ML19309A123

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Interim Deficiency Rept on Effects of Piping Break Outside Containment.Inservice Insp Program Described
ML19309A123
Person / Time
Site: Rancho Seco
Issue date: 05/01/1973
From:
SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
References
2852, NUDOCS 8003260840
Download: ML19309A123 (44)
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Text

{{#Wiki_filter:. ~ 66 3 /W e m ~ g(QT f nLud r[pd 3 3 APR3 01973, 2 i t S U.S. ATOMIC ENERat 'k COMMISSI:,16 Regule:ory Mad Sectice / I Cu 9 Regulatory File Cy. l nd.as ciur tw&/-N - t i. INTERIM REPORT ON EFFECTS OF A PIPING BREAK OUTSIDE THE CONTAINMENT I l !~ i RANCHO SECO NUCLEAR GENERATING STATION i r-I { MAY 1,1973 i f,. i {~ ...r.s u 1:e ':{ j:]',]': 7 Ebi ~ gu016 ~~~ 1 ('t m I "003200ft,rg Eu o 7 2852 bee

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.7.. t. - g- . a., TABLE OF CONTENTS - 7 Introduction. 1 Question 1........................... 4 Question 2. 6 Question 3...................... 8 Question 4. 9 t Question 5........................... 11 Question 6........................... 14 n Question 7.... 15 Question 8................ 15 Question 9................ '. 6

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Question 10... 16 Question 11.. 17 Question 12. 19 Question 13........................... 20 Question 14. 22 Question 15............... 23 Question 16........................... 23 l i Question 17....................... 24 Question 18............... 25 Question 19........................... 26 ' I Question 20. 27 Question 21........................... 28 T* w .9 \\ i w.' e -m-n ~ .v v

I INTERIM REPORT ON EFFECTS OF A PIPING BREAK OUTSIDE THE CONTAINMENT .i INTRODUCTION This interbn report covers the effects of a piping system break outside the containment. A double ended rupture of the main steam or feedwater line has been considered. The effects of a pipe break including pipe whip, '"~ jet stream impingement, environmental effects and compartment flooding are covered. I In analyzing the steam or feedwater line break accident, loss of offsite power is assumed to occur concurrently with pipe failure. In analysis of the accident single failure of an active component is assumed to occur during subsequent cooldown to cold shutdown conditions. The question numbers in this report refer to the app'. cable sections of the AEC requirements for information on the subject of pipe break outside the containment transmitted to SMUD on December 14, 1972. Inservice Inspection The District will perform 100 percent inservice inspection of all circumfer-r' ential welds in the main steam and feedwater lines to allow installation of the terminal end restraints and guard pipe restraints after fuel loading. Actual stress values are much lower than allowable. For the main steam line the maximum calculated stress is 0.460 of allowable. For the main feedwater piping the maximum calculated stress is 0.536 'f allowable. 10 CFR-50 Paragraph 50.55a provides the code requirements for the design fabrication inspection and installation of piping systems. As indicated in this regulation, it would have been acceptable for the District to use ANSIB31.1 for the piping under consideration. Instead, for added con-servatism, the District elected to use 1969 ANSIB31.7 Nuclear Power Piping Code. In addition, the District imposed other stringent requirements on the fabrication and installation of these piping systems to assure the quality of welds and material and provide additional confidence in the integrity of there systems. The additional requirements imposed include: 1. Radiograph of all welds on feedwater main (code only requires L, radiograph on section from stop valve to OTSG). 2. Material test reports for physical and chemical characteristics. ,1'" 3. Certified stress analysis. 4. Review and maintenance of r. Welding procedure qualifications r-t C . f9' .y 1

. r' [- r; L. b. Welder qualification records c. Weld repair charts r. d. Welder identification to welds I'- e. Heat treat charts '[" All piping systems have been fabricated by organizations with quality i-assurance programs meeting the requirements of 10 CRF 50 Appendix B. The piping systems, 215 inch and larger were f abricated by either of two manu-

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facturers in the Los Angeles area. The 2 inch and smaller pipe was fab-j ricated by the site erection contractor who also performed all field crection work and assembly welding. All three organizations have been surveyed by 7.. ASME and are current holders of "N" stamp authorization. Prior to contract award these contractors were audited by the SMUD Quality Assurance Director. Subsequent audits by him have verified performance in accordance with their respective manuals. All pipe and fittings have been procured under requirements for furnishing heat numbers and material chemical and physical test results and non-i destructive examination, as require'd to meet the project specifications. These specifications incorporate all requirements of ANSI B31.7. e. p . t, Throughout all of the fabrication and site erection work control has been exercised over all welding operations. Strict weld rod control has been observed and all welding has been performed in accordance with welding procedures approved by the engineer. The engineer specified the procedure used for each field welded joint. Non-destructive examination of all welds has been in accordance with ANSI B31.7 for the classification of system involved. Prior to the acceptance of piping systems they are subjected to all require-ments of the SMUD Quality Assurance Program. This includes complete docu-mentation of non-conforming items and disposition in accordance with the Manual. Assurance of all Code requirements is met by the State of California who are performing site inspection. A dynamic analysis of the main steam line was made for steam hammer affect upon turbine stop valve closure.

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During hot functional testing and power tracsient tests, piping will be monitored for displacement and movement and s tresses will be evaluated and .c compared with calculated values. The use of strain gages and scratch plates where applicable will be used to verify that the as-built plant.is in L-agreement with the design. Strain gage determinttions will be made on safety valve mounting welding nozzles. 4*% i L.J q 2 ij

f-l, i f' L.: For piping systems, where modifications have not been completed by power operation an inservice surveillance program will be implemented until the modifications are complete. Work will continue during olant operation on 6. a timely basis to complete the modifications. At least once per shif t, a visual inspection of all high energy lines where the modifications have not been completed will be conducted for the detection of leakage or enrion-mental conditions that might adversely effect the integrity of those lines. If the plant is shutdown and cooled down for any other reason for a three month interval, the insulation will be removed from the affected piping and the piping will be inspected for signs of structural distress and leakage. For any modification that is not completed by the end of the I. first refueling, the af fected welds will be x-rayed during the first refueling. Because of conservative design methods an. thorough inspection o these systems during construction, the Districe believes that their inservice inspection program will be adequate to insure that catastrophic failures will not occur. i. 1. 4w

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-4 a4 ..,.4wm= e.r. + es b#-*-% o ' f> QUESTION 1. The. systems (or portions of systems) for which protection against pipe whip is required should be identified. Protection from pipe whip need not be provided if any. L of the following conditions will exist: [' (a) Both of the following piping system conditions i,, are met: (1) the service temperature is less than 200 F; I and (2) the design pressure is 275 psig or less; or

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(b) The piping is physically separated (or isolated)- from structures systems,' or components important to safety by protective barriers, or restrained t from whipping by plant design features, such as concrete encasement; or r-l (c) Following a single break, the unrestrained pipe b' movement of either and of the ruptured pipe in any possible direction about a plastic hinge formed at the ne.arest pipe whip restraint cannot impact i l_ any structure, system, or component important to safety; or . r, (, (d) The internal energy level associated with the shipping pipe can be demonstrated to be insuffi-cient to impair the safety function of any struc-j ture, system, or component to an unacceptable s, level. f ( ANSWER: The following systems are required for cold shutdown. Portions of these r- .{ systems requiring pipe failure protection are identified in table 1. I' Class I batteries & busses both a-c & d-c i NS busses & switchgear t 'I 4 Control room emer. air conditioner Reactor protection system NSCW system including all cables to P-462 A&B L. NSRW system including all cables to P-482 A&B ~~ Decay heat system including all cables to P-261 A&B and to room coolers e,- 4 - CJ b

7 sh Emer. FW system including all cables to P-319 & FV-30801 r-l Auxiliaries for diesel generators & rooms including cables & lines to F.O. Pumps P-888A-D and all cables to fans A-544A-D I" Make-up system including power to P-236 & P-238A&B and L.O. pumps and. room 1-coolers A529A-E Ductwork for emer. control room air conditioner and power cables ~ Class I. structures housing or supporting impactees Suction from BWST to HPI, DH & RB spray pumps Boric acid supply to suction of HPI pumps including cables to heat tracing and pumps P-705A&B i. Borated water storage tank Condensate storage tank INSTRUMENTATION Instrucentation source range detectors Instrumentation rod position indication TE-21023A, 24B RC inlet temp. PT-21092, 38, 40, 42, 43, 39, 37 reactor coolant pressure ' [' LT-20503A, B&C pressurizer level L. LT-20504C&D Stm. gen, level [, Question la above as modified by the following criteria defines high energy systems: 'T Definition of a high energy line - Lines which do not exceed 200 F or 275 psig do not need to be considered high energy lines. In this case, the pipe whip and environmental effects resulting from breaks need not be considered. If a line exceeds 200 F and 275 psig, it is a high energy line and both the local and environmental effects in pipe whip must be us investigated.. Lines which exceed either the temperature limit or the pressure limit but not both simultaneously, were evaluated considering ,e.l the environmental effects due to cracks but not evaluated for pipe whip effects. -The following diagram illustrates this: !L

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n (2) any intermediate locations between terminal ends where the primary plus secondary stress intensities Sn (circumferential or longitu-dinal) derived on an clastically calculated ,g.j basis under the loadings associated with one-half safe shutdown earthquake and oper-4 exceeds 2.0 S ational plant conditions m i,t for ferritic steel, and 2.4 S for m <f austeritic steel; (3) any intermediate-locations between terminal ends where the cumulative usag'e factor (U)6 '~ derived from the piping fatigue analysis and ll based on all normal, upset, and testing "4 plant conditions exceeds 0.1; and 'l~ (4) at intermediate locations in addition to j! those determined by (1) and (2) above, selected on a reasonable basis as necessary to provide protection. As a minimum, there .i should be two intermediate locations for each piping run or branch run. { ' (b) ASME Section III Code Class 2 and 3 piping breaks should be postulated to occur at the following locations in each piping run or branch run: I}- (1) the terminal ends; (2) any intermediate locations between terminal ends where either the circumferential or il longitudinal stresses on an elastica 11y calculated basis under the loadings asso-ciated with seismic events and operational

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plant conditions exceed 0.8 (Sh+S) OY A the expansion stress exceed 0.8 S ; and A

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(3) intermediate locations in addition to these v determined by (2) above, selected on reason-able basis as.necessary to provide protec-tion. As a minimum, there should be 1 two intermediate locations for each piping run or branch run. .!^ 40perational plant conditions include normal reactor operation, upset con-ditions (e.g., anticipated operational occurrences) and testing conditions. 1 SS is the design stress intensity as specified in Section III of the ASME m Boiler and Pressure Vessel Code, " Nuclear Plant Components." 6U is the-cumulative usage factor a. specified in Section III of the'ASME l Boiler and Pressure Vessel Code, " Nuclear Power Plant Components." ) 7

(~ t(- f ANSWER: (a) Question 2a is not applicable to Ranch Seco. There are no ASME Section III Code Class I piping outside the containment structure. 'f (b) Pipe break locations were postulated for high energy lines at the 'I terminal ends and at least two intermediate points at the most adverse locations where the relative stress values were the highest. There are no locations on lines considered where the f stress values exceeded 0.8 (Sh + S ) or 0.8 S,. For the most a highly stressed line the calculated valves are 0.524 (Sh + Sa) and 0.542 S,. QUESTION 3. The criteria used to determine the pipe break orientation at the break locations as specified under 2 above should be equivalent to the following: 8 [ (a) Longitudina1 breaks in piping runs and branch j runs, 4 inches nominal pipe size and larger, and/or (b) Circumferential9 cracks in piping runs and branch runs exceeding 1 inch nominal pipe size. ANSWER: l The consideration given for breaks in which plant structures, systems, and components important to safety could be affected included full circumferential l breaks in high energy lines in 1 inch and larger sizes; and, in addition, longitudinal breaks in p1pe 4 inches nominal pipe size and larger. L 7S is the stress calculated by the rules of NC-3600 and ND-3600 for h l, Class 2 and 3 components, respectively, of the ASME Code Section III Winter 1972 Addenda. SA is the allowable stress range for expansion stress calculated by rules I of NC-3600 of the ASME Code, Section III, or the USA Standard Code for Pressure Piping, ANSI B31.1.0-1967. 0 j Longitudinal breaks are parallel to the pipe axis and oriented at any point around the pipe circumference. The break area is equal to the effective s cross-sectional flow area upstream of the break location. Dynamic forces . ),_ resulting from such breaks are assumed to cause lateral pipe movements to the direction normal to the pipe axis. 9Circumferential breaks are perpendicular to the pipe axis, and the br'eak area is equivalent to the internal cross-sectional area of the ruptured ? pipe. Dynamic forces resulting from such breaks are assumed to separate the piping axially, and cause whipping in any direction normal to the pipe axis. 3-() 8 U~

QUESTION 4. A summary should be provided of the dynamic analyses applicable to the design of Category I piping and associated supports which determine the resulting load-ings as a result of a postulated pipe break including: (a) The locations and number of design basis breaks on which the dynamic analyses are based. I (b) The postulated rupture orientation, such as a I circumferential and/or longitudinal break (s), for each postulated design basis break location. 1. (c) A description of the forcing functions used for the pipe whip dynamic analyses including the direc-tion, rise time, magnitude, duration and initial conditions that adequately represent the jet stream dynamics and the system pressure difference. l (d) Diagrams of mathematical models used for the dyanmic analysis. (e) A summary of the analyses which demonstrates that i unrestrained motion of ruptured lines will not damage to an unacceptable degree, structure, sys-tems, or components important to safety, such as the control room, r ANSWER: 1 l l \\ b (a) Design basis break locations are discussed in the answer to Question 2 J and are identified on table 1 and figures '1, 2, 3, 4 & 5. t s (b) Postulated break orientations are discussed in the answer to Question 3. (c) The pipe whip forcing function used in the analysis for all. pipe l o restraints other than guard pipe restraints is derived in reference (1) as modified by reference (2). Jet stream dynamics y are approximated in the reference document by calculating a jet thrust force based on the pressure difference inside and outside the break using appropriate impact factors and thrust multiplication factors i for single or two phase flow from the break. l The reaction force acting on the pipe caused by momentum change of fluid flowing through the pipe can be expressed as a function of the 'l s (1)BN-TOP-2 Design for Pipe Break Effects - Bechtel Corporation, September 20, 1972, Revision O. } (2)AEC Letter to Bechtel Corporation. DeYoung to Allen, February 27, 1973. LJ t. s 9 Lv \\

p CASE CROSS SEC' MATHEMATICAL MODEL AT FORCE NUMBER F = L w C R \\t ( ~ = a =f l l fg c, }p k A S x+ g C : * =- F = L G) i 2 K u te w

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_\\ ION DESCRIPTION OF MODEL 3 (R) IN THE ELASTIC ANALYSIS THE LENGTH (L) IS SMALL AND CONSEQUENTLY THE STRESS IN THE PlPE WILL NOT EXCEED THE ELASTIC LIMIT. IN THE PLASTIC ANALYSIS HINGES I ARE ALLOWED TO FORM ALONG THE PlPE. IN ALL CASES IN WHICH HINGES ARE ALLOWED TO FORM THE DECISION HAS BEEN BASED ON THE ACCEPTABILITY OF ANY ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET OR ITS REDIRECTION. IN THE ELASTIC ANALYSIS THE LENGTH (L) IS SMALL AND CONSEQUENTLY THE STRESS IN THE PIPE WILL NOT EXCEED THE ELASTIC LIMIT. IN THE PLASTIC ANALYSIS A PLASTIC HINGE IS ALLOWED TO FORM ANYWHERE ALONG THE LENGTH OF THE PIPE. IN ALL CASES IN WHICH HINGES ARE ALLOWED TO FORM THE DECISION HAS BEEN BASED ON t' THE ACCEPTABILITY OF ANY ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET OR ITS REDIRECTION. UNLIKE THE PREVIOUS CASES WHERE GUILLOTINE FAILURES WERE POSTULATED, THIS CASE INVOLVES A SLOT RUPTURE IN THE LONGITUDINAL DIRECTION. IN THE ELASTIC ANALYSIS THE LENGTH (L) WILL BE SMALL AND PIPE STRESSES WILL BE IN THE ELASTIC RANGE. IF THE LENGTH (L) BECOMES QUITE LARGE, A PLASTIC ANALYSIS WILL BE PERFORMED. IN ALL CASES IN WHICH PLASTIC HINGES ARE ALLOWED TO FORM THE DECISION HAS BEEN BASED ON THE ACCEPTABILITY OF ANY ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET OR ITS REDIRECTION. A SLOT RUPTURE IN THE LONGITUDINAL DIRECTION IS POSTULATED TO OCCUR AT POINT C. IN THE ELASTIC ANALYSIS THE LENGTH (L) WILL BE SMALL. IN THE PLASTIC l ANALYSIS THE LENGTH (L) CAN BE LARGER AND PLASTIC HINGES CAN OCCUR ALONG THE LENGTH OF THE PIPE. IN ALL CASES IN WHICH HINGES ARE ALLOWED TO FORM THE DECISION HAS BEEN BASED ON THE ACCEPTABILITY OF ANY ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET V OR ITS REDIRECTION. f L

PIPE BREAK OUTSIDE C0t i CASE MATHEMATICAL MODEL CROSS SEC NUMBER AT FORCE F L + A $? ~ 1

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) lTAINMENT-MATHEMATl CAL MODELS IN DESCRIPTION OF MODEL R) IN THE ELASTIC ANALYSIS THE LENGTH (L) AND THE RESTRAINT GAP (A) ARE SMALL. IN THE PLASTIC ANALYSIS THE' LENGTH (L) AND THE RESTRAINT GAP (A) CAN BECOME LARGER AND A PLASTIC HINGE WILL OCCUR AT THE POINTS OF MAXIMUM MOMENT. (i.e., ' f A, B, OR, A AND B.) THE PRINCIPAL OF VIRTUAL WORK REQUIRES THAT THE WORK DONE DURING PLASTIC STRAINING BE EQUAL TO THE WORK DONE BY THE JET FORCE (F). IN ALL CASES WHERE PLASTIC HINGES ARE ALLOWED TO FORM THE DECISION HAS BEEN BASED ON THE ACCEPTABILITY OF ANY ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET OR ITS REDIRECTION. IN THE ELASTIC ANALYSIS THE LENGTH (L) IS SMALL AND CONSEQUENTLY THE STRESS IN THE PIPE WILL NOT EXCEED THE ELASTIC LIMIT. IN THE PLASTIC ANALYSIS THE LENGTH (L) CAN BE LARGER AND PLASTIC HINGES CAN OCCUR AT A, B, OR, A AND B. THE VIRTUAL WORK PERFORMED BY THE JET FORCE WILL BE EQUAL TO THE COMBINED TOTAL WORK PERFORMED BY THE HYDRAULIC SWAY COMPRESSOH AND THE PLASTIC HINGES. IN ALL CASES IN WHICH HINGES ARE ALLOWED TO FORM THE DECISION HAS BEEN BASED ON THE ACCEPTABILITY OF ANY ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET OR ITS REDIRECTION. IN THE ELASTIC ANALYSIS THE LENGTH (L) IS SMALL AND CONSEQUENTLY THE STRESS IN THE PIPE WILL NOT EXCEED THE ELASTIC LIMIT. IN THE PLASTIC ANALYSIS THE h LENGTH (L) CAN BE LARGER AND PLASTIC HINGES CAN OCCUR AT A, B, OR, A AND B. THE VIRTUAL WORK PERFORMED BY THE JET FORCE WILL BE EQUAL THE TOTAL WORKED STORED IN THE PLASTIC HINGES. IN ALL CASES IN WHICH HINGES ARE I ALLOWED TO FORM THE DECISION HAS BEEN BASED ON THE ACCEPTABILITY OF ANY / ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET OR ITS REDIRECTION. IN THE ELASTIC ANALYSIS THE LENGTH (L) IS SMALL AND CONSEQUENTLY THE STRESS I IN THE PIPE WILL NOT EXCEED THE ELASTIC LIMIT. IN THE PLASTIC ANALYSIS THE LENGTH (L) CAN BE LARGER AND A PLASTIC HINGE WILL OCCUR AT THE POINT OR POINTS OF MAXIMUM MOMENT.THE POINT OR POINTS OF MAXIMUM MOMENT MAY OR MAY NOT BE LOCATED AT THE RESTRAINTS. IN ALL CASES IN WHICH HINGES ARE ALLOWED TO FORM THE DECISION HAS BEEN BASED ON THE ACCEPTABILITY OF ANY ATTENDANT PIPE WHIP AND CHANGING ENVIRONMENTAL EFFECTS CAUSED BY THE SWEEP OF THE JET OR ITS REDIRECTION. l \\ l

PIPE BREAK OUTSIDE CONTAINMENT-MATHEMATICAL MODELS 0 MATHEMATICAL MODEL N R 3T. O R IMPINGEMENT PHASE EFFECT ON GUARD PIPE LOAD EMPLOYS COMPONENT USED THRUST FACTOR IN MATHEMATICAL .440 IMPINGEMENT _ /

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"(1MP1NcEMEg y-l / v 10" i SECTION A-A g g 1 D s IMPINGEMENT AREA FOR IMPACT PHASE MODEL PLEASE SEE NEXT SHEET

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t ~' PIPE BREAK OUTSIDE CONTAINMENT-MATHEMATICAL MODELS MATHEMATICAL MODEL DESCRIPTION NUMB R F0 R IMPACT PHASE EFFECT ON GUARD PIPE RESTRAINT LOAD EMPLOYS B COMPONENTS USED IN MATHEMATICAL .1SD MODEL h

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IMPACT AREA a q j/ s., R IMPACTk / AREA 1.44D / R,r N / x g I IMPINGEMENT L 1 B y u AREA v N 1 SECTION 0-8 1 I I ,1 D IMPACT AREA I IMPINGEMENT AREA

PIPE BREAK OUTSIDE CONTAINMENT-MATHEMATICAL MODELS C S MATHEMATICAL MODEL DESCRIPTION N M8 R FO C R THRUST PHASE EFFECT ON B GUARD PIPE ~ I LOAD EMPLDYS THRUST FACTOTt iI ~ s ii + l:Il THRUST i x I AREA I REACTIONd ,r~ - i AREA Y \\ D Rg D F' }I g' B THRU T AREA )I h h SECTION A-A g lWy-g g n k) l E IMPINGEMENT -. o l,i AREA 3 FR g. ] g .i B B r h q [ ip REACTION AREA D ~ SECTION 0-8 = = FOR IMPACT PHASE = = PLEASE SEE NEXT SHEET

7 q __q .) PIPE BREAK OUTSIDE CONTAINMENT-MATHEl:ATICAL MODELS MATHEMATICAL MODEL C OSS S CT DESCRIPTION N R R IMPACT PHASE FE E COMPONENTS USED EFFECT ON GUARD IN MATHEMATICAL PIPE RESTRAINT e MODEL C LOAD EMPLOYS IMPACT FACTOR j ' R g 1 r IMPACT _m IMPACT AREA L I l E, l g AREA i f Ii / F R,R C

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F i / r r, r ,M 3 .L_h F SECTION C-C 31 30' IMPACT 1 IMPINGEMENT d' AREA i_ },, _ _,i. AND ROTATIO f I j REACTIONS g ~, AREAS G j' y G y FpR n T SECTION E-E, E'-E' D2 2 . A' 5 .g 3 IMPACT AREA l (A FINIT ELEMENT ANALYSIS REQ 'D HERE) SECTION G-G

.b q., l \\ QUESTION 5. A description should be provided of the measures, as j p applicable, to protect against pipe whip, blowdown jet { and reactive forces. 1 I (a) Pipe restraint design to prevent pipe Lhip impact. (b) Protective provisions for structures, systems, and components required for safety against pipe whip and blowdown jet and reactive forces

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(c) Separation of redundant features; (d) Provisions to separate physically piping and other components of redundant features; and -[ (e) A description of the typical pipe whip restraints and a summary of number and location of all restraints in each system. ANSWER: h, 'l-(a) Measures to protect against pipe whip blowdown jet and reactive forces will be guard pipe restraints. (See figure 6' for typical device), pipe whip restraints and structural shielding. I L 1. Guard Pipe Restraints: j Restraints alone are sufficient to protect against pipe whip, blowdown jet and reactive forces. In some cases, the combined effects associated with postulated pipe breaks require partial encasement to protect adjacent components. An articu-lating encasement can be provided to redirect the fluid and control L pipe movement. The guard pipe restraint may be in ta11ed in-side of the normal insulation with small clearances between it and the main pipe or spacers may be provided and the encasing ,f, device installed outside of the normal insulation. In either case, the impact is substantially reduced by limiting the distance through which the pipe accelerates. Impact, jet forces and internal pressure dictate the design of the installation. Generally, the impact is through a distance less than one inch, .i and though the pressures are less than the main line pressure i by a significant amount because of vent.ing no credit for reduc-tion is taken. The unbalanced jet forces are controlled by the { placement and sizing of ports. In most cases, the unbalanced ,1 jet forces are less than 10 percent of the design basis value and well within the capacity of the normal supports and restraints. '} 'f kJ 11

p <~ Articulation and some clearancesare required to allow the main t pipe to expand and contract. Articulation is provided by the f, use of loose couplings to tie the sections together. Straight 3' runs are used to provide the anchorage points, since straight j portions of the main pipe are not stressed as highly as tees

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Structurally, therefore, the piping retains con-tinuity through the encasement device even through the main pipe is completely severed or split open. Typics.1 details l[ of this device are shown on figure 6. I Detailed calculations will be provided for guard pipe coupling sleeves, attachment rings welded to the main lines and other ,y. special components. Attachments welded to the main line and .I.. the guard pipe restraint will be designe'd supported and installed so as to not introduce significant strain concen-I trations during normal upset or faulttid conditions on the encapsulated section of pipe. i .( Axial displacement and motion within the sleeve following a postulated circumferential pipe break will be precluded by ~ suitable attachments to the main line or at the terminal end restraint. ./ The encapsulation sleeve will be designed to withstand the dynamic forces of internal pressurization resulting from the j escape of high energy fluid at the postulated pipe break location assuming complete pipe severance and axial separation. The encapsulation sleeve will be provided with open vent and drain pipe nipples which extend beyond-the pipe insulation as a means of monitoring the encapsulated pipe section for any leaks which might develop in service, b The design of the encapsulation sleeve will permit either its removal by machinery or flame cutting techniques for the replacement of encapsulated pipe section in the event leaks develop which require repair or replacement of the pipe. n. The guard pipe restraint system will be designed, constructed ) and erected as a quality and seismic Class 1 item, f Guard pipe restraints will be designed and constructed in si accordance with ANSI B31.7 Class 2 as noted below: (a) Material mill test reports giving chemical and j physical data will be provided. l (b) Radiographic examination will be made of all t, possible subassembly welds. l. ' l-(c) Dye penetrant or magnetic particle inspection will. be made of each pass of installation welds and l welds which could not be radiographed during such lu assembly. l7 l : i > 12

~ t' ~ 1 i 'n .) (d) Stress levels will meet the requirements for o emergency conditions. J' (e) Only qualified and approved welding procedures will be used in fabrication and installation. (f) Only code materials will be used. (- t In addition, strict control will be exercised over all weld rod in accordance with documented standards. 2. Pipe Whip Restraints f Postulated severance at the terminal anchor on the main steam and feedwater lines causes severe pipe whip. To prevent pipe whipping, attachments will be made to the existing penetration [1 sleeve. (See figure 6.) The restraint will be attached to the containment sleeve, which is a structural element rather than a pressure part of the containment in the Rancho Seco Unit 1 design. t' These sleeves are designed to accept the load of such failure in the main steam and feedwater. The restraint is attached j to the straight pipe outside the containment opening by means L. of a ring. (b) Protective provisions are described on table 1. ~ (c)&(d) Separation of redundant features was taken into account as described in the answer to Question 11 which outlines the steps involved in the analysis. Where adequate separation was not provided the solutions b are identified in table 1. {. (e) Typical pipe whip restraints are described in the answer to item 11. L; Restraints will be used on two systems: The main steam and main feedwater lines. Locations of these restraints are shown on [s figures 1 and 2. 1 E \\t ,L f U l 1 13 j [j lt

I %) QUESTION 6. The procedures that will be ~ used to evaluate the ^f structural-adequacy of Category I structures and to design new seismic Category I structures should be provided including: 1 ~; (a) The method of valuating stresses, e.g., the work-ing stress method and/or the ultimate strength method that will be used; jr I (b) The allowable design stresses and/or strains; and (c) The load factors and the load combinations. ANSWER: 'I (a) The method of evaluating stresses for existing structures will be accomplished by the use of the ultimate strength method. a l (b) Stresses are as follows: y-- 1 Concrete ( (a) Auxiliary Building above grade f'c=3000 psi j (b) Auxiliary Building below grade f'c=4000 psi L, (c) Strains limited to 0.003 in/in (d) Reinforcing steel f y=60,000 psi 2 Structural Steel q, I (a) ASIM A-36 Steel fy=36,000 psi -("i-(c) The load combinations and load factors are as follows: .l. 1 C 0 2 (1 0.05) D + 1.0R 2 C 2 (1 0.05) D + 1.OR + 1.0E' c = Required capacity to resist factored loads d = Dead load of structure, piping and equipment E' = Design bases earthquake

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R = Force or pressure on structure due to rupture of any one pipe. t-D'l i u V 14-O

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i! Yield capacity reduction factors: -i 0 = 0.90 for concrete in flexure 0 = 0.85_for tension, shear, bond, and anchorage in concrete 4 0 = 0.75 for spirally reinforced concrete compression members I' 0 = 0.70 for tied compression members ~l. 0 = 0.90 for fabricated structural steel .r-

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0 = 0.90 for reinforcing steel in direct tension 0 = 0.90 for welded or mechanical splices of reinforcing steel I O = 0.85 for lap splices of reinforcing steel l !_ c QUESTION 7. The structural design loads, including the pressure and temperature transients, the dead, live and equipment loads; and the pipe and equipment static, thermal, and dynamic reactions should be provided. ANSWER: The main steam lines and feedwater lines will be restrained at the containment penetration. 1. The design loads from the main steam line reaction is 1200 kips and from the feedwater line reaction is 671 kips. r

['

QUESTION 8. Seismic Category I structural elements such as floors, ,( interior walls, exterior walls, building penetrations l, and the buildings as a whole should be analyzed for eventual reversal of loads due to the postulated

p accident.

'l ANSWER: I .1 Seismic category I structural elements such as floors, interior walls, exterior walls, building penetrations and the building as a whole were j analyzed for reversal of loads due to a postulated accident using the (, ultimate strength method. .t b 7 '), ,v ( {' 15 m

1 QUESTION 9. If new openings are to be provided in existing structures, the capabilities of the modified structures to carry the design loads should be demonstrated. 7l_

I ANSWER:

In the event new openings are required in the existing. structure the i~ structural elements will be analyzed by the ultimate strength method. The -{. elements will be checked for flexure, longitudinal and transverse shear.

{;

QUESTION 10. Verification that failure of any structere, including

(.

nonseismic Category I structures, caused by the accident, will not cause failure of any other structure in a man-ner to adversely affect: (a) Mitigation of the consequences of the accidents; and

('i (b)

Capability to bring the unit (s) to a cold shutdown condition. ANSWER: 1 No structural failure will result from any postulated pipe rupture. The compartment above the diesel generator rooms structure will be subjected to a pressure of 0.83 lbs per square foot which will not create any struc-tural failures. In the compartments in the Auxiliary Building basement the reinforced concrete walls will be subjected to a pressure of 27.4 lbs per square foot and at this pressure the walls are capable of withstanding this load with no adverse effect.

L

See table 1 for specific areas of interest. .iL t_ i 1 t .L. p i (- - 1,1 t 16 u

-i -QUESTION 11. Verification that rupture of a pipe carrying high f" energy fluid will not directly or indirectly result in: ,f (a) Loss of required redundancy in any portion of the protection system (as defined in IEEE-279), i* Class IE electric system (as defined in IEEE-308), engineered safety feature equipment, cable penetra-- tions, or their interconnecting cables required to mitigate the consequences of that accident and place the reactor (s) in a cold shutdown condition; or f- {' (b) Environmentally induced f ailures caused by a leak or rupture of the pipe which would not of itself result in protective action but does disable pro-y tection functions. In this regard, a loss of q redundancy is permitted but a loss of function is not permitted. For such situations plant shutdown j~' is required. 1 ANSWER: A. Each system including piping, valves, equipment, electrical power, control cables and switchgear was evaluated for.high energy line I-f ailures using the Guidelines given in the " General Information L-F.? quired for Consideration of the Ef fects of a Piping Break Outside Concainment." Each line was assumed to fail in LL. ecc 1mity of any equipment required for cold shutdown. Each potential problem ras evaluated on the basis of the guidelines, i.e.: if a failure of a line could damage a piece of equipment and the failure of the 3 line did not result in protective action as defined in item ll.b single failure was not assumed and credit was taken for redundancy. t. B. The areas where rupture of a pipe carrying high energy fluid might ( 4 result directly or indirectly in loss of function of equipment L included in (A) and (B) above have been determined and are listed as i tems 3,4,8,9,10,14,15,16,17,19,20,21,231,23J, 23L, 23M,25,27, and 29 on table 1. The equipment exposed to the pipe rupture accident j i and the action required to avoid damage to this equipment in those areas is also given on table 1. The steps involved in the analysis were as follows: 1 From AEC letter and errata i L_ a. Develop guidelines t (. Ib f, 17 m

r ,1 The following assumptions were made to develop the guide lines for the analysis: s 1 (1)- The most stringent requirement for shutdown was postulated to'be the double ended rupture of one main steam line with a turbine stop valve stuck open on the uneffected line such that both steam generators blow down through the single break. I (2) Loss of offsite power was assumed to occur following the accident. ,[ (3) Cold shutdown capability would be required with one 't - stuck rod. ~ (4) For all line breaks which would result in a reactor trip-single failure of an active component was assumed. For all line breaks which would not result-in a reactor trip cradit was taken for redundancy [ of active and passive components.

lg b.

List of impactees c. List of impactors 2 a. Develop marked up P&I diagrams of impactees and impactors. l~ b. Develop list of electrical equipment required for cold ' shutdown. 3 Using information from Item 2, develop general arrangement drafts

t from electrical and piping area drawings.

4 Transfer information from 3 in simplified form to vellum using only one elevation and general arrangement as a guide. 5 Make overlays of impactees and impactors. I 6 Using guidelines, overlays and drawings developed in 3, list -potential problem areas. i L_ 7 Make isometrics of potential problem areas. 8 Investigate isometrics and firm up list of problem areas. 9 Independently consider (assuming reasonable operator action)

f Floodit.g from t ailure of high energy lines or any other source.

a. 3, L. fI - t; o l}! r 18

LJ

r e O< "f b. Impingement-effects where high energy lines have large . c- 'l energy reservoirs. c. (1) Environmental effects on items from 2b above for

i impactees.which have large energy reservoirs and
t high temperatures.

~(2) Compartment pressure effects for impactees which have r- .f large energy reservoirs and high temperatures. 10 Investigate problem areas and' develop solutions. QUESTION 12. Assurance should be provided that the control room ,~ will be habitable and its equipment functional after a steam line or feedwater line break or that the capability for shutdown and cooldown of the unit (s) will be available in another hatitable area. g, ANSWER: f There are five paths that steam could travel through the Auxiliary Building to reach the control room as follows: P h Main steam line break'which is,outside of the Auxiliary Building. Path 1 Steam enters Auxiliary Building at Ele.v. 40'. Steam must travel 212 lineal feet and must pass through 5 doors. Path 2 Steam enters Auxiliary Building at Elev. 20'. Steam must travel 138 lineal feet and must enter stairway No. 2 at 1 ~, Elev. 20' and pass through 5 doors. it Path 3 Steam enters Auxiliary Building at grade. Steam must travel 188 lineal feet and must enter stairway No. 2 and pass through 7 doors. Auxiliary steam line break which is in the basement at Elev. -20 feet. ~ Path 4 Steam must travel up stairway No. 6 to grade, down the corridor to stairway No. 2 to Elev. 40'. Must travel 282 lineal feet and pasc through 4 doors. ,c Path 5 Steam must travel up stairway No. 5 to grade and enter stair-way No. 2 exit at Elev. 40 feet, a distance of 215 lineal feet and pass through 5 doors. r The emergency air conditioning intake is located on the roof of the Auxiliary i j; Building on the South side and steam must travel a distance of 224 feet .l before entering the intake. The normal control room air conditioner intake. ' 7 -d .r} 19 oJ

q_ i ~ is located on the East end of the Auxiliary Building. Intakes )- will not be affected by steam from any high energy line. 4 Considering the large volume of the building in which the steam can dissipate before reaching the control room, the control room would be habitable'and the equipment would remain functional.

(

QUESTION 13 Environmental qualification should be demonstrated by test for that electrical equipment required to func-tion in the steam-air environment resulting'from a high energy fluid line break. The information required [ for our review should include the following: 't (a) Identification of all electrical equipment neces-sary to meet requirements of 11 above. The time - after the accident in which they are required to operate should be given. ~ (b) The test conditions and the results of test data showing that the systems will perform their intended function in the environment resulting from the postulated accident and time interval of the accident. Environmental conditions used for the tests should be selected from a conservative evaluation of accident conditions. (c) The results of a study of steam systems identify-ing locations where barriers will be required to i prevent steam jet impingement from disabling a t protection system. The design criteria for the barriers should be stated and the capability of r the equipment to survive within the protected environment should be described. s (d) an evaluation of the capability for safety related electrical equipment in the control room to function in the environment that may exist following a pipe break accident should be pro- ~ vided. Environmental conditions used for the evaluation should be selected from conservative calculations of accident conditions. I;i (e) An evaluation to assure that the onsite power i distribution system and onsite sources (diesels and batteries) will remain operable throughout the event. ANSWER: ' l' (a) The following electrical equipment is required to mitigate the consequences of the steam line break accident and to place the 7 reactor in a cold shutdown condition: a -[T 20 L

C Equipment: P-236 Make-up pump P-238A&B HPI pumps P-261A&B DH pumps I P-472A&B NSRW pumps P-482A&B NSCW pumps c G-886A&B Diesel gen., including local'and remote controls j-A-544A-D Diesel gen, room exhaust fans ( P-888A-D Diesel gen. fuel oil pumps '~ P-319 Motor Driven Aux. FW pump FV-30801 Steam to Aux FW pump P-318 A-529A-E Emerg. operating pump room coolers P-705A&B Boric acid pumps - Class I batteries & buses d-c & a-c - Nuclear services buses U-545 Control room Emer. air conditioner t-Instrumentation: }'- - Source range detectors - Rod position indication TE-21023A, 24B R.C. Inlet temp. PT-21092, 38, 40, 42, 43, 39, 37 RC pressure LT-20503A, B&C Pressurizer level LT-20504C&D Steam generator' level ~' (b) Item 21 table 1 is the only case where electrical equipment jg will be subjected to custained temperature. Electrical equip-ment required to place the reactor in a cold shutdown condition that may be exposed to a steam environment is the following: j P-236 MU pump motor P-238 A&B HPI pump motors P-261 A&E 2H pump motors P-705 A&B boric acid pump motors i The motors for the MU pump, HPI pumps, and DH pumps have class F insulation systems which can withstand a total continuous tem-

g perature of 130 C (40 C ambient, 90 C temperature rise).

In jl - addition, these motors are provided with weather protected type I i enclosures. The motors for the boric acid pumps have Class B insulation systems which can withstand a total continuous tem-perature of 125.C (40 C ambient, 85 C temperature rise). These motors have totally enclosed fan cooled enclosures. Refer to 9 item 21 on table I and to. answer to question 20. For this case ~ a 6 inch auxiliary steam line with a crack equal to the cross sectional area was assumed to break. This break would not j require the operator to shut down. If the above mentioned pumps j..

a c

fl. 21 L ~

T~ l' are running they may be damaged. However the redundant pump ~ would not be running and would not be damaged if the building was heated to 212 F and would be available should the operator elect to shutdown. Alarms are provided in the control room which i, would warn the operator of a failure of the auxiliary steam line. These alarms would be low auxiliary steam pressure'and high Auxiliary Building-pressure. p (c) Refer to items 8, 9,14, 15, 16, 17, 19, 20 and 23J on table 1 on the pipe break problem areas tabulation. (d) As explained in answer to question 12 it is not credible that steam ejecting from postulated ruptured pipes will enter the control room and affect safety related equipment. (e) Steam ejecting from postulated ruptured pipes will not enter the electrical switchgear rooms and the battery rooms. (Refer to table 1 item 3). Steam from postulated pipe failures will not enter the diesel generator rooms (Refer to table 1 item 25). i QUESTION 14 Design diagrams and drawings of the steam and feedwater lines including branch lines showing the routing from containment to the turbine building should be pro-vided. The drawings should show elevations and include the location relative to the piping runs of safety i related equipment including ventilation equipment,.- intakes, and ducts. i i ANSWER: i 1. The following drawings were previously provided for AEC review. A. Mcrked up P&I diagrams with impactors and impactees. B. Overlays

I, LINES CONSIDERED IMPACTEE L.

1. Main FW, steam & 6. NSRW & CR AC aux. stm. ,[ 7. Emer FW, DG & NSCW 2. Cond., LD, aux. bir FW & N2 8. DH system 3. HPI & MU pump disc. 9. Elec 4. DH disc. 10. DH & MU Suct from BWST & BA pump 5. Emer. FW I 'l 22 'l-

b OVERLAYS MADE & INVESTIGATED 1 on 6,7,8,9,10 2 en 6,7,8,9,10 3 on 8,9,10

i '

4 oc 8,9,10 J, 5 on 7

r~

C. Isometrics of potential problen areas. 2. The following drawings have been submitted previously and will be sub-mitted with the final report; plan and cross sections of area between 7' - spent fuel building and reactor building (area of concern for failure of main steam and main feedwater affecting equipment required for cold shutdown). QUESTION 15. A discussion should be provided of the potential for flooding of safety related equipment in the event of failure of a feedwater line or any other line carrying i high energy fluid. y ANSWER: In the event of the failure of the feedwater line, in the area, bounded by the Reactor Building, Fuel Storage Building, the transfer tube struc-ture and the Auxiliary Building, the area will be drained by the utiliza-tion of the tendon access gallery. The gallery can store 175,830 gallons of water. An overflow will be provided by utilizing a tendon gallery shaft to drain the excess of 175,830 gallons to the plant drainage sys-tem. The tendon gallery will be pumped dry using the existing pump. Con-crete curbs will be installed to protect the Auxiliary Building from flood-ing through the doors. The louvered wall at the diesel generator room will be replaced with a wall to prevent flooding of the diesel generators. In the event of a tank rupture in the yard area north of the Reactor i-Building the existing plant drainage system is adequate to drain the area without build-up of water which could endanger class I systems. See l' table 1 for specific areas of interest. u QUESTION 16. A description should be provided of the quality control and inspection programs that will be required or have

j_

been utilized for piping systems outside containment. 'l ANSWER: L. All piping systems have been fabricated by organizations with quality assurance programs meeting the requirements of 10 CRF 50 Appendix B. i_ The piping systems 2h inch and larger were fabricated by either of l F' u_ IEl 23 LL)

. e-two manufacturers in the Los Angeles area. The 2 inch and smaller pipe was fabricated by the site erection contractor who also performed all field erection work and assembly welding. All three organizations have been surveyed by ASME and are current holders of "N" stamp authorization. E' Prior to contract award these contractors were audited by the SMUD Quality Assurance Director. Subsequent audits by him have verified performance in accordance with their respective manuals. r-All pipe and fittings have been procured under requirements for furnishing heat numbers and material chemical and physical test results and non-destructive examination, as required to meet the project specifications. I_ These specifications incorporate all requirements of ANSI B31.7. Throughout all of the fabrication and site erection work control has been exercised over all welding operations. Strict weld rod control has been observed and all welding has been performed in accordance with welding procedures approved by the engineer. The engineer specified the procedure used for each field welded joint. p. Non-destructive examination of all welds has been in accordance with ANSI B31.7 for the classification of system involved. r' i,_ Prior to the acceptance of piping systems they are subjected to all require-ments of the SMUD Quality Assurance Program. This includes complete (~ documentation of non-conforming items and disposition in accordance with the Manual. Assurance of all Code requirements is met by the State of

  • g" California who are performing site inspection.

QUESTION 17. If leak detection equipment is to be used in the proposed modifications, a discussion of its capa-bilities should be provided. ANSWER: b, No new leak detection system is required as a result of the analysis. Although no credit was taken in the analysis, the main steam line failure protection as described in section FSAR 14.2, would reduce the blowdown for failures of the main steam lines and branches and main feedwater failures between the isolation valves and the Reactor Building. {. L. g t. g, bJ ~' 24

4 - g. j i-QUESTION 18. A summary should be provided of the emergency procedures

r-that would be followed after a pipe break accident,

}" including the automatic and r.nual ooerations required to place the reactor unit (r) in a cold shutdown condi-t ion. The estimated tim s -following the accident for

p.

all equipment and perse.inel ooerational actions should '[- be included in the procedure summary. I j ANSWER: The worst pipe break accident considered is the double ended rupture of a 36 inch steam line between the steam generator and the steam stop valves with concurrent failure of one turbine stop valve to close in the unbroken steam line causing the blowdown of both steam generators. The following procedure will be employed in this incident. The response of the reactor coolant system to this accident is shown graphically on sSAR Figure 14.2-1, sheets 3 and 4. II' '{. About 15 seconds after the accident, the safety features systems actuate; the reactor is tripped. The systems are secured as soon as the control !r-room operator realizes there is a steam line break and not a LOCA. The j~ operator secures appropriate pumps after the accident. The high pressure injection pumps will be secured to prevent the reactor coolant system from going solid and causing undesirable pressure transients. ~ L. The reactor coolant pumps are tripped at the start of the accident because it is assumed that off-site power is lost at that time. Automatic feed-water isolation on low steam line pressure causes the steam generators to blow dry. At one minute after the accident, the plant is in the following condition: Average RC temperature 540 F RC Pressure Saturation Steam Generators Empty Control Rods Inserted Both diesel generators are available to supply the power to both auxiliary r-feed pumps. Since the steam generators are dry no heat can be removed from '~ the system. The operator will start the motor driven auxiliary feed pump as soon as possible to remove energy from the reactor coolant system caused by the high decay heat rate immediately after shutdown. A small feed rate ,m-is used until natural convection begins and the operator determines the cool-down rate. Cooldown by this means, with the steam venting to the atmosphere

{"

through the break, continues until reactot coolant temperature reaches 280 F. When reactor coolant reaches 280 F, cooldown is shifted to the decay heat removal system. a_ il7 e 25

L;

r' t. During cooldown the makeup pump is operated as necessary to add borated water to the system to maintain proper level in the pressurizer. Adding water from the BWST and concentrated boric acid from tha boric acid 4i pumps directly to the suction of the HPI or MU pump (required only if a stuck rod is assumed) as the plant cools down insures sufficient neg-l ative reactivity to maintain at least a 1 percent AK/K subcritical margin. t.. Once the reactor is being cooled by the decay heat removal system, cool-I' down is continued to less than 200 F, and the reactor is maintained in a cold shutdown condition. A discussion of a steam line break which does not rupture piping on the { other steam line is given in subsection 14.2 of the FSAR. QUESTION 19. A description should be provided of the seismic and quality classification of the high energy fluid piping systems including the steam and feedwater pioing that run near structures, systems, or comoonents important [" to safety. I ANSWER: Operating { Temo/ Pres Seismic / Quality F / psia f 1. Main steam lines and branches out to 1/1 595/925 the first remote operated valve. l 2. Main feedwater between isolation valves 1/1 464/1000 [ and Reactor Building. 3. Main feedvater upstream of the iscla-2/2 464/1000 tion valves. 4. Auxiliary Boiler F.W. pump discharge. 2/2 212/350 t. IL Ihe above list does not include project class of lines considered for impingement, temperature and flooding or list of high energy lines con-sidered which will not affect equipment required for cold shutdown. l I l IL. 26

J

r* L

(
s: i..

l,, QUESTION 20. A description should be provided of the assumotions,. methods, and resultsaof analyses, including steam ,(. ~ l generator blowdown, used to calculate the pressure and temperature transients in compartments, pipe tunnels, intermediate buildings, and the turbine

(

building following a pipe rupture in these areas. The equipment assumed to function in the analyses should be identified and the capability of systems ll required to function to meet a single active com-ponent failure should be described.

j~

ANSWER: r' Analysis of compartment pressure or temperature transients were restricted

{

to those areas where a pipe failure could damage equipment required for cold shutdoun. These areas are identified in table 1 as items 26, 27 and 29. The only case where equipment could be affected by a pipe

l break is item 27 which is a postulated failure of a 6 inch auxiliary steam line in the Auxiliary Building.

Failure of this line would not result in protective action as defined in ll(b) such that failure of an active com- .i ponent is not required for the analysis. Electrical equipment required d _. for cold shutdown which may be affected by impingement are identified in schedule item 3 and temperature affects were investigated in table 1 item 21. This steam line is not required for normal operation and olant shutdown would not be required if the line is isolated outside of the Auxiliary

f.,

Building. Electrical equipment in the Auxiliary Building required for cold shutdown would not be damaged by the resulting 212 F temperature except in the case of a high pressure injection, low pressure injection or boric acid pump unless the motor was running for an extended period of [ time in this environment. If a motor did fail the redundant pump would be-used to shutdown. t-j The energy releases used for the compartment pressure analysis were critical crack size as defined by question item 2 for table 1 item 26, - ', ~ double ended rupture of the main steam or feedwater line for table 1 item 29 and a crack equal to the area of the pipe for table 1 iten 27. L All postulated failures were analyzed for these effects on structures and in no case was a secondary missile (spelled concrete, doors etc) il generated with enough energy to affect the equipment required for cold shutdown. ]-% 7 d' 27 p

e- ' f, s. s, QUESTION 21. A description should be provided of the methods or s. analyses performed to demonstrate that there will be no adverse effects'on the primary and/or secondary containment structures due to a oipe rupture outside these structures. .i. ANSWER: g Main steam and feedwater lines will be restrained to prevent impact on the secondary containment structures. l See table 1 for specific areas of interest. . t. 41 4 4 i t (_. l' . L. . 1

Ls e

b .m

t:

28 L

p ._.__m. i 7 .J i l ITEM PRO 1 + 1. SteamLinetd Breaking NS i 2. Main Feedwat -4 Lines Damag>. Stacks l-3. Aur.iliary SG Indication e i switchgear g 4. Steam to Ems 3 Affecting Po it 5. Letdown Lins Raw Water Lft 6. Letdown Aff@ L and from R.V i i l 7. Letdown AffG From BW Stog l 8. Letdown Line Power to HPI Pump and Fan Make-Up Pump 9. Letdown Line Both Boric A 10. Aux. Boiler Power to Bot PRELIMINARY - ALL PLANS SU1 e

q TABLE 1 . POSTULATED PIPE FAILURE OUTSIDE OF CONTAINMENT STRUCTURE EM TYPE OF ANALYSIS SOLUTION Emsr. F.W. Pump Pipe Whip & ImpinFement No Solution Required for Pipe l Law Water Piping Whip. High Stress Points and Terminal Ends are Not Near Small Piping - Main Header is Larger and has Thicker Uall than Steam Line - Small Piping to be Supported for Impingement pr and Main Steam Pipe Whip & Impingement Pipe Whip - Restrain Terminal kg Diesel Generator Critical Crack Size Only Ends & Install Guard Pipe Restraint to be Considered on MS & FW - Install Additional Supports for Impingement (See Item 23) cm Affscting CRD Impingement and Temperature Line to be Abandoned and Source-id electrical to come from Line 61250-6" poma

r. F.W. Pump Pipe Whip, Impingement and Reroute Cable in Uneffected Area rar to NSCW Pump Temperature Affecting NS Impingement Install Additional Supports for u

Impingement

ting NSRW to Impingement No Solution Req'd. Letdown is

, Emer. Coolers 2 " w/.375 in, wall & NSRW is 10" w/.365 in wall. Calculations Not Rea'd. pting Pump Suction Impingement Existing Hangers Are Adequate kg2 Tank to Prevent Damage Affseting 'B' Impingement and Temperat.ure Encase Letdown Line Pump, Mskeup p cud 'A' Power to Affseting Pc. r to Impingement and Item 8 also solves 9 Sid Pumpa Temperature .W._ Lins Affecting' Impingement and Pipe Whip No Problem. Conduit is 10' Apart ._ B:ric Acid Pumps at all points JECT TO FURTHER REFINEMENT BY DESIGN OR ANALYSIS l l t i

ITEM P1 11. Letdown Lini Pump Disc L-Suction Lini 12. Aux. Boiler of Boric Ac' 13. R.C. Pump @ Decay Heat ' 14. HPI and Maki Affecting ' Pump and 'B 15. HPI and Mab Affecting ' Make-Up Pum: Pump 16. HPI and Mak Affecting B and 'B' Pow 17. HPI and Mak Affecting P Acid Pumps 18. HPI and Mak Breaking Din Acid Pumps 19. D.H. Discha to Both Bor j 20. 'A' D.H. Di Power to DB l' 21. Aux. Stm Li I Affecting E 22. Any Line Br Both Decay PRELIMINARY - ALL PLANS S -1

) TABLE 1 POSTULATED PIPE FAILURE OUTSIDE OF CONTAINMENT STRUCTURE WBLEM . TYPE OF ANALYSIS SOLtiTION i Affecting Boric Acid Impingement Add Supports to Boric Acid 3 ; & HPI Pump Line for Impingement !3 9.W. Breaking Discharge Pipe Whip and 3" Line Impingement will d Pumpo Impingement not Effect 2" Line. Restrain F.W. line for pipe whip protection tal Supply Breaking Impingement Install Supports on D.H iucticn Line to Withstand Impingement t-Up Pump Discharge Impingement Remove Power Cable from Tray and L' Powsr to Make-Up Install in Pipe Thru MU Pump Room P war to HPI Pump t-Up Pump Discharge Impingement Encase 2-4" Lines 23822 & 23620 L' cr 'B' Power to for 25' from MU Pump Room Wall i and 'B' to HPI t-Up Pump Discharge Impingement See Item 15 for Fix >wsr to Make-Up Pump er to HPI Pump e-Up Pump Discharge Impingement See Item 15 for Fix >wir to Both Boric t-Up Pump Discharge Impingement Install Hangers on Boric Acid ,sharge of Both Boric Line for Impingement rgs Affseting Power Impingement Reroute Power for Pump 'B' Lc Acid Pumps from affected area. l hchargsAffecting'B' Impingement Encase Cable Tray M34B1 in 3/8"

Pump cud Room Fan Thick Formed Channel for Approx.

15 ft. - Encase Cable Tray M39AD8 with Steel Plate Brsck in Aux. Bldg. Temperature .o Solution Required - Motors Good

cc. Fquipment for 212 F Continuous not Running.

Yking-cud Flooding Flooding - Investigated Large No Solution Required k5:tPumps Sources of Water and Operator Would Have Adequate Warning IJBJECT TO FURTHER REFINEMENT BY DESIGN OR ANALYSIS e - -1 .I

l i ) ITEM PRO 23. Main Steam and Main'F.W. and A. Breaking N Removal Co Surge Tank B. Breaking N Dic.sel Gen C. Breaking E from BWST Connection D. Creaking Pump Sucti f E. Breaking ' A' & ' B ' ' l I F. Breaking i from Undes. l G. Breaking Return to. H. Affecting' to NSCW % I.Affectingd and Indicat' RoomforDil J. Affecting' Power to Steam In1 l K. Damaging Water Ta-L. Affecting Cableto( Grounding M. Affecting' Neutral @ o 9 r ' [ PRELIMINARY

TABLE 1 }' POSTULATED PIPE FAILURE OUTSIDE OF CONTAINMENT STRUCTURE BLEM TYPE OF ' ANALYSIS SOLUTION Brcn:hes and . Pipe Whip Install Guard Pipe Restraint over br:nthao in Area 7.. -Portions of Both Lines & Restraints at' Terminal Ends BCW to' cnd from DH ' Impingement-Add Hangers as Required plcre and Branch to SRW to and from Impingement Add Hangers as Required arctors for both systems PI'& DH Pump Suction Impingement Add Supports as Required Warming Pump Disc pcnt Fuel Cooling Impingement' Line is not close enough on from BWST to be Affected by Impingement mer. F.W. Pump Disc. Impingement Reroute Loop A - Provide Supports for Loop B iesel F.O. Supply Impingement No Impingement Effects ground Tank Resel F.O. Overflow Impingement No Impingement Effects Underground Tank Trcy cud Power Cable Impingement & Temperature Reroute Cable (See Item 4) mp 'A' sy Containing Control Impingement & Temperature Reroute Cable in encased on Cable to the Control Conduit cel Gsn. A Conduit Conc *ning Impingement & Temperature Reroute through duct bank hsr.. F.W. Pump Turbine to manhole near pump it Mav. londsnscte and Borated ' Impingement No Problem with Critical Crack Impingement Effects Conduit Containing Impingement &' Temperature-Reroute See Item M liocal Gen.' Neutral Recictors 'A' &-'B' Diccol Generator. Impingement & Temperature Move Resistors munding Resistors A&B o \\ d.L PLANS SUBJECT TO FURTHER REFINEMENT BY DESIGN OR ANALYSIS ' i

r = y ITEM PROB.l J 24. Flooding of Die GearRooms Room, Aux. Build Between Fuel Sts Building 25. Steam Effects oa and Electrical - Generator Rooms Line or F.W. Liq Between Fuel Sto; Building 26. Main Steam Line Line Pressurizi Above Diesel G 27. Aux. Steam Line Aux. Building 28. Rupture of Larg 450,000 Gas Capy J 29. Main Steam Line and Condensate-Breaking in Turb t 0 i i t e i 30. Faili e of One ( 'nd Failure of if to Close. PRELIMINARY - ALL PLANS SUBJE ?,

m % e TABLE 1 POSTULATED PIPE FAILURE OUTSIDE OF CONTAINMENT STRUCTURE H TYPE OF ANALYSIS SOLUTION $ G:ncrotor Flooding - Worst Case F.W. Provide Adequate Drain to ing, Switch-Line Break Tendon Gallery Rai e Doors l,ina Brock and Install l' High Curbs Laga & R: actor Gngina Intake Temperature & Moisture Close East Wall of Rooms and uip in Diesel Provide Cooling and Combustion rom Storm Air from Roof at Elevation 60 ft. Breck 'ago cnd Reactor sanch or F.W. Pressure - Critical Crack No Remedial Work Required - Compart-Compartment Gize Calculation for Either ment is vented and will not fail. rotor Room Line 'ailuro in Pressure and Temperature - No Solution Required - Steam will Failure was Assumed to be a vent out thru Normal Vent Ducts Crack Equal to the Pipe Cross and Temperatures will not Affect Sectional Area Elec. Equipment stor:gn Tanks Flooding or Undermining No Solution Required - Water Level ity Redundant Equipment Required Will not be High Enough to Effect for Shutdown Shutdown Equipment. All tanks are Seismic Category I 'r Main F.W. Flooding No Solution Required - Water will na & Brcnches all be Contained in Condenser Pit. na Building Temperature No Solution Required - Steam will be Exhcusted to Atmosphere and Cannot Reach Equipment Required for Cold Shutdown Pressure No Solution Required - Calculation Was Not Made but Vent Areas are Large and Exterior Panels will relieve pressure. Panels are not heavy Enough to Cause Damage to Equipment Required for Shutdown dn Stccm Line Inability to Supply Emergency Add Second Motor Driven F.W. Pump rbina Stop Valve Feedwater if the Motor Driven Capability with Power Supplied F.W. Pump is Inoperative from Diesel Generator 'B' E TO FURTHER REFINEMENT BY DESIGN OR ANALYSIS s 1

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