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Regulatory Docket File pgp;.a,Ur Day 2 -{@ | |||
NORTliERN 5W.'tt.S POWER CCMPAln MCNTICELLO NUCLEAR GENE 9A'11hr PLANT l | |||
SECOND RELGAD SUBMITTAL REVISION 1 FEBRUARY 8, 1974 Prepared By Information Supplied By Northern States General Electric Power Company Company 9105070342 740200 PDR ADOCK 05000263 P PDR | |||
i Table of Co. :a $ | |||
p | |||
.yagg. | |||
: 1. Introduction 11 I 2. Sunnary 2-1 1 | |||
: 3. Mechanical Design 31 3.1 General Design Description 3+ 1 | |||
' 3.2 Hechanical Design Bases 3$ | |||
3.3 Results from Mechanical tiesigu Evalestions 1 15 | |||
; 3.4 fuel Operating and Developmental Exptrience 3-17 1 Referencca Section .1 3 V6j j | |||
i 4. Thermal Hydraulic Charact eristica 41 4.1 ruel Assembly Hydraulic Anolitis 4*l l | |||
: 4.2 Fuel Assembly fliersnal-Hedt aulic Evaluation 4 '. | |||
4.3 Results of Thermal-Hydraulic .inalysis 45 References - Section 4 48 l - | |||
i | |||
$. Nucler.r Characteristics 51 l' 5.1 Introduction 5-1 a 5.2 Bundle Nuclear Dt.scription 5-1 5.3 Analytical Hethode 5-11 1 5.4 Experience with GE Nuclear !bdels 5-15 a 5.5 Nuclear Characteristics of the Core 5-16 References - Section 5 5 5 19 i | |||
6, Safety Analysis 6-1 6.1 Model Applicability to 8x8 Fuel 6-1 6.2 Results of Safety Analysis 66 References - Section 6 6-66 i 7. Technical Specifications 7-1 t | |||
i l | |||
ilt t | |||
r | |||
- ~ . . _ _ _ . _ _ _ _ | |||
_ _ _ . _ . . _ __ _ _ Revised 2/8/74 | |||
* 1 l | |||
l I | |||
J List OF ILLUSTRATIONS rigure Title h ! | |||
21 Reference Core loading tbnticello Cycle 3 2-2 31 8x8 Reload Fuel Assembly 32 . | |||
32 tbnticello R2 Reload Puel Lattice 33 l 33 8x8 Power Spike Versus Cap Size - Normal Operation 3 26a 34 9x8 Power Spike Versus Cap Sire - Cold Zero void condition 3 26b 3-5 Percent of Rods in Core with X t Peak KW/FT (X a X, 4X=.2 3 26d l 3-6 8x8 Power Spike Penalty vs Axial Position - Normal Operation 3 26e | |||
+ | |||
Cladding Average Temperature at A Puel Column Axial Cap 3 26g 37 38 Clad Critical Collapse Pressure Ratio Versus Time 8x8 Reload fuel 3-26h 5-1 llo t Average Void Infinite Lattice K versus gg Exposure 5-3 5-2 2.63 7x7, 6.62 8x8 Infinite Lattice Kag versus In channel 5-6 , | |||
Void Fraction Zero Exposure 5-3 AK void Comparison 7x7 versus 8x8 from 0.40 Void to other Voids 5-7 5-4 200 FMd/t Doppler coefficients Uncontrolled 5-9 5-5 10,000 tMd/t Doppler coef ficient s Uncontrolled Solo 5-6 Sversus Exposure Comparison 2.62 wt'. 8x8, 7.63 wt% 7x7 5-12 Average Voids, Uncontrolled i | |||
57 tiaximum Local Peaking versus Exposure Comparison 5-13 l | |||
6-1 Doppler Reactivity Coef ficient versus Average Fuel 6-13 Temperature as a Punction of Exposure and Moderator Condition , | |||
6-2 Accident Reactivity Shape Functions for Cold Startup 6-14 jf= 0.0054 6-3 Scram Reactivity Function for Cold Startup 6-15 l | |||
6-4 Cladding Temperature versus Time for the Recirculation 6-20 , | |||
Line Break with railure of the LPCI Injection Valve (llPCI + 2CS + ADS) AEC Assumptions 2 | |||
6-5 Performance of ECCS with railure of 11PCI for Small (0.02 f t ) 6-22 l | |||
l Liquid Break (2CS + 4LPCI + ADS) AEC Assumptions 6-6 6-23 Cladding Temperature vergus Time for a Small Break with Failure of IIPCI (0.02 ft Break) (4LPCI + 2CS + ADS) t i | |||
1 iv ' | |||
Revised 2/8/74 | |||
_ - ___ , _ _ . - _ _ _ . _ _ . _ . . _ . . . _ _ - - _ . . ~ _ _ _ , - _ _ _ _ _ . _ _ | |||
LIST OF !!11tSTRAT10NR (Continued) 112ME1 Title lage 67 Peak Cladding Temperature Spectrum for Single Failure 6-24 f Conditions AEC Assumptions 6-8 Emergency Core Cooling System Performance Capability 6 26 j 6-29 h 6-9 Performance of ECCS for Fbin Steam Line Break Inside the Drywell with all ECCS Operating (HPCS + 2CS + 4LPCI + ADS) | |||
AEC Assumptions 6 10 Core flow and Pressure rollowing a Recirculation Line Break 6 30 6-11 Performance of ECCS with Failure of LPCI Injection Valve for 6-32 ; | |||
the Design Basis Recirculation Line Dreak (HPCS + 2CS + ADS) I AEC Assumptions 1 | |||
6-12 Minimum Critical Heat Flux Ratio for DBA at Honticello 6 13 | |||
? | |||
6 13 Cladding Temperature versus jime for an Intermediate Break 6-35 with Failure of HPCI (0,1 ft Break) (4LPCI + 2CS + ADS) | |||
AEC Assumptions 6-14 Quality versus Time for DBA at Monticello 6-36 ! | |||
6-15 Heat Transfer coef ficient for a small Break (0.02 f t2) 6-37 (4LPCI + 2CS + ADS) AEC Assumptions 6-16 Heat Transfer Coefficient for an Intermediate Break 6-38 ! | |||
(0.1 ft2) (4LPCI + 2CS + ADS) AEC Assumptions (, | |||
6-17 tbnticello Heat Transfer coef ficients for DnA with LPCI 6-39 Injection Valve Tailure - AEC Analysis No. 1 (2CS + HPCI + ADS) 6 18 Performqnce of ECCS with Failure of HPCI for Intermediate 6-40 (0.1 f t') Liquid Break (2CS + 4LPCI + ADS) AEC Assumptions . | |||
6-19 Power Cen= ration Following a Design Basis Recirculation Line 6-41 i Break Accident i | |||
: 6-20 Ef fects of Exposure on Maximum Cladding Temperature 6 43 $ | |||
; 6-21 Puel Rod Perforation Data 6 44 l 6-22 Distribution of Internal Pressure Within Rods 6-45 | |||
! 6 23 Percent Rod Perforation versus Break Area (AEC Assumptions) 6 46 ' | |||
6 24 8x8 Reload Fuel Rod Identification 6 47 i | |||
6-24a Fbnticello Reload 2 (8x8) CECAP III MAPLHCR and PCT vs Exposure 6 48b l 6-25 Puel Type Locations 6 52 l | |||
l | |||
[ | |||
v , | |||
Revised 2/8/74 | |||
. - - - - . - . . . - , - - . . ~ | |||
LIST Or I!LUSTRATIONS (Continued) rigurs Title Pans Base Rod Pattern for RWE 6 53 6 26 6 27 tbnticello Composite HCliFR versus Rod Position Limiting 6-$4 Rod Withdrawal 6 26 MCHTR versus Rod Position Limiting Rod Withdrawal 7x7 6-5$ | |||
6 29 MCHER versus Rod Position Limiting Rod Withdrawal 8x8 6-56 6-30 Peak liest Flux and Relative Power versus Rod Position 6-57 Limiting Rod Withdrawal !boticello DOC 3 6-31 RHM Response - Channels A and C 6 58 6-32 RBM Response - Channels B and D 6-59 6-33 Scram Reactivity curve 6-62 6-34 Control Rod Drive Scram Times 6-63 l | |||
l l | |||
1 1 | |||
vi Revised 2/0/74 | |||
: 1. I_liTRODUCTION This docuinent provides the technical basis of the license submittal for the second reload of Northern States Pcuer/Monticello l' nit 1. Presented herein is a description of the new fuel and the results of the evaluation of the refueled core for the February,1974 outage. | |||
The fuel at ttis site available for loading at the outage will be 116 Reload-2 fuel bundles, which are 8x8 bundles with an average enrichment of 2.62 wet of U-235. There will also be availabic i .r reinsert 7 initial core bundles discharged at the end of Cycle 1 with an average exposure of about 7400 mwd /t. | |||
The objective of this outage is to remove the 44 temporary control cur-taina which remained in the core during Cycle 2 and to load the core so as to assure the availability of the plant at high power for approximataly an annual cycle. | |||
Sections 3, 4, 5 and 6 of this document, dealing with the subjects of reload fuel mechanical design and reloaded core thermal-hydraulic and nuclear charac-teristics, and safety analysis, present descriptiore of design criteria, methods and results from design calculatior.s and safety evaluations and represents com-plete information for the review of fuel assenbly and core design. | |||
1-1/1-2 Revised 2/8/74 | |||
A more detailed summary of Cencral Electric experience with BWR Zi rc a lo y- c l ad P02 pellet fuel, inc luding recent production and development data, has been documented (see Reference 3) . | |||
3.4.4 Analysis of Fuel Densi fication Ef f ects This section presents results of the ef fects of densification in the 8x8 reload fuel as determined from application of the models described in Reference 7. | |||
3.4.4.1 Power Spiking Analysis 8x8 Puel Lattice An analysis of potential local power spikes due to axial gaps in fuel pellet column for GE IMR's employing an Sx8 fuel lattice design has been performed. This analysis employs the same method and basic assumptions that were reported in Reference 7. Important aspects of this analysis are noted as follows: | |||
: 1) The equation employed to calculate maximum gap size is that noted in Reference 7: | |||
. 965 - f4 + 0.0025 L g , | |||
2 _ | |||
where L = maximum axial gap length L = fuel column length fi = mean value of measured initial pellet density (immersion - SV) 0.0025 = allowance for irradiation induced cladding growth and axial strain caused by fuel-clad mechanical interaction | |||
: 2) The magnitude of the power spike versus gap size for fuel rods of the 8x8 design is shown in Figure 3-3 for normal opercting conditions, and Figure 3-4 for cold zero void conditions. | |||
3-26 i Revised 2/8/74 I | |||
FIGURE 3-3 - | |||
I 8x8 PO'4ER SPIKE VERSUS GAP SIZE - UOPRAL OPERATION . | |||
a | |||
* i i . | |||
* E3D RELATIO15 HIPS I | |||
-- 7.0 _ | |||
w y __ . 6,0 . . .. -. . _ _ . - . . _ . . _ _ _ _ ___ -_.- - - - | |||
~ ~ - - ~ ' | |||
p 5.0 - - - - - | |||
' _ . . . . . _ _ _ _ . . - - - - -- - - - The effect of a gap in Ecd 1 | |||
, jr ' | |||
on adjacent pellet i. | |||
_1_,0 n. . . . | |||
~~ - | |||
t- Effect of gap in Rod 2 on Red 1 | |||
_3 .0 ... _ .__ _ _ _ . . _ . . . . _ _ . _ _ . . . _ . . _ | |||
- ~ ~/ p .- | |||
/ ,, . . ' | |||
/ ---- | |||
Effect of gap in Rod 3 on Rod 1 2 0 -' ~' - - ~* ~ - ~~--'[-~' ---- | |||
i .- | |||
Effect of gap in Rod 4 cn Rod 1 | |||
'(; l.0 ~ : | |||
~ ~ ~ | |||
/ | |||
^ ^ ~' . . , | |||
'y ~ | |||
=_.. | |||
Effect of gap in Rod 5 on Rod 1 | |||
. , / ,., - | |||
.im> 4 .. *. | |||
} | |||
2.0 3.0 4.0 5.0 6.0 70 - - | |||
1.0 1 | |||
; ~R . . ._ | |||
r U . . _. ._ _ . _ _ _ . . _ _ . _ _ . _ | |||
! _--_-.__-_.-._s-._ - . . . . . . | |||
4 y | |||
1 2 | |||
s FIGURE 3 4 . . . . ._ . .._. | |||
8x8 PO*4ER SPIKE VERSUS GAP SIZE - COLD ZERO VOID CONDITION . | |||
R03 RELATIONSHIPS 4 | |||
-.. .--- . .. . . . . - - . . ~ - -. | |||
..- Effect of a gap in Rod el on | |||
~ | |||
p adjacent pellet LP/P , __ _, _, __ | |||
g,____ _ _ _ _ , , | |||
w 7 i. | |||
, h c; - -- | |||
/ | |||
) @- ..-..--./. | |||
6 - . . - . . - _ _ | |||
.--- - - - ,. ._ Effect of a gap in Rod e2 | |||
/ - ~ - | |||
on Rod dl j ,- , | |||
5 - -- | |||
7' - -- : | |||
f- - - - -- - - - - | |||
j | |||
/ -- - - - /- --f._ _ - _ . .. . _ | |||
j .....g , --_-....-.,/._----------..-. -- | |||
r ,- | |||
/ | |||
/ .. . | |||
i -- - 3 - | |||
- - 7 '- - - - - - - - - | |||
- - ; -- Effect of a gap in Rod e3 on i / . | |||
.__., . Rod #1 e ' | |||
: x - | |||
2 - . - | |||
a . , | |||
l < .' s . ,s"', | |||
._ _ _ ' - _ _ . <[- - Effect of a gap in Rod *4 on Rod l {, . | |||
1 . 1. - ' | |||
. _ _ _--_2--__ - | |||
;. _-a . | |||
Effect of a gap in Red -5 on Rod | |||
. u .. ' ~ - __ -L_ - -- - . | |||
) | |||
.i - - _ _ | |||
, , '.*./ - _ | |||
M S* | |||
1 2 3 4 5 6 | |||
~ | |||
~ ~~ | |||
- ~~ ~~'-~~~~~ ~ ~ | |||
GAP SIZE (CM) . | |||
1 | |||
5 4 | |||
i 3) The core power histogram employed was for a 13.4 W/f t maximum design linear heat oper-i ation rate; see Figure 3-5. | |||
e RESULTS l | |||
; Normal Operation n e results from this analysis are shown in Figure 3 6 with I | |||
d initial fuel density as a parameter. The line shown for an initial l | |||
fuel density of 957. T.D. is considered to be most representative consider- | |||
] | |||
a ing current GE data on manufactured fuel pellet densities as a function of axial position, is the required margin which must be maintained during normal operation between the actual peak operating condition and the peak i | |||
design IJICR; i.e., 13.4 W/f t. Maintaining this margit. will assure, with i better than 95% confidence, that no more than one rod will exceed the design 1 | |||
, peak lilGR due to the random occurrence of power spikes resulting from axial j fuel column gans. Consistent with GE's position on densification, previously discussed in Reference 4 and its supplements, the results of this analysis are considered to be a very conservative representation of the power peaking penalty required to accommodate potential axial fuel column gaps during normal operating conditions in GE IMR's. | |||
Accident Effects - | |||
l Since the results of the power spiking analysis for normal operation will be utilized to limit bundle power to ' assure that the random occurrence of power spikes will not result in exceeding the design peak LilGR, it is not believed necessary to separately consider power spikes in the analysis of 1 transients or accidents which have as an initial condition some form of normal operation, n e control rod drop accident is unique in the respect that it begins at the cold condition, and is not affected by normal operating power level. Further, the existence of fuel column gaps can result in power 3-26c Revised 2/8/74 | |||
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E' r XV ' X'/ + X > 1.3 /l1X NV]d X 111114 3'dO3 I11 50_ ' 3011J33)l3d 6d Revised 2/8/74 | |||
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' I I 3 26e Revised 2/8/74 | |||
spiking in the cold condition during a control rod drop which should thus be considered in the evaluation of this accident. For this purpose, a separate power spiking enlysis has been performed using the same assumptions as indi-cated above, but employing a power spike versus gap size calculated to occur in the cold condition with zero voids (Figure 3-4). This analysis i | |||
was performed for a conservative maximum gap site calculated employing a pellet average immersion density of 94.5% T.D., and a position near the top of the core in order to maximize the power spiking effect. This analysis j yicided a 997 probability that any given fuel rod would have a power spike o f ( 57. . | |||
3.4.4.2 Cladding Creep Collapse Using the same conservative bases presented in References 7 and 8. the critical pressure ratio; i.e. , ratio of collapse pressure to actual coolant pressure, was calculated. Figure 3-7 presents the clad mid-wall temperature versus time for the 8x8 reload fuel. No credit is taken for internal gas pressure due to released fission gas or volatiles. The internal pressure due to helium backfill at I atmosphere during fabrication is considered. 1he fuel chacteristics for creep en11 apse calculations are j as follows: | |||
Clad 0.D., in. 0.493 Clad 1hickness, in. 0.034 A 0.003 peak LilGR, KW/ft 13.4 Fast Flux >l mev, n/cm -sec 4.37 x 10 13 2 | |||
Figure 3 8 gives the calculated critical pressure ratio. As evidenced by the curve, the calculated critical pressure ratio is always >1.0. | |||
3.4.4.3 Increased Linear liest Cencration Rate The following expression was employed to calculate the dec. ease in fuel column length due to densification in calculation of a penalty in linear heat generation rate: | |||
3-26f : | |||
Revised 2/8/74 | |||
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..__.r-------------- r- - -- r - - - - - - - - - - | |||
: o. o. o. o. | |||
., M N m ' | |||
011V>l 3?ifiSS3dd lV3111)O i | |||
3-26h Revised 2/8/74 l | |||
at = 0.965 - ft t 2 | |||
where a L = decrease in fuel column length L = fuel column length fi = mean value of measured initial pellet density (immersion - .5%) | |||
The length reduction due to densification as calculated by the above equation requires knowledge of the mean immersion density (pt) obtained from the QC data. A correction 0.5% T.D. is applied to convert the immersion density to a geometric density. The mean pellet immersion density for Monticello 8x8 fuel is 95.44% T.D. This results in a L = Ot 965 - (.9544 - 0.005) = 0.0156 e 0.008 L 2 2 or A L = 0.87 L | |||
Due to thermal expansion, an ax8 pellet normally expands in going from the cold to hot condition, on amount equal to 1.2% for a pellet at 13.4 KW/ft. | |||
This increase in length from the cold to hot condition is not taken credit for in either design calculations or in the process of core performance analysis during reactor cperation. The cold pellet length is assumed for r | |||
these conditions. | |||
i Therefore, the decrease in pellet length due to densification is more than l | |||
i of fset by pellet axial theimal expansion. | |||
l 3-261 Revised 2/8/74 | |||
a e REFERENCES - SECTION 3 | |||
: 1. UAPD TM 263, "Ef fects of High Bornup on Zircaloy Clad, halk U02 Plate Fuel | |||
/ | |||
Element Samples," September 1962. | |||
: 2. WAPD TM-629, " Irradiation Behavior of Zircaloy Clad ruel Rods ContaintLg Dished End UO 2 Pellets," July 1967. | |||
: 3. Williamson, H. H. , and Ditmore, D. C. , " Experience with IWR Puel Through September 1972." May 1972 (NFDO-10505). | |||
) | |||
i j 4. Ditmore D. C. , and Elkins, R. B., "Densification Considerations in IMR Fuel Design and Performance," December 1972 (NEDM 10735) . | |||
J | |||
: 5. Miner, M /... " Cumulative Damage in Fatigue," dournal of Applied Mechanics, | |||
,12. Transactions of the ASME, H,1945. | |||
:: 6. O'Donnel W. J. , and Langer, B. F. , " Fatigue Design Basis for Ziraloy 1 | |||
, Components'," Nuclear . Science and Engineering, 20, 1964. | |||
I | |||
: 7. llinds, J. A. (General Electric) letter to V. A. tbore (USAEC), | |||
" Plant Evaluations With CECAP-111," December 12, 1973. | |||
i | |||
: 8. NEDM-10735 Supplement 6, "Puel Densification Ef fects on General 1 Electric Boiling Water Reactor Fuel," August 1973. | |||
J i | |||
1 l | |||
3-26j Revised 2/8/74 | |||
: 4. THERMAL-HYDRAULIC CHARACTERISTICS 4.1 ITEL ASSIMBLY HYDRAULIC ANALYSIS 4.1.1 Core Pressure Drop. Hydraulic 1.oads, and__ Correlations The (1w distribution to the fuel asse%1ies is calculated on the assump-tion that the pressure drop across all fuel assemblies is the same. This assuv-tion has been confirmed by measurements of the f1w distribution in modern to11-ing water reactor as reported in References 1 and 2. The components of bundle pressure drop considered are f riction, local, elevation, and acceleration. Pres-sure drop measurements made in operating reactors confirm that the total measured core pressure drop and calculated core pressure drop are in good agreement. | |||
There is reasonable assurance, therefore, that the calculated f1w distribution throughout the core is in close agreement with the actual f1w distribution of an operating reactor. | |||
4.1.1.1 Friction Pressure Drop i | |||
; Friction pressure drop is calculated using the model relation 2 | |||
w fL AI f | |||
* 2gp 2 | |||
I TPT 3 H^ch where t.P g = friction pressure drop, poi, v = mass f1w rate. ' | |||
g = acceleration of gravity, p = vater density, D | |||
H | |||
= channel hydraulic diameter, Ag = channel f1w area. | |||
L = length, f a friction factor, and | |||
= two phase friction multiplier. | |||
(UF 4-1 Revised 2/8/74 j | |||
o wtDE W6DE COHNER 1 2 3 4 6 e 7 e I, 9 10 11 17 13 14 16 16 17 18 19 20 Pt 22 23 24 26 M 37 70 79 30 31 37 33 34 36 38 37 38 39 40 41 42 43 44 45 46 47 as 49 to 61 67 63 H 66 W 07 f4 60 to 61 62 83 M i | |||
i 1, i l , | |||
i i | |||
! Figure 6-24. 8x8 Reload Fuel Rod Identification * | |||
; t i | |||
l- | |||
.6-47 l > | |||
i i Revised 2/8/74 | |||
.t | |||
, e t Confomance Wi th int erim Accent st. Lee Criteria. | |||
In t% analyses discussed above there have been no deviations f rom the evaluation rodel described in Appendix A, Part 2 of the AEC Interim Policy Statement. | |||
Ef fects_of ECCS operation on the Core. The mechanical ef fects of ECCS opera-tion on the core, reactor coolant system and LCCS are those associated with the thermal ef fect of injecting water into these systems which is cooler than these systems and components. These thermal stresses have been considered in the design of the core, reactor coolant system and ECCS. | |||
There are no nuclear effects resulting from ECCS operation, since all con-trol rods are inserted and the reactor remains subcritical during the injection of the cooler ECCS water. | |||
There are no chemical additives in the ECCS water and therefore no chemical effects on the core, reactor coolant system or ECCS. | |||
tan Timeu. The system time delays assumed in the LOCA accident are as followns 6-48 Revised 2/8/74 | |||
Maximum Allowable Time From Maximum Time Delay After Signal Receipt Until the Pumps Receipt of Signal Until All llave Reached Rated Speed Valve Motion is Complete System (sec) _ (sec) _ | |||
llPCI 30 30 CS 30 30 LPCI 43 43 ADS - | |||
120 6.2.2.4.4 Densification Effects figure 6-24a provides plots of Maxicnim Average Planar Linear llent Generation Rate (MAPLilGR) and peak cladding temperature (PCT) versus exposure, for Monticello 8x8 reload fuel. | |||
The 1.0CA analyses were performed using the approved Interim Acceptante Criteria }bdel with gap conductance values as calculated with the new GFGAP III nodel with AEC modifications. | |||
6-48a | |||
__ ------ Da"!**d 7 /R /74 | |||
figon (,74a , | |||
. 4. It01tl(Li10 l'LI HAls P (1:rM) | |||
(,Lf.Al' I I 1 | |||
! IMi't.IICH AND l'Cf VL 1 /M r,Uli Cul:VL 1%Xll10'1 ALLO',!Aal L 10 L IM 1:ll.0'.I Cui !:L.lf li.C!tilCAl. | |||
SPLLlfilA110:1 Llitll ful: LilGR CURVC a . ImXil'uti A' LLOUAclI: tilTil GCGAP 111 A:10 AEC 110D!l JChl1015 I? | |||
tb s | |||
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i ' | |||
g j j | |||
* I s s W i i ; i . | |||
t | |||
:_s 10 5 10 15 20 25 30 i , j- , AVERAGE PLANAR EXPOSURE - GUD/ TON 5 .l.0 15 20 25 30 | |||
. 2300 , | |||
l , | |||
i d | |||
. c | |||
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!d 2200 i j | |||
., 8 - | |||
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2 4 | |||
w | |||
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:;. 210J , i 1 i ! | |||
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i I | |||
li. | |||
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1 | |||
: l. l j l l 9 t l | |||
6-48b Revised 2/8/74 | |||
6 WEDS-20016 | |||
: 11. Monticello Nuclear Generating Plant, First Reload License Submittal. | |||
February 1973, | |||
: 12. Hillstone Unit 1, FSAR Amendment 14, Dkt. 50-245. | |||
: 13. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel, Supplement 6, 7, and 8, Composite," August 1973 (NEDH-10735). | |||
: 14. "Results of Transient Reanalysis for Monticello Nuclear Generating Flant with End-of-Cycle Core Dynamic Characteristics," February,1973, | |||
: 15. "Monticello - Safety Valve Setpoint increase Analysis," Change request dated September 3, 1973. | |||
: 16. "Monticello - Cycle 2 Scram Reactivity Considerations, Analyses and Modifi-cations," October 1973. | |||
: 17. Hinds. J. A. (General Electric) let ter to V. A. Moore (USAEC) . | |||
" Plant Evaluations With GECAP III." December 12 1973. | |||
l 6-67 | |||
_ _ . _ _ _ _ o - < c ! M a> M . _ .- . . _ . - - | |||
l | |||
: 7. .TITHNICAL SpErl.TjQTJyjj There are four areas of the Technical Speci fications af f ected by the preceeding information. Changes made necessary hv the reactor pressure relief system nodifications discussed in Section 6.2.4 will be outlined in tiie forth-coming submittal on that subject. The formal request for Technical Specifiestion changes will be a separate. subsequent submittal. Specifications affected by this submit tal include the following: | |||
Section 2 - The heat flux of a 7x7 fuel assembly operating up to 17.5 kw/f t results in a 3.08 total peaking factor. Changes should reflect the use of Bx8 fuel operating up to 13.4 kw/f t resulting in a 3.04 total peaking factor. | |||
Section 3.3.C - The transient analynis (Section 6.2.4) was done based on a control rod scram time to 907. insertion of 3.5 seconds rather that 5.0 seconds as presently allowed. The Specification will be changed accordingly, pection 3.5.K - The 8x8, R-2 fuel will have unique properties for consideration of postulated fuel densification phenomena. Since the AEC staff nodel requires the use of measured pellet theoretical density, this information can not be finalized until the fuel is fabricated. | |||
Section 5.2 The facility description states that fuel assemblies have 49 tuel rods ecch. This must be changed to allow the use of 8x8, 63 fuel rod assemblies. | |||
7-1 Revised 2/8/74 | |||
_ _________ _ _ _ _ _ _ _ _ . _ _ _ . _ _J}} | |||
Latest revision as of 02:25, 13 May 2020
| ML20090M552 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/08/1974 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20090M550 | List: |
| References | |
| NUDOCS 9105070342 | |
| Download: ML20090M552 (24) | |
Text
{{#Wiki_filter:. . Regulatory Docket File pgp;.a,Ur Day 2 -{@ NORTliERN 5W.'tt.S POWER CCMPAln MCNTICELLO NUCLEAR GENE 9A'11hr PLANT l SECOND RELGAD SUBMITTAL REVISION 1 FEBRUARY 8, 1974 Prepared By Information Supplied By Northern States General Electric Power Company Company 9105070342 740200 PDR ADOCK 05000263 P PDR
i Table of Co. :a $ p
.yagg.
- 1. Introduction 11 I 2. Sunnary 2-1 1
- 3. Mechanical Design 31 3.1 General Design Description 3+ 1
' 3.2 Hechanical Design Bases 3$
3.3 Results from Mechanical tiesigu Evalestions 1 15
; 3.4 fuel Operating and Developmental Exptrience 3-17 1 Referencca Section .1 3 V6j j
i 4. Thermal Hydraulic Charact eristica 41 4.1 ruel Assembly Hydraulic Anolitis 4*l l
- 4.2 Fuel Assembly fliersnal-Hedt aulic Evaluation 4 '.
4.3 Results of Thermal-Hydraulic .inalysis 45 References - Section 4 48 l - i
$. Nucler.r Characteristics 51 l' 5.1 Introduction 5-1 a 5.2 Bundle Nuclear Dt.scription 5-1 5.3 Analytical Hethode 5-11 1 5.4 Experience with GE Nuclear !bdels 5-15 a 5.5 Nuclear Characteristics of the Core 5-16 References - Section 5 5 5 19 i
6, Safety Analysis 6-1 6.1 Model Applicability to 8x8 Fuel 6-1 6.2 Results of Safety Analysis 66 References - Section 6 6-66 i 7. Technical Specifications 7-1 t i l ilt t r
- ~ . . _ _ _ . _ _ _ _
_ _ _ . _ . . _ __ _ _ Revised 2/8/74
- 1 l
l I J List OF ILLUSTRATIONS rigure Title h ! 21 Reference Core loading tbnticello Cycle 3 2-2 31 8x8 Reload Fuel Assembly 32 . 32 tbnticello R2 Reload Puel Lattice 33 l 33 8x8 Power Spike Versus Cap Size - Normal Operation 3 26a 34 9x8 Power Spike Versus Cap Sire - Cold Zero void condition 3 26b 3-5 Percent of Rods in Core with X t Peak KW/FT (X a X, 4X=.2 3 26d l 3-6 8x8 Power Spike Penalty vs Axial Position - Normal Operation 3 26e
+
Cladding Average Temperature at A Puel Column Axial Cap 3 26g 37 38 Clad Critical Collapse Pressure Ratio Versus Time 8x8 Reload fuel 3-26h 5-1 llo t Average Void Infinite Lattice K versus gg Exposure 5-3 5-2 2.63 7x7, 6.62 8x8 Infinite Lattice Kag versus In channel 5-6 , Void Fraction Zero Exposure 5-3 AK void Comparison 7x7 versus 8x8 from 0.40 Void to other Voids 5-7 5-4 200 FMd/t Doppler coefficients Uncontrolled 5-9 5-5 10,000 tMd/t Doppler coef ficient s Uncontrolled Solo 5-6 Sversus Exposure Comparison 2.62 wt'. 8x8, 7.63 wt% 7x7 5-12 Average Voids, Uncontrolled i 57 tiaximum Local Peaking versus Exposure Comparison 5-13 l 6-1 Doppler Reactivity Coef ficient versus Average Fuel 6-13 Temperature as a Punction of Exposure and Moderator Condition , 6-2 Accident Reactivity Shape Functions for Cold Startup 6-14 jf= 0.0054 6-3 Scram Reactivity Function for Cold Startup 6-15 l 6-4 Cladding Temperature versus Time for the Recirculation 6-20 , Line Break with railure of the LPCI Injection Valve (llPCI + 2CS + ADS) AEC Assumptions 2 6-5 Performance of ECCS with railure of 11PCI for Small (0.02 f t ) 6-22 l l Liquid Break (2CS + 4LPCI + ADS) AEC Assumptions 6-6 6-23 Cladding Temperature vergus Time for a Small Break with Failure of IIPCI (0.02 ft Break) (4LPCI + 2CS + ADS) t i 1 iv ' Revised 2/8/74 _ - ___ , _ _ . - _ _ _ . _ _ . _ . . _ . . . _ _ - - _ . . ~ _ _ _ , - _ _ _ _ _ . _ _
LIST OF !!11tSTRAT10NR (Continued) 112ME1 Title lage 67 Peak Cladding Temperature Spectrum for Single Failure 6-24 f Conditions AEC Assumptions 6-8 Emergency Core Cooling System Performance Capability 6 26 j 6-29 h 6-9 Performance of ECCS for Fbin Steam Line Break Inside the Drywell with all ECCS Operating (HPCS + 2CS + 4LPCI + ADS) AEC Assumptions 6 10 Core flow and Pressure rollowing a Recirculation Line Break 6 30 6-11 Performance of ECCS with Failure of LPCI Injection Valve for 6-32 ; the Design Basis Recirculation Line Dreak (HPCS + 2CS + ADS) I AEC Assumptions 1 6-12 Minimum Critical Heat Flux Ratio for DBA at Honticello 6 13
?
6 13 Cladding Temperature versus jime for an Intermediate Break 6-35 with Failure of HPCI (0,1 ft Break) (4LPCI + 2CS + ADS) AEC Assumptions 6-14 Quality versus Time for DBA at Monticello 6-36 ! 6-15 Heat Transfer coef ficient for a small Break (0.02 f t2) 6-37 (4LPCI + 2CS + ADS) AEC Assumptions 6-16 Heat Transfer Coefficient for an Intermediate Break 6-38 ! (0.1 ft2) (4LPCI + 2CS + ADS) AEC Assumptions (, 6-17 tbnticello Heat Transfer coef ficients for DnA with LPCI 6-39 Injection Valve Tailure - AEC Analysis No. 1 (2CS + HPCI + ADS) 6 18 Performqnce of ECCS with Failure of HPCI for Intermediate 6-40 (0.1 f t') Liquid Break (2CS + 4LPCI + ADS) AEC Assumptions . 6-19 Power Cen= ration Following a Design Basis Recirculation Line 6-41 i Break Accident i
- 6-20 Ef fects of Exposure on Maximum Cladding Temperature 6 43 $
- 6-21 Puel Rod Perforation Data 6 44 l 6-22 Distribution of Internal Pressure Within Rods 6-45
! 6 23 Percent Rod Perforation versus Break Area (AEC Assumptions) 6 46 ' 6 24 8x8 Reload Fuel Rod Identification 6 47 i 6-24a Fbnticello Reload 2 (8x8) CECAP III MAPLHCR and PCT vs Exposure 6 48b l 6-25 Puel Type Locations 6 52 l l [ v , Revised 2/8/74
. - - - - . - . . . - , - - . . ~
LIST Or I!LUSTRATIONS (Continued) rigurs Title Pans Base Rod Pattern for RWE 6 53 6 26 6 27 tbnticello Composite HCliFR versus Rod Position Limiting 6-$4 Rod Withdrawal 6 26 MCHTR versus Rod Position Limiting Rod Withdrawal 7x7 6-5$ 6 29 MCHER versus Rod Position Limiting Rod Withdrawal 8x8 6-56 6-30 Peak liest Flux and Relative Power versus Rod Position 6-57 Limiting Rod Withdrawal !boticello DOC 3 6-31 RHM Response - Channels A and C 6 58 6-32 RBM Response - Channels B and D 6-59 6-33 Scram Reactivity curve 6-62 6-34 Control Rod Drive Scram Times 6-63 l l l 1 1 vi Revised 2/0/74
- 1. I_liTRODUCTION This docuinent provides the technical basis of the license submittal for the second reload of Northern States Pcuer/Monticello l' nit 1. Presented herein is a description of the new fuel and the results of the evaluation of the refueled core for the February,1974 outage.
The fuel at ttis site available for loading at the outage will be 116 Reload-2 fuel bundles, which are 8x8 bundles with an average enrichment of 2.62 wet of U-235. There will also be availabic i .r reinsert 7 initial core bundles discharged at the end of Cycle 1 with an average exposure of about 7400 mwd /t. The objective of this outage is to remove the 44 temporary control cur-taina which remained in the core during Cycle 2 and to load the core so as to assure the availability of the plant at high power for approximataly an annual cycle. Sections 3, 4, 5 and 6 of this document, dealing with the subjects of reload fuel mechanical design and reloaded core thermal-hydraulic and nuclear charac-teristics, and safety analysis, present descriptiore of design criteria, methods and results from design calculatior.s and safety evaluations and represents com-plete information for the review of fuel assenbly and core design. 1-1/1-2 Revised 2/8/74
A more detailed summary of Cencral Electric experience with BWR Zi rc a lo y- c l ad P02 pellet fuel, inc luding recent production and development data, has been documented (see Reference 3) . 3.4.4 Analysis of Fuel Densi fication Ef f ects This section presents results of the ef fects of densification in the 8x8 reload fuel as determined from application of the models described in Reference 7. 3.4.4.1 Power Spiking Analysis 8x8 Puel Lattice An analysis of potential local power spikes due to axial gaps in fuel pellet column for GE IMR's employing an Sx8 fuel lattice design has been performed. This analysis employs the same method and basic assumptions that were reported in Reference 7. Important aspects of this analysis are noted as follows:
- 1) The equation employed to calculate maximum gap size is that noted in Reference 7:
. 965 - f4 + 0.0025 L g ,
2 _ where L = maximum axial gap length L = fuel column length fi = mean value of measured initial pellet density (immersion - SV) 0.0025 = allowance for irradiation induced cladding growth and axial strain caused by fuel-clad mechanical interaction
- 2) The magnitude of the power spike versus gap size for fuel rods of the 8x8 design is shown in Figure 3-3 for normal opercting conditions, and Figure 3-4 for cold zero void conditions.
3-26 i Revised 2/8/74 I
FIGURE 3-3 - I 8x8 PO'4ER SPIKE VERSUS GAP SIZE - UOPRAL OPERATION . a
- i i .
- E3D RELATIO15 HIPS I
-- 7.0 _
w y __ . 6,0 . . .. -. . _ _ . - . . _ . . _ _ _ _ ___ -_.- - - -
~ ~ - - ~ '
p 5.0 - - - - -
' _ . . . . . _ _ _ _ . . - - - - -- - - - The effect of a gap in Ecd 1
, jr '
on adjacent pellet i. _1_,0 n. . . .
~~ -
t- Effect of gap in Rod 2 on Red 1 _3 .0 ... _ .__ _ _ _ . . _ . . . . _ _ . _ _ . . . _ . . _
- ~ ~/ p .-
/ ,, . . '
/ ----
Effect of gap in Rod 3 on Rod 1 2 0 -' ~' - - ~* ~ - ~~--'[-~' ---- i .- Effect of gap in Rod 4 cn Rod 1
'(; l.0 ~ :
~ ~ ~
/
^ ^ ~' . . ,
'y ~
=_..
Effect of gap in Rod 5 on Rod 1 . , / ,., -
.im> 4 .. *.
} 2.0 3.0 4.0 5.0 6.0 70 - - 1.0 1
- ~R . . ._
r U . . _. ._ _ . _ _ _ . . _ _ . _ _ . _
! _--_-.__-_.-._s-._ - . . . . . .
4 y 1 2
s FIGURE 3 4 . . . . ._ . .._. 8x8 PO*4ER SPIKE VERSUS GAP SIZE - COLD ZERO VOID CONDITION . R03 RELATIONSHIPS 4
-.. .--- . .. . . . . - - . . ~ - -.
..- Effect of a gap in Rod el on
~
p adjacent pellet LP/P , __ _, _, __ g,____ _ _ _ _ , , w 7 i.
, h c; - --
/
) @- ..-..--./. 6 - . . - . . - _ _
.--- - - - ,. ._ Effect of a gap in Rod e2
/ - ~ -
on Rod dl j ,- , 5 - -- 7' - -- : f- - - - -- - - - - j
/ -- - - - /- --f._ _ - _ . .. . _
j .....g , --_-....-.,/._----------..-. -- r ,-
/
/ .. .
i -- - 3 -
- - 7 '- - - - - - - - -
- - ; -- Effect of a gap in Rod e3 on i / .
.__., . Rod #1 e '
- x -
2 - . - a . , l < .' s . ,s"',
._ _ _ ' - _ _ . <[- - Effect of a gap in Rod *4 on Rod l {, .
1 . 1. - '
. _ _ _--_2--__ -
;. _-a .
Effect of a gap in Red -5 on Rod . u .. ' ~ - __ -L_ - -- - .
)
.i - - _ _
, , '.*./ - _
M S* 1 2 3 4 5 6
~
~ ~~
- ~~ ~~'-~~~~~ ~ ~
GAP SIZE (CM) . 1
5 4 i 3) The core power histogram employed was for a 13.4 W/f t maximum design linear heat oper-i ation rate; see Figure 3-5. e RESULTS l
- Normal Operation n e results from this analysis are shown in Figure 3 6 with I
d initial fuel density as a parameter. The line shown for an initial l fuel density of 957. T.D. is considered to be most representative consider- ] a ing current GE data on manufactured fuel pellet densities as a function of axial position, is the required margin which must be maintained during normal operation between the actual peak operating condition and the peak i design IJICR; i.e., 13.4 W/f t. Maintaining this margit. will assure, with i better than 95% confidence, that no more than one rod will exceed the design 1 , peak lilGR due to the random occurrence of power spikes resulting from axial j fuel column gans. Consistent with GE's position on densification, previously discussed in Reference 4 and its supplements, the results of this analysis are considered to be a very conservative representation of the power peaking penalty required to accommodate potential axial fuel column gaps during normal operating conditions in GE IMR's. Accident Effects - l Since the results of the power spiking analysis for normal operation will be utilized to limit bundle power to ' assure that the random occurrence of power spikes will not result in exceeding the design peak LilGR, it is not believed necessary to separately consider power spikes in the analysis of 1 transients or accidents which have as an initial condition some form of normal operation, n e control rod drop accident is unique in the respect that it begins at the cold condition, and is not affected by normal operating power level. Further, the existence of fuel column gaps can result in power 3-26c Revised 2/8/74
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spiking in the cold condition during a control rod drop which should thus be considered in the evaluation of this accident. For this purpose, a separate power spiking enlysis has been performed using the same assumptions as indi-cated above, but employing a power spike versus gap size calculated to occur in the cold condition with zero voids (Figure 3-4). This analysis i was performed for a conservative maximum gap site calculated employing a pellet average immersion density of 94.5% T.D., and a position near the top of the core in order to maximize the power spiking effect. This analysis j yicided a 997 probability that any given fuel rod would have a power spike o f ( 57. . 3.4.4.2 Cladding Creep Collapse Using the same conservative bases presented in References 7 and 8. the critical pressure ratio; i.e. , ratio of collapse pressure to actual coolant pressure, was calculated. Figure 3-7 presents the clad mid-wall temperature versus time for the 8x8 reload fuel. No credit is taken for internal gas pressure due to released fission gas or volatiles. The internal pressure due to helium backfill at I atmosphere during fabrication is considered. 1he fuel chacteristics for creep en11 apse calculations are j as follows: Clad 0.D., in. 0.493 Clad 1hickness, in. 0.034 A 0.003 peak LilGR, KW/ft 13.4 Fast Flux >l mev, n/cm -sec 4.37 x 10 13 2 Figure 3 8 gives the calculated critical pressure ratio. As evidenced by the curve, the calculated critical pressure ratio is always >1.0. 3.4.4.3 Increased Linear liest Cencration Rate The following expression was employed to calculate the dec. ease in fuel column length due to densification in calculation of a penalty in linear heat generation rate: 3-26f : Revised 2/8/74
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at = 0.965 - ft t 2 where a L = decrease in fuel column length L = fuel column length fi = mean value of measured initial pellet density (immersion - .5%) The length reduction due to densification as calculated by the above equation requires knowledge of the mean immersion density (pt) obtained from the QC data. A correction 0.5% T.D. is applied to convert the immersion density to a geometric density. The mean pellet immersion density for Monticello 8x8 fuel is 95.44% T.D. This results in a L = Ot 965 - (.9544 - 0.005) = 0.0156 e 0.008 L 2 2 or A L = 0.87 L Due to thermal expansion, an ax8 pellet normally expands in going from the cold to hot condition, on amount equal to 1.2% for a pellet at 13.4 KW/ft. This increase in length from the cold to hot condition is not taken credit for in either design calculations or in the process of core performance analysis during reactor cperation. The cold pellet length is assumed for r these conditions. i Therefore, the decrease in pellet length due to densification is more than l i of fset by pellet axial theimal expansion. l 3-261 Revised 2/8/74
a e REFERENCES - SECTION 3
- 1. UAPD TM 263, "Ef fects of High Bornup on Zircaloy Clad, halk U02 Plate Fuel
/
Element Samples," September 1962.
- 2. WAPD TM-629, " Irradiation Behavior of Zircaloy Clad ruel Rods ContaintLg Dished End UO 2 Pellets," July 1967.
- 3. Williamson, H. H. , and Ditmore, D. C. , " Experience with IWR Puel Through September 1972." May 1972 (NFDO-10505).
)
i j 4. Ditmore D. C. , and Elkins, R. B., "Densification Considerations in IMR Fuel Design and Performance," December 1972 (NEDM 10735) . J
- 5. Miner, M /... " Cumulative Damage in Fatigue," dournal of Applied Mechanics,
,12. Transactions of the ASME, H,1945.
- 6. O'Donnel W. J. , and Langer, B. F. , " Fatigue Design Basis for Ziraloy 1
, Components'," Nuclear . Science and Engineering, 20, 1964. I
- 7. llinds, J. A. (General Electric) letter to V. A. tbore (USAEC),
" Plant Evaluations With CECAP-111," December 12, 1973.
i
- 8. NEDM-10735 Supplement 6, "Puel Densification Ef fects on General 1 Electric Boiling Water Reactor Fuel," August 1973.
J i 1 l 3-26j Revised 2/8/74
- 4. THERMAL-HYDRAULIC CHARACTERISTICS 4.1 ITEL ASSIMBLY HYDRAULIC ANALYSIS 4.1.1 Core Pressure Drop. Hydraulic 1.oads, and__ Correlations The (1w distribution to the fuel asse%1ies is calculated on the assump-tion that the pressure drop across all fuel assemblies is the same. This assuv-tion has been confirmed by measurements of the f1w distribution in modern to11-ing water reactor as reported in References 1 and 2. The components of bundle pressure drop considered are f riction, local, elevation, and acceleration. Pres-sure drop measurements made in operating reactors confirm that the total measured core pressure drop and calculated core pressure drop are in good agreement.
There is reasonable assurance, therefore, that the calculated f1w distribution throughout the core is in close agreement with the actual f1w distribution of an operating reactor. 4.1.1.1 Friction Pressure Drop i
; Friction pressure drop is calculated using the model relation 2
w fL AI f
- 2gp 2
I TPT 3 H^ch where t.P g = friction pressure drop, poi, v = mass f1w rate. ' g = acceleration of gravity, p = vater density, D H
= channel hydraulic diameter, Ag = channel f1w area.
L = length, f a friction factor, and
= two phase friction multiplier.
(UF 4-1 Revised 2/8/74 j
o wtDE W6DE COHNER 1 2 3 4 6 e 7 e I, 9 10 11 17 13 14 16 16 17 18 19 20 Pt 22 23 24 26 M 37 70 79 30 31 37 33 34 36 38 37 38 39 40 41 42 43 44 45 46 47 as 49 to 61 67 63 H 66 W 07 f4 60 to 61 62 83 M i i 1, i l , i i ! Figure 6-24. 8x8 Reload Fuel Rod Identification *
- t i
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.6-47 l >
i i Revised 2/8/74
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, e t Confomance Wi th int erim Accent st. Lee Criteria. In t% analyses discussed above there have been no deviations f rom the evaluation rodel described in Appendix A, Part 2 of the AEC Interim Policy Statement. Ef fects_of ECCS operation on the Core. The mechanical ef fects of ECCS opera-tion on the core, reactor coolant system and LCCS are those associated with the thermal ef fect of injecting water into these systems which is cooler than these systems and components. These thermal stresses have been considered in the design of the core, reactor coolant system and ECCS. There are no nuclear effects resulting from ECCS operation, since all con-trol rods are inserted and the reactor remains subcritical during the injection of the cooler ECCS water. There are no chemical additives in the ECCS water and therefore no chemical effects on the core, reactor coolant system or ECCS. tan Timeu. The system time delays assumed in the LOCA accident are as followns 6-48 Revised 2/8/74
Maximum Allowable Time From Maximum Time Delay After Signal Receipt Until the Pumps Receipt of Signal Until All llave Reached Rated Speed Valve Motion is Complete System (sec) _ (sec) _ llPCI 30 30 CS 30 30 LPCI 43 43 ADS - 120 6.2.2.4.4 Densification Effects figure 6-24a provides plots of Maxicnim Average Planar Linear llent Generation Rate (MAPLilGR) and peak cladding temperature (PCT) versus exposure, for Monticello 8x8 reload fuel. The 1.0CA analyses were performed using the approved Interim Acceptante Criteria }bdel with gap conductance values as calculated with the new GFGAP III nodel with AEC modifications. 6-48a __ ------ Da"!**d 7 /R /74
figon (,74a ,
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6 WEDS-20016
- 11. Monticello Nuclear Generating Plant, First Reload License Submittal.
February 1973,
- 12. Hillstone Unit 1, FSAR Amendment 14, Dkt. 50-245.
- 13. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel, Supplement 6, 7, and 8, Composite," August 1973 (NEDH-10735).
- 14. "Results of Transient Reanalysis for Monticello Nuclear Generating Flant with End-of-Cycle Core Dynamic Characteristics," February,1973,
- 15. "Monticello - Safety Valve Setpoint increase Analysis," Change request dated September 3, 1973.
- 16. "Monticello - Cycle 2 Scram Reactivity Considerations, Analyses and Modifi-cations," October 1973.
- 17. Hinds. J. A. (General Electric) let ter to V. A. Moore (USAEC) .
" Plant Evaluations With GECAP III." December 12 1973.
l 6-67 _ _ . _ _ _ _ o - < c ! M a> M . _ .- . . _ . - -
l
- 7. .TITHNICAL SpErl.TjQTJyjj There are four areas of the Technical Speci fications af f ected by the preceeding information. Changes made necessary hv the reactor pressure relief system nodifications discussed in Section 6.2.4 will be outlined in tiie forth-coming submittal on that subject. The formal request for Technical Specifiestion changes will be a separate. subsequent submittal. Specifications affected by this submit tal include the following:
Section 2 - The heat flux of a 7x7 fuel assembly operating up to 17.5 kw/f t results in a 3.08 total peaking factor. Changes should reflect the use of Bx8 fuel operating up to 13.4 kw/f t resulting in a 3.04 total peaking factor. Section 3.3.C - The transient analynis (Section 6.2.4) was done based on a control rod scram time to 907. insertion of 3.5 seconds rather that 5.0 seconds as presently allowed. The Specification will be changed accordingly, pection 3.5.K - The 8x8, R-2 fuel will have unique properties for consideration of postulated fuel densification phenomena. Since the AEC staff nodel requires the use of measured pellet theoretical density, this information can not be finalized until the fuel is fabricated. Section 5.2 The facility description states that fuel assemblies have 49 tuel rods ecch. This must be changed to allow the use of 8x8, 63 fuel rod assemblies. 7-1 Revised 2/8/74 _ _________ _ _ _ _ _ _ _ _ . _ _ _ . _ _J}}