ML20062E196

Suppl Reload Licensing Submittal for Browns Ferry Unit 1, Reload 2, Revision 1

ML20062E196
Person / Time
Site:	Browns Ferry
Issue date:	11/30/1978
From:	Freemon A GENERAL ELECTRIC CO.
To:
Shared Package
ML20062E190	List: ML20062E186 ML20062E196
References
NEDO-24136, NEDO-24136-R01, NEDO-24136-R1, NUDOCS 7812050251
Download: ML20062E196 (33)

Text

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r NEDO-24136' Class I Revision 1 November 1978 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY NUCLEAR PLANT UNIT I RELOAD 2 Prepared: f h% A. M. Freemon, Engineer Operating Licenses II Approv d: [,/ W-R. . Brugge, Mger Operating Licenses II 80iltNG WATER HE ACTOR PROJECTS DEPAR TMENT eGENERAL ELECTRIC C SAN JCSE. CALIFORNIA 95125 GENER AL h ELECTRIC

NEDO-28136 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for The Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory Commission (USNRC) The for amending TVA's operating license of the Browns Ferry Nuclear Unit 1. information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the tLae this report was prepared. The only undertakings of the General Electric Company respecting information .3',, in this document are contained in the contract between The Tennessee Valley Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Browns Ferry Nuclear Plant Units 1 and 2, dated June 17,1966, and nothing contained in this document shall be construed as changing said centract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any repre-sentation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any respcnsibility for liability or damage of any -kind which may result from such use of such information. .] I

. ~ NEDO-24136 1. PLANT-UNIQUE ITEMS (1.0)' Items different from or not included in Reference 1: Rotated Bundle Analysis Procedure: Appendix A Recirculation Pump Trip Feature: Reference 2 Total Number and Capacity of Safety / Relief Valves: Appendix B Total Number of Safety Valves: Appendix B Total Number of Safety / Relief Valves assumed operable in analysis: Appendix B Fuel Loading Error LHGR: Appendix B 2. RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0) Fuel Type Number Number Drilled Irradiated Initial Core Type 2 260 260 Irradiated Initial Core Typo 3 180 180 Irradiated 8DB274L 144 144 Irradiated 8DB274H 24 24 New 8DRB265L 88 88 New 8DRB265H 68 68 TOTAL 764 764 l 3 REFERENCE CORE LOADINO PATTERN (3 3.1) Nominal previous cycle exposure:

'),592 mwd /t.

Assumed reload cycle exposure: 15,360 mwd /t. Core loading pattern: Figure 1. l I l- '( ) refers to areas of discussion in Reference 1. 1 l

NEDO-24136 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 200C (3.3 2.1, i.e., 3 3.2.1.2) BOC k,77 1.113 Uncontec11ed 0.954 Fully Controlled 0.983-Strongest Control Rod Out 0.000 R. Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, ak STANDBY LIQUID (DNTROL SYSTEM SHUTDOWN CAPABILITY (3 3.2.1.3) 5 Shutdown Margin (Ak) (20CC, Xenon Free) pfm 600 0.038 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3 2.1.5 AND 5.2) EOC3 EOC3-2000 mwd /t Void Coefficient N/A'(t/% Rg) -7.569/-9 462 -8.368/-10.46 Void Fraction (%) 40.18 40.18 Doppler Coefficient N/A (t/%0F) -0.19635/-0.18653 -0.1858/-0.17658 Average Fuel Temperature (OF) 1521 1521 Scram Worth N/A ($) -38.71/-30.97 -36.94/-29.552 l Scram Reactivity versus Time Figure 2a Figure 2b l l l l l 'N !bclear Input Data . A = Used ~1n Transient Analysis 2

NEDO-24136 7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) EOC3 EOC3-2000 mwd /t Exposure 7x7-8x8 8x8R 7x7 8x8 8x8R Peaking factors 1.24 1.22 1.22 1.24 1.22 1.22 (local, radial 1 355 1 383 1.517-1 371 1.442 1 579 and axial) 1.4

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1.4 1.4 1.4 1.4 1 R-Factor 1.08 1.09 8 1.051 1.08 1.09 8 1.051 Bundle Power 5.723 5.837 6.397 5.791 6.086 6.659 (MWt) J Bundle Flow 117.88 108.96 109.40 117.42 107.24 107.77 i (103 lb/hr) Initial MCPR 1.22 1.27 1.28 1.20 1.21 1.22 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2) t Recirculation Pu::p Trip 9 CORE WIDE TRANSIENT ANALYSIS RESULTS (5.2.1) hw+a Core Flow Q/A P3t Fy CPR Plant tearstset Ern sure (1) (1) (1 '(EM) (1 M2?) (pstg) (psis ) ?t7 9x!/8u8R Raspense Lc44 Rejection dit%at Sypass ECC.) 104.5 t00 261.4 106.9 1201 1229 0.15 0,20 rig 2re 3a E003 000CWs/t 104.5 100 203.0 103.3 1110 1215 0.09 0.14 Figure 3b Less of TCCor Fee $ water 104.5 100 113 1 117,7 13g3 ggg g,j3 g,tg - Figure 4 Hest;rg Feesweter Contreiler 20C.3 104.5 tCO 163.1 5C6.9 1154 -tt$5 0.09 0.14 Figare 5a Failurg Ecc3 200CW1/t 104.5 100 137.t 106.7 1152 Ite$ 0.06 3.08 Figure 53 i i 3

NEDO-24136 10. LOCAL ROD WITHDR AWAL ERROR (WITH LIMITING INSTRUMEYr FAILURE). J TRANSIENT

SUMMARY

(5.2.1) i Rod Position Rod Block (Feet ACPR LHGR Limiting l R eading Withdrawn) 7x7 8x8 8x8R 7x7_ 8x8 8x8R-Rod Position 104 4.50 0.08 0.10 0.12 16.0 14.4 13 3 Figure 6 105 5.00 0.09 0.11 0.13 15.9 14.3 13 3 Figure 6 106 5.50 0.10 0.12 0.14 15.8 14.2 13 3 Figure 6 107 6.50 0.11 0.13 0.17 15 3 13 9 13 3 Figure 6 108 10.00 0.16 0.15 0.23 18.2 16.4 15.4 Figure 6 s 109 11.00 0.18 0.19 0.25 18.4 16.2 15.8 Figure W

11. OPER ATING MCPR LIMIT (5.2)

E0c3-2000 mwd /t Boc3 to to EOC3 EOC3-2000 mwd /t i 1.27 (8x8/8x8R fuel) 1.22 (8xB/8x8R fuel) 1.22 (7x7 fuel) 1.20 (7x7 fuel)

12. OVERPRESSURI2ATION ANALYSIS

SUMMARY

(5 3) - i s/ i Power Core Flow P31 P Plant y Transient (%) (%) (psig) (psig)

Response

MSIV Closure 104.5 100 1238 1273 Figure 7 (Flux Scram) 4 i t t , - - ~ - e ,v v ..e

NEDO-24136

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability: Decay Ratio, x2 x, 0.871 / (Natural Circulacion - Rod Block Power) Channel Hydrodynamic Performance Decay Ratio (Natural Circulation - Rod Block Power) 7x7 channel 0.255 8x8 channel 0.415 8x8R channel 0 308

14. LOSS-OF-COOLANT-ACCIDENT RESULTS (5.5.2)

F 8DRB265H 8DRB265L LOCAL LOCAL i EXPOSURE MAPLHOR P.C.T. OXIDATION EXPOSURE PAPLHOR P.C.T. OXIDATION (mwd /t) (kW/ft) (OF) FRACTION (mwd /t) (kW/f t) (CF) FRACTION 200 11.5 1707 0.004 200 11.6 1711 0.004 1000 11.6 1698 0.004 1000 11.6 1700 0.004 5000 11.9 1681 0.003 5000 12.1 1692 0.003 10,000 12.1 1666 0.003 10,000 12.1 1663_ 0.003 15,000 12.1 1688 0.003 15,000 12.1 1683 0.003 20,000 11.9 1687 0.003 20,000 11 9 1683 0.003 j 25,000 11 3 1639 0.003 25,000 11 3 1637 0.003 ) 30,000 10.7 1580 0.002 30,000 10.7 1579 0.002

15. LOADINO ERROR RESULTS' (5.5.4, Appendix B)

Limiting Event: Rotated Bundle 8DRB265H MCPR: 1.07 'Using new Rotated Bundle Analysis Procedures described. n Appendix A.

• • Includes added penalty of 0.02 imposed by NRC.

5

- - ~ NEDO-24136

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Doppler Reactivity Coefficient: Figure 9 Accident Realtivity Shape Functions: Figures 10 and 11 Scram React'.vity Functions: Figures 12 and 13 Plant Specific Analysis Results Parameters not bounded: Accident Reactivity (cold) Scram Reactivity (cold) Resultant peak Enthalpy (cal /gm): 171.6 REFFRENCES T: 1. General Electric Boiling Water Generic Reload Fuel Application, NEDE-24011-r, Revision 3, March 1978. ' 2. General Electric Boiling Water Reactor Increased Relier Valve Simmer Margin Evaluation for Browns Ferry Nuclear Plant Unit 1, Contract 77K64-822849 e ,+# 1 } 6

i NEDO-24136 60 3 313 13 I 3 3 0 3 3 3 313 3 I 3 3 2l4 2, 4 3 4!2 4 2 7 2 2'7 ga 3 3 3 l 2_ 2l3 2 6 6 6 2 7 2 2 2 7 2 7 2g6 56 2 {4 3P 4 3PI4 3P 4 P 4 3P 4 3P 4 2 54 g2 k 2l6 7 2 7 2 7 2 6 6l2 7l 2 7 2 7 2 6l 2 3 52 - 2 Ql 2_ 2 l_3 2_7 l 3P 4 3P 4 P 4 2 2 4 3P l 4 3P 4 7 50 - - 48 3 2l6 3]P 2 6 2 5 51 2 6l2 6 2 7 2 6 3P 6l 2 3 6 2i 7 2g 6 2! 4 3Pl4 3P k 4 3P 4 2 4, 2l 4 28 46 - 12 4l2 4l 2 4 ' 3P 4 3P 4 3P 4 2 2l6 2,4 3l 6 6 35h 6 3P 7l3P 7!3 3 3 44 3 3 7 P 3P 6 .' __'I _ 42 'lk2_ S.' %.'_'_L2_2b'._2 15_ _ bP 6 3P 3P 71 2 3 214 214 2l4 24 23 _3. 4 !2 3Pl6 2l7 28 6 2Pl5 562P 6 2 7 2 6 3 2 7' 3P, 7 3P 6 40 l 2lS l 4] 2 5l2 4 2 412 38 3 4 2 4 I 2 4 2 4 2 2 4 3 4 4 2 2 4 2 4 3P 3P 7 l 3P l2 3 36 3 2l6 3Pl 6 7 2l6 7 2 6l2 2l 2 34 3 7 I 2 4I 2 4 4 3 4 2 4 2 4 4 2 I4 2 4 3 4 2 4 2l4 2 7 3 2 6 $*2 6 2 S 2 6 2 6 2' 3 2l 2 32 2 6 2 l5 216 2 2 6 30 3 2 6 2l 6 2l5 2!6 2 5 2! 2 2 2 6 2 5 5{ 2 6 l 2 6 2; 3 2 6 2 28 3 7l h 4[3 4l2 4 2 4 2 2 4 7 4 3 4 2l 4 2 4 2g 7 3 7! 2 26 3 2 6 3P 7 3P ! 6 3 4 2l6 2 7 2l6 6 2 6 2 4 3P 6 3P 7 3P 6 2 3 7 [4l2 4 2 4 4 2 5 l2 2l 4 2' 5 2 4 2 4 2 4 2'4 3 24 6!3P 7 l 3P 7 2 3 1,,2 17 3P 7,3P{6 6 2 7 2 6 2P 5 6 22 2 2. 6 __[4 2l 4 3 2l4 20 3 442 4 g 2l 4 l 2 4 2 4l 2 5 2 4 2 2l 4 2 4 4 2 5 3! 7 3P 7l3Pl 7 3P. 6 2l 6, 4 2 6 6 3 2 6 2 6 3P 7 3P. 7 l3P 18 3 7 3 3 16 - 2 4l2 4 2 4 3P 4 2P 4 3P l 4 2 2 4 3P 4 3P 4 3P 2 [ 2l 4 21 3 3 _2 6 3Pl t; l 2 7 2 6l2 6 l 2 Q,5 ! 2 _ 7 l 2,_6 _2l3l2 14 6 2 6 2 3P 6 2_ 3_ 82 2l312 7 3P 4 3P 4 l 3P 4 3Pl 4 l2 2 3P 4 3P 4 3P 4 3P 7 2j,7 217 2 212 6 f2 , _ _7[2 !.'. 2_t 642J 7 2 7 2 6 06 3l2l4 3P l 4 P 4 3P 2' 2 4 2Pg 4 3Pl 4 3F 4 2 2[3l] 7l2 7 2 6 6 6 2 6 2 7 {7 31 2 06 2 2g7 2l4 2l 4 h3 04 3l3 4g2 4g2l7 i 02 l3 3l3 3l3l3l3 3l 3

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01 03 05 07 tyJ 11 13 15 17 19 21 23 25 27 23 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 @ LPnM LOCATION (Letter indicates TIP machine) PUEL TYPE g LPfiM LOCATION 4 Common location for als TIP macn.nes) 2 - INITIAL COR E TYPE 2 @ IRM LOCATIOri$ (Letter irwiscates IRM channel) 3 -INITIAL CORE TYPE 3 1 SRM LOCATIONS ILetter indicates $RM channel 4 -80274 L 48 SOURCE LOCATIONS 5 - 80274-H I 6 - 80R B265 L 7 - 8CR D266H l Figure 1. Reference Core Loadirs Pattern I 7

HEDo-24136 d6 100 618 CRO tN PERCENT

====== -== NOMIN AL SCR AM CURVE IN 1-8) --- SCR AM CURVE USED IN AN ALYSl3 / / / [ 70 /6 = /' so l' n a .1 f I C 8 I s /8 N p so l ll a 2 E x f, 4o f 15 / x /'ll J 10 // m / /, // i ,0 .s d _ a4 f f 0 C 0 1 2 3 4 TIME (sed Figure 2a. Scram Reactivity and Control Bod Drive Specificattors - ECC3 8

NEDO-24136 1 100 45 C - 678 CRO IN PERCENT 1 - NOMINAL SCR AM CURVE IN 1-$1 ~ 2 -SCRAM CURVE USED IN ANALYSIS 40 CONTROL ROD DRIVE VERSUS TIME SCRAM REACTIVITY VERSUS TIME 80 35 70 30 2 M = 25 3 2 ,9 m C g G 1 U 2 5 20 g 40 15 30 10 20 ~ 10 l l l 0 1 2 3 4 TIME (sec) Figure 2b. Sc: a:s Reactivity and Control Rod Drite Specificatiens EOC3 - 2000 Edd/t 9

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1. ROO PATTERN IS 1/4 CORE MIRROR SYMMETRIC UPPER LEFT QUADRANT SHOWN ON MAP

2. NO. INDICATE NUV8ER OF NOTCHES WITHORAWN OUT OF 48. BLANK IS A WITHORAWN ROD.

3. ERROR ROO IS (26,39)

Figure 6. Limitirs R'4E Pcd Pattern 15

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NEDO-24136 1.2 ULTIM ATE PERFORM ANCE CRITE RI A 1.0 0.S NATURAL -[ CfRCULATION OC 0.6 (c ><v w0 105% RCD LINE 0.4 ~ 0.2 ~ 0 o 20 40 so ao too 120 POWER (%) I l Figure 8. Decay Ratio 17

NEDO-24136 s 0 s, -10 w f /

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9 / / 2% 5~ / yv.,. / l f# "A / / / -20 p / s 25 / BOUNDING VALUE FOR 280 C4UC COLD ~ ~ -= 8OUNDING VALUE FOR 290 CAUG HS8 CALCULATED V ALUE - COLD CALCULATED V ALUE - MS8

== I -30 s I I I I I I _3, Q 400 800 1200 1600 2000 2400 FUEL TEMPER ATURE ('Cl l Figure 9 Doppler Reactivity Coefficient Coc:parison for RDA 18

NEDO-24136 24 20 l 3 3 16 9 4 3 9 w 4 I 12 2 U< 5 f% __m"**** / // a / 80VNOtNG VALUE FOR 280 CAUG CALCULATED VALUE 4 a b // 0 0 4 8 12 16 20 ROO PCSITION tft OUT) Figure 10. RDA Reactivity Shape Function at 200C 19

NEDO-24136 i i I 24 2: SOUNOING VALUE FOR 280 CAUG - - CALCULATED VALUE 3 16 e 4 i C w

• 1 12

~ N 2 I 8 ) ~ / [ 4 / / / I I l I l o 0 4 8 12 16 20 ROO POSITION (f t CUT) Figure 11. RDA Reactivity Shape Function at 2860C 20

u_ NSDO-24136 80 70 BOUNDING V ALUE FOR 200 CAUG - - CALCULATED VALUE G f 50 9 5 8 5) d. 40 D 5 C N [ r 30 l / I 10 / I I 0 O 2 4 6 8 to ELAPSED TIME (soci Figure 12. Scram Reactivity Function at 20 C 0 21

NEDo-24136 32o too ~ 'N, g I so S 8 a 1 3 l / 4 c,o I E I Oe I 1 ao / f souNoiNo vAtut son zoo cAUG - - cALcutATEo vAtut 2o / / / - __ W l l o_ o 4 6 a to ELAP$to TIME tsec) Figure 13 RDA Scra:: Reactivity Function at 2860C 22

t NEDO-24936 (1) The channel fastener assemblies, includir4 the spring and guard used to maintain clearances between channels, are located at one corner of each fuel assembly adjacent to the center of the control rod. (2) The identification boss en the fuel assembly handle points toward the adjacent control rod. (3) The channel spacing buttons are adjacent to the control. rod passage area. (4) The assenbly identification numbers which are located on the fuel assembly handles are all readable from the direction of the center of the cell. (5) There is cell-to-cell replication. Experience has demonstrated that these design features are clearly visible so that any misloaded bundle would be readily identifiable during core loading verification. Figures A-1, A-2 and A-3 denote a normally loaded bundle, a 1800 rotated bundle, and a 900 rotated bundle, respectively. Actual experienet-," References A-1 and A-2)- has demonstrated that the probability of a rotated bundle is low. The new analyses procedure results show that the minimum CPR for the most limiting-rotated bundle in the core is greater than the safety limit. REFERENCES A-1 Letter, R.E. Engel (GE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," MFN-219-77, June 1, 1977. A-2 Letter, R.E. Engel (GE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," MFN-457-77, Nove=ber 30, 1977. A-2

NEDO-24136 APPENDIX A NEW BUNDLE LOADING ERROR EVE!TT ANALYSES PROCEDURES The bundle loadire error analyses results presented in Section 15 in this supplement are based on a new analysis procedure for the rotated bundle loading error event. The use of this new analysis procedure is discussed below. 1 A.1 NEW ANALYSES PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error event analyses results presented in this supplement are based on the new analyses procedure described in References A-1 and A-2. Bis new method of performing the analyses is based en a more detailed analysis model, N which reflects more accurate analyses than that used in previous analyses of.this i event. The principle difference between the previous analyses procedure and the new analyses procedure is the modeling of the water gap alx.g the axial length of the bundle. The previous analyses used a uniform water gap, whereas the new analyses utilize a variable water gap which is representative of the actual condition. The effect of the variable water pp is to reduce the power peaking and the R-factor in the upper regions of the limiting fuel rod. This results in the calculation of a reduced ACPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represerts the actual geometry. In the new analyses, the axial alignment of a 1800 rotated bundle conservatively ignores the presence of the channel fastener. We more limiting condition of assumig that the spacer buttons are in contact with the top guide is assumed. There is no known loading that cculd bend or break the channel spacer button during the insertion of a 1800 rotated bundle, since both the top guide and spacer button are chamfered to provide lead-in. For a pmperly assembled bundle, na mechanism exists which could invalidate the assumption that a 1800 rotated bundle leans to one side. It should be noted that proper orientation of bundles in the reactor core is readily ver* fied by visual observation and assured by verificaticn procedures during core lead. T. Five separate visual irdications of proper bundle orientation exist: A-1

l NEO-24136 Jl Oy, O /s

=A

'O x .D 6 $e v i k_O 'O O O N \\ / '\\ / N h( N O o o, o A / 1 ,A P, %l f y 0 'hO <l NOTE: Bundle numbers are for illustrative purposes only. Figure A-1. Normal Leading A-3

1 NED0-24136 I C ISO Rotnion - Bundle U 23:3 Il / _ _N \\ EO l o']. g / 4 'C e x 9 P, w h> \\ v l O 'O / i \\ / I / \\ l ( / \\_ _l 0 '0 O-O x s .o s' f 1 '+ f) \\/ O /\\'O L l ll l i NOTE: Bundle numbers are for illustrative purposes only. Figure A-2. Rotated Bundle, 180 Degree Rotation l A-4

NEDO-215136 i i 90 Rocacion - Sundle LJ 2343 ,/ n O( O( l / 4 >9 n) n 'a# f' v V O 'O O 'O M i \\ / \\ / I N / N I y / N .O (o O, O A D 4 3 4 0, A e ~? ) \\[ OOO h NOTE: Bundle nu=bers are for illustrative purposes only. I Figure A-3 Rotated Bundle, 900 Rotation A-5/A-6

I NEDO-24136 + t APPENDIX B Safety / Relief Valves j Number - Setpoint Installed / Analyzed (psig) j 4/3 1105 + 15 4/4 1115 + 1% i 5/5 1125 + 1% Total Capacity of Safety / Relief Valves 2 Installed: 82.6% Analyzed: 76.246% + h Total Number of Safety Valves: 0 5 . Fuel Loading Error LHGR: _16.2 kW/ft i -1 1 4 l i .i t 't 1 1 i i-B-1/B-2 4

'NEDO-24136 a (1) The. channel fastener assemblies, including the spring and guard used to maintain clearances between channels, are located at one corner of each fuel assembly adjacent to the center of the ' control rod. (2) The identification boss on the fuel assembly handle points toward the adjacent control rod. (3) The channel spacing buttons are adjacent to the control rod passase area. (4) The assembly identification numbers which are located on the fuel assembly handles are all readable from the direction of the center of the cell. (5) There is cell-to-cell replication. Experience has demonstrated that these design features are clearly visible so that any misloaded bundle would be readily identifiable during core loading verification. Figures A-1, A-2 and A-3 denote a normally loaded bundle, a 1800 rotated bundle, and a 900 rotated bundle, respectively. Actual experience (References A-1 and A-2) has demonstrated that the probability of a rotated bundle is low. The new analyses procedure results show that the minimum CPR for the most limiting rotated bundle in the core is greater than the safety limit. REFERENCES A-1 Letter, R.E. ?nsel (GE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," MFN-219-77, June 1, 1977. A-2 Letter, R.E. Engel (CE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," MFN-457-77, November 30, 1977. A-2}}