coolant pressure of 2200 psia.
4.4.
Material Desian The ch=h1 cmpatibility of all possible fuel-claddirg-coolant-asscebly interactions for batch 9A, 9B and 9C fuel assemblies is identical to those for the present fuel.
4.5.
Ceeratim brerience Mhk & Wilcox operating experience with the Mark B 15x15 fuel assembly has verified the adequacy of its design.
As of October 31, 1987, the 4
following experi'ence has been accumlated for eight B&W 177 fuel asserbly plants usirg the Mark B fuel a-bly:
cumlative Current Pax FA bumun,)Mi/rtU(a) net electric (b) outrut.MWh_
3 Reactor Cvele Incore Discharued Oconee 1 10 45,908 50,598 66,183,044 Ooonee 2 9
40,580 41,592 60,968,626 i
Oconee 3 10 33,290 39,701 60,843,663 Three Mile Island 6
26,090 33,444 29,469,976 Arkansas Nuclear 8
51,540 47,560 51,626,035 t
One, Unit 1 Pancho Seco 7
26,242 38,268 39,045,954 Crystal River 3 6
35,350 31,420 38,512,798 Davis-Besse 5
36,960 32,790 25,236,663 (a)As of October 31, 1987.
(b)g, og p, - *+r 31, 1986.
4-3 B&W Fuel Comparty
Table 4-1.
Fuel Desian Parameters and Dbnensions Batd17B Batch 8A Batd1 BB Batch 9A Batd19B Batd19C Fuel assembly MK-B4 E-B4 E-B4 MK-B4 MK-B4 E-B4Z type Number of 25 12 64 15/21 4
36 assemblies Fuel rod O.D.
O.430 0.430 0.430 0.430 0.430 0.430 ncainal, in.
Fbel rod I.D.
0.3 71 0.377 0.377 0.377 0.377 0.377 noininal, in.
Undensified 142.25 141.80 142.25 142.25/141.80 141.80 141.80 active fuel
{
1ergth, in.
Fuel pellet 0.3695 0.3686 0.3695 0.3695/0.3686 0.3686 0.3686 O.D. (Mean), in.
Ebel pellet 94 95 94 94/95 95 95 initial density (Naun.), % T.D.
Initial fuel 2.85 2.95 2.85 2.85 2.95 3.63 3
enrie-ent, Wt. % U235 g
=
g Average Btarnup 15700 16700 14900 0
0 0
Boc, m d/stu Cladding
>35000
>35000
>35000
>35000
>35000
>35000 collapse time, m n Estimated 28500 21200 21200 11000 11000 11000 residence time mM, IDc
Table 4-1.
Mael Desim Paraumeters and Dimansions (Cont'd)
Batd17B Bist d 1 8 A Batdt 88 Batd 9A Batch 9B Batdt 9C Nauninal linear 5.73 5.74 5.73 5.73/5.74 5.74 5.74 heat rate at 2568 PMt, W/ft Average fuel 1399 1400 1399 1399/1400 1400 1400 t# at. ore at nouninal UR (BOL),
or Mininsuun UR to 20.5 20.5 20.5 20.5 20.5 20.5 melt, W/ft a
5m i
1
5.
!UC1 EAR DE:SIGi 5.1.
Ihysics Characteristics Table 5-1 lists the core physics paramters of design cycles 6 ard 7.
We values for both cycles were generated usirg the tooDIE ccde.10 Figure 5-1 illustrates a representative relative pcuer distribution calculated with the ll for the beginnirg of cycle 7 at full power with equilibritu FDQ07 code xenon ard ncniral red positions.
We longer design life ard differences in the shuffle pattern and BPRA 1 cad-irg create the differences in the physics parameters betvoen cycles 6 ard 7.
We BDC critical boron con:entrations for cycle 7 are higher because the additioral reactivity remly for the larger cycle is not ccrpletely offset by the burrable poison.
The differences in control rod worths between cycles are due to charges in radial flux ard burnup distributions.
21s accounts for the smaller stuck ard ejected rod worths in cycle 7 ccrpared to cycle 6 values.
Chleulated ejected rod worths ard their adherence to criteria are considertd at all tires in life ard at all pcuer icvels. in the devolc5 rent of the rod position liraits presentcd in section l
8.
All safety criteria awriated with the control rod worths are not. Se adequacy of the shutdan rargin with cycle 7 stuck red korths is denonstrated in Table 5-2.
We follcuirg cxanservatisms were applied for the s W.n calculaticris:
1.
Poison raterial depletion allcuarce.
2.
10% urcertainty on not rod worth.
3.
Flux rtdistriintion pemity.
Flux rcdistribution kas accounted for since the chutdown aralysis was calcu-latcd usirg a tst>-dinensicml redel. We reference fuel cycle shutdown rar-gin is presented in the MI-1 cycle 6 relcad report.4 l
5-1 B&W Fuel Company L
5.2.
Analvtir-21 Imut The cycle 7 incors
==iw calculation constants to be used for caputiry core power distributions were prepared using the same methods as for the reference cycle.
5.3.
Char== in Nuclaar Desian The core design charges for cycle 7 are the upgrading of the core power to 2568 R, and the in::rease in cycle lifetime to 445 DTD. The calculational methods used to obtain the important nuclear design parameters for this cycle were the same as those used for the referenos cycle.
The operating limits (Technical Specifie:ations chan; pes) for the reload cycle are given in section 8.
5-2 B&W Fuel Company
Table 5-1.
'IMI-1 Cvele 71hysian Par =*mrs(a) cvele 6(b) cyei. 7(c)
Cycle length, DTD 425 460(d)
Cycle tuniup, IRI,/stU 13,129 14,395 Average oors burmp - IDC, Mti/mtU 20,f88 23,112 Initial core loading, atU 82.1 e',1 Critical boren - B3C, ppa (no Xe)
HZP(e), group 8 inserted 1418 1641 HFP, group 8 inserta:1 1253 1515 Critical boren - IDC, syn (eq Xe)
HZP, group 8 inserted 237 218 HFP, grcqp 8 inserted 0
0 Control rod worths - HFP, BDC, % Ak/k Group 7 1.07 1.07 Grcup 8 (=v4==)
0.21 0.20 Centrol rod worths - HFP, IDC, % AP/k Gru p 7 1.10 1.10 Group 8 (=vi==)
0.22 0.21 Max ejected rod worth - }CP, % AP/k(f)
BOC 0.47 0.30 IDC 0.32 0.32 Max stuck red worth - HZP, % Ak,/k BOC 2.31 1.63 IDC 2.17 1.96 Ptwer deficit, HZP to HFP, % AP/k IOC
-1.39
-1.37 IDC
-2.23
-2.28 Doppler coeff - Icc,10-5 (ap-)
1004 power (no Xe)
-1.49
-1.53 Dwler oceff - IDC,10-5 (gyjyjop) 100% power (eq Xe)
-1.84
-1.80 Moderator coeff - HFP,10~4 (AP/P/ F)
C IOC (no Xe,1515 ppn, group 8 ins)
-0.40
-0.06 IDC (eq Xe, 0 ypn, group 8 ins)
-2.64
-2.73 5-3 saw ru s company
Table 5-1.
fCbnt'd)(a) cvele 6(b) cyci. 9(c) noren worth - HFP, gmV4 Ak/k j
Xenon worth - HFP, % Ak/k i
BOC (4 EFPD) 2.60 2.59 IDC (equil.)
2.73 2.73 Effective delayed neutron fraction - HFP 7
IOC 0.00644 0.00638 EDC 0.00524 0.00519
(*) cycle 7 data r for the conditions stated in this report; a
the cycle 6 cv.ve m nitions are identified in reference 4.
(b)Sased on 290 EFIC at 2535 Wt, cycle 5.
(C) Values calculated at 2568 Wt; haeai on 425 DTD at 2535 Wt, cycle 6.
(d)All and of cycle values were calculated at 460 ETPD: the design cycle 7 length is 445 15 DTD.
l
(*)lCP denotes hot zero power (532 F Tavg); HTP denotas h% full 0
0 power (581 F Tavg)*
(f) Ejected rod worth for groups 5 through 8 insertad.
I r
h I
5-4 B&W Fuel Company i
l l
l Table 5-2.
Shutdown Maruin Calculation for 'IMI-1 Cycle 7 l-BOC. b k/k EOC. h P/k(a)
Lvgilable Rod Worth 7btal rtd worth, HZP(b) 8.870 9.501 Werth reduction due to burntp of l
poison raterial
-0.420
-0.420 Maciann stuck rod, HZP
-1.628
-1.956 l
Net worth 6.822 7.125 Less 10% uncertaiity
-0.682
-0.713 Tbtal available worth 6.140 6.412 i
Recuined RM Worth Power deficit, HFP to HZP 1.370 2.282 Max allowable inserted rod worth 0.298 0.460 Flux redistribution 0.239 0.633 Total required worth 1.907 3.375 Shutdoun Maruin
~
7btal available minus total required 4.233 3.037 l
j EZE: Required shutdown margin is 1.00% ak/k.
l (a)460 EFPD.
(b)HZP denotes hot zero power (532 F Tavg); HFP denotes hot full 0
0 power (581 F Tavg)-
l l
l 5-5 B&W Fuel Company
,,,,.,y,
,..----.-rv-, -,, - + - - - -
.---r-.
..-.-.--,y..
Figure 5-1.
B0C (4 EFPD), Cycle 7 Two-Dimensional Relative Power Distribution -- Full Power, Equilibrium Xenon, APSRs Inserted
- 8 9
10 11 12 13 14 15 H
1.07 1.22 1.22 1.23 1.15 1.24 0.80 0.38 p-. -
X 1.17 1.26 1.25 1.25 1.03 1.05 0.40 1.22 1.28 1.15 1.23 0.81 0.32 L
1.26 1.30 1.02 0.91 M
1.11 1.06 0.43 N
0.50 0
P R
- Calculated results from two-dimensional pin-by-pin PDQ07.
Inserted rod group no.
X.XX Relative power density 5-6 B&W Fuel Company
6.
'IHEmAL-HYIFAULIC EESIN
'Ihe thermal-hydraulic design evaluation supporting cycle 7 operation utilized the methods ard models described in references 4, 8,12 and 13 as supplemented by reference 5, which implements the NC (reference 14) CHF correlation for analysis of the Mark BZ fuel a = mbly.
'Ibe analyses presented in Section 5 of reference 5 dercroLr. ate that charges in the flw parameters resulting frm the incorporation of Zircaloy spacer grids do not sagnificantly impact the thermal-hydraulic characteristics of the Mark BZ cc e relative to the Mark B care values. Inplementation of the Mark BZ fuel assemblies into existing reactors, however, is perfcuned on a batch basis, with the transition cycles having both Mark BZ and Mark B fuel ammblies,
'Ibe Mark BZ fuel a=combly has a slightly higher pressure drop than the Mark B assembly due to the higher flow resistance of the Zircaloy grids.
'Ihe presence of Ma::k-BZ fuel a~mblies in a predminantly Mark B core, will, therefore, tend to divert sme flow frm the more restrictive Mark BZ accomblies to the Mark B fuel.
'Ihis creates the need to consider a "transition core penalty".
'1he aucunt of coolant flow reduction and consequently, the magnitude of the transition penalty is dependent on the numbe.r of Mark BZ assemblies (with the smaller number of Mark BZ assemblies being more limiting).
'Ihe core bypass fraction is dependent on the number of unplugged guide tubes, which is in turn dependent on the number of burnable poison reds l
(BPRAs) ard cuhul rod assablies (CRAs), since these cct:ponents restrict flow through the control red guide tubes (CPGrs).
For thermal-hydraulic analysis, the rest limiting case is that with the higher bypass ficw fraction, or s:nller number of BPRAs.
l h design basis chosen for cycle 7 therral-hydraulic analyses was a full Mark BZ core, containing 40 BPPAs, for which the core bypass f1cw is 8.8%.
'Ihis design configuration was used to calculate the 1.77 IUBR value shown on Table 6-1.
h actual cycle 7 core configuration consists of 141 Park B 6-1 B&W Fuel Company
)
l l
a==amblies, 36 Mark BZ fuel eecamblies and 68 BEAs, with a core bypass flow of 7.6%.
'Ihe DE for this configuration, using the same core coniitions presented in Table 6-1, is 1.80.
Cenparison of the DiBRs for the design and actual core configurations shows that the design configuration is conservative for cycle 7 DE analyses and that a transition core penalty is, therefore, not mewy.
'Ihe design core configuration has been used in DE calculations perfortned for tho evaluation of core operating and safety limits and in the evaluation of transient DE margins.
Table 6-1 provides a summary cx:nparison of the Da analysis parameters for cycles 6 ard 7.
'Ihe pressure-temperature safety limits have been recalculated usire the BC CHF correlation in the 12NXF15 crossflow analysis.
'Ihe limits are outside the juncture of the low pressure and high teaperature limits, thus allowing for the elimination of the DiB-based variable low pressure trip setpoint.
No rod bow penalty has been considered in the cycle 7 analysis based on the justification provided by reference 16.
6-2 B&W Fuel Company
~ ~ -,
-,-_~. _ _ _. -. _. _
Table 6-1.
Maxinrn Desian Conditions Cycles 6 and 7 Cycle 6 Cycle 7 2535 MRt 2568 MWt Design power level, MWt(a) 2568 2568 System pressure, psia 2200 2200 Reactor coolant flw, gpn 374880 374880 Core bypass fi m, %(a) 8.4 8.8 Da m, rial 4 ng Crossflow Crossfles Peference design radial-local pcrv.:r peaking factor 1.71 1.71 Reference design axial flux shape 1.65 cosine 1.65 cosinc Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.01 Flow area 0.98 0.97 c)
Active fuel length, in.(D) 141.8 141.8 Avg heat flux at 100% power, 2
103 M -ft 174 174 Vax heat flux at 100% power, 2
103 Btu /h-ft 492 492 QiF correlation B&W-2 BWC 01F correlation DIB limit 1.3 1.18 l
Mininn DE at 112% power 2.08 1.77 (c) at 102% power (d) 2.37 2.01(c)
(a)Used in the analysis.
(b) Cold ncminal stack height.
(c) Calculated for the instnrn?.nt guide tube subchannel which is limitire for the Mark-BZ fuel assemblies.
(d)'Ihis represents initial condition DE for accident analyses.
6-3 B&W Fuel Company
_. ~ _
7.
ACCIDENT AND 'IRANSIDff ANALYSIS 7.1.
General Safety Analysis Each ESAR accident analysis has been examimi with respect to changes in cycle 7 paramters, includiry the power upgrade to 2568 Mft, to determine the effect of the cycle 7 reload and to '.asure that' thermal performarce during hypothetical transients is not degraded.
'Ih2 effects of fuel densification on the ESAR accident results have been evaluated and are reported in referen 8.
Since batch 9 reload fuel accamblies antain fuel rods whose theoretical density is higher than those considered in the reference 8 report, the conclusions in that reference are still valid.
A ocr:parison of the calculated cycle 7 radionuclide source inventory to the cycle 3 inventory confirmed that there are no increases that would significantly increase doses.
Generation of the cycle 7 radionuclides was based on the increase in rated power to 2568 Mft frcxn the circle 6 power of 2535 Mft.
'Ihe radiological dose consequerces of the accidents presented in Chapter 14 of the FSAR were re-evaluated for the cycle 7 reload report.
In order to account for future cycle design variations a 10% factor was applied to the calculated cycle 7 sources for additional conservatism.
In addition, the re-evaluation also incorporated more current plant data and information.
'Ihe result is that the accident doses for cycle 7 are higher than the cycle 6 doses.
All cycle 7 accident rh are well below the dose acceptarre criteria of 10CFR100. 'Ihe cycle 7 analysis is therefore expected to provide a conservative dose evaluation for future cycles.
7.2.
Accident Evaluation
he key parameters that have the greatest effect on determining the outccne of a transient can typically be classified in three major areas:
core 7-1 B&W Fuel Company
themal parameters, themal-hydraulic parameters, aM kinetics parameters, including the reactivity fauYhack ocefficients and control rod worths.
Core thermal properties used in the FSAR accident analysis were design operating values based on calculational values plus uncertainties.
Themal parameters for fuel batches 7, 8, and 9 am given in Table 4-1.
The cycle 7 themal-hydraulic maxirum design corditions are cxmapared with the previous cycle 6 values in Table 6-1.
These parameters are osmon to all the accidents considend in this report.
The key kinetics parameters frcn the FSAR and cycle 7 am ccanpared in Table 7-1.
A generic IIX:A analysis has been performed for the B&W 177-FA lowered loop IGS using the Final Acceptance Criteria ECCS evaluation model; this study is reported in reference 17 ard updated per references 18 and 19.
This e
analysis is generic in nature since the limiting values of the key parameters for all plants in this category were used.
Furthermore, the combination of the average fuel temperature as a function of linear heat rate and the lifetime pin pressure data used in the 10CA limits analysis is conservative ocanpared to those calculated for this reload.
Thus, the analysis and the IDCA limits mported in references 17,18, and 19 provide conservative msults for the operation of IMI-1 cycle 7 fuel.
Table 7-2 shows the bounding values for allowable IDCA peak linear heat rates for IMI-1 cycle 7 fuel.
It is concluded frtn the examination of cycle 7 core themal and kinetics properties, with reg ect to acceptable previous cycle values, that this core reload will not adversely affect the 7MI-1 plant's ability to operate safely durirg cycle 7.
Considering the previcusly aompted design basis used in the FEAR and subsequent cycles, the transient evaluation of cycle 7 is considered to be bounded by previously aampted analyses.
The initial conditions for the transients in cycle 7 are bounded by the FSAR, the fuel densification report, ard/or subsegaent cycle analyses.
7-2 B&W Fuel Company
Table 7-1.
Ca parison of Key Parameters for Accident Analysis FSAR and densification Predicted Parameter retort value value Doppler coeff (BOC), Ak/k/ F
-1.17 x 10-5
-1.53 x 10-5 0
D:ppler coeff (EOC), Ak/k/ F
-1.33 x 10-5
-1.80 x 10-5 0
Moderator coeff (BOC), Ak/k/ F
+0.5 x 10-4
-0.06 x 10-4 0
Moderator coeff (EOC), Ak/k/ F
-3.0 x 10-4
-2.73 x 10-4 0
All-rtxl group worth (HZP), % Ak/k 10.0 8.87 Initial boron concentration (HFP), gra 1200 1515 Boren reactivity worth (HFP),
75 123 ppm /1% A):/k Max. ejected rod worth (HFP), % Ak/k 0.65 0.26 Droppcd rod worth (HFP), % Ak/k 0.46
<0.20 Table 7-2.
Boundirq Values for Allowable IDCA Peak Linear Heat Rates for 'IMI-1 Allcwable Allowable Allowable Core Peak UE, Peak IER, Peak Um Elevation, 0-1000 Mid/mtU, 1000-2600 P.Ad/mtU, after 2600 MM/mtU, ft
}M/ft
)M/ft
}9/ft 2
14.0 (a) 15.0 15.5 4
16.1 16.6 16.6 6
16.5 18.0 18.0 8
17.0 17.0 17.0 10 16.0 16.0 16.0 (a) An allowable peak UR of 14.0 }M/ft is based on FIECSET credit.19 7-3 B&W Fuel Company
8.
PR3IOSED }ODIFICATIGG 'IO TEONICAL SPECIFICATIOtG
'Ihe Technical Specifications have been revised for cycle 7 cperation for charges in guer peaking and control red worths.
'Ibe cycle 7 design l
analysis basis includes a 1cu leakage fuel cycle design, mixed Mark B/ Mark BZ core, gray APSRs, crossflcu analysis, and pcuer level upgrade frun 2535 l
fat to 2568 vat.
'Ihe IOCA linear heat rate limits used to develop the 1
Technical Specification Limitirg Conditions for Operation in::lude the inpact of }URID-0630 claMirg swell and rupture mcdel, and inplement the crtdit frcxn FIECSET analyses.19 A
cycle 7
specific analysis was corducted to generate Technical Specification Limitire conditions for Operation (rod irdex, axial power inbalance, and quadrant tilt), based on the rethodolcgy described in reference 20.
'Ihe effects of gray APSR repositionirg were includcd in the analysis.
'Ihe burnup-dependent allcuable IDCA linear heat rate limits used in the analysis are provided in Figure 8-18.
'Ihe analysis determined the selective loadiry of the batch 9A 2.99 wt% asserbly (see Section 3 for description) in core location K-12 based on rargin to the core therral design criteria and IOCA linear heat rate limits.
'Ihe analysis also determined that the cycle 7 Technical Specifications provide protection for the overpcuer ceniition that could m,m durirg an overocoliry transient because of nuclear instrumentation errors, ard verified that a pcuer level cutoff hold was not required for cycle 7.
Technical Specification section 3.5.2.4 was revised to ac acdate pcuer level-deperdent gaadrant tilt setpoints.
'Ihe analysis verified the acceptability of a full syntetrical incore system quadrant tilt sctroint of 4.12% when therral pcuer is greater than 50%, and a setpoint of 6.83% khen therral pcuer is less than or cqual to 50%.
'Ibe reasurement system-irdependent red position ard axial pcuer irtalarce limits determined by the cycle 7 analysis were error-adjusted to generate alarm setpoints for pcuer operation. 'Ihe error adjusted alarm setpoints are 8-1 B&W Fuel Company
provided in Figures 8-6 throtx3h 8-17.
Revisions to the affected Ibchnical Specifications text sections are being subnitted to the 100 under separate cover.
'Ibe revised pressure-temperature curves that result frun the tue of the BC correlation in the LY1DCP code are shcun in Figures 8-1 ard 8-3.
'Ihe revised protection systan rayimm allcvable setpoints in Figure 8-4 reflect the renoval of the variable lcu pressure trip.
Based on the amlyses and Technical Specification revisions Amibed in this report, the Final Acx:eptance Criteria ECCS limits will not be exceeded, nor will the thermal design critaria be violated.
The follcuing pages contain the revisions to the Technical Sp:cification figures.
1 8-2 B&W Fuel Company
Figure 8-1.
Core Protection Safety Limit 1
2400 2200 ACCEPTABLE e,
OPERATION o.
J
$ 2000 0
UNACCEPTABLE a
((
OPERATION 8
$ 1800 u
1600 580 500 620 640 660 Reactor Outlet Temperature, OF 8-3 B&W Fuel Company
Figure 8-2.
Core Protection Safety Limit?
TMI-I Cycle 7 w
Thermal Power Level, %
I 120
(-43.8,112) 1 (37.8,112)
ACCEPTABL E 4 PUMP
~
OPERATION
- 100
(_ -43.8,89.3) 2 (37.8,89.3)
AC EPT L
(-58.5,80.4) 80 (53.0,80.4)
OPERATION
(-43.8,62.0) 3 (37.8,62.0)
(-58.5,57.8)
ACCEPTABLE 60 (53.0,57.8) 2,3, & 4 PUMP OPERATION 40
(-58.5,30.4) e (53.0,30.4) 20
~
I f
f I
I I
I I
l I
I I
I I
I 70 50 30 10 0
10 20 30 40 50 60 70 80 Axial Power Imbalance, %
1 Curve Reactor Coolant Flow (lb/hr) 6 1
139.8 x 10 6
2 104.5 x 10 6
3 68.8 x 10 8-4 B&W Fuel Company
/\\
l Figure 8-3.
Core Protection Safety Bases 2400 -
2200 1
f
\\
~
3 3
E 3 2000 c.
tc 5
E e
1800 ff 1600 580 600 620 640 660 Reactor Outlet Temperature, OF Reactor Coolant Flow Curve (lbs/hr)
Power PumpsOperating(TypeofLimit) 1 139.8 x 106 (100%)*
112%
Four Pumps (DNBR Limit) 2 104. 5 x 106 (74.7%)
89.4%
Three Pumos (Quality Limit) 3 68.8 x 106 (49.2%)
62.0%
onePumpinEachLoop(QualityLimit)
- 106.5% of Cycle 1 Design Flow 8-5 B&W Fuel Company
Figure 8-4.
Protection System Maximum Allowable Setpoints 2500 P = 2355 psig 2300 T = 618.8 F m
ACCEPTABLE OPERATION 2100 0
E o.
{
P = 1900 psig E
1900 0
UNACCEPTABLE g
0PERATION t
a 1700 1500 i
i i
i 540 560 580 600 620 640 Reactor Outlet Temperature, OF 1
s d
B&W Fuel Company g,g I
Figure 8-5.
Protection System Maximum Allowable Setpoints
~
for Axial Power Imbalance, TMI-1 Cycle 7 Thermal Power Level, %
- - 120
(-30.0,108)
(24.5,108)
ACCEPTABLE I
m = -1.854 m = 1.900 l 4 PUMP 100 i
l 2
l OPERATION l
l l(-30.0,80.6 )
(24.5,80.6 )
ACCEPTABLE 80
(-50.0,70.0)
MP _
(45.0,70.0)
OP RAT I
l(-30.0,53.1 )
6f24.5,53.1) i l
l ACCEPTABLE -
2,3, & 4 (45.0,42.6 )
(-50.0,42.6 )
l PUMP 40 l.
I OPERATION l
l l
l
(-50.0,15.1 )
l 20 m.
9 i
g l
0 (45.0,15.1 )
l u
u u
u
,a
- l E l a
i i
i i ii i
i i
i
, 70 50 30 10 0 10 20 30 40 50 60 70 80
\\
Axial Power Imbalance, %
8-7 B&W Fuel Company
Figure 8-6.
Rod Position Setpoints for Four-Pump Operation From 0 to 40 +10/-0 EFPD, TMI-1 Cycle 7 I
(300,102)
(92.5,102)
(275.9,102) 100 SHUTDOWN
( 7.5,90) 90 MARGIN
~
NOT ALLOWED LIMIT 80 (249.5,78) 70 RESTRICTED a
E 60
~
~
50 (38.5,48)
(201.5,48) 5-40 E
30 l
l 3
I j
20 PERMISSIBLE i
(0,11.5)
~
10 ' -
0
( 0,,2. 6 )
~
0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0
25 50 75 100 l
l I
i Group 7 0
25 50 75 100 l
i I
I I
Group 6 0
25 50 75 100 i
f f
f 1
Group 5 8-8 B&W Fuel Company
l Figure 8-7.
Rod Position Setpoints for Four-Pump Operation From 40 +10/-0 to 100 +10/-0 EFPD, TMI-1 Cycle 7 (300,102)
(92.5.102) 100 (270.1.102) i_
90 SHUTOOWN (265.5,90)
MARGIN NOT ALLOWED LIMIT 80 (249.5,78) 70 E
RESTRICTED g
60 N
50 (38.5,48)
(201.5,48) 40 g
E 30 z
3 20 PERMISSIBLE (0,11.5)
~
10 ' -
0,2.6) 0 0
25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0
25 50 75 100 1
I I
I I
Group 7 0
25 50 75 100 1
I l
l l
Group 6 0
25 50 75 100 l
l i
I I
Group 5 8-9 B&W Fuel Company
Figure 8-8.
Rod Position Setpoints for Four-Pump Operation After 100 +10/-0 EFPD, TMI-1 Cycle 7 (300,102)
I9*'
)
(266.5,102) 100 90
~
SHUTDOWN MARGIN LIMIT (249.5,78) y 73 RESTRICTED NOT ALLOWED g
N 60
~
(116.5,48)
(201.5,48) 40 c.
t 30 3
3 s
j 20 PERMISSIBLE 10 (58.5,13)
I,0,2. 7 ),
0 0
25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn l
0 25 50 75 100 l
l 1
l I
l l
Group 7 0
25 50 75 100 I
i l
_uroup b 0
25 50 75 100 I
I I
l l
Group 5 E-10 B&W Fuel Company i
im
I Fi gure 8-9.
Rod Position Setpoints for Three Pump Operation From 0 to 40 +10/-0 EFPD, TMI-1 Cycle 7 100 90 (300,77) 80 (93.2,77) r 70 (273.5,67) 5 NOT ALLOWED SHUTDOWN MARGIN 60 y
LIMIT (249.5,58) 50 RESTRICTED 40 h
(38.5,36)
(201.5,35.5) 2 30
?g 20
.2 3
10,
(0.8.6)
PERMISSIBLE O<
(O'l f) i 0
25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0
25 50 75 100 i
f I
l i
Group 1 0
25 50 75 100 t
I t
t I
Group 6 0
25 50 75 100 t
f f
f 1
Group 5 8-11 B&W Fuel Company
Figure 8-10. Rod Position Setpoints for Three Pump Operation From 40 +10/-0 to 100 +10/-0 EFPD, TMI-1 Cycle 7 I
100 90 (300,77)
~
(93.2,77)
(270.3.77)-
( 66.5,67)
~
h NOT ALLOWED SHUTDOWN MARGIN e
60 h
LIMIT (249.5,58) t 50 RESTRICTED f
a f
40 3
(33.5,36)
(201.5,35.5)
[
30 z
20 3
PERMISSIBLE (0,8.6) 0, 0,1.4) 0<
0 25 50 75 100 125 150 175 200 225 250 275 300 l
Indicated Rod Index, % Withdrawn O
25 50 75 100 l
l f
f I
Group 7 0
25 50 75 100 I
I f
I l
Group 6 l
0 25 50 75 100 l
i I
I i
Group 5 8-12 B&W Fuel Company L..
l Figure 8-11.
Rod Position Setpoints for Three Pump Operation After 100 +10/-0 EFPD, TMI-1 Cycle 7 100 90 (300,77) 80 (198.5,77)
(266.5,77)
SHUTDOWN (266.5,67)(
70 h
140T ALLOWED R N 7
y 60 (249.5,58) ed RESTRICTED g
50 a
40 g
(116.5,36)
(201.5.35.5)
S 30 3 20 E
E 10 (58.5.9.7)
PERMISSIBLE 0,
(p. 2.0),
0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0
25 50 75 100 t
I i
I l
Group 7 0
25 C3 75 100 1
I I
I i
0 25 50 75 100 i
t i
l I
Group 5 9
8-13 B&W Fuel Company
c' 4
Figure 8-12.
Rod Position Setpoints for Two-Pump Operation From 0 to 40 +10/-0 EFPD, TMI-1 Cycle 7
- g 100 90 80 70 SHUTDOWN 60 MARGIN (300,52)
LIMIT
[
50 (94.5,52)
(276.2,52)-
o NOT ALLOWED (273.5,44) 40 b
(249.5,38) g RESTRICTED c.
30 f
20
(
5'
)
(201.5,23)
E 5
10 PERMISSIBLE 0.5.7)
(0,0.3) 0,
i i
i e
0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn O
25 50 75 100 l
l l
t l
l Group 7 l
0 25 50 75 100 t
1 I
I I
Group 6 0
25 50 75 100 i
f f
I I
Group 5 8-14 B&W Fuel Company
Figure 8-13.
Rod Position Setpoints for Two-Pump Operation From 40 +10/-0 to 100 +10/-0 EFPD, TMI-1 Cycle 1 100 90 80 70 a
E SHUTDOWN 60 E
(270.5.52)(-300,52)
W
'IM (94.5,52) 50 D
NOT ALLOWED (266.5,44) 40 g
RESTRICTED (249.5,38)
E
?
30 k
20 (38.5.24)
(201.5.23) 3
.5 10 PERMISSIBLE
_(0,5.7)
(0,0.3) '
0' i
i e
i i
t i
0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0
25 50 75 100 1
I f
f I
Group 7 0
25 50 75 100 l
I i
i i
Group 6 0
25 50 75 100 I
f f
f 1
Group 5 l
8-15 B&W Fuel Company
Figure 8-14.
Rod Position Setpoints for Two-Pump Operation After 100 +10/-0 EFPD, TMI-1 Cycle 7 m
100 90 80 y
70 5
g) 60 (266.5,52)(300,52) t (200.5,52) 50 SHUTDOWN (266.5,44),
t NOT ALLOWED MARGIN l
40 LIMIT RESTRICTED (249.5,38) c 30 3
.5!
I
- 4)
(201.5.23) 20
~
lE 10 PERMISSIBLE (0,1.3)
(58.5,6.5) 0' i
i i
e i
e i
e i
e i
0 25 50 75 100 125 150 175 200 225 250 275 300 l
Indicated Rod Index, % Withdrawn l
0 25 50 75 100 q,
I f
f I
I Group 7 0
25 50 75 100 i
e i
i i
Group 6 0
25 50 75 100 i
e i
f i
Group 5 8-16 B&W Fuel Company
~
Figure 8-15.
Axial Power Imbalance Enve'. ope for Operation From 0 to 40 +10/-0 EFPD, TMI-1 Cycle 7 Indicated Power, % of 2568 MWt
. 110
(-13.9,102)
T(19.7,102)
-100
(-14.1,92)
<(19.7,92)
- 90
(-22.0,80)<
80
, (25.8,80)
- 70
- 60 RESTRICTED PERMISSIBLE RESTRICTED REGION OPERATING REGION
- 50 REGION 40
. 30
- 20 10 l
I I
f I
f I
l f
I
-50 30 -20
-10 0
10 20 30 40 50 Indicated Axial Power Imbalance, %
8-17 B&W Fuel Company
Figure 8-16.
Axial Power Imbalance Envelope for Operation From 40 +10/-0 to 100 +10/-0 EFPD, TMI-1 Cycle 7
.)
Indicated Power, % of 2568 MWt
-- 110
(-19.7,102)
(21.6,102)
~
(-20.8,92)
(21.8,92)
(-24.9,80)<
80
,(27.8,80) 70 l
RESTRICTED PERMISSIBLE -
60 RESTRICTED REGION OPERATING REGION REGION 50 40 30 1
I 20 10 I
t I
f f
I t
i 1
l
-50 30 10 0
10 20 30 40 50 Indicated Axial Power Imbalance, %
8-18 B&W Fuel Company
J Figure 8-17.
Axial Power Imbalance Envelope for Operation Af ter 100 +10/-0 EFPD, Ti11-1 Cycle 7 Indicated Power, % of 2568 MWt
- 110
(-22.6.102)
- 100
(-22.8,92)
(22.8,92) 90
(-27.8,80)<
80 i(28.7,80) 70
- 60 RESTRICTED PERJIISSIBLE RESTRICTED REGION OPERATING REGION
- 50 REGION
. 40
- 30
- 20 10 I
I l
i I
I I
I I
l 40 -30 10 0
10 20 30 40 50 Indicated Axial Power Imbalance, %
8-19 B&W Fuel Company
Figure 8-18.
LOCA Limited Maximum Allowable Linear Heat Rate TMI-1, Cycle 7 l
20 s
a 3
18 e
5 f./*/*
~
1 k
16
/
8
//
/
/
/
l I
14 3
- E E
E "x
- 2 12 0-1000 mwd /mtU 1000-2600 mwd /mtU After 2600 ftWd/mtU 10 I
t i
e i
0 2
4 6
8 10 12 Axial Location From Bottom of Core, ft, 8-20 B&W Fuel Company c
9.
SDRIUP HOGRAM - HH5ICS TESTING
'Ihe planned startup test pswtaru associated with core performance is outlined below.
'Ihese tests verify that oore performance is within the assucptions of the safety analysis ard provide information for ocntinued safe operation of the unit.
9.1.
Precritical 'ntsts 9.1.1.
Control Rod Trio ' Nest Precritical control rod drop times are recorded for all control rods at hot full-flow corditicos before zero pow.r physics testing begins.
Acceptance criteria state that the red drop time frm fully withdrawn to 75% inserted shall be less than 1.66 seconis at the conditions above.
It should be noted that safety analysis calculat!.ns are based on a rod drop from fully withdrawn to two-thirds inserted.
Since the most acx: urate position indication is obtained frm the zone reference switch at the 75%-inserted position, this position is used instead of the two-thirds inserted position for data gathering.
9.1.2.
RC Flow Reactor coolant flow with four RC pumps running will be measured at hot shutdown oorditions.
Acceptance criteria require that the measured flow be within allowable limits.
9-1 B&W Fuel Company
9.2.
Zero Power Ihysics hts 9.2.1.
Critical Boron COirmitaation once initial criticality is achieved, equilibrium boron is obtained and the critical boren concentration determined.
'Ibe critical boren concentration is calculated by correctire for any rod withdrawal required to achieve equilibrium boren.
'Ibe acceptance criterion placed on critical boron concentratico is that the actual boron concentration nust be within 100 ppn boron of the predicted value.
9.2.2.
Tencerature Reactivity Coefficient
'Ibe isothermal HZP te@erature coefficient is measured at approximately the all-rods-cut configuration.
During charges in te@erature, reactivity feedback may be - W sated by control rod movement.
'Ibe change in reactivity is then calculated by the su:nmation of reactivity (obtained frcra f
a reactivity calculator strip chart recorder) associated with the te@erature charge.
Acceptance criteria state that the measured value shall not differ frcra the predicted value by more than 0.4x10~4 AW F.
O
'Ihe nederator coefficient of reactivity is calculated in conjunction with the temperature coefficient neasurer:ent. After the tem erature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is added to obtain the rederator coefficient.
'Ihis value nust not be in e e n of the acceptance criteria limit of +0.5x10-4 AM.
9.2.3.
Control Rod Group / Boron Peactivity Worth control rod group reactivity worths (groups 5, 6, and 7) are measured at hot zero power corviitions using the boron / red swap method.
'Ihis technique consists of establishirg a deboration rate in the reactor coolant system and cunpensatirq for the reactivity charges frcra this deboration by irsertirg control rod grcups 7, 6, and 5 in incremental steps. 'Ihe reactivity charges that n'9n-during these taeasurements are calculated based on reactimeter data, and differential rod worths are obtained frcra the measured reactivity worth versus the change in rod group position. 'Ihe differential rod worths of each of the controlling groups are then sumed to obtain integral red group worths.
'Ibe acceptance criteria for the aantrol bank group worths are as follows:
l l
l 9-2 B&W Fuel Company l
1.
Individual bank 5, 6, 7 worth:
nrrdie*=d value - maa-mid valge I
x 100 $ 15 gy, 1
1 2.
Suns of groups 5, 6, and 7:
gradic*ad value -====>M value1 measured value l x W < 10
'Ihe boren reactivity worth (differential borm worth) is measured by i
dividing the total insertad red worth by the boren chanye made for the red worth test.
'Ihe acceptance criterim for measured differential boren worth is as follows:
1 1.
oradicted value -====itud value x 100 $ 15 measured value
'Ihe prailcted rod worths and differential boren worth are taken frun the PIM.
9.3.
Power Fm alation 'Insts Y
9.3.1.
Core Sv=aatry 'hnst
'Iha purpose of this test is to evaluate the sysseetry of the core at low power during the initial power escalation following a refueling.
Symnetry evaluation is based on j ' ore quadrant power tilts during escalation to the iMamadi_ ate power level.
'Ihe core symetry is acceptable if the absolute values of the quadrant power tilts are less than the error adjusted alarm limit.
9.3.2.
Core Power Distribution Verification at Imamadlata Power Invel (IPU and 100% FP With Neninal Control Red Position Core power distribution tests are performed at the IPL and 100% full power (FP).
Equilibrium xenon is established prior to both the IPL and 100% FP tests.
'Ibe test at the IPL is essentially a check on power distribution in i
the core to identify any abnormalities before aaralating to the 100% FP plateau.
Peaking factor criteria are applied to the IPL oore power distributim results to determine if additional tests or analyses are required prior to 100% FP operation.
'Ihe following acceptance critaria are placed on the IPL and 100% FP tauts:
1.
'Ihe worst-casa nav4== IHR mast be lecs than the IOCA limit.
9-3 B&W Fuel Company
2.
'1he mininum DM nust be greater than the initial condition DE limit.
3.
h value obtained frcan extrapolation of the mininum DE to the next power plateau overpwer trip setpoint nust be greater than the initial condition DM limit, or the extrapolated value of imbalance nust fall outside the RPS power /irbala:ce/ flow trip envelope.
4.
h value cbtained frtan extrapolation of the worst-case maxinum IER to the next power plateau overpower trip setpoint nust be less than the fuel melt limit, or the extrapolated value of imbalance must fall outside the RPS power /imbalan:e/flw trip envelope.
5.
'Ibe quadrant power tilt shall not eM the limits specified in the Technical Specifications.
6.
'Ihe highest reasured and predicted radial peaks shall be within the follwing limits:
credicted value - neasured value x 100 more positive than -5 reasured value 7.
h highest neasured and predicted total peaks shall be within the following 1imits credicted value - measured value 100 nore positive than -7.5 measured value Itens 1, 2, and 5 ensure that the safety limits are raintained at the IPL and 100 %FP.
Items 3 and 4 establish the criteria shereby m1ation to full power ray be acccr:plished without the potential for exrvwHng the safety limit.a at the overpwer trip setpoint with regard to DE and linear heat rate.
Items 6 and 7 are established to detemine if measured and prailcted power distributions are consistent.
9.3.3.
Incore Vs. Excore Detector Inbalance Correlation Verification at the IPL_
Imbalances, set up in the core by control rod positioning, are read sinultaneously on the incere detectors and excore power range detectors.
'Ihe excore detector offset versus incere deta: tor offset slope rust be greater than 0.95.
If this criterion is not met, gain anplifiers on the 9-4 B&W Fuel Company
exoore detector signal processing equipment are adjusted to provide the required gain.
9.3.4.
'Immerature Reactivity Coefficient at 100% FP
'Ibe average reactor coolant tarperature is decraad ard then increased by about 5 F at constant reactor power.
'Ibe reactivity a-iated with each 0
tarperature change is obtained frun the charge in the controllirq red gluxp positicn.
Controlling red group worth is measured by the fast insert / withdraw method.
'Ibe teraperature reactivity coefficient is calculated frun the measared charges in reactivity ard tenperature.
Acceptance criteria state that the moderator tenperature coefficient shall l
t be negative.
9.3.5.
Power Doooler Reactivity Cbefficient;_at 100% FP
'Ihe power Doppler reactivity coefficient is calculated frun data reccItied during control Itxi worth measurements at power usirxJ the fast insert / withdraw method.
'Ihe fuel Doppler reactivity coefficier.t is calculated in conjunction with the power Doppler coefficient neasurenent. 'Ibe power Domler coefficient as measured above is cultiplied by a precalculated conversion factor to obtain the fuel Doppler coefficient.
'Ihis measured fuel Doppler coefficient rust be rcre negative than the acceptara) criteria limit of -0.90 x 10-5 gyp.A, 14.
Procedure for Use if Acceptance Criteria Not Met If acceptarce criteria for any test are not met, an evaluation is performed before the test prtxJram is continued.
Further specific actions depend on evaluation results. 'Ihese actions can irclude repeating the tests with nore detailed attention to test preregaisites, added tests to search for arxxnalies, or design personnel perfornirg detailed analyses of potential safety problems because of parameter deviation.
Power is not m1ated until evaluation shcus that plant safety will not be cuw JM by such e lation.
J 9-5 B&W Fuel Company
- 10. REFEREN3S 1.
Bree Mile Islard Nuclear Station, Unit 1, Final Safety Analysis Recoit, USNRC Docket No. 50-289.
2.
BPRA Retainer Design Report, BAW-1496, hW4 & Wilcox, Dfniburg, VA, May 1978.
3.
J.H.
Taylor (B&W) to S.A.
Varga (NRC),
letter "BPRA Retainer Reinsertion," January 14, 1980.
4.
'Ihree Mile Islard Unit 1, Cycle 6 Reload Report, BAW-1977, hWk &
Wilcxx, Dfndburg, Virginia, October 1986.
5.
Rancho Seco Cycle 7 Reload Report - Volume 1 - Mark BZ Fuel Assembly Design Report, BAW-1781P, hWk & Wilcox, Dfnchburg, Virginia, April 1983.
6.
Rancho Seco Nuclear Generating Staticn - Evaluation of Mark BZ Ebel Amammbly Design, U.S. Nuclear Regulatory hk= ion, Washington, D.C.,
November 16, 1984.
7.
F1.y m to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084P. Rev.
2, hWk & Wilcox, Djntburg, VA, October 1978.
8.
'Ihree Mile Islard Unit 1 Fuel Densification Report, BAW-1389, mWk &
Wiloax, Dfnchburg, Virginia, June 1973.
9.
TACO 2: Fuel Performarce Analysis, BAW-10141P-A. Rev. 1, hWk &
Wilcox, Dfnchburg, Virginia, June 1983.
- 10. N0002 - A Multi-Dimensional 'No-Group Reactor Sim21ator, BAW-10152A, hWk & Wilcox, Dfnchburg, Virginia, June 1985.
- 11. *Wk & Wilcox Versicn of PDQ User's Harml, BAW-10117P-A, B'Wk &
Wilcox, Dfnchburg, Virginia, Jaruary 1977.
- 12. Final Safety Analysis Report (Updated Version), 'Ihree Mile biland Nuclear Station - Unit 1, USNRC Dxket No. 50-289.
l
- 13. 'Ibermal Hydraulic crossflow A;plications, BAW-1829, hWk & Wilcox, Dfnchburg, Virginia, April 1984.
10-1 B&W Fuel Company r
- _ _ ~ _... _.,.. _.. ~ _ _ _ _.
- 14. IMC Cbrrelation of Critical Heat Flux, BAW-10143P-A, mWk & Wilcox, Lyniburg, Virginia, April 1985.
- 15. I2NXI: Core Transient 'Ihermal Hydraulic ProTram, BAW-10156-A, mWk &
Wilcox, Dfnchburg, Virginia, February 1986.
- 16. Ebel Red Bowing in mWk & Wilcox Fuel Design, BAW-10147P-A, Rev.1, mhk & Wilcx*, Lyrrhburg, Virginia, May 1983.
3, mWk & Wilcox, Lyniburg, Virginia, July 1977.
- 18. TACD2 Ioss-of-Coolant Accident Limit Analysis for 177-FA Iowered loop
- Plants, BAW-1775. Rev.
O, mWk & Wilcox, Dfnchburg, Virginia, February 1983.
- 19. Bounding Analytical Aewwnt of NURD3-0630 Models on LOCA kW/ft Limits with Use of FIICSET, BAW-1915P, mWk & Wilcox, D/nchturg, Virginia, May 1986.
- 20. Normal Operating Controls, BAW-10122-A. Rev.
1, mWk & Wilcox, Lyniburg, Virginia, May 1984.
i l
i 1&2 B&W Fuel Company
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