ML19343A983
| ML19343A983 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/31/1980 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19343A975 | List: |
| References | |
| NEDO-24855, NUDOCS 8011240381 | |
| Download: ML19343A983 (23) | |
Text
_ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ - - _ _ - - - - _ - _ _ - - - - -
- Arr^cumeur 1
, N EDO-24355 9 80NE0072 CLASS I AUGUST 1980
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QUAD CITIES-1 NUCLEAR GENERATING PLANT , . .
EXTENDED BURNUP FUEL PROGRAM N l
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DISCLAIMER O" RESPONSIBILITY This document was prepared by or for the General Electric Company. Neither the General Electric Company nor any of the contributore to this document:
A. Makes any varranty or representation, express or implied, uith respect to the accuracy, completenese, or usefulnese of the information contained in this document, or that the use of any information diectosed in this document may not infringe privately ovned righte; or B. Accumes any rceponsibility for liability or
- damage of any kind uhich nay result from the use of any information diccioced in thie document.
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- 1. PROPOSED PROGRAM The decision by the federal government to delay nuclear fuel re-processing and plutonium recycle has necessitated development of designs that reduce uranium ore requirements and fuel cycle costs.
An increase in discharge fuel burnup is one method which can be used to reduce uranium needs and improve fuel cycle costs. Increase of hydrogen-to-uranium ratio by use of annular fuel pellets also has potential to improve uranium utilization and performance. Current licensing analyses for General Electric reload fuel are performed for peak pellet exposures up to 40,000 mwd /ST.* The Quad Cities-1 extended burnup Advanced Fuel Test Bundle (AFTB) program is one of several anticipated programs whereby.
2 lead burnup bundles will be extended to peak-pellet exposures of 50,000 mwd /ST or higher.
'nformation to be obtained from these programs will be used to systematically determine the impact on fuel reliability and weigh the advantages of extended burnup relative to other uranium utili-zation improvement methods.
The Quad Cities-1 fuel bundles also contain 5 annular pellet fuel rods in each bundle which have been fabricated with the same process as adjacent solid pellet fuel rods.
The two fuel assemblies to be operationally extended were first inserted into Quad Cities at the beginning of Cycle 2 (Reload 1) in 1974. These assemblies have been under irradiation in a joint EPRI/GE/ Commonwealth Edison program and have been the most carefully pre-cheracterized and measured fuel assemblies ever irradiated in a BWR.'~#hhe primary objectives and accomplishments under this program to date are summarized in Tables 1 and 2. Table 3 is a summary of some of the key measurements that have been made on this fuel. The program to date has significantly contributed to improved reactor performance and reliability by providing important data that has aided inLconfirming or improving design methods and predictions.
No fuel failures or defects have oomuTed in these fuel assemblies
- ST indicates short ton.
1-1
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O NEDO-24855 TABLE 1 QUAD CITIES 1 FUEL ASSEMBLY PERFORMANCE PROGRAM OBJECTIVES:
9 OBTAIN NUCLEAR DESIGN AND FUEL PERFORMANCE DATA FOR UO AND UO -pug FUEL RODS 2 2 2 9 VERIFY SAFETY, PERFORMANCE AND ECONOMICS OF UO AND UO2 -PuO FUEL'IN BWRs 2 2 8 EXPAND DATA BASE FOR FUEL BUNDLE NUCLEAR AND MECHANICAL PERFORMANCE THROUGH ON-SITE MEASUREMENTS 9 OBTAIN BENCHMARK DATA FOR METHODS VERIFICATION FROM MEASUREMENTS INVOLVING PRE-CHARACTERIZED FUEL BUNDLES WITH WELL-DEFINED IRRADIATION HISTORY e VERIFY AND COMPARE SOLID AND ANNULAR FUEL PELLET PERFORMANCE e OBTAIN FUEL ISOTOPIC COMPOSITION DATA AS A FUNCTION OF BURNUP FOR UO , UO -Gd O AND 2 2 23 UO2 -PuO 2 FUEL 1-2 k
. TABLE 2 OC-1-FUEL PERFORMANCE PROGRAM MAJOR ACCOMPLISHMENTS S
FIRST FUEL WHICH HAS BEEN LOCATED ON CORE CENTER AND CA SCANNED PIN-BY-PIN EVEhY CYCLE, COUPLED WITH PERIODIC 1/8 CORE SCANS, TO GIVE ACCURATE ROD AND PELLET POWER MEASURE-MENTS WITH WHICH TO BENCHMARK FISSION GAS RELEASE PRED S
FIRST IRRADIATION OF 12' LONG ANNULAR PELLET RODS.
9 FIRST SIDE-BY-SIDE IRRADIATION OF ANNULAR AND SOLID PELLETS FABRICATED BY SAME PROCESS S
FIRST DETAILED SPATIAL (AXIAL AND RADIAL WITHIN PELLETS)
GADOLINIA, BURNUP, AND HEAVY ELEMENT MEASUREtiENTS ON LOW BURNUP, UNDEPLETEC GADOLINIA PINS 9
I FIRST FAST AND THERMAL AXIAL AND RADIAL FLUX MEASUREMENT A SIMULATED ONE STUCK CONTROL ROD COLD CRITICAL IN AN OPERATING REACTOR S
FIRST FISSION GAS MEASUREMENTS ON A Gd PIN 4
FIRST HIGH BURNUP DATA PROTOTYPIC OF BWR/5-6 PEAK KW/FT O 9
FIRST DETAILED CHANNEL DIMENSIONAL MEASUREMENTS 9-FIRST DETAILED ROD GROWTH MEASUREMENTS FOR SIDE-BY-SIDE IRRADIATION OF TWO TYPES OF CLADDING i
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. NEDO-24855 TABLE 3 QUAD CITIES 1 FUEL PERFORMANCE PROGRAM PROGRAM FOCUS:
IRRADIATION OF ADVANCED FUEL TEST BUNDLES (AFTBs)
OCl CYCLE OF OPERATION 2 3 4 5 6 AFTB IRRADIATION HISTORY 7/74-1/76 3/76-3/77 5/77-1/79 3/79-9/80 1/81-9/8 O Bundle Sxp. (Ctr.)
8.3GhD/ST 13.6GhD/ST 21.4GC/ST 428GsD/ST %34GO/S O Peak Pellet Exp. 14.4GO/ST 21.2Gh'D/ST 31.lGhD/ST %40GC/ST %50GhD/S O Peak Pellet LHGR 12.9 kw/ft 11.6 kw/ft 11.0 k2/ft 10 kw/ft 9 kw/f AFTB DUNDLE HISTORY DATA O OCD&M Edits 2nthly Weekly Weekly Weekly Weekly O Process Computer hly (Daily Edits Available) - --
0 FADAS System ?b No R) Yes Yes AFTB EOC SITE MEASUREMENTS O AFTB y-Scans 5 Bundles 5 Bundles 5 Bundles 5 Bundles 3 Bundle O ROD y-Scans 2 Bundles 2 Bundles 2 Bundles 2 Bundles 2 Btndle O Rod Profilometry 28 Ibds --
27 Ibds 27 Pods 27 Pods O Rod NDT 28 Ibds 27 Bods 27 Pods 27 Pods 27 Pcds O Rod Length 28 Ibds 77 Pods N75 Bods Selected Seltxde<
O Channel Dimensions 5 Channels 5 Channels 5 Channels 5 Channels 3Cberne:
O Fuel Isotopics 15 Ibds - --
15 Pods 15 Ibes O Fission Gas 2 Pods - -
15 Pods 15 Rxis O Axial Flux Meas. -
Yes -
Proposed -
0 Neutrography 8 Fods - -
Proposed Pmpora 1-4 t /
,. 1 during their irradiation in the ^2ad Cities-1 reactor. Additional fuels and nuclear measurements will be made on these fuel assemblies during the forthcoming refueling _ outage, scheduled to start in September 1980.
.It is inter ded that two of the four centra 1' assemblies currently in the reactor core will be extended for an additional cycle, provided the measurements on these fuel assemblies continue to indicate satisfactory reliabic performance.
Theme two fuel assemblies will be loaded in the central. centerline core cositions whien thev nave occupied throuch-out tnelr operating nistory.- Tnis-location nas contrioutea to tne value and reliability.of the fuel and nuclear measurements.
Measurements which will be made on one or both of these fuel assemblies during the refueling outage include:
o Pin-by-pin gamma scan at several axial planes to measure. local planer power distribution e Global gamma scan to measure average axial power distribution e Fuel bundle and rod visual inspection e Rod length e Sipping to detect any defects It is. anticipated that some minor bundle adjustments or modifications may be necessary to accommodate the differential irradiation growth of the fuel rods. Such adjustments also have been made during the previous two refueling outages to assure proper interface of the rods with the upper tie plate. The adjustments will be based upon detailed measurements to be made during the examination prior to Cycle 6.
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NEDO-24855 After Cycle 6, the integrity of the extended burnup fuel bundles will be again ascertained and the fuel assemblies will be dischmged.
Both fuel bundles will exceed 40,000 mwd /ST peak pellet exposure during cycle 6, with the highest exposure annular fuel pellets exceeding 50,000 mwd /ST.- Other detailed fuels, nuclear, isotopic and burnup measurements and hot cell examinations also are tentatively scheduled to be completed at the end of Cycle 6 and after discharge.
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- 2. FUEL MECHANICAL DESIGN ANALYSIS Burnup-dependent fuel mechanical design analyses for the extended burnup bundles have been performed for conditions which meet or exceed
- expected Cycle 6 operating conditions. Models and assumptions used in_these analyses are those. documented in Reference 1. The results are given below.
2.1 FUEL ROD THERMAL ANALYSIS Safety evaluations are performed and measured against established.
conservative safety criteria. The consequence of calculating values which exceed such criteria is that fuel failure must be assumed to occur. For plant normal and abnormal operation, this is not permissible. Fuel 1 failure is defined as a perforation of the cladding.which would permit the release of fission products to the reactcr coolant.- The mechanisms which could cause_ fuel damage in '
-reactor abnormal operation transients are: (1) rupture of the fuel rod cladding due to strain caused by relative expansion of the UO 2
pellet, and (2) severe overheating of the fuel rod cladding caused by inadequate cooling.
Avalueof1%plasticstrainoftheZircaloycladdinghasbeen ,
established as.the' safety limit below which fuel damage due to over-straining of the fuel claddingLis not expected to occur. The Fuel Cladding Integrity Safety Limit' ensures that fuel damage resulting from severe overheating of the fuel rod cladding . caused by inadequate cooling, is avoided. Of these criteria,.only:the linear heat genera-
" tion rate associated with the lt plastic strain safety limit is-affected by increased fuel burnup.- Analyses performed for the extended cournup_ fuel bundles resulted in' values .of 20.0 kw/ft at a peak-pellet exposure of 50,000 mwd /ST for UO2L fuel r ds, 18.9 kw/f t at _45 ,000 mwd /ST
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.for urania-gadolinia rods and 21.0 kw/f t at 57,000 mwd /ST for the i
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NEDO-24855 annular pellet fuo3' rods.-
All values include the 2.2% power' spiking penalty-documented in Reference 1. These linear heat generation rata values are used during specific evaluations of transients due to single operator error or equipment malfunction to ensure that the safety limit is not exceeded.
2.1.1 Fuel Cladding Temperatures .
The cladding surface temperature is calculated using the cladding surface heat flux at a given axial position on a fuel element'in conjunction with the overall cladding-to-coolant film coefficient.
The modals used are noted in Reference 1. The inside, average, and outside cladding temperature during normal operation at the end of Cycle 6 are calculated not to exceed 797*F, 766*F, and 734'F, respectively.
2.1.2 Fission Gas Release The amount of fission gas released during a time increment is i calculated based on the fission gas generated and fission gas release fraction.
The calculated maximum fission gas release traction in the extended exposure fuel rod with the most limiting peaking factors is
! less than 15%.
2.1.3 Incipient Center Melting The fuel is designed so that fuel melting is not expected to occur i
during normal steady state full power operation including extended burnup levels.
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NEDC-24855 2.2 FUEL ASSEMBLY MECHANICAL EVALUATIONS The fuel assembly is evaluated by analyses, tests and experience to demonstrate fuel assembly structural integrity. When analyses are used to demonstrate structural integrity, resulting stress and/or strain levels are compared to the associated mechanistic limits documented in Reference 1. Results of the fuel rod mechanical analyses of the normal and transient loads for the extended exposure fuel are given below.
2.2.1 Cladding Creep Collapse The measured ovality of the AFTB fuel rods has been small to peak pellet burnups of 31,000 mwd /ST as shown in Table 4. This data, when applied to creep collapse evaluations, demonstrates that cladding creep collapse is not expected to occur in the event of a maximum over power transient through the end of Cycle 6.
2.2.2 Stress Evaluation Fuel rod stress analyses of the extended burnup bundles were performed with the model documented in Reference 1 for operation through Cycle 6.
These analyses showed that the fuel design ratios were well below 1.0.
2.2.3 Deflection Evaluation The operational fuel rod deflections considered are a result of manu-facturing tolerances, flow-induced vibration, thermal effects, and axial load. Deflections of the extended burnup fuel rods were combined and compar.ed to fuel rod-to-fuel rod and fuel rod-to-channel spacing deflection limits. This comparison demonstrated that the fuel rod clearance criterion was met.
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l NEDO 24855 '
TABLE 4 MAXIMUM OVALITY COMPARISON AFTB FUEL RODS POST IRRADIATION 1976-1979 Rod Serial Post-trradiation Post 4rrediation Amiel Location 1976 Number 1979 3 Change Bundia/ Rod (In.) - (mils) (mils) 1976-1979 AZS 0124 GEB 162/F4 6 0.3 BSG 1147 1.0 07 GEB 162/F5 137 1.6 2.0 AZN 0128 GEB 159/F7 97 04 1.9 2.0 BSB 0138 GEB 159/G1 9 01 1.1 1.6 BSE 0103 GEB 159/E1 47 0.5 1.1 21 BOG 0768 GEB 159/G- 1.0 13 2.5
( BSG 0735 GEB 159/C2 2.6 0.1 to 1.6 BSA 0139 GEB 159/A1 2.2 0.6 tot 08 BSB 0153 2.6 1.8 GEB 159/B1 9 2.5 2.7 AZN 0120 GES 159/D2 .
0.2 129 1.0 1.5 0.5 Rod Post-frradiation Post-Irradiation Serial Anlai Location Pre trradiation 1976 A Change 1979 Number Sundle/ Location A Change (in.) (mils) (mits) (mlis) (mils) Pre-1979 VP 053 GEB 162/DS 144 0.7 1.1 + 0. 4 0.5 VP 014 GEB 162/E5 - 0.2 58 03 1.3 + 1.0 1.0 + 0.7 VP 001 GEB 162/G4 132 0.7 1.1 + 0.4 1.0
' VP 039 GES 162/G7 + 0.3 124 0.6 0.8 + 0.2 1.6 + 1.0 VP 017 GEB 159/D5 22 1.4 2.0 +06 1.9 + 0.5 VP0'O GEB 159/D6 57 1.7 24 + 0.7 2.1 + 0. 8 VP 051 GEB 159/E4 76 1.3 2.6 - 1.3 VP 054 GEB 159/F4 30 + 1.7 60 2.3 3.7 VP 022
-16 4.0 + 1.7 GEB 159/F5 52 1.1 16 + 0.5 1.2 + 0.1 2-4'
- 4 NEDO-24855-2.2.4 Fatigue Evaluation
, The cyclic loads considered in cladding fatigue analysis are coolant pressure and thermal gradients. The analysis performed for the
^
i higher burnup fuel was based'on previous and projected operating cycles
.through the end of. cycle 6, maximum and minimum pressures, and the
. stresses determined ~in Subsection 2.2.2. The cumulative fatigue damage was calculated to be less than the allowable fatigue limit.
2.3 FUEL-ROD CORROSION, HYDRIDING AND FRETTING WEAR CONSIDERATIONS 2.3.1 Potential for Hydriding The potential for hydriding is discussed in Reference 1 and is not affected by higher fuel burnup.
2.3.2 Fuel Element Energy Release During cycle 6 these fuel assemblies are predicted to operate with very large boiling transition margins.
Significant boiling transition is not possible at normal operating conditions or under conditions of abnormal _ operational transients because of the thermal margins at which the fuel ~is operated at the high fuel burnups. It can, there-fore,.be concluded that the energy release and potential for a chemical reaction:is not an-important consideration during normal operation or-abnormal transients. The insignificant energy released in the event of b' oiling transition reported in Reference 1 does not change because of extended-fuel burnup.
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NEDO-24588 2.3.3 Fretting Wear and Corrosion As discussed in Reference 1, no significant fretting wear or corrosion has been observed throughout continuing fuel surveillance programs, including the specific visual inspections which have taken place at the ends-of-cycles 2 through 4 for the extended burnup, AFTB fuels.
Increased burnups are not expected to significantly change the observed results. The fuel bundles which will operate to higher burnups will also be visually examined before loading in Cycle 6.
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- 3. _ IMPACT ON RELOAD ANALYSES All of the models documented in Reference 1 are applicable for use with higher fuel burnups. Some inputs into these models are exposure-dependent, however, which, in turn, are reflected in calculated results.
A description of these exposure-dependent changes is given below.
3.1 NUCLEAR EVALUATIONS The nuclear etaluations are comprised of two analyses: lattice and core.
Most of the lattice analysis is performed during the fuel bundle design process. The results of these single bundle calculations are red.uced to " libraries" of lattice reactivities, relative rod powers, and few group cross-sections as functions of instantaneous void, exposure, exposure-void history, control state, and fuel and moderator temperature.
Because of this exposure dependence of these results, the libraries have included greater burnups beyond expected operation as noted below. The core analysis is unique for each reload.
It is performed using :he above lattice " libraries" in the months preceding the reload to demonstrate that the core meets all applicable safety limitr, The effects of higher fuel burnups are thus reflected in the 7are analysis results through use of the expanded " libraries".
3.1.1 Reactivity Traditionally, bundle reactivities have been expressed in terms of K.
(i.e. , the neutron multiplication of an infinite array of like bundles) .
This lattice reactivity is a function of lattice average enrichment, gadolinia loading, void fraction, hydrogen-to-uranium ratio and exposure.
.90t reactivity of the advanced fuel test bundles decreases by 0.05 AK.
from 35,000 mwd /ST to 45,000 mwd /ST.
3-1 k
NEDO-24855 3.1.2 Local Peaking Factors For a given lattice at a given void fraction, the maximum local peaking factor will occur at different fuel rods as the exposure increases. This is due to the different cepletion and generation rate of the various fissile nuclides in each fuel rod. Calculated maximum local peaking factor for the extended exposure fuel bundles increases by 0.04 from 35,000 mwd /ST to 45,000 mwd /ST.
The local peaking factor does vary with void' fraction, and this dependence is taken into account in the calculations used to assign local oaaking factors to each axial segment of the fuel. The above
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values are for O.40 void, as this is the typical average bundle void fraction.
The local peaking factor is a nominally calculated value. The nominally calculated values are appropriate, for conservative representa-tion of fuel mechanical-thermal performance.
3.1.3 Doppler Reactivity The Doppler coefficient is of prime importance in reactor safety. The Doppler coefficient is a measure of the reactivity change associated with an increase in the absorption of resonance-energy neutrons caused by a change in the temperature-of the material in question. The Doppler reactivity coefficient provides instantaneous negative reactivity feedback to any rise in fuel tamperature, on either a gross or local-basis.
Maximum and minin.am calculated Doppler coeffi-cients at several burnups are shown in Reference 1. The two bundles will have negligible effect on the core Doppler coefficient of reactivity.
3-2 L
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NEDO-24855 3.1.4 Void Effect The void reactivity coefficient affects stability and transient response.
The overall void coefficient is always negative over the complete operating range, since the BWR design is undermoderated. The void reactivity coefficient is not affected appreciably by extension of the burnup of the two AFTB fuels. The steam void reactivity coeffi-cients will be less negative an6 therefore more favorable in extended burnup operation due to reduced fuel assembly power and steam voids at extended burnup conditions.
3.2 STEADY-STATE HYDRAULIC ANALYSIS Cure steady-state thernal-hydraulic analyses are performed using a model of the reactor core, which includes hydraulic descriptions of orifices, lower tieplates, fuel rods, fuel rod spacers, upper tie plates the fuel channel and core bypass flow paths. Model details are documented in Reference 1.
The flow distribution to the fuel assemblies und bypass flow paths is calculated on the assumption that the pressure u.cp across all fuel assemblies and bypass flow paths is the same.
An iteration is performed on flow through each flow path (fuel assemblies and bypass paths), which equates the total differential pressure (plenum to plenum) across each path and matches the sum of the flows 'chrough each path to the total core flow. This analysis is insignificantly affected by extended burnup of the AFTB fuel.
3.3 REACTOR LIMITS DETERMINATION Limits on plant operation are established to assure that the plant can ba safely operated and not pose any undue risk to the health and safety of the public.
This is accomplished by demonstrating that radioactive releast from plants for normal operation, abnormal operational transients and postulated _ce'i ents meet applicable regulations in which conserva- !
tive limits are documented. This conservatism is augmented by using 3-3
. s NEDO-24855 conservative evaluation models and observing limits which are more restrictive than those documented in the applicable regulations.
These observed operating limits and methods ur,ed to determine if the limits are met are documented in Reference 1.
3.3.1 Fuel Cladding Integrity Safety Limit The generation of the Minimum Critical Power Ratio (MCPR) limit requirr s a statistical analysis of the core near the limiting MCPR e
condition.
Bounding statistical analyses have been performed which provide conservative safety limit MCPRs for operating BWR plants.
These safety limit MCPRs conservatively apply for all reload cycles including equilibrium cycle. Insertion of the two low-powered extended burnup AFTB fuel bundles dor,s not change the conclusions of these bounding analyses.
3.3.2 MCPR Operating Limits The MCPR operat ng limit is established to ensure that the fuel cladding integrity safety limit is not exceeded for any moderate frequency transient.
This operating requirement is obtained by addition of the absolute, maximum AMCPR value for the most limiting transient from rated conditions postulated to occur at the plant to the fuel cladding integrity safety limit. Higher fuel burnups are are reflected in the nuclear input data. However, due to the high burnup, these fuel assemblies will operate with relatively low power and will not be near MCPR operating limits.
3.3.3 Vessel Pressure ASME Code Compliance The Quad Cities pressure relief system is comprised of thirteen safety /
relief valves, nine of which are aEsumed to gyrabe in the analysis.
To assure that the peak allowable pressure of 110% of the vessel design pressure is not exceeded, the raost severe isolation event with indirect 3-4
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.' NEDO-24855 scram and credit for subsequent valve operation is evaluated. The event which satisfies this specification is the closure of all main steamline isolation valves (MSIV) with' indirect (flux) scram. The model used to analyze this event is described in Reference 1. Greatm burnup of the AFTB fue] has negligible effect on reactor pressure response.
3.3.4 Stability Analysis Two types of stability are examined utilizing a linearized analytical model. First, is the hydrodynamic channel stability of one or more types of channels operating in parallel with other channels in the core.
Second, is the reactivity feedback stability of the entire reactor core which also involves power oscillations. The assurance that the total plant is stable and, therefore, has significant design margin is demonstrated analytically when the acceptable performance limit of decay ratio less than 1.0 or a damping coefficient greater than 0.0 is met for each type of stability. These criteria must be satisfied for both ususi and unusual operating conditions of the reactor that may be encountered in the course of BWR plant operation.
The analysis is performed using the models documented in Reference 1 at the most limiting condition, which usually occurs near the end of cycle, with power peaking toward the bottom of the core. The most sensitive reactor operating condition is that corresponding to natural circulation flow and a power level equal to or greater than the rated rod pattern power level. Extended fuel burnup for the two AFTB fuel assemblies has negligible effect on core stability.
3.3.5 Accident Evaluations Accidents are events which have a projected frequency of occurrence of less than once in every one hundred years for every operating BWR. The l
broad spectrum of postulated accidents is covered by six categories of 3-5 i
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a design basis eventn.
These events are *.he control rod drop, main steam-line break, loss-of-coolant, refueling, recirculation pump seizure, and fuel assembly loading 3 accidents. Consequences of these events
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with the low-powered extended exp,osure fuel bundles are not as great as lower burnup bundles. cifowever, the MAPLHGR values for the AFTB fuel have been extended to an average planar burnup of 50,000 mwd /ST.
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- 4. REFERENCES
- 1. " Generic Reload Fuel Application, "NEDE-240ll-P-A, (August 1978. )
- 2. "Monticello Nuclear Generating Plant Extended Fuel Program" NEDO-24202 (July 1979)
- 3. RM Tilley, TP Choong, RL Crowther, HD Kosanke, NH Skarshaue, RA Wolters (GE E6n Jose), BA Zolotar (EPRI), "GE/EPRI Quad Cities 1 Nuclenr and Fuel Performance Measurements Following Third Irradiation Cycle", Trans. Am. Nucl. Soc., 33, 841 (1979).
3 RL Crowther, JJ Haskins, RD Reager, RE Smith, RM Tilley, RA Wolters (GE San Jose), BA Zoltor (EPRI), "Gadolinia Spatial Depletion Measurements in BWR Fuel", Trans. Am. Nucl. Soc., 28, 746 (1978)
R Petrie (Comm. Ed. ) , BA Zolotar (EPRI) , "GE-EPRI Quad Cities 1 Cold Critical Measurements", Trans. Am. Nucl. Soc., 28,795 (1978).
- 6. TP Choong, TA Keys, RD Reager, RE Smith, RM Tilley, RL Crowther, RA Wolters (GE San Jose) , BA Zolotar (EPRI), " Isotopic Measurements of BWR Irradiated in Quad Cities 1", Trans. Am. Nucl. Soc.,
30,'688 (1978).
- 7. RA Wolters, RL Crowther, D Sheppard, " Planning Support Document for EPRI LWR Fuel Performance Programs", EPRI Fuel Performance Contractors Meeting, Dallas, Texas, April 14, 1977.
- 8. RM Tilley,.RD Reager, RA Wolters, RE Smith, TP Choong, "Burnup and Transuranium Element Composition in Irradiated UO2, UO2-PuO2 Rods from the Quad Cities 1 Reactor", WF-7906-34, June 1979.
- 9. WE Baily, RL Crowther, SY Ogawa, RA Proebstle, et al, "Use of Plutonium Fuel in Boiling-Water Reactors, Interim Report, EPRI 72-2, June 1975.
O. RL Crowther,.CC McNeely, RM Tilley, RA Wolters 'GE San Jose),
BA Zolotor (EPRI), " Predicted Nuclear Characteristics and Core Nuclear Environment at Quad Cities Nuclear Power Station 1 for Cylce 1 and Beginning of Cycle 2", Final Report, October 1976.
1 4-1
- j. ,
NEDO 24855 11.
Martin B Cutrone, George F Valby '"" 3an Jose), Robert N Whitesel (EPRI), " Gamma Scan Measurements en Juad Cities Nuclear Fower Station 1 Following Cycle 2", NP-214, Final Report, July 1976.
- 12. NH Larsen, GR Parkos, O Raza, " Core Design and Operating Data for Cycles 1 and 2 of Quad Cities 1", NP-240, November 1976.
- 13. DW Merth (GE San Jose), BA Zolotar (EPRI), " Gamma Scan Measurements at Quad Cities Nuclear Power Station Unit 1 Following Cycle 3",
NP-512, July 1977.
BA Zolotar (EPRI), " Mixed-Oxide Fuel Bundle Performance Following for Cycle 2 Operation of Quad Cities Nuclear Power Station Unit 1, NP-553, January 1977.
- 15. DO Sheppard, NH Skarshaug (GE San Jose), BA Zolotar (EPRI), " Fuel Inspection Data Summary Report for Initial Mixed-Oxide Examination at Quad Cities 1", NP-550, April 1977.
- 16. DO Sheppard, NH Skarshaug, RA Wolters (GE San Jose) , E1 Zolotar (EPRI), " Mixed Oxide and Uranium Fueled Surveillance at Quad Cities Nuclear Power Station Unit 1 Following Cycle April 1977 3", NF-551,
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" Mixed-Oxide and Uranium Fuel Surveillance at Quad Cities Nuclear Power Station Unit 1 following Cycle 4", April 1979.
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21.
JT Ma (GE San Jose) , BA Zolotar (EPRI) , " Core Design and operating Data for Quad Cities 1 Cycle 4", October 1979/
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