ML19276H547

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Forwards Response to 790713 Request for Info Required for Review of Implications of TMI incident.Twenty-four Oversized Piping & Instrumentation Drawings Encl for Central Files Only
ML19276H547
Person / Time
Site: Browns Ferry  
Issue date: 11/27/1979
From: Mills L
TENNESSEE VALLEY AUTHORITY
To: Thomas C
NRC - TMI-2 BULLETINS & ORDERS TASK FORCE
References
NUDOCS 7911290392
Download: ML19276H547 (55)
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Text

{{#Wiki_filter:% ~~' ,. ~ - ~7 400 Chestnut Street Tower II November 27, 1979 Mr. Cecil 0. Thomas Bulletins & Orders Task Force U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Thomas:

In the Matter of the ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 In T. A. Ippolito's July 13, 1979, letter to H. G. Parris the Staff requested information required for NRC's review of the implications of the TMI-2 incident regarding operating and near term OL BWR's. On July 12, 1979, representatives of the BWR Operating Plant Owners' Group and the NRC's Bulletins & Orders Tack Force (B&OTF) met to discuss the request before it was formally issued. This meeting resulted in a deferral of the responses to several plant unique Systems Group questions. The enclosure to this letter provides our response to these deferred questions for Browns Ferry Nuclear Plant. Very truly yours, TENNESSEE VALLEY AUTHORITY q L. M. Mills, Manager Nuclear Regulation and Safety Enclosure 0 } \\ 7 9112 9 0s $t 1421 173

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PLANT S c o m s b <w UNIT (S)l 1 s% i PLANT-SPEC FIC SYSTEH INFORMATION p(_ f' General Water ~ Sources Instrumentation and Control Frequency of Safety Seismic Safety Seismic Safety Seismic System and System Classification Category Classification Category Classif. Category Component Tests 1. RCIC No I 5 1 5 I O '" O # 2. Isolation C'ondenser Oo<s %A .t e 01 w. 3. IIPCS Does ~o 4 [o n.11 4. IIPCI i 1 5 1 1 6 %.s <. k e d 5. LPCS ( t o e, sqc-y 5 1 5 1. s i % % a c. b e d 6. LPCI 5 1 5 d n W.a k e t} 7. ADS C 1 DH A DNA 5 i t W mob e J 8. SRV ko i bNIN O N l\\ No 1.L. 6Ostbed 9. RilR (including shutdown cooling, bl o i N* DNA N* g 'A suppressionpool]cooli steam condensing g o es u o -N uppjm 6-containment spray, mode 5 1 5 1 5 i O "tb <d -10. SSil 5 1 5 1 5 1- % Wa ok <J 11. RBCCil Nc i 5 i Ne i achc L .,12, CRDS - 5 1. ONn D d th 5 1 wo4e 3

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IC IC C/M High r-acter pressre functienel tect 4.1.A -6 4.1.A-6 IC IC O/M High dryvell pressure f.tnetional test 4.1.A-7 4.1.A 7 !C IC O/M Reacter icv vater level functienal test ~~ (LIS-3-203 A-D, Sv. #2-3) h.1.A-8. - 4.1.A IC IC O/31 High water level in scra= dischar6e tank functional test h.1.A-9 4.1.A-9 IC IC O/M Turbine cendenser icw vacuu=~ functional test 4.1.A-10 4.1.A-10 ' CP' T IC.'. ON Main stec.n line high readiatics funcf.icnal tent 4.1.A-11.,. 4.1.A-11. CP 'IC O/M Main steam lica isolatica valte cicsure ..:. s. :. .p,:- .7., 7.,, functional test.. g, .j,

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p A L-f. ? TECH SPEC RELATED TEST REOUIREMENTS _ SYSTEM FREf'UENCY COMPONENT TESTS 1. RCIC !.onthly SI 4.2.B RCIC Turbine steam line high flow (FT) SI 4.2.B RCIC Steam line space high temp. (FT) SI 4.5.F (See 1.b) - RCIC Pump operability (Reactor Steam) SI 4.5.F (Sec 1.c) - RCIC Mov. Operability SI 4.5.F (Sec 1.d & e) - RCIC Flev Test Quarterly SI 4.2.B RCIC Turbine steam line high flow cal. SI 4.2.B RCIC Steam line space high temp. cal. SI 3.2.7 - Testing of HPCI & RCIC vacuum relief check valves (Units 2 & 3) SI 3.2.8 - Testing of RPCI & RCIC vacuum relief check valves (Unit 1) temi-Annual SI 4.2.B-40 -- RCIC Logic Functional ~ 0/0C SI 4.5.F (Sec 1.a) - RCIC Si=ulated automatic actuation SI 4.5.F (Sec 1.d & e) - RCIC Flow rate at 150 psig SI 4.2.3-40 (Sec 4.5) - RCIC Timer Calibration 0/10 years SI 3.3.10 - ASME Section II Hydrostatic testing of RCIC Cold Shutdown SI 3.2.2 - RCIC Mov. Operability SI 3.2.2 - RCIC Testable check valve test Post Maintenance SI 4.5.F (Sec 2.d & e) - RCIC Flow using aux., stea= 2. SYSTEM FREOUENCY COMPONENT TESTS RPCI Monthly SI 4.2.3 Condensate storage tank low level functional \\42\\ \\80

}- 9 A U c. 6 TECH SPEC RELATED TEST REOUIREMENTS 2. SYSTEM FREOUENCY COMPONENT TESTS j02CI IHonthly SI 4.2.B Suppression chamber high level functional SI 4.2.B HPCI turbine steam line high ficw ft. SI 4.2.B HPCI Steam line space high temp. functional SI 4.5.E (Sec 1.b) - HPCI pump operability (Reactor Steam) SI 4.5.E...e 1.c) - HPCI Mov. Operability SI 4.5.E (Sec 1.d & e) - HPCI Flow Rate Quarterly SI 4.2.3 Condensate storage tank low level calibration SI 4.2.B Supp. chamber high level calibration SI 4.2.B HPCI Turbine steam line high fiaa calibration SI 4.2.B HPCI Steam line space high temp. cal. SI 3.2.7 - Testing of HPCI & RCIC vacuum relief check valves (Unit 2 & 3) SI 3.2.8 - Testing of HPCI & RCIC vacuum relief check valves (Unit 1) Semi-Annual SI 4.2.B HPCI Simulated auccmatic actuation 0/0C SI 4.5.E (Sec 1.a) - HPCI Simulated automatic actuation SI 4.2.B-42 (Sec 4.5) - Timer Calibratien 0/10 years SI 3.3.9 - ASME Section XI Hydrostatic testing of HPCI Cold Shutdcwn SI 3.2.2 - HPCI Mov. Operability SI 3.2.2 - HPCI Testable check valve test Post Maintenance SI 4.5.E (Sec 2d & e) - HPCI Flew using aux. steam 0/0C SI 4.5.E (Sec 1.d & e) - HPCI Flow rate at 150 psig )k2

p A (r c. cg 3. SYSTEM FREOUENCY COMPONENT TESTS LPCI Semi-Annual ST 4.2.B LPCI Auto sequencing ti=er normal Power functional SI 4.2.B LPCI Logic functional SI 4.2.3 Logic (Contain=ent Spray) logic functional SI 4.2.3 LPCI (Break Detection) logic functional SI 4.2.B LPCI Auto initiation inhibit functional 0/0C SI 4.2.3-45 (Sec 4.6) - LPCI Auto sequencing times diesel power calibration SI 4.2.B-45 (Sec 4.7) - LPCI Break detection timer cal. SI 4.2.B-45 (Sec 4.6) - LPCI Logic ti=er calibration (Sec 4.7) - LPCI (Contain=ent Spray) Logic ti=er calibration SI 4.5.3 (See 1.a) - LPCI Staulated automatic actuation 4. SYSTEM FREOUENCY COMPONENT TESTS ADS Monthly SI 4.2.B Reactor low water level functional SI 4.2.B Reactor, low water level functional SI 4.2.3 Drywell high pressure functional Quarterly SI 4.2.3 Reactor low water level calibration SI 4.2.3 Reactor low water level calibration SI 4.2.B Drywell high pressure calibration Semi-Annual SI 4.2.B ADS Logic functional 0/0C SI 4.2.B-44 (Sec 4.4) - ADS Logic timer functional SI 4.2.B-44 (Sec 4.4) - ADS Ti=er calibration 5. SYSTEM FREQUENCY COMPONENT TESTS SRV Continuously SI 4.6.D Monitor integrity of the relief / safety valve bellows. 0/0C SI 4.6.D Bench test one safety valve and one half of the relief valves. SI 4.6.D Manual open each relief valve until thermocouples downstream indicate steam flow. SI 4.6.D At least one relief valve shall da disassembled and inspected. 1421 182'

p A L-c. \\0 6. SYSTEM FREOUENCY COMPONENT TESTS RHR Daily SI RHR Loop A discharge pressure instru=ent check (PI-74-51) SI RHR Loop B discharge pressure instrument check (PI-74-65) Monthly SI 4.2.3 RHR Pu=p discharge pressure functional SI 4.5.B (See 1.b) - RER Pump operability SI 4.5.B (Sec 1.c) - RHR Mov. operability SI 4.5.B (Sec 1.d) - RHR Pump flow test SI 4.2.B RHR Area cooler ther=cstat functional Quarterly SI 4.2.B RHR Pu=p discharge pressure calibration Semi-Annual SI 4.2.B RHR Loop A discharge pressure calibration SI 4.2.B RHR Loop B discharge pressure calibration SI 4.2.B RHR Area cooler therm. calibration SI 4.5.B (Sec 1.a) - RHR Simulated automatic actuation 0/10 SI 4.2.A Group 2 (RER Isolation - Actuation logic functional test 0/5 years SI 4.5.3 (Sec 2.a) (2b) - Air test on the drywell and torus headers and no:zles. 0/10 years SI 3.3.8.A - ASME Section XI Hydrostatic testing of the RHR system SI 3.3.8.3 - ASME Section XI Hydrostatic testing of the RHR system SI 3.3.8.C - ASME Section XI Eydrostatic testing of the RHR system SI 3.3.8.D - ASME Section XI Hydrostatic testing of ~ the RHR system SI 3.3.8.E - ASME Section XI Hydrostatic testing of the RHR system SI 3.3.3.F - ASME Section XI Hydrostatic testing of the RHR system SI 3.3.8.G - ASME Section XI Hydrostatic testing of the RHR system 1421 183

v 2 ge v( \\\\ 7. SYSTEM FRE0VENCY COMPONENTS TESTS SSW Quarterly SI 4.5.C.1 - RHRSW & EECW Valve operability test SI 4.5.C.2 - EECW System functional test SI 4.5.C.3 - RHRSW Pump and header oper flow test Semi-Annual SI 4.2.B RHRSW Initiate logic functional SI 4.2.B RHRSW Ti=er functional and calibration Yearly SI 4.5.C.3 - RHRSW Pump flow rate test SI 4.5.C.4 0/3 years SI 3.3.13A - (Sec 5.1) - ASME Section XI hydrostatic testing of the RHRSW system SI 3.3.13A (Sec 5.1) SI 3.3.13A (Sec 5.2) - ASME Section XI hydrostatic testing of the RHRSW system. SI 3.3.13.3 (Sec 5.2) 8. CORE SPRAY Daily SI 2 _ Core spray sparger to RPV d/p instrument check SI 2 _'C,cre spray discharge pressure knytrument check Monthly SI 4.2.3-24. Core spray sparger to RPV d/p functional test SI 4.2.B Core spray pump discharge pressure functional test SI 4.5.A _ Core spray pump start functional test (starts area coolers) SI 4.5.A _ Core spray system pump operability SI 4.5.A _ Core spray system MOV operability Quarterly SI 4.2.B Core spray sparger to RPV d/p calibration SI 4.2.B-21_ Core spray pump discharge pressure calibration SI 4.5.A . Core spray system and pump' flow rate test Semi-annual SI 4.2.B-39. Core spray auto sequencing timers functional test )hS\\

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g a 3 c. M - PLANT [wi0 .j t. NIT (S) /y DESIGN REQU:REMDITS FOR CONTAINMENT ISOLAT!CN SAF.R!ERS Question: Discuss the extent to which tha quality standards and seismic design classification of the containment isolation provisions follow the reccamendations of Regulatory Guides i.26, " Quality Group Classifications and Standards for Water, Steam, and Radioactive-Water-Containing Components of Nuclear Power Plants", and 1.29, " Seismic Design Classification". .isJCI [ g,;yj,3 k,,. g,3 3,i ..m,),)diinti's n a N Responss: /' /?))' (r't< 2 .sI 0') ') .f'. / i /W.Rj l', / 90 7 [70/TS li J d ) C/1 '. ' ' ~,

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Y -rummt m4, . =. -. - - . - +. - T l' I 3 T* 3 c 7,3y uW f PLANT S r ow w s _';n 7(s)_ \\)1,g Yeeeg PROVIS!ONS FOR TEST:'d Question: Discuss the design provisions for testing the operability of the isolation valves. e c3 s t, $ .D m b s o ~ sS9

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SI 4.7.D Primary Containment Isolation Val ys t This surveillance instruction is used to com ply with requirements of technical specification 4.7.D. The following table cross-references the surveil' lance requirements with surveillance instructions performed to satisfy these requirements: Table 4.7.D Surveillance Requirements Surveillance Surveillance Item Frequency Requirement Instruction 4.7.D.I.a Once/ operating cycle Test the operable isolation valves that are SI 4.7.D.1.a-1 power operated and automatically initiated for closure times. '.7.D.1.a once/ operating cycle Test the operable isolation valves that are See belou (1) power operated and automatically initiated for simulated automatic actuation.

4. 7. D.1.b (1)

Once/ quarter Fully close and reopen all no'rmally open SI 4.7.D.1.h-1 and power operated isolation valves except ECCS HOV Operability HSIV's. tests 4.7.D.1.h(2) Once/ quarter Trip MSIV's individually with reactor power SI 4.7.D.I.h-2 <75% and verify closure time. 4.7.D.I.d Once/ operating cycle Check operability of reactor coolant system SI 4.7.D.1.d instrument line flow check valves. 4.7.D.1.c TVice/ week Exercise HSIV's one at a time by partial closure and subsequent reopening 4.7.D.2 Daily-whenever an In each line having an inoperable valve, isolation valve listed record the position of at least one valve, ~(( in table 3.7.A is which is in the mode corresponding to the y, a y psy inoperable isolated condition, '8$ gg$", (1) Die simulatad automatic actuation is performed in accordanca with the channel and logic testa. The p logic tcat overlaTs the channel teat at the channel ralay, thereby establishing continuity of signal

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T r.. Po*" .i PLANTED-.em hn,, UNIT (S)/ 1., y T }k O D CODES, STANDARDS, AND GUICES ~ @ o lu &3 Cuestion: Identify 'he codes, standards, and guides applied in the design of the containment isolatien system and system components. Response:,'j'; ), y,, yf 7- 'j,- i ,,,y, pi nig, j f f(g/'b'i f;H.f / 5 ' k'til W.L' clwf.)L*.i,)i cr fi f!u Kis aiMC Gr' Odf.S.$ihy J. ll2' syn'ini's,3lpln./...a a n*// D , q.i /..: 2.j f ut:/: n'd w ;re clnyw.i 5,n::r nie en.nr aw,m ce' c.m t r,r :<'< * ' ^*,)-! C;wjr.icri.v 3,). ; ei:: 1i u u.: =, .i.,,;f;,,1) ebi 8.',/i<! j %::":r /.1...- r. : ih:!.

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9 A ve is PLANT Browns Ferry UNIT (S) 1, 2, 3 N0? MAL OPERATING MODES AND ISOLATION MODES Question: Discuss the normal operating modes and containment isolation pro-vision and procedures for lines that transfer potentially radio-active fluids out of the containment. Resoonse: The containment isolation system is provided to prevent release of radioactive materials from the containment during the course of an accident. Pipes that penetrate the drywell and connect to the nuclear steam supply systen are double-valved, as are those that open into the drywell free space. In the case of lines forming a part of the reactor coolant pressure boundary, one valve is located inside the drywell and the other outside, as close as practical to the contain-m.ent. Normally closed double valves in lines to the drywell free sence, such as the vacuum relief fron the Reacter Building and the suppres-sion chamber nitrogen makeup header, are both located outside the containment. In lines forming a closed loop inside the drywell, one re-mote, manually-controlled, cotor-operated valve Ls generally provided outside the drywell. Inlet lines fron the engin'eered safety systems which penetrate the primary containment do not incorporate automatic isolation valves, since operation of these systems is essential foll) wing a design basis accident. Similarly, the feedwater lines to the reactor do not include power operated isolation valves downstream of the EPCI and RCIC injection points, thus assuring use of these lines for post-accident core flooding. However, the inlet lines mentioned above incorporate check valves to assure containment of radioactive =aterials. Isolation Valves The isolation valves which make up the PCIS have been defined into three categories: 1. Class A isolation valves are in pipelines that co=municate directly -rith the reactor vessel and penetrate the pri=ary containment. These lines generally have two isolation valves in series: one inside and one outside the pri=ary containment. 2. Class 3 isolation valves are in pipelines that do not com= uni-cate directly with the reactor vessel but penetrate the primary containment and cocmunicate with the pri=ary containment free space. These pipelines generally have two isolation valves in series, both of them outside the primary containuent.

9. g c I L. 3. Class C isolation valves are in pipelines that penetrate the primary containment but do not communicate directly with the reactor vessel, the pri=ary containment free space, or the environs. These lines require one isolation valve located outside the primary containment. Isolation Functions and Settines The functions that initiate" automatic isolation are discussed as follows: 1. Reactor vessel low water level A low water level in the reactor vessel could indicate that either reactor coolant is being lost through a breach in the nuclear system process barrier or that the normal supply of reactor feedwater has been lost. Reactor vessel low water level ic'.ciates closure of various Class A and Class 3 valves. The closure of Class A valves is intended to either isolate a breach in any of the pipelines in which valves close or conserve reactor coolant by closing off process lines. The closure of Class B valves is intended to prevent the potential escape of radioactive materials from the primary containment through process lines which are in co=munication with the pri=ary containment free seace or suppression pool. Two reactor vessel low water level isolation trip set-tings are used to complete the isolation of the pri=ary containment and the reactor vessel. The first reactor vessel low water level isolation trip setting, which occurs at a higher water level than the second setting, initiates closure of certain Class A and Class 3 valves in najor process pipelines except the main steam lines. The main steam lines are left open to allow the re= oval of heat from the reactor core. The second and lower reactor vessel low water level isolation trip setting completes the isolation of the primary containment and reactor ves-sel by initiating closure of the main steam isolation valves and any other Class A or Class 3 valves that re-quire isolation. The first low water level setting, which is coinciden-tally the same as the reactor vessel low water level scram setting, was selected to initiate isolation at the earliest indication of a possible breach in the nuclear system process barrier yet far encugh below normal opera-tional levels to avoid spurious isolation. Isolation of the following pipelines is initiated when reactor vessel low vater level falls to this first setting. RER reactor shutdown cooling supply RER reactor head soray }} {hh Reactor water claanup

gn g e 19 Drywell equipment drain discharge Drywell floor drain discharge Drywell purge inlet Drywell main exhaust Suppression chamber exhaust valve bypass Suppression chamber purge inlet Suppression chamber main exhaust Drywell exhaust valva bypass Suppression chamber drain RHR-LPCI to Reactor (in shutdown cooling mode) RER flush and drain vent to suppression chamber Drywell purge and vent outlet Drywell makeup Suppression chamber =akeup Exhaust to standby gas treatment The second and lower of the reactor vessel low water level isolation settings was selected low enough to allow the renoval of heat from the reactor for a pre-determined time following the scram and h,igh enough to complete isolation in time for the operation of core standby cooling systems in the event of a large break in the nuclear system process barrier. This low-low water level setting is low enough that partial losses of feedwater supply would not unnecessarily initiate full isolation of the reactor, thereby disrupting nor=al plant shutdown or recovery procedures. Isolation of the following pipelines is initiated when the reactor vessel water level falls to this second setting. All four main steam lines Main steam line drain Reactor water sample line RCIC steam line drain RCIC condensate drain EPCI steam line drain 2. Main Steam Line Isolation The valves listed belcw will be isolated by feur different signals which indicate possible fuel failpres or pipe ruptures in the vicinity of the steam lines: All four main steam lines Main steam line drain Reactor water sample line A. Main steam line hi;h radiation 3. thin steam line space te=perature high C. Main steam line high flow D. Lcwer pressure at turbine inlet 1421 200

18 9 ~ g e. 3. Primary containment (drywell) high pressure. High pressure in the drywell could indicate a breach of the nuclear syr:em process barrier inside the dry-well. The automn::ic closure of various Class 3 valves prevents the release of significant amounts of radio-active material from the primary containment. Upon detection of a high drywell pressure, the following pipelines are isolated: RHRS shutdown cooling supply RHRS reactor head spray Drywell equipment drain discharge Drywell floor drain discharge Traversing in-core probe tubes Drywell purge inlet Drywall main exhaust Suppression chamber exhaust valve bypass Suppression chamber purge inlet Suppression chamber nain exhaust Drywell exhaust valve bypass Suppression chamber drain RHR-LPCI to reactor (in shutdown cooling mode) RHR flush and drain vent to suppression chamber Drywell purge and vent outlet Drywell makeup Suppression chamber makeup Exhaust to standby gas treatment The primary containment high pressure isolation set-ting was selected to be as low as possible without in-ducing spurious isolation trips. 4. HPCI and RCIC Isolation Provisions Various protective features exist to detect pipe breaks in the HPCI and RCIC piping, and to initiate automatic isolation of those systems. A. HPCI/RCIC high steam flow B. HPCI/RCIC steam line low pressure C. HPCI/RCIC equipment space high temperature The isolation setpoint is chosen at a pressure below thatwheretheHPCI/turbinecanoperateefficiently. Ac#(_ 5. Reactor building ventilation exhaust high radiation. High radiation in the reactor building ventilation exhaust could indicate a breach of the nuclear system process barrier inside the prinary containment.51ch would result in increased airborne radioactivity N tis in the primary containment exhaust to the secondary 1421 101

g4 g < l 't containment. The automatic closure of certain Class 3 valves acts to close off release routes for radioactive caterial from the primary containment into the secondary containment (reactor building). Reactor building venti-lation exhaust high radiation initiates isolation of the following pipelines: Drywell purge inlet Drywell main exhaust Suppression chamber exhaust valve bypass Suppression chamber purge inlet Suppression chamber main exhaust Drywell exhaust valve bypass Drywell purge and vent outlet Drywell makeup Suppressson chanber =akeup Exhaust ;o standby gas treatment The high radiation trip setting selected is far enou;h above background radiation levels to avoid spurious isolation, but low enough to provide timely detection of nuclear system process barrier leaks inside the pri-mary containment. Because the primary containment high pressure isolation function and the reactor vessel icw water level isolation function are adequate in effecting appropriate isolation of the above pipelines for gross breaks, the reactor building ventilation exhaust high radiation isolation function is provided as a third re-dundant method of detecting breaks in the nuclear system process barrier significant enough to recuire automatic isolation. 6. Cleanup system space and floor drain high temperature. High temperature in the cleanup system space or the floor drain could indicate a break in the cleanup sys-tem. The auto =atic closure of certain Class A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier. When high temperature occurs, the cleanup system is isolated. The high temperature isolation setting was selected far enough above the anticipated normal area temperature to avoid spurious operation, but low enough to provide timely detection of a cleanup system line break. The following pipelines are isolated: Reactor water cleanup from reactor Reactor water cleanup return \\h \\

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!!OIt!AL OPERATING ?t0 DES During normal operation only those process lines in service will have open isolation valves; as main steam, feed water, and reactor water cleanup. Systems not in service are secured, with their respective isolation valves maintained closed. Technical specifications have strict re-quirements to =aintain primary containment isolation capability. 1421 103}}

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