ML20235T482
| ML20235T482 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/02/1987 |
| From: | Auve S, Buckley J, Willie Lee PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20235T472 | List: |
| References | |
| PECO-FMS-0003, PECO-FMS-3, NUDOCS 8707220154 | |
| Download: ML20235T482 (65) | |
Text
{{#Wiki_filter:- _ _ _ _ _ - _ _ _ _ _ _ _ _ l 1 l STEADY-STATE FUEL PERFORMANCE - l METHODS REPORT Prepared By: - fc/ [o /Z/,/47 J / P'. Backley // Date' F/igineer r Fuel Management Section Reviewed By: S. A. Auve' [ Date 30 [ Engineer Fuel Management Section Reviewed By: ' s 7 [7 H W . Lee JUhte ' Senior Engineer Fuel Safety & Technology Branch Fuel Management Section Approved By: L. F. Rubino W 2 Date 7 Engineer-in-Charge Fuel Management Section OPERATING LICENSE DPR-44 AND DPR-56 Philadelphia Electric Company Nuclear Operations Nuclear Support Department 2301 Market Street 8707220154 870713 Philadelphia, PA 19101 ADOCK 0500 7-PDR P
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I l 1 DISCLAIMER This document was prepared by Philadelphia Electric
' Company and is believed to be true and accurate to the best of Cito knowledge and information. This document and the l
information contained herein are authorized for use only by - g .. ;p Philadelphia Electric Company and/or the appropriate Subdivisions within the U.S. Nuclear Regulatory Commission for; , rcview purposes., With regard to any unauthorized use whatsoever, Philadelphia Electric Company and its officers, directors, cgsnts, and employees assume no liability nor make any warranty or representation with regard to the contents of this document or its accuracy or completeness. i
ACKNOWLEDGEMENT The author would like'to thank Yankee Atomic Electric
' Company.for' developing and qualifying the FROSSTEY code and for.
making it available to the utility industry. Also I would like
.to thank Scott Auve' for his review and comments during the
.3 preparation of this report. .;3 0 ii
)
ABSTRACT Steady-state fuel performance analysis of the Peach Bottom ' Atomic Power Station, Unit 3, Cycle 7, has been performed using th2 FROSSTEY computer code. The Philadelphia Electric Company's ability to use the FROSSTEY code to predict fission gas release and fuel temperature, and by implication gap conductance,.is demonstrated through comparisons of predicted quantities for various test reactor rods to their measured data. iii _________________________D
TABLE OF CONTENTS PAGE DISCLAIMER ................................................. 1 ACKNOWLEDGEMENTS ...... .................................... 11 e o anASSTRACT ................,.................................. iii
- o ,n TABLE OF CONTENTS .......................................... ~iv LIST OF TABLES ............................................. V LIST OP PIGURES ............................................ vi-
1.0 INTRODUCTION
......................................... 1 1.1 Purpose ....................................... 1 i
-1.2 Description of PROSSTEY ....................... 2
{ 2.O PROSSTEY COMPARISONS TO MEASURED TEST REACTOR DATA ....- 4 2.1 Description of Test Reactor Rods ............... 4 2.2 Comparison of the PROSSTEY Code Results to Measured Test Data ............................ 7 3.0 PEACH BOTTOM 3 CYCLE 7 - FROSSTEY APPLICATION ........ 9 3.1 Core Average Gap Conductance .................. 9 l 3.2 Hot Channel Gap Conductance ................... 10 3.3 Fuel Temperature .............................. 11 4.0
SUMMARY
AND CONCLUSIONS .............................. 12
5.0 REFERENCES
........................................... 53 iv
LIST OF TABLES TABLE NUMBERS DESCRIPTIONS PAGE Table 1 Test Reactor Rods: Measured Data Available 13 ) Table:2 Test Reactor Rods: Fission-Gas Release Cossparisons 14
, Table 3 Test Reactor. Rods: Fuel Temperature Comparisons 15 Table 4 Peach Bottom 3 Cycle 7 Gap Conductances 16 Table 5 Peach Bottom 3 Cycle 7 Hot Channel Gap Conductances 17 1
O l i v
LIST OF FIGURES FICURE NUMBERS- DESCRIPTIONS % PAGE's
< w,, ' Figure 1 Halden Test Assembly IPA-513 Rod 1 18 o . Fuel Temperature Comparison M#
Figure 2 Halden' Test Assembly IFA-513 Rod 2 19 Fuel Temperature Comparison '39-Figure 3. Halden Test Assembly IPA-513 Rod 6 20 Fuel Temperature Comparison Figure-4 Halden Test Assembly IFA-431 Rod 1 21 Fuel Temperature Comparison
' Figure 5 Balden Test Assembly IFA-431 Rod 2 22 Fuel' Temperature Comparison 1 Figure 6 Halden Test' Assembly IFA-431 Rod 3 23 Fuel Temperature Comparison . Figure 7 Halden Test Assembly IFA-431 Rod 5 24 Fuel Temperature Comparison Figure'8 Halden Test Assembly IFA-431 Rod 6 25 Fuel Temperature Comparison Figure 9 Halden Test Assembly IPA-527 Rod 1 26 Fuel Temperature comparison Figure 10 Halden Test Assembly IPA-432 Rod 1 27 Fuel Temperature Comparison . Figure 11 Balden Test Assembly IFA-432 Rod 2 28 Fuel Temperature Comparison-Figure 12 Halden Test Assembly IFA-432 Rod 3 29 Fuel Temperature Comparison Figure 13 Halden Test Assembly IFA-432 Rod 5 30 Fuel Temperature Comparison Figure 14 Halden Test Assembly IFA-432 Rod 6 31 Fuel Temperature Comparison L Figure 15 Test Rod Data Fuel Temperature 32 Comparison Including IPA-432 Rod 6 Figure 15-A Test Rod Data Fuel Temperature Comparison 33 Not Including IFA-432 Rod 6 vi L
. m i
e' k J 1 LIST OF FIGURES-FICURE-NUMBERS: - DESCRIPTIONS: ' v.PAGE Figure 16 Batch 8 Average Gap Conductance for- 1 34 Peach Bottom 3 Cycle.7 , Figure'17 '
' Batch 9 Average Gap Conductance For 35 Peach Bottom 3 Cycle 7
.Figuref18 Batch.10 Average Gap Conductance For 136' Peach Bottom 3 Cycle 7
. Figure-19 Batch ll Average Gap Conductance For 37 Peach. Bottom.3 Cycle 7 Figure 20 Batch 12 Average Gap Conductance For 38 Peach Bottom 3 Cycle 7 Figure 21 Batch.13LAverage Gap Conductance For 39' Peach-Bottom 3 Cycle 7
-Figure-22, Core Average Gap Conductance For- 40 Peach Bottom 3 Cycle 7 Figure 23 Batch 8 Average Hot Channel Gap' 41 Conductance For Peach-Bottom 3 Cycle 7 I
Figure 24 Batch 9 Average Bot Channel Gap 42 Conductance For Peach Bottom 3 Cycle 7 Figure 25 Batch 10 Average Hot Channel Gap . 43 Conductance For Peach Bottom 3 Cycle 7
-Figure 26 Batch 11 Average Bot Channel Gap 44 Conductance For Peach Botton'3 Cycle 7 j 1
Figure 27 Batch 12 Average Bot Channel Gap 45 Conductance For Peach Bottom 3 Cycle 7
! Figure 28 Batch 13 Average Hot Channel Gap 46 Conductance *or Peach Bottom 3 Cycle 7
' Figure 29 Peach Bottom 3 Cycle 7 Batch 8 47 Volume-Average Temperature vs. Burnup
, Figure 30 Peach Bottom 3 Cycle 7 Batch 9 48 Volume-Average Temperature vs. Burnup Figure 31- Peach Bottom 3 Cycle 7 Batch 10 49 Volume-Average Temperature vs. Burnup vil
LIST OF FIGURES FIGURE NUMBERS DESCRIPTIONS PAGE
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l Figure 32 Peach Bottom 3 Cycle 7 Batch 11 50 l Volume-Average Temperature vs. Burnup Figure 33 Peach Bottom 3 Cycle 7 Batch 12 51 i volume-Average Temperature vs. Burnup t Figure 34 Peach Bottom 3 Cycle 7 Batch 13 52 Volume-Average Temperature vs. Burnup , vili
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INTRODUCTION 1.1 PURPOSE The purpose of this report is to demonstrate the-3 Philadelphia Electric Company's (PEco) proficiency in the use of the computer code PROSSTEY(3) for calculation.of i the fuel rod temperatures and fuel rod pellet to cladding gap conductances which are primarily used in safety analyses. This report is not attempting to requalify PROSSTEY as the PECo version is identical,to the version i previously qualified by Yankee Atomic Electric Company and
. reviewed by the NRC staff (2).
The following information is included in this report:
- 1. Comparisons to the measured results of various test fuel rods in order to demonstrate PECo's competency in the use of the PROSSTEY code.
- 2. Summaries of the PECo methods used in calculating fuel temperature, core average gap conductance and hot channel gap conductance.
- 3. Results of sample cases for the analysis of the demonstration cycle, Peach Bottom 3 Cycle 7.
I
'l.2 Description of FROSSTEY FROSSTEY (Fuel Rod Steady-State Thermal Effects) is a s, w computer _ code developed by. Yankee Atomic Electric Company to predict-the thermal performance of fuel rods for use in safety analyses of light water reactors. PROSSTEY is -
specifically designed to provide fuel rod pellet-to-cladding gap conductance, fuel rod temperature - distribution, fuel rod dimensional changes, fission gas release, internal gas pressure, and stored energy predictions as a function of fuel rod operating history. Additionally, the above parameters may be evaluated as a function of fuel rod power level at any point in the fuel rod operating history. The parameters of primary importance are fission gas release and fuel rod temperature. Accurate prediction of these parameters against measured data will show PECo's ability to utilize the code in prediction of all other parameters due to their interdependence. The use of PROSSTEY in the PECo reload analysis effort will be limited to the calculation of fuel temperatures as well as hot channel and core average gap conductances for use in transient analyses.
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FROSSTEY considers a cylindrical fuel rod as composed of a discrete number of axial segments. For each axial-segment, PROSSTEY calculates values for the thermal parameters by solving the one-dimensional, radial, steady-state temperature-dependent heat transfer equation with i thermal hydraulic boundary conditions specified by the user. Durnup dependencies are modeled and updated at each exposure step. At the completion of each axial slice calculation, the new fuel dimensions are used to update the assumed gap conductance. The axial slice gap conductance is determined to be converged if the assumed and calculated values agree to within 1%. For each exposure step, the code calculates the rod fission gas release using the values calculated for each axlal segment. The gas mixture is updated on each outer iteration until the gas inventory values also agree to within 1%. FROSSTEY was developed and qualified by the Yankee Atomic Electric Company (1,3). FROSSTEY has been reviewed by the NRC and was approved for performing t licensing calculations for the Vermont Yankee Nuclear Power Station (2), q 1 F t 1 3
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'2.0 FROSSTEY COMPARISONS TO MEASURED TEST REACTOR DATA I
Various test reactor rods (Table 1) were modeled for the. FROSSTEY computer program. The basis for choosing these rods was that (a) they were BWR rods, (b) the measured data was readily ) l
.available, and (c) that they were not previously-modeled"ih" Yankee =-
vgj% Atomic Electric Company's Qualification ReportIII. ,sym gr 2.1 Description of Test Reactor Data aW" The following are short summary descriptions of the i experiments and their purposes. 2.1.1 Studsvik Inter-Ramp Rods The objective of the program was to i experimentally investigate the failure propensity of- ' typical, unpressurized 8x8 BWR fuel rods when l subjected to fast power ramps. For those rods that
, survived the ramp. *., fission gas release data was
! j obtainedbypunctgringtherods.
The rods were fairly short (15.8") and were base-irradiated alternately at high (10-12 kw/ft) and low (8-10 kw/ft) power levels for 300 or 500-plus days. After the base-irradiation, the rods were subjected to a fast power increase (ramped) to high power levels and held at the high power levels for 12 hours or until failure occurred. The rods used by i FEco are those$that survived the ramps.
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r.. - 1 2.1. 2 ' 'Monticello BBEP/GE Rods-The main objective.of the High.Burnup Effects Program (HBEP) is to obtain good quality fuel. performance data, with an emphasis on fission gas-release,Lat high burnups. The. rods modeled by PBco,:8D10-1, 8D10-2'and 8D14-2 are short. rod segments manufactured by.GE,' and'had been base-irradiated-in'the'Monticello reactor to burnup levels of approximate 1y'30 Gwd/MTU. After base-irradiation, 8D10-1 was shipped to Studsvik and was bumped (a power increase to high power level)!in the R2 reactor to simulate a power transient then punctured to obtain fission gas release data. The rods 8D10-2'and' 8D14-2 were not bumped.and were punctured to obtain; fission gas release data after base-irradiation. 2.1.3 Peach Bottom 2 Rods This data was generated from an EPRI-sponsored BWR Fuel Rod Performance Evaluation j Program. The' main objective was to obtain fuel performance data including fission gas release, rod dimensional change and cladding oxide thickness i measurement of modern BWR fuel. 1 5
The bundles LJLTA-2 and LJLTA-3 were symmetrical bundles within the core. Fission gas release measurements were made on the rod F3 from , bundle LJLTA-2 at the end of two cycles (14.37
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mwd /KgU rod exposure) and on the same rod'(F3) from bundle LJLTA-3 at the end of.three cycles (25.70[ mwd /KgU rod exposure). They were typical unpressurized GE 8x8 fuel rods. 2.1.4 Halden Test Assemblies IPA-513, 527, 431 and 432 These were NRC sponsored experiments. The experiments were performed at the Halden reactor. The main objectives of these tests were to measure fuel centerline temperatures and internal pressures (or fission gas release). These objectives were accomplished by placing thermocouple and pressure sensors within the fuel rods. The details of these rods are contained in References 10 through 15 and incorporated herein by reference. 6
I: 2.2 Comparison of the PROSSTEY Code Results to Measured Test-Data The.results of the comparison between PBCo's' FROSSTEY calculations and the measured test data are reported both in tabular and graphic. form. Table 2 presents the fission gas release comparisons for all the. test reactor rods which fission gas release data was available. The'results are presented both with and without Halden test assembly'IFA-432 rod 6 as it had a low density, unstable' fuel which is not typical of modern BWR fuel designs and therefore had significant densification and-relocation during its irradiation. Due to this unstable fuel design, the FROSSTEY fission gas release prediction did not match as well as the other test rods. Halden test assembly IFA-431 rod 6 also had an unstable fuel design, however the results of the PROSSTEY calculations were not separated out due to their limited impact on the overall results. The measured release value for IFA-432 rod I was estimated as per Reference 15, page G.3. The fission gas release comparison mean predicted to measured difference for all the test rods was 1.29%. Without IFA-432 rod 6 the
'i mean predicted to measured difference is reduced to 0.39%.
l, 7
l-Table 3 presents the fuel temperature comparisons for all the test reactor rods for which thermocouple' data is available. .The measured thermocouple data has been corrected for thermocouple decalibration. These comparisons are also shown graphically in Figures 1 thru
- 14. Statistics were generated for the fuel temperature comparisons and are contained in Table 3. The overall mean difference for all the rods was 162.6 O F higher:than;**
measured. Without the Halden test assembly IFA-432 rod 6 results, the mean difference drops to 56.0 O F over-prediction.- A plot of'all fuel temperature comparisons'is presented in Figure 15. The results without Halden test assembly IFA-432 rod 6 are presented in Figure 15-A. The comparisons show that the PECo FROSSTEY predictions of fuel temperature and fission gas release agree very well with the measured data. 8 l _ _ _
3.0 PEACH BOTTOM 3 CYCLE 7 PROSSTEY APPLICATION Peach Bottom Unit 3 Cycle 7 was selected to demonstrate ; PECo's reload licensing application of PROSSTEY to obtain core average gap conductances, hot channel gap conductances and fuel 3 temperatures. In PECo's application of PROSSTEY, each bundle in the core was assigned to a batch according to its fuel type and when it was loaded into the core. Each batch was then separated into a typical UO2 rod and a typical UO -Gd 2 023 (gad) rod for which PROSSTEY input models were developed. These fuel rod input models along with an application specific power history were then input into PROSSTEY to obtain the required gap conductance or fuel temperature values. 3.1 Core Average Gap Conductance The core average gap conductance is used in the transient analysis system model and affects the amount of stored energy which potentially can be released in a plant transient. As such smaller values of this parameter are generally considered conservative for licensing evaluation of CPR-limited events. 1 I l 9
For Peach' Bottom 3 Cycle 7, the core average gap conductance was calculated at three cycle statepoints (Table 4) using the weighted average of each batch in the core. The power history was batch dependent covering each cycle that the batch was in the core. A Haling power shape. o was utilized to give best estimate results. The results are presented in Table 4 and plotted in Figures 16 through 22. 3.2 Bot Channel Gap Conductance The hot channel gap conductance is used in the transient analysis hot channel model to predict the change in transient CPR for the most limiting bundle in the core. As such, larger values of the hot channel gap conductance tend to be more conservative since the change in transient CPR will be more rapid and larger. The hot -channel bundle is identified as the bundle with the maximum fraction of limiting CPR or highest power. Hot channel gap conductances is calculated for each fuel type on a cycle independent basis. The power history is fuel type dependent and utilizes a 1.4 chopped cosine power shape. The hot channel gap conductance is determined for each fuel type required for the transient analysis hot channel model by: 1 0 1 10
3 n o
- 1) determining the exposure range.seen by the limiting bundle for each'of the three' cycle statepoints (Table 5), each statepoint being representative of an exposure range in the cycle of interest, and 2)' ' extracting the maximum value of gap conductance over the range from the cycle independent analyses.
The hot channel bundle exposure was found for' various exposure points in the cycle utilizing PECo's version of the computer code SIMULATE-EI4),-developed for.three-dimensional steady. state nodal analysis of light-water. reactors. The results of each batch's hot channel gap conductance analysis are presented in Figures 23 through 28 and-the core wide hot channel gap conductance results are presented in Table 5. 3.3 Fuel Temp y ture Fuel temperatures were calculated for each batch in the core over the batch lifetime. The power histories were batch dependent and utilized a Baling power shape. From these PROSSTEY runs, a fuel temperature can be found at any point in the batch's lifetime. Also at any point in the batch lifetime, the fuel temperature during a power ramp could be calculated. Figures 29 through 34 illustrate the fuel temperature for each batch versus exposure. I 11
I 1 4.0
SUMMARY
AND CONCLUSIONS { PROSSTEY fuel rod modEls were developed by PECo for various I test reactor rods as well as for each batch type in Peach Bottom i Unit 3 Cycle 7 in order to perform steady-state fuel performance analysis of the rods. A number of comparisons to measured test reactor data were made in order to demonstrate PECo's proficiency in utilizing the computer code PROSSTEY., The results demonstrate that PROSSTEY can be used accurately by PECo to determine gap conductances, both core average and hot channel, as well as fuel temperatures for use in core reload design and licensing calculations. 12
TABLE 1: Test Reactor Rods: Measured Data Available Fission Fuel Test Reactor Rod identifier Gas Tennp . 4 Studsvik Inter-RoTp Rod HR2 X HR4 X HR5 X LS2 X-TR1 X LR1 X MonticeiIo HBEP/GE Rod 8D10-1 X 8D10-2 X 8D14-2 X Peoch Botiorn 2 LJLTA-2 & -3 Rod F3 X Ha l den Test Asserrb l y IFA-513 Rod 1 X Rod 2 X Rod 6 X i Halden Test Assembly IFA-527 X Ha l den Tes t Asserrb l y IFA-431 Rod 1 X X Rod 2 X X Rod 3 X Rod 5 X ; Rod 6 X X Ha l den Tes t As serrb l y IFA-432 Rod 1 X X Rod 2 X Rod 3 X Rod 5 X Rod 6 X X 13
m TABLE 2: Test Reactor Rods: Fission Gas Release Comparisons Test Reactor Rod (Pre et d E Meahured) Studsvik inter-RaTp Rod HR2 0.36 i Rod HR4 -3.07 - - Rod HR5 0.74 Rod LS2 -0.94 Rod TR1 1.19 Rod LR1 -0.50 ! Mean -0.37 Monticello HBEP/GE 8D10-1 1.50 8D10-2 4.66 8D14-2 5.81 Mean 3.99 l Halden ' Test AsseTblies IFA-431 Rod 1 0.88 IFA-431 Rod 5 2.34 IFA-431 Rod 6 4.31 IFA-432 Rod 1 -4.89 I IFA-432 Rod 6 14.79 Mean 3.49 i Mean W/out IFA-432 Rod 6 0.66 i Peach Bottan 2 Rod F3 Bundle LJLTA-2 1.06 Bundle LJLTA-3 -7.53 Mean -3.24 l l Overall Mean Difference - 1.29 W/out IFA-432 Rod 6 - 0.39 14 l - - - - - - - - - -
1 a l l TABLE 3: Test Reactor Rods: Fuel _ Temperature Comparisons , I Halden Test Final Local Number Mean FROSSTEY Std. Assemblies of Overarediction Dev. l Burnup) (Mwd /t l Points (Deg. F) (Deg. Fj IFA-431 1 UTC* 5348 0 15 -31.52 38 75 LTC 3859.0 15 -73.13 - 39.32 2 UTC , 5552.0 15 99.73 78 26 LTC ! 4002 0 l 15 ! 26.47 48 96 3 UTC 5794.0 l 15 138.72 35.33 LTC 4170 0 1 15 71 74 20 68 5 UTC 5371.0 l 15 -13.55 74 62 LTC 3834 0 j 15 32.98 61.75 6 UTC 5251.0 l 15 403.62 188.44 LTC 3823.0 15 I 370.98 204 97
! {
'lFA-432 1 UTC 9914 0 i 20 i -107 24 99 18 LTC 24689.0 . 105 -15.95 120 89 2 LTC 24234 0 ! 105 155.21 ~153.99 3 UTC i 26858 0 l 77 -9.54 94.55 LTC 24879.0 105- - 54.76 102.31 l218 04 5 UTC l 7060 0 l 11 l 28 68 LTC i 25067.0 , 105 l 163.64 l200.40 6 UTC 8289.0 16 841 40 121 59 LTC i 22330.0 I 90 l 987 08 I 95.60 IFA-513 1 UTC 11521.0 ! 22 -79.55 178 02 LTC 9483.0 l 22 14 57 174 97 2 LTC 9213.0 , 22 7 69 83.91 6 UTC 11140 0 I I 18 -168 99 338 73 LTC 22 9460 0 -30.09 239.82 IFA-527 1 UTC 382 0 7 -15.86 50.34 LTC 555.0 7 -86 08 92.26 l Overall Summory l 904 l 162.61 340 40 l I Overcil Summary - 798 . 56.02 180 55 (w/out iFA-432 ! ; rod 6)
, j l o UTC - Upper Thermocouple LTC - Lower Thermocouple 15
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TABLE 5: Peach Bottom 3 Cycle 7 Hot Channel Gao Conductances State Hot Channei Bundie Gap Conductance Point Batch Exposure btu ; Nttrber (Mwo/sT) Hr-Ft2 *F I i I BOC' 10 , 10022 4123.5 ; j i i : ETPL i 13 ! 8402 4055.0
-2000 l l EOFPL I 13 ! 10738 ! 4097.5 l l
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- BOC Beginning-Of-CycIe io End-Of-FuIl-Power-Life (ETPL) minus 3000 MND/ST EOFPL-2000 ETPL-3000 MNJ/ST to ETPL-2000 NWJ/ST !
ETPL ETPL-2000 MAD /STtoETPL i 17 l
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i
5.0 REFERENCES
- 1. ' Methods for the Analysis of Oxide Puel Rod Steady-State Thermal Effects (PROSSTEY), Code Qualification and '
Application', YAEC-1265P, June, 1981. i
- 2. ' Safety Evaluation of the Yankee Atomic Electric Company, I Topical Reports YAEC-1249P and YAEC-1265P, PROSSTEY, l Methods for the Analysis of Oxide Puel Rod Steady-State I Thermal Effects', U.S. NRC, June 1984, Washington, D.C.
- 3. ' Methods for the Analysis of Oxide Puel Rod Steady-State Thermal Effects (PROSSTEY), Code /Model Description Manual',
l YAEC-1249P, April, 1981. 4
- 4. SIMULATE-E: User's Manual, EPRI NP-2792-CCM.
1
- 5. E. Larsson, 'The Inter-Ramp Test Puel Rods; Fabrication and Characterization Report', STIR-4, February 1979.
I
- 6. J. Gy11ander, 'BIRP1; Pre-Ramp Irradiation Report', STIR-16, June 1977.
- 7. J. Gy11ander, 'BIRP3; Pre-Ramp Irradiation Report', STIR-22, June 1977, i
J
- 8. J. Gyllander, 'BIRP2; Pre-Ramp Irradiation Report', STIR-31, April 1978.
1
- 9. H. Mogard, et. al., ' Final Report of the Inter-Ramp Project', STIR-53, August 1979.
- 10. R.L. Young, ' Fabrication, Pre-Irradiation Characterization and Irradiation Bistory of GE Rods - Task 2A, Task 2B 1 (Partial) and Task 2C', EBEP-06(2GI), April 1981. j
- 11. G. Ronnberg, B. Nilsson and A. Brisling, ' Power Bumping of GE Rods - Task 2C (Group 1)', HBEP-08(2G3), April 1981. l
- 12. D.M. Boyt, 'NDT Examinations of GE Rodlets - Task 2B and Pre-Bump Task 2C (Group 2)', EBEP-34(2G6), February 1984.
53
s
;!g .
J J
- 13. D.M. Boyt, 'NDT Examination of GE Rods-Task 2A and Prebump Task 2C (Group 1)', HBEP-07!2G2), April 1981.
- e
- 14. D.M. Boyt and L.A. Hanson, 'Postirradiation Examination of
.GE Rods-Task 2A and Task 2C (Group 1)', HBEP-13(2G4),
November 1981.
- 15. J.L. Daniel, ' Archive Fuel Characterization-Task 2 and Task HBEP-12(2/3P3), January 1982.
3',
- 16. D.D. Lanning et al, ' Qualification of Fission Gas Release *
, Data from Task 2 Rods', HBEP-25(2P4~) Revision 2, October 1984. t
- 17. R.D. Grimoldby and B. Crossley, ' Pre-Irradiation
}.. Characterization of BNFL Rods-Tack 2A', HBEP-04 (2B1),
- April 1981.
/
- 18. C. Winne, ' Irradiation Bistory of BNFL Rods-Task 2A', EBEP-05(2B2), April 1981.
- 19. D.K. Dennison and D.O. Sheppard, ' Task A: The Pre-Irradiation Characterization of Selected Lead Test Assembly Fuel Rods and Channels', General Electric NEDC-21528 Class 1, March 1977.
- 20. D.O. Sheppard, ' Task C: First Cycle Interim Site ,
Examination of the Lead Test Assembly Fuel and Channels at Peach Bottom 2', General Electric Report NEDC-23719, October 1977.
- 21. D.O. Sheppard, ' Task C: Second Cycle Interim Site Examination of the Lead Test Assembly Fuel and Channels at Peach Bottom 2', General Electric Report NEDC-24609, February 1979.
/ 22. J.E. Gonser, T.C. Rowland and D.K. Dennison, ' Tasks C and E
)
- Third Interim Site Examination of the Lead Test Assembly Fuel and Channels at Peach Bottom 2', General Electric Report NEDC-25440 Class 1, July 1981.
)
- 23. T.C. Rowland, 'BWR Fuel Rod Performance Evaluation P rog r aia ' , EPRI NP-4602, May 1986.
54 f .,., ).-
- 24. E.R. Bradley, et. al., ' Pre-Characterization Report for Instrumented Nuclear Fuel Assembly IFA-513', NUREG/CR-1077 PNL-3156 R-3, November 1979.
- 25. E.R. Bradley, et. al., ' Data Report for the Instrumented Fuel Assembly IFA-513', NUREG/CR-1838 PNL-3637, August 1981.
- 26. M.E. Cunningham, et. al., ' Pre-Characterization Report for Instrumented Fuel Assembly (IFA)-527', NUREG/CR-2168 PNL-3824 R-3, July 1981.
- 27. M.E. Cunningham, et. al., 'End-of-Irradiation Data Report for the Instrumented Fuel Assembly (IPA)-527', NUREG/CR-2600 PNL-4201 R-3, May 1982.
l l
- 28. C.R. Hann, et. al., ' Test Design, Pre-Characterization, and Fuel Assembly Fabrication for Instrumented Fuel Assemblies IFA-431 and IFA-432', NUREG/CR-0332 BNNL-1988 R-3, November
, 1977.
- 29. D.D. Lanning and E.R. Bradley, ' Irradiation History and Interim Post Irradiation Data for IFA-432', NUREG/CR-3071 PNL-4543 R-3, March 1984.
- 30. D.D Lanning and E. R. dradley, ' Retained Fission Gas Measurements from IPA-432 Rods 1 and 6', PNL-SA-12217, May 1984.
l 55
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