ML20196F605

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Forwards Comments in Response to 770414 Memo Requesting Review of SEP Draft List.Deletion List Forwarded by 770419 Memo Also Reviewed But No Comments Necessary
ML20196F605
Person / Time
Issue date: 05/02/1977
From: Butler W
Office of Nuclear Reactor Regulation
To: Ziemann D
Office of Nuclear Reactor Regulation
Shared Package
ML20196F442 List:
References
FOIA-87-854, TASK-03-01, TASK-03-02, TASK-03-11, TASK-03-12, TASK-05-01, TASK-05-02, TASK-05-03, TASK-05-04, TASK-05-05, TASK-05-13, TASK-07-02, TASK-07-03, TASK-07-04, TASK-07-05, TASK-08-01, TASK-08-02, TASK-3-1, TASK-3-11, TASK-3-12, TASK-3-2, TASK-5-1, TASK-5-13, TASK-5-2, TASK-5-3, TASK-5-4, TASK-5-5, TASK-7-2, TASK-7-3, TASK-7-4, TASK-7-5, TASK-8-1, TASK-8-2, TASK-RR NUDOCS 8803040130
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{{#Wiki_filter:C ( MAY 2 s 1977 MEMORANDUM FOR: D. Ziemann, Group Leader, Systematic 1 Evaluation Group, OT, DOR l PROM: W. R. Butler, Chief, Plant Systems Branch, DOR

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP) DRAFT COMPREllENSIVE LIST / 1 Members of the Plant Systems Branch, OT/ DOR, have reviewed the first draft of the SEP Comprehensive List as directed i by your memorandum of April 14, 1977. Our comments are written into the appropriate pages of the list. We have also reviewed the Deletion List transmitted by your memorandum of April 19, 1977. We have no comments on this list. il W. R. Butler, Chief Plant Systems Branch Division of Operating Reactors yww=. v.-:~_- .e...~ _g s \\ 1 As 4 8903040130 880229 PDR FOIA CONNOR87-854 PDR t )

3-1 j, .o a a ~; } i. III-l Classification of Structures, Components and Systems (Seismic and 'l Quality) l i A. Seismic Requirements For Auxiliary Systems B. Fuel Pool Cooling System Clatsification C. Classification of Electrical Systems III-2 Wind and Tornado Loadings O A. Design uind Protection B. Tornado Wind and Pressure Drop Protection C. Effect of Failure of Structures not Designed for a Tornado on Safety of Category I Structures, Systems and Components i D. Tornado Effects on Emergency Cooling Ponds i III-3 External Water Level (Flood) i A. Effects of High Water Level on Structures B. Structual Consequences of Failure of Underdrain Systems III-6 Missile Generation and Protection A. Tornado Misstiles B, Turbine-Missiles - bCr.; prds..d b, h-/f w 5 lo. d f b M C 'cf A ~ ~L ' ) U A-S il tu oc C. Internally Generated Missiles (,Inside and Outside ontainment) D. Site Proximity Related Missiles (Including Ai raft) j ..s E. Pump Flywheel Integrity F. Mic.sile Overall Effects \\ J

~ ..i G. Corrosion H. Reactor Pressure Vessel Supports I. Reactor Cavity Subcompartment Analyses to Determine Reactor Vessel Support Capability J. Stress Corrosion Cracking of High Strength Supports / Restraints Fracture Toughness of Component Supports K. Inservice Inspection for Supports and Bolting of LilR Class 1 h L. Components M. MSRV Line Restraints III-10 Pumps and Valves Integrity A. Operability B. Testability C. Inservice Inspection D.-Seismic -Quali fication-h,c,(.c!-}c.a r; yo fjj (2 (,' E. Passive Mechanical Valve Failures F. Thermal-Overload Protection. fcr Motors of Motor Operated Valves G. Motor-operated Valves in ECCS Accumulator Lines H. Pump Flytiheel Integrity I. Manual Valve Position Requirements J. Surveillance Requirements on' BUR Recirculation Pumps and Discharge Valves b77 ph I.1 ew

3-5 (;3 a s.- III-ll Miscellaneous Mechanical Component Integrity A. Qualification of Safety Related Equipment l B. Inservice Inspection l l C. Effect of Faulted Conditions on Component Integrity l III-12 Environmental Qualification of Safety Related Equipment (including Mechanical and Electrical Equipment) g, A. Temperature, Pressure, Humidity l B. Chemistry l t (: C. Seismic I D. Radiation i E. Flooding of Equipment Important to Safety F. Operability of Components Under Accident Loads f G. Mechanical System Reliability

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A 51 42 ,' } g V-1 Compliance with Codes and Standards (10 CFR 50.55a) V-2 Applicable Code Cases --s / f f.,?; f ) G. (,4 fr~cw."' c..~. r/ (?..q CEf.( l g e p(c % V-3 RCS Cverpressurization Protection J '.y.c ) [E,0 S A. Design Basis Transient

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/ ,I. B. Protection System Design Criteria ~ C. Reactor Vessel Overpressure Protection - PWR h D. Qualification of Steam SRV For Dynanic Loads E. Safety / Relief Valve Requirements V-4 Piping Integrity A. Materials and Fabrication B. Corrosion C. Fracture Toughness D. Inservice Inspection E. Flaw Evaluation Criteria F. BWR Bypass Line Cracks G. Stress Corrosion Cracking in BWR Piping H. Control of the Use of Sensitized Stainless Steel I. Pipe Rupture Protection V-5 RCPB Leakage Detection

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d-4 a (~"3 Y-11 High Pressure - Low Pressure Interface A. Isolating Low Pressure Systems Connected to the Reactor Coolant Pressure Boundary B. Requirements for Isolation of High and Low Pressure Systems C RHR -Interlock' Requirements-

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[O C I i f V-12 Reactor k'ater Cleanup System - BUR's A. Chemistry of Boiling Water Reactor Primary Coolant gl l l i l l - e, t 3 i l la*k lA 1977

- - w. - n. - - - b-b a ~ ' ( l'. VI-12 Main Steam Isolation Valve Leakage - Regulatory Guide 1.96 - 55iR's l A. MSIV Reliability - PWR l B. MSIV Failure Investigation i ~ i C. Main Steam Line Isolation Seal System D. MSIV Uesign Dssis j VI-13 Other Engineered Safety Features Aspects Testing of Reactor Trip System and Engineered Safety Features, /A / 4 t2 TI4 O f4 0 t w A. I B. Shared Safety and Service Systems for Multiple Unit Facilities h.f. C. Effect of Failure in Non-Safety-Related Systems During Design Basis Reviews l. O i i n?R 1 1 1977

9 o 7-1. i. ' VII-1 Reacter Trip System (IEEE-279) A ig/2 cc .6-Cf Cb [Ib A. Comon:.er. Applications--RPS- /t.so - J B. Maltiplexing In Control and Safety Systems C. Automatic Resetting of Reactor Trip System Trip Bistable Relays

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0g5fc~wS cm O'~- 0 w ~ !s' t%lg% c J VII-2 ESF Actuation System - f 3 Tjofc wow.S,r.%'cM, C A. On-Line Testability of ECCS Actuation System and ' a Component Availability W

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[ 6.O '.-c[:a s.'-?, k,f[ .s bw ['.~/ l c R C-- rI.. ),t.it j.s. A.dC.[Acb-.} k 'v l VII-3 Systems Required for Safe Shutdown 1 c; ca, rre.f kra.c./- Ta r dwr.S O.w c(. O,/ 8 A. Safe Shutdown Systems / / L w $0rc t.h % Q $ [ f*Cv.e I r3 $ u.b.a t% B. Remote Shutdwon C,6 Ai$t Mck O[L Sd'L. f,y 4 9t* m &,,.ch 3,- C. Method of Bringing a PWR from a High Pressure Condition J l to Low Pressure Cooling Assuming the Use of Only Safety Grade Equipment VII-4 Safety Related Control and Instrument Potter A. Transfer of Safety Related Equipment From fiormal to Emergency Power and Return ( r, _... c,[,9, Tju Tj 4 3 ch,, 8. Loss of Plant Air Systems (Effect on Plant Control and I fionitorina) VII-5 Instrumentation to Follow Accident A. Post-Accident Instrumentation

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81 VIII-l Offsite Power Systems (GDC. Reliability and Stability) A. P_otential-Equipment ~ Failures Associated with Degraded ( cs r//.q/ C w r.(_ b rc~s 'TE)* [ De T d ( Gr.id _Vol tage '- ( b""' f ' ' O,1 _ - Cf._ k B. Loss-of-Offsite Power Subsequent to Manual Safety Injection Reset Folicwing a LOCA ,- / .. _S "..* r.'i_ *a ( cn*Wr.<5. -+ ' t Y ' ;j!-- ~L D d c u t C. Load Break Switch - - ~ - - (b<.d h e- .ht-J.c 4[>t' r:~/ c b.X' ,34 VIII-2 Onsite Emergency Power Systems / d'cl c / [l '/,:A /t8' k_cu., A7-c.gf A. Load Break Switch--- ~~~~~ [Cw'A i f ri f 9 e g B. Emergency Power for Two or More Reactors at the Same DV' Cd & c /.', , C ' * ~ - a',' L<{ ( (/ u Site C. Diesel Generator (Qualification, Reliability, Operation) g,Q,7{* VIII-3 Emergency DC Power Systems A. Station Battery Capacity ' Test Requirements VIII-4 klectricalContainmentPenetrations U t%. '. hd :- ; h h iot t~. Gt iul-iG.~ - Q, p C i't ; :. c.t 1 :: s G ~. -

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U A. Failure of Containment Penetrations from Electrical Faults insid2 containment during LOCA VIII-5 Loss of All A-C Power.. J. [ g.[. 3, c.;,.s. [, < d ,. Z C s 1 r,c.~ h kh ( i\\, 'e* 9 pyy

9-2. D. Ccoling Water Makeuo Requirements E. Ultimate Heat Sink and Safe Shutdown Systems IX-4 Process Auxiliaries ^ A. Chemical and Volume Control System 1. Boron Addition System 2. Baron Precipitation (Post LOCA) g 3. tieed For Improved Equipment B. Standby Liquid Control System (8k!R) IX-5 Ventilation Systems A. Control Room Ventilation System B. Spent Fuel Pool Area Ventilation System C. ESF Equipment Ventilation System r ~

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(.C.1 D h . ~~,.. I IX-6 Fire Protection ./ (!,,k{,,'{ 5 Wf^ I IX-7 Diesel Gencrator j b{ [- h dets 3 N' 7 A. Requirements / ) m l -),.g.,.? g. ' y,E bkSOM Of $~L i B. Qualification c(c n% Of ,of r <. [ O I O'A_ C. Reliability 4 / e, e D. Operation g.r.. ( g q c.,, E. Lock Out / F. Fuel Oil g/ .; lill IX-8 Communicaton and Lighting Systems t

\\t o-d r '<. i a l t X-1 Condensate'and Feedwater System l s X-2 Auxiliary Feedwater Systcm l A. Auxiliary Feedwater System Pump Drive and Power ( ~ Diversity B. Water Hamer in Feedwate r System C. 2.[,w 3 $,.( (j g,.; w i % k.':'-._ '5. (c.., [- o ( b[ h [f c, } {,l.,' g i l I i i l l I l I i i ~ O! 1 i i 4 I i ) i l 4 i i } i APR 141977 i i l

15-1 1, -s\\*, \\! XY-1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow and Inadvertent Opening of a Steam Generator Relief or Safety Valve XV-2 Spectrum of Steam System Piping Failures Inside and j Outside of Containment (PHR) I XV-3 Loss of External Load, Turbine Trip, Loss of' Condenser Vacuum, Closure of Main Steam Isolation Valve (BUR), and Steam Pressure Regulator failure (Closed) O! XV-4

6. Loss of flon-Emergency A-C Power to the Station Auxiliaries b.Lossof0.s.t 6. t.. G' q r.-a'/ b C G c. 3 f

XV-5 Loss of Normal Feedwater Finw i XV-6 Feedwater System. Pipe Breaks Inside and Outside 1 Containment (PWR) I i XV-7 Loss of Forced Reactor Coolant Flow Including Trip of Pump and Flow Controlled Malfunctions i i Reactor Coolant Pump Rotor Seizure and Reactor Coolant h I XV-8 ~ Pump Shaft Break 1 i XV-9 Uncontrolled Control Rcd Assembly Withdrawal Frcm a Subcritical or Low Power Startup Condition XV-10 Uncontrolled Control Rod Assembly flithdrawal at Power , ; d.*? j 3 i

~m O .{ l XV-ll Control Rod Misoperation (System Malfunction or Operator Error) t l I XV-12 Startup of an Inactive Loop or Recirculation loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BUR CORE FLOW RATE XV-13 Chemical and Volume Control System iblfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant (PWR) O i XV-14 Inadvertent Loading and Operation of a fuel Assembly in an Improper Position i f XV-15 Spectrum of Rod Ejection Accidents (PWR) Xhl6 ___ Spectru:a.of. Rod Drop Accidents (BWR) ~ hoy,rt,[ b [Od O G A " C U- / y - ( R SC S) XV-17 Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction That Increases Reactor Coolant Inventory XV-18 Inadvertent Opening of a PHR Pressurizer Safety / Relief ~ Valve or a BUR Safety / Relief Valve XV-19 Failure of Small Lines Carrying Primary Coolant Outside Containment Radiological Consequences of Steam. Generator Tube"Failure (PWR) XV-20 W ffR 14 3

h0-8 a C1, .n XVI-l Technical Specifications A. Comparison to Standard Tech. Specs. (STS) 8. Review Calculation Bases - Quantitative vs. Qualitative C. Operation with Equipment Out of Service (LCOMS) D. Compatibility with Design of Plant ~ E. Action to be Taken in Response to Fuel Failures (Level of Activity Before Action is Required; Monitoring) F. Instrument Trip Setpoints in Standard Technical Specifications G. tiuclear Uncertaintie; in Tech Specs H. LUR Fuel Cladding Design Limits I. Analyses and Reduction of In-core Measurements J. Maintenance and Inspecticn of Plants K. Allowable ECCS Equipment Outage Periods (,. O vs. g>e, na-e!n-G w R Eg(,7 c h 0 wYsy

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