ML20216J974

Forwards Follow Up to 900530 Presentation on Rmov Boards 2D & 2E & Mod of Response to NUREG-0737,Item II.B.2, Plant Shielding

ML20216J974
Person / Time
Site:	Browns Ferry
Issue date:	11/13/1990
From:	Wallace E TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM NUDOCS 9011190169
Download: ML20216J974 (8)

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O TENNESSEE VALLEY AUTHORITY

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j CHATTANOOoA. TENNESSEE 374o1 l

$N 157B Lookout Place NOV 131990 i

U.S. Nuclear Regulatory Commission ATTN Document Control Desk Washington, D.C.

20555 Centlement In the Matter of

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Docket Nos. 50-259 Tennessee Valley Authority

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50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - FOLLOW-UP TO THE TVA PRESENTATION OF MAY.30, 1990 ON RMOV BOARDS 2D AND 2E AND MODIFICATION OF TVA RESPONSE TO NUREG 0737. ITEM II.B.2, PLANT SHIELDING i

On May 30, 1990 TVA met with members of the NRC staff to present justification for the removal of 480 volt RMOV boards 2D and 2E from the 10 CFR 50.49 list at i

BFN. ' Enclosure 1 of this letter is to provide clarification of the information I

presented on May 30,-1990.

In addition, based on reviews of the actions necessary to address the RMOV board issues, TVA considers it necessary to modify the information previously

' presented to NRC on NUREG 0737. Item II.B.2, Plant Shielding. As discussed in-of this letter, TVA consi'ders this item to remain closed.

t Finally, Enclosure 3 provides a list of the specific commitments made by TVA

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.in this letter.

Very truly yours, NESSEE VALLEY AUTHORITY

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w E. G. Wallace, Manager Nuclear Licensing and Regulatory Affairs Enclosures cc See page 2 A*

w 9011190169 901113 PDR ADOCK 050002'59 PDC p

g An Equal Opportunity Employer i

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+ j U.S. Nuclear Regulatory Commission g {g g cc (Enclosures):

.Ms. S.-C. Black, Deputy Director 3

Project Directorate 11-4 U.S. Nuclear Regulatory Commission

.j One White Flint, North 11555 Roew-ille Pike, i

Rockville, Maryland 20852 p

NRC Resident' Inspector Lt Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35609-2000 Mr. Thierry M. Ross, Project Manager U. S. Nuclear Regulatory Commission One White Flint, North c

11555 Rockville-?ike Rockville, Maryland 208 l'

Mr. B. A. Wilson, Project Chief U.S.; Nuclear Regulatory Commission Region Il-101 Marietta Street, NW, Suite 2900 Atlanta Georgia 30323

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ENCLOSURE 1 is Oubject: FOLLOW-Up TO THE TVA PRESENTATION OF MAY 30, 1990 ON RMOV BOARDS 2D AND 2E i

TVA stated in the presentation of May 30, 1990 that the only accident that would produce a harsh environment at the location of the RMOV boards is a RWCU system liigh' Energy blue Break (llELB) outsido primary containment.

Following such an event, the RMOV boards would be conservatively assumed to fait due to an elevated temperature greater than 140*F.

The failure of the i

RMOV boards would result in an inability to initiate the Low preocure Coolant j

Injection-(LPCI) ' mode of Residual lleat Removal (RHR).

Ilowever, core cooling i

would be provided by the core spray system in addition to operation of one loop of RilR in the suppression pool cooling mode.

Both loops of core spray are qualified for operation following a RWCU'line

-break outside containment. This provides the redundant low prosauro core

. cooling capability-required to meet the safety design basis, without the use of the LpCI mode of RllR.

In addition, operation of one loop of RHR in the suppression pool cooling mode would be required.

Following a.RWCU line' break, one loop of suppression pool cooling is required-to remove the heat from the suppression pool that would result from lipC1/RCIC operation and/or safety / relief' valve blowdown. The suppression pool cooling i

mode of.RilR is normally initiatied from the control room..llowever, for this particular event manual' actuation will be required based on the postulated

. failure of the RMOV boards following a RWCU line break.

The manual. actuation would require entry into the reactor bullding following j

. isolation of: the line break and requic as evaluation under the-criteria of NUREC 0737, item II.B.2, plant Shielding.

The manual actions will include

.deenergizing the breakers associated with the R11R torus cooling valves on the

'2D and 2E RMOV boards in the-reactor building and manually opening one of t'ae-two RilR torus cooling valves located in t.he reactor building'(see Enclosure

' 2). - Procedures for the manual actions will be implomonted prior to the restart of, Unit'2.< -In addition, minor modification of the RMOV boards is

' required.:

lThe valve control. circuits on 'RMOV. boards 2D and 2E-are being modified to ensure a RWCU line-break outsido containment would-not result in a failureLof the valve control circuits on the RilR=LpCI injection valves. These valves maintain the high pressure / low pressure interface betwoon= the reac' or coolant system and the RHR system and mot remain closed when the high pressure interlock. signal is present. This modification is not' completed at this.

I time, but is scheduled for completion prior to the restart of Unit 2.

In' addition, the BFN Final Safety Analysis Report:(FSAR) requires revision to (clarify =that RMOV board 2D and:2E valvo operability is not~ required to support <LpC1 operation following a RWCU line' break outside containment. This

. revision will be implemented in the next scheduled update of the BFN FSAR.

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Enclogura 1 i

page 2 of 2' One. final required action is related to the RitR pump mini-flow valves. These

, valves are required to open for. protection of the RHR pumps during low-flow a

condit.lons. Valve operability-cannot be assumed in the event.- of a RWCU line break.- Therefore. procedural controls will be-implemented prior t.o the rest, art of Unit. 2..These procedural controls will provide for operator

, action to ensure that. if the RHR pumps start following' a RWCU line break, they will-not, cont.inut to run.unicas a pump discharge flow path has been established.

Based on the follow-up information presented, TVA concludes t. hat, the 480 Volt.

RMOV boards 2D and 2E are not required for postulated RWCU line breaks and do l

not require qualificaLion per 10 CFR 50.49.

TVA considers this-issue to be closed.

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4 ENCLOSURE 2 4[

Subject:

_ AMENDED RESPONSE TO NUREC 0737 ITEM 11.B.2

Reference:

1.

TVA Letter,. L. M. Mills to 11. R. Denton dated January 17, 1980 2.

NRC SER D. B. Vassallo to H. G. parris dated March 8, 1983 INTRODUCTION h

In respo tse to NUREG' 0737 Item 11.B.2, TVA lot.ter from L. M. Mills to H. R.

Denton dated January 17, 1980 stated that modifications to plant shielding were not necessary at Browns Ferry Nuclear plant (BFN). This position was-based on the TVA' evaluation of vital access areas,:which showed that no post-accident vital area requiring occupancy exists within the reactor building (see Attachment, Reference 1). ' NRC Safety Evaluation Report dated

. s#

March 8,'1983 concurred with this position and concluded that the recommendations'of NUREG 0737 Item 11.B.2 had been met.- llowever, recent

- evaluat.Lons performed during the Environmental Qualificat.lon (EQ) effort have shown the need to clarify t.his posit lon in the event of a postulated Reactor Water Cleanup' (RWCU); system line break outside containment..

BACKGROUND.

e The RWCU system is, required during power operation to maintain reactor water purity.- primary coolant = is piped.by t.be RWCU system t.hrough the primary

= cont.altunent.' and.into the reactor building for processing and then is returned i

to the primary coolant, system.

High Energy.Line Breaks (llELBs) of the RWCU system are detected by 1 environmentally qualified' temperature detection inst.rumentation.

Isolation of-l

.the~ affected piping is performed by primary containment isolat. ion valves c

. located inside.and out. side t.he primary containment.

These valves are seismic-class 17and are required to be-capable of closing within 30 seconds of a detected'RWCU~line break by t.he BFN Technical Specifications'(TSa).- Routine j

.surveillances are performed to verify this closure t.ime.

During _ t.he _ t ime interval required to isolate _ the RWCU. Ilne,' 27,900 lb-m of '

primary coolant would discharge-into the reactor building. >With a total pressure vessel inventory.of 546,300 lb-m,_the loss of 27.900'1b-m is negligible and would'not result in either~ Reactor _ protection System-(RPS)

-act.uations or a turbine trip.

Therefore,,in accordance with. Standard Review 3

Plan 3.6.1,.such an event.is not required to be' analyzed concurrent with a loss of-of fait.e' power. _ TVA analysis. has - shown that this loss would be sensed

- as a plant transient and the feedwater system would respond to restore the-plant. to normal condit.lons.

Therefore, core uncovery and_any subsequent fuel-n failure would not occur.

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page 2 of 4 5

KVALUATION i

Following a RWCU line break outside containment, automatic isolation of RWCU illnes would occur.- In addition, the control room would receive annunciation Q

of the RWCU isolation. Normal p'. ant shutdown would begin when warranted by reactor coolant chemistry. Normal shutdown would allow the core cooling.

requirements to be provided utilizing the condenser.

However, if the i

. condenser was unavailable the core cooling requirements would be provided by i

the mnergency core cooling syst. ems.

The low pressure core cooling requirements can be met by either LpCI (one

loop /one pump) or core spray (one loop /two pumps) in addition to operation of one loop of RHR in the suppression pool cooling mode. Both loops of. core spray are qualified for operation following a RWCU line break outside containment.

Therefore, redundant low pressure core cooling capability is

.providedLby either loop of.the core spray system without operation of RHR in the LpCI mode.

Operation of,one loop of RHR in the suppression pool cooling mode is' required following an RWCU line break outside containment if the reactor is isolated.

Normally, the suppression pool-cooling mode of RHR could be remotely actuated. However, the. motor control centers for actuation of suppression

. pool cooling following a RWCU.line break are not environmentally qualified and must be conservatively postulated to fail.

Therefore, suppression pool

, cooling test be initiated manually when the suppression pool temperature 1

exceeds the 95'F limit specifled in the emergency operating instructions. The et

' manual initiation:of suppression pool cooling requires entry into.the reactor

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. building to stroke'one of the two suppression pool cooling valves.

This manual 1 action'has been~ evaluated with regard to the whole body dose received by the'individua1 performing this-action.

1 INUREG-0737,) paragraph II.B.2(1), states'"The minimum radioactive source term

.should be equivalent to the source terms recommended in Regulatory Guides 1.3, 1.4, 1.7 and' Standard Review. plan 15.6.5'," and. defines the percentage of total core inventory that should be. considered for.nobio gases, halogens, and others.-~Thelsource terms required by.the above mentioned documents are-

.directly related to Loss of. Coolant Accidents'(LOCAs) and do not apply in the

event 'of. a HELB, such as a-RWCU line break outside primary containment.

NUREG-0737, paragraph 11.B.2(4),~ defines non-LOCA HELD source terms..However, this; paragraph-is specific.in stating that these values are to be used to

~ determine the EQ of safety-relatef equipment. -Therefore, the stated source-terms do not necessarily apply to the determination of personnel exposure.

NUREG-0737,: paragraph:11.B.2(3),stefines the' total allowable dose to be used When evaluating personnel exposur e.

The critoria states "... the dose to

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.personnelLshould notibe in excess of S rem Whole-body, or its equivalent to any part of the body for'the durat ion of' the accident." This paragraph allows theldctermination of dose rates.-to be made on a case-by-case basis.

Specifically, NUREG-0737 states "Therefore, allowable dose = rates will be based

'upon expected occupancy,-as well as the radioactive source terms and shielding." However, ein order to provide a general design objective, we.are providing the following dose rate criteria with alternatives to be documented l

on a case-by-case basis.

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l Enciczura 2 1

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l previous guidance was provided in NUREG-0578, TMI-2 Lessons Learned Task Force i

Status Report -and Short Term Recommendations.

NUREG-0578, Section 2.1.6.b.

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Item 2'(Discussion), states that "af ter an accident in which significant core l

damage occurs, the radiation source terms may approximate those of Regulatory j

Guides 1.3 and 1.4"

. NUREC-0578, Section 2.1.6.b, item 3, (position), states that each licensee 15 should perform a radiation and shielding design review with the assumption of L

a post-accident release of radioactivity equivalent to that describe' in Regulatory Guides 1.3 and 1.4.

The design review should identify locttlon of

' vital areas and equipment in Which personnel occupancy may be unduly 12mited.

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LAdditional reviews'of NUREG-0588, Interim Staff position on Environmental Qualification of Safety-Related Electrical Equipment,. and 10 CFR 50.49 were

' performed to determine if specific saidance was provided for personnel entry into vital areas. These regulatory-documents were specific to equipment

-qualification and did not provide specific guidance relevant to personnel

~ entry into t1u) reactor building following a RWCU line break.

f TVA has~ completed the design review for BF1.

The original position concluded that entry into the reactor building was not required to mitigate the

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consequences of an accident.

Implicit in NUREG-0578 and in NUREG-0737 is the

-condition'that~the source term reflect the potential core damage for the L

events. The initial position taken by TVA-(Reference 1), in Which no i

post-accident' vital areas requiring occupancy exists within the reactor building, remains unchanged for, accidents normally assumed to result in, core damage.

i TVA analysis shows that for a RWCU line break outside containment fuel dr.nage I

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s" will not occur and therefore the actual source term will be dependent or. the

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expected normal reactor coolant activity..not total core inventory.' Of the 27,900 lb-m of reactor coolant released,' icas than-40 percent flashea to steam

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and becomes airborne.

This= normal reactor coolant ' source term for-steam was

,obtained in'accordance with. ANSI /ANS 18.1, 1984.. This source resulted'in a Ewhole-bodyj dose-of less than one millirem to the individual, making the

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ci 1 required entry.~ However, the' calculation of-the total'whole body dose to the individual performing the mission also took into account the dose received' i

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from contained sources (CRUD traps). The highest dose l rate froa. contained

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L aources was 270 mrem / hour. Therefore..since the mission'will;lastabout one i

hour the individual performing the mission will not receive whole body dose i

greater,than about 270 meem. Utilization of this maximum doseirate from-j contained'uources meets the: intent of providing!a dose rate map.

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' CONCLUSION y

(Based on'the reviews of NUREG-0737, NUREG-0578, NUREG-0588, Land 10 CFR 50.49, TVA: considers.NUREG-0737, Item II.B.2. to remain closed, However, due to'the.

i' required entry into'the i "or building,- the -previous response (Reference 1) to NUREG-0737, item 11.B.z.,= requires clarification.

NUREG.0737', paragraph 11.B.2(3) allows personnel entry into vital areas to be addressed on.a case-by-case basis. ~1n'the case of a SWCU line break, personnel entry into the reactor building to initiate suppression pool cooling will not result in a i

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_, :s Enclorura 2 Page 4 of 4 personnel-dose greater t.han 270 mrom whole body when evaluated wit.h an

' ANSI /ANS 18.1 source t,orm for normal reactor coolant. (see below). This li L personnel dose is' wit.hin the guidance provided in NUREG-0737, item II.B.2.

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l-NORMAL REACTOR COOLANT l

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SOURCE TERMS:

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TVA concludes _ that the analysis _ allowing ent.ry into the reactor building to d

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- init.iate suppression pool cooling is consistent. wit.h the objectives of j

1 HUREG-0737, Item II.B.2.

However, TVA will complete any. required E

modificat. ions during _ t.he Cycle 6, refueling out, age - to ensure ent.cy into the j

_ reactor building following an RWCU line break will no longer be required, i>

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- AT N'Uf1 MENT TO ENCLOSURE 2 c,

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~400 Cheetenst Street Tower 11 January 17, 1980 4

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Mr. Barold R. Decasti..Direetxe offime of mesh Rometer Regalatien U.S. maa$mdfr Regulatory e==4maton vae g tes, DC-20SSS Deer Mr. Destent=

!a the Matter of the

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Docket Noe.30-139 Teasessee Valley Authority

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S0-260

$0-296 Realased;1e our resposee to your October. 30, 1979,. letter ta all operating resetore la vidah you rwussted doeismentation of the motheda 2

seed for implesammestion of the NUREG-0578 short-term requiremente.

3 Also,1 in response to your. Jasmaary 2. - 1980, Confirmatory order letter

. te E. C. Parris regarding Browns Ferry unit 1, unit 1:ia pressetly -

7 la eoid elastdown for a refueltag outage and will be in full eos,11amme

.with the numap 0578 short-tarei roquirements before restart-as indiaated a

is my Hovember-16,:1979 1etter.to you.

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Very.truly vours.

TERER8835 VA*AIT ADTRORITT i

C3-1..-M. Milla,-Manager O

Nuclear Reguistion _and Saf ety

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i ARMS. 640 CSJ2-1.,.

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n. Culver, 249A HBB-K E.'A. Belvin, ROB-H J.'R. Calhoun,.716-F.B-C

'A. W. Crevasse, 401 UBB-C

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'O.-F.-Dilworth, W10C126-C-K H. S. Sanger. Jr., E11833 C-K

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R. F. Sullivan, Browns Ferry -. Resident Inspector F.- A. 8sesapanski,. 417 UBR-C

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...;,. s 2.1.1 DEAGENCY POWEA SUPPL.T

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TVA Response e4 '

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i Pressuriser Bester Power Supply l

This position is not directly applicable to Browns Perry; bolllog water reactors de not use a pressuriser system to establish and esistals primary

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systes pressure.

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Natural circulation in the 8WR, as discussed in NEDO-24708, is strong and i.

. inherent in all eff-normal modes of operation as long as sufficient vessel laventory is maintained, and natural circulation is.todependent 6f any.

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powered system.

This is because, even in norsel operation - the BWR is ;

j essentially an augmented estural circulation eachine. Because - the BW operates. in.all modes with both = liquid and stese in the reactor pressure

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vessel, saturation conditions are always : enintained irrespective of systes pressure.

Thus, there is no need for emergency power to esintsia natural circulatica or to keep the system pressurized.

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. Power Supply for' Pressurizer Relief and

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'ock Valves and Pressurizer 3'.

All' esin: stese' line' relief valvec that. are paa of the automatic

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depressurtration' system (ADS) are steplied with sai'ty grade, diverse contro! power. !fotive air La supplies' to these velves-by redundant ' air

compressors,. which

are energized by diverse safety grade power sources.

The., ADS valves also ' have accumulators, which. allow:-sultiple remote o

operations even in the absence of control air. free the compressors.

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BWR's do not hav'e block valves.

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.The reactor vessel level indication instrument channels.for safety

-system activation and cor.crol-are M

. power.

already -powered by emergency.

-O' For the reasons stated,above, there is no need for ' action in respoose to

<J,o z JfRC position 2.' t. l.-

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2.1.2 PERTORMANCE TESTINC POR" ret 1EI' AND SAFETY VALVE 5 j

1 TVA Respones j

The conditions under which BWR relief and saf ety valves would be empacted

-to esperience single-phase liquid or two-phase flow are divided into two

categories t: (1) high-pressure conditions, and (2) low-pressure conditions.

The high-pressure conditions would result from s failure to shut off the high-pressure ECC$' systems and the feedwster system prior to putting water into the main steen lines. The low-pressure conditions would result from

' intentionally estabitshing alternste shutdown cooling through the relief valves'with makeup.from the LPCI mode of RHR, or from the insdvertent 1

' overfilling of the vessel from the low-pressure ECCS.

' As a result of flRC's llovember 14, 1979, response to the B'.'R Owners Group generic position, the Owners Group is preparing a test program to sddress those.

i conditions which could tesult in single-phase liquid or two-phase flow throur,h the, relief and safety valves at low-pressure conditions. The I

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high-pressore conditions alli'be addressed by either high-pressure

' te sting under two-phans flow conditions or by installation of high j

0; lteliability, high level-trips for hir,h-pressure ECCS"and reactor feedvater O'

This trip-will proclude-the conditions of two-phase flow through i

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t'te relief and safety-valver at high pressures. tve believe tnat tuiw W'

,froposition~will satisfy tie intent of tiRC's November 14, 1979, comument s.

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'ly January 31, 1980,- the JetR Owners Cruop will present their program f or the high level tripe and for testing relief and safety velves under the low-preaaurn conditions. This presentation will-include the neope of testing O

and analysis to be perioread, and a schedule 'or completion of testing and: analysis.: Wo coussit to accept the negotiated MRC/0wners Group resolution of this item.- gasolution of this item will be discussed in the uoconing seatingi(Januaty 31, 1980) with NRC.

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4i 2.1.3.a DIMCT INDICATION of M!,IET AND SATETT. VALVE Po8! TION TVA Response i

TVA is providing saaebiguous pain control room indication of valve posittee by use of an acoustic nonttoring systee on the relief and safety valves.

An alors is provided in the control room in conjuaction with the system.

The new valve position indication is single traia, using a class it power supply.-' As a backup individus1 valve positions will be deter 1ained using the existing temperature sensor located in the discharge piping downstream of each valve. The use of the primary and backup valve position indicators is discussed in the appropriate emergency procedures.

The acoustic monitoring system is qualified as. seismic class I, and-is environmentally qualif ted for the appropriate temperature and pressure.

Qualification for high radiation fields is currently being conducted.

The, required modifications have been completed -for Browns Terry unit 2

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and 3. and _will be= completed for Browns Ferry unit 1 during the refueling 3.g outage scheduled to begin in early January.

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.2.1.3.b: INSTRLMDf7Af!ON FOR INADEQUATE CORE COOLING

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TVA Response-CE bas examined this requirement on a generic basis for the BWR Operating

0wners Croup.. The conclusions of this study are documented in NEDO 24708, which - bas been recently. submitted to FRC. A need for addittomat instrumentation has not baen identifled.

Developeent of improved procedures _ to be used by the operator to recognize inadequate core - cooling with currently available instrumentation is discussed in the TVA response to iters 2.1.9.

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2.1.4 CONTAINMENT ISOLATION TTA Pnettion The containment isolation system at Browns Ferry is designed to prevent the

-release of radioactive material to the envircnment after an accident while ensuring that systene important for post-accident mitigation are operational.

Systems were. evaluated and containment isolation provisions provided based

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.on,the followingi I

l 1.

Non-essential Systems - These systems are not required for post-accident mitigation and are isolated automatically upon receipt of a primary containment L

isolation signal (PCIS) or are provided with manual valves which are locked closed when containment integrity.is required.

3 2.

Essential Systems - These systems are required for post-accident mitigation and are -not isolated automatically upon receipt of a PCIS. However isolation N

of.these-lines, if required, is possible from the main control room. The i

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following systems are classified'as essential, as-a result of their accident

.i mitigation.tunctions c3 (1)' Standby 1,1 quid Control- (St.C)

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.(2): Reactor Core teolation Cooling (RCIC) t 4-

'(3) 'llish Pressure Coolant injection (HPCI)

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- (4) - Residual Heat Removal - 1,ow Pressure Injsetion sad Containment y.

Cooling Modes (RHR) 1 6

'(5), Core Spray (CS) j Or c(6): Containment Atmospheric' Dilution (CAD) 3

- i Each-line penetrating primary containment hag been reviewed to ensure that (1)

'isointion of the line was based on its need to be:in service post-accident and

~(2);that each containment-~1 solation valve received the proper isolation signal.

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The BrownsErerry PCIS are' provided by diverse and redundant saf ety grade equipment.

Browns. Ferry complies with SRP 6.2.4 by isolating on.Lin' general.-(a) low reactor level or (b)-high dryvell pressure.

The PCIS setpoints were chosen such that

' ~ isolation will: occur ' prior to or at the tine of ECCS. initistion. There are.

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geveral other isolation modes in addition to the main PCIS logici For example, sain steam 11ne iso,. ion valves -will also close as a result of high steam Line i

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-2.1.6.b Plant Shieldina Review

-TVA Response'

' TVA plants are specifically designed to mitigate major design basis events

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Iwith no access outside the NCR being required. With this goal in mind, the plants were not specifically designed for any access autside the main

.eostrol room.-

j TVA has performed a'. shielding review for Browns Ferry Nuclear Plant for i

vital' access of the plant where access may be desirable af ter an accident.

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These vital areas include the spreading room, computer roce, auxiliary l instrument rocas, DC equipment rooms, bettery rooms, battery board roome end the conmounications room in the control building, the shutdown board rooms in the reactor building, and the diesel generator building. Shielding for normal operation in these areas is adequate for required access after an-accident..

l Access to other areas 'of-the' reactor building on elevation 593.0' and below

. is. severely limited because of exposed' RHR system and Core Spray system L

1'piping. Ilowever, no vital access acess' have been identified in these areas.

Access to' areas of the reactor building on elevation 621.25' and _above would I

y not:be limited by contained sources: however, access may be limited by l

airborne: activity for Large accident cases.

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An the current plant design allows ~11mited access to vital areas dincussed -

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-above, no plant modifications are foreseen at thin time.

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2.1.7.a ALTTO INITIATION OF AUXILIARY FEEDWATT.R TVA Response Although Browns Terry does ' ot have an auxilia ry feedveter systes, n

portions of the Brovus Ferry ECCS network perform coeparable functions.

These ECCS systees are safety grade and meet the intent of the NRC's positions and clarifications.

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r-2.1.7.b AUXILIARY TTIDWATT.R FLOW INDICATION TVA Response As noted in the response to 2.1,7.a portions of the Browns Ferry ECC3 network s e rve similar functions for post-secident recover 7 as the auxiliary feedveter system for FVR's.

The flow indications for these systems are safety grade and meet the intent o! NRC's positions and clarifications.

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1 2.1.8.a-Improved Post-Accident samptly capability TVA Response l.

The design and operational review of the reactor coolant '.nd containment "steospheric sampling systems and analysis facilities ha e been completed.

Due'to the location of1 the current samplinet systems. access is !!aited for

' 'large accident cases. Using the results of this review, engineering analysis-of post-accident sempling facility.needs have begun. The routine sampling f ac ilities, representative of-TVA plants, have been inspected to f amiliarire la pstential-TVA consultant with the existing sampling capabtlities. Two

.pos sible contracting. options for assistance -needed by TVA to provide a post-ace, dent sampling capability have been developed and analysed fe effectiveness.

~A. decision vill be made shortly on the option that best. meets TVA ne.'a.

Work

'wil' then proceed to locate the new sample station and procure the facilisfes an i equipment needed for this post-accident sampling capability. At that ties a-. letailed schedule for implementation will be available, l

17-1 Until the-design _ modifications are complete, procedures have been devised to J

evaluate the~ primary coolant systes activity depending on the accessibility of the ~ sampling stations.for particular degraded conditions, 19 '.

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2.1.8.b increased Range of Rsdiation Monitors TVA Fosition TVA will provide redundant safety inh-range noble ass effluent monitors.

A method or methods of sampling effluent particulates and todine will be chosen and redundant particulate and lodine effluent monitoring systems qualified to the present state-of-the-art will be implemented..

L TVA will provide redundant safety r,rade radiation monitors to meet MRC's high-range monitor requirements. --These monitors will be isolated f rom the containment staosphere yet not located outside the containment shielding.

Exposure rates at selected detector location from netivity within the

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containment will be high enough such that instrument readings can be correlated with containment activity throughout the course of the accident.

~ 4 In the interim, procedures have been developed an practical to estimate release rates if existing effluent instrumentation goes off scale, J

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IMPROVED INPLANT 10 DINE INSTRtJMENTATION TVA R'sponse e

Browns Terry has portable lowyolume air samplers, each equipped with a particulate filter followed by a charcoal adsorber. to collect iodine-isotopes. -The particulate filter will be counted in the health physics.

a laboratory for gross activity and-the charcoal adsorber sent to the radiochemical laboratory for a gasma isotopic analysis for radioactive lodines. If necessary as necessitated by a high-gross activity, the particu: ate filter will also be sent to the radiochemical laboratory for an isotopic analysis.- The primary difference in obtaining inplant airborne isotopic concentrations for accident and routine operating conditions is-the time required for sampling. A shorter sample time could be necessary for accident conditions because of the presence of high

-isotopic ' concentrations.

4 The plant has procedures for sampling and analysis of inplant air spaces I

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-incorporated in the Health Physics 1,aboratory Instruction Manual and the

- Radiation Control Instruction Manual.

Plant : bealth physics technicians are required in complete a formal.

O training program plus-receive inplant training which includes the-use of bestth physics procedurps and instrumentation.

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l 2.1.9 Transient and A;cident Analyses j

i TVA' Response s

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Small Break Loss-of-Coolant Accidents The small break analyses have been completed as a revision to NEDO-24708. : This document has been sent to NRC staf f.

The con-n siderattons sxpounded in the subject HEDO have been incorporated into plant emergency operating procedures.

2.

Inadequate Core Coolina '

The analyses for-inadequate core cooling, procedures development.

j and operator training, will be completed according to the schedule i

= agreed upon by the BWR Owners Group and the B&OTF. Also, the reactor water level instrumentation in the main control has been i

distinctively: coded to _ alert the operator to which are-reliable

'under LOCA conditions as recommended by General Electric's guidelines a

presented in SIL-299.

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-Transients and' Accidents-tt-The transients and,' accidents analyses, procedures development, and i

. operator training vill be completed according to the schedule agreed '

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upon byLthe BWR Owners Group and the B&OTF.

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2.1.9 CONTAINtiDIT PRESSURE INDICATION TVA Response TVA vill provide new instrumentation capable of meeting Regulatory Guide 1.97 (revision 1), requirements on this position. This system will i

be divisionalized with division I sad II, each having e -5 pois to +5 pois and e 0-300 peig differential pressure tressaitter.

The division !

- treassitters will interface with a control room pressure recorder. The division 11 treassitters -v111 interface with a control room (edicator.

Both divisions will be powered by divisionalised emergency onsite power.

We espect to have this modification completed by January 1,1981.

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' 2;1.9-Containment Hydromen Monitor

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' TVA Response

. A Hayes-Republic hydrogen concentration monitoring system will be installed.

A continuous. indication of hydrogen concentration in the containment atmosphere ehall'be provided in the control room.

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. Heasurement capability shall be provided over the range of 0-20 percent

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i hydrogen concentration through a containment pressure range of 12 pois to maximum containment design pressure (56 peig).

- The.. hydrogen _ concentration monitoring syntes vill meet the design and qualification provisions'of_ Reg Guide 1.97.

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TVA Response TVA will provide new instrumentation capable of meeting Regulatory Guide.l.97.(revision 1), requirements on this posittee. The system wil.'

be divisionalised with division I and !! each having a range of at least the lovest ECCS suction point to.five feet above normal water level. The division I transmitter will interf ace with a control room level recorder, and division II with a level indicator. Both dtvisions will be powered by j

divisionalised emergency onsite power.

This_ additional system will require two torus penetrations per division (four total), ubich presently do not exist. We expect to have this modification coepleted by January 1, 1981.

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TVA Response TVA concurs in the consensus of the SWR Owners Group that the Browns Terry estetins. reactor venting capability is fully satisfactory.

The justification for our position was presented to NRC during the October 11. -

p 1979. topical meeting and is also contained to the !Mt Owners Group

.i submittal-in response to NUREG 0578. The existing greves Ferry reactor iventing systes is represented accurately in the Bb1t Owners Group submittat.

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2.2.1.a SMITT SUPER 7180R'S RESPONSIBILIT!f.S TVA Response 1.

'TVA's-administrative procedures, shif t supervisor job description,

-and training progrees strongly emphasise the primary saaagement responsibility of the shift supervisor.

In addition, periodic retraining acts to reinforce his consand reopensibilities. While these estating possures provide a high level of eenfidence that the i

shift supervisor has primary sensemat responsibility for safe operettoa of the plant, TVA will annually issue a essagement 1

directive which emphasises this assignment of responsibility.

2a.. Pisat administrative procedures have been reviewed to ensure that they clearly define the authority and respoestbilities of each

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position on shift. The duties and responsibilities of the shif t i

supervisor, as specified la the job description, are consistent with 0

_ position stateneat 2a.

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2 b.' The shift crew 'ta TVA plants consists of the following: (1) a shift

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. engineer who has a SRO license and who has overall responsibility for

~j the plant when higher level "ta-Line" sanageeset personnel are not 3

present, (2) as assistaat shif t engineer (also has a Sao license) for 9

sach unit who has supervisory responsibility for all normal, abnormal, and emergesey activities on his assigned unit, (3) a unit operator (with as R0 license) for each unit, sad (4) other personnel as appropriate. The. duties-of the shif t supervisor as, discussed in '

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NURIG-0578 'are performed by = the assistant shift engineer.on'each

. unit. For purposes of our response, we will' use-the ters assistant

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shif t engineer for shift supervisor.

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s The assistant shif t engineer's normal work station is in the control-7 room, but be periodically aiskes inspections of plant. equipment. He-

'will-immediately go to the control room during emergency sif.uations.

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Re remains = in the: control room at all times during accident U

situations, to direct the activities of the unit operator unless formally relieved of this function by -the shif t engineer. The shift engineer.- eey, in _ turn. - be formally relieved by the as sistant operations supervisor or the operations supervisor (both also hold

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an SRO license).'

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In the event that the assistant shif t engineer (shif t supervisor) is absent, the unit operator will be_ the lead operator en the unit to

>which he is assigned.: An addit.cosi licensed ' operator will be available in the control complex to act as an assistant to the unit coerator in abnormal or energency situations.

3.

The shif t engineer and assistant shif t engineers have received such i

trainin3 4.

TI+ adminf etrative duties of the shift supe rvisor will be ptriodies11) reviewed by the senior of ficer of TVA responsible for plant operat ions. Administrative functions that detract free or are i

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d subordieste to ensuring safe operation of the pleet have been -

assigned to ether employees. A clerk has been assigned to the shift

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engineer's of fice os each shif t to. perform adatsastrative details formerly done by the shif t. engineer and assistant shift engineers.-

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._2.2.1.b SHITT TECWi! CAL ADVISOR TVA'is using esperienced staff nuclear engineers to meet this requirement.

These eagineers will be os skif,t continuously at the plast site.

- The Shift Technical Advisors have college degrees is engineering, and have received additional simulator _ training concerning soreal and off-rersal operations in order to enhance the. accident and operating essesseest function at the plast..- This person will be assigned other dettes when his duties as shift Technical Advisor, provided that his avettability is not compromised.

TVA believes that a multi disciplined review group is necessary to adequately investigste !.ER's.

TVA's Nuclear taperieoce Review' Panel presentlyLreviews all licensee event reports. When applicable, results of the review are incorporated in TVA's Operator and Shif t Technical Advisor training - and ' requalification programs.

In addition, periodic training sessions are conducted for each shif t crew. The mater 1_al covered.during 1

'in these - sessions - include, but= is not limited to,11censee event reports.

operator - errors, recent equipment problems, changes to-technical

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- specifications, and generet plant, status. The Shift Technical Advisors

'have additional-responsibilities in being cognizant of the results of this

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enented shif t and relief turnover precedures that provide

..t,3 9 assurance that the oneselag shift possesses adequate knowledge of.

" v.g r.t critical plant states Laforsetten and system availability. A checklist er 7

similar hard copy is completed by offgoing and oncoming shifts at esek shift turnover.

' This checklist includes critical plaat parseeters and allowable limits, "N

availability and proper stigament of s tfety systems, and a listing of

~7 esfety systee components in a degraded mode aloeg with the length of time

'q is that eede. -This checklist is signed by the offsetog unit operator sad

-e-4 the oneoeing assistaat shif t supervisor and unit operator. All shift

,A personnel respeasible for the status of critical equipeest have relief checklists for oncoming and of f going shif ts that include any core cooling equipment under esintenant: er test that would degrade a safety systee.

In addition, a systee will be estabitsbed to evaluate the effectiveness of the turnover procedures.

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TTA h.e estabitsbed plast odotaistrative precedures to teatret necess to

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. ' procedure else addressee identificattee of authertse ' perseasel.

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Plant adelaistrative precedures ' ave anse been lessed defletes the s

' respoestb(11ttes under both sermal sed accident 'ceedittees.

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2,2.2.b onette Technical Support Center _

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The enette Technical Support Center (TSC) has been estabilehed in the i

Shif t Engineer's Of fice end the relay room in the powerhouse contret bay.

i The plane for manning the TSC are prescribed in the Browns Ferry Radio-togical toersoney Flas.

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4 Phones have been provided in the Shif t Engineer's Of fice to consweicate

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j the Chattanooga Division georgency Ce:.i;r.

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plete set of functionet plant drawings and necesse:7 technical information.

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This center is located in the control bay, and therefore, meets the same j

habitability requirements of the main control room.

Is TVA is planning to upgrade the TSC by January 1, 1981.

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radiation, high steen flow, or high steam line tunnet teetierstute. the primary containment ventilation systen isolates on reactor building high

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radiation. The MPC1 and RCIC systese have instrumentation to detect pipe breaks within their own flow patho g and to subsequently teotste the syntes.

TVA te modif ying the isolation logic on 3$ valves on each unit to meet the require.+nt e nf it en (4). The design to such that resetting the sain primary containment isolation signets will not result in the automatic reopening of these isolation valves. The logic modification has been completed on unite 2 and 3, and will be completed for unit i during upcoming refueling outage

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in early January, 1980.

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2.1.S.a DLD!CATED ETDR0 GEN C0KTR01, PENETRATIONS j

TVA Response post secident. bydrogen control for Browns Ferry is provided by inerting j.

the primary containeemt durteg normal operation. Af ter an accident, leet.

ters combustible gas concentrations are maintained by the containment i

atmospheric dilution (CAD) systes. This systes is designed to purge small l

quantitles of the containment atmosphere to the standby gas treateest i

system while adding makeup nitrogen to the containeest. The CAD systee seets NRC regulations on redundancy and single f ailure criteria.

i During CAD systee operation, containment steosphere is vented in a controlled manner to the standby gas treatment system through a two inch line. Tallure of one vent path will not disable the CAD system, since a redundant vent line is available.

For the reason stated absve, there is no need for action 1: response to NRC position 2.1.5.a.

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2.1.3.c INSTALLING NYDROCEN RICOMBINERS IN 1.k1t's TVA Response Tbio requirement is set applicable to Browns Terry. The trowes ferry dest'en utillies inerted contat' ament design for combustible ses centret.

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2.1.6 a Systems Integrity for High Radioactivity TVA Resposee TVA has impleeented e leak detection and reduction progree for syatene outside primary containment which could contain highly radioactiv fluide

'during a serious accident or transient.

F.ach designated syeten vill be inspected and will include sessurement of actual leslage with the system operating when possible.

Identified leakage pathe v111 be repelred in order to maintain as-low-ee practical levels.

Also, as requested in NRC's letter of October 17, 1979, to Operating plants, (North Anna probles) TVA has investigated Browns retty syntes, to ensure similar release paths se described do not exist. Although no pathways were' identified, the review recorumended some minor changes to

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leprove our radweste handling procedures.

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o ENCt0SURE 3 I

Subject:

COMMITMENTS REQUIRED TO ADDRESS Tile TVA PRESENTATION OF HAY 30, 1990 Three types of commltments were made in Enclosure 1 of this letter.

The first was related to procedural controls, the second involved immediate plant modificationn and the third involved plant modifications during the Cycle 6 refueling outage.

A list of these commltments is provided below.

First, procedural controls will be implemented for the manual actions required to open the RitR torus cooling valve, to deenergize the associated breakers, and to ensure an Rl(R pump discharge flowpath has been established. The E

procedural controls will be implemented prior to the restart of Unit 2.

Second, the valve control circuits on RHOV boards 2D and 2E will be modified to ensure a reactor water cicanup (RWCU) line break outside contaltunent would not result in a failure of the RilR low-pressure coolant injection (1.pCI) valve control circuits prior to the restart of Unit 2.

Third, TVA will complete any required plant modifications during the Unit 2 Cycle 6 refueling outage at UFN.

The modifications will ensure that encry into the reactor building following a reactor water cleanup linebreak will not be required following Unit 2 Cycle 5 operation.

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