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Category:REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)
MONTHYEARML20113H4961992-04-30030 April 1992 Rev 0 to Licensing Topical Rept Nutech Feedwater Nozzle Bypass Leakage Monitoring Sys,App:Results for Browns Ferry Unit 2,June-Dec 1991 ML20059L4961990-08-31031 August 1990 BWR Owners Group NUREG-0578 Implementation: Analyses & Positions for Plant-Unique Submittals ML20235Z0491987-03-17017 March 1987 Nonproprietary Topical Rept, RDS-1000 Radwaste Dewatering Sys ML20113H5051986-09-0505 September 1986 Rev 0 to Licensing Topical Rept Nutech Feedwater Nozzle Bypass Leakage Monitoring Sys,App:Results for Browns Ferry Unit 1 ML20113H4991986-09-0505 September 1986 Rev 0 to Licensing Topical Rept Nutech Feedwater Nozzle Bypass Leakage Monitoring Sys App:Results for Browns Ferry Unit 3 ML18030B0411985-12-31031 December 1985 Radiological Impact of Ventilation Damper Closing Time During Design Basis Fuel Handling Accident. ML20128A0421985-05-31031 May 1985 Rev 2 to LOCA Accident Analysis for Browns Ferry Nuclear Plant Unit 2 ML20112F6931984-08-31031 August 1984 Core Region Matls Info (Units 1,2 & 3) ML20077C8841983-05-31031 May 1983 Rev 1 to LOCA Analysis for Browns Ferry,Unit 1 ML20069J1721982-08-31031 August 1982 Errata & Addenda Sheet 5 to NEDO-24088-1, LOCA Analysis for Browns Ferry Nuclear Plant,Unit 2 ML20069J1691982-07-31031 July 1982 Errata & Addenda Sheet 4 to NEDO-24088-1, LOCA Analysis for Browns Ferry Nuclear Plant,Unit 2 ML18025B8221982-07-23023 July 1982 BWR Transient Analysis Model Utilizing Retran Program, Second Part ML20028G3681982-05-31031 May 1982 Safety Review of Browns Ferry Nuclear Plant Unit 1 at Core Flow Conditions Above Rated Flow During Cycle 5. ML17276B8021981-12-31031 December 1981 BWR Transient Analysis Model Utilizing Retran Program. ML20039B7011981-12-0909 December 1981 Errata to LOCA Analysis, Sheet 1 ML18025B7101981-09-21021 September 1981 Method for Development of One-Dimensional Core Kinetics Data for RETRAN-02. ML20041E4721981-05-31031 May 1981 Plants,Units 1,2 & 3 Single Loop Operation. ML20126J6151981-04-30030 April 1981 Errata 2 to LOCA Analysis for Browns Ferry Unit 1 ML18030B2231980-09-30030 September 1980 DC Power Source Failure for Browns Ferry Nuclear Power Station,Units 1,2 & 3. ML20147D6701980-09-0808 September 1980 In-Vessel Neutron Spectral Analysis ML19270F6121979-01-31031 January 1979 Verification of TVA Steady-State BWR Physics Methods. ML19270F6141978-06-0101 June 1978 Three-Dimensional LWR Core Simulation Methods. ML19270F6131978-04-15015 April 1978 Methods for Lattice Physics Analysis of Lwr'S. 1992-04-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML18039A9041999-10-15015 October 1999 LER 99-010-00:on 990917,automatic Reactor Scram on Turbine Stop Valve Closure Occurred.Caused by High Water Level in Main Steam Moisture Separator 2C2.Unit 2C2 Reservoir Level Transmitter & Relays Were Replaced & Tested Satisfactorily ML18039A8981999-10-14014 October 1999 LER 99-009-00:on 990915,manual Reactor Scram Was Noted Due to EHC Leak.Caused by Failure of Stainles Steel Tubing Connection.Removed Damaged Tubing & Connection Plug ML18039A8951999-10-0808 October 1999 LER 99-008-00:on 990905,HPCI Was Inoperable Due to Failed Flow Controller.Caused by Premature Failure of Capacitor 2C3.Replaced Controller & HPCI Sys Was Run IAW Sys Operating Instructions ML18039A8751999-09-30030 September 1999 LER 99-005-00:on 990901,SR for Standby Liquid Control Sampling Was Not Met.Caused by Deficient Procedure for Chemical Addition to Standby Liquid Control.Revised Procedure.With 990930 Ltr ML20217F9671999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML20212E6341999-09-23023 September 1999 Suppl to SE Resolving Error in Original 990802 Se,Clarifying Fact That Licensee Has Not Committed to Retain Those Specific Compensatory Measures That Were Applied to one-time Extension ML20212D3831999-09-20020 September 1999 Safety Evaluation Supporting Proposed Rev to Withdrawal Schedule for First & Third Surveillance Capsules for BFN-3 RPV ML20212B8561999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Browns Ferry Nuclear Plant.With ML18039A8821999-08-31031 August 1999 Increased MSIV Leakage Tech Spec Change Submittal - Seismic Evaluation Rept. ML18039A8391999-08-0606 August 1999 BFN Unit 2 Cycle 10 ASME Section XI NIS-1 & NIS-2 Data Repts. ML20210N1221999-08-0202 August 1999 Safety Evaluation Accepting Licensee Request for Relief from ASME B&PV Code,Section XI Requirements.Request 3-ISI-7, Pertains to Second 10-year Interval ISI for Plant,Unit 3 ML20210R0931999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A8201999-07-26026 July 1999 LER 99-004-00:on 990625,facility Core Spray Divisions I & II Inoperable at Same Time Due to Personnel Error.Electrical Supply Breaker to Core Spray Division II Pump 3B Returned to Normal Racked in Position ML18039A8171999-07-20020 July 1999 LER 99-007-00:on 990623,discovered That SR for Monitoring of Primary Containment Oxygen Concentration Had Not Been Met. Caused by Failure of Operators to Adequately Communicate. Required Surveillances Were Performed.With 990720 Ltr ML18039A8161999-07-19019 July 1999 LER 99-006-00:on 990618,noted That Main Steam SRV Exceeded TS Setpoint Tolerance.Caused by Pilot Vlve disc-seat Bonding.Util Replaced All 13 SRV Pilot Cartridges with Cartridges Certified to Be Witin +/-1%.With 990719 Ltr ML20209J0771999-07-16016 July 1999 Safety Evaluation Concluding That Licensee Provided Adequate Information to Resolve ampacity-related Points of Concern Raised in GL 92-08 for BFN & That No Outstanding Issues Re GL 92-08 Ampacity Issues for Browns Ferry NPP Exist ML18039A8121999-07-12012 July 1999 LER 99-005-00:on 990617,ESF Actuation & HPCI Declared Inoperable.Caused by Personnel Error.Reset HPCI & Returned Sys to Operable Status with 25 Minutes.With 990712 Ltr ML20209H4381999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A8101999-06-28028 June 1999 LER 99-004-00:on 990530,safety Features Sys Actuations Occurred Due to RPS Trip.Caused by Failure of MG Set AC Drive Motor Starter Contractor Coil.Licensee Placed 2B RPS Bus on Alternate Feed & Half Scram Was Reset ML20196F8811999-06-23023 June 1999 Safety Evaluation Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves ML18039A8071999-06-14014 June 1999 LER 99-003-00:on 990515,automatic Reactor Scram Due to Turbine Trip Was Noted.Caused by Failure of Mechanical Trip Cylinder to Latch When Hydraulically Reset.Operations Crew Stabilized Reactor Following Scram ML18039A8021999-06-14014 June 1999 LER 99-002-00:on 990501,SRs for Single CR Withdrawal During Cold SD Were Not Adequately Implemented.Caused by Procedural Inadequacy.Revised Applicable Plant Surveillances.With 990614 Ltr ML18039A8011999-06-14014 June 1999 LER 99-001-00:on 990515,automatic Reactor Scram Occurred Due to Tt.Caused by Failure of Mechanical Trip Cylinder to Latch When Hydraulically Reset.Operations Crew Stabilized Reactor Following Scram.With 990614 Ltr ML20196B8051999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A7791999-05-0606 May 1999 LER 99-003-00:on 990408,declared Plant HPCI Sys Inoperable Due to Loose Wire.Caused by Failure to Properly Tighten Screw at Some Time in Past.Loose Wire Was Tightened ML18039A7761999-04-30030 April 1999 Revised Surveillance Specimen Program Evaluation for TVA Browns Ferry Unit 3. ML20206R0731999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Bfnp.With ML18039A7561999-04-23023 April 1999 Bfnp Risk-Informed Inservice Insp (RI-ISI) Program Submittal. ML18039A7671999-04-0808 April 1999 Rev 0 to TVA-COLR-BF2C11, Browns Ferry Nuclear Plant Unit 2 Cycle 11 Colr. ML18039A7461999-04-0707 April 1999 LER 99-001-00:on 990308,determined That Two Trains of Standby Gas Treatment (SGT) Were Inoperable.Caused by Trip C SGT Blower Motor Breaker.Initiated Shutdown of Plant,Reset C SGT Blower Motor Breaker & Declared Train Operable ML20205N8341999-04-0101 April 1999 Part 21 Rept Re Automatic Switch Co Nuclear Grade Series X206380 & X206832 Solenoid Valves Ordered Without Lubricants That Were Shipped with Std Lubrication to PECO & Tva.Affected Plants Were Notified ML20205F9341999-04-0101 April 1999 Safety Evaluation Authorizing Licensee 990108 Relief Request PV-38,from Requirements of ASME BPV Code Section XI IST Testing,Valve Program for Plant,Units 1,2 & 3 ML20205T5441999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Bfnp.With ML20205S0601999-03-31031 March 1999 Rept on Status of Public Petitions Under 10CFR2.206 with Status Change from Previous Update,990331 ML20205S0661999-03-31031 March 1999 Rept on Status of Public Petitions Under 10CFR2.206 with No Status Change from Previous Update,990331, Atlas Corp ML18039A7361999-03-11011 March 1999 Rev 4 to TVA-COLR-BF2C10, Bfnp,Unit 2,Cycle 10 Colr. ML20204C7891999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A6951999-02-19019 February 1999 LER 99-002-00:on 990122,LCO Was Not Entered During Calibration Testing of 3D 480 Volt Rmov Board.Caused by Personnel Error.Tva Has Briefed Operations Personnel to Preclude Recurrence of Event.With 990219 Ltr ML18039A6871999-02-12012 February 1999 LER 99-001-00:on 990114,Unit 3 HPCI Was Noted Inoperable. Caused by Oil Leak on Stop Valve.Corrective Maint Was Performed to Repair Oil Leak.With 990212 Ltr ML18039A6931999-02-0303 February 1999 Rev 3 to TVA-COLR-BF2C10, Bfnp Unit 2 Cycle 10 Colr. ML18039A6941999-02-0303 February 1999 Rev 1 to TVA-COLR-BF3C9, Bfnp Unit 3 Cycle 9 Colr. ML18039A6671998-12-31031 December 1998 LER 98-004-00:on 981202,SR Intent Was Not Adequately Implemented.Caused by Procedural Inadequacy.Revised Procedures to Provide Proper SR Implementation.With 981231 Ltr ML18039A6661998-12-31031 December 1998 Ro:On 981215,HRPCRM 2-RM-90-273C Was Declared Inoperable. Caused by Downscale Indication.Containment RM Will Be Utilized as Planned Alternate Method of Monitoring Until Hrpcrm 2-RM-90-273C Can Be Returned to Operable Status ML20199K8951998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Browns Ferry Nuclear Plant.With ML20199F2721998-12-31031 December 1998 ISI Summary Rept (NIS-1), for BFN Unit 3,Cycle 8 Operation ML18039A6471998-12-15015 December 1998 LER 98-007-00:on 981116,unplanned ESF Following Loss of 4kV Unit Board 3B Occurred.Caused by Temporary Energization of Lockout Relay on 4kV Unit Board 3B When Resistor on Relay Monitoring Lamp Circuit Shorted.Replaced Resistor ML18039A6371998-12-0707 December 1998 LER 98-006-00:on 981116,MSSR Valves Exceeded TS Setpoint Tolerance.Caused by Pilot Valve Disc/Seat Bonding. Installed SRV Pressure Switches During Unit 3,cycle 8 Outage.With 981207 Ltr ML20199F2791998-12-0303 December 1998 Bfnp Unit 3 Cycle 8 ASME Section XI NIS-2 Data Rept ML20198D9621998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Bfn,Units 1,2 & 3. with ML18039A6071998-11-12012 November 1998 LER 98-005-00:on 981014,mode Changes Not Allowed by TS 3.0.4 Were Made During Reactor Startup.Caused by TS LCO 3.0.4 Not Being Properly Applied.Training Info Memo Re Proper Application for TS LCO 3.0.4 Was Prepared.With 981112 Ltr 1999-09-30
[Table view] |
Text
/'EDO-24266
'k pl d
~ l ~ ENCLOSURE 80NED272 Class I September 1980 V ~
,yi DC POWER SOURCE FAILURE FOR I IL BROWNS FERRY NUCLEAR POWER STATION UNITS 1, 2 AND 3 8b04080233 8b032b PDR ADDCK 05000259 P PDR NUCLEAR POWER SYSTEMS DIVISION ~ GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA95125 GENERAL 'LECTRlC 4
DISCLAIMER OF RESPONSIBILITY This document was prepared by or for the General Electric Company. Neither the General Electric Company nor any of the contnbutors to this document:
'I A. Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this docu-ment. or that the use of any information disclosed in this document may not infringe privately owned rights; or B. Assumes any responsibility for liability or damage of any kind which may result from the use of any information disclosed in this document.
NEDO-24266 CONTENTS
~Pa e
- 1. INTRODUCTION l-l
- 2. CONCLUSIONS 2-1
- 3. BREAK ANALYSIS 3-1 3.1 Discharge Breaks 3-1 3.2 Feedwater Line Breaks 3-2
- 4. REFERENCES 4-1
NEDO-24266 TABLES Tab le Title ~Pa e 3-1 Calculated PCT Results for a Recirculation Discharge Line Break With a DC Power Source Failure 3-3 3-2 Calculated PCT Results for a Feedwater Line Break With a DC Power Source Failure 3-3 ILLUSTRATIONS r
~PX ure Title g )
~Pa e 3-1 Browns Ferry 1, 2 and 3 0.3 ft Recirculation Discharge Break DC Power Source Failure (No ADS) 3-4 3-2a Browns Ferry 1, 2 and 3 0.3 ft2 Recirculation Discharge Break DC Power Source Failure (No ADS) (8x8 Fuel Type) 3-5 3-2b Browns Ferry 1, 2 and 3 0.3 ft2 Recirculation Discharge Break DC Power Source Failure (No ADS) (7x7 Fuel Type) 3-6 3-3 Browns Ferry 1, 2 and 3 0:25 ft2 Feedwater Line Break DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes 3-7 3-4 Browns Ferry 1, 2 and 3 0.25 ft2 Feedwater Line Break DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes 3-8 v/vi
1 4 V r
j h
NEDO-24266
- 1. INTRODUCTION The purpose of this study is to investigate, in a generic manner, the implications of a direct current (DC) power source failure for Browns Ferry Nuclear Plant Units 1, 2, and 3. In addition, this study provides bounding peak cladding temperatures (PCT) as a function of break area for small breaks.
ECCS analyses were performed for the most severe break locations (the recirculation discharge line where LPCI injects and the feedwater line where HPCI injects), assuming that the DC power failure also disables the automatic depressurization system (ADS).
The analyses were performed with the 1977 approved GE LOCA analysis models SAFE03, REFLOOD05 and CHASTE05, and considering the LPCI modification as described in Reference l.
This study is based on system availability following a DC power source failure defined by TVA for the three units.
NEDO-24266
- 2. CONCLUSIONS The general conclusions reached from this study are:
(1) The conservative bounding PCT calculated for small break LOCAs with the DC power source failure that also disables the Automatic Depressurization System for Browns Ferry Units 1, 2, and 3 is less than 1900'F.
(2) For large break LOCAs, the maximum PCT for any unit is not affected by a DC power source failure (Reference 2).
(3) The MAPLHGR
~l for any unit is not affected by a DC power source failure (Reference 2).
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v' NEDO-24266
- 3. BREAK ANALYSIS For the Browns Ferry Units 1, 2 and 3 plants with the LPCI modification, the most severe DC power source failure that will also disable the ADS will result in one Core Spray system, three LPCI pumps injecting in both recirculation loops, and the HPCI (1CS + 3LPI2 + HPCI) all remaining operable (3LPI2 = 3 LPCI pumps in two recirculation loops).
.The worst postulated line break locations are the recirculation discharge line where two of the three LPCI pumps infect coolant to the reactor (1CS + 1LPCI + HPCI) and the feedwater line where the HPCI injects coolant (1CS + 3LPI2), because these break locations will result in the maximum degradation of the remaining emergency cooling capability.
'I Analysis was performed conservatively assuming total loss of the ECCS flow in the broken lines regardless of break size.
3.1 RECIRCULATION LINE DISCHARGE BREAKS jg Ten break 4
sizes were analyzed between 0.05 ft 2 and 0.50 ft with one gqre Spray system, one LPCI pump (in)ecting into the unbroken recirculation loop)', and the HPCI assumed operating.
u I
As seen from the results in Table 3-1, no core uncovery is predicted for break sizes <0.15 ft . The CHASTE05 code has been used only when the ultraconserv-ative REFLOOD results exceeded the PCT limits, since the CHASTE05 code more accurately predicts PCT.
The highest PCT calculated with the CHASTE05 code was 1864'F for 7x7 fuel at break size of 0.3 ft2 . The PCT for 8x8 fuel at the 0.3 ft break size was calculated to be 1693'F. Water level, pr'essure and PCT plots for this limiting break size are shown in Figures 3-1 and 3-2.
NEDO-24266'.2 FEEDWATER LINE BREAKS Small feedwater line breaks with the total loss of both HPCI and ADS are similar in behavior to the steamline break outside of the containment case discussed in the Final Safety Analysis Report (FSAR), in that the reactor vessel will not depressurize to the operating point of the low pressure ECC systems in the short-term without manual action to depressurize the reactor.
\
lt Manual actuation of four-out-of-six total ADS valves was assumed at 10 minutes after the instant of the break initiation, which is consistent with the above steamline break case assumption of 10-minute operator action time. The assump-tion of four-out-of-six valves is conservative because recent guidelines (THI)
P instruct the operator to open as many valves as there are in the ADS, which, in this case, is six valves. The assumption of four valves resulted from the fact that, after the assumed subject DC power source failure, only four of the '~
ADS valves can be operated manually; but there are many additional SRVs that could be'manually actuated by the operator.
Thirteen break sizes, ranging from 0.01 ft2 to 0.50 ft2 , were analyzed with the SAFE03 and REFLOOD05 computer programs, and PCTs were conservatively calculated.
with the REFLOOD05 small break model. As seen from the results in Table 3-2, the maximum calculated PCT with REFLOOD05 was 1880'F 0 at a break size of 0.25 ft 2 The water level, pressure and PCT plots for this 0.25 ft 2 break size are shown in Figures 3-3 and 3-4.
3-2
NEDO-24266 Table 3-1 I CALCULATED PCT RESULTS FOR A RECIRCULATION DISCHARGE LINE BREAK, WITH A DC POWER SOURCE FAILURE (BROWNS*FERRY UNITS 1, 2 AND 3)
PCT PCT Break Area (REFLOOD) (CHASTE)
(ft2) ('F) ( F) 0.05 core does not uncover 0.07, core does not uncover 0.09 core does not uncover 0.11 core does not uncover 0.13 core does not uncover 0.15 core does not uncover 0.20 1840
- 0. 30 2307 1864
- 0. 40, 1753 0.50 1686 Table 3-2 CALCULATED PCT RESULTS FOR A FEEDWATER LINE BREAK WITH A DC POWER -SOURCE FAILURE
'l 4 f lg (BROWNS FERRY UNITS 1, 2 AND 3)
PCT
= Break Area (REFLOOD)
(ft2) ('F) '7 5 M 0.01 1221 0.03 1233 0.05 1194 0.07 1151 0.10 1126 0.15 1269 0.20 1413 0.25 1880 0.30 1813 0.35 1586 0.40 1414 0.45 1311 0.50 1211 3-3
NEOO-24266 1.0 1 VESSEL PRESSURE IIysie) 2 8AF 18.03 ft 48 3 TAF 30.03 ft 0.9 45 4 WATER LEVEL(R) 0$ 40 0.7 35 o
x 3 06 30.03 3
llf 0.5 25 tt:
I g 0.4 1ILoa 0.3 12 0.2 0.1 0
100 200 300 400 500 600 700 800 900 1000 TIME {sec)
Figure 3-1. Browne Ferry 1, 2 and 3 0.3 ftl Recirculation Diecharge Break DC Power Source Failure (No ADS)
+))
3-4
I,
~
NEDO-24266 20 19 16 17 16 15 14 O
x 13 0
12 CC PEAK CLAOOING P 11 TEMPERATURE REFLOOO K
w 10 HYC St Ih .fi> F X
w 8
Q O
O Y 6 6
h ~ 10,000 h ~25 h ~ 0.0 0
0 100 200 ,
300 400 TIME Isec)
Figure 3-2a. Browns Ferry l, 2 and 3 0.3 ft Recirculation Discharge Break DC Power Source Failure (No ADS) (8x8 Fuel Type) 3-5
NEDO-24266
$ %l S 20 19 18 16 o 14 13
/
7L.
PEAK CLADDING TEMPERATURE cc I
~+
12 11
/ CHASTE OS HTC Bcu/hr ft> F g 10 I-Q 9 I
z a 8 r
O Y
t 6 f h ~ 10,000 h ~25 h~O 0
0 300 400 TIME bee)
Figure 3-2b. Browns Ferry 1, 2 and 3 0.3 ft2 Recirculation Discharge Break DC Power Source Failure (No ADS) (7z7 Fue] Type) 3-6
NEDO-24266 1.0 1 VESSEL PRESSURE (psis) 2 8AF 18.03 ft 48 4
0.9 3 TAG 30.03 ft 45 4 WATER LEVEL tff) 08 40 07 35 X
W 0.6 30.03 CC D
g 0.5 1 4 25 0.4 18.03 0.3 0.2 0.1 0
0 100 200 300 400 500 600 700 800 900 1000 TIME bee)
Figure 3-3. Browns Ferry 1, 2 and' 0.25 ft Feedwater Line Break DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes 3-7
NEDO-24266 ~ ~ ~
k 19
- C 18 17 PEAK CLADDING 16 TEMPERATURE REFLODD 05 15 HTC Biu/hr.ft> F o 'l4 X
13 us 12 K
D 11 K
10 UJ I 9 Q
8 Il O
7 Y 6 EL
~ ~ 3 h ~ 10,000 h 0 25 h & 0.0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 TIME Isscl Figure 3-4. Browns Ferry 1, 2 and 3 0.25 ft Feedwater Line Break DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes
))
3-8
NEDO-24266
- 4. REFERENCES
- 1. ,
"Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 51 to Facility Operating License No. DPR-33, Amendment No. 45 to Facility Operating License No. DPR-52, Amendment No. 23 to Facility Operating License No. DPR-68, Tennessee Valley Authority, Browns Ferry Nuclear Plant, Units Nos. 1, 2, and 3".
~
- 2. Letter, R. E. Engel to P. S. Check, "D.C. Power Source Failure for BWR/3 and 4", dated November 1, 1978.
I \
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