ML18025B361

Annual Operating Rept,1980.

ML18025B361
Person / Time
Site:	Browns Ferry
Issue date:	02/27/1981
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18025B362	List: ML18025B361 ML19341D376
References
RH-81-1-BF-1, RH-81-1-VF-1, NUDOCS 8103050511
Download: ML18025B361 (40)

Text

TENNESSEE VALLEY AUTHORITY Division of Nuclear Power ANNUAL OPERATING REPORT BROWNS FERRY NUCLEAR PLANT J.

January 1;- 1980 December 3l,. 1980 Docket Numbers 50-259, 50-260, and 50-296 License Numbers DPR-33, DPR-52, and DPR-68 Submitted by As ant Director of Nuclear Power Submitted by

/'ower Plant Superintendent

TABLE OF CONTENTS

~Pa e Critical Systems, Structures, and Component Tests and Experiments for 1980 1- 19 Plant Modifications Summary 20 32 Fatigue Usage Evaluation 33 Chall,enges to or Failures of Main Steam Relief and Safety Valves 34 - 35 Occupational Exposure Data 36 37

CRITICAL'SYSTEMS STRUCTURES AND COMPONENT TESTS AND EXPERIMENTS FOR 1980 80-01 Temporary storage of low level radwaste in drums/boxes (containing LSA material) on trailers within the Browns Ferry Nuclear Plant (BFNP) security area until the waste can either be shipped to an offsite burial facility or transferred to an authorized on site storage facility.

s Unreviewed Safet uestion Determination

~uestton Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report increased? Yes No X Justification Temporary storage of low level radioactive waste in trailers until it can either be .

shipped to an approved commercial burial facility or transferred to an authorized site storage facility does not impact any of the safety considerations previously evaluated in the FSAR. These trailers are not stored in a location that 'could affect any CSSC equipment or its function. Temporary storage of low level radio-

}

active waste in trailers will not directly or indirectly affect the nuclear safety of the plant.

~notion, Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created2 Yes No X .

Justification The FSAR considered accidents which have a potential for serious releases to the environment but does not and is not intended to cover every routine activity associated with day-to-day plant operation. During normal shipping operations, it is not unusual to have four or five trailers onsite, as well as the reactor cleanup'nd low level casks. As long as the number of trailers onsite does not exceed these "normal" levels, the accident potential is no different than for normal operations.

~uestton Is the margin of safety as defined in the basis for any technical specification reduced2 Yes No X Justification The present BFNP technical specifications impose no limits on solid radioactive waste storage or handling. It should be noted, however, that the consequences of releasing all of the activity which is temporarily contained in the trailers would be far less severe than either the release resulting from the seismic event associated with the loss of integrity of the offgas system or the release resulting from a main steam line rupture. Both of these events have been analyzed in the FSAR from a radiological standpoint and determined to pose no significant hazards to public health and safety.

s? a, s

$ t NV > ~

80-02 Continued operation of Browns Ferry Nuclear Plant utilizing the three 0

, outside condensate'storage'anks'.

Unreviewed Safe{: uestion Determination

? s uestion Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety, Analysis Report (FSAR) increased?

2 t I

s ?, ~

Yes No X h

'Justificatio'n The header supplying core cooling water to the HPCI and RCIC systems is normally aligned to receive water from the condensate storage tanks. However, the'design basis for these two systems (FSAR Section 6.2, Item ll, Attachment A) is met by provisions for alterna'te? alignment to the'pressure'uppression ' pool" (torus).

?V, 1 jL' ue'stion"

'Is the possibilit'y for 'an accident or malfunction'of a different type than any.

'evaluated'previously'in the-Final Safety Analysis Repor't created'? Yes No X Justification The results of our analysis indicate that if the postulated release were to occur, radionuclide concv'entiati'ons at 'th'e'aresublic" water supply (Champion?Paper Company

's ae'M 233.0) would'be 'less than 2 pere'ent"of'PC w "defined in 10CFR20. Thus, such a release would result in negligible environmental impacts.

~uest1on Is the m'argin of safety'as defined'nhe"basi's'or 'any t'echnical 'specification X -.'""

reduced? r Yes No 1,

Justification'" """ " ' s t  ??? t The condensate storage and supply systems are not addressed in the BFNP technical specifications. t r

n ~a 80-03 Verify flow from core spray pump lA following wood block pieces introduced into the sy'tem.

Unreviewed'Safet 'uestion Determination-Based on the'res'ults of- the'safety'v'a1'uation,

'th'ere'is'no det'ectable flow. blockage it of-'-the'nit has been determined that

'1?system l,core spray. spargers.

'Followi'ng'erifica't'ion"of flow'through the pump'A.minimum flowlines, it,may be furthery'concl'uded tha't no 'other's'ignificant flow blockage is present in the core spray" sys tern'.

V s

The core spray nozzles are suffer,ciently small to conclude that any debris entering the reactor vessel through the nozzles would not create an accident of a cKfferent type than evaluated previously in the safety analysis report. Also, the maxgin of safety as defined in the basis for the technical'specification is not reduced.

Therefore, the consequence of an accident of malfunction of equipment important to safety will not be increased.

As a result of these determinations, it is concluded that operation with, the existing core spray system status does not constitute an unreviewed safety question as defined in 10CFR 50.59.

80-05 To allow use of the Visicorder Oscillograph to determine the cause for spurious trips of the scram channels recently observed at Browns Ferry unit 2. The test instrument will be used to monitor voltages at various points internal to the Reactor Protection System Circuitry.

Unreviewe'd Safet uestion Determination A review has been performed to verify that the test does not constitute an unreviewed safety question as presented in 10 CFR 50,59.

The isolation amplifiers described in the test are sufficient to isolate the test circuitry from the RPS logic circuits.

Applicable IEEE standards regaxding testing, isolation, and separation of redundant channels are not violated. We do, however, make the following recommen-dations:

1. Considering the safety significance of the RPS, diligent care should be employed in installing, monitoring, and ultimate removal of test instrumentation.

An instrument engineer or instrument mechanic foreman should make a daily check of the equipment status.

2. For each auto-scram channel (i.e., Al, A2, Bl,'2) that has test points connected, a test of the circuitry (half-scram) shall be performed following test equipment installation. Utilization of the test switches (5A-52) appears to be a convenient way to accomplish this.

3. If both logic trains within a scram channel have test equipment installed .

(i.e., Al and A2 or Bl and B2), the test described in step 2 above should be repeated daily.

80-06 STEAR BF-80-06 has been written to allow installation of a voltage recorder on RPS buses A and B in order to determine if abnormal RPS voltage has contributed to recent scrams of Bxowns Ferry Unit 2.

Unreviewed Safet uestion Determination A review has been conducted to verify that the test does not constitute an unreviewed safety question as defined in 10CFR 50.59.

Installation of the test instrumentatJon does not degrade thu opernblllty c>l:

the RPS. There are no postulated failures that can affect the RPS in an unsafe manner. The recording voltmeter was installed such that it is protected by the existing voltmeter fuses.

80-07 Monitor vibration levels on HPCI turbine and inlet steam pipe on unit 2.

This STEAR actually is concerned with the installation of some temporary accelerometer mounts via glue or special clamps.

Unreviewed Safet uestion Determination

~session Is the probability of occurrence, of the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis

'eport increasedf Yes No X Justification o

This STEAR is concerned with the temporary installation of some ac'celerometer

.mounts o'nly.

~neutron Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created'es No X Justification This STEAR is concerned with the temporary installation of some accelerometer mounts only.

Question Is the margin of safety as defined in the basis for any technical specification reduced'es No X Justification This STEAR is concerned, with the temporary installation of some accelerometer mounts only.

80-08 Install recording pressure transducer at 1 PT 73-4 to record supplemental data during performance of SI 4.5.E.1.D and E.

Unreviewed Safet uestion Determination uestion Is the pxobability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report increased? Yes No X

80-08 (Continued)

Justification Attachments to pipe and equipment are made without welding and have no effect on pressure boundary or seismic analysis of system. t uestion Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created7 Yes No X Justification Test does not alter design or operation of system in any way. Therefor'e, there is no possibility of, creating a new or different type of malfunction.

~ues cion Is the margin of safety as defined in the basis for any technical specification reduced? Yes No X Justification Attachments to pipe and equipment are made without welding and have no effect on pressure boundary or seismic analysis of system.

Test does not alter design or operation of system in any way. Therefore, there is no possibility of creating a new or different type of malfunction.

80-09 Proposes temporary modifications to two containment isolation valves in an attempt to reduce the closure time of the valves to '2.5 seconds or less.

This will be done by replacing the present solenoid actuators with larger actuators that allow faster venting of the purge valve pneumatic 'actuators.

~cotton Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Rep'ort increased'? Yes No X Justification The proposed tests are to be performed with the unit in cold shutdown and with the ventilation system out of service. The plant will be returned to its pretest configuration prior to startup. Containment isolation is not required when the reactor is in cold shutdown.

e question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created'es No X

80-09 (Continued) s n I Justification integrity and the ability to isolate the containment are not r'equired  !'ontainment during-,cold.,shutdown,.when,these<,tests wilIberun., An accident or malfunction 'of .

a type not previously evaluated is'ot',created

~uestion Is the margin, of safety,. as defineds in,the basis for any technical specification reducedt., Yes ...; No . X Justification The. proposed test,does not,,change the operating plant configuration. The new solenoids will reduce valve,cl'osure times. pr'oviding incieased "safety'm'argixis;(The "

solenoid valves will'e creinstalled prior'o 'operation" of th'e unit.')"'"'"" .'riginal a@is<<so/3  ! y <<a k<<$ jp'$

80-10 Rl Obtain vibration, position, and speed Y Fats '!<<e'- '(! of various HPCI'ystem comp'onentsud via temporary instrumentation.

Unxeviewed Safet uestion Determination dYO;3',Y'! f.<<u Of<<,gtf>>sf >'Juk,','s!<<Y.<<'b ", 'l,,is-',f rf 't;"s

$s the,,probability,cof, occurrence or the consequences of an accident or malfunction of equipment important to 'soafety "previously evaluated itn"the'Pinal Safety'A'nalysis '--

Report increasedt Yes No X""

s Justification

<< 'v Th$ s, STEAR 8010.

!4'<<>>-<<S<<O- P <<at ~,t,f, '/<<!P "tie ., "

Rl is, concerned','with'he'emporary"instrumentatio'n 7 S YQ Y" S 'y ', !, !VS of'HPCIoso that

<<g!steered.during;normal operability"test's and heat'-upotradn'si'ends'Y" (Cold "'-'ataymay,,be HPCI.witp,.hot.,xeactor) , This instrumeiitati'~ri Wll not directly, or indirectly affect the nuclear'afety or/cathe plant.

~action Is, tge., possibility .for,,an, accident or 'm'alf unctifon of a'dif percent 'type'than'any '-

'" ':: No'."='X' evaluated priviously in the Final Safety Anliysis'Reps'r't"c'reatedt'"Yes

... ~ YY<<

Justification This, STEAR 8010,Rl isconcerned with the temporary instrumentation of HPCI so that data may;,be,.gathered,,during,.normal, operability 'te'st's'nd hoeat"-up"'"tran!si'ent's';""'"

(Cold,HPCI. with.hot,reactor.);,In .my, opinion~ 'this,insttxumentation will"not'tdix'ectly or indirectly affect the nuclear safety of'he plant.

"'uestion Is the;margin of,safety, as.:~defined in the d dl Y N X ,

basis fog any

!i...;", .!.",: v!i! ra:.f~r.!" i,.

technical specification

80-10 (Continued)

Justific5tion This STEAR 8010 Rl is concerned with the temporary instrumentation of HPCI so that data may be gathered during normal operability tests and heat-up. transients.

(Cold HPCI with hot reactor.) This instrumentation vill not directly or indirectly affect the nuclear safety of the plant.

80-11 To install an isolator pad on EEL pump and verify that mounted structural critical of the machine is reduced to an acceptable level. This is to be verified by "Rap" test before and after installation and also by vibration measurements before and after installation.

Unreviewed Safet uestion Determination This STEAR is only to be performed with pump out of service. Therefore, it is judged nonsafety related.

80-12 To demonstrate the viability of interfacing the Rusco MAC/540 Cardentry controller with a host computer for the purpose of storing, processing, and reviewing plant vital area access data as required by NRC IE Bulletin No. 79-16.

Unreviewed Safet uestion Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report increased'? Yes No X Justification The Rusco cardentry system is not addressed in the FSAR and is not directly or indirectly related to a nuclear safety system.

~uestion Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created'? Yes No X Justification I

Since the Rusco cardentry system is neither directly nor indirectly connected to any nuclear safety system, there is no increased possibility for an accident or malfunction of safety related equipment.

t

80-12 (Continued)

Question Is thh'ar'gin 'of safety -'as 'defined'in "the 'basis 'fox any technical specification, reduced? .Yes' No" .X ".,

Justification The Rusco cardentry system is not addressed in the technical specifications and item.

therefore'will not reduce the margin of safety for any technical specification F

" a "I s e

80-13 The change will add instrume'ntation to'"monitor the sera'm disch'arge volume pressure during a reactor scram for test purposes. The scram to be monitored will be a scheduled scram.

Unreviewed Safet uesfion Determination es

~caution Is the probability of occurrence or the consequences of an accident or malfunction of equipment'mpor'hant'tosafety? previously tevalua'ted in'he Safety'-Analysis Report

.increased';"Yes.""'.'No

'ustification The change adds pressure instrumentatiOn to the'eactorc'o'olant'.pressure?boundary.

Failures of such instrumentation have been analyzed and fourld to be acceptable 0

in the FSAR. The materials, fabrication, and installation will meet the same requirements as similay components that are a part of the reactor coolant pressure boundatrv'.'

e

~uastfon Is the possibility for an accident or malfunction of a different type than any previously in the Safety Analysis Report cceated? Yes No X 'valuated Justification The materials, fabrication, and installation will meet the same requirements as similar components that are apart of the reactor coolant pressure boundary. The of,'the pressure instiumentatioyn 'does'"not" result in y'potential accidents

addition

'anr or malfunctions that were"not already presen6 with" other'pressure ins'trumentation 'n the reactor coolant pressure boundary.'?destine 4

"Is the margin of'safety 'as'defined'in"'the'asis for any"" technical spe'cification reduced'es No X Justification No existing technical specification is affected by this change.

80-15 The objective is to provide a temporary alarm to panel 9-6 in the un1t control room for high water detected in the'RD headers.

l I

Unreviewed Safet uestion Determination tjuestton Is the probability of occurrence or the consequences of an accident or malfunction of 'equipment important to safety previously evaluated in the Final Safety Analysis Report inoreased2 Yes No X Justification This STEAR is concerned with temporary installation of "informat1on only" alarms.

~nest%on Is the possibility for an accident of malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created2 Yes No X Justification This STEAR is concerned with temporary installation of "information only" alarms.

2juestion Is the margin of safety's defined in the basis for any technical specification reduced2 Yes No X Justification This STEAR is concerned with temporary installation of "1nformation only" alarms.

80-16 The objective is to provide a temporary alarm to panel 9-6 in the unit 2 control room for high water detected in the CRD headers.

Unreviewed Safet uestion Determination question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report increased2 Yes No X Justification This I

STEAR is concerned with temporary installation of "informat1on only" alarms.

Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created2 Yes No X Justification This STEAR is concerned with temporary installation of "information only" alarms.

v ~

80-16 (Continued) uestion Is the margin of safety as defined in 'the basis for any technical specificati'on reduc'ed2 Yes No R s

Justification This STEAR is concerned with temporary installation of "information only" alarms.

80-17 The objective is to provide a temporary alarm to panel 9-6 in the unit 3 control room for high water detected in the CRD headers.

Unreviewed.,Safet uestion Determination

~uestdon Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previou'sly evaluated in the Final Safety Analysis Report increased2 Yes No R Justification h

a This STEAR is concerned with 'temporary installation of "information only" alarms.

~caution Is the, possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report cr'eated? Yes No X Justification This STEAR is concerned with temporary installatio'n of information only" alarms.

~uestfon Is the margin of safety as defined in the basis for any technical specification reduced? Yes No X v

Justification This STEAR is concerned with temporary installation of "information only" alarms.

80-.18 The change will cut the caps off of the clean radioactive waste (CRW) drain stubs that receive the scram discharge header (SDH) vent lines Md to cut the drain, lines a few inches above the, stub. Wire mesh would be u'sed to keep debris,out of the drains. The Purpose of the change is to give additional positive venting of the system.'

80-18 (Continued)

Unreviewed Safet uestion Determination

~uestion The probability of occurrence or the consequences of an accident or malfunction of equipment'mportant to safety previously evaluated in the Safety Analysis Report increased2 Yes No X Justification Steam releases to the reactor building have been analyzed and found to be acceptable.

~nest ton Is the possibility for an accident or malfunction of a different type than any evaluated in the Final Safety Analysis Report created2 Yes No X Justification No accident or malfunction of a different type from that analyzed in the FSAR.

tluestion Is the margin of safety as defined in the basis for any technical specification reduced2 Yes No X Justification No existing Technical Specification is affected by this change.

t 80-19 This CRD scram discharge header test will be conducted to determine the flexibility of the scram discharge header for unit 3 for use in a seismic evaluation as required for IE Bulletin 79-14.

Unreviewed Safet uestion Determination uestion Is the probability of occurrence or'he consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased2 Yes No X Justification Per discussions with EN DES, adherence to the precautions listed in the Special Test Request will assure that no connections to the CRD scram discharge header will t

be overstressed during the test. The integrity of the CRD scram discharge header will not be threatened by this test.

~usst1on i

Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created2 Yes No X

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80-19'Continued) rc ~ - I /)" c" Justifilcation

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CRD'cramd'bosch'ar'geahead'er integrity'illrnot~be.,threatened'.if) the, ape'cial, requi'rement is. adhered to. Based'n this requirement,< no other'ossibilities' are foreseen.

~uestfon U " )""'t "<>>>> )'u <<)r>>, t j,,<<, T c K>>

Is the margin of safety as defined in the basis for any technical specification<<<<

reduced'es No X Justificati'on'".k >> ='<>>'>>" >"'>")

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No technical specification requirements are affected.

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80-20 ~

To obtain load current feeding from breakers on the attached breakers'ist, usi'ng clamp-on ammeters. These currents will be used by EN DES to study undervoltage analysis on I6'C buses.

>> >>>>t-r C lljt>> <<'>> >>' I'jll'r " ' r'r j<<c j

<< ~

1 Unreviewed Safet uestion Determination <<

uestion c

Is the probability of "occurrence 'or'.the)consequences<ofan,accident. or, malfunction.

of equipment important to safety previously evaluated in the Safety Analysis Report increased'es No X I

Justifi>>cr"atc'ion'"') o="'"'-':>n<<~' '."" <<<<"-i"-v -l,o c"'er discussion with. EN> DES", tthere )wild; be; no,.mechanical interface "in perforaing the sub]lect STEAR. The integrity of the system is not compromised by using the clamp-on'mmeter., This devide induces no transient;- or; additi'ona3. Load: to>,the, breakers.

No breaker will be interfered with during the test.

~ "T question.

'<<'1 Is the " po'ssib'i'lity">for 'an a'ccident or'.mal'function of a different 'type than ' -

any evaluated previously in the Safety Analysis Report created2 Ye's, N'o Justification

,The."use"of the clamp-'on<<ammeteroes not alter <<the design;)or change, the, load on the af fetcuted>>.I&C 'buses.'c) " 6'-<'

Question r .. ~, r c ~

Is'he margin of safety as defined'n the basis for any technical specification reduced'es No X c>>.<<I, I

80-20 (Continued)

Justification No technical specification bases are affected by performing the sub)ect STEAR.

80-21 Install annubar flow measurement device to verify adequate RBCCW flow to the reactor building equipment drain sump heat exchanger on unit 1 as over-heating problems in the sump have been experienced.

Unreviewed Safet uestion Determination

~uestion

's

'the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased? Yes No X Justification The drywell interior ambient temperatures have been found to be higher than predicted.

In order to determine the effectiveness of the RBCCW system, special testing is planned to determine the actual heat loads in the drywell and the capability of the RBCCS system for heat removal. To properly monitor the system parameters at various points,,it is necessary to install temperature and flow sensors. Temperature wells and flow connections must be added to the system piping so that the sensors can be used.

All cutting, welding, and material selection will be done in accordance with the piping classification requirements for the portion of the system piping affected.

Also, the seismic qualification of the affected parts of the RBCCW system will be examined to assure that the qualification is not changed by this modification.

Therefore, implementation of this ECN will not affect the RBCCW system's integrity or performance. Any changes found to be necessary following the testing will be evaluated separately and are not covered by the scope of this USED.

~ues tton Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created'es No .X Justification No other possibilities are foreseen.

~uestton Is the margin of safety as defined in the basis for any technical specification reduced'7 Yes No X Justification No technical specification basis is affected.

80-24 To investigate vibration/noise on" the recirculation system.

0 Unreviewed Safet uestion Determination tjuesiion e

s

-'s ,2 the probability of occurrence of the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report increased2 Yes No X Justification

, ~ s This test'involves the installation of passive accelerometers by mounting'he instruments at selected locations on the'ecirculation 'system piping using straps.-

The construction of the straps to be used is shown on Figures '2-5. The mou'nting blocks are'ade of carbon steel and the 'pipe clamps are spade of stainless steel.

Permanent insulation at these locations will have to be removed and replaced with

temporary'~<insul'ation with' minimum 6" thickness of calcium silicate. There will be:no 'penet'ration,'of any, safety-.related system pressure boundary and no additional "str'e's's'pla'c'ed 'on',.the systems'. "As a xesult, the probabi'lity of occurrence, or'he" 0

consequence of"an accident,"or, of. malfunction of equipment important"to safety previously evaluated in the'SAR .has not been increased.

2 s t

~uestion ':

.! s II Is the2 possibiH.ty for. an accident or malfunction. of a different type than any e'valu'ate'd"previously 'in! the, Final2>afety Analysis Report'reated'es '- No" <

X if 2, 's )

'1 Justification The system pair@ vibrations will be monitored during a normal plant startup with only brief hold periods at the vaxious recircu'lation pump speeds to obtain';

sufficient d'at'a; 'If any abnormally high vibration is detected during the period of data collection;.. the tlength of the hold period't that condition will be minimized. In view of the fact that no special plant'onditi'ons will be for the test, the possibility for an accident or malfunction of a different 'eeded type than any evaluated previously in the FSAR has not been created.

e ~

~uestton es Is th'e 'maxgin'f safety as defined in the basis for any technical specification teduoedi Yes No X

i 80-24 (Continued)

Justif ication The test consists of monitozing the recirculation system reactions using passive instruments attached to analysis equipment outside containment with temporary cables connecte'd thxough spare electrical containment penetrations. There appears to be enough available connections on penetration No. 100A EC to use for this test. This penetration was used in STEAR 7903 and is located above the drywe11 entrance. The system reactions are monitored during a normal plant staxtup period using standard procedures with no special modifications or opera-tional changes. With the temporary insulation in place over the areas where the acc'elerometexs are installed, no significant increase in containment temperatures should be encountered. The test instrument, cables, and mounting materials were sihXected to ensure that they are compatible with system materials and can with-stand the extreme environment. In view of these facts and the above considerations on the evaluations in the FSAR, the margin of safety as defined for the basis for any technical specification has not been reduced.

80>>25 Due to a discrepancy between the GE/MAC and Yarway water level indications, thezmocouples are to be installed to verify the temperature used in water level calibration.

Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report tncressed'l Yes No R Justification This STEAR attaches thermocouples to the Yarway instrument columns; it will in no way affect the level indication. It is a repeat of the start-up test to verify initial water level calibration. This STEAR will not directly or indirectly affect the nuclear safety of the plant.

~estion Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Repoxt created'es No X Justification This STEAR attaches thermocouples to the Yarway instrument columns; it will in no way affect the level indication. It is a repeat, of the start-up test to verify initial water level calibration. This STEAR will not directly or indirectly affect the nuclear safety of the plant.

80-25 (Continued)

'h r

~coction in the basis for technical specification'=-

, u Is the margin of safety as defined any reddced?...Yes::.'.....No ..X. h JuStifiCatiOn i' . ,=.~ s:s.?;? ..

.". '...'-s '. ... ,>.:i'?

> .. Y,< ..., ... . ', *

'Ah "3 sl. ifsd?r i,hh?j ",, e+.$ ', ?)'t f, Ih, . .

This STHAR attaches';thermocouples ?to'the.,Yarway instrument columns'; it will in h "h .'- i i,

no way affect'he- level;,indication. It is,a repeat of the' star't-.up test to verify "'

initial water>1eveQ;cali/ration. ,This,STEAR .will not dire'ctly or indirectly affect the nuclear isafetyi.of?;the'plant t s"

II h 80-. 26 )X-,.Gather"danae,necessary, to.,evaluate, whether a regulated "or 'uninterruptabl'e source to the,.RUSCO system is, required. 'powe'r h

II Unreviewed Safet uestion Determination

~destine

's the-probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Pinal Safety Analysis Report increesed? Yes No 'X

. Justification  ? s hh >

The'RUSCO" cardentry sy'stem;is not addxessed,in the,PSAR and is'not'ire'c'tly'"

or indirectly related to a nuclear safety system.

~ueseian Is the po'ssibility, for'n accident or, malfunction of a different typ'e than any evaluated previously;-in;,the-.Pinal,, Safety Analysis Repo'rt'reated'7"'-Yes' -No" X Justification hh

, ~

• Since the RUSCO cardentry system is neither directly or indirectly connected to '"

any nuclear safety system, there is no increased possibility for an accident ".=,,",.

~ s or malfunction, of, safety",related, equipment...

~coction Is the margin of safety as defined in the basis for any technical specification.h;",'.,

reduced'.,'Yes.)?:".,- No -,X'.;.;, " - ~

~ ."

Jus tifica tion.',.'.:,- i The RUSCO cardentry system is not addressed in the te'chnical'spsecifications'nd ?'h therefoxe will not reduce the margin of safety for any technical specification item.

l : 80-27 The purpose of this test is to obtain vibration and acoustic noise data from the outboard MSIV's to detect differences between the valves during normal plant operation to determine the source of continual failure of the valves during leak rate testing.

~notion Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis

'Report ineressed'? Yes No R .

Justification This test involves the installation of one accelerometer/clamp on one yoke rod of each of the four outboard MSIV's. The pressure boundary will not be penetrated and the movement of the bottom spring seat will not be inhibited by the location of the transducers or their instrumentation cables. Thus, the possibility of occurrence of an accident or malfunction of equipment is not increased.

~tention Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created2 Yes No X Justification This test involves the installation of one accelerometer/clamp on one yoke rod of each of the four outboard MSIV's. The pressure boundary will not be penetrated and the movement of the bottom spring seat will not be inhibited by the location of the transducers or their instrumentation cables. Thus, the possibility of occurrence of an accident or malfunction of equipment 'is not increased.

Question Is the margin of safety as defined in the basis for any technical specification -.

reduced2 Yes . No X Justification This test involves the installation of one accelerometer/clamp on one yoke rod of each of the four outboard MSIV's. The pressure boundary will not be penetrated and the movement of the bottom spring seat will not be inhibited by the location of the transducers or their instrumentation cables. Thus, the possibility of occurrence of an accident or malfunction of equipment is not increased.

80-28 The test will require the opening and closing of the generator breaker to record the response of the turbine generator during such transients. The test will be run in a cooperative effort with GE.

80-28 -"'(Continued) '

I r 2 Unreviewed" Saf et ues tion 2Determination

-~uestion

~, '!

Is the probability of occurrence or the consequences of an accident or malfunction'f

'equipment import'ant'o safety previously evaluated in the, Final Safety Analysis Repo'rt in c're'seed  ? 'Yes '~ No .. X 2

Justification I'he test involves the mounting of speed detectors and the turbine housing to monitor the turbin'e shaft: 'speed and the opening and closing of power circuit breakers. No safety-releated equipment will be affected or involved. As a result, "the probability of "occurrence, er the consequence of an accident,

or of malfunction'f equipment- important to safety previously evaluated in the FSAR haa nOt'een inCreaSedep e

~us stion Is the possibility for an accident or malfunction of a different type than any evaluated pr'eviously in'"th'e Final Safety Analysis Report created'7 Yes No X Y e Justification The transients resulting from the operation of the power circuit breakers will be very short lived and will result in a slight increase in, the torsional stress on'the ~turbine generator shaft. As a result, the. possibility for an,

<ri'l acciden't 'or malfunction of a'ifferent type than any evaluated previously fn'he FSAR has not been created.

~ue st i on Is the margin of safety as defined in the basis for any technical specification s

reduced2 Yes No X r

Justification )> ~

r r The operation of the power circuit breakers and the subsequent recording of torsional stresses on the turbine shaft will not reduce the margin of safety as 'defined- in th'e basis 'for any technical specification.

r2" I

sr >e2r will require " opening and s ~

80-29 Thisetest closing the West, Point 500-kV PCB 5204 to record the response of the turbine generator during such transients. The test will be run in a cooperative .effort with GE.

i,r <

Unreviewed'Safet uestion-Determination Question Is. the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Ansiysis Report increesed2 Yes No X

80-29 (Continued)

Justification The test involves the mounting of speed detectors and.the turbine housing to monitor the turbine shaft speed and the opening and closing of power circuit breakers. No safety-related equipment will be affected or involved.

As a result, the piobability of occurrence, "o'r the consequence of an accident, or of malfunction of equipment important to safety previously evaluated in the FSAR has not been increased.

Question Is the, possibility for an accident or malfunction of a different type .than anyevaluated previously in the Final Safety Analysis Report created'es No X Justification The transients resulting from the operation of the power circuit breakers will be very short lived and will result in a slight increase in the torsional stress on the turbine generator shaft. As a result, the possibility for an accident or malfunction of a different type than any evaluated previously, in the I

FSAR has not been created.

~uestion Is the margin of safety as defined in the basis for any technical specification reduced2 Yes NO N Justification The operation of the power circuit breakers and the subsequent recording of torsional stresses on the turbine shaft will not reduce the margin of safety as defined in the basis for any technical specification.

Modification

.ECN P0291 and 3 ',

January 1, 1980 Primary Containment;-,Uriits 1,'2,"

~

Plant Mode.fications

>> December.

Revised 'contro1

'alves-'as S

"required 31, 1980 Safet Evaluation

~ << ~ P<<~ <<<<<<,<<

"circuits'on various'coritain<<ment isolation to'prevent the valves'from automatically

'" 'reoperiing when'primary'containment- isolation'ogic is

""reset~'~ 'The ECN"'was 'completed'he S

2 modification was a result of NRC IE Bulletin 79-08 and.

NUREG-0578.,The containment isolation valves have'only.

mitigative functions, not preventive functions. Failure

.or'misoperation 'of-'the" containment isolation'valves wiU. not

'cause"an<<accident 'but"wi11"'in'cre'ase'the 'consequences of the accident,,The possibility for an accident of a different type than any previousIy evaluated in the FSAR was'not created. by the change.

- Residual'" ."Replaced.d'amaged vessel-'supports'on RHR'heat"exchanger'3B.

ECN P0292 Heat U

Reimval'ystem--t'"-' '-'he The ECN w'as'completed" modifications were necessary

'E. h...

to'estore

~

the heat",-."

exchanger structural supports, piping, and piping supports to a condition where it meets the original design'. criteria, thus resto. ring the heat exchanger to an operable and safe

"" '"'" 'c'ondition." Restoring'he" items to"the o'riginal design" removed the possibility of an .accident. different'm"any evaluated in I the FSAR.

DCR '1608 - .Radioactive Xnstalled threaded bushing for low-level cask cover head Waste Cask '-'"'Comon "

' ' bolts for'radioactive'w'aste-'ask'S'-'33>>'180.'"--The DCR was couagwe Me~ ~

The modification restores the integrity of the cask and does not affect any other system or component.

ECN L1845 <<RHR-LPCI Constructed 'the new battery room in the unit 3 diesel Modification - Unit 3 generator buiMing. Only a small portion of the work covered. by the ECN was accomplished.. This portion of the ECN will not affect the nuclear safety of the plant nor degrade 'the structure of the diesel generator building.

~ ECN P0205 - Reactor Revised the recirculation pump trip logic .from a two<<out;of-Recirculation System- four logic to a two-out-of-two logic for turbine stop valves Unit 3 to allow for reactor protection system surveillance testing.

The ECN was completed.

No other system is affected or altered by the modification.

The probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR was not increased.

Plant Modifications Summar January 1, 1980 - December 31, 1980 Modification Safet Evaluation Summar ECN L2106 - Demineralized Added block valves and test vents in the demineralized Mater Unit 1 water supply piping to allow testing of primary containment isolation valves 1383 and 1192. The ECN was completed.

Modification will, help assure the containment-boundary integrity. The probability or consequences of an accident or safety equipment malfunction will not be increased.

ECN P0079 Feedwater Changed feedwater system logic presently fed from System - Unit 1 I&C bus B to I&C bus A in order to prevent a unit scram.

The ECN was completed.

The modification made the system more reliable by assuring that the three logic channels are powered from separate sources, two of which would not be lost by any single event less than main generator trouble or trip.

1 ECN's P0301 and P0302-Primary Containment Unit 2 Drilled and tapped bolt holes in drywell equipment hatch X-1B at 6:00 o'lock and 9:00 o'lock locations to 1-1/8" 4 and rep'laced the 1" bolts with 1-1/8" bolts.

The ECN's were not completed.

The new bolts were of equal or better structural strength than the existing bolts. Also, it was confirmed that the reduction in the amount of the material due to drilling did not impair the structural or seismic qualification of the hatch.

ECN P0072 250V DC Replaced GE type CLF Class J fuses with Gould Elect.

System - Units 1 and 3 Manufacturing Company Shawmut Class J fuses in 48V and 24V DC power systems. The ECN was completed.

The new fuses are- rated for operation << 300 volts DC or less and therefore are fully qualified for the intended service. They increased the safety of the system by ensuring that fully qualified fuses are used.

ECN P0306 4kV Shutdown Provided a temporary second off<<site power source to the Board Unit 3 unit 3 4kV shutdown boards through the 4kV bus tie board.

This 'was accomplished by lifting wires around interlocks and jumper interlocks to allow closing breaker 1642 without tripping breaker 1612 or breaker 1622. Rl of the work plan returned system to normal as modification was only temporary. The ECN was completed.

The modification only affected the AC auxiliary power system associated with the distribution of off-site power. It did not affect the availabU ity or operation of the plant's diesel generators or the distribution method for emergency AC power.

Plant Modifications Summar January 1, 1980 December 31, 1980 Modification Safet Evaluation Summar ECN P0236 Core Spray Adjusted spring load on hangers as necessary to provide Residual Heat Removal adequate support for Velan swing check valves. Hangers Reactor Core Isolation involved were RCIC hangers H-29,'-31, and H-34; core Coolant Units 1 and 2 spray hangers H-ll and H-31; and RHR hanger H-13. The ECN was completed.

The modification assures integrity of the systems and assures compliance to design load. Normal operation of these systems was not affected nor any other systems.

ECN L2125 Main Steam Removed three Bergen-Paterson snubbers from main steam Unit 1 line B in the drywell and two Bergen-Paterson snubbers in parallel'arrangement on condensate bypass, line in turbine bulding and installed in their places E-systems hydraulic snubbers of the same rating as those removed.

The ECN will not be completed until these snubbexs are removed.

Snubbers were installed on a temporary basis for test.

The E-systeps hydraulic shock and sway arrestors have piston movement, lockup, bleed, and rating identical to the replaced Bergen-Paterson arrestors.

ECN P0242 Main Steam Removed packing bleed-off valves from main steam

. Unit 1 isolation valves and capped bleed-off lines. The ECN was completed.

Removing the bleed-off line valves and the capping of the lines did not alter the MSIV steam packing function as it is to decrease the likelihood of a steam 1eak along the stem nor did it affect MSIV performance.

ECN P0284 Main Steam Installed acoustic valve position monitoring system Units 1 and 2 for unit 1 main steam safety and relief valves. The system proyides information on the valve positions to the operator in the main control room and inputs the information to the plant computer. The ECN was completed.

This modification is an NRC requirement. The changes did not impact the operability of the valves; therefore, the margin of safety was not reduced.

Plant Modifications Summar January 1, 1980 December 31, 1980 Modification Saf et Evaluation Summar ECN P0224 Reactor Added various test connections to the recirculation Recirculation System pump seal water lines between valves68-507 and 508; Unit 1 508 and 550;t522 and 523;68-523 and 555. The ECN was completed.

The modification was required to assure compliance with NRC requirements. Normal 'operation of the s'stem was not affected. The test connections meet the same requirements with regard to piping class and seismic ~

status as the existing system.

DCR CC l/16 Neutron Replaced LPRM assemblies that had reached the end-of-life Monitoring Units' and 2 criteria. Performed pre- and post-installation electrical tests on LPRM detectors. Reuter Stokes model number RS-C6-1100-214, Westinghouse model number WL 24086 and General Electric model number 63C4179G027 (NA-200 Breeder Type) were used for the replacements. The DCR core component was completed.

The detector assemblies installed have specifications comparable to existing assemblies, longer lifetime than the existing detectors, and had no affect on plant safety performance, limits, or margin as defined by the operating license.

ECN L1692 RHR Service Installed flanged connections to RHRSW lines on 3B RHR Water Unit 3 heat exchanger. The ECN was not completed.

This modification allows removal of the heat exchanger head without having to cut and reweld large diameter service water lines. Operations and reliability of the RHR,or the RHR service water systems were not affected".

ECN P0150 Unit Preferred Modified unit preferred 16 sets and controls by 120-AC System Units 1 installing local and remote voltmeters to monitor and 3 battery armature voltage difference. Installed zener diodes, resistor, fuse blocks, and- associated wiring.,

The ECN was completed.,

The additions did not affect any other system and greatly improved the reliability of the present system.

Plant Modifications Summar January 1, 1980 - December 31, 1980 J

Modification Safet Evaluation. Summar ECN P0089 Associated Installed 'junction box with method for connecting Electrical Equipment portable test equipment both inside and outside Unit 1 unit 1 drywell. Pulled the cables and conduit from penetration EG to the junction boxes.

The only time the covers will be off the junction boxes and the wires energized will be when the unit is shut down and leak rate testing is underway. The ECN was completed.

Implementing, the ECN did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

ECN P0317 Primary Replaced existing bolting material with ASME SA 193 Containment Unit 1 bolts and'SME SA 194 nuts for penetrations X200A, X200B, and. X-6. The ECN was completed.

The replacement nuts and bolts are comparable to the existing ones, and the hardness tests assured that proper hardness was maintained. Using the new nuts and bolts did not reduce the functional abilities of the penetration hatches involved.

ECN $ 03L6 High Pressure Relocated four stud holes on unit 2 HPCI turbine Coolant In)ection System-coupling bearing pedestal purchased from Hope Creek Unit 2 Nuclear Plant and installed the pedestal. The ECN.

was completed.

The replacement pedestal has the same capabilities as the existing- pedestal in its undamaged state. Thus, the modification did not adversely affect its capabilities.

ECN L1382 Reactor Installed two restraints on each RBCCW riser pipe to the Building Closed Cooling drywell coolers in unit 1 drywell. The ECN was completed.

Water Unit 1 The additional restraints will help assure that RBCCW does not fail during system operation or a seismic event.

The modification could only improve the system.

0

~ ~ ~

Plant Modifications Summar January 1, 1980 December 31, 1980 Modification Safet Evaluation Summar ECN P0312 Feedwater Modifications were made to feedwater check valves 3-554, System Unit 1 3-558, 3-568, and 3-572 to allow the disc support.

bracket to be moved a slight amount, to align the disc seat with the valve seat driller and threaded holes into the disc to aid in lowering 'the disc into the valve without damaging the seats. The ECN was completed.

The modification did not reduce the integrity of the ~

valve body and disc. The modification allows temporary attachment of an eye bolt for lifting and positioning the disc in a safe manner during the valve repair process.

ECN'P0339 48'OV Shutdown Installed new bus from transformer TSIB to the bus Board Unit 1 of the normal feeder breaker of 480V shutdown board '1B.

The ECN was completed.

The new bus has at least equivalent capabilities and:

qualifications as the existing bus in its undamaged state. Therefore, the margin of safety was not reduced.

ECN P0164 Service Air Adds isolation valve inside unit 1 drywell between System Unit 1 valve 33-785 and the first tee; added isolation valve between 33-785 and FCV-33-10, and added vents and test connections. The ECN was completed.

Implementation of this ECN assured the service air ..

system has,been properly documented, qualified isolation valves and provided means of testing the leak tightness of the isolation valves. This represents an improvement to plant safety.

ECN P0300 Control Rod Installed flush connections in end caps of control rod Drive 'Unit 1 drive discharge header, west side only. The ECN,was

'ompleted.

The modification did not change the scram di'scharge header overall volume, the level switch setpoints or the level switch functions. The flush line discharge will be processed through the contaminated radwaste system. Based on these considerations, no other possibilities are foreseen.

Unit 1

'emoved ECN P0155 Main Steam -existing 10" wide flange beam and erected new 10" wide flange beam at higher elevation,to eliminate interference due to changing three stage Target Rock Relief Valves to two stage Target Rock Relief Valves.

Only a small portion of the work covered by the ECN -was accomplished.

This portion of the modification did not affect the availability or normal operation of the system.

Plant Modifications'ummar January 1, 1980 December 31, 1980, Mod ification Saf et Evaluation Summar ECN P0142 Primary Installed test,'vents on the suction and discharge .

Containment Units 1 of the drywell torus differential pressure compressor and 2 to allow leak rate testing of primary containment isolation valves FCV-64-31, 64-34,64-139, 64-140, and 84-20. The ECN was complet'ed.

The test vents installed are of the same TVA pipe class as the lines they connect to. The testing made available by this modification adds assurance that the'rimary containment isolation valves in the system perform the isolation functions.

DCR Core Component //15 Replaced existing fuel assemblies with pressurized Fuel Storage and Handling 8X8R fuel assemblies. The DCR core component is a Units 2 and 3 continuing modification until this type of .fuel .

assembly is used up..

The fuel assemblies which were replaced, were no longer capable of; providi'ng full reactor power. The new pressurized -8X8R fuel is the vendor's current design ECN' L1383 and L2104-Residual Heat Removal

'Sysem - Unit 3 and will provide superior performance.

Installed in the RHR.

an additional isolation valve, FCV-74-150, loop cross connect to backup existing valve FCV-74-46. Added a 2-inch drain line between o

the new manual, valve, FCV-74-150, and the existing motor operated valve, FCV-74-46., The ECN's were completed.

The modification will not interfere with the RHR operation.

The valvedoes not,significantly decrease pressure, as water flows through, it. The possibility of a'ifferent type of accident or malfunction was not created by this modification.

DCR 1540 and ECN P0154 Installed temporary. neutron and gamma dosimeters to Containment -'rimary monitor radiation, spectra in the reactor beltline,region Unit 3 for research purposes. The dosimeters were installed in November. of 1978 and were removed in August of 1979.

The ECN and the DCR were completed.

f r The dosimeters wereinstalled, removed, and were not required, for, any.-safety .function. Therefore, the probability or consequences of an accident or safety equipment malfunction was not increased.

'26-

0 I Plant Modifications Summar January 1, 1980 December 31, 1980 Modification Safet Evaluation Summar ECN P0373 - Standby Replaced SLC pump suction drain valve, 2-63-505; with a Liquid Control Unit 2 1-inch Hancock valve of similar construction. The ECN was completed.

The new valve has the same seismic and code classification as the old valve and it also has a higher pressure rating. The margin'f safety defined in the Tech Specs was not reduced.

F ECN P0350 - Feedwater Reversed polarity on diode IN 4499 in panel 9-9; System Units 2 and 3 TB1-5, and 2-1, and 2-9, points D and E. The ECN was completed.

This modification was necessary to ensure proper operation of the flow indicators for the reactor feedwater pumps flow rates. The indicators are only used for operational information. This change did not introduce any new modes of failure for the system.

ECN P0338 Reactor Core Replaced existing rupture disk, Fike Metal Products Isolation Cooling System Model, 8-PLHOV, with an improved model, Fike Model Units 2 and 3 No. 8-C-PVC. The ECN was completed.

The replacement disk did not affect the design of the exhaust portion of the RCIC system. There was no change in setpoint and the new disk uses a stainless steel vacuum support plate. The replacement disk did not adversely affect the RCIC system.

ECN P0283 Reactor Building Removed existing ventilation ducting in the steam, Ventilation Unit 2 vault back to the east wall, installed grill and, replaced 18" x 18" duct outside the steam vault back to the main branch line and replaced with 18" x 27" ducting;.

The ECN was completed.

The modification increased the air flow through the steam tunnel thus reducing the temperature and the probability of a spurious MSIV closure. The change did not affect the temperature sensors or the setpoints in any way.

ECN P0078 RHR Service Modified the RHR Service Water pump discharge check Water Common valves to use spring-assisted disc closure.'he ECN was totally completed.

The functions of the valves are unchanged except that it will better p'erform by adding the springs as they will assist disc closure.

4~

Plant Modifications Summar I

January 1, 1980 December 31, 1980 I

Modification Saf et Evaluation Summar

'ECN L1931 Control Rod Decontaminated the CRD scram headers using hydro-laser Drive Unit 1, process and installed piping for connecting inboard and outboaid 'headers together. The installation of "A" lo'op on A and B headers was deleted by ECN P0300 prior to work being accomplished. The ECN was not

,completed.:

The modifica'tions did not affect the ability of the CRD system to operate as designed. Therefore, the margin of safety'as not reduced.

~'

ECN P0353 Control Rod Rerouted the,l-inch CRD scram header vent to DRW floor Drive Units 1, 2, and 3 drains. Cov'ered the DRW floor drain with a splash shield. Capped the 4-inch DRW drain pipes and sealed with weld. The"ECN was completed.

The modifications were performed to prov'ide additional positive venting of the system. Based upon bounding analysis and equipment review, the modifications did not have a significant adverse impact on plant nuclear safety.

ECN P0277 Main Steam Replace the existing transducer mounting studs at Unit 2 each MSRV tailpipe with a vendor supplied replacement.

The ECN was completed.

The mounting studs are a'axt of the monitoring system and do not interfere with the relief valve operation in any way. Therefore, the margin of safety is not affected.

l DCR 1884 Main Steam Replaced fiber bat insulation with calcium silicate Units 1 and 2 block insulation on main steam and feedwater lines in the steam vault room. Insulation from 2-1/2 inches up to 6 inches was used. The DCR was completed.

The modification decreased the normal heat load in the steam tunnel and the probability of a spurious MSIV closure. The increase in pipe stress and hanger load due to the added insulation is essentially insignificant.

There is no probelm with seismic qualification.

DCR 1755 Main-Steam Installed new floating stellite stems in MSIU's, Unit 2 2-1-26 and 2-1-15, that failed the local leak rate test. The'CR was completed.

Substitution of the disc matexial provides a disc less susceptible to steam cutting of the suxfaces. Operation of the valves should not be affected and leak tight integrity improved. The max'gin of safety was not reduced.

0 Modification Plant Modifications Summar January 1, 1980 December 31, 1980 Safet Evaluation Summar ECN P0148, Unit Preferred Separated the power supplies for reactor manual control 120V AC System Units 1 and feedwater controls by utilizing a spare breaker and 2 in panel 9-9. Feedwater control power circuit breaker 602 was used for reactor manual control and CRD hydraulic instrument power circuit breaker 601 was used for The ECN was completed. 'eedwater.

The modification provides independent breaker protection to the two systems, thereby preventing a fault in'one immediately affecting the other system resulting in signal generated. The power supply to the two a'cram systems was not altered.

ECN P0331 - High Pressure Installed bolt and dowel steel guide blocks to unit 2 Coolant Injection Unit 2 HPCI sole plate. The ECN was completed.

The replacement of the weld on the guide blocks with the bolt and dowel did not alter the design or function of the governor and pedestal sliding feet.

ECN P0320 Radwaste Installed six monitoring wells outside of the low-level radwaste storage perimeter. Only a small portion of the work covered by the ECN"was accomplished.

These wells are located outside the security fence and in no way affect plant operation.

ECN P3000 Primary Replaced existing GEMAC transmitters with Foxboro Containment Unit 1 transmitters for PT-64-50, 51, and 67. The ECN was completed.

The replacement items are compatible with the existing equipment, perform the same functions, and meet -the-requirements specified for the actions performed.

Their use does not adversely affect existing analyses; ECN P0384 Primary Increased 1/4-inch NPT on top cover of the operator Containment Unit 2 for 76-24 to 1/2-inch NPT. The ECN was not completed.

The isolation valve will still perform the functions ',

as before with a more rapid closure time that does,.

not affect the valve's function.

h P2084 High Pressure I'CN Modified HPCI restraint R-16 to allow for clearance Coolant Injection System restraint and adjacent disk. The ECN was not, 'etween Unit 2 completed.

This modification did not change the system design. It adds assurance that the- system can meet the original design requirements for seismic qualifications.

Plant Modifications Summar January 1980 December 31, 1980 Modification Saf et Evaluation Summar ECN L2091 - Main Steam- Increased the existing set pressure of the main steam Units 2 and 3 relief valves by 25 psig, from 1080, 1090, and 1100 psig to 1105, 1115, and 1125 psig. Also, increased the set pressure of the main steam safety valves from 1230 psig to 1250 psig. The ECN was completed.

Implementing the ECN did not reduce the margin of safety as defined in the basis for any Tech. Spec. NRC approval was obtained prior to the unit being returned to operation.

DCR 1250 Main Steam Replaced existing DC pilot solenoid coils with 270V DC Unit 1 rated coils. for the MSIV's in unit 1 drywell and steam vault. The DCR was completed.

The replacement of the coils should improve operational reliability. and will not affect the safety-related aspects of, the MSIV's ability to close when signaled and closure times. The'margin of safety was not reduced.

I I

ECN P0046 - Reactor Provided a source of cooling water for the RWCU pump Water Cleanup Unit 1 seals. The source of the cooling water is from the CRD system during unit operation and from the RHR system whe the unit is shut down. Only a portion of the work covered by the ECN was accomplished on this work plan.

The instrumentation portion of the work was covered under a separate work plan.

The modification will help reduce seal failures and improve the reliability and availability of the RWCU pumps'eplaced ECN P0337 High Pressure turbine exhaust line rupture disk, Fike Metal Coolant Injection Units Products Model 16-PLHOV, with an improved rupture 2, and 3 disk, Fike Metal Products Model 16-CPV-C. The ECN was completed.

The replacement disk did not adversely affect the HPCI system. There is no change in the setpoint, capacity, function, or quality of the component. System reliability was not reduced.

ECN's F0052 and P0116- Replaced the existing feedwater spargers. Removed Feedwater System feedwater nozzle bore cladding and heat affected Units 2 and 3 base metal. Machined feedwater nozzle safe end to accept the dual piston ring seal, interference fit spargers with forged tee design and orificed elbow discharges. The ECN's were completed.

Implementing the ECN's reduced the probability of forming cracks and increased the assurance of maintaining vessel integrity. Analysis performed by GE concluded that the modifications were, within 0

cpde requirements. No Technical Specification safety margin was reduced.

Plant Modifications Summar January 1, 1980 December 31, 1980 Modification Safet Evaluation Summar ECN P0271 and FCR-134 Removed packing leak-off valve from valve 69-1,and Reactor Water Cleanup installed a 3000 psi stainless steel pipe plug into Unit 1 the packing leak-off threaded hole.

Valve function was not altered by this modification.

No Technical Specification safety margin 'was affected."'

Other valves may be modified on an as-needed basis at a later date under this ECN.

ECN P0321 - High Pressure Replaced the set screw that attaches the overspeed Coolant Injection trip mechanism'iston to the tappet and ball holder Units 1 and 2 with a stainless steel cotter pin. The ECN was completed.

1 The modification did not alter the design or intent of the overspeed trip assembly of the HPCI system. This is a supplier recommended product improvement associated with the HPCI turbines.

ECN P0157 - Core Spray Removed the lube oil coolers in the upper reservoir Units 1, 2, and 3 of the core spray pump motors and capped the inlet and outlet EECW pipes to the coolers. Only a portion, of the work covered by the ECN was accomplished.

Removal of the oil coolers had no adverse effect on the pump motor seismic qualifications. No Technical Specification basis was affected.

ECN P0201 - Unit Pr'eferred Changed normal feeder for board 9-9, panel 5, from .the 120V AC System Unit 1 plant nonpreferred bus to the unit preferred bus..

Changed feeder responsible for previous unit trips from the plant preferred bus to. the unit preferred distribution panel on board 9-9. The ECN was completed.

The probability of a unit trip due to loss of'ower to critical loads was greatly reduced by this 'modification.

The probability of occurrence o'r the consequences of'an =

accident important to safety was not increased.

ECN P0319 - High Pressure Fabricated and installed an angle plate to Coolant Injection Unit 'HPCI governor end pedestal. was completed.

reinforce'racked 2 The ECN This modification was performed as a temporary-fix until a new pedestal can be installed. The repaired pedestal was verified fully capable of supporting its design loads and assures HPCI system operability.

Plant Modifications Summar January 1, 1980 - December 31, 1980 Modification Safet Evaluation Summar ECN P0267 Residual Heat Provided chain driven operators for valves HCV-74-49, Removal Reactor Water 55, 69, and HCV-69-500. The ECN was completed.

Cleanup Unit 2 The addition"of the operators did not affect the seismic ~

integrity of the system. The. operators allow operation of the valves without the use of specia'1 scaffolds and did not affect the process barriers whatsoever.

ECN P0289 Various (1) Repaired and modified HPCI restraint R-32 in unit 3; Systems Units 1, 2, and 3 ,(2) Installed modified support R16 on RHRSW piping in unit 1; (3) Removed existing lugs to HPCI restraints, Rl and R2. Rewelded two new lugs to each restraint 3" x 1/2" x 12" in unit 1; (4) Installed R-10 on radwaste sump pump discharge and removed Rl and existing hanger P-6 in unit 1; (5) Added channels to complete restraint RCW R-16 (RBCCW H-4) in unit 1; and (6) Added additional bracing to RBCCW restraint, R-57, in the unit- 2 drywell.

Only a portion of the work covered by the ECN was accomplished; These modifications to the system will, ensure that safety-related piping conforms to the seismic de'sign criteria and will not be damaged by a seismic event. Modifications were performed per NRC Bulletin 79-14.

ECN P0093 Primary (1) Fabricated box beams to be used as supports for Containment Units 1 and 2 quenchers to be installed in unit 1 suppression chamber; (2) Fabricated quencher and. quencher support collars that will be added to the safety relief valve discharge lines in unit 2; and (3) Fabricated quenchers. for MSRV tailpipes, quencher support collars for unit 1 MSRV discharge, and the plate material for reinforcement of the torus vent headers.

Only a portion of the work covered by the ECN was accomplished. Only prefabrication work was done for the long-term torus integrity program. No work has been done in the plant. Therefore, no system in the plant has been affected by the change.

ECN P0214 - Main Generator Installed generator breaker and associated equipment to

'Unit 2 the main generators isolated phase buses. The ECN was completed.

The addition of the generator breaker has no affect upon normal operation of this or any other system, yet the use of an additional auxiliary power supply.

it allows 0

FATIGUE USAGE EVALUATION n

The cumulative usage factors for the reactor vessels are as follows as of December 31, 1980:

Usa e Factor Location Unit l Unit 2 Unit 3 Shell at water line 0.00497 0.00381 0.00315 Feedwater nozzle 0.24040 0.1608l 0.11746 Closure studs .0.09766 0.13426 0. 09816 0

-33>>

CHALLENGES TO OR FAILURES OF MAIN STEAM RELIEF AND SAFETY VALVES 0

Unit l 05-'06-80 at 0520, the reactor isolated as required by RPS. All 13 MSRV's operated normally.

05-07-80 at 1239, the reactor isolated as required by RPS. One MSRV operated normally with the remaining MSRV's not being required to respond.

06-17-80 at 0654, the reactor isolated as required by RPS. Two MSRV's operated normally with the remaining MSRV's not being required to respond.

07-22-80 at 2108, the reactor isolated as required by RPS. Three MSRV's operated normally with the remaining MSRV's not being required to respond.

09-01-80 - at 1341, the reactor isolated as required by RPS. Three MSRV's operated normally with the remaining MSRV's not being required to respond.

09-24-80 at 2222, the reactor isolated as required by RPS. Six MSRV's operated normally with the remaining MSRV's not being required to respond.

R Unit 2 04-15' at 0715, the reactor isolated as required by RPS. No MSRV's were required by RPS. No MSRV's were required to respond. 'However, one MSRV, simmered (opened and closed) for a short while after the scram. The following unit startup indicated the valve is again leaking. The valve simmered until repaired during fall refueling outage.

'04-22-80 at 1056, the reactor isolated as required by RPS. One MSRV operated normally this was the same valve which continued to simmer until R.O. The remaining MSRV's were not required to respond.

06-28-80 at 0852, the reactor iso'lated as required by RPS. The same simmering MSRV responded t'o relieve pressure while the remaining MSRV's were not required to respond.

08-16-80 at 0539, the reactor isol'ated as required by RPS. Seven MSRV's operated normally with the remaining MSRV's not being required to respond.

0

Unit 2 (Continued) 12-,'27-80 at 0227, the reactor isolated as required by RPS. Seven MSRV's operated normally with the remaining MSRV's not being required to respond.

There were no failures of the MSRV's to respond as required for all three units during 1980.

N NU'(BER OF PERSONNEL GREATER THAN 100 HREH TCTAL HAhl-REH GREATER THAN 100 tlREH STATION OTICITY COIITRKCTYURKEKS STATION UTIL'ITT CONTRACT MORKERS TS WORK AWD J()B FUNCTION E'IPLDYEES EHPLUYEES AND OIHFRS EHPLOYEES EHPLOYEES AND OTHERS REACTO< OPERATIONS AND SURVEILLAttCE HA It(TEtkAt1CE ()IECH.T HAtftTEWANCE (CLEC ~ ) 095 OP ERAT IUWS 0."te HEALTkt PHYSI ICS RESULTS LINET~K MEIII 12 ENGINEERIt(G (SUPERE)

OTAi-:R 5, .5995 ROUT It(E HAINTENANCE C

~EATII H<INTENANCE HE H ~ )

HAINTEt(ANCE NA (ELEC ~ )

2 77 9

150 36.6'695 TE 559 9'5 HEALTH PHYSICS 31 35 12 ~ 2 099 1856 RES(ICTS ( 1NS (ELECT s ) 3 b5.9 ENGINEERING (SUPER E ) 41 166 (5097 10692 195 OTHER 2 tee 2 It(SERVICE INSPECTION

~KTRTETTXE~E HAltETEWANCE )

OKCRAT TORA HEALTH PHYSICS.

KEEOETS I ffiSTR ENGINEERING (SUPER, ) 1 ' Oe2 0 HER SPECIAL HAINTEttAtfCE NA ILITETIANCE TM~%. 37 652 MAlttTEttANCE (ELEC 9 ) 26 10 '

OPERATIONS HEALTH PHYSICS RESULTS ( INS~s )

ENGlt(EERING (SUPERe ) 092 Oef9 OTHER WASTE PROCESSlt(G PADWASTE 12 2~

HA INTEt(AtlCE (ELEC ~ )

HEALTH PHYSICS RESULTS ( It(5TROOHEH)

E"Glt(EERII(G (SUPERE )

OTHER

'REFUEL IWG H INTENANCE (ME(H ~ )

HAItiTEttnttCE (ELEC.)

OPERATIOttS 109 HEALT5( PHYSICS EEL'OCT S fiffST~EKT 6 Ef(GlttEERIttG (SUPER, ) 2 093 leb 093 OTN R OTHER KE IITTEIT HA INTEttANCE (ELEC ~ )

OPERA'T If)tiS HEALTH PHYSICS RES~ULTS I NSTIDRTK I

\ FNI:INFFff tNf: I Cll(IFf( ~ 'I,

-;, MAINTENANCE (MECH+)

(ELEC+)

354 874 176 12 0

IS IeO 37 ~ 1 604o3 106 '

-9o5 MAINTENANCE 81 OoO P 0 68 F 6 0 ' OeO HEALTH PHYSICS 31 5 35 ,12m 2 0' 18' RE ( TR CHEM 3g 0 0 69 ' 0 0

~ '

ENGINEERING (SUPER') 176 12 4 ~2 109 ' 2~ 1 2 9 0 3 0 4 '

r G ANO T 779 1233 68 409(6 821 ~ 8 34 ~ 4

T