ML20140G470

Forwards 10CFR50.59 Evaluation of MSLB Reanalysis & Effect on Isolation Valve Cubicle & Rcb

ML20140G470
Person / Time
Site:	South Texas
Issue date:	06/13/1997
From:	Alexion T NRC (Affiliation Not Assigned)
To:	NRC
References
TAC-M98914, TAC-M98915, NUDOCS 9706160259
Download: ML20140G470 (79)

Text

{{#Wiki_filter:. l Juns 13, 1997 MEMO T0: PD IV-1 File FROM: Tom Alexion ORIG SIGNED BY: Project Directorate IV-1 Division of Reactor Projects III/IY Office of Nuclear Reactor Regulation

SUBJECT:

LICENSEE'S 10 CFR 50.59 EVALUATION OF MAIN STEAM LINE BREAK (MSLB) REANALYSIS AND EFFECT ON ISOLATION VALVE CUBICLE i (IVC) AND REACTOR CONTAINMENT BUILDING (RCB) (TAC NOS. M98914 AND M98915) I requested the licensee to provide the above subject document. The purpose of (his memo is to place this information in the public document room. Docket Nos. 50-498 and 50-499

Attachment:

As stated DISTRIBUTION: Docket File PUBLIC (PDR) WBeckner TAlexion Document Name: STP98914.00C PD4 l b N'PD4-1 o pA f 0FC NAME TAlexionk WBeckner DATE b/O /97 b/13/97 COPY [YEd/NO YES/NO 0FFICIAL RECORD COPY l l \\ t NRC Hlf CENTER COPY g,na

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l p or r UNITED STATES l lf j NUCLEAR REEULATORY COMMISSION l o 's WASHINGTON. D.C. soseH001 5, *... + g/ i June 13,1997 HEMO TO: PD IV-1 File A FROM: / yam Alexion Project uirectorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

SUBJECT:

LICENSEE'S 10 CFR 50.59 EVALUATION OF MAIN STEAM LINE BREAK (MSLB) REANALYSIS AND EFFECT ON ISOLATION VALVE CUBICLE (IVC) AND REACTOR CONTAINMENT P,UILDING (RCB) (TAC N05. M98914 AND M98915) i I requested the licensee to provide the above subject document. The purpose of this memo is to place this information in the public document room. Docket Nos. 50-498 and 50-499

Attachment:

As stated l l l l l l l l l

r i Plant Operations Review Committee OPAP01-ZA-0104 i Rev.O j l Page 10 of 10 PORC Review Cover Sheet OPAP01-ZA-0104-1 (Page 1 of 1) (Sample) Originating Document No. 054b U'OOl3 Revision No. O ()J. V(G 2. hytK_ Mth oh & his h l'i% TITLE brcb (> M 2t/c, 61 $6 ) The PORC has reviewed this item and has determined that (eheck as appropriate): ) It does oes NOT involve an UNREVIEWED SAFETY QUESTION. It .does does NOT adversely impact plant nuclear safety, d it es does NOT adversely impact the health and safety of plant personnel or the public. It does does NOT require further review by the Plant Mgr, the NSRB, or other individuals / groups. Plant Manager NSRB other (specify below) Unit i 2 REMARKS ) i The PORC recommends this item for: APPROVAL DISAPPROVAL OTHER PORC MEETING NO. '00 [ DATE f A\\ Completed by PORC SMretary This form, when completed, SHALL be retained in accordance with the retention requirements of the originating

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OPGP05 ZA-0002 Rev.O P:ge 44 of 50 10CFR50.59 Evaluations Form 1 10CFR50.59 Screening Form (Sarnple) Page 1 of 4 must D UNTT WI DPROCEDURE D PLANT MOD 15'ICATION OECNP .D DCN O UNIT #2 O BOTH D UFSAR ON D OTHER, ORIGINATING DOCUMEN1 NO.Licenisnc Chance Notice 1962 REV. NO.0

SUMMARY

DESCRlirrlON OF CHANGE:UFSAR chantes revisine the analysis of the main steam line break in the TVC and RCB REASON FOR CllANGE ReviIed analysis to support ICO 93 0004. *MSLB Blowdown Model for the TVC PRELIMINARY SCREENING YES NO

1. Does the proposed change represent a change to the Plant Technical Specifications?

D E

2. If an Unreviewed Safety Question known is known to be associated with the subject change, then further screening is not required; refer to IP 1.19Q.

If "Yes refer to IP-1.19Q. Further screening is not required. Does the proposed change represent: YFS NO

3. A change to correct a typographical, editorial or drafting error?

O E

4. A change which is identical to and addressed in its entirety by an existing approved 10CFR$0.59 O

R Screening /USQE7

5. A procedure change in which the format or text changed without changing actions or intent?

O E

6. A spare or replacement part/ component change with an equivalent part/ component?

O Pi (See Section 2.16 for a definition of equivalent) If all answers to the above questions are *NO" perform the final screening and mark N/A in the approval blocks below. If the answer to any question (3) through (6) is "YES" a final screening is not necessary. Sign approval blockt below and discard pages 2 through 4. Provide an explanation / justification and references if any of items (3) through (6) is answered "YES". Prepared by: N/A Originator Date Approved by: Ilb Section Supervisor Date

OPGP05-ZA-0002 Rev.O Pege 45 of 50 10CFR50.59 Evaluations Form i 10CFR50.59 Screening Form (Sample) Page 2 of 4 ORIGINATING DOCUMENT LCN 1962 FINAL SCREERING In response to the questions below,if the change involves something that is not described in the SAR and is not part of the licensing basis as shown by a review of NRC-published documents, then "NO"is appropriate. However, this decision must be clearly documented with adequate technicaljustification. The phrase "not part of licensing basis" implies that the subject matter was not, used by NRC to issue or maintain the operating license or amendments; and is determined by examination of the Licensing Docket and the following (as applicable): Safety Analysis Report Training and Qualification Program Environmental Report Final Environmental Statement Fire Hazards Analysis Repon Safety Evaluation Report Physical Security Plan Standard Review Plan Safeguards Contingency Plan Conespondence Operations Quality Assurance Plan Emergency Plan Previously Approved USQ Evaluations YFs No 1. Does the subject of this review involve a change to the facility as described in the Safety ht; D Analysis Report? 2. Does the subject of this review involve a change to the procedures as described in the D 15, Safety Analysis Repon? 3. Does the subject of this review propose the conduct of tests or experiments not described D t in the Safety Analysis Report? 4. Does the proposed change affect conditions or bases assumed in the Safety Analysis ( D Repon or safety-related functions of equipment / systems, even though the proposed change does not entail any physical change in existing structures, systems, or procedures as described in the SAR7 If any answer is affirmative, complete the screening form and perfonn an Unreviewed Safety Question Evaluation. If all answers are negative, no Unreviewed Safety Question Evaluation is required. All questions require adequate technical justification. W -

OPGP05 ZA-0002 Rev. O Page 4s of so i i 10CFR50.59 Evaluations Form 1 10CFR50.59 Screening Form (Sample) Page 3 of 4 OR]GINATING DOCUMENT:LCN 1962 TECHNICAL JUSTIF]CA710N SHOULD INCLUDE THE SATITY IMPLICA710NS OF THE CHANGE AND OT11D1INFORMATION SUPPORTING ALL ANSWERS. BRIEFING DESCRIPTION AND TECHNICAL JUSTTFICATION OF THE CHANGE: The twonosed chance is a revision to the UFSAR mass and enerry release ster:m line break analysis for the IVC and RCB JCO93 0004 Revision 5 contains additional information concernitin this chance (use additional pages as necessary) Interdiscipline Coordina$on Required? ll$.Yes D No If yes, obtain appropriate concurrence. O Risk and Reliability Analysis D Thermal Hydraulics D Reactor Engr. 1/1//ff KCivil DM h D Elect. O I&C D EQ D Other Prepare by: A'> 1

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'Driginator D' ate 3/3r /fi' Approved by: %as

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OPGP05-ZA-0002 Rev.O Page 47 of 50 e 10CFR50.59 Evaluations Form 1 { 10CFR50.59 Screening Form (Sample) Page 4 of 4 ORIGINATING DOCUMENT LCN 1962 l l The following documents / attributes have been reviewed as part of the 10CFR50.59 final screening process. l Documents Sections Reviewed UFSAR 3.6. A. 6.2.1.3 l Technical Specifications Safety Evaluation Report (SER) 6.2.1.2 Fire Protection (FHAR) Environmental Report (ER) Security Plan Emergency Plan Offsite Dose Calculation Manual (ODOM) Final Environmental Statement Core Operating Limits Report (COLR) Operations Quality Assurance Plan Other See JCO 93 0% Rev 5 Attributer. Check if Reviewed En vironmental Qualification.................................................. X S ei s mic Desig n........................................................... Personnel Radiation Exposure.................................. Mi ssil e Prot ecti on......................................................... Containme nt I nte g rity....................................................... Single Failure Criteria Electrical Separation (RG 1.75)................................................ H ea vy lea d s............................................................. High Energy Line Break Accident Analysis..........................,............ X Control Room Habitability................................................... I nternal Flood ing.......................................................... Pl ant Chemistry........................................................... H u m a n Fact ors........................................................... Probabilistic Safety Assessment................................................ Other NOTE: If Attributes are identified in the originating document this section need not be completed. 4 i c, -_..._n.- - - - ~. -. - n :, m._ r -- w n-u _w =--r

OPGP05-ZA-0002 Rev.O P:ge 48 of 50 10CFR50.59 Evaluations Form 2 Unreviewed Safety Question Evaluation Form (Sample) Pag 61 of 3 Unreviewed Safety Question Evaluation # 95-0013 Revision No. O ORIGINATING DOCUMENT: LCN 1962 REV. NO. O NOTE: Attach 10CFR50.59 Screening Form or License Compliance Forrn to this USQE. NOTE: Use additional sheets as necessary to provide the bases. A.] I Does the subject of this evaluation increase the probability of occurrence of an accident previously evaluated in the Safety Analysis Report? O YES Q NO Bases: The revised UFSAR chantes represent a change to the safety analysis and not a plant or procedural chance that would introduce an accident initiator that would increase the probability of an eccident tweviously evaluated in the Safety Analysis Report. Does the subject of this evaluation increase the consequences of an accident previously evaluated in the Safety Analysis Report? D YES flNO Bases: Th? revised analysis presented in the UFSAR changes demonstrates that all acceptance limits are satisfied Therefore there is no increase in the consecuences of an accident previously evaluated in the Safety Analysis Report !!! Does the subject of this evaluation increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report? D YES (NO Dases: The revised UFSAR chances represent a change to the safety analysis and not a plant or procedural chance that would introduce an accident initiator that would increase the probability of occurrence of a malfunction of eculement important to safety previously evaluated in the Safety Analysis Report. l IV Does the subject of this evaluation increase the consequences of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report? O YES WO Bases: The revised safety analysis presented in the UFSAR changes demonstrates that the acceptance limits of the IVC and RCB structures are not exceeded The analysis does not impact the Eouipment Oualification in these subcompartments. No other components, systems. or structures are impacted by this analysis. Therefore, there is no increase in the consecuences of a malfunction of ecuipment important to safety previously evaluated in the Safety Analysis Report. l 1

OPGP05-ZA-0002 Rev.O Page 49 of 50 10CFR50.59 Evaluations I ] Form 2 Unreviewed Safety Question Evaluation Form (Sample) Page 2 of 3 A.2 I Does the subject of the evaluation create the possibility of an D YES $ NO accident of a different type than any previously evaluated in the Safety Analysis Report? 1 bases: The revised analysis presented in the UFSAR chanres demonstrates that all accentance limits are l satisfied for an accident already addressed in the UFSAR. Therefore, the chante does not increase the l possibility of an accident of a different type than anY previouslY evaluated in the Sarery Analysis Report. 11 Does the subject of this evaluation create the possibility of an C YES 52 NO different type of malfunction than any previously evaluated in the Safety Analysis Repon? Bases: See response to A.21 l l A.3 1 Does the subject of this evaluation reduce the margin of safety D YES % NO l as defined in the basis for any Technical Specincations? l Bases: The revised ana1YSis presented in the UFSAR change demonstrates that all acceptance limits are l satisfied Therefore, there is not a reduction in the margin of safety as defined in the basis for any Technical specifications. i l l l 0 hr 9% % .9 "M & sh *

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OPGP05-ZA-0002 Rev.O P:ge 50 of 50 10CFR50.59 Evaluations Unreviewed Safety Question Evaluation Form (Sample) Page 3 of 3 Form 2 B.

1. X All of the above questions were answered NO; therefore, the originating document does rot involve an Unreviewed Safety Question.

o 2. One or more of the above questions was marked YES; therefore, the originating document involves an Unreviewed Safety Question. The originating document, as presented, shall NOT be implemented without prior approval by the NRC. Provide a recommendation for disposition of the Unreviewed Safety Question below. Refer to IP.1.19Q for processing licensing amendments. Further processing of this form to the PORC, Plant Manager and NSRB is a required. RECOMMENDED DISPOSITION: Approve the USOE and Assoieste UFSAR Chance PREPARED BY: N M Y ~ RIGINATOR Elate b. 3hI/ 9.s' APPROVED BY: .A-- DEPfiENT MANAGER Date I N~ "O M 3 ) 5~ PORC MEETING NO. 'Date MOMm J//9f APPROVED BY: ~ HCAl4T MANAGER Date REMARKS: ),C 93-004. Revision 5 contants additional infromation concerning this evaluation. O 1 i w,. ,e 4 8 k

l l Plant Operations Review Committee OPAP01-ZA-0104 l Rev.O Page 10 of 10 i PORC Review Cover Sheet OPAP01-ZA-0104-1 (Page 1 of 1) (Sample) Originating Document No. UCO O3* M Revision No. 6 _, TITLE MSLB Oludou M.Je! for TVc The PORC has reviewed this item and has determined that (check as appropriate): It does does NOT involve an UNREVIEWED SAFETY QUEST 10N. / It does does NOT adversely impact plant nuclear safety. ) / \\ It pes does NOT adversely impact the health and safety of plant personnel or the public. ff l It does does NOT require erreview by the Plant gr, the NSRD, or other individuals / groups. Plan E. NSRB cther (specify below) Unit 1 2 REMARKS l l The PORC excommends this item for: PROVAL DISAPPROVAL OTHER PORC MEETINO NO. @0 $ Completed by [ DATE-1 I'ORU fecrehry This form, when completed, SHALL be retained in accordance with the retention requirements of the originating [ document. J e Td.E ' M "T5- 'MT ' '- TW C NI

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Justification for Continued Ooerstion OPGP05-ZN-0005 Rev.0 Page 14 of 15 l JCO APPROVAL COVBR SHEET (TYPICALT OPGP05 Z9-0005-01 1 (Page 1 of 1) j JCO Approval Cover Sheet Initiation Date: Expiration Date; (Unit 1) PJlA (Unit 2) uiA

Subject:

M 4 t. G Bt. s w p bott bit ht e JCO No. %-o==4 Revision No. S~ ] Applicable Units: 1 4% Summary: M Sc.o is m \\wo,. Oc4 Arel cd t s bh.t ded., % A m..e.., L A h emiu m 3&s:s k, c '-> refd b h + E d e 6-An' arm // Prepared by/Date:). b, / hkf N bl Cb o -m,. i Concurrence With/Date: ums - < eue, 1//#1 /GEpahh'g'e'r,' Nuclear nsing / / Date Nvc4 oAAAr. / Recommended by PORC Meeting NoJDate: / Approved /Date: _ M e 4> rf'2<, s / J/3//ff V ' Plant Manag Unit! ' Date Appmved/Date: d.

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/ 71/Y Plant Mansk. J' it 2 [ /Date v t Affected Stmeture, System, and Components (SSCs, Unit 1: TdG 4Th% Unit 2: %,1 As Om i f2 o B +tn d u r A 4 Unit 1: Unit 2:

References:

SPR MS-vu f SPR 4 W 44 4 r r 1 LCTS LCTS w.%-Gs-This Form, when completed, SHALL be retained for the life of the plant. l l ..,. _ - _ _ - = ,,., = =

JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. R;visi:n 5 ~. TABLE OF CONTENTS REAS ON FOR REVISION......................................'. 1 EXPI RATION DATE........................................... 2 1.0 ID ENTIFI CATI ON............................................ 2 2.0 O PE RATI O N................................................. 3 3.0 S AFETY FUN CTIO N........................................... 3 4.0 S A FETY A NA LYS IS........................................... 5 4.1 Calculations Reviewed....................................... 6 4.2 Pressure / Temperature Analysis.................................. 6 4.3 Stmetural Analysis of IVC Main Steam Line Break and Results of IVC S h ort-Terrn Eval u ation....................................... 7 4.4 Results of RCB Short-Term Main Steam Line Break Evaluation.......... 8 4.5 Results ofIVC & RCB Long. Term Main Steam Line Break Analysis 8 4.6 Results of IVC & RCB Main Feedwater Line Break Evaluation........... 9 4.7 Results of IVC & RCB EQ P/I' Analysis Evaluation.................. 10 5.0 REQUIRED COMPENSATORY ACTIONS.......................... 10 6.0 CORRECTIVE A CTI ON........................................ 10 7.0 RE PO RTI N G................................................. 11 8.0 REFEREN CES................................................ 12 ATTACIIMENT - A, JCO OPElt.70NAL IMPACT STATEMENT.......... 14 } m__._..,_., . - -.. u_ m _ m m. _ _ _ _ _-,m 99

l a. l l JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. R:vislan 5 REASON FOR REVISION l The purpose of Revision 5 is to close this JCO. The JCO is being closed because the Information contained in Licensing Change Notice 1962 and USQE 95-0013 shows that the MSLB issue does not result in an Unreviewed Safety Question. No further actions are required for this JCO. O l l .c - - o. . -. uw y LY -- 21-- = '

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~ 1 JCO 93 0004, MSLB BLOWPOWN MODEL FOR IVC. Revisi:n 5 EXPIRATION DATE Not applicable for closure. I 1.0 IDENTIFICATION SPR 93-2415 was originated because of concern that the Main Steun Line Break (MSLB) blowdown model for the Isolation Valve Cubicle (IVC) does not account for moisture carry-over from the steam generators; therefore, the blowdown may be non-conservative. This issue was discovered during a review of NRC Information Notice 93-55, " Potential Problems with Main Steamline Break Analysis for Main Steam Vaults / Tunnel." Note that this issue does not affect tae Equipment Qualification (EQ) temperature analysis for equipment located within the IVC. As part of the investigation of SPR 93-2415, Plant Analysis has identified that the mass and energy releases used in other subcompartment pressure / temperature (P/T) analyses are not conservative. The following calculations were reviewed and found to be impacted: NC-7012,

• IVC P/T Analysis MSLB,"

NC-7023. " Main Feedwater Line Break P/r Analysis in IVC," NC-7028. "MSLB Subcompartment Analysis," and NC-7048. " Main Feedwater Line Break Subcompartment Analysis in RCB." In particular, the mass and energy releases used in these analyses did not consider the hot zero power conditions with the MSIVs open. The following non-conservative assumptions were used in the IVC P/r analysis (NC-7012): the initial steam generator pressure was assumed for full power conditions (1100 psia); the limiting initial pressure would occur at zero power conditions (1266 psia including instrument uncertainties); and only steam release from the affected steam generator was considered; backflow steam from the other three steam generators and the main steam piping was not considered. This JCO is being perfomied in accordance with STP procedure OPGP05/Z.N-0005 l " Justification for Continued Operation" Revision 0 3-11-94. l l

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JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. Revist:n 5 i 2.0 OPERATION, The four main steam lines passing through the WC structums are housed in separate cubicles isolated from one another by two foot thick reinforced concrete walls. The main function of the structure is to isolate the main steam lines from one another so that a break in one steam Jine cannot affect adjacent main steam lines. Immediately following a postulated main steam line break in the WC, a very rapid pressurization of the affected WC cubicle would occur. The pressure buildup would be limited by the blowout panels located at the top of the WC. A AP of 0.6910.14 psid causes the IVC blowout panels to give way and limit pressure buildup. However, due to the dynamic nature of the event, the blowout panels cannot relieve pressure quickly enough to avoid a pressure spike substantially in excess of 0.69 psi during the first second following the postulated rupture. Preliminary analysis for this JCO shows that the pressure would peak at 13.1 psid. The RCB operates as a passive structure which protects the safety-related equipment within it. 3.0 SAFETY FUNCTION 10CFR50, Appendix A, General Design Criterion (GDC) 4, Environmental and dynamic effects design bases, discusses the essential safety and regulatory criteria of the IVC and the RCB. GDC 4 states: " Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These stmetures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping." n+ x? a.m.m . uw.s 3 1 - ~ - -

JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. R:vist:n 5 Section 3.6.1, "Piant Design for Protection Against Postulated Piping Failure in Fluid Systems Outside Containment," of the STP Safety Evaluation Repon (SER), and its supplements, also discusses safety and regulatory criteria to be met by the IVC and its subcompanments. Section 3.6.2, " Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," of the STP SER, and its supplements, discusses safety and regulatory criteria to be met by the RCB and its subcompartments. Section 6.2.1.2, "Subcompanment Analysis," of the STP SER, and its supplements, discusses safety and regulatory criteria to be met by various RCB subcompartments. The following criteria were used or considered in the various analyses discussed in Section 4.0 of this JCO: ] ANSI /ANS-56.10-1982, "Subcompartment Pressure and Temperature Transient Analysis in Light Water Reactors," provides guidance for subcompanment analyses including shon-term pressure and temperature transients to which subcompartments will be exposed as a results of postulated line breaks. Also, determination of long-term pressure and temperature transients resulting from both normal and abnormal occurrences are discussed. The standard considers subcompartments located both inside and outside of containment. ANSI /ANS 58.21980 & 1988, " Design Basis For Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture," provides guidance on the design basis for protection of light water reactors from the following potentially adverse effects of a postulated pipe rupture: pipe whip, pipe internal loads, jet impingement, companment. pressurization, environmental conditions, and flooding. 10CFR100 Appendix A, Section VI.(a)(1), discusses engineering design as it applies to a = Safe Shutdown Eanhquake. Specifically, it is stated that *[I]t is permissible to design for strain limits in excess of yield strain in some of these safety-related structures, systems, and components during the Safe Shutdown Eanhquake and under the postulated concurrent conditions, provided that the necessary safety functions are maintained." m _...-zo... . om.i 4 , _ _ = - - ~ - - - = ~ . ~ -, ~

- m JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. R;vist:n 5 General Civil / Structural Design Criteria SA360SQlOO1 governs all stmetural analysis at STP. Table 11 of the Criteria defines the required design load combinations and load amplification factors. 4.0 SAFETY ANALVSIS Plant Analysis and DED performed several analyses and cvaluations as a result of this issue. The following actions were completed for this Safety Analysis: twenty Nuck r Calculations and one Civil Calculation were reviewed for impact; a pressure / temperature analyses were performed for the IVC and the RCB; the existing structural analysis for the IVC was reviewed to determine the impact of the revised pressures; shon-term and long-term main steam line break evaluations and analyses were made for the IVC and the RCB; feedwater line break evaluations were made for the IVC and the RCB; and equipment qualification pressure / temperature evaluations were performed for the IVC and the RCB. 4.1 Calculations Reviewed The following calculations were reviewed with respect to this analysis and found to noJ, be impacted: NC-7007. "MSLB - Containment Pressure / Temperature Analysis," NC-7017. " Mass / Energy Release From SGBD Line Break in IVC," NC-7018. "AFW Turbine Driven Pump Steam Supply PR Analysis," NC-7019. "AFW Line Break P6 Analysis in IVC " NC-7020, "SGBD - P/T Analysis in IVC," NC-7030, "MSLB Forcing Function Analysis for MS1001/MS1002," NC-7038.

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NC-7042, "MSLB Forcing Function Analysis for MS1003/MS10N," NC-7044. "lVC EQ MSLB," NC-7047, "MSLB - Containment Pressure / Temperature Analysis for Split Breaks," NC-7055. " Inflatable Seal Equipment Qualification Temperature," NC-7057, " IVC P/T Analysis in Five-way Restraint Area," w-_- - : - m m...n . m,..; 5 gg g m x - : e:x -- _,:,' x v -:1- : - w w v-meatn r 2 m-- ~c.rr, v r a:m u -,, rr r ~. v =w ~

JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. R3 vision 5 j NC 7061. " IVC Corridor P/T Analysis at El.10' Due to SGBD Line " NC-7062. " IVC Pressurization Following a Break in AFW System at low Power," l NC-7063. " Electrical Equipment Qualification Inside Containment," and NC-7064. " Post-Accident Polar Crane Temperatures." l The following, calculations were reviewed with respect to this analysis and found to be impacted: NC-7012. " IVC P/T Analysis MSLB," NC-7023. " Main Feedwater Line Break P/T Analysis in IVC," NC-7028 "MSLB Subcompanment Analysis," NC-7048. " Main Feedwater Line Break Subcompartment Analysis in RCB," and CC-6251 " IVC Reanalysis." 4.2 Pressure / Temperature Analysis Thermal Hydraulics performed analysis to determine the impact of the revised assumptions on the IVC and RCB P/T analysis. The analysis was performed using the RETRAN-03 computer code to determine the mass and energy releases which were then used by the. GOTHIC computer code to determine the pressure response of the IVC & RCB subcompanments. RETRAN-03 is a RELAP4-type code and meets the criteria specified by ANSI /ANS 56.10-1982 and SER Section 6.2.1.2. The GOTHIC models were benchmarked against the results from Bechtel's COPDA models with acceptable results. The computer a models were revised as necessary to conform with the applicable portions of ANSI /ANS-58.2-1988, " Design Basis For Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture" and ANSI /ANS 56.10-1982, "Subcompartment Pressure and Temperature Transient Analysis in Light Water Reactors." The revised analysis makes the following assumptions which are significantly different from those used in the original analysis: the initial pressure is at 1266 psia which is the saturation pressure at the zero power a temperature of 567 F (plus 7 F for instrument uncertainty); the original analysis used the full power pressure of 1 00 psia; the RETRAN model includes all steam sources including the three unaffected steam generators and the steam headers; the original Bechtel model did not include all l steam sources; and the impact of steam moisture carry over was considered. 1 m 6 l

~ JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. Revisi:n 5 The pressure response of the subcompartments is divided into two distinct phases, the short-term response and the long-term response. The shon term response identifies the initial peak pressure used in the dynamic pressure loading analysis of surrounding stmetures. Short-term analysis includes approximately the first 0.5 seconds of the transient. The fluid phase of the short-term portion of the transient is 100% quality steam because the moisture carry-over from the steam generators does not reach the break location until after the initial pressure peak has been experienced. The second phase (long-term response) identifies the pressure response associated with the j moisture cany-over from the steam generator. For the purposes of these analyses, a 4% l quality steam was assumed as discussed in ANSI /ANS standard 58.2-1980, Appendix E. (This is the ANSI standard referenced in IEN 93 55.) 4.3 Structural Analysis ofIVC Main Steam Line Break and Results ofIVC Short-Term Evaluation Using the results of the short-term IVC pressure analysis, DED evaluated the impac't on the structural analysis and design of the IVC contained in calculation CC-6251. The existing ( analysis is based on calculated pressure peaks at nine different locations (nodes) in the IVC, j ampli0ed by appropriate load factors (as required by General Structural Design Criteria i SQ1001). In comparing the calculated new pressure peaks to the nine peaks used in the current analysis, the maximum pressure in the limiting node (the break node) for the IVC decreases from 14.0 0 psid to 11.5 psid. However, the maximum building stress due to dynamically applied I pressure loads depends on the shape of the pressure versus time function, not just on the magnitude of the peak pressure. This effect is incorporated by use of " dynamic load factors" to amplify the pressure peaks. New dynamic load factors, which are higher than the ones used in the current analysis, have been calculated by Sargent & Lundy. Using the new l dynamic loading factors, it is possible that one local region of the IVC may be slightly overstressed by as much as 10%, however stme: ural integrity of the IVC will not be affected for the following reasons: ) if a localized area is overstressed by 10% in this concrete configuration, the stresses = will be redistributed, and thus the loads would be carried by other structural areas or j members; l 7 i _rm.__.--_ . _ =,.

JCO 93 0004, MSLB BLOWDOWN MODEL FOR IVC. R;vist:n 5 the IVC building design does not consider in situ material strengths. The design strength of the concrete (f/) is 4000 psi whereas actual strength of the concrete is in excess of 6000 psi. Actual strength of Grade 60 reinforcing steel is typically 5 to 10% above the minimum required 60 ksi. Unlike the IVC break node, other IVC nodes experienced increases in peak pressure. However, based on a review of the existing analysis, DED has concluded that the maximum stress in the building is strongly correlated to the maximum pressure in the break node, and the fact that the non-peak pressures have increased is of relatively little significance. Accordingly, the IVC structure will maintain its overall structural integrity as a result of the new pipe rupture loads. Therefore, the revised P/T analysis does not affect continued safe g operation. 4.4 Results of RCB Short Term Main Steam Line Break Evaluation l Results of the pressure analysis on the RCB short-term main steam line break show that the new calculated peak pressures are slightly higher than those originally calculated by Bechtel. This small increase has been evaluated and found to be within the existing margin of the j structure. Therefore, there is no adverse impact on the RCB stmetural analysis resulting from a short-term main steam line break. 4.5 Results of IVC & RCB Long-Term Main Steam Line Break Analysis As discussed in IEN 93-55, moistum canry-over from the steam generators may result in higher peak pressures in the subcompartments. Due to the high pressure in the main steam lines, choked flow would occur in the event of a postulated break. Under choked flow conditions, moisture in the steam lines would decrease the enthalpy of the break fluid but increase the break flow rate. The overall effect is an increase in the energy released from the break because the decrease in enthalpy cannot compete with the increase in break flow rate. The analysis of the long-term response with moisture carry-over for the limiting subcompartments t. hows that the calculated pressures in the IVC & RCB exceeded the peak pressure calculated for the short-term analyses. The following assumptions were used for the long-term analyses using moisture carry-over: u-_- . m.. 3 g --z

JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. Revisi:n 5 steam generator pressure was.ald constant at 1266 psia, with no credit taken for = depressurization of the steam generators; The impacted steam generator depletes its water mass approximately 23 seconds = after the break occurs. Calculation of steam generator water mass deplection time assumes the following: (1) the full mass of water in the steam generator at hot zero power plus 10% for uncertainties, (2) AFW flow is added coincidnet with'the bmak at the runout flow of 1210 GPM, and (3) 4% quality steam; After the mass in the effected steam generator is depleted, the mass and energy release from this generator is significantly reduced. At this time, the MSIVs are also assumed to close. This is conservative because the a MSIV closure is expected to occur at approximately 15 seconds based on a low steam line pressure signal; and The mass flow rate addition from AFW is not sufficient to produce moisture = carry-over. This is based on the fact that the AFW pump mnout flow of 160 lbm/sec is significantly less than the main steam mass flow rate of 1175 lbm/sec at full power conditions. At full power conditions, moisture carry-over is not an issue. The long-term effects of the IVC pressure response for the limiting subcompartment (Node 7) p is less than the peak pressure for the Bechtel analysis. The pressure reponse has been evaluated by DED and found acceptable. For the RCB, the peak pressure differential between i the main containment subcompartment and the break subcompartment is less for the long-term response than the short-term response. Therefore, the long-term effects of a main steam line break in the IVC and RCB are bounded by the short-term effects. 4.6 Results of IVC & RCD Main Feedwater Line Break Evaluation The main feedwater line break analyses for the IVC (NC-7023) and the RCB (NC-7048) were reviewed. All sources of feedwater were considered in the analyses, but full power conditions were assumed. Assuming full power conditions serves to increase the enthalpy of the fluid considered in the break yet decrease the initial assumed pressure. Although the sensitivity of these effects were not evaluated by Bechtel, the results of the current pmssure analyses show i .c--- - _. mmm... . m., 9 r _- - __ = _ m _,, ,__x__w- . =-- --., _ - - 71 g

JCO 93 0004, MSLB BLOWDOWN MODEL FOR IVC. Revisi:n 5 l that the feedwater line break is bounded by the main steam line break. The peak pressures for the full power cases are shown below. Umiting Feedwater une Break Peak umiting Main Steam Une Break Peak Differential Pressure (Current Ardysh) Difhr:r.tla! Pressure (Current Analysis) !VC 4.3 psid (Node 11) 14.0 psid (Node 7) RCB 1.2 psid (Node 7) 13.7 psid (Node 3) 4.7 Results ofIVC & RCB EQ P/T Analysis Evaluation A review of the EQ P/T analyses for the IVC (NC-7044) and the RCB (NCs-7007,7047, 1 7055,7063, and 7064) show that the mass and energy releases calculated by Westinghouse using the LOFTRAN code properly assumed steam from all sources and at the correct initial pressures. if moisture carry-over were considered for EQ analyses, the subcompanment temperatures would be reduced. Therefore, the EQ P/T analyses remain bounding. 5.0 REQUIRED COMPENSATORY ACTIONS There are no compensatoy actions required as a result of this JCO. 6.0 CORRECTIVE ACTION

1. Upon receipt of the final pressure / temperature analyses from Nuclear Fuel and Analysis, DED Civil / Structural Group will reevaluate the IVC structure for the effects of the pressure loading.

2. Nuclear Fuel and Analysis and DED will incorporate the analyses and evaluations discussed in Section 4.0 into the STP design basis.

These actions were completed on March 30,1995. Completion of these corrective actions removed the necessity for this JCO. I i l l EM- -..~.MMM

• M1
• M
• S$229) l

i t JCO 93 0004, MSLB BLOWDOWN MODEL FOR IVC. Revisi:n 5 ) l 7.0 REPORTING This section addresses the need for reporting under 10CFR50.72 and 10CFR50.73 as well as potential 10CFR21 concerns. 1 -10CFR50.72: There are no required reportable concerns pursuant to 10CFR50.72. 10CFR50.73: The IVC structure will maintain its overall structural integrity as a result of the new pipe rupture loads. 10CFR50.73(a)(2)(ii) applies, but no structural system will be degraded or in an unanalyzed condition as a result of the new pipe rupture pressure loading provided. Therefore, this item is not reportable. 10CFR2h There are no known 10CFR21 concerns as a result of this JCO. ,e s-_-- .. m m. .om. g r . ~_ ~

JCO 93-0004, MSLB BLOWDOWN MODEL FOR IVC. R:vist:n 5

8.0 REFERENCES

l

1. 9A010SS1009; " Specification for Controlled Release Roofing for the Isolation Valve Cubicle;" Revision 4.

2. Calculation RC-6548; " Pipe Stress Analysis Calculation"; Revision 6.

i

3. NUREO-0781; " Safety Evaluation Report Related to the Operation of South Texas Project, Units 1 and 2;" April 1986.

4. NUREG-0781; " Safety Evaluation Repon Related to the Operation of South Texas Project, Units 1 and 2;" Supplement 1; September 1986.

5. Station Problem Report 93-2415.

6. NRC Information Notice 93-55; " Potential Problems with Main Steamline Break Analysis for Main Steam Vaults / funnel."

7. Calculation CC-6251; " IVC Reanalysis;" Revision 3.

8. ANSI /ANS-56.101982, "Subcompartment Pressure and Temperature Transient Analysis in Light Water Reactors."

9. ANSI /ANS 58.2-1980, " Design Basis For Protec' tion of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture."

10. ANSI /ANS 58.2-1988, " Design Basis For Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture."

11. Regulatory Guide 1.142, " Safety Related Concrete Structures for Nuclear Power Plants."

12. 5A360SQ1001 General Civil / Structural Design Criteria.

13. B&R TRD 2A700GP003-C; Show Cause Report on Concrete Structures; " Review of Safety-Related Concrete Structures Including Embedments."

l u+ - g

JCO 93*0004, MSLB BLOWDOWN MODEL FOR IVC. Revisi:n 5

14. NUREG-0781, " Safety Evaluation Report Related to the Operation of South Texas

{ Project, Units 1 and 2. l i i .i 8 .h 1 2 h 1 i 4 .? I 1 l V-OWMJCO a %Um.easa46 58#414J c-W

MSLB BLOWDOWN MODEL FOR IVC. Revisi:n 5 JCO 93-0004, ATTACHMENT - A JCO OPERATIONAL IMPACT STATEMENT e JCO No. 93-0004. Revision 3 JCO

Title:

hiSLB Blowdown Model for IVC To maintain the validity of this JCO, Plant Operations must do the following: A, There are no actions required of Plant Operations in order to maintain the 1. validity of this JCO. 9 .c m c... re --...

4 OPGP05-ZN-0004 Rev.I rage 27 or 32 Changes to Licensing Basis Documents and Amendments to the Operating License Data Sheet 1 Licensing Document Change Request (Sample) l Page 1of1 Change Number i9 62 Date s/2A A r t Originator CkA iMLs Dept AJALA / Change Description N x <Amm uu M k.TVC E l Jnitiating Documentation TC O %.-c el/ l 6' USQE Number 36"c613 Unit (s) Affected: Unit I / Ura. Implementation Status: Unit 1 Completion Date i Unit 2 Completion Date 1 Reviewed and Approved by Supervising Engineer, Nuclear Licensing Datt Reviewed and Approved by (ER, UFSAR 2.1, 2.2, 2.3) Manager, Technical Services Department Date l Reviewed by (OQAP changes only) General Manager, Nuclear Assurance Date 1 l This fonn when completed, shall be retained for the life of the plant. -..,,_ -: _ w m., _w_ __ _ u - - -

STFECS UFSAR AFFENDIX 3.6.A Iso 1ATION VALVE CUBICLE SURCOMPART1GET ANA1.YSIS 3.6.A.1 Desien Features. Ne Isolation Valve Cubicle (IVC) is located between the containment and Turbine Generator Buildings (TCBs) on the north side of the containment. Figures 1.2 21 through 1.2-25 provide the plan and elevation views of this area. The IVC consists of four cubicles with each cubicle designed to accommodate equipment and piping pertaining to each of the four trains of the steam feedvater system, thus meeting the train separation criteria. At lover levels (between El. 10 ft-O in, and 34 ft-0 in.) each train has an auxiliary feedvater (AW) pump. Three of the pumps are motor-driven while the fourth is turbine. driven. Watertight doors assure the separability of the auxiliary pump cubicles from one another in the event of flooding of any one of the cubicles due to a pipe break. Main steam (MS) and main feedvater (W) l pipes run through the IVC above El. 34 ft-0 in extending from the Containment penetrations to the fivi.vay bending torsional restraints mounted between tvo walls on the north and of the IVC. The main steam isolation valves (MSIVs), MS safety valves, main feedvater isolation valva (HTIV), etc. are located in l this compartment. A sloped metal roof covers the top of the IVC. The roof will lif t of,f the affected cubicle in the event of a pressure build up due to a pipe break in one of the cubicles. The A W pump cubicles relieve their pressure build up in the event of a AW pipe break through the opening at El. 34 ft-0 in. from whence it is eventually vented to the atmosphere via the roof in the IVC. j 3.6.A.2 Desien Evaluation. T M--s c e;; r r 7:tr;;.. -.1...;; ;;r:- dete--i.;d

ie i'... C0;;m s.nyuces coas. was.-vf....

Je e;: ;;9~ in l P e t.ie.. '. 2.170.h The MS and W piping in this compartment is designed to the break exclusion criteria, stated in Section 3.6.2.1, for those portions of the piping passing through the primary containment and extending to the first pipe whip restraint past the first outside isolation valve. Accordingly, mechanistic pipe breaks are not postulated in the MSIV/MFIV piping. However, to provide an additional level of assurance of operability of safety-related equipment in this compartment, the building structures and safety-related equipment are designed to environmental conditions (pressure temperature and l flooding) that.vould result from a break, equal to one cross-sectional area of the MS and main W piping. Adequate venting is provided to limit the pressurization of the cubicles to below the design pressures of the wall. The following cases were analyzed to datermine the worst environmental conditions for the IVC. I 1. Main steam line break (MSI.8) equivalent to the area of a single area rupture l 2. Main W line break due to a single area break 3. AW line double-ended break in the AW cubicle i 4 Double. ended steam generator (SC) blevdown IIine break in common corridor area l 3.6.A.1 Revision O l,

l STPECS UFSAR In general the calculated maximum pressures resulting from an MSLB are greater than those calculated for the other postulated break types. There is one i exception. A break in the SC blowdown line results in the highest pressura calculated for the common corridor area north of the AW pump cubicles at 2

1. 10 ft.

f 3 e uit (pr le

• Q K

-teru' Ivd ,/ J ~ r "i MA w e calcu ed 's t metho o are 3.6 9, ..A 6 and" . e hg,a FWlle bD W lld h d i The%odalization schem seleoced for the IVC model is shown in Figure 3.6.A-1. The common corridor area is not part of this model. The nodal boundaries have ) l been selected wherever there are flow restrictions (such as grating plat-l forms). As mentioned, the roof of the IVC is covered by built-up metal panels. The differential pressures at which these panels lift is 0.8 psig. The weight of these panels is 3 lbs/ft'. Ihe panel is assumed to move parallel to its original position (note the panel has a small slope away from the Containment Building) till it clears the sidewalls of the IVC. Once the panels clear the walls, it is assumed to lift away from the path of the flow of the steam air mixture to the atmosphere. Thus, this movement of the panels above its nominal position creates movable nodes 10 and 11 shown in Figure

3 3.6.A-1A The node volume and junction parameters of the IVC are given on h>

Table 3.6. A-2A. Node 10 and 11 have variable properties as the panel moves above its nominal position. The vent area and the volume of these nodes are ven in Tables 3.6.A-3 and 3.6.A 4 The nodalization model selected for the common corridor area is shown in Figure 3.6.A-5. Ihe node and junction parameters of the common corridor area 1 of the IVC are given in Table 3.6.A-6. Results of the cases which yield maximum pressures in the various nodes of an IVC cubicle,, including the associated AW pump room are presented in Figure 3.6.A-3. In MSI.B case 1 ell ',h.;?;... is assigned to node 6, while in MSLB case 2 al M 1-g oe/ to N Ne 7. In the AN break casp a44.Mowdoom is assi ed to node 2.- L. y -k pressures for the limit ng ca $ in each node are indicated in Table 3.6. A-2 kgMg / Feak pressures for the SC blowdown line break in the common corridor area of the IVC structure are presented in Table 3.6.A 6. For generating the equipment qualification temperatures and pressures of the IVC a simpler 3-node model of the IVC has been used and the volume and junction properties were inputted into a modified COPDA code named FLUD (see Section 3.6. A.3 for discussion of FIi!D). The simplified model consists of 3-nodes with node 1 being the auxiliary pump room between E1.10 ft and 32 ft, node 2 is Lcrveen El. 34 f t and 55 fi-6 in. and node 3 occupying space above 55.5 ft. Out of the various casar, considered, MSLB produced.the limiting temperatures and pressures in the IVC. The long ter M rdm -used in the analysis is presented in Table 3.6.A-5 and the te rature profiles are given N O$$ I I 3.6.A-2 Revision 2 l l l

INSERT A The pressure analysis for the Main FW line break, AFW Line break, and steam generator blow down line break were modeled using the COPDA computer code. Details of the code are given in section 6.2.1.2.3. Short term mass and energy releases were calculated using the methodology of References 3.6-9,3.6.A-6, and 3.6.A-7. The MSLB subcompartment pressure analysis was determined using the GOTHIC 4.0 HLP-001 (Reference 3.6.A-8) compater code. Details of the code are given in Section 3.6.A.6. The short term mass and energy releases for theIVC subcompartment pressure analysis were determined using the RETRAN-03 computer code, which is discussed in Section 6.2.1.4.7. INSERT B For the MSLB analysis, the nodalization scheme is presented in Figure 3.6.A.-1B. The nodalization is similar to that of the main feedwater line and AFW line break with the exception of the modeling of the movable panels. The position of the movable panels are modeled as a function of IVC pressure using the STEM TRAVEL code (Reference 3.6.A-9). The node and junction parameters of the IVC are given on Table 3.6.A-2B. e ,e O (FSARVvC WP951) S!2V95 .. - =. -...-. 4,. m n., ....wa....-

STPEGS UFS U o.d @'Z%b in Figure 3.6. own has been obtained using Westinghouse 14f7RAN code (Ref. 3.6. A-5). 3.6.A.3 FLUD. A Comeartment Differential Presgure Analysis Code. This describes the computational procedure and the analytical techniques used in i FLUD. The analytical basis for COPDA is described in Reference 6.2.1.2-2. The set-up of initial conditions, the determination of the thermodynamic state point at subsequent time increments, and computation of energy and mass transport between one time step is discussed in Sections 3.6. A.3.1, 3.6. A.3.2 end 3.6. A.3.3 fer FLUD. Selection was made of the control volume and flow path configuration that resulted in the best representation of the pressure transients in the compartments along the flow paths from the break. The major differences between FLUD and COPDA (Ref. 3.6.A-4) are the use of steam table curve fits (Sectioc 3.6. A.3) instead of table look-ups, the equation of state which is a first-order virial expansion (discussed in Section 3.6.A.3.1) and the capability of wall heat transfer calculation. The fluid flow equations (compressible equations, HEM model and integrated momentum equation) used in COPDA have been reproduced in the FLUD code. It may be observed from the FUUD flowchart in Figure 3.6.A-2 that the calculational pro,cedures for FLUD and COFDA are very similar. 3.6.A.3.1= E'austion of Sta11 - This section describes how FLUD determines the thermodynamic state for each compartment in a system of interconnected compartments. The thermodynamic system (compartment) is assumed to be in equilibrium. The states assumed by the air-steam-water mixture can be described in terms of thermodynamic coordir.ates, P V, and T referring to the mixture as a whole. The equation of state is derived from a first order virial expansion as presented in Reference 3.6.A-1. Using the molecular theory of gases, the following equation of state for an air-steam mixture is obtained assuming negligible air-steam molecular interaction: P-(M.R, + M.R,) T + (M,)2 R. T B,(T), (1bf/ft) s V V (Eq. 3.6.A-1) where the temperature dependence of the second virial coefficient for steam R,(T) is given by (Ref. 3.6. A-2). B,(T) - 0.0330 - 75.3137 10 s.usent

• x 20 * + 1.uoes T

(Eq. 3.6.A-2) Equation 3.6.A-1 can be rewritten as the sum of the partial pressure of air P, and the partial pressure of steam P, where 2 P, - M.R.T. (1bf/f t ) - 0.37043 T. (psia) (Eq. 3.6.A-3) V v, 3.6.A-3 Revision 0 i ._m

sTPEcs urSAR. E,w atmospheric mass flow rate mass condensation rate 3.6.A.5 Thermodynamie Prooerties of Steam. Vater. and Afr - FLUD uses steam, air, and water properties for various thermodynamic calculations which are performed during each step. The thermodynamic variables needed in FLUD j calculations are: specific internal energy of air e.(T) P..t(i) saturation pressure of steam saturation specific volume of steam v.,t(T) specific internal energy of steam e,(T P) specific volume of water v.,(T) specific internal energy of water l e,(T) saturation temperatura of steam T.s(P) saturation temperature of steam T,.t(v) saturation specific internal energy of steam e 1(T) saturation specific enthalpy of steam h,,t(T) I hg,(P) enthalpy of vaporization of steam i The ' unknown" quantities that can be used to calculate the above nine. vari-ables are the macroscopic compartment thermodynamic variables pressure, specific volume, and temperature, P v, and T, respectively. ~ The air and water properties e.(T), v,(T), and e (T) are calculated by fitting polynomials to data in the steam and gas tables (Refs. 3.6. A 2 and 3.6. A-3). The air property e (T) was found to be adequately represented by a linear fit. This is no doubt due to the good " ideal gas" behavior of air. Thus, e.(T) -aT (Eq. 3.6.A-30) The water properties v,,(T) and e (T) and the steam properties h..s(T), e.(T), and e,,s(T) are very nearly straight line functions, but small variations were l accomdated by using third order spline polynomial fits of the general form: property (T) - a, + a2T + m3T2 + a3Ts (Eq. 3.6.A-31) j For example, for hg,(P); I + a P's (Eq. 3.6.A-33) } hg,(P) - a, + a3P + m2 P2 a The accuracy of the curve fits range between 0.01 percent and t. percent for the various properties. 3.6.A-9 Revision 0 ~.. -,

i INSERT C 3.6.A.6 GOTIflC 4.0 HLP-001_. GOTHIC 4.0 HLP-001 (Reference 3.6.A-8) is a state-of-the-art program that solves the conservation equations for mass, momentum and energy for multi-component, multi-phase l flow. The phase balance equations are coupled by mechanistic models for interface mass, l energy and momentum transfer that cover the entire flow regime from bubbly flow to lI film / drop flow, as well as single phase flows. The interface models allow for the possibility of thermal nonequilibrium between phases, and unequal phase velocities. GOTHIC includes full treatment of the momentum transport terms in multi dimensional models, with an optional one-dimensional turbulence model for turbulent shear and turbulent mass and energy diffusion. Conservation equations are solved for three fields: Steam / gas mixture Continuous liquid Liquid droplet a The program calculates the relative velocities between the separate but interacting fluid fields, including the effects of two phase slip on pressure drop. The program also calculates heat transfer between phases, and between surfaces and the fluid. Liquid droplets are transported in the vapor / gas flow. In addition, a mass balance is solved for a solid ice phase. The somewhat simplified model for the ice phase does not provide for transport of ice, and if ice exists within a model, the ice temperature is set to a constant value by code input. Ice remains at its initial temperature throughout a transient. Ice can only change phase to liquid; i.e., there is no direct ice-to-vapor phase change. The three fluid fields may be in thermal nonequilibrium in the same computational cell. For example, saturated steam may exist in the presence of a superheated pool and sub-cooled drops. The code can model extremely dry noncondensable gases (down to steam partial pressures of 0.001 psia) and has a temperature range from -50 F to 8540 F. The steam /r,as mixture is referred to as the vapor phase and is comprised of steam and, optionally, up to eight different noncondensable gases including air and hydrogen. Mass balances are solved for each component of the steam / gas mixture, thereby providing the volume fraction of each type of gas in the mixture. Solution of the equations is based upon nodalization of the region ofinterest, with the principal element of a model being a computational volume. The program features a flexible noding scheme that allows computational volumes to be treated as lumped parameter, one, two-or three-dimensional, or any combination of these within a single model. As a i minimum, a GOTHIC model consists of at least one lumped parameter volume. Subdivision l FSARVVC WF95-0 l 3/W95 k u -w l

l of a volume into a one, two-or three-dimensional mesh is based on orthogonal coordinates. Adjacent cells in a subdivided volume communicate through parameters defined by discretization of the governing equations. Separate volumes communicate through what are referred to as junctions or flow paths. A separate set of momentum equations are solved for junctions. GOTHIC has been verified by HL&P against the applicable portions of ANSI /ANS 56.10-1982 (Reference 3.6.A-10) for subcompartment pressurization analysis. 3.6.A.7 STEM TRAVEL STEM _. TRAVEL (Reference 3.6.A-9) was developed by HL&P to facilitate the use of GOTHIC in subcompartment pressure / temperature (P/I) transient analysis. The code calculates the panel height corresponding to a given MSLB subcompartment pressure profile generated from the GOTHIC computer code. The vent paths are modeled by an equivalent panel height, and in tum, translated into an equivalent stem travel. GOTHIC generates a new pressure profile btsed on the new stem travel profile. Iterations between the two codes are done manually until convergence between the stem travel and pressure profiles is obtained. The panel height is obtained from the goveming equation for dynamic vent paths as discussed in Appendix E of ANSI /ANS 56.101982, "Subcompartment Pressure and Temperature Transient Analysis in Light Water Reactors," Reference 11. d3 o oue_ p,,, y H ,p zpu_p de pA, where: M mass of panels = current vertical panel displacement, ~ h = fully open vent area A, = weight per unit area of the panels, P., = static pressure at current time in the region beneath the panel, P, = time t = static pressure in the region above the panel. P,, = mass flow rate W = fluid mass density p = USARWC WP95-1) anws

The equation is simplified to the following form in STEM _ TRAVEL: h =h,+ v s + EE ( l. "~"'

• s' + #'"' *~'** ~'* s )

^ 3 o m 6 x 2 current vertical panel displacement, where h = panel displacernent at beginning of time step, h, = panel velocity at beginning of time step, v, = time step sizo,4t. T =. gravitational constant, g, = mass per unit area of the panels, m = l P,,, weight per unit area of the panels, = static pressuro at current time in the region beneath the panel, Fu = l Pg, = static pressure below the panel at beginning of time step, and P,, = static pressure in the region above the panel. f i I i l (FSARVVC WP95 4 3f1&f95 .. <e

• m

c..,

s,. ...c.. s .s ,,,w,

STPECS UFSAR REFERENCES Aeoendix 3.6.A: 3.6.A-1 Reif, F.J. Fundamentals of Statistical and Thermal Physics, McCraw-Hill Book Co., p. 183. 3.6.A-2 Kennan, J.H. et al. Steam Tables, John Wiley & Sons. Inc., New York, 1969. 3.6.A-3 Keepan, J.H., and J. Kaye, Cas Tables, John Wiley & Sons, Inc., New York, 1948. 3.6.A-4 Bechtel Topical Report BN-TOP-4 Rev. 1. October 1977, "Subcompartment Pressure and Temperature Transient Analysis". This report was approved by the NRC in February, 1979. 3.6.A-5 Burnett, T.V.T., et. al., *14FTRAN Code D2heription". VCAP-7907-P-A (Proprietary Class 2), VCAP-7907-A, (Proprietary Class '3), April 1984. 3.6.A-6 Bilanin, V.J., *Ceneral Electric Mark III Pressure Suppression Containment System Analytical Mode", General Eleccric Topical Report NEDO 20533, June 1974 3.6.A-7 Sharma, D.F., " Technical Descripton - Annulus Pressurization Load Adequacy Evaluation", General F.lectric Topical Report NEDO 24548, January 1979. QS WYh ~ 3.6.A 10 Revision 0 M _- J U.1_ E 'W F WWWW M WE ' E.E W E EE E M

i l 1 l INSERT D 3.6.A-8 George, T.L., et al, " GOTHIC 4.0 HLP-001, Containment Analysis Package," Developed by Numerical Application,Inc. for the Electric Power Research Institute, September,1993. 3.6.A-9 Futschik, M.W., " STEM TRAVEL", developed by Houston Lighting and Power Company, February,1995. l l l 3.6. A-10 ANSI 56.10-1982, "Subcompartment Pressure and Temperature Transient, l Analysis in Light Water Reactors." l l l l l l l l l l 63ARVVC WP95-1) 4 3/24/93

I j STPECS UFSAR l l TABl.E 3.6.A-1 i M_AIN STEAM LINE BREAK BLOVDOWN (An enth py of 1200 Beu/lba is conservat.ively assumed yfroughout). / Stea Generator Time A or B (secs) (1b/seci l 0.0 0.0 0.0066 8700.0 0.01 6090.89 0.013 5763.63 0.025 5531.81 [ 5227.25 0.05 O.10 ,/ 4890.86 0.125 / 4972.72 i s l 0.15 ' 5190.87 s 0.20 50,36.35 0.25 '/ 4899.95 0.30 4809405 l 0. 4763.6 l 0; O 4718.15 0.45 4718.15 0.50 4672.70 h95 Tb-hd "N 1 I l 1 3.6.A-11 Revision 0 i

fADtt3.6.A-2O IYC StfBCCrent> tut NCDAL_0ttCilPf!04 Fo<- A.'s P u L t < y INFO L.', t,qd SC Palne199 t.w Bre k r Inittel_C m ottl m Flow Flow Cateuteted Volu=* Yojtm fem. Pressure Punidi t y flow Arge coefft-t/A Peak Press. Sreek a twe ft

• F role Percent Path ft elmt f(I Mle in u

1. 3188.$ 103.0 14.7 30 21.65 sista 2 2. 1977.$ 10$.0 14.7 30 2 1 15.45 0.73 0.05 21.30 mtts 3. $$30.95 103.0 14.7 30 32 54.16 0.82 0.18 20.62 ptte 4 2558.$ 105.0 14.7 30 4 3 210.3T 0.80 0.024 20.66 msts 4 5 64.60 0.92 0.26 4 6 236.84 0.82 0.02 to u 1453.0 105.0 14.7 30 5 3 115.72 0.80. 0.035 23.31 nste Q $ 7 131.90 0.817 0.04 g m Y 6. 2221.7 105.0 14.7 30 6 7 56.52 0'.91 0.27 22.99 psta e [ 6 8 195.42 0.80 0.0$ g 7. 1262.26 10$.0 14.7 30 7 9 90.29 0.79 0.09 28.87 nste 8. 7937.44 10$.0 14.7 30 8 9 80.55 0.25 0.07 19.81 R$tt 8 10 172.5 0.65 0.038 9 ' $448.47 105.0 14.7 30 9 11 178. 0.72 0.04 19.26 R$te 8 9 12 15.$ 0.54 0.48 10. (Pteese see febte 3.4.A 3 for detalte for thle nede.) 11 (Please see table 3.6.A-4 for detelte for thle node.) 12. 1.0E22 105.0 14.7 30 p r E:: PJ

d TAott 3.6.A.2 h tYc stmcMAttMut umAt etscalPitow Fev. % 5 %,t.ws BA I a __ fnftlet Cervittfees flow itew Cateviated coefft-t/A Peek Press. treek veju 4, fewp. Pressure meldity flow Arge ] volu e nWr ft

• F eefe Percent Pat %

ft etent ft este tyre 1 i 1. 3588.5 -10$'.0 1cf t,o 14.7 30' b 21.65 Mste 2 1 2. 1977.3 tCS;0 g do14.7 30 20 2 1 75.45 0:78 t>.~8 0.05 21.30 ests I 3. $$30.95 10%.0 g d o 14.? 30 ';Lo 3 2 St.16 0.82 c b'5 0.18 20.62 mste i I 4 2$38.5 10hc joi.o 14.7 30 d'o 4 3 2to.3r 0:co o.b' O.e24 20.66 mtte 4 S 64-60 F J i 0:92 0. 9 0.26 l 4 6 256.84 0.82 0.02 (A ts S. 1453.0 Scho jo{,o14.7 Mpo S 3 115.72 c.to o.Bl 0.033 23.31 Rsta Q 5 7 13f:90 0.177 o.Bt 0.04 g

• e gg

[ Y 6. 2221.7 t053)cAl.014.7 3@ & 7 5M323W 0:tt c>.E or2r o.-# 22.99 aste [ 6 8 195.42 0.80 0.05 y 1 7. 1262.26 sono tcN.ot4.7 30-7 7 9 90.29 0.79 0.c9 2a.ar asts I 8. 4952.11 los A g g o tt.7 -30 y 8 9 80.53 tot. 5f1' 0.e5 o.e440.or o.on 19.83 mste 81 ~31.% 8 40'39 172.5 0.65 0.038 $Q oGo.BS 9 54*s:c7 tono 1o9.o 14.7 3o2O ,.st-ar ira. o.72 0.04 19.26 asta 8 0.48 9 12 } P 15.5 0.54 ff. 3P (PC a: ::: t+u: ?.".;.-3Mtette tehle-eiede.) SI A Id{.o 1 p tL %.5,% Qgh c.M.b.9C f ce.. = mm., 9,-] me mn~s,- Pb iM o I l"7 th l I 12'.1f t.cs32 Josse W7 y a pA lo4s ci, j O l U ea 1

l l ATMO5PHERE 12 14 16 15 MOVABLE MOVABLE COMPARTMENT COMPARTMENT j 11 10 i 13 12 11 9 8 10 9 l 8 7 6 l i 7 6 5 s 4 i 4 3 a 3 2 2 2 j 1 1 3 i 1 SOUTH TEXAS PROJECT 2 UNITS 1 & 2 1 NODE AND JUNCTION DIAGRAM OF THE IVC i fi:A GcT9A Figure 3.G.A-1 A Revision 0 j

1 Atmosphere Junction 14 (HVAC Junction Junctbn l Relief 13 2 Opening) Junction 11 Junction Junction 10 g Junction 8 Junction Junction 7 6 Junction 5 Junction Junction 4 3 3 Junction 2 2 Junction 1 1 uooe no waw wAcab o p TOLWC 7 o F G o r H T-C P s ov>:.3 6.h-\\B Riorn=*3 C w..,.., m,

i i _q"~ M i Z O 1 G_ M \\ W 1 t' l l Z \\, .c' W -; N l T E: }s D O-l M *'z: J \\ m ;- N .s

6. J W 8-Ni bM

. 3 V Q_ 0, '\\ y s m u / \\ N l a >y O N y s 6 \\ Y N, _o y: / 'N k N \\ CD i X J t W N y _f N w w h J O o I-- N o a e s e o u w (visa) wnss3wd scoN l l l SOUTH TEXAS PROJECT l yNITS 1 & 2 PRESSURES IN IVC DUE O MSLB NODE 1 (BREAK NODE CASEE)I ( Sheet 1 of 9 Figure 3.6.A-3 Revision 0 i ! a.nm... ~... I

. muusis I r l l 1 ak in Node 6. A ^- 'w m e. i ~ co.........+. y ..................e......... .e........., .......)'.......4'........g:.........:........1,........., .y s m a n n .a ,.........y.........g.......g....... #.,.........y......... n g a. v m.......... w g o 1 o. 1 e l a --.......p....... g.. .........y........y......... w m m _......... + $ L.......p......4..........:..................;.......... l l c -... iia i ,,,,,a i,...',, i i,ia , i i iiiii'..ii.in 1000) "v.001 0.01 0.1 1 10 100 SOA s eWi t 4. 0 m.t. 9 01 09/22/95 18:22.$4 i l l I r t L l i l l I t l l \\ '.e' l .. w.. gay, men..e h a w e.' %. fA.. - ,.+t. v. A sA. San- -.

• • • w

STP: MSLB IN IVC, PRESSURE RESPONSES ADSOLUTE PRESSURE IN NODC 2 ~ 2 A / 21 i gn / l ' >, / ,/ /N _vy g-l,/ / h j 1 / / / p 16 I / / ? m l / n 2, o Y b "c cI / O [5 'Z I 14 p 2 ym 0 0.2 0.4 / 4 z e x i TlWE (SEC) g E a >)

x..

wz u o N]o -fg c,cc (j,h 6 g V-Oo e g-R s: m -b;e = o a I o

l l Baco IVCSPP Hods 1 with Bratk in Nodo 6. Wad Mar 22 19 18:06 1995 GOTHIC Varaicn 4.0 HLP-001 - Octcber 1994 Break in Node 6 ,.PR2 w.,~'~"' N - v v v ^ m I tom N i 1 .......$....... 4.........:....... g n n m u .......y...............;........y........y......... 1 I e.9 ) F2,oo I g l e u ) ........;......4.... I o \\ e s. n ) \\, ou l $ -........ D. s y...... .j........g........g.........p........ r g 1 .~ .s.........>...............:.........:.........:.......... I. t \\ l '. v _,{, _ t t ttt r' i 1 o t t ttt! 1 i t o tett? I t t i e tti! I t t e ttit! 1 t o t e t t! .031 0.01 0.1 1 10 100 1000 -'m G en t t 4. 0 X1.7 0 01 03/22/95 10:22 54 l l i l e 5 1 i l 5 l .o.. ..a r l

O STP: MSLB IN IVC, PRESSURE RESPONSES A_sso!.uTE PRESS _URE IN NODE 3 _m s_ 1 20 \\N ^ 19 \\ in S Um. 17 16 s z, o ( E C 15 ) my 'o I ? Wc .C _B$ Z m5m q m-l 14 0 0.2 0.4 ii > f m >X 2X-l TlWE (SEC) W2$ Am ~ " - oO go g ~

5"3 ro8 Lplam a.h esc Al Frc, re3 I

s. oo c_ Rg m ma O [ 2-O d U3 d

i l l B200 IVCSPP Hod 31 with Br= k in Nodo G. I Wad M2r 22 19:17:48 1995 GOTHIC Varcion 4.0 HLP-001 - Octcber 1994 l l l l Break in Node 6 PR3 a Dw ~ t u i t I e g .r....... 3........ 3......... g n u c .,4 o 3 gj ........p.................:........:.........:.......... - {'.3 ^ s e 1 u s I a m O ou m m -......3........<........<.........:.........y........ s. w ,y 1 V ~ 1 i f f it!!! 1 i f il t ff! f f I f ffit f 1 i l f f f!! t 1 1 1 1111! t f f I f f t? 4.001 0.01 0.1 1 10 100 1000)j w..... ~ v {sec) y - so;xte 4.0 m.cci 03/22/95 18.22.54 I I l e i L

O e 't STP: MSLB IN IVC, PRESSURE RESPONSES AssoLUTE PRESSURE _th_ NODE 4 s 2o. W s m \\,, l i .),, = O Z 16 2! g 3 53 0 C CJ Om l lm mg -4 15 - I I i A C C ^$m 2 -4 HSE j> 14 o o.2 o.4 / 2*8 /

• z am l

Re *Q ~ / _.. ~ m gm g ! f coa d. cr e 9 o m mn o O5 -4 l o i

Beso IVCSPP Model with Crack in Neds 6. l W:d Mar 22 19:17:55 1995 l GOTHIC Varcion 4.0 HLP-001 - Octcb r 1994 1 Break in Node 6 PR4

~ l m

g........>........:....... 4........ ,[~.,%. a e < y.- m -........p..............3.........y........y......... \\,3,,, - n u ,3 1 n s v p. i g .......t.......4.................:.........:......... o u

3 l

n t n t o y y.......;....... 3..........y........y........ u on \\. i y................4.........;.........:.........:.......... ( \\ s s T f f f f it t!' f f f 1911!! I f f littt! I f f f f t tt! I f f f itt!! f f f f f f f' i k !\\ 4.001 0.01 0.1 1 10 100 1000 - Time (sec) ... ~. - ~. ~. d eTxit 4.0 x17-e s t 03/12/95 18:22:54 1 e f l l i 1 l I i I

y y 4 1 1 l STP: MSLB IN IVC, PRESSURE. RESPONSES '1 cmumwit_eatssunt is soot s I 23 - i l 3 m l ? 21 E b 2o y V VN h 3 = s N E l e. ja I w A (D ' 17 - Sa o C gg _ i -w am p mg -4 j' P

• C.C I

-@"$ Zq 15 pg qm E 7: 3 CD y 14 A (/) 0 0.2 0.4 m@h I 688 mm Tiut(ste) -wm 33 m_ -~ 3 hN CC h-i wE m g mp o SS -1 o

m-. Bace IVCSPP Model with Br@ak in Nodo 7. Wad Har 22 19:37 57 1995 GOTHIC Version 4.0 HLP-001 - October 1994 Dr d in N.h 1 n~ b a ........;........4.........{.........'.,.........;.......... 4 9,'yarc h g .....:.p....... ........{. .......;.........]......... = l ~.......'. ..J...... I k i .......p........<........<.........:.........p......... 4 i. l = t i e nint i i i r e nd i e i nin/ i sittin' i e i n ini' i s iten .. c o s c.s1 e.1 1 to sco sooo - J 4 eTxt e 4. 0 m 0 01 03/22/f5 19:2? 01 I t s ll

STP: M$LB IN IVC, PRESSURE RESPONSES ADSOLUTE PRES $URE IN NODE 6 23 22 - ( 21 - j l l ^ 20 - A, y I W t/ N 19 - 0 1s -~ E 5 17 .n { M S z, o h 8$ C 16 a m m -l 'P mE C I w m 15 2 --{ - m p qm Z X 2 M) J 14 1 = o o.2 o.4 e+.g;; aM m Tiut (stc)

gg >2

-gh_c,j Igo4 i 2 -m m ~'C-O o L m3 m mp O o em -4 1

Base IVCSPP Modal with 2 rook in Nodo G. Wed Mar 22 19:17:51 1995 GOTHIC Version 4.0 HLP-001 - October 1994 Break }rs Node 6 PR6 d A cs i ( g .......;....... 4........ 4.......

1. L eq l

,g g ......q.........p........ s nn, t ~. - \\ t @........ ;..... 4...... ou

s

s. -

e o g ........g........ ........g.........:.........y........ \\. c o t ........;....... 4........ 4..........:.........:........ 4 \\ l i i s. v ,,,,,,,e ,,,ii,a i, i i,,it ,,,, iia,,,,,,,i- ,,,,,,,s i 4.001 0.01 0.1 1 10 100 1000,.i Time (sec) /' ..s seTwse 4.0 m.est es/22/95 18:22 54 -RIG. 4.2 1 i / I I I l 1 i i i J l i l l i I l l l i I l i h 1 $1MWE'2We.es. pew.-*==.ei. maw s e= ., *w. .v..

STP: MSLB IN

IVC, PRESSURE RESPONSES e" 'E'" ' ' " " ' 5 "'" ' ' " "

'7 2. 1 28 = 27 - 26 25 - a# { 24 23 Y 22 E 21 - E \\ 2o \\ w O 19 - 18 - E (D z, O C 17 ,w d mg 16 ? Mc.C 1 mj$ 2 -4 15 im = mII 5 'A 2 (n >>( 0 0.2 0.4 TIME (SEC) @mC ~~K y p c.C W i E az m g mp O ti m -i o 9

Daco IVCSPP Modal with Brack in Nodo 7. Msd Mar 22 19:38300 1995 GOTHIC Veraion 4.0 HLP-001 - Octcber 1994 i e red. im n.a.1 FA7 'M s( ~ l (

..........;.......4........4..........:.........:.........

F047 1 ( ( l ,1 g -........ s 3 i n -.......a............... ~ u t s. .6........ ..;.......4........4.........:.........:.......... ) . i i i i m* i ....in' i, i i iiii' i...... ' i e i nitii . i i ii ii, . cot e.ot e. a se too 2000 TIM (ses) ~.~/ v-(eTxt t 4.o XLP-c ol es/22/95 19:27.e1 1 l su t- * !- .J. 2 - 3 1 I l l l s.a -. : ~. - .... a... -o ono. vww -. e.,. I l l

P STP: MSLB IN IVC, PRESSURE RESPONSES sot.UTE PRESS)RE IN NODE 8 20 - 19 ? [ 18 V U S 17 'N E wo 16 A M E s 53 o om C j

• 8 d

15 ES.C I .) I m zy u a,E Qm pm

• E 7: _.

CO X 14 0 0.2 0.4 a emc R* T v/v' N - Tiut (SEC) y ~OO u x / b'fCb mm x k>f t, ). CR C-r. R5: m

• U O

2* -i o m. m. .A m.

Bac@ IVCSPP Mod 31 with BrCck in Nodo 7. Wad Mar 22 19:37 52 1995 l GOTHIC Version 4.0 HLP-001 - October 1994 j Press ter p.t..a eas sw- -- A, ,g p....... 4......... q.........;......... p......... , }~; g8 .......g...... g........g....... ........y......... ~ j ( / s, .......;........g... ........y........y........ e, g e. i i., i a

-........y...

... 4........ 4..........:.........y......... s s ~ e t t s t a t ti' t s t i t t t'? e e t eetti? e etteen? t t t i t u t' s a t t t tt!

• • w.001 0.01 0.1 1

10 100 1000 a M* Q o m t 4.0 m,.00, 0,,,,,,, ,,,,,.0, l l \\ l I l --._m E -r l b

STP: MSLB IN IVC, PRESSURE RESPONSES 19.5 A s ~ '\\ ) 19 .l18.5 2 18 \\ l 17.5 1-a 17 \\ Es ~ yf.16 o m ] E O l $3 0 { 1 s.s ) ta Om C b M$ d -- ) ( b Sc I 15 ^hM 4 / 0 E Qm i 14.s ) g (A 0 0.2 0.4 g y ag@ e' *m Tiuc (stc) eme

w % ~ ~,,

-4 N k 4 W'. C-1 ' E: Rs: hi 8 mE O O* -4 o

i Baco IVCSPP Modol with Breck in Nods 6. Wsd Mar 22 19:17:59 1995 GOTHIC Vorsion 4.0 HLP-001 - Octcbar 1904 l l l Break in Node 6 l PR9 gO cs _.......;.......4........4..........:.........v........, g y g. g l n m......... n m s........;........;.........y........y........ n o. s v g .......;...... 4.........:.........v........v......... o .u 2 t e n (. o g .~ .......;...... 4........ 4..........'.........v........ j i. t. ~ i t' t t tet i' t t t otti e tiett t t Itt i i ttatt tt1 3 %s... u

s. Time (sec)

/ s..._...._-__ d eTxt c 4.0 XLt.4 01 03/22/95 18 22 54 I i l l l 1 i l l l

I STPECS UFSAR i blowout panels are used, thus the flow are. is assumed to be constant with I respect to time. eH oC-h\\.M d h '4 blw.b N N T1 Genera 1Me subcompartment pressure transients were l determined using the COPDA computer code (Ref. 6.2.1.2-1). The COPDA code employs a finite difference technique to solve the time dependent equations I for the conservation of mass, energy and momentum. This code and the assumptions inherent to it are fully explained in Reference 6.2.1.2-2. loss coefficients utilized were based on the formulations of References 6.2.1.2-3 and 6.2.1.2-4 p k' I l Nodalization of each subcompartment was based on the physical arrangement of l the interconnected subcompartment and the structure, equipment, piping, ventilation ducting, floor grating, and other physical obstructions to flow. By appropriate selection of node boundaries based on the physical arrangement, pressura differences within a node are minimized while pressure differences between nodes are maximized. The LOCA blowdpwn model used to calculate the short-term mass and energy release rates,for all primary system ruptures, including the surge line break and the pressurizer spray line break, is fully described in Reference 6.2.1.2-5. The mass and ener5y release data are presented in Table 6.2.1.2-1.

h. g.n ?The-@EJAP>$~c~oTe-(Ke~f'. D.1.2.)T~w~~a~s Iis]e'd'i~oTa[1'cL'1' ate' the s m

~ ~ blowdown of t'he main' steam li,ne and ma n feedwater line.,The mass' and'ener y f jleaseratesfordhesecup91nesar~provided ab ed.2 p K1 j tdovr j ine breaI blowdown vasgl'culate -using t er n_Nacional-ScandaIds nstitute (XNd) methoaoibly of.iferen 2-7. The ANSI methodology for subcooled blowdown from a pipe break results in a decompression vave propagating through the system at sonic velocity with the pressure behind the wave corresponding to saturation pressure of the liquid. Because of the very low compressibility of subcooled water, subcooled blo.*down cannot be sustained for more than a few milliseconds and the total mass release under subcooled blowdown conditions is quite small. Following this extremely short-term initial phase, the pressure vill correspond to saturation pressure. The blowdown for saturated and subcooled water conditions is determined using Henry Fauske and Hoody relationships and is given in Table 6.2.1.2-1. 6.2.1.2.3.2 Resetor cavity - No pipe breaks are postulated in the reactor cavity and inspection toroid. 6.2.1.2.3.3 Steam Generator subconmartment - SG subcompartment design pressure.is deteruined by breaks in the pressurizer surge line and SI accumulator injection lines. The blowdown for these breaks is presented on j Table 6.2.1.2 1. The noding of the SC comparraents is shown on Figure j 6.2.1.2-3. The node and junction diagram is shown on Figure 6.2.1.2-11. The flow parameters were evaluated to account for all obstructions such as cable. i tray supports and various small-sized piping. The principal obstructions within the SG loop compartments are the SG and reactor coolant pumps. The NRC approved COPDA computer program (Ref. 6.2.1.2-2) was used to perforra the subcorepartment analyses. The modeling for the COPDA program requires that i 6.2-11 Revision 0 1 i .--,_____m --__ ~

INSERT 1 The main steam and feedwater subcompanment pressure transients were determined using the GOTHIC 4.0 computer code. A description of this computer code is presented in Section 3.6.A.6. INSERT 2 The RETRAN-03 computer code was used to calculate the short term mass and energy release of the main steam line. The RELAP 05 computer code (Ref 6.2.1.2-6) was used to calculate the shon term mass and energy release of the main feedwater line. The mass and energy release rates for the main feedwater line is presented in Table 6.2.1.2-1. Letdown line break mass and energy release was calculated using the American National Standards (ANSI) methodology of Reference 6.2.1.2-7.

I STPECS UPSAR 6.2.1.2.3.6 Main Steam and Feedwater Line Subcomcartnents - The main steam and feedwater line subcomparta'ents are shown on Figure 6.2.1.2-6, with a node and junction diagram given on Figure 6.2.1.2-14. A double-ended main steam line rupture (8.1 ft ) was assumed to occur in either Node 1 or 3. The 2 calculated peak pressure occurred in Node 3. A double-ended rupture (2.837 ft ) of the main feedvater line was assumed to 2 occur in either Node 5 or 7. The calculated peak pressure occurred in Nede 7. [pptedr3 Node and junction parameters utilized in the analyses are given in Tables 4 6.2.1.2-13 and 6.2.1.2-14 Plots of calculated pressures are given on, Figures 6.2.1.2-25 and 6.2.1.2-26, while calculated and design values are compared in Table 6.2.1.2-13. Mass and energy release races are provided in Table 6.2.1.2-1. The-eass-and suoig,y. 1..s maces use cultantMMIT1Tig RE1:AP-5 analysi Q,g 1, - 6.2.1.2.3.7 Recenerative Heat Exchancer Subcompartment - A double-ended rupture of the CVCS letdown line is the limiting break in the regenerative heat exchanger subcompartment. A node and junction diagram is given on Figure 6.2.1.2-15. The nodal model initial conditions, control volumes, vent areas and corresponding flow coefficients and inertial terms are given in Tables L.2.1.2-15 an4 6.2.1.2-16. The calculated subcompartment pressure response is shown on Figure 6.2.1.2-27. Calculated and design presseces are compared in Table 6.2.1.2-15. The blowdown rate for the CVCS letdown line break is calculated using ANSI 58.2, Appendix E2, methodology (Ref. 6.2.1.2-7) and applying that to a one-dimensional Henry-Fauske model for saturated liquid. Mass and energy release rates are shown in Table 6.2.1.2-1 (Refer to Section 6.2.1.2.3.1 for more details). Plant operation is assumed to be in the heat-up mode. The break is assumed to occur at the inlet to the regenerative hear, exchanger. The break area is 0.0884 ft2 for each end of the double-ended break (0.1768 ft2 total area). There are no significant restrictions to forward flow, but the reverse flow is restricted by the CVCS letdown orifices 2 (0.00166 ft ) located immediately downstream of the regenerative heat exchanger. In addition, the reservoir of reverse flow is limited since high energy fluid conditions extend only to the letdown heat exchanger. 6.2.1.2.3.8 Radionetive Pipe Chase Subcomnartment - A double ended rupture of the CVCS letdown line is the limiting break in the radioactive pipe chase subcompartment. A node and junction diagram is illustrated on Figure 6.2.1.2-16. The flow model initial conditions, control volumes, inter-compartment flow paths, and corresponding flow coefficients and inertial terms are listed in Tables 6.2.1.2-17 and 6.2.1.2 18. The calculated subcompartment pressure response is shown on Figure 6.2.1.2-28. The calculated and design pressures are compared in Table 6.2.1.2-17. The blowdown rate for the CVCS letdown line break is calculated using ANSI 58.2, Appendix E2 methodology and applying that to a one dimensional Henry-Fauske model for saturated liquid. Mass and energy. release rates are given in Table 6.2.1.2-1 (Refer to Section 6.2.1.2.3.1 for more details). Plant operation is assumed to be in the heatup mode. The break is assumed to occur at the Containment penetration. The break area is 0.0884 f t* for each end of double-ended break (0.1768 ft' total l area). A significant restriction to forward flow is the CVCS letdown orifices (0.00166 2 f t ) located immediately downstream of the regenerative heat exchanger. For reverse flow, the letdown heat exchanger reduces the line temperature to 115*F ard a pressure reducing valve. immediately downstream of 6.2-13 Revision 0 e=_ ~- =. -,, -- u- ---.w

INSERT 3 l Results of the analysis show that the main steam line break bounds the main feedwater line I break case. l INSERT 4 The mass and energy releases were calculated using the RETRAN-03 computer code as discussed in Section 6.2.1.4.8. E' l <--. < - -.. _. - m m. _ >. _ _ _ - _., - _., - - - - - _.. _,. =.. _. __.- -. -, _ _ _

= STPECS UFSAA 2.- Failure of Main Feedvater Pump Trip No credit is taken for feedvater pump crip and coastdown in calculating main feedvater addition prior to feedvater line isolation. Therefore, this failure has no effect on the results presented. 3. Failure of a Main Steam Line Isolation Valve Failure of an MSIV is assumed to increase the volume of steam piping 8 which empties into the Containment by 5,570 ft. This case is included in the analysis. I The effects of this failure on calculated Containment pressures and i temperatures were compared with the effects of the failure of one Containment spray train. Vith respect to the'maxisrms. Containment pressure, calculations showed that the adverse effects of' a main steam ,line isolation valve failure were considerably less than that of one Containment spray train failure. Vith respect to the maximum Containment temperature, no significant difference was found. beeveen the two failures. 4 Failure of one Containment Heat Removal' Train The worst single failure following a steam line break is the failure of one of the three redundant CIG.S trains. The spray actuation is assumed to occur 69.1 seconds followins the time at which the Containment pressure reaches 0.2.0 psig. The fan cooler actuation is assumed to occur 30.0 seconds after the Containment pressure reaches itu setpoint. I 6.2.1.4.6 This section is not used.

"1.,2

., est m - p .n 3.. 68

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..u n ,~, l l INSERTS j The RETRAN-03 (Ref. 6.2.1.2-8) computer code was used for the main steam line pipe break analysis. The RETRAN-03 computer code has been' verified by HL&P for calculating short-term mass and energy release rate data following a postulated Main Steam Line Break for STPEGS. RETRAN-03 also meets the requirements of ANSI /ANS-56.10-1982 to perform this function. RETRAN-03 is a best-estimate transient thermal hydraulic code designed to analyze operational transients, small break loss-of-coolant accidents, anticipated transients without l scram, natural circulation, long-term transients, and events involving limited nonequilibrium conditions in light water reactors. RETRAN-03 is a derivative of the RELAP4 code referred to in ANSI /ANS-56.10-1982. RETRAN-03 is the result of a program sponsored by the Electric Power Research Institute since 1975 to analyze thermal-hydraulic transients. Earlier versions of RETRAN 03 (e.g., RETRAN-02) have been reviewed by the NRC and received a Safety Evaluation Report. Major assumptions of the main steam line break analysis are as follows: 1. The initial conditions for the main steam system are at zero power operating conditions plus instrument error (1266 psia and 574*F). 2. The postulated high energy line double-ended rupture is assumed to reach maximum opening area within one millisecond of break initiation. Each pipe end discharges through a break area equal to the internal cross-sectional area of the pipe. 3 During the transient, the SG pressure and temperature are assumed to remain constant at the initial conditions. This is conservative, because the actual SG pressure would decrease during this event. 4. The quality of the moisture carryover is conservatively assumed to be 4% and is assumed to continue until the mass of the affected steam generator (including AFW flow)is depleted. The 4% quality assumption is taken from Appendix E of ANSI 58.2-1980. 5. The analysis continues for greater than 20 seconds until the water mass in the affected steam generator (included AFW flow) is depleted. After the mass in the effected steam generator is depleted, the mass and energy release from this generator is significantly reduced. At this time, the MSIVs are also assumed to close. This is conservative because the a MSIV closure is expected to occur at approximately 15 seconds based on a low steam line pressure signal. t l 6. AFW flow begins at the time of the break at the runout flow of 1210 GPM. 7. A sink volume is maintained at a constant pressum of 14.7 psia. _._m________. l l l l 8. Smooth commercial steel pipe, with a relative roughness of 0.00015ft is assumed for all piping. 2 9. A throat with an area of 1.38 ft is assume to limit the mass and energy release from the SG side. The feedwater line break analysis was performed using the RELAPS/ MODI (Ref. 6.2.1.2-6) computer program. Assumptions used in this analysis include: 1 1 i l ,_v. 1-,- ,.. =.,... -,. .m . -., - -, - - = - - _ c _, .. = = c-_ a _ i - x... __ 1 l STPECS UFSAR l l l REFERENCES l Section 6.2: I 6.2.1.1-1 Bechtel Power Corporation Computer Code: COPATTA User's Guide, Volume I, " Practical User's Guide", Volume II, " Theoretical i User's Guide" 1974. 6.2.1.1 2 Bechtel Power Corporation Topical Report No. BN-TOP-3, Rev. 4, " Performance and sizing of Dry Pressure Containments", March 1983. 6.2.1.1-3 Slughterbeck, D. C., Review of Heat Transfer Coefficients for Condensine Steam in a Containment Buildine Followine a loss of Coolant Accident, IN 1388, September 1970. 6.2.1.1-4 Uchida, H., A. Ogama, and Y. Togo, " Evaluation of Post-Incident Cooling Systems of Light,-Vater Power Reactors", Proceedines of the Third International Conference et the Peaceful Uses of Atomic Enerev, Volume 13. Session 3.9, United Nations. Geneva (1964). s 6.2.1.1-5 Tagami, Takashi, " Interim Report on Safety Assessments and Facilities, Establishment Project in Japan for Period Ending June 1965 (No.1)". 6.2.1.2-1 Bechtel Power Corporation, "COPDA Compartment Pressure Design Analysis", (Bechtel Computer Code),1973. 6.2.1.2-2 Bechtel Power Corporation, "Subcompartment Pressure and l Temperature Transient Analysis", Topical Report No. BN-TOP-l 4, (Rev.1), October 1977. 6.2.1.2-3 Crane Co., " Flow of Fluids", Technical Paper No. 410, 1969. 6.2.1.2-4 Idel' Chik, I.E., " Handbook of Hydraulic Resistance Coeffeients of Local Resistance and of Friction", AEC.TR-6630,1966. l 6.2.1.2-5 Shepard, R.M., H.W. Massie, R.H. Mark and P.J. Doherty, " Westinghouse Mass and Energy Release Data for Containment Design", WCAP-8264-P-A, Proprietary (June 1975) and WCAP-8312-A Revision 1, Nonproprietary (June 1975). ) 6.2.1.2-6 RELAP 5/ MOD 1 Code manual Volume 1: System Models and Numerical Methods, NUREd/CR-1826, ECC-2070, 1980. l l 6.2.1.2-7 American Nuclear Standard, " Design Basis for Protection of Light { Water Nuclear Power Plant Against Effects of Postulated Pipe Rupture", ANSI /ANS-58.2-1980. I T-.psack N ~~* t 4 6.2-58 Revision 0 _.,-.-----L. INSERT 6 6.2.1.2-8 Peterson, C.E. et al, "RETRAN-03 MOD 01 HLP-001, " developed by Computer Simulation and Analysis, Inc., for the Electric Power Research Institute, July,1991. e 4 l O TABLE 6.2.1.1-8 t i THERM 0 PHYSICAL PROPERTIES OF STRUCTURAL HEAT SINKS k FOR IACA AND MSLB ANALYSIS Thermal Volumetric Heat l Conductivity capacity Material (Bru /hr-f t

• F)

(Bru/fts,*p) 3 Amercote 90 .0375 / 49.9 (organic) Dineccote 6 0.633 21.67 (2norganic) Nutech Paint 0.1258 '28.29 Air 0.0174 0.0103 Carbon Steel 25.0 54.0 ~ Concrete' O.8 30.0 Stainless Steel 9.4 54.0 l l l ( i 6.2-74 Revision 1 i , M [.'. .wmaa.. .'E . = =- l j STPECS UFSAR TABLE 6.2.1.2-1 (Continued) $}{QRT-TERM MASS AND ENERCY RELEASE RATES Ft)R SUBCOMPAR'IHENT ANALYSES I J. MAIN STEAM LINE DOUBLE ENDED CUILIATINE BREAK USED FOR MAIN STEAM LINE SUBCOMPARTMENT ANALYSIS (Blovdm.m inch Ame arbit= w/ = 5 N 1 na l Time (s)- Ma s s 31ow.11btn/s ec) Enerry Flow (Bru/sec) A_ Avr Enthalov (Bru/lbm) l I l O '.* 0.0 0.0 0.0 l 0.0066-1.7400E+04 2.0880 1200.0 0.'01 481E+04 1.4jl E+07 1200.0 l 0.013 1.1527E 4 4832E+07 1200.0 0.025 1.1063E+04 1.3276E+07 ** 1200.0 0.05 1.0454E+04 1.2545E+07 1200.0 0.1 9.7810EM 3 1.17370E+07 1200.0 0.125 9.9450E+03 Q34E+07 1200.0 0.15 1.0381E+04,. 1.245'7E+0 1200.0 0.20 1.0072Ef04 1.2086E+07 1200.0 1 0.25 9.7.999E+03 1.1760E+07 1200.0 0.30 976181E+03 1.1542E+07 \\1200.0 l 0.35 y ' 9.5272E+03 1.1433E+07 1200.p. 0.40 9.4363E+03 1.1324E+07 1200.0% 0.45' / 9.4363E+03 1.1324E+07 1200.0 0.5 9.3454E+03 1.1214E+07 1200.0 T@@ DE"Y# i

• An enthalpy of 1200 Btu /lbm was assumed conservatively throughout the transient, i

I 6.2-129 Revision 0 I N 15 STPECS UFSAR TABLE 6.2.1.2-13 MAIN STEAM LINE AND FEEDWATER LINE SUBCOMPARTMENTS ANALYSTS Feak Time to Peak Net Differential Differential. Design Design Volume Pressure Pressure Pressure Margin i 3 Node (ft ) (esid) h eci (esid) (t) l 1 6,030.42 14rH.13. o 3 0.019 30.15 1.% % 4 1 3 1.'A. 2 16,982.55 1,44- 'l.1 ;. 0.055 15.00 3 5,332.80 13r67 )3,39 0 G17-o.o18 30.15 1.2G-9 I I ~1.8 ] 4 '.9,518.67 1-*t-1.9 3 0.M D *SN 15.00 5 6,766.11 6Fr S.Ob 0,.03 0.53l 14.25 130-9 lEnl.b 6 15,748.10 L 63 ).63 0 044 o.e'45-15.00 7 5,958.95 644L{.y S.434 o.olo 14.25 .ta1.3 ici15 8 19,100.75 L 69 176 0 444 0 *W 15.00 9 35,385.71 W O. 00 Q 050 25.50

1. O 10 32,975.70 0,4409o Gre50o.*S4 25.50 11 3.18 x 105 12 50,653.39 0.-e} Q plo 0ree-negligible N.A.

13 48,905.96 0050.08 OM negligible N.A. 14 58,770.56 0.04 O. A o 0 09-negligible' N.A. Large cargin exists. I Initial conditions for all nodes are identical. Temp. - 120*F, Press. - 14.7 psia, and relative humidley - 50% 6.2-147 Revision 0 .w s,, - - , ~ ~ . - w e e m.~.-, -., - s.- ~ FABLE 6.2.1.2-14 1 MAIN STEAN tlNE AND FEtDVATER $URCONDARTFEWT ANALYSIS JUNCTION DESCt!P1104 ", N J 'N

• Wh t_wr-e erm' wodas Vmt length /

Weed toss CoeffIclent Areg Area Flow from To fft 1 (ft'I) kontraction kupanslen "Crstine Efotel Coefflefect 1 2 154.92 0.23 0.1962 1.b 1.1 0.91 1 68.85 0.23 0.365 1.0. .365 0.85 1 5 204.43 0.025 0.3163 1.0 0.3 1.6163 0.78 2 6 555.83 0.012 0.2787 1.0 0.3 1.5787 0.79 2 11 .79 0.0064 0.2787 1.0 0. 1.5787 0.79 2 12 208. 0.61 0.165 1.0 1.165 0.92 3 4 189.34 0.16 0.138 1.0 1.138 0.93 3 7 190.72 .028 0.306 1.0 0.3 1.606 0.78 4 8 676.39 0.0 0.2791 1.0 0.3 1.5791 0.79 0 m 4 9* 468.04 0.0052 0.2791 1 0.3 1.5791 0.79 g m 4 12 36.00 0.76 0.453 1.0 1.453 0.83 c 7 5 6 179.22 0.22 .196 t0 1.196 0.91 g 5 5 7 71.65 0.19 0. 1.0 1.365 0.85 es 5 9 204.43 0.17 0.316 1.0 0.3 1.616 0.78 6 9 555.83 0.0099 0.278 1.0 0.3 1.5787 0.79 6 13 192.49 0.604 0 5 1.0 1.165 0.92 7 8 219.03 0.15 0.1378 1.0 1.1378 0.93 7 10 190.72 0.019 0.306 1. 0.3 1.606 0.78 8 10 673.08 0.009 0.279 1.0 0.3 1.579 0.79 8 13 39.82 0.81 0.443 1.0 1.443 0.83 9 10 361.35 .6 0.111 1.0 1.111 0.95 9 11 752.93 0.004 0.335 1.0 1.635 0.78 9 14 225.13 0.50 0.1837 1.0 0.3 1.1837 0.91 [ 10 11 645. 0.0043 0.348 1.0 0.3 1.648 0.77 10 14 .70 0.665 0.425 1.0 0.3 1.425 0.83 p 12 11 796.85 0.002 0.34 1.0 0.3 1.64 0.78 3 12 13 1667.57 0.004 0.2862 1.0 0.3 1.586 0.79 13 14 403.35 0.004 0.448 1.0 0.3 1.748 0.75 14 11 755.98 0.0024 0.403 1.0 0.3 1.703 0.76 i l STPEGS UFSAR TABLE 6.2.1.2-14 MAIN STEAM LINE AND FEEDWATER SUBCOMPARTMENT ANALYSIS JUNCTION DESCRIPTION (STEAM LINE BREAK ANALYSIS) Junction Nodes Vent Area Inertia Loss Hydraulic Friction From To (f tal Length (ft) Coeff. Diameter (ft) Length 1 1 2 154.92 35.630 1.1962 4.4662 2' 2 1 3 68.85 15.840 1.3650 3.1864 8' 3 1 5 204.43 5.111 1.6163 0.1296 2" l 4 2 6 555.83 6.670 1.5787 0.1296 2' 5 2 11 539.79 3.455 1.5787 D.1296 2' 6 2 12 77.16 47.070 1.3763 3.2540 2' 7 3 4 189.34 30.290 1.1380 4.4847 2' 8 3 7 190.72 5.340 1.6060 0.1296 2' 9 4 8 676.39 6.764 1.5791 0.1296 2' i 10 4 11 468.04 2.434 1.5791 0.1296 2' 11 4 12 36.00 27.360 1.4530 1.0588 8' 12 5 6 136.80 30.096 1.2686 2.7959 2' 13 5 7 58.60 11.130 1.4007 2.4298 2' 14 5 9 204.43 3.475 1.6160 0.1296 2" 15 6 9 555.83 5.503 1.5787 0.1296 2" 16 6 13 192.49 116.300 1.1650 3.9342 8' 17 7 8 219.03 32.850 1.1378 S.7274 2' 18 7 10 190.72 3.624 1.6060 0.1296 2' 19 8 10 673.08 6.058 1.5790 0.1296 2" 20 8 13 39.82 32.250 1.4430 1.8776 8' 21 9 10 361.35 93.950 1.1110 3.5896 2' 22 9 11 752.93 3.012 1.6350 0.1296 2* 23 9 14 108.71 54.355 1.3473 5.39:7 2' 24 10 11 645.72 2.777 1.6480 0.1296 2* 25 10 14 65.70 43.690 1.4250 2.1366 8' 26 12 11 796.85 1.594 1.6400 0.1296 2" 27 12 13 1667.57 6.670 1.5860 0.1296 2' ) l 28 13 14 403.35 1.613 1.7480 0.1296 2* l 29 14 11 755.98 1.814 1.7030 i 0.1296 2" i 6.2-148 r j i ..-,.......c~.mm,.... (\\

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