ML18283A521

Affidavit of Jack R. Calhoun

ML18283A521
Person / Time
Site:	Browns Ferry
Issue date:	07/21/1976
From:	Calhoun J Tennessee Valley Authority
To:	Atomic Safety and Licensing Board Panel
References
Download: ML18283A521 (39)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

TENNESSEE VALLEY AUTHORITY, ) Docket Nos. 50-2

) O- 6O (Browns Ferry Nuclear Plant )

units 1 and 2) )

AFFIDAVIT OF JACK R. CALHOUN Jack R. Calhoun, being duly sworn, deposes and says:

business address is Tennessee Valley Authority, 702 Edney Building, Chattanooga, Tennessee; I 'am employed by the Tennessee Valley Authority as the Chief of the Nuclear Generation Branch, Division of Power Production. I am familiar with these proceedings and have personal knowledge of the matters contained herein.

alifications I have been continuously employed by the Tennessee Valley Authority since 1949. Prior to that time I served for eight years in the United States Navy. Part of this time I was an Electrical Officer on the light cruiser VSS Oklahoma City and the aircraft carrier USS

'I Saratoga and was qualified as'Engineering Officer-of-the Watch at sea on both ships.

I have I received the Bachelor of Science degree in electrical engineering from Tennessee Technological University in 1949. During this period I was the Executive Officer and ELectronics Officer of the U.S. Naval Reserve Electronics Warfare Company located at Cookeville, Tennessee.

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I began my employment with TVA in 1949 as a student in the steam generating plant operator training program at the Watts Bar Steam Plant and later became an instructor in that program. I was transferred to the Johnsonville Steam Plant in 1952 as a unit operator and later assumed the position of an electrical engineer. In 1954 I was placed in charge of all electrical maintenance at the Johnsonville plant.

In 1958 I became assistant plant superintendent at the 1,500-MW Shawnee Steam Plant at Paducah, Kentucky.

In 1960 I became superintendent of the Experimental Gas-Cooled Reactor (EGCR) at Oak Ridge, Tennessee. During this period. I attended the Oak Ridge School of Reactor Technology., In 1961 I spent five months at the Berkeley Nuclear Power Station in Bristol, England, assisting in the startup of that reactor. While at Berkeley I completed. the reactor operator training course on a nuclear plant simulator used to train all reactor operators for the Central Electricity Generating Board..

In 1963 I was appointed assistant ProJect Manager of the Experimental Gas-Cooled Reactor and was responsible for assisting the proJect manager in all phases of technical and operational work.

From 1963 to 1966 I was a member (for reactor operation) of a panel created by an agreement between the United Kingdom Atomic Energy Authority and the United States Atomic Energy Commission to exchange information on gas-cooled reactors. As a member of this panel, I twice traveled to England to investigate and. to observe the operation of the British Advanced Gas-Cooled Reactor in preparation for the startup of EGCR.

From February 1966 to, February 1968, I held the position of Assistant to the Chief, Power Plant Maintenance Branch, Division of Power Production in TVA. I assisted in the engineering'and coordina-tion of the electrical and mechanical maintenance of all TVA steam and hydro plants. I was also responsible for the op'eration and maintenance planning relating to future TVA nuclear power plants.

From February 1968 to July 1971, I held the position of Plant Superintendent of the Browns Ferry Nuclear Plant in Athens, Alabama.

4 From July 1971 to April 1974, I was nuclear operations coordinator; and,in April 1974 my title was changed. to Chief, Nuclear Generation Branch. In this position I am responsible for and in charge of staffing, startup testing, and operations of all TVA nuclear power plants, including the Browns Ferry Nuclear Plant, units 1 and. 2. I

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am also responsible for the coordination of the restoration and modifications activities, including fire protection improvements, of the Browns Ferry Nuclear Plant, units 1 and. 2, following the March 22, 1975, fire.

I am presently a member of the Advisory Council at Pennsylvania State University'advisor.'to the Nuclear 'Engineering Department ) and serve as Vice Chairman, Reactor Operations Division, Ameri'can Nuclear Society.

I am familiar with %his proceeding and have personal knowledge of the matters stated. herein.

Iii Statement The modification and restoration of. units 1 and, 2 in accordance with TVA's "Plan for Evaluation, Repair and Return to Service of Browns Ferry, Units 1 and, 2, (March 22, 1975, Fire)" have been substantially completed. Permission- has been granted to lodd fuel in units 1 and 2.

All control rods have been fully inserted and electrically disarmed throughout fuel loading which is now complete for'nit 2. Unit 1 currently has 659 fuel assemblies loaded. After refueling is complete on each unit, TVA proposes to conduct the following subcritical testing, l

which is a part of the,startup retest program: Co'ntrol Rod.=Drive Syst'e m tests (Startup Test No.' ) scheduled at zero reactor pressure. The control rod drive tests proposed are position indication, insert/withdraw time, coupling, fricti'on, and scram testing at zero reactor pressure. The pro-posed testing is a portion of the startup test program previously approved by the Nuclear Regulatory Commission, and will be conducted. as described in the Browns Ferry Final Safety Analysis Report, Section 13.5 (pages 13.5-18 and. 13.5-19) and Table 13.5-5 (Attachment 1), except for those changes discussed in Part XI, Section D, of TVA's "Plan for, Evaluation, Repair and Return to Service of Browns Ferry Units 1 and. 2, (March 22'975'ire)"

(Attachment 2 ).

The purpose of conducting the Control Rod Drive System tests will be to determine initial operating characteristics of the Control Rod Drive System and, to ensure that no control rod interference exists in the fully loaded core. On successful completion of these tests, TVA will install the reactor vessel head which will reduce the startup retesting period by approximately ten days when permission is granted to operate units 1 and 2.

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This testing will require the operability and: use of each control rod., one at a time. At all times during'this testing, the, remaining rods not in use will be fully inserted, valved out, and electri-cally disarmed; the RHR system will be aligned to cool the core in the reactor vessel; and all valves in lines which could.,drain the reactor vessel and the RHR system in this mode will be disabled in the position that will not dr'ain the reactor. These conditions are in accordance with the conditions stipulated when permission was granted to load fuel in units 1 and 2.

Justification is presented below for TVA's position that this testing can be performed with reasonable assurance that the health and safety of the public will not be endangered., with no reliance placed on fire damaged equipment to maintain the reactor subcritical or to mitigate the consequences of an accident.

Prior to commencing the proposed testing, a verification that the fuel in each core is loaded correctly in the position it occupied prior to the fire is made by'comparing videotapes of the loaded core with core maps generated. following initial fuel loading of units 1 and 2. An affidavit by R. G. Cockrell (attachment 4 ) shows .that this verification has been completed on the unit 2 core and that it will be conducted prior to the testing of unit l.

Permission to load fuel in units 1 and 2 included the require-ments that all control rods be fully inserted, valved, out, and elect'rically disarmed; the,RHR system be aligned to cool the core in,the reactor vessel; 5

and all valves in lines which could drain the reactor vessel and. the RHR system in this mode be disabled in the position which will'not drain the

'eactor. In an affidavit o'f Thomas V. Wambach, of the Nuclear Regulatory Commission, dated May 5, 1976, Mr. Wsmbach made an evaluation 'of the safety of units 1 and 2 in the above described conditions, and concluded "Chat the core could be kept adequately cooled without reliance on any system which has been restored'after fire damage or whose design has been modified in the restoration work. " For the proposed testing the only plant condition that differs from the conditions required for fuel loading is that each control rod will be made 'operable, one at a time. At a11 times while a control rod is in use, all 184 rods not in use will be fully inserted, out, and electrically disarmed. During the proposed l'a1ved testing, the RHR system will be aligned to cool the core in the reactor vessel, and all valves in lines which could .drain the reactor vessel and the,RHR system will be disabled in, the position that 'will not drain. the reactor.

A safety evaluation for the worst accident that could occur with one control rod operational and the=-resulting effects on the margin to c'riticality and ability to cool the core is presented below.

Evaluation of Effects of Placin One Control Rod in eration NRC has previously approved TVA's loading of fuel into the unit 1 and unit 2 cores provided all control rods are fully inserted,, valved out, and electrically disarmed. Upon completion of this operation, all fuel will be in its correct'location-in the unit 1 and 2 cores, and all control rods will be fully inserted, valved out,'nd electrically, disarmed. The only change in plant conditions between those',described above and those

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0 that would exist during the performance of the proposed control rod. tests is that one control rod will be operational. All other control rods (184) will be fully inserted, valved out, and electrically disarmed.

This provides'ositive assurance that these 184 rods will not be with-drawn from the core. In order to adversely affect the ability to cool the core and the ability to keep it covered, as described in Mr. Wambach's affidavit dated May 5, 1976, a malfunction of the fire damaged equipment placed. in operation would have to be of such severity to make the reactor critical. The evaluation of the failure of the fire damaged systems used to conduct the tests proposed, in this affidavit for each unit has been made and is presented below.

Systems needed to conduct the proposed testing are as follows:

(1) Source Range Monitoring System, (2) Control Rod. Drive Hydraulic System, and (3) Reactor Manual Control System.

Unit 2 Anal sis For unit 2, none of the components or equipment necessary for the conduct of the proposed testing were damaged by the fire or modified as a result of the fire. Therefore, the safety evaluation as described in Mr. Wambach's affidavit dated May 5, 1976, is valid and applicable, and this testing can be conductedwith no reliance on fire damaged equip-ment. To prove that the reactor will always remain subcritical during the proposed. testing, the following safety analysis was performed,, in which the worst accident occurs which results in maximum reactivity insertion. The worst case accident is the failure of the one operational control rod such that it fails, in the fully withdrawn position. For

this control rod system failure to result in the maximum increase in core reactivity, this safety analysis assumes that the failed rod is the analytically strongest rod (rod 26>>07), as identified by the General Electric Company (see attachment 3). Unit,2 core average exposure is currently 2,165 MWd/t, and from attachment 3; the shutdown margin with the analytically strongest rod withdrawn is 3.15$ Ak/k.= This value was determined by General Electric Company's improved calculational methods recently reviewed by NRC. These methods result in calculations even more accurate than methods described in the Browns Ferry FSAR. The General Electric Company calculations are based on input data on the Browns Ferry core characteristics provided by TVA. In TVA's unit 2 technical specifica-tions, a value of 0.38$ Ak/k has been assigned to account for uncertainties in fuel content and uncertainties in calculating the analytically strongest rod. This value'ust be subtracted from any analytical determination of shutdown margin. After applying this 0.38$ Ak/k uncertainty value, the core will still be subcritical E

by at least 2.77fo Ak/k in the worst .case accident, one in which the analytically strongest rod. is fully withdrawn from the core. The withdrawa1 of any rod other than 26-07 will similarly not result in criticality. In the event that any accidental single rod movement occurs, there will still be no criticality in the reactor core and the test can I

be conducted without danger. Because the reactor will remain subcritical in the worst case accident, this.testing can be,,

conducted without endangering the safety of the public, with no reliance placed on fire damaged equipment or systems to maintain the reactor subcritical or to mitigate the consequences of an-accident.

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Unit 1 Anal sis For unit 1, as above for unit 2, the conditions that will exist during the proposed testing will be the same as those required. for fuel loading with the exception that one control rod at a time will be operational.

The ability to cool the core and to keep it covered will not be affected.

as long as the reactor remains subcritical throughout testing. Some of the systems to be used in the unit 1 testing were damaged by the fire and have been restored. For this analysis it will be assumed that this equip-ment fails and no credit will 'be taken for its ability to;mitigate the consequences of the worst case accident which could. occur during this testing. "

Effects of the failur'e o'f each of these systems, on the margin to criticality during the worst accident, one in which the operational 1

control rod fails in the fully withdrawn position, are analyzed. below.

l. Source Ran e Neutron Monitorin S stem With all rods not in use fully inserted, valved out, and electrically disabled, it is onlyC possible to fully withdraw one rod f om the c'ore at a time, which would, be the worst possible accident during this testing. From 'the attached Shutdown Margin Curve (Attachment 3) for 5,750 MWd/t exposure, the shut>>

down margin with the analytically'trongest rod withdrawn is 2.105 hk/k.

Applying 0.385 hk/k for uncertainties in fuel content and uncertainties'n calculating the analytically strongest rod, the'eactor will reamin subcritical by at least 1.72$ Lk/k. Since the reactor remains subcritical..

during the worst possible accident, the SRM "system will not be needed to serve any safety actuation function during this testing. Therefore, this testing can be conducted without endangering the he'alth and safety of the public, even if a complete failure of the SRM system occurs.

0

2. Control Rod Drive CRD $ draulic S stem - Portions of the h

electrical cables for the CRD hydraulic system for unit.l were damaged by the fire. The worst case accident that could occur if this system failed to function properly'would be the full withdrawal of the one operational control rod. All other rods will,remain inserted because of the fact that they will be va1ved out and electrically disabled.

Assuming the worst ca'se accident in which the analytica1ly strongest.,

rod fails in the fully withdrawn position, the reactor xemains subcritical by at least 1.72$ Ak/k. Therefore, it is concluded. that this testing can 'be conducted without endangering the health and safety of the public, with no reliance placed on the proper operation of the CRD hydraulic system to maintain the, reactor subcritical or to mitigate the consequences of an accident, provided all 'other rods are fully inserted, valved out, and electrically disarmed.

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3. Reactor Manual Control stem - With the mode switch in the refueling mode, the reactor manua1 control system prevents withdrawal of more than one rod at a time. However, portions of this system were fire.

P damaged, by the By valving out and electrically disarming all rods other than the one in use in the fully inserted: position, reliance on this system to ensure only one, rod is withdrawn at a time is eliminated. Under this condition, the worst case accident that could occur if this system failed would. be the full withdrawal of the one operational control rod.

Assuming that this fully withdrawn rod is the analytically strongest rod, the reactor remains subcritical by at least 1.72fo Ak/k. Therefore, it is concluded that this testing can be conducted without endangering the health 10

and safety of the public, without reliance on the proper operation of the Reactor Manual Control System to maintain the reactor subcritical or to mitigate the consequences of an accident, provided all other rods are fully inserted, valved. out, and electrically disarmed.

~Summar Based on the above analyses, which show that the reactor will remain subcritical during the worst case accident, it is concluded that the proposed testing can be conducted without danger of uncovering the core or risk of serious accident, providing that (1) all control rods other than the one being tested are disabled. in the fully inserted position, (2) the RHR system is aligned to cool the core in the reactor vessel, and (3) all valves in lines which could drain the reactor vessel and the RHR system in this mode are disabled in the .position which will not drain the reactor. For this analysis, no credit was taken for fire damaged and restored equipment to maintain the reactor in the subcritical condition or to mitigate the consequences of the worst accident which could occur during the testing.

JacR'. Calhoun Subscribed and sworn to me a~is;2/~'ijcc ..- i97<

)

Nota y Public commission expires -D" 7

0 0 Attachment j.

5FNP SS TEST NUMBER I) CONTROL ROD DRIVE SYS! EM 'escrfptlon The CRD tests pciformed during Phases II through IV Purpose of the startup test program are designed cs an extension of The purposes of the Control Rod Drive System test the tests performed during the prcopcrational CRO system are fa) to demonstrate that the Control Rod Drive ICRD) tests. Thus, after it is verified that ail control rod drives System operates properly over the full range of primary operate properly vvhcn installed, they cre tested pcriodi.

coolant temperatures and pressures from ambient to cally during hcatup to assure that there ls no significant operating, and fb) to determine the Initial operating binding caused by thermal expansion of the core characteristics of the entire CRO system. components. A list of all co~~ol rod drive tests to be performed during stertup testing ls given below.

CONTROL ROD DRIVE SYSTEM TESTS Reactor Pressure with Core Loaded Test Accumulator pslg fkg/cms )

Description Pressure Preop Tests 0 ~ 600 (422) 800 1682) Rated Position Indication ~ ll Normal Times inter tNitMrawn all 4e Coupling alf ageee Friction ~ 11 4e Scram Normal all ~ ll 4e 4e all Scram Minimum 4e Scram Iero Scram IScrem Discharge 4 {fullcore Volume High I.cvel) scram)

Scram Normal 4ee

~ valve refers to the four stewart cRD'e as determined from the normet eccvmv4tor peeeeure scram test et ambient reactor pressure.

ytuovehovt the procedure, she rove steweet cRtye" Implies the tour slowest compeubl ~ eetth rod earth rrcnimlser end cRD sertvenoe relsviremenre.

"scram times of the four eloweet cRD's eNn be determined et tttts, end t00% of rs!ed power dvrlrea p4nrevd reactor scremr.

"'Erteblhh tnh4gy shet thh check Ie normal operatina procedure.

ti4 NOTE: Sing! e CRO scrams should be performed with the charging valve closed fdo not ride the charging pump head).

1$ % 18

0 BFNP-e4 Criteria TEST NUMBER 6 SRM PERFORMANCE AND CONTROl. ROD SEQUENCE Levef 1 Each CRD must have e normal withdraw speed less Purpose than or equal to 3.6 Inches per second (9.14 cm/sec), The purpose of this test ls to demonstrate that the indicated by' full 12 foot stroke in greater than or equal to operational sources, SRM instrumentation, and rod 40 seconds. withdrawal sequences provide adequate information to The mean scram time of all operable CRD's must not achieve criticality and increase power in a safe and efficient 54 exceed thc following times: (Scram time Is measured from manner. The effect of typical rod movements on reactor the time the pilot scram valve soienoids are deenergizecl) power will be determined.

Scram Time Scram Time (Seconds) (Seconds) Description Vessel Dome Vessel Dome The operational neutron sources will be installed and Pressure Pressure Percent 0950 psig C950 psig source range monitor count rate data will be taken during Inserted (6(L9 kg/cms ) (66.9 kg/sms) rod withdrawals to critical and compared vilth stated 6 0.375 OA75 criteria on signal and signal count.to noise count ratio.

20 0.90 1.100 A withdrawal sequence has been calculated v3hich 50 2.0 2.0 completely specifies control rod vrithdrawais from the e4) '90 3.5 35 all-rodsin condition'to the rated power configuration.

Critical rod patterns will be recorded periodically as the

.The mean scram time of the three fastest CRD's reactor Is heated to rated temperature.

ln a two by two array must not exceed the following times: Movement of rods in a prescribed sequence is (Scram time is measured from the time the pilot scram monitored by the Rod Worth Minimizer eno the Rod valve soienoids are deenergized.) Sequence Control System, which will prevent outwf.

sequence withdrawal and insertions. 's Scram Time Scram Time the withdrawal of each rod group is completed (Seconds) (Seconds) during the power ascension, the electrical power, steam Vessel Dome Vessel Dome Pressure Pressure flow, control valve position, and APRM response will be Percent WSO psig C950 psig 5 recorded.

Inserted (66.9 kg/cms) (66.9 kg/sms) Date will be. obtained to verify the relationship between core power and first stage turbine pressure to 5 0.398 0.504 Insure that the RSCS properly fulfills its intended function 20 0.954 1.166 up to the required power level.

,60 2.120 2.120 e4) 90 3.800 3.800 Criteria Level 2 Each CRD must have a normal insert or withdravm Level 1 speed of 3.0+0.6 inches per second (7.62 sma,b1.52 There must be a neutron signal count.to noise count cm/sec), indicated by a full 12.foot stroke ln 40 to ratio of at least 2 to 1 on the required operable SRM's or 60 seconds. Fuel Loading Chambers.

With respect to the control rod drive friction tests, if There must be a minimum count rate of 3 55 the, differential .pressure variation exceeds 15 psid counts/second on the required operable SRM's or Fuel (1 kg/cm') for a continuous drive in. a settling test must be Loading Chambers.

performed, in which case, the differential settling pressure The IRM's must be on scale before the SRM's exceed shou'Id not bc less than 30 psid l2.1 kg/cms) nor should it the rod block set point.

vary'by more than 10 psid (0.7 kg/crn ) over a full stroke. Rod Sequence Control System shall be operable

'he Scram times with normal accumylator charge should as specified in the Technical Specifications.

fall within the time limits indicated on Figure 5,3-1 of the Starlup Tc'st Instructions. ~

T 13.6-19

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0 Attachment 2 I Part XI Section D l Page 8 TZST NUMBER 5 CONTROL ROD DRIVE SYSTEM Recovery Plan BPIP

'1. Deviation from urnoae description and 'criteria 11/u/75

a. Purpose - The FSAR calla'for demonstration of'RD system operation over the full range of primary coolant temperatures and pressures from ambient to operating. 'etermination of initial operating charac-teristics of the entire CRD system is also required. For the purposes of startup retesting it will be sufficient to determine initial operating characteristics by friction and scram testing at sero reactor pressure after fuel loading (the preop tests as 1iited will also be performed prior'o fue1 loading). Scram times will also be measured at rated t'emperature and pressure during heatup and/or low power retesting.

b, Description - According to .the FSAR the periodic CRD testing during heatup is done to assure that there ia no significant binding caused by thermal expansion of the core components. Since the tnermal expansion characteristics have already been proven, they Qi33. not require periodic

.testing during heatup for thet retest p'rogram. The control rod drive.

tests which will be required during startup retesting are position,

't indication, insert/withdraw times, coupling fricti.on, aero reactor pressure; and scram testing-of and scram all CRD's at rated testing

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, temperature and pressure, Additional initial atartup testing with accumulator pressures 1'arious and with,repeated confirmatory tests for selected rode has demonstrated expected design response and expected repeatability; therefore an extended retesting is not needed. The testing program will adhere to all technical specification requirements.

c. Criteria - No change.

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Part XI Section D Page 9 Recovery Plan

'BFNP TEST NUMBER 5 - CONTROL ROD DRIVE SYSTEM (Continued) 11/13/YS

2. Deviation from table 13.5-5 fre uenc STI 5 will only be performed during open vessel testing, heatup, and at 15-40X power as described in the purpose and description (see l.a and l.b above). The change cf the upper limit at 15-35X to 15-40X is consistent with technical speyification requirements and rod sequence control system limitations. Further testing at test conditions 2E, 3E, and 4E is not needed for reasons given in l.b above.

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'Attachment 3 ' ~

GENEBAL '~i"~ ELECTRIC POWER SYSTEMS "SALES OPERATlON GENERAL ELECTRIC COMPANY i '.........

CHATTANOOGA, TENNESSEE 37402 ~

832 GEORGIA AVENUE Phone (615) 894-2550 l I July 12, 1976 Mr. J. R. Calhoun Tennessee Valley Authority 702 Edney Building Chattanooga, TN 37401

Subject:

Browns Ferry.l and 2 Shutdown Margin GE Letter No. CF-78

Dear Jack:

Attached you will find. a copy"of the shutdown margin curve for Browns Ferry 1 and 2. This curve assumes.: strongest control rod withdrawn (Core Coordinates 26-0'7),cold (20'C) xenon-free condition.

Very truly yours,

~l'";.C"=iVGD W. E. Buist Q2'2 m j~(I'Jol Generation Sales Engineer WEB/lg Attachment

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GENERAL ELECTRIC POWER SYSTEMS I

SAJ.ES OP E RAT I 0 N GENERAL ELECTRIC COMPANY......... 832 GEORGIA AVENUE CHATTANOOGA, TENNESSEE 37402, Phone (615) 894-2550 July 19, 1976 Mr. J. R. Calhoun Tennessee Valley Authority 702. Edney Building Chattanooga, Tenn. 37401 Browns Ferry 1 and 2 Shutdown Margin GE Letter No, CF-80 4

Reference:

, (1) GE Letter No. CF-78 (2) NEDE-20913-P, "Lattice Physics Methods" NEDO-20939, "Lattice Physics-Methods Verification" '3)

(4) NEDO-20953, "3 Dimensional BWR Core Simulator" (5) NEDO-20946, "BWR Simulator Methods Verification"

Dear Jack:

The following should be added to our letter of July 12 (Reference No. 1):

"Control rod 26-07 is the analytically strongest rod through-out Cycle 1. This cur've was derived using methods documented in references 2-5. "

Very uly yours, RECEIQQ) 27edg~

Co-.n-W. E. Buist

~

Generation Sales Engineer .IJL >"'76

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Attachment 4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of TENNESSEE VALLEY AUTHORITY Docket Nos. 50-259 50-260 (Browns Ferry Nuclear Plant units 1 and 2)

AFFIDAVIT OF ROBERT G. COCKRELL Robert G. Cockrell, being du1y sworn, deposes, and says:

business address is Tennessee Valley Authority, Division of Power Produc-tion, 727 Edney Building, Chattanooga, assigned to Browns Ferry Nuclear Plant, Athens, Alabama. I am employed by the Tennessee Valley Authority as a nuclear engineer by the Division of Power Production. I have personal knowledge of the matters stated herein.

alifications B.S. Engineering Science Tennessee Technological University M.S. Nuclear Science and Engineering Virginia Polytechnic Institute and. State University I was hired by the Tennessee Valley Authority in the spring of 1975 and spent two months in the Plant Engineering Branch central office in Chattanooga before being sent to the Browns Ferry Nuclear Plant. I worked for approximately five months in the Browns Ferry Nuclear Plant Quality ASsurance Staff as a Quality Assurance Engineer. I was then

transferred to the Power Plant Results Section to prepare for fuel loading and startup shift coverage. During this period, I was designated cognizant engineer on several prospects including fuel assembly upper tie plate replacements, fuel assembly lower tie plate drilling for bypass flow holes, and operational startup source installation. For the past month, I have been standing coverage as a shift nuclear engineer for fuel loading on Browns Ferry units 1, 2, and 3.

Fuel Loadin Verification In'order to ensure the proper loading of fuel assemblies at Browns Ferry, the fully loaded cores are inspected to verify correct placement and orientation of all fuel ass'emblies.

The core of. each Browns Ferry reactor contains 764 fuel assembly bundles. Each of these'undles is permanently identified by a number etched into the assembly bail (handle), which is visible from above the fuel assembly.

Fuel bundles must also be checked for the proper orientation with respect to the control rod within each control cell ( 2 x 2 bundle array around a control rod). There are several ways to verify this or'ientation, but two methods are normally used. Each fuel bundle has a channel fastener in one corner which should point toward the center of the control rod when properly oriented. The operators use this method during fuel loading since it is easy to see this fastener from a distance.

Secondly, when properly oriented, the fuel bundle identification number on a fuel assembly handle (bail ) is always upright when viewed from the center of the control cell.

Upon completion of fuel loading, a video camera attached to the refueling boom is lowered to slightly above the fuel bundles. A monitor and. videotape machine are also connected to the video camera.

The fuel bundles are then scanned row by row in a prescribed manner producing a videotape that shows the identification number and orientation P

of every fuel bundle.

Browns Ferry units 1 and 2 were originally loaded to a pre-planned array specified in the startup program. (Browns Ferry Nuclear Plant FSAR Section 13.5. ) Verification that the cores were originally loaded correctly was made, as required by the Browns Ferry FSAR Section 13.5.1.2. This was achieved by making a vt,deotape of the serial numbers of the loaded fuel. A core loading map of the serial numbers was made from this tape. This core loading map was compared to the preplanned map and verified identical by a Tennessee Valley Authority employee and a General Electric startup engineer. The maps and tapes are available for inspection and have been audited by Nuclear Regulatory Commission and Tennessee Valley Authority Quality Assurance personnel for both units.

After Browns Ferry unit 2 was reloaded., videotapes of the core were made by Tennessee Valley Authority employees on July 19, 1976.

To ensure independent review of the tapes, I was not involved in the taping process itself.

Subsequently, I reviewed the tapes made on July 19, 1976, on a television monitor and observed. every fuel bundle, noting the bundle location, identification number, and orientation. Since the camera was

fixed, fuel bundle orientation was determined by observing the alignment of the identification numbers. As I observed each fuel bundle, I copied the identificat;ion number and orient;at;ion on a blank core map. Aft.er t;he t;apc were viewed in ent;iret,y and tho core map complet;ed, I compared my map against the core loading map, which was made at t;he. completion of the original fuel loading.' compared the two maps on a bund1e-by-bundle basis and found them to be identical. On this basis, I conclude that Browns Ferry unit 2 is loaded to the exact configuration established prior to the March 22, 1975, shutdown.

At the completion of fuel loading on Browns Ferry unit 1, I will follow the same core verification procedure.

Robert G. Cockrell Sworn and. subscribed. before me this ~ >'ay of .6 ., 1976.

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