ML20081K822
| ML20081K822 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/31/1983 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20081K795 | List: |
| References | |
| 79NED288, NEDO-24195-A-05, NEDO-24195-A-05-ERR, NEDO-24195-A-5, NEDO-24195-A-5-ERR, NUDOCS 8311100274 | |
| Download: ML20081K822 (23) | |
Text
NUCLEAR ENERGY BUSINESS OPER ATIONS o GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 im U
GENER AL $ ELECTRIC APPLICABLE TO:
~2
PUBLICATION NO 79NED288 ER N M And ADDENDA T.1. E. NO.
TITLE Ceneral Electric Reload no. 5 Fuel Application for Oyster OATE July 1983 Creek NOTE: Correctallcopies of the applicable ISSUE DATE Aunust 1979 publication as specified below.
REFERENCES INST RUCTIONS PfRAG APH N E) (COR RECTIONS AND ADDITIONS) 01 Title Page Replace with new Title Page.
02 Page viii Replace with new page viii.
03 Page viii-a Replace with new page viii-a.
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04 Page x Replace with new page x.
05 Page 5-30 Replace with new page 5-30.
06 Page 5-31 Replace with new page 5-31.
07 Page 5-32 Replace with new page 5-32.
08 Page 5-33 Replace with new page 5-33.
09 Page 5-34 Replace with new page 5-34.
10 Page 5-35 Replace with new page 5-35.
11 Page 5-36 Replace with new page 5-36.
12 Page 5-37/5-38 Replace with new pages 5-37 and 5-38.
13 Page 5-41c Insert new page 5-41c.
14 Page 5-50 Replace with new page 5-50.
15 Page 5-51a Replace with new page 5-51a.
16 Page 5-59c Insert new page 5-59c.
17 Page 5-70 Replace with new page 5-70.
18 Page 5-71 Replace with new page 5-71, 19 Page 5-72 Replace with new page 5-72, 20 Page 5-73 Replace with new page 5-73.
O y 21 Page 5-74 Replace with new page 5-74 22 Page 5-82 Delete page 5-82.
8311100274 831028 PDR ADOCK 05000219 PAGE 1 Of 1 P PDR
. l- NEDO-24195 79NED288 Class I August 1979 Amendment 5 I- July 1983 i
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! GENERAL ELECTRIC
- RELOAD FUEL APPLICATION 4
FOR l-4- OYSTER CREEK
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NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPA?4Y t SAN JOSE. CALIFORNIA 95125 1.
I GENER AL $ ELECTRIC-
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/ LIST OF ILLUSTRATIONS (Continued)
' (])
Figure Title Page 5-6 Increase in Heat Generated as a Function of Distance 5-65 from the Gap 5-7 Idealization of , Flux Spike 5-66 5-8 q ' as a Function of a General Profile 5-67 5-9 Central Peak Heat-Flux Distribution (TS No. 65) 5-67 5-10 Central Peak Heat-F1,ux Distribution (NS No. 76) 5-68 5-11 Critical Quality Ve'rsus Boiling Length for Tests 65 and 76 5-68 5-12 Damping Coefficient Versus Decay Ratio (Second Order Systems) 5-69 5-13 (Deleted) 5-70 5-14 (Deleted) 5-71 5-15 (Deleted) m 5-72 g 5-16 (Deleted) 5-73 G(,/ 5 -17 (Deleted) 5-74 5-18a Water Level Inside the Shroud and Reactor Vessel _
Pressure Following a 4.66 ft2 Recirculation Discharge .,
Line Break, Emergency Condenser Failure 5-75a q 5-18b Water Level Inside the Shroud and Reactor Vessel Pressure Following a 1.0 ft2Recirculation Discharge -
m Line Break, Emergency . Condenser Failure 5-75b _h 5-18e Water Level Inside the Shroud and Reactor Vessel Pressure Following a 0.3 ft2 Recirculation Discharge _,
-Line Break, Emergency Condenser Failure 5-75c q
~
5-18d' Water Level Inside the Shroud and Reactor Vessel Pressure Following a 0.10 ft 2 Recirculation Discharge ., ,
Line Break, Emergency Condenser Failure 5-75d q e 5-19a.1 _ Peak Cladding Temperature Following a 4.66 ft' Recirculation Line Discharge Break, Emergency Condenser .,
Failure (LBM) (E >20,000 mwd /ST) 5-76a.1 q 5-19a.2 Peak Cladding Temperature Following a 4.66 ft Recirculation Line Discharge Break, Emergency Condenser g Failure (LBM) (E <1000 mwd /ST) 5-76a.2 ,g 5-19b . Peak Cladding Temperature Following a 1.0 f t 2 ,,
Recirculation Line Discharge Break, Emergency Condenser m Failure (LBM) 5-76b .2 b(' -
19c Peak Cladding Temperature Following a 0.3 f t _
Recirculation Line Discharge Break, Emergency Condenser Failure (LBM) 5-76c -
_2 -
viii
, _ - . . __ _ _ _ - ~ . _ . . . _ _ . _ -
i NEDO-24195 LIST OF ILLUSTRATIONS (Continued).
Figure Title Page i
~
2 5-19d . Peak Cladding Temperature Following a 0.3 ft Recircu-lation-Line Discharge Break, Emergency Condenser Failure-(SBM) 5-76d Peak Cladding Temperature Following a 0.10 ft 2-
. ~5-19e-Recirculation Line Discharge Break, Emergency Condenser Failure-(SBM) 5-76e i- .
~
, 5-20a Fuel Rod Convective Heat Transfer' Coefficient at the -
i Highest. Power Axial Node for a 4.66 ft2 Recirculation -
1/80 Line Discharge Break (LBM) 5-77a - 9779 5-20b Fuel Rod _ Convective ' Heat Transfer Coef ficient at the
. -Highest Power Axial Node for a 1.0 ft2 Recirculation 1/80-Line . Discharge Break (LBM) 9/79
- 5-77b _
5-20c Fuel Rod Convective Heat Transfer Coefficient at the -
4 ' Highest' Power Axial Node for a 0.13 ft2 Recirculation
.Line Discharge Break (LBM) 5-77c _
9/79 g s.
21a : Normalized Core Average Inlet Flow Following a Maximum "
Recirculation Line Discharge Break (4.66 fe z) 5-78 5-21b' ~ Normalized Core Average Inlet Flow Following a Maximum
- . Recirculation Line Discharge Break (1.0 f t2) 5-78a
.- 5-21c Normalized Core Average Inlet Flow Following a Maximum U
Recirculation Line Discharge Break (O.3 ' f t 2) 5-78b 5-22a Minimum Critical Power Ration Following a 1.0 f t2 Recirculation Line Discharge Break 5-79; 5--22b Minimum Critical Power Ratio Following a 0.3 ft2 Recirculation Line Discharge Break 5-79a 23 Normalized Power Versus Time 5-80 5-24 P.eak Cladding Temperature Versus Break Area 5-81 9/79 t
- _R s +
l vili-a
, __m- . . _ . . . ~ . - . - . _ . _ _ _ _ . _ - _ , _ . . _ . . , _ . . _ _ _ , _ _ _ _ . - - - _ . .
NEDO-24195 LIST OF TABLES (Continued) l Table Title Py 5-12 LOCA Analysis Figure Summary 5-59 5-13 Single Failures Considered in the Oyster Creek LOCA Analysis 5-59 _
m 5-14a MAPLHGR Versus Average Planar Exposure O (Fuel Type: P8DRB239) 5-59a -
5-14b MAPLHGR Versus Average Planar Exposure (Fuel Type: P8DRB265L) 5-59a 5-14c MAPLHGR Versus Average Planar Exposure (Fuel Type: P8DRB265) "
5-59b 5-15 Four-Loop Operation MAPLHGR Multipliers 5-59c m
R
_~
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O
NEDO-24195 5.5.1 Control Rod Drop Accident Evaluation D
Q There are many ways of inserting reactivity into a boiling water reactor; how-ever, most of them result in a relatively slow rate of reactivity insertion and therefore pose no threat to_ the system. It is possible, however, that a rapid removal of a high worth control rod could result in a potentially significant excursion; therefore, the accident which has been chosen to encompass the consequences of a reactivity excursion is the control rod drop accident. y The dropping of the rod results in a high local reactivity in a small region of the core and for large, loosely coupled cores, significant shifts in th'e spatial power generation during the course of the excursion. Therefore, the method of analysis.must be capable of accounting for any possible effects of the power distribution shif ts.
Analysis of this accident is performed at various reactor operating states; the key reactivity feedback mechanism affecting the shutdown of the initial prompt f'}
m power burst is the-Doppler reactivity coefficient. Final shutdown is achieved by scramming all but the dropped rod. The methods utilized to evaluate the rod drop accident have been updated on a continuing basis to reflect improvements -
n in analytical capability (References 5-13, 5-14, 5-15 and 5-16). f Since the lattice cross sections are homogenized and the reactivity charac-teristics of all BWR lattices are similar, the fuel lattice has no effect on the excursion model used in the analysis of the RDA or on the reactivity feed-back effect due to Doppler which is used in the analysis. The number of fuel pins failed due to the RDA is dependent on the fuel pin (local) power peaking factors in the bundle and the final peak fuel enthalpy in the core. The local peaking factors are known from the lattice design calculations, and peak fuel enthalpy can be determined from the RDA analysis.
Homogenized bundle cross sections and nuclear constants are calculated using standard lattice design techniques. Because the bundle cross sections, which are produced from the lattice calculations and which are used in the RDA f~ excursion model are homogenized, the RDA excursion model does not recognize kq the lattice type used to produce the bundle cross sections.
5-30
7._
- (}- A mixture of . fuel types in a reloaded core presents no analytical problem.
The homogenized cross sections and nuclear constants used to represent each 1
-fuel bundle in the RDA analysis are calculated using methods which have
/ _previously.been used for lattice designs from 6x6 to 11x11 geometry and in mixed cores. Local power peaking at RDA conditions is explicitly calculated.
5.5.1.1 Sequence of Events The sequence of events and approximate time of occurrence for this postulated accident are described below.
Approximate Event Elapsed Time (a) Reactor is at a control rod pattern corresponding to a maximum incremental rod worth. -
~
(b) . Rod worth minimizer (RWM) or operators are functioning
[) within constraints of Bank Position Withdrawal Sequences (BPWS). The control rod that will result in the maximum incremental reactivity worth addition at any_ time in core life under any operating condition f$
while employing the BPWS becomes decoupled from the control rod drive. -
~
(c) Operator selects and withdraws the drive of the decoupled rod along with the other control rods assigned to the Banked-Position group such that the g proper core geometry for the maximum incremental rod 0 worth exists. -
'(d) .Decoupled control rod sticks in the fully inserted position. -
l (e) Control rod becomes unstuck and drops at the maximum velocity determined from experimental data (3.11 fps). O
_ {a]
- 5-31
NEDO-24195 Approximate
'(m) , Event Elapsed Time (f) Reactor goes on a positive period and initial power 1 1 sec burst is terminated by the Doppler Reactivity :
Feedback. _
(g) APRM 115.7% power signal scrams reactor (conserva-tive; in startup mode, APRM scram would be operative
+ IRM). -
(h) Scram terminates accident. 1 5 see To limit the worth of the rod which could be dropped, the RWM -~ -
is used-below 10% power to enforce the rod withdrawal sequence. The RWM is programmed to follow the Bank Position Withdrawal Sequences (BPWS), which are generally defined in Reference 5-13.
The rod drop accident design limit restricts peak cnthalpies in excess of (Oj 280 cal /gm for any possible plant operation or core exposure.
5.5.1.2 Model Parameters Sensitivities Although there are many input parameters to the RDA analysis, the resultant peak fuel enthalpy is most sensitive to the following input parameters:
(1) cteady-state accident-reactivity shape function; (2) total control rod reactivity worth; (3) maximum interassembly local power peaking factor, pP . Pp repre-sents the maximum local peaking factor normalized over the four bundles surrounding the dropped control rod. Mathematically, 4 BP P
p
= Max P 4 i = 1,2,3,4 (5-14)
L' ,
w [ BP (d. <j=1 ,
5-32
., . - - , _ . _ ._. ~_. , - --_ .-,. _ - - , - . _ _ . - - , _-
NEDO-24195 O- where C'
subscript i refers to one of the four bundles surrounding the control rod P = local peaking factor for the ith bundle BP = integrated power over the Ith bundle with the control rod withdrawn.
(4) delayed neutron fraction; (5) scram reactivity shape function; (6) doppler reactivity feedback; and (7) moderator temperature.
O P.od drop velocity is assumed to be that justified by the statistical evaluation in the appendix to Reference 5-14 (i.e., the maximum velocity of 3.11 f t/sec was used).
The conservative times tabulated below were used in developing the scram reac-tivity curves for the 280 cal /gm design limit boundary. - - -
Time from Deenergization of
% of Rod Insertion Scram Solenoid Valve (sec) 5 0.475 20 l'.10 50 2.0 90 5.0 5.5.1.3 Basic Conditions for Bounding Analysis To meet the RDA design limit of 280 cal /gm, the input parameters discussed in Subsection 5.5.1.2 are combined to meet three basic conditions. These 5-33
NED0-24195 i
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) are: (1) the accident reactivity characteristics; (2) the Doppler reactivity feedback; and (3) the scram reactivity characteristics.
The above conditions must all be satisfied in order to conservatively stay within the 280 cal /gm design limit boundary. If any one of the bounding conditions is not met, then a more detailed plant-specific evaluation would have to be performed to demonstrate compliance with the design limit.
Because so many parsmeters are involved in the determination of the resultant peak fuel enthalpy due to a control rod drop accident, it is not realistic to set a specific value of maximum control rod worth that would be used in Tech-nical Specifications. In the past, a local peaking factor was applied as the upper limit and, based on this local peaking value, a " maximum allowable" control rod worth was set. In reality, a core may exceed both the value for rod worth and the local peaking factor value yet still meet all the boundary requirements. Therefore, no specific control rod worth requirement will be set other than those described above.
i /
5.5.1.4 Analytical Methods Techniques and models used to analyze the rod drop accident (RDA) are docu-mented in References 5-14, 5-15, 5-16. The information in these documents has been used for the development af fesign approaches to make the consequences of RDA acceptable. Where safety analyses and resulting technical specifications -
were previously established with the old approaches, the information in the above referenced reports is not easily applied. The purpose here is to bridge that gap by using the information and techniques developed to provide a technical basis with the current design bases safety philosophy applied in the RDA area.
Control rod drop accident results from BPWS plants based on the above condi- g tions have been statistically analyzed and documented in Reference 5-37. 70 The results show that in all cases the peak fuel enthalpy in an RDA would be much less than the 280 cal /gm design limit even with a maximum incremental s rod worth corresponding to 95% probability at the 95% confidence level. Based
]
5-36
n-NED0-24195
) on these results, it is expected that an RDA will not result in limiting peak fuel enthalpies, and therefore plant and cycle specific RDA analyses are not performed for BPWS plants.
p
(,
x- /
- e 5.5.1.5 Effect of Fuel Densification Localized power spikes due to axial gaps in the fuel column would result in ,
a proportional increase in the calculated peak fuel enthalpy. Current rod drop accident (RDA) analyses indicate that the peak enthalpy occurs approxt-mately 18 inches from the top of the core in a fuel bundle adjacent to the dropping control rod. Qualitatively, it should also be recognized that this axial spiking effect is very localia-d and only one or two fuel pellets of _
a very small number of fuel rods will be af fected by a rod drop accident.
L.)
A 5-37
NEDO-24195 (g/ The effect of axial gap formation due to fuel densification on the rod drop accident results is discussed in Reference 5-17. Based on this evaluation, it has been established chat there is a 99% probability that increased local peaking in any fuel-rod due to the formation of axial gaps will be less than 5%. This effect has been accommodated by adjusting the local peaking factor.
5.5.1.6 Results and Consequences Based on the preceding analysis, it was conservatively determined that 837*
fuel rods of the P8x8R configuration would reach a fuel enthalpy of 170 cal /gm, which is the enthalpy limit for eventual cladding perforation. Safety analysis reports written before the development of the model and techniques reported previously, and those used to predict the 837 failures, resulted in the fail-ure of approximately 330 fuel rods for the 7x7 fuel. Based on these new models and assumptions, the resultant number of failures for a 7x7 core would be 650 fuel roda. If the conservative assumption is made that the fractional plenum activity for the P8x8R fuel is the same as for the 7x7 fuel, the resultant increase in activity released from the P8x8R fuel and the subsequent
(}
~~
radiological exposures relative to 7x7 analysis for the failure of 330 rods {$
is (837/330)* (49/62) = 2 times the 7x7 analysis. As noted in the FSAR, even if the radiological exposures are increased by a factor of two, the effects are still orders of magnitude below those identified in 10CFR100.
l
- Includes 10% allowance for uncertainties in the calculation.
l l
5-38 i
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5.5.2.11 Four-Loop Operation LOCA Analysis For the four-loop operation case, the same core power was used since 100% flow can be obtained with four recirculation loops. However, under these conditions the broken loop in the LAMB analysis was modeled with a higher flow, and the remaining loop in LAMB was modeled with a corresponding decrease in ~. low. Therefore, initial core flow remained constant.
The ef fect of the reduced core flow coastdown time affects the large break calculations by yielding an earlier boiling transition time. There is no impact of reduced core flow on small breaks. The effect of the reduced inven-tory (idle loop isolated case only) on small breaks is an earlier core g uncovery. There is no impact of reduced inventory on large breaks (i.e., ?
uncovery occurs so fast that insignificant delay is calculated). For the transition break (i.e., 0.3 ft ), the reduced core flow coastdown and inven-tory have a negligible (i.e. , <20*F) ef fect on the calculated PCT. Calcula-(^] tions performed show that a 2% reduction in MAPLHGR limits conservatively U compensates for the reduction in boiling transition time for large breaks.
A 2% reduction in MAPLHGR limits also conservatively compensates for the reduction in core uncovery time for small breaks as shown in previous calcu-lations (Reference 5-36). Note that no reduction is necessary for MAPLHGR limits set by the limiting small break (5000 to 15,000 MWD /ST exposure range) if the idle loop is not isolated.
MAPLHGR multipliers for four-loop operation are given in Table 5-15.
. 'tv gI 5-41c
f- ,
5.6 REFERENCES
5-1 General Electric B:,R Thermal Analysis Basis (CETAB): D1ta, Correlation and Design Application, January 1977 (NEDE-10958-PA and NEDO-10958-A) .
5-2 Basis for 8x8 Retrofit Puel Thermal Analysis Application, September 1978 (NEDE-24131).
5-3 R. B. Linford, Analytical Methods of Plant Transient
- nluations for the General Electric Boiling Water Reactor, February 1973 (NED0-10802).
5-4 R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BVR Amendment No. 2, June 1975 (NED0-10802-01) .
5-5 R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR Amendment No. 2, June 1975 (NEDO-10802-02) .
5-6 Generation of Void and Doppler Reactivity Feedback for Application to BWR Design, December 1975 (NEDO-20964).
5-7 J. A. Woolley, "Three Dimensional BWR Core Simulator," January 1977 (NEDO-20953A).
5-8 Analytical Model for Loss-of-Coolant Analyses in Accordance with 10CFRSO n Appendir K, January 1976 (NEDE-20566-P and NEDO-20566) .
f I L;
5-9 R. L. Bolger, Commonwealth Edison Company, letter to E. G. Case, Deputy Director Office of Nuclear Reactor Regulation, USNRC,
Subject:
Dresden Station Unit No. 2 Proposed Amendment to Facility Operating License No. DPR-19 to Permit Power Coastdown from 70% Power to 40% Power, NRC Docket No. 50-237, dated June 6, 1977.
5-10 0. Glenn Smith, W. M. Rohren, Jr., and L. S. Tong, " Burnout in Steam-Water Flows with Axially Nonuniform Heat Flux " ASME Paper 65-WA/HT-33, November 1965.
5-11 H. S. Swanson, J. R. Carver, and C. R. Kakaral, "The Influence of Axial Heat Flux Distribution on the Departure from Nucleate Boiling in a Water-Cooled Tube," ASME Paper 62-WA-297, November 1962.
5-12 ." Stability and Dynamic Performance of the General Electric Boiling Water Reactor," NEDO-21506, January 1977.
~
5-13 C. J. Paone, Banked Position Withdraval Gaquence, January 1977 m
-(NEDO-21231). R
_1
- 5-14 C. J. Paone and J. A. Woolley, Rod Drop Accident Analysis for Large Boiling Water Reactors, Licensing Topical Report, March 1972 (NEDO-10527).
!3 V'
5-50 .
- l. _) 5-30' Letter, J. F. Quirk (GE) to T. P. Speis (NRC), "0DYN Improvements",
October 13, 1981.
5-31 Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Change in GE Methods for Analysis of Cold Water Injection Transients", September 30, 1980.
5-32 Letter, R. C. Tedesco (NRC) to G. G. Sherwood (GE), " Acceptance for g Referencing General' Electric Licensing Topical Report NEDO-24154/
~
NEDE-24154P", February 4, 1981.
5-33 Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment Methods for Determination of Operating Limits", January 19, 1981.
~
5-34 Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), "End of Cycle Coastdown Analyzed With ODYN/TASC", September 1, 1981.
5-35 Letter, H. C. Pfefferlen (GE) to D. G. Eisenhut (NRC), " Correction of ODYN Errors", June 8, 1982. ,
~~
5-36 General Electric Company Model for Loss-of-Coolant Analysia in Accordance uith 10CFRSO, Appendix K, Amen &nent No. 2 - One Recirculation m Loop Out-of-Service, General Electric Company, Revision 1. July 1978 2 (NEDO-20566-2).
5-37 Letter, R. E. Engel (GE) to D. B. Vassallo (NRC), " Control Rod f- .s Drop Accident," February 24, 1982.
Q) n n
v 5-51a
.- - - ~ _ . - . - - - - _
. Table 5-15 FOUR-LOOP OPERATION MAPLHGR MULTIPLIERS (All GE P8x8R. Fuel Types)
MAPLHGR Idle Loop Condition Exposure Range -Multiplier Unisolatied 5000 1 E 115,000 mwd /t 1.00 Unisolated E <5000 or E >15,000 mwd /c 0.98 Isolated All Exposures 0.98 6
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