ML20116P049
| ML20116P049 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/16/1992 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20116P047 | List: |
| References | |
| NUDOCS 9211240273 | |
| Download: ML20116P049 (184) | |
Text
{{#Wiki_filter:-. - .. _ _ ___ _ _. . _ 2.1~ SAFETY LIMITS f BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-1 correlation and the W-3 correlation for conditions outside the range of WRB-1 correlation. The DNB correlations have R142 been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would B at a particular core locatien to the local heat flux, is indicatiwe~ti fttle ma fr tcr BH . wOf @ BN8 LNn The DNB design basis is as follows: thhcJasLbe,.atAcast rcent probability that the minimum DNBR of the limitino rod durina Condition I and II events is greater than or equal to the DNB imiVof t R142 wme [ginndised4theJRB-L4r VJforfrat' i his,.spplic lon) NB he cafrelatjofi[n
- prfela JNBR limit is e'stablished Masp ron ee ire,4pplica ee rimen),a'l datrsef/
such that there is a 95 percent probability with 95 percent to fidence that DNB not occur when the minimum DNBR is at thegDNBR limit. y@p
; J THE WlV _3he curves of Figure .
2 1-1 show the loci of n s of THERMALpoiPOWER, t
" Reactor Coolant System pressure and avarage temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
N The curves are based on an enthalpy hot channel factor,. F3H, specified in R159 the Core Operating Limit Report (COLR) and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in Ffg at reduced power based on-the expression: RTP Fh=F gy, ppg (1.p)) where P = THERMAL POWER , R159 RATED THERMAL POWER FhP=theFfg limit at RATED THERMAL POWER (RTP) specified in the COLR, and PF 3g=thepowerfactormultiplierforFfH specified in the COLR.
.SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19, 114, 138, 155 noacx osoo 92112402739211g1 goa giccocm..
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9 , POWER DISTRIBUTION LIMITS BASES M ( pesu Dt/dR/ L, W Fuel red bowing reduces the value of DNB ratio. Marginhasbeenretainedj between the DNBR value used in the safety analysis M/g] and theWRPy_r' R142 [drpelat1orfllimit #2AJM to completely offset the rod bow penalty. The applicable value of rod bow penalty is referenced in the FSAR. RIST Margin ;n excess of the rod bow penalty is available for plant design R142 flexibility. M The hot channel factor Fg (z) is measured periodically and increased by a cycle and height dependent power factor, W(z), to provide assurance that the limit on the hot channel factor, F (z), is met. W(z) accounts for the effects 9 of normal operation transients cnd was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(z) R155 function is specified in the COLR. 3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. , Radial power distribution measurements are made during startup testing and wmk periodically during power operation. The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F isn reinstated by reducing the power by 3 percent from RATED THERMAL POWER fDr each percent of tilt in excess of 1.0. 3/4.2.5 DNB PARAMETERS The limits en the DNB related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety R142 analysis DNBR limit throughout each analyzed transient. The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. 1 SEQUOYAH - UNIT 1 B 3/4 2-4 Amendment No. 19, 138, 155 l 001 h bl J
^ 2.1 SAFETY' LIMITS _
BASES-2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures pecause of the onset o.' departure. from nucleate boiling (DNB) and the resultant sharp reduction in heat transfe* coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure .'. ave been related to DNB through the WRB 1 correlation and the W-3 correlation for conditions outside the range of WRB-1 correlation. The DNB correlations have- R130-been developed to predict the DNB flux and the location of DNB for axially uniform and non uniform heat flux distributions. The local DNB heat flur, ratio, DNBR, defined as the ratio of the heat flux that would cause DN t a particular core location to the local heat flux, is indicative e .. rai pasaw %n <so a u n r. . The DNB design basis is as follows: there mu Aa g ast 9 fand ' i probability that the minimum DNBR of the limitina rod durina Cifn ,d @ition_ercen W II events is greater than or equal _to the ONBR mi R130 lbeing Afsed _(Afte WRS-1pf W ' cofr 6latig in- ispp%fthef'MBcofr6lati plicatjef1). JMe cor - lat,dn DNBR limit is established ased on t.he en_ re applicablVexperJfnental dataA)et such that there is a 95 percent probability with 95 percent co f1dence that DNB will not occur when the minimum DNBR is at the4DNBR limit. - surs Tth 1? hecurvesofFigure2.1-1showthelocialpointsofTHERMANWER, R104-Ifeactor Coolant System pressure and average temperature for which the minimum DNBR-is no less than the safety analysis DNBR limit, or the average entha)py at
-the vessel exit is equal to the enthalpy of saturated liquid. R130 The curves are based on an enthalpy hot channel factor, F g, specified in R14 the. Core Operating Limit Report (COLR) and a reference cosine with a peak of '
1.55 for axial power _ shape. An allowance is included for an increase in N R21= F AH at-reduced power based on the expression: _
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P Fh=F gy ppg (y_p)) where P = THERMAL POWER , l -RATED THERMAL POWER -R14: F P=theFhlimitatRATEDTHERMALPOWER(RTP)specifiedinthe L COLR, and PI AH
= the power factor multiplier for F specified in the COLR.
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- SEQUOYAH - UNIT 2 8 2-1 Amendment No. 21,.104, 130,
.146 March 30, 1992
I i POWER DISTRIBUTION LIMITS p o n r4 BASES w aA. i m i Fuel rod bowing reduces the value of DNS ratio. Margin has been retained
. between the DNBR value used in the safety analysis [K38tland theDgg R130
[gefrpfatJ<fnllimitlW1})1 to complGtely of fset the rod bow' penalty'. The applicable value of rod bow penalty is referenced in the FSAR. R146 Margin in excess of the rod bow penalty is availeole for plant design R130 flexibility. The hot channel factor F M(z) is measured periodically and increased by Q R21 a cycle and height dependent power factor W(z), t( a . vide assurance that the limit on the hot channel factor, F (z), is met. W w accounts for the effects 9 of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(z) function is specified in the COLR. R146l 3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. ; Radial power distribution measurements are made during startup testing and ' periodically during power operation. R2 mmir The two hour time allowance for operation with a tilt' condition greater than 1.02 but less than .1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing 9 the power by 3 percent from RATED THERMAL POWER for each percent of tilt in excess of 1.0. 3/4.2.5 DNB PARAMETERS The limits on the DNB related pr.rameters assure that each of the para-meters are maintained within the normal steady state envelope of operation . assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety R130 analysis DNBR limit throughout each analyzed transient. The 12 hour periodic surveillance of these parameters through instrument R21 , readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. l I SEQUOYAH - UNIT 2 B 3/4 2-4 Admencment No. 21, 130, herch 30, 1992
g de_A _ A -c *.JM, - - - ENCLOSURE-2' PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE-SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-92-14) DESCRIPTION AND JUSTIFICATION FOR DEPARTURE FROM NUCLEATE BOILING RATIO (DNER) LIMITS BASES CHANGE f f
Dessrintion_of_ Change TVA proposes to modify the Sequoyah huclear Plant (SQN) Units 1 and 2 technical specifications (TSs) Bar,es Sections 2.1.1, 3/4.2.2, and 3/4.2.3. Bases Section 2.1.1 will have the reference to the departure from nucleate boiling (DNB) correlation limit replaced with the design departure f rom nucleate boiling ratio (DNBR) limit that will be calculated using the new Mini-Revised Thermal Design Procedure (Mini-RTDP). Additionally, the Bases Section 3/4.2.3 will delete the DNBR value of 1.38 used in the safety analyses and the WRB-1 correlation limit of 1.17. This will be replaced with a reference to the design DNBR limit. Renaon_LotChane-SQN is requesting this change as a result of using the Mini-RTDP tc determine the DNBR margin as identified by Sefety Evaluation Check List 91-451, Revision 2. This is a new methodology being applied at SQN. The standard thermal design procedure, which has been used to show acceptable conformance to DNBR limits for Updated Final Safety Analysis Report, Chapter 15 accident analysis for the current and previous SQN cycles, will be replaced with the Westinghause Electric Corporation Mini-RTDP starting with the Cycle 7 coce design. The Mini-RTDP method generates additional DNBR margin through a statistical combination of uncertainties. Use of this method provides sufficient DNBR margin to offset the DNBR margin utilized when the core-peaking factors are increased. For future reload packages, SQN may elect to use a different staff-approved thermal design procedure. The proposed TS wording was established to provide that flexibility.
,luatification_for Change The Mini-RTDP was reviewed and approved by NRC, and a staff evaluation was issued in 1989. This method conservatively satisfies the design criterion that protects against DNB in a pressurized water reactor core while providing additional DNBR margin. A description of the Mini-RTDP, along with the specific values of the design DNBR limit, the safety analysis DNBR limit, and the DNBR correlation limit, is provided in the marked up Updated Final Safety Analysis Report (UFSAR), Section 4.4.1.1, in the attached safety evaluation check list. There is no value in maintaining the specific limit values in the bases as they are design related in nature and are best reflected in the UFSAR. In addition, any changes would be addressed in each unit's reload analysis and a 10 CFR 50.59 review performed.
Additionally, similar changes were approved by NRC on May 1, 1992, for Beaver Valley Power Station Unit 2. Note that, although the Unit 2 analysis for the present Cycle 6 operation does not use the Mini-RTDP, the proposed wording for the basis is still valid. 1
. ,y, -y . _ , . . . _ m . ._ , ATTACHMENT TO ENCLOSURE 2! TENNESSEE VALLEY AUTHORITY j i SEQUOYAH UNITS l'AND 2 INCREASE FdH TO 1.62 ANb INCREASE Fq TO 2.4 WITH l I Mini-RTDP FINAL SAFETY EVALUATION (SECL-91-451, REVISION 2' ; i
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l Weringhouse Electric Corporation Energy Systems Q L' m,g g hir. J. T. Robert 92TV *-G-0073 Nuclear Fuel Economics ET-NSL-OPL-1-92-500 Tennessee Valltj Authority October 27,1992 1101 Market Street BR 6N 60A Ref. 1) 92TV*-G-0070 Chattanooga, Tennessee 37402-2810 Tennessee Valley Authority Sequoyah Units I and 2 Increase FdH to 1.62 and increase Fq to 2.4 with Mini RTDP Final Safety Evaluation (SECL-91-451. Revision 2)
Dear Mr. Robert:
The revised safety evaluation, SECL-91-451, Revision 2, addressing increasing FdH from 1.55 to 1.62 and increasing Fq from 2.32 to 2.4 with mini RTDP is attached This revision incorporates TVA's comments, and addresses a mixed core that includes once burnec,350 psi backfill pressure standard fuel with inconel grids. , Except for the NOTRUMP Bessel Function Potential Issue (PI-92-006), the status of the dispositioning of the LOCA-related potential issues tr1nsmitted via Reference 1 continues to be applicable. The NOTRUMP Bessel Function potential issue has been resolved and information regarding this resolution is attaded. The effects of this resolution is to increase the Peak Clad Temperature (PCT) 11*F for the Smali Break LOCA. This is shown on the PCT Rackup in the Safety Evaluation (Page 40). _ If you have any questions do not hesitate to contact us. Very truly yours,
/k $./
N. R. Metcalf Project Engineer Mktg. & Customer Projects LVTomasic/cid Attachment cc: Im Evans - (W) Chattanooga Sales . C765 LYT!!tIrN2
TVA SEOUOYAH I
- NOTRUMP BESSEL FUNCrlON ERROR' j l
1
- Resolution
-l 1
i INTRODUCTION , Westinghouse has completed its evaluation of an issue affecting the NOTRUMP small break LOCA ' Evaluation Model. This information is being provided to allow affected utilities to assess individual report.ng requirements which may exist due to changes in Peak Cladding Temperature (PCT) in their small heak LOCA analysis results. BACKGROUND i During a recently completed effort, anomalous behavior was noted in the NOTRUMP runs. This behavior was eventually traced to an error in SUBROUTINE BESSJO which calculates Bessel Function values used . ' during the transient solution. During the time before this error was corrected, convergence anomalies _ were observed in NOTRUMP. It has been determined that this error was introduced in Cycle 23 of the - NOTRUMP code and that only analyses performed with this version of the code are affected. Subsequent reruns with a corrected version of NOTRUMP (cycle 24) showed that the convergence abnormalities were 1 Indeed the result of the Bessel Function error, and that the standard convergence criteria used for Evaluation Model calculations continue to be valid when the corrected code is used. EFFECTS OF ISSUE : The effect of this issue on Sequoyah Unit 1 (TVA) has been determined by a plant specific calculation to - be a change of + 1l'F. This result should be evaluated to determine reporting requirements under 10 CFR 50.46, 4 A i i t a C761 LYT /100292
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SECL-91-451, Rev. 2 Customer Reference No(s). N/A Westinghouse Reference No(s). N/A WESTINGilOUSE NUCLEAR SAFETY SAFETY EVALUATION CIIECK LIST (SECL)
- 1) NUCLEAR PLANT (S): Sequovah Units 1 and 2
- 2) SUBJECT (TITLE); Increased F Delta H from 1.55 to 1.62, increased FO From 2.32 to 2.4, With Mini RTDP
- 3) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59(b) has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.
Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed. CHECK LIST - PART A - 10CFR50.59(a)(1) 3.1) Yes 1 No _ A change to the plant as described in the FS AR? 3.2) Yes _ No 1 A change to procedures as described in the FSAR? 3.3) Yes _ No 1.A test or experiment not described in the FS AR? 3.4) Yes 1 No _ A change to the plant technical specifications? (See Note on Page 2.)
- 4) CHECK LIST - PART B - 10CFR50.59(a)(2) (Justification for Part B answers must be included on page 2.)
4.1) Yes _ No 1 Will the probability of an accident previously evaluated in the FSAR be increased? 4.2) Yes _ No 1 Will the consequences of an accident previously evaluated in the FS AR be - increased? 4.3) Yes _ No.X_ May the possibility of an accident which is different than any already evaluated in the FSAR be created? 4.4) Yes _ No.Z Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased? 4.5) Yes _ No .X_ Will the consequences of a caalfunction of equipment important to safety previously evaluated in the FSAR be increased? 4.6) Yes _ No _X. May the possibility of a malfunction of equiptaent important to safety different than any already evaluated in the FSAR be created? 4.7) Yes _ No 1 Will the margin of safety as described in the bases to any technical specification be reduced? Page 1 I I
. - . -. . -. = - SECL-91-451, Rev. 2 NOTES: i If the answ r to any of the above questions is unknown,-indicate under 5.) REMARKS and explain below. If the answer to any of the above gestions in Part A (3.4) or Part B cannot be answered in the negative, based on written safety evaluation, the change review would require an application for license amendment as required by 10CFR50.59(c) and submitted to the NRC pursuant to 10CFR50.90.
- 5) REMARKS:
Note for Part A item 3.4, F-DELTA 41 is in the CL,LR for both units. Technical Specification bases changes are provided for Units I and 2. FOR FSAR UPDATE kction: Pages: Tables: Figures: Reason for / Description of Change: Marked-up affected FSAR and Technical Specification and COLR marked pages are enclosed The answers given in Section 3, Part A, and Sectica 4, Part B, of the Safety Evaluation Checklist, are based on the attached Safety Evaluation. SAFETY EVALUATION APPROVAL LADDER: Nuclear Safety Preparer: D. ec t Date: #'EO 82 h*D l o , 3 o - q 7_,
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Nuclear Safety Reviewer: L. V. Tomas, ic Date: Coordinated with Enginects: J. Doman Date: (Signatum on me) Coordinated with Engineers: R. Anderson Date: (signature on nie) Loordinated with Engineers: F. Baskerville Date: (Signature on Ele) Page 2
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4 SECL-91-4Sl; Rev. 2. Table of Contents . PJII9 1.0 Introduction .5 2.0 Non-LOCA Evaluation 5 2.1 The Effects of an increase Fon 5 2.1.1 Partial and Complete Loss of Forced Reactor Coolant Flow (UFSAR 15.2.5 & 15.3.4) 6 2.1.2 Single Reactor Coolant Pump Locked Rotor - Rods in DNB (UFSAF 15.4.4) 7 2.1.3 Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature (UFSAR 15.2.6) 7 2.2 The Effects of an increase F, 8 2.2.1 Single Reactor Coolant Pump Locked Rotor - Fuel / Clad Temperature (UFSAR 15.4.4) 8 2.2.2 Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly - Ejection) (UFSAR 15.4.6) 8 2.3 Utilization of the Mini Revised Thermal Design Procedure 9 2.4 Conclusions 9 2.5 References 10 3.0 LOCA Evaluation 27 3.1 Large Break LOCA (UFSAR 15.4.1) 27 3.2 Small Break LOCA (UFSAR 15.3.1) 28-p 3.3 . Blowdown Reactor Vessel and Loop Forces (UFSAR 6.3.2.2) 28 l
- i. - 3.4 Post LOCA' Longterm Core Cooling Suberiticality Requirement (UFSAR -15.4.1) 29 3.5 Hot Leg Switchover To Prevent Potential Boron Precipitation (UFS AR 6.3.2.2) 29 Page 3
l SECL-91-451, Rev. 2 l 3,6. LOCA Conclusion 29-l 3,7 ' L'A References - 30 4.0 Thermal Hydraulle Design Evaluation 30-4.1 T/H Conclusion 31 4.2 T/H References 31
)l 4.3 F Increase o 31 4.4 FqConclusions 32 4.5 FqReferences 32 5.0 Technical Specification Basis Changes for Mini-RTDP 32 6.0 Assessment of Unreviewed Safety Questions 35 7.0 Conclusion 36 8.0 LOCA PCT Rackups 37 9.0 List of UFSAR, Tech Spec & COLR Mark-ups 41 9.1 UFSAR Mark-ups 41 9.2 Tech Spec & COLR Mark-ups 42 10.0 Appendix: UFSAR, Tech Specs, & COLR Mark-ups 43 10.1 UFSAR Mark-ups 44 10.1.1 Non-LOCA 44 10.1.2 LOCA 120 l-l-
10.1.3 Chapter 4 (Fuel) -150 i 10.2 Tech Spec Mark-ups -166 10.3 COLR Mark-ups 171 1 Page 4 l' 1
SECL-91-451, Rev. 2 ' l .0 - Introduction Revision 2 addresses a mixed core of the following fuel types: 1) once burned, 350 psi backfill pressure . standard (inconel grids); 2) fresh,275 psi backfill pressure V-5H (zire grids); 3) fresh,100 psi or greater backfill pressure IFBA (zirc grids). Prior revisions addressed a mixed cores identical to the one described above except that it did not consider the once burned,350 psi back pressure standard fuel with inconel grids. TVA's comments were duscussed and incorporated in Revision 1. l The above revisions do not alter the discussions, bases, or conclusions of the original safety evaluation and do not represent an unreviewed safety question. The original issue of this safety evaluaton addressed the impact of increasing Fon from 1.55 to 1.62 and I increasing Fq from 2.32 to 2.40 using the Mini Revised Thermal Design Procedure (Mini RTDP) on the l UFSAR Chapter 4, Thermal and Hydraulic Analysis, Chapter 6 and 15 LOCA and non-LOCA accident analyses for Sequoyah Units I and 2. 2.0 Non-LOCA Evaluation :l l i This section summarizes the non-LOCA reanalyses and evaluations performed for the Sequoyah Unit 1 i and 2 increased Fon, increased Fq, and Mini RTDP implementation. The increase in the design limit . l value of the nuclear enthalpy rise hot channel factor, Fon, is from 1.55 to 1.62. The increase in the i design limit value for the nuclear heat tiux hot channel factor, Fq , is from 2.32 to 2.40. j 2.1 ne Effects of an increase Fmn in general, an increase in Fon results in a decrease in Departure from Nucleate Boiling Ratio (DNBR) for a given set of thermal-hydraulic conditions. On this basis it would be expected that all transients for -q which DNBR is calculated would be affected. However, the margins obtained through the use of the. ' WRB-1 DNB correlation allow for the increased peaking factor without changing the core thermal limits. Therefore, only those transients which explicitly incorporate a value of Fog in the calculation of the thermal-hydraulic conditions existing at the time of minimum DNBR require reanalysis. These are the. 1 Partial Loss of Flow, Complete Loss of Flow (including Reactor Coolant Pump (RCP) Underfrequency), RCP Locked Rotor and Startup of an Inactive Loop at an Incorrect Temperature. .; 1 i Page 5 4 l l
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SECL-91-451, Rev. 2 He following non-LOCA accident analyses were not reanalyzed because Fon is not an explicit analysis assumption: UFSAR Chaptet 15.2.4 Uncontrolled Boron Dilution 15.2.8 Loss of Normal Feedwater 15.2.9 Loss of Offsite Power (Station Blackout) 15.2 10 Excessive Heat Removal Due to Feedwater System hialfunctions 15.2.11 Excessive Load increase 15.2.12 Accidental Depressurization of the Feactor Coolant System 15.2.13 Accidental Depressurization of the biain Steam System 15.2.14 Spurious Operation of the Safety injection System at Power 15.3.3 Inadvertent Loading of a Fuel Assembly ir.o an Improper Position The following non-LOCA accident analyses were not reanalyzed because the increase in Fon does not change the transient conditions and sufficient DNBR margin exists t ' maintain the same DNBR limit used in the current licensing-basis safety analyses: UFS AR Chgler 15.2.1 RCCA Withdrawal from Subcritical 15.2.2 RCCA Withdrawal at Power 15.2.3 Rod Cluster Control Assembly hiisaligmnent 15.2.7 Loss of External Load and/or Turbine Trip 15.3.2 hiinor Secondary System Pipe Breaks 15.3.6 Single RCCA Withdrawal at Full Power 15.4.2 hiajor Secoadary System Pipe Rupture 15.4.6 Rupture of a Control Rod Drive hiechanism Housing (Rod Cluster Control Assembly Ejection) A summary of the non-LOCA design basis calculations that were performM for Sequoyah Units 1 and 2 at an increased Foa of 1.62 follows. 2.1.1 Earth) and Complete loss of Forced Reactor Coolant Flow (UFS AR 15.2.5 & 15.3.4) The Partial Loss of Flow accident is an ANS Condition II event. The Complete Loss of Flow accident is an ANS Condition Ill event. Both of these transients have been reanalyzed in support of N-loop operation with the increased Foa. The Partial Loss of Flow transient assumes the coastdown of two RCPs during 4-loop, full-power operation while Complete Loss of Flow transient assumes the c..stdown of 4 RCPs. The analyses have incorporated the increased design Fon of 1.62 in the determination of the thermal-hydraulic conditions existing at the time at minimum DNBR. Page 6
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SECL 91-451, Rev. 2 The results of these two transients are shown in Figures 3.1-1 through 3.1-4 and 3.15 through 3.1-8, I respectively. The flow coastdown transients are shown in Figures 3.1 1 and 3.1-5, For both transients, , the FACTRAN code [1] is used to calculate the heat flux tran ient bised upon nuclear power and flow from LOFTRAN [2]. The Partial Loss of Flow transient is terminated by a low Reacter Coolant System (RCS) loop flow reactor trip. The Complete Loss of Flow transient is terminated by reactor trip on reactor coolant pump undervoltage. In both cases, the DNBR safety analysis limit is not violated at the i design Fon of 1.62. Herefore, the safety analysis DNBR limits are met and the conclusions of the UFSAR remain valid. 1 The complete loss of forced reactor coolant flow from a pump frequency decay in all four RCPs was also reanalyzed for the increased Fou. The transient assumptions for this complete loss of flow case are , identical to the complete loss of flow case above except for the flow coastdown. The Underfrequency analysis asrumed a constant frequency decay rate of 5.0 Hz/second. The transient is terminated by l~ reactar trip on RCP underfrequency. The transient results indicate that the safety analysis DNBR limit is not violated for an Fos design limit of 1.62. Therefore, the safety r.nAsis DNBR acceptance criterion is met for this loss of flow event. It is determined that the underfrequency event is the limiting loss of flow case for these analyzed conditions. The recommended UFS AR markups for the Partial and Complete Loss of Forced Reactor Coolant Flow . accidents are included in the Appendix. ; 2.1.2 Single Reactor Coolant Pumn Locked Rotor - Rods in DNB (UFSAR 15.4.4) Reactor Coolant Pump Locked Rotor - Rods in DNB is analyzed to determine the percentage of fuel rods l
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in the core that experience DNB during the accident. The Locked Rotor accident is postulated as an instantaneous seizure of one reactor coolant pump rotor at full-power conditions with all four loops in i operation. Flow through the reactor coolant pump is rapidly reduced leading to an initiation of a reactor trip on a low flow signal. The Locked Rotor - Rods in DNB transient was reanalyzed to incorporate a full-power Foa design limit of 1.62. The FACTRAN code [1] is used to calculate the heat flux transient based upon nuclear power and flow from LOFTRAN [2] Enough DNBR margin is available to maintain the number of rods in - DNB to less than 10%. Therefore, for an increased Fon, the Locked Rotor - Rods in DNB analysis adheres to safety analysis limits and is bounded by previous radiological dose release analyses. l. 2.1.3 Startuo of an Inactive Reactor Coolant Looo at an incorrect Temperature (UFSAR 15.2.6) ; The Startup of an Inactive Loop transient is an ANS Condition II event analyzed to demonstrate that'the DNB design basis is met. The transient has been reanalyzed incorporating the increased Foa consistent with a full-power design limit of 1.62. The FACTRAN code [1] is used to calculate the heat flux ~ ; transient based upon nuclear power and flow from LOFTRAN [2] The transient results are shown in - Figure 3.1-9 through Figure 3.1-12. The reactor trip is assumed to occur on low coolant loop fWi wnen the power range neutron flux exceeds the P 8 setpoint. The P-8 retpoint is conservatively assumed to ' l be 84 percent of rated power which corresponds to the nomin.at setpoint plus 9 percent for nuclear Page 7 I
,I,
SECL-91-451, Rev. 2 instrumentation errors. The DNDR safety analysis limit is not violated at the increased Fon of 1.62.- Therefore, the safety analysis DNBR !imit is met and the conclusions of the UFSAR remain valid. UFSAR markups for the Startup of an Inactive Loop event analysis are included in the Appendix. 2.2 The Effects of an increase F9 The peaking factor Fo is the ratio of maximum local or " hot spot" rod power to average rod power. The full power design limit Fq is explicitly assumed in two UFSAR non-LOCA events. UFSAR Chanter 15.4.4 Single Reactor Coolant Pump Locked Rotor - Fuel / Clad Temperature 15.4.6 Ruptare of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)
- These two events are adversely affected by an increase in the full-power design limit Fe and require evaluation to ensure that cladding integrity and fuel melting at the " hot spot" are mainW*ied within the -
applicable safety analysis limits. 2.2.1 Single Reactor Coolant Pumn Locked Rotor - Fuel / Clad Temperature (UFSAR. 15.4.4) The Lockeo Rotor event is classified as a Condition IV event. The UFSAR analysis for the Locked Rotor ^
- fuel / clad temperature transient was conservatively analyzed with a 3.0 Fq including allowances for calculational uncertainty and nuclear power peaking due to densification. The results of the current -
UFS AR Locked Rotor analysis show that the maximum clad temperature at the core hot spot is 2026*F. _ This is less than the limit of 2700*F. The amount of Zr-water reaction is small, calculated to be 0.70% ) l by weight. Because these UFSAR fuel and clad temperature transient results were analyzed with a 3.0 Fq, the increase in Fo to 2.40 is bounded and the applicable fuel / clad temperature safety criteria continue to be met. The UFSAR conclusions for the Locked Rotor analysis continue to remain valid. 2.2.2 Ruoture of a Control Rod Drive Mechanism Housine (Rod Cluster Control Assembly Eiectice , (UFS AR 15.4.6) The Rod Cluster Control Assembly (RCCA) Ejection event is classified as a Condition IV event for which conservative criteria are applied to ensure that there is little or no possibility of fuel dispersal in the cociant, gross lattice distortion, or severe shock waves. These criteria are, i
- 1. Average fuel pellet enthalpy at the hot spot below 225 cal /gm for unitradiated fuel and 200 cal /gm for irradiated fuel.
f
- 2. Peak reactor coolant pressure less than that s.hich would cause stresses to exceed the faulted condition -
stress limits. 3 Fuel melting will be limited to less than 10% of the fuel volume at the hot spot even if the average fuel pellet enthalpy is below the limits of criterion (1) above. Page 8 I o
SECL-91-451, Rev.- 2 Ei i To conservatively bound the Fo increase to 2.40, the UFS AR BOL and EOL full-power RCCA Ejection cases . vere reanalyzed with a 2.625 Fe including allowances for calculational uncertainty and nuclear power peaking due to densification. The FACTRAN code t ]l was used to calculate the fuel and clad - transient based upon nuclear power from the TWINKLE code {3]. A detailed discussion of the method of analysis can be found in Reference 4. The nuclear power transient and hot spot fuel, average fuel, and clad temperature versus time for the EOL full-power case (which is the limiting case in terms of the fuel melt criterion) are presented in Figures 3.2-1 and 3.2-2. A summary of parameters used in both of the full-power rod ejection cases and the results of these cases, are presented in Table 3.2-1. He analysis results demonstrate that the i .! applicable UFSAR fuel melting ard stored energy limits are not exceeded for the increased Fq . Therefore, the UFSAR conclusions for the RCCA Ejection analysis remain valid. jl! i UFSAR markups for the RCCA Ejection analysis are included in the Appendix. l 2.3 Utilization of the hiini Revised Thermal Design Prxedure The hiini Revised Thermal Design Procedure (hiini RTDP) is described in Reference 5. htini RTDP . is similar to the current fixed value DNB design basis metnodology in that initial condition assumptions , for power, flow, temperature, pressure, and bypass flow are :.ssumed to be at their extreme values when ' used in the plant transient analyses, The hiini RTDP differs from the fixed value DNB design basis methodology in that it statistically combines peaking factor uncertainties, et al, with the DNB correlation ~ i ur'ecrtainty. The statistical convolution of uncertainties results in a net increase in DNBR margin for any 4 event which uses the hiini RTDP. . As hiini RTDP results in a net increase in DNBR margin, its implementation does not adversely affect any of the UFSAR events, Therefore, the UFSAR conHusions remain valid with the implementation of the hiini RTDP. 2.4 Non-LOCA Conclusions An increased Fos from 1.55 to 1.62 and an increased Fq from 2.32 to 2.40 has been evaluated utilizing hiini RTDP to determine the effects on Sequoyah Units 1 and 2 Chapter 6 and Chapter 15 accident analyses. The following non-LOCA transients were reanalyzed for the increased Fon and F q : Partial and Complete Loss of Flow-(including RCP Underfrequency), RCP Locked Rotor - Rods in DNB, Startup of an Ihactive Loop, and full-power RCCA Ejection. As previously demonstrated in this safety evaluation, all applicable acceptance criteria for these events have been satisfied and the conclusions presented in the UFSAR still remain valid. He increased Fos and Fqwill have no impact on the remaining nou-LOCA transients because sufficient DNBR margin is used to maintain the safety analysis DNBR limits. Thus the proposed increase in Fon to 1.62 and increase in Fqto 2.40 does not constitute an unreviewed safety question and the non-LOCA-accident analyses, as presented in this report, support the proposed change. Page 9 d
-'m, , .w-- - - , , - , . . , , _ . , , .,,9, _ . .g. _
A SECL-91-451, Rev. 2 l 1, 2.5 Non-LOCA References
- 1) Hargrove, H. G., "FACTRAN A Fortran IV Code for Thermal Transients in a UO2 Fuel Rod,"
WCAP-7908-A, December 1989.
- 2) Burnett, T. W. T., et al., "LOFTRAN Code Description", WCAP-7907-P-A (Proprietary),
WCAP-7907 A (Non Proprietary), April 1984.
- 3) D. H. Risher, Jr., R. F. Barry, " TWINKLE '- A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979-P-A (Proprietary), WCAP-8028-A (Non Proprietary), January 1975.
- 4) D. H. Risher, Jr., "An Evaluation of the Roa Ejcion Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision I-A, January,1975.
- 5) S. Ray, " Mini Revised Thermal Design Procedure (Mini RTDP)," WCAP-12429-A, October 1989.
i i ( I-E l l Page 10
, ,, . ,i~ . _ , . , , , - - - _ - .. .,,-. .
.. _ _ . . _ =_. . . _ . . . .. ~ . _ _ . . . _ , ._
SECL 91-451, Rev. 2-1 Table 3.2-1 Summary of Rod Ejection Analysis Re'sults I Parameter BOL HFP EOL.HFP Total Core Peaking Factor 7.11 7.88 Ejected Rod Worth, (pcm) 200 210 Maximum Fuel Pellet 4121 4056 l Average Temperatt're, ('F) Maximun Fuel Pellet 4971 4879 Centerline Temperature, (*F) Maximum Clad- 2319 2267 Average Temperature, (*F) Maximum Fuel Enthalpy, 181 177 , (cal /gm) Maximum Fuel 7.0 8.7 Centerline Melt, (%) 2 Page 11
..x --..c. - . . . - . . -.--.--.2. ..-..._.-..-_.....--.-..-a,....:...
Figure 3.1-1 Flow Transients for Partial Loss of Flow, All loops Oporating, Two Loops Coasting Down
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5 - .- 5 i I '
= !
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.J 4 .
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fr __--_______--------.-- --- - - - ' _ - -~
Figure 3.1-2 '\ Nuclear Poter All loops Ope and ating,Pressurt2er Pressure Two Loops Coasting Down offor Partial lo Floa, 1 4-l 1 E ! , 1 E ; 5 i i y j-b B6 l. 2 p 6 - I .)
*4 I 0
a l i' Y 2-0. 0 1 2 3 4 5 j 6 7 8 TrWC 9 'O (SEC) 2400. --
-a 2300. ,- ? = 2200.+ '
a 5
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A u 1 F 2000, 3-, 1900. , L 1800. .
- 0. 1, 2. t' l
- 3. 4, S. 6, 7 8.
TtuC 9. 10. (SEC) l, - 1 I -
'4
-l
Figure 3.1-3 Average and Hot Channel Heat Flux Transient for Partial loss' of Flow, All loops Operating, Teo Loops. Coasting Down 1.4 1.2 - 1 d I. bp
$ acM .S<
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= c' s
e N .4-
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TlWE (SEC) 1.4 1,1 j, T' db v:
@m h .8" y ..-
.2- -
00 1 2 3 4 5 6 7 8 IIidt 9 19 (KC) (4 i
Figure 3.1-4 r)NBR versus Time for Partial loss of Floc, All Loops Operating, Teo Loops Coasting Doan 1 26 --- I 2.4 t I 2.2 1. i.a 1.6 1.4 x 4 1.2 -- 0 1 2 3 e s e-~ y t'4 a , go (350)
/
zn
=.. .- . . . . - -. . - .. .. .
Ff Reacter Vessel Flow.versu,ure 3.1-5 i s Time for All Loops Operating, All loops Coasting Down, Complete loss of Flow s s - d z 12 j 3
~
1 M 8 u !
=
2 S - o
.J
.a j 6- -
m 5 '% f 4< ,
.. Q l u
7 , , 4' I o 0 1 2 3 4 l 5 5 7 8 TiWE 9 io (SEC) i 1 l. l
..v-i.
(. " l
Figure 3.1 Nuclear Power Transient and Pressurifer Pressure Tra., stent for - All Loops Operating, All Loops Coasting Docn, Complete Loss of Flow i E 2 r g . i u
$ I p-si 5 s. ,
d a t l j 2 g
%i .
- 100 "
200
=
i ' W 3 S: a 100 b y :: . t900.
'IUC o, t. 2, 3- T 6. 9- **
' 7', y g '(SEC 1\
o
-. . . . - . . - . . . _ . - - .. - - . . - . - - - _ . ~ - . . . . . - - . - _ . - . . - . . .
1 Figure 3,1-7 i Average and Hot Channel Heat Flux Transients for All loops i Operating, All 1. cops Coasting Docn, Complete Loss of Floa- ' 1& , 1 2+
- i 1.
I e i w W S- , a ' w - W .2a 6" at w a: - W
~
t o ; k , 2 0' o 1 2 3 4 5 6 7 8 9 'O flWE (SEC) 1.4 12 " g, a lt { 8 " E5w J
.6" p-
=c
.y .4"
.2 "
^ 00 1 2 3 4 5 6~ 7 8 9 10
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.:._. = - - .
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Figure 3.1-8 DNBR versus Time for All Loops Operating Allloo95CoastingDown,CompletelossofFioe i 4 l i 4 5- , i 2.41 22
~
2. 1 8-1.6< 1.4< O 1 2 3 4 5 6 7 8 9 '3 riac (sec) 1 9
Nuclear Power versus Time, Startup of an inactive Reactor Coolant Pump I 1 4 1 1 2< - S h 1 , g S. 5
.6 - -
h 8 U 4, , 5 t 2 3. 0 5 to 15 20 23 ~) i TlW (SEC)
- '~ ' ~ ,, ,_,._.a-+>
Figuro 3.1-10 Core Heat Flux versus Time, Startup of an inactive Reactor Coolant Pump 3 4_- I i
-4
?z .,
'5 , l 8
w g s . D l h 6- I w k , y 4., i I 2 ,, l
- o. I O 4 I 6 8 10 12 to 16 IS 7814 ($(C) 23 i
I l
' figure 3.1-11 Core Average Temperature versus Time. ] Startup of an inactive Reitter Coolant Pump l l 700 680 5 660-d a 640-E
" 620 E! .
w , 600-u g Se0- - a w $60-E
" 540-520-00 0 5 10 15 20 25 30 flWC (SEC) i 5
1 O
Figure 3.1-12 Pressuriger Pressure versus Time. Startup of an inactive Reactor Coolant Pump 2400 -- 2J00. 2 E. 2200. m 2100.
.Y 2000 1900 1500.
O, S- 10. 15, 20. flW 25' 30' (SCC) l
-- - - - _ _ _ _ , _ ...s.. ---~~~~-^ , , , . - . - -
~~~*'9tblu 4*s*60 versus (ime for Startup of an inactive Re__ _
actor Coolant Loop
, 1 L
25-i 2.4 ., t 2,2- i u 2. m I s< 16 - f 4 , 12 0 1 2 3 4 5 6 7 l frWE 5 9 ' O (SEC) i d 1 1
, . - - - - - - . - . . . , . . . _ - . , - . . - - . . - _ . . . , . . m- - . - ----, - . . - - , , , . , . _ _ , , - - - , - , .
_ . ..tigure 3,2-1_ . - Nuclear Power Transient, HFP EOL Red Ejection Accident 3 - 25 - u , g .
?
u c 3 1.5
- f 1.
5-O. I 2 3 4 5 7 6 ilME 9 'O (SEC)
"t/ Chin -
l t o
Figure 3.2-2 Hot Spot Fuel, Average Fuel, and Clad Tempurature versus Time, EOL HFP Rod Ejection Accident 6000 5000 Nelting 4800'F Fuel Center Temperature ; C 4000-w 5 j 3000 Fuel Avg Temperature 7 w I m 2000-C1ad Temperature 1000
- i. !
O - - - 0 1 2 3 -4 5 6 7 5 iiME (SEC) ,,,,,,,,. i f e T* -=..---;-... ._
- - - - - - - - . _ - - - - . - - . - - - . _ _ - - - . . - . - _. -- _ ~ . - . _ _ . SECL-91451, Rev. 2 3.0 1.OCA Evaluation Revision 2 addresses a mixed core of the following fuel types: 1) once burned, 350 psi backfill pressure standard (inconel grids); 2) fresh, 275 psi backfill pressure V-5H (zire grids); 3) fresh,100 psi or greater backfill pressure IFDA Similar fuel with higher burn ups would be bounded by these fuel types. The following LOCA related accidents will be con:idered: large break and small break LOCA; reactor vessel and loop blowdown forces; hot leg switchover to preclude boron precipitation; and post LOCA long term core cooling minimum flow and subcriticality. It must be shown that the higher peaking factors will not increase the probability of occurrence or the consequences of any previously analyzed accident. And that they will not lead to the possibility of an accident different from any previously analyzed. The preceding revision addressed the increase in peaking factors for a mixed core condition identical to the one described above--except that it did not consider the once burned, standard fuel. Because of its inconel grids which generally result in approximately a 100'F penalty in the large break LOCA analysis, the standard fuel is likely to bound all other fuel types. 3.1 Larce Break LOCA Analysis - FSAR CHAPTER 15.4.1 A large nipture of the reactor coolant system (RCS) piping is a hypothetical event postulated to demonstrate that the calculated performance of the emergency core cooling system is adequate to mitigate the consequences of such a scenario. The effect of increased core peaking factors during a hypothetical large rupture of the RCS piping is examined to ensure that the bases and assumptions of the calculation remain valid, and that a conservative approach will yield values that conform to 10CFR 50.46 standards. Following a large rupture of the cold leg RCS piping, the RCS depressurizes in approximately 30 seconds to a pressure nearly equal to the containment pressure. During this time the core flow reverses and the core is cooled by a two-phase mixture flowing down through the core, up the down:omer and out the break. When the reverse core flow period ends, end of bypass occurs, and the lower plenum can begin filling with cold safety injection water. After the lower plenum fills, and the bottom of the core is reached, the process of reflooding the core begins. The peak cladding temperature (PCT) for large break LOCAs occurs during the reflooding portion of the transient N elevations near or above the mid-plane of the core (6 feet) for the Westinghouse ECCS Evaluation Models. This is due in part to the fact that the chopped cosine power distribution has been demonstrated to be limiting (see Addendum 1 to Reference 1). A large break LOCA anrj ysis for Sequoyah Nuclear Plant Unit I was performed using the 1981 Evaluation Model with BASH (Reference 1). The analysis assumed a reactor power level of 102% of 3411 Mwt, Fq=2,40, FAh=1.62, and uniform 10% steam generator tube plugging. The DECLG limiting discharge coefficient of CD =0.6 was analyzed. The PCT calculated in this analysis was 2%9'F for a mixed core in which the V-5H fuel is bounding (that is, no standard fuel is present) (Reference 6). With a taixed core including the once burned, standard fuel, the limiting PCT has been determined to be 2169'F, With respect to the 2069" F PCT of the current licensing basis analysis, this represents a 100'F analysis penalty due to the inconel grids. When cortpared to the current licensing basis PCT of 2013*F, this figure also represents a penalty of approximately 160'F for the increase in Fq and Fah. Until such Page 27 _ . _ __ _. - . - _ . _ _ _ _ _ _ - _ _ ~ _ . - ._ _
SECL-91451, Rev. 2 time as the standard fuel is removed from the core the limiting PCT will remain at the figure of 2169. Conformance to 10CFR 50.46 limits is documented in the margin utilization documentation provided. 3.2 Small Break LOCA Analysis - FSAR Cltanttt.Ehl A small rupture of the RCS piping is postulated to demonstrate that the calculated performance of the ECCS design is adequate to meet the requirements of 10CFR 50.46 for these more realistic LOCAs. In the Westinghouse small break LOCA ECCS Evaluation hiodel (Reference 2), the ECCS performance is a function of the break size, core power level and operational performance. For the Westinghouse small break LOCA ECCS Evaluation hiodel, the primary system response to a small rupture of the RCS piping is typically a rapid depressurization to a pressure equal to the hot leg saturatlou pressure. Usually, the break energy removal capability in conjunction with the secondary heat removal capability exceeds the decay heat production, and the RCS will depressurize to a pressure slightly above the secondary pressure. This ensures that the steam generator secondary sides continue to remove decay heat, producing a condition of quasl<quilibrium pressure at which the primary system tends to stabilize prior to the venting of steam through the broken leg loop seal. Following the venting of this steam, core boil-off may continue, possibly exceeding the safety injection mass flow rate and resulting in a boil-off core uncovery transient. The depth and duration of uncovery can be influenced by several parameters (e.g., initial power level, break size, safety injection flowrates, etc.). Reference 3 contains the results of several analyses for typical break sizes, power levels and system capabilities in Westinghouse PWRs for the Westinghouse small break LOCA ECCS Evaluation hiodel. The current licensing basis small break LOCA analysis for Sequoyah Units 1 and 2 was performed using the 1985 NOTRUh1P model (Reference 2). The analysis assumed a reactor power level of 102% of 3411 hiwt, Fqn2.7, FAh=1.7, and uniform 15% steam generator tube plugging. The analysis determined the limiting break size to be a 3 inch diameter cold leg break. Note that this analysis bounds any mixed core at peaking factors of Fq=2.4, FAh= 1.62. Conformance to 10CFR 50.46 standards is maintained as indicated in the margin utilization sheet provided. In addition to this safety evaluation, this rackup also reflecu the resolution of the NOTRUhiP Bessel Function Potential Issue (PI-92 006) which results in an increase of 11" to PCT for the Small Break LOCA. 3.3 Blow @wn Reactor Vessel and Ieon Forces - FSAR Chapter 3.9 The blowdown hydraulic forcing functions resulting from a loss of coolant accident are considered in Section 3.9.1.5 (Analysis hiethods Under LOCA L oadings), and Section 3.9.3.5 (Blowdown Forces Due to Cold and Hot Leg Break) of Volume 4 of the Sequoyah Units I and 2 FSAR. Neither the mixed core condition described above nor the increased peaking factors will have a significan effect on the LOCA blowdown hydraulle loads or on the results of the LOCA hydraulic forces calculuions. I l Page 28 l
. - - - - . _. .. .~. . -_ -- - . -
l SECL-91-451, Rev. 2 3.4 Post LOCA lenzterm Core Cooline Suberiticality Rtquirement - FSAR CIMI'rER 15.4.1 The Westinghouse licensing position for satisfying the requirements of 10CFR Part 50 Section 50.46 Paragraph (b) Item (5) "Long Term cooling" is defined in WCAP-8339 (Reference 4, pp. 4-22). De Westinghouse commitment mat the reactor will remain shutdown by borated ECCS water residing in the .omp following a LOCA seference 4). Since credit for the control rods is not taken for large break LOCA, the borated ECCS water provided by the RWST and Accumulators must have a concentration that, when mixed with other sources of water, will result in the reactor core remaining suberitical assuming all control rods out (ARO). This requirement is not affected by the increase in core peaking factors or the particular type of mixed core. 3.5 Ilot 142 Switchoffr To Prevent Potential 11oron PreclD ilation - FSAR Chapter 6.3.3.2 During a large break LOCA the plant switches to cold leg recirculation after the RWST switchover setpoint has been reached. If the break is in the cold leg there is a concern that the cold leg injection water will fall to establish flow through the core. Safety injection entering the broken loop will spill out the break, while SI entering the intact cold legt will circulate around the downcomer and out the break. With no flow path established through the core the fluid in the core remains stagnant. As steam is produced in the core from decay heat, the boron associated with the steam will remain in the vessel. Thus, as water is boiled off with no circulation present in the core, the boric acid concentration increases. The boron concentration in the vessel will increase until the solubility limit of the boric acid solution is reached, at which time boron will begin to precipitate. As the boron precipitates, it may plate out on the fuel rods, which would adversely affect their heat transfer characteristics. The purpose of the hot leg recirculation switchover time analysis is to provide the time at which hot leg recirculation must be established to prevent boron precipitation in the core. Neither an upgrade in core peaking factors nor mixed core as described above will affect this calculation. The current time for hot leg switchover remains applicable. 3.6 LOCA Conchulon The evaluation performed for the LOCA-related accidents applies to a mixed core design in which once l burned, 350,si fill pressure, standard fuel has been established as the bounding fuel type. The LOCA-l related accidents within the scope of Safeguards Analysis have been examined to determine whether or i not conformance to 10CFR 50.46 criteria can be demonstrated with the mixed core design at increased j core peaking factors of FAh=1.62 and Fq=2.40. As detailed in the preceding discussion it has been concluded by Westimghouse that this change will not increase the consequences of any previously malyzed accident. The included margin utilization documentation demonstrates conformance to 10CFR 50.46 standards for the mixed core design at increased peakmg factors. Page 29 L
SECL-91-451, Rev. 2 3.7 LOCA Referents
- 1. WCAP 10266-P-A Rev. 2, with Addenda, Desspiata, J. J., et al., "The 1981 Version of the Westinghouse ECCS Evaluation hiodel Using the BASH Code", hiarch,1987,
- 2. WCAP 10054-P-A (Proprietary), WCAP 10081 (Non Proprietary), Lee, N., et al., 'f
" Westinghouse Small Break ECCS Evaluation hiodel Using the NOTRUhiP Code", August, ,
1985.
- 3. WCAP-Ill45-P-A, WCAP-il372 (Non-Proprietary), Rupprecht, S. D., et al., " Westinghouse Small Break LOCA ECCS Evaluation hiodel Generic Study with the NOTRUh1P Code",
October,1986.
- 4. Westinghouse Technical Bulletin NSID TB-86-08, " Post-LOCA Long-Term Cooling: Boron Requirements," October 31,1986.
- 5. WCAP-8301, Bordelon, F. hi., et al., "Locta-IV Program: Loss-of-CoolantTransient Analysis,"
June,1974.
- 6. TVA 92-021, 01/31/92, "Large Break LOCA Analysis Worst Case Break with BASH Final Analysis Results."
4.0 Thermal Hydraulle Desien Evaluation The proposed change to the Sequoyah Units 1 and 2 Core Operating Limits Report (COLR) Section 2.6 which impacts the DNBR calculations is the value of FE enthalpy rise hot channel factor, determined from the following equation: FN = 1.62 [(1.0 + 0.3 (1.0 - P))] where P = Thermal Power / kated Power F5 = hteasured value of FAH obtained by using the moveable incore detectors to obtain a power distribution map with approximate uncertainties. The measured radial peaking factor limit increase from 1.5s to 1.62 has a direct impact on DNBR calculations. The thermal-hydraulic design method for Sequoyah Units 1 and 2 is the mini-Revised Thermal Design Procedure (hiini-RTDP), Reference 1, which replaces the previous Standard Design Procedure. The mini-RTDP approach is a statistical approach which combines the uncertainties on the nuclear, thermal and the fuel fabrication paraneters with the uncertainties on THINC-IV and the transient codes. The resulting overall DNBR uncertainty is then combined statistically with the DNB correlation Page 30
SECL-91431, Rev. 2 i uncertainty to denne the design limit DNBR. The implementation of the minl RTDP procedure generates additional DNBR margin to offset the 4.5% increase in FAH, Therefore, the currere core limits in the ! Technical Specification, Figure 2.1 1, and the DNBR limits for the FSAR Reference 2 analysis remain ' valid for Sequoyah Units I and 2. He limit on the nuclear enthalpy rise hot channel factor, F[y, will take the following form in Section 3[ 2.6 of the Core Operating Limits Report (COLR): .I tl ; F[y .5; 1.62 [(1 + 0.3(1-P))] ; where P = Thermal Power / Rated Hermal Power, and ) F[u = Measured valued of F[y obtained by using the movable incere detectors to obtain a power distribution map with appropriate uncertainties. De increase in the measured radial peaking factor limit will allow additional flexibility for fuel management and for determining core loading patterns. Cycle specific reload core analysis performed in accordance with the methodology described in References I and 2 demonstrates that the new radial peaking factor limit is met. No changes to the current methodology are required as a result of this change. 4.1 T/H Conclusion in summary, the effect of increasing F[y from 1,55 to 1.62 at rated thermal power has been l ac:ommodated in the safety analyses by generating additional DNBR margin by utilizing the mini RTDP procedure. He current DNB-related safety limits, including the core limits in Figure 2.1-1 of Sequoyah Units I and 2 Technical Specifications, remain valid. 4.2 I/H Referengga
- 1. S. Ray, " Mini Revised Thermal Design Procedure (Mint RTDP)," WCAP-12178-P A, October 1989.
- 2. Sequoyah Nuclear Plant Final Safety Analysis Report, USNRC Docket No. 50-327/328.
l i- { Page 31 E .
t SECL-91451, Rv. 2 4.3. E InnsAtt ne limit on the heat flux hot channel factor, Fq(z), will take the following form in the Section 2.5 of the COLR: Fn(z) 1 (2.40/P) * (K(z)) for P > 0.5, and F9 (z) 1 (4.80) * (K(z)) for P 10.5 where P = Thermal Power / Rated Thermal Power, and K(z) =the function obtained from Figure 3 of the COLR for a given core height location. : The increased total peaking tactor limit will allow additional flexibility in fuel management and core operation as well as accommodate the increased radial peaking factor limit. The increased radial peaking factw discussed above will result in increases in the total peaking factor. Fe(z), experienced in the core. - Many cycles of cores representative of the Sequoyah Units indicate that the new increased Fn(z) limit will be met for Sequoyah reload cores operating with the increased radial peaking factor F[a limit. Actual Sequoyah reload cores will employ the usual methods of enrichment variation and burnable absorber usage to ensure compliance with the new COLR peaking factor limits. Cycle specific reload core analysis - performed in accordance with the methodology described in References 1 and 2 will demonstrate that the new total peaking factor limits will be met. 4.4 FO Conclusion The increase in nuclear enthalpy rise hot channel factor and heat flux hot channel factor will be met on a cycle specific basis. No changes to the current methodology will be required. The increased total peaking factor limit will have no impact on other key _ safov parameters used as input to the FSAR Chapter 15 accident analyses. 4.5 FO References
- 1. Sequoyah Nuclear Plant Final Safety Analysis Report - USNRC Docket No. 50-327 and 50-328
- 2. Davidson, S. L. (Ed.), et al., " Westinghouse Reload Safety Evaluation Methodology,"
WCAP-9272 P-A, July 1985. 5.0 Sequoyah Units 1 and 2 Technical Specification Basis Chances for Minl-RTDP The following Basis changes in the Sequoyah Units 1 and 2 Technical Specifica ions reflect the implementation of the Mini-RTDP method: Page 32
.. - . _ - _ _ - - . - - . - . _ . - ~ . - - .-. _-_ -
1 SECL 91-451, Rev. 2 5.1 Section 2.1 S AFETY LIMITS (nace B 21) Reference to the correlation limit is replaced with the design limit which is calculated using the Mini-RTDP procedure. 5.1.1 Basis for the change: The DNB design basis is such that there is at least a 95 percent probability that DNB will not occur on the limiting fuel rod during normal operation, operational transients, and any transient conditions arising from faults of moderate frequency (Condition I and 11 events) at a 95 percuit confidence level. This criterion is met by limiting the minimum DNBR to a 6esign limit DNBR. The value of the design limit DNBR depends on the thermal design method selected. To produce margin to offset penalties such as those due to rod bow and transition core, and for core design flexibilty, the design !!mit DNBR values are increased to values designated as the safety analysis DNBR limit. The safety analysis DNBR limits are used when performing the Thermal Hydraulics and reactor safety analysis. In the Standard Thermal Design Procedure (STDP) the design limit DNBR is set equal to the correlation DNBR limit. Then, all design parameters are treated in a conservative way frw a DNBR standpoint; that is, uncertainties are added to all parameters to give the lowest minimum DNBR. , The Mini Revise Thermal Design Procedure (Mini RTDP), which as reviewed and approved by NRC, and a staff evaluation ' was issued in 1989 conservatively satisfies the design criterion that protects against DNB in a PWR core, while providing DNBR margin. In the mini-RTDP, the uncertainties on nuclear and thermal parameters, the fuel fabrication parameters are statistically combined with the uncertainties on THINC-IV and the transient code. The resulting value is then combined Statistically with the DNB correlation uncertainty to define the design limit DNBR. The rest of the parameters such as; reactor power, flow, temperature, pressu.e and bypass flow are excluded from the statistical combination process. The appropriate dervative plant initial condition assumptions are used for these parameters. Therefore, the design limit DNBR defined by mini-RTDP is different from the DNBR limit of the correlation used, I. S. Ray, ' Mini Revised Thermal Design Procedure (mini-RTDP)". WCAP-il2178-P-A, October 1989. I Page 3'4 l
SECL-91-451, Rev. 2. 5.2 Section 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY liOT CHANNEL FACTORS (page B 3/4 2-4) The following sentence: Margin has been retained between the DNBR value used in the safety analysis (1.38) and the WRB 1 correlation limit (1,17) to complety offset the rod bow penalty. Is replaced by: Margin has been retained between the DNBR value used in the safety analysis and the design DNBR limit to completly offset the rod bow penalty. 5.2.1 Dmes for chang The argument presented to support the change in Section 2.1 applies. l l Page 34 l u l L- .
}
P SECL-91-451, Rev. 2 ) 1 1 1 6.0 Assessment of Unreviewed Safety Ouestions ; i
- 1. Will the probability of an accident previously evaluated in the SAR be increased? -
. No. As addressed in this safety evaluation, all non LOCA transients affected by en increased Fou ",
and Fo were reanalyzed utilizing the NRC approved Mini RTDP (WCAP-12i78 P A) or evaluated and found to adhere to the safety analysis acceptance criteria. The increased Few and Fqhave no impact on the remaining non LOCA transients. Therefore, the probability of an accident occurring that is already evaluated in the SAR will not increase. Since the total core peaking factors affect only 1/4 ft. length of one fuel rod, the hot channel peaking factors affect one fuel rod in rod heat up models, and all analysis acceptance criteria continue to be met, these changes will not increase the probability of the occurrence of the LOCA.
- 2. Will the con.Uquences of an accident previously evaluated in the SAR be increased?
No. Per the discussion presented in the Evaluation section, all the applicable non-LOCA acceptance criteria are still met for the transients evaluated and for the events reanalyzed. Additionally, no new limiting single failure is introduced by the proposed change. Therefore, there is no potential for an increase in the consequences of an accident previously evaluated in the SAR. . The increase in core peaking factors would not adversely affect the safeguards systems actuations or the accident mitigation capabilities important to LOCA events. This conclusion is based on the fact that the following countermeasures intended to limit the , consequences of a LOCA (as described in the Sequoyah Nuclear Plant FSAR) would not be compromised, a Reactor trip and borated water injection complement void formation in causing rapid reductbn of power to a residual level corresponding to fission product decay heat. , b. Injection of borated water provides heat transfer from core and prevents excessive clad temperature. Therefore, the consequences of a LOCA will not be increased.
- 3. May the possibility of an accident which is different than any already evaluated in the SAR be created?
No. Increasing the Fon and Fq does not introduce a new accident initiator mechanism. Thus, no e new accident will be created. Page 35
I SECL 91-451. Rev. 2 4 %ill the protabi!ity of a malfunction of equipment important to safety previously evaluated in the S AR be increased? No. Increas!ng the Fos to 1.62 and increasing the Fo to 2.40 will not adversely affect the operation of the Beactor Protwtion System, any of the protection setpoints, or any other device required for accident mitigation. increased core peaking factors wili not increase the probability of the malfunction of any equipment imprtant in safety as concerns the LOCA.
- 5. Wili the tort,equences of a malfunction of equipment important to sdety previously evaluated in the -
SAR be increased? No. As discussed in the responses to questions 2 and 4, there is no possibility of increasing the consequences of a malfunction of equipment for an increase in Fon and Fqas defined in the attached safety evaluatioe. Increased core peaking factors will not increase the consequences of the malftmetion of any equipment important to safety as concerns the LOCA.
- 6. May the possibility of a malfunction of equipment important to safety different than already evaluated in the SAR be created?
No. As discussed in question 4, an increase in Fou and an increase in Fq will not impact any other equipment important to safety.
- 7. Will the mergin of safety as described in the bases to any technical specification be reduced?
No. As c'.iscussed in the safety evaluation, the proposed increase in Fon andqF will not invalidate any of the non-LOCA conclusions presented in the UFSAR accident analyses. Thus, there is no reduction in the margin of safety. , Increased core peaking factors will not create the possibility of a malfunction of equipment ,: important io safety as concerns the LOCA different than that already evaluated in the FSAR. I i Since the calculated PCT of 2069'F is within the limit of 2200*F set by 10 CFR 50.46, an increase in core peaking factors does not reduce this margin of safety as defined in the bases to any technical specifications. 7.0 Conclusions The preceding evaluation demonstrates that all safety analyses acceptance criteria have been met ' and supports the Tech Spec and COLR changes provided.- i l Page 36 e- + - - w- .e p. . . ~ w . m - - . , , , - v4.tr #
- . - - - - .. . . . . . . . - -- - - . . . . . . . . . . - ~ _ = - - . . . . . __ _ , _ _ . . . .
( .; 4' SFCL-91451 Rev. 2 b f t I
,t 8.0 LOCA PCT Rack-Ups .
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STANDARDIZED
SUMMARY
FORMAT SHEET FOP REPORTING OF 10CFR50.46 MARGIN UTILIZATION LARGE Dr,EAK LOCA Pl. ANT NAME: SEOUOYAH Pl. ANT UNIT I UTILITY NAME:.IENNESSEE VALLEY AUTHORITY A. ANALYSIS OF RECORD PCT = 2169 'F Comments: Evaluation Model: B ASH . FQT = 2.40 , FAH = L62_, SGTP = 10 %, Other; once burnt. standard fuel. 350 osi fill cressure B. PRIOR LOCA MODEL ASSESSMENTS - 1991 APCT = + 0*F (Permanent Assessment of PCT Margin - Letter #: TVA 91 181 )
- 1. SG-TUBE SElSMIC/LOCA ASSUMPTION APCT = + 20'F C. CURRENT LOCA MODGL ASSESSMENTS - 02/1992 (Permanent Assessment of PCT Margin - Letter #: )
~ D. 10CFR50.59 SAFETY EVALUATIONS (Permanent Assessment of PCT Margin) APCT = + 0'F E. CURRENT LOCA MODEl. ISSUES (Temporary Use of PCT Margin):
- 1. LB LOCA POWER DISTRIBUTION ASSUMirTION APCT= NOTE _L
- 2. CORE AVERAGE ZlRC-WATER REACTION APCT= MOTE 2
- 3. BOL IFBA IMPACT ON SAFETY ANALYSIS APCT= NOTE 3 F. OTHER LOCA RELATED MARGIN ALLOCATION (Temporary Use of PCT Margin):
- 1. ECCS FLOW INCONSISTENCIES (1989) APCT= NOTE 4
- 2. ECCS FLOW MEASUREMENT INACCURACY ('./ 4) APCT= NOTE 5
- 3. COLD LEG STREAMING TEMPERATURE GkeD'EN T APCT= + 10*F G. OTHER MARGIN ALLOCATIONS (Temporary Use of PCT Margin):
1 ANALYSIS MARGINS USED: APCT= + 0*F
- 2. PLANT MARGINS USED: 5% SGTP (SECL-88-417 Rev.11 APCT= - 20'F l 3. FUEL MARGINS USED: APCT= + 0*F LICENSING BASIS PCT + MARGIN ALLOCATION PCT = 2179 'F Natu:
l 1. O'F PCT Margin allocated to dat:. on the basis of the core design axial offset.
- 2. An additional 0.7% of Zr-H O2 margin allocated for all plants for reasonable assurance of safe operation within the licensing basis.
No ECCS Analysis PCT margin is allocated for this issue on the basis that, more likely than not, 4 the issue will be resolved without any change to the ECCS analysis results or the ECCS Evaluation l Model. l- 3. No margin allocated to date because of high burn-up currently on the Cycle 6 IFBA fuel, i 4. The ECCS Analysis of record has addressed this issue be modelling safety injection pump line i flow imbalances as specified by TVA.
- 5. No ECCS Analysis PCT margin is allocated for this issue; no specific safety evaluation has been perfer ned.
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. . . _ . - .. - . - . . - _ . - - - - . - . . . . . - . - . . ~ . - . _ . - . - . .
STANDARDIZED SUhih1ARY FORh1AT SHEET FOR ; REPORTING OF 3CFR50.46 hiAP. GIN UTILIZATION LARGE BREAK LOCA i PLANT NAh1E: SEOUOYAH PLANT UNIT 2 UTILITY NAh!E: TENNESSEE VALLEY AUTHORITY A. ANALYSIS OF RECORD PCT = 2069 Comments: Evaluation hiodel: BASH , FQT = 2.40 . F4H = 1.62 , SGTP = 10 %, Other: fresh. generic. V-5H fuel 275 osi fill pressure .- B. PRIOR LOCA h10 DEL ASSESSh1ENTS - 1991 APCT = + 0'F (Peemanent Assessment of PCT hiargin Letter #: TVA-91-181 )
- 1. SG-TUBE SFIShilC/LOCA ASSUh1PTION APCT= + 20'F C. CURRENT LOCA h10 DEL ASSESShiENTS - 02/1992 (Permanent Assessment of PCT hiargin - Letter #: ._)
- 1. LB LOCA BURST & BLOCKAGE ASSUh1PTION APCT= + 0'F D. 10CFR50.59 SAFETY EVALUATIONS (Permar. cat Assessment of PCT Margin)
- 1. Letter: SECL-90-537 Issue: Loose Fuel Paits APCT= + 45'F E. CURRENT LOCA MODEL ISSUES (Temporary Use of PCT hiargin):
- 1. LB-1.OCA POWER DISTRIBUTION ASSUMPTION APCT= NOTE 1
- 2. CORE AVERAGE ZIRC WATER REACTION APCT= NOTE 2 F. OTHER LOCA RELATED MARGIN ALLOCATION (Temporary Use of PCT Margin):
- 1. ECCS FLOW INCONSISTENCIES (1989) APCT= NOTE 3
- 2. ECCS FLOW MEASUREMENT INACCURACY (1990) APCT= NOTE 4
- 3. COLD LEO STREAMING TEMPERATURE GRADIENT APCr= + LQ'F G. OTHER MARGIN ALLOCATIONS (Temporary Use of PCT Margin):
- 1. A.NALYSIS MARGINS USED: APCT= + 0*F
- 2. PLANT MARGINS USED:.1% SGTP (SECL-88-417 rev.1) APCT= - 20*F
- 3. FUEL MARGINS USED: APCT= + 0'F LICENSING BASIS PCT + MARGIN ALLOCATION PCT = 2124 'F Hates:
- 1. 0*F PCT Margin allocated on basis of the core design axial offset.
- 2. An additional 0.7% of Zr-H O2 margin allocated for all plants for reasonable assurance of safe operation within the licensing basis.
No ECCS Analysis PCT margin is allocated for this issue on the basis that, more likely than not, the issue will be resolved without any change to the ECCS analysis results or the ECCS Evaluation Model.
- 3. The ECCS Analysis of record has addressed this issue be modelling safety injection pump line flow imbalances as specified by TVA.
- 4. No ECCS Analysis PCT margin is allocated for this issue; no specific safety evaluation has been performa!.
Y
- - . .. - - - _ _ . -- _- -- . - . ~ . - l STANDARDIZED
SUMMARY
FORMAT SHEET TOR REPOkTING OF 10CFR50.46 MARGIN UTILIZATION SMALL BREAK LOCA , PLANT N AME: SEOUOYAH PLANT UNIT 2 WOUNDS UNIT 11 UTILITY NAME: TENNESSEE VALLEY AUTHOillTY A. ANALYSIS OF RECORD PCT = 2105 'F i Comments: Evaluation Model: NOTRUMP_, FQT=7.70 -, FAH =L2L, SGTP =._lL%, 4 B. PRIOR LOCA MODF.L ASSESSMENTS - 1991 APCT= + 0'F (Permanent Assessment of PCT Margin - Letter #: ) C. CURRENT LOCA MODEL ASSESSMENTS - 02/1992 (Temporary Assessment of PCT Margin - Letter #: )
- 1. SB-LOCA ROD INTERNAL PRESSURE ASSUMIYTION APCTo + 0'F ,
- 2. SB LOCA BURST AND BLOCKAGE APCT= + 26'F (NOTE 5)
- 3. SECONDARY SIDE MODELING IN SB/ INPUT CORREGIONS APCT= - 147'F
- 4. SB-LOCA NOTRUMP DESSEL FUNCTION APCT= + ll'F D. 10CFR50.59 SAFETY EVALUATIONS (Permanent Assessment of PCT Margin)
- 1. Letter: SECL-90-537 issue:_ Loose Fuel Parts APCT = + 37'F .
E. CURRENT LOCA MODEL ISSUES (Temporary Use of PCT Margin): .
- l. CORE AVERAGE ZIRC-WATER REACTION APCT=EOTEI
- 2. RCCA INSERTION ASSUMPTION IN SB LOCA APCT= NOTE 2 F. OTHER LOCA RELATED MARGIN ALLOCATION (Temporary Use of PCT Margin):
- 1. EtCS FLOW INCONSISTENCIES (1989) APCT= NOTE 3
- 2. ECCS FLOW MEASUREMENT INACCURACY (1990) APCT= NOTE 4
- 3. COLD LEG STREAMING TEMPERATURE GRADIENT APCT= + 2*F-G. OTHER MARGIN ALLOCATIONS (Temporary Use of PCT Margin):
- 1. ANALYSIS MARGINS USED: APCT= + 0'F
- 2. PLANT MARGINS USED: APCT= + 0*F
- 3. FUEL MARGINS USED:_, APCT= + 0'F LICENSING BASIS PCT + MARGIN ALLOCATION PCT = 2034 'F H214
- 1. An additional 0.7% of Zr-H O2 margin allocated for all plants for reasonable assurance of safe operation within the licensing basis.
2, No ECCS Analysis PCT margin is allocated for this issue on the basis that, more likely than not, the issue will be resolved without any change to the ECCS analysis results or the ECCS Evaluation Model.
- 3. The ECCS Analysis of record has addressed this issue be modelling safety injection pump line flow imbalances as specified by TVA.
- 4. No ECCS Analysis PCT margin is allocated for this issue; no specific safety evaluation has been performed. .
- 5. PCT margin allocated on basis of the core design axial offset.
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SECL 91-451, Rev. 2 9.0 List of UFSAR, Tech Spec and COLR Mark-Ups 9.1 UFSAR Markups 9.1.l List of Figures 9.1.2 For the Startup of an inactive Reactor Coolant Purnp the following changes were made: Reference 3 updated, Table 15.21 (sheet 3), and Figures 15.2.6-1 through 15.2.6-4 were replaced. 9.1.3 For the Complete Loss of Flow event the following changes were made: Text markups, Reference 8 updated, Table 15.3.4-1 revised, and Figures 15.3.4-1 through 15.3.4-3 were replaced. 9.1.4 For the Partial Loss of Flow eveat the following changes were made: Text changes, Reference 3 update, Table 15.2-1 (sheet
- 2) revised, and Figures 15.2.51 through 15.2.5-3 were replaced.
9.1.5 For the RCCA Ejection event the following changes were . made: Text changes, References 25 and 29 updated, Tables 15.4.1-12 (sheet 3) and 15.4.6-1 were revised, and Figures 15.4.6-1 through 15.4.6-2 were replaced. 9.1.6 For the " Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident) " event, changes were made to the text of section 15.4.1, Tables 15.4-1-1,15.4.1-3,15.4.1-6 and 15.4.1-7 and Figures 15.4.1 1 through 15.4.1-20. 9.1.7 For the Thermal and Hydraulic Design, changes were made to : Tables 4.1-1 (sheet 1),4.3.2-2 (sheet 1) and 4.4.2-1 (sheets 1 and 2) Pages 4.3-18, 4.4-1, 4.4-la, 4.4-8,- 9, -24, -31, 38, -39. i Page 41 I 4_
I SECL-91451, Rev. 2 I i l 4 9.2 Affected Technical Specification and COLR Technical Specification Technical Specifications l Page B 2-1 Page B5/4 2-4 COLR -- (Unit 1) Page 2 of 10 ' Page 3 of 10 COLR , (Unit- 2) Page 2 of i1 , Page 3 of 11 i i T t Page 42 . 1 y,e- - ,. , -. w,,, 4 m y .~ .w_ . . . , y c, .- .. , , y . r , , , w , . ,. , , , --,
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i 10.0 Appendix: UFSAR Markups, Tech spec Markups. COLR Markups l, I t i E l e e t 9 9
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SECL 91-451, Rev. 2 l l 10.1 UFSAR Markups 10.1.1 Non LOCA G (
_ 4 _ . _ _. ._ . _ -_ . _ . . . - _ . - . ._ . ._ . - __. _ _ _ i SON-G l LIST OF FIGURES (Continued) Number Ilfle I 15.2.2-7 Effect of Reacti ity Insertion Rate on Mtnteue ON3R for a Rod ! Withdrawal Accident from 107. Power
- I 15.2.3-1 'l Transient Response to Orooped Rod Cluster Control Asseecly i 15.2.4-1 RCs Boron concentration versus Tlas - 80L Ecut11brium XE and Clean Initial Conditions l I
15.2.4-2 K ,, vs Time - Ecullibrius XE and Clean Conditions Following Trip from fyll Power l i 15.2.a-3 Nuclear Power (Detector Indication) After Trip versus Time
.2.5-1 All Loops Operating. Two Loops Coasting Down, Flew Coast versus Tire 15.2.5-2 A, s operating. Two Loops Coasting , lux Transients Rifbt.t 15.2.5-3 All Loops oper g Tw Looo sting Down, DN8R versus Time >
! L xrt A 15.2.5-4 All But one Loop ng, apops Coasting Down Flow Coast - l ' OQwn versus T ( . 15.2.f-5 A one Loop Operating, Two Logs Coastl I i
- 15. All 8et One Loop Operating, Two-Loops Coasting Down, DN8 us l Tlas-t Wyu n.2.e; n.a. a :a.a:.;::;. , ,,e=
- wi6 fgM f I3.2.k2 CI..oi. In sousdi; l_ G CeM L;i I,.U .C..; E '.Q'TZ IIZIG _
l 15.2.7-I Loss of Load Accident With Pressurtzer Spray and Power Operated Relief Valves, 8eginning of Ltfe 15.2.7-2 Loss of Load Accident, WIth Pressurizer Spray and Power Operated i Relief Valves, 8eginning of Lifa i 15.2.7-3 - Loss of Load Accident With Pressurtter Spray and Power Operated 1 Re*1ef Valves, End of Ltfe , 15.2.7-4 Loss of Load Accident. With Pressurizer Spray and-Power Operated l Rollef Valves End of Life l ' l e'* 15-10 0113F/COC4 03 I . - -- - ... - . . . - - - . - - .
UFSAA Inserts - LIST 0F FIGURES insert A: ' 15.2.5-1 Partial Loss of Forced Reactor Coolant Flow, Reactor Vessel and Loop Flow vs. Time 15.2.5-2a Partial- Loss of Forced Reactor Coolant, Flow, Heat Flux vs. T'.:ae (Hot Channel) 15.2.5-2b Partial Loss of Forced Rt tor Coolant Flow, Heat Flux vs. Time (Average Channe 15.2.5-2c Partial loss of Forced Reactor Coolant Flow, Nuclear Power vs. Time 15.2.5-3 Partial loss of Forced Reactor Coolant Flow, DNBR vs. Time Insert B: 15.2.6-1 Startup of an Inactive Reactor Coolant Puer,, Active Loop 1 Flow vs. Time;, Core Flow vs. Time 15.2.6-2 Startup of an Inactive Reactor Coolant Pump, Core
, Average Temperature vs. Time;, Nuclear Pewer vs. Time '
15.2.6-3 Startup of an Inactive Reactor Coolant Pump, Heat Flux vs. Time,-Pressurizer Pressure vs. Time 15.2.6-4 Startup of an Intetive Reactor Coolant Pump, DNBR vs. Time tc
~. - .. . ~
i SON-S QST OF RGURES (Condnued) Nmnhar Ihla 35 Landing a MaqWon 2 Ae::Wi into a ,"G.1 Position Near Core Ponchery , t-15.3.4 1 N Loops Operreng, M Loops Coesting Down, Coastdown versus Time kpbC(.15.3.4 2 M Loops , M Loops Coasting , w,% au T-g--{ 15.3.4-3 M Loops OperstW, M Down, DNGR versus Time i 15.3.4-4 M but One Lofop 0 6, M Coseting Down, Mow c As. versus Time 15.3.4 5 M One Loop Operating, M Loops Consting
, Mux Trenaients 15.3. ONOR versus Time M But One Loop Opersdng AE Loope Consting Down 15.4.1 1 Comperiment Pressure 15.4.1 2 (
Ik8 Pressure DECLG C.=0.6 - l 15.4.1 3 Core Roweste DECLG, C.=0.8 i 15.4.1-4 Cold Lag Accumuistor Rowrses - DECLG, C.=0.6 b^ q 15.4.1 5 Core Preneurs Drop DECLG, C.=0.8 15.4.1 6 Break Mass Rowrsen DECLG, C,=0.6 i. 15.4.1 7 amak Energy Mowrote - DECLG, C.=0.G 15.4.1 8 Nuestaed Core Power - DECLG, C.=0.6 - 15.4.1 0 Care and Downsomer Uguid Levels DECLO, C.=0.8 15.4.1 10 com Inte Ruid Veoocity DECLG, C.=0.6 las input to the thenre ana#ysis codel 15-14 kl
UFSAA Inserts - LIST OF FIGURES Insert C: 15.3.4 1 Complete Loss of Forced Reactor Coolant Flow, Reactor Vessel Flow vs. Time 15.3.4-2a Complete Loss of Forced Reactor Coolant Flow, Nuclear Power vs. Time 15.3.4-2b Complete loss of Forced Reactor Coolant Flow, Heat Flux vs. Time (Average Channel) 15.3.4-2c Complete loss of Forced Reactor Coolant Flow, Heat Flux vs. Time (Hot Channel) 15.3.4-3 Complete loss of Forced Reactor Coolant Flow, DNBR vs. Time W6
SON-4 usT OF FicuREE (Cononued) Number Ittis 15.4.2-4 Trenament Response to Steam Une Broek Downstroem of Flow Measuring Nozzle Mth Safety injection and Without Off Site Power (Case el 15.4.2 5 Transient Response to Stosm Une Broek at Exit of Steam Generator With Safety inpoction and Mthout Off Site Power (Case d) 15.4.2 6 Deleted by Amendment 8 15.4.2 7 Deleted by Amendment 8 15.4.2 8 Main FeocNino Rupture Accident Core Aversee Temperature Praesurtrer Pressure and Water Voeums as a Function of Time 1E.4.4-1 As Loops Operating, One Locked Rotor Pressure Varous T':N 15.4.4 2 A5 But One Loop Operating 1 Locked Rotor Pressure versia, T1rne 15.4.4 3 As Loops Operating, One Locked Rotor Core Row versus Time *- 15.4.4 4 ' As Loops Operating 1 i N Rotor Rum Transients versus p 15.4.4-5 As but One Loop Operating, O< J Locked Motor Core Row versus Tlme 15.4.4-6 As tut One Laop Operating, One Lacked Motor Heat Rux versus Time 15.4.4-7 As Loops Operating, One Locked Rotor Time versus Clad Temperature
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m UFSAR' Inserts - LIST OF FIGUR15 Insert 0:
-15.4.6-1 Rod Cluster Control Assembly Ejection, Nucle'ar Power vs.
Time (EOL, HFP) 15.4.6-2 Rod Cluster Control Assembly-Ejection,< Fuel and Clad-Temperature vs. Time (EOL, HFP). 15.4.6-3 Nuclear Power versus Time for V5H, EOL, HZP 15.4.6-4 Fuel and Clad Temperature versus Time for V5H, E0L, HZP
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SON 4 Ramenvwv Caefflements A tonserveWwefy large absolute value of the Doppler only power coefficient is used (See Table 15.1.2 2). The total integrated Doppler reacovity from 0 to 100% power is assu to be 0,016 A. The lowest absolute magnatude of the moderstor temperature coefficient 10.0
- A is assumed since this results in the maximum hot soot heat flua dunne the initiat part of the transient when the mensmum DNSR is reached.
Aow Cematdown . The flow coastdown analysis is based on a momentum balance around each reactor coolant loco
/Q and across the reactor core. This triomentuar bahnce is com4Mned with the cononutty equation, a pump momentum balance and the pumer characterisocs and is based on high estimates of system pressure losses.
w / S g BandI8 ob , ,4 & . b The cad'ated 6uence of . vents is shown on Tabie 15.2.i for b anaeysed. ngures 15 pow.2$ 1 through 18.2.53 show the loop coastdowns, the core flow coastdowns, the nuc k er.coastdowns and the avwege and hot channel heat flux coastdowns for M of the esses. l pmenemum DNgR for each of the cases is not less than the safety analysis limit. N ,, l b 15.2.5.3 Canchiniana a A The anefysis shows that the DNgR wiE not decrease below the safety analysisJimit at any time dwing the t enesent. Thus there wie be no cladding damage and no release of fisaeon d products to-the Reactor Comient System. {( l 15.2.6 Rewium of An ' = ' ; ?- m r'- ' u 5 - 15.2.6.1 Idensification of Cauana and Amendant Damerioden if the plant le operating with one pump out of service, eere le reverse flew through the inactive loop duJ to the pressure efference across the reacter venast. The said leg temperature in an inactive loop is idendeel es the osed leg tempererJre of the aedwo loose (the reacter core inist temperature). If the reester is operated at power, there _le a temperature drop across the steam generseer in the innesive leap and, wide the reveres flow, to het les temperature of me inactive : loop is lower then the reacter core inist temperature. Administredwe precedures require that the unit be brought to a lead of less then 25% of fue power prior to suer $ig a pump in an inscWwe leep in order to bring me inactive loop het leg toneersture - l closer to $p cort best temperature. Stordng of an ide reacter cocient pump without I~ 51 15.2 19
. -- - .-. .- - - - - - - - - . - - _ - - ~ . . - -
SON-4 brtnging the inacdve loop hot les temperature close to the core inlet temperature would rudt in the inpoc6en of cold water into the core wf*:n causes a rapid reactivity anaertion and subsequent poww incremen. This event is closarned as an ANS Condicon ll inodont (a feuk of moderets frequency). Should the startue of an inactive reactor coolant pump scodont occur, the transaent will be 2 terminated automa0cally by a reactor tne on low coolant loco flow when the power range neutron flux Itwo out of four channels) exceeds the P 8 setpoent, whech has been prowous)y reset for three loop opersoon. g 15.2.6.2 Anahan of EMaets and ConMua Mathed of Anaiveig This transient is anatyred try three dienal computer codes. The LOFTRAN Code (4) is used to calculate the loop and core now, nucieer power and core pressure and temperature transsents follovnno the startup of an idle pump. FACTRAN (3T is used to ceioulete the core heet Aux traneaent based on core now and nuclear power from LOFTRAN. The THMC Code (see Section 4.4) is then used to calculate the DNOR durin0 the trenaient based on syneem conditions (pressure, temperature end flowl ceiculated by LOFTRAN and heet Aux as calcudeted by FACTRAN. Plant characteristics and initial condhions are discussed in Section 15.1.2. In enter to einein conservseve resuks for the startup of an insedve pump acc6 dent, the fenowing assumptions are ~ made: ( '
- 1. Inielal condions of meatmum core power and reactor osoient everage temperatures and minimum reacter coolant pressure resuhin0 in minimum inidei margin se DNS. These values are consistent with the maalmum senedy state power level aAowed wth three loose bi opersdon. The high inkiel power elves the gresenet temperature eptforence between the core intet temperature and trJ inactive loop hot le0 temperature.
- 2. Fotowing inhiedon of startup of the ido pump, flow in the insedve loop reverses and accelerates to its nomined fus now value in approalmedoy 7 seconds.
- 3. l A C.. 'di arge moderator density coefficient (see Seeden 15.1.88.
- 4. A conservedvely smet labsolves vetus) negedve Despier power costnedent (see Seedon 1 5.1.61.
j' i
- 5. The inhief rescaer ocelant loop flows are at the appropriate values for one pump out of .
servies. .
- 6. The ressaar trip le assumed to occur on low cooient loop How when the power rance neutron '
fluz amanada the P 8 seapoint. The P-4 sequesit :s conservatively aneumed to be 84 percent of rated power which corresponds to the nomenal setpoint plus 9 percent for nucoser instrumenteden emers. 5 1 15.2 20
_ __ _ _ _a_ i SON 8 i Meauha The resuhe feeswmc the startup of an idle pung with the above listed assumotions are shown in Figures 15.2.41 through 15.2.6-4. _ As shown in these curves, dunne the first part of the transient, the increase in core flow with cooier water resuita in an ircrease in nuclear power and a decrease in core average water ternperature. The minernum DNM durine the transient is consideratW greater than the safety analysis limet. See Seccon 4.4 for a descnocon of the ONBR design basss. Reactive; mi for the indctive loco startue accident is due to the decrease in core water tomtveture. Durtne ?w tronaaent, this decrosse le due both to the increase in reactor coolant flow ary, as the inactive loot, flow reverses, to the colder water entering the core from the het leg side tr older temperature aide der to the start of the transiend of the steem generator in the inactive . l< op. Thus, the reactivity k'aertion rete for this transient changes wwth time. The resuttant core n eclear power trenaient, con,wted with consideration of been anoderster and Doppler reacdvity fa wsneck effects, is shown ons icure 15.2.6 2. Tho caiculated sequence of events 8er this accident le 9%vn on Table 15.21. The trenaient res 4ts ieustreted in Figures 15.2.61 vier #i 16.2.6-4 lndcate that a stabstred pient condhion, wr h the reactor tripped, is approached rapa9y. Plant cooldown may subsequentty be achieved by fe aowmc normal shutdown procedures. 15.2.8.3 Canctuminna The to thetrenaient DNS safetyresuhe show anstysis Amk.that the core is not adverseh affected, i.e., there '7 considerable nee .
- r. -
15.2.7 Lama Of Essamal Elantrical Land AnsuOr Twtdna Trts 15.2.7.1 lentineanian of Cauana and Ameidam 1 4 Y ua or on d pieni an ,=uh from inn eisc.ieeno.d or from a iurbin. tne.
, g7 For either case the reactor cooient power is avatable for sendnued operosion of plant components such as l The case of of e5 AC power (seeden bischoud is anelveed in Subsecton 15.2.9, generseer need, en kneedets feet eteswo of the turtene
.(Q M g
. contrei velves we esaur. This sudden redueden in ensam nour, resuhing in an increase in preneurs and esmooreams annem As e resust, the heet transfer rees in me steem senerator is seusine the rosesor :n., -- no se stes, which in tum causse coolant empension, insurge ,and MCS preneurs etes .
Fora Qe messear would be tripped afrecdy lurdens below __ , - 50% powert from a dortved from the antine aueneens ce preneurs (Weednehouse Turtene turtune stop The tuttine esop volves close on lose of autosses et pressure actuated by of a et poemade antine trip aionais. Turtenerirto initieden alonais include:
. Generosor Tate
- 2. Low condenser vacuum b
15.2 21 r -- -na . - , - < , - - , . . --r- + ---r
SON-8 4 6. Turtnne Land Turtnne load was assumed constant und the electro hydraulic pove the throele velve wide open. Then tutt>ne load drops as s'. sam pressure drops. 7. Reector Trip Reactor Trip was initiated by low pressurmer pressure assumed at a conservettvely low value of 1775 pse. ' 4 E11hlt1A The transent response is shown in Figures 15.2.141 and 15.2.14- 2. Nuctear power starts decreassne immediately due to boron inpocoon but steam flow does not decrease untd 1 into nuclear thepower transsent causes Twhen the turtune throttle valve goes wide open. The mamatch betwee pressurtzer water level, and pressuruer pressure to drop. The low pressure trip set poir.t is reached at 64 seconds and rods start moving into the core at 66 second After trip, pressures and temperatures slowly rise since the turtnne is tripped and the reactor is producane some power due to deleyed neutron flasens and decay host. h(VI 15.2.14.3 Canciunions Results of the anahele show that sounous safety injection with or without enmedista reacto presents no hazerti to the integrity of the Reactor Coolant System. A DN8 of ratio fiseson products le never to the reactorlesscoolantthansystem. the initial value. Thus there wE be no claddne /y w e reeaar doe. o= = i,nroeiseiety, m iow ,eener = w. e aliusted. T=
= =wn. and - .=e= coowown e,.y dn. ,ecove,y f,.m = ino,dem.
s 15.2.15 References 1. W. C. Gangloff, 'An Evolustion of Antioloseed Operational Trenesents in Westinghouse Pressurimod Weter Reactors," WCAP 7486, May 1971. 2. D. H. Risher, Jr. - R. F. Geny, 'TWWGCLE A Mule Cimensional Neutron Kinetics Comoveer Code,' WCAP 79MA (Proprietary), WCAP 402h 96en Proprietary), January 1975. .g (h \ e vdnto + Meet A L a-+wu,*i, ru ,.=, A r . . c1 : r ._ T _ . , ac, r : ra =m. fn !"70.-
- 4. >
Bumea, T. W. T., et eL, '1.0FTRAN Code Descripelen", WCAP 7907 P A (Proprietary), WCAP. 7907 A 96en Proprietary), Aart 1884 g-4 15.2-44
-u.--.
. . - . . . ... . . - . . - . .~ .- -. . - - . . _ - - - . . .- - .- . - -
9 FSAR 15.2.6 - Startup of an Inactive Reactor Coolant Loop
-Inserts for reanalysis due--to' increased FAH Insert A : Change Reference 3 to the following: -
- 3. Hargrove, H. G.,
- Transient:; in a UO"FACTRAN - A Fortran '!V Code fer Thermal 2 Fuel Rod," WCAP-7908-A, December 1989. .
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O i SQN 4 I TAOLE 15.21 (sheet 3) (Canenwee TIME SEQUENCE OF EVENTS FOR CONDmON 11 EVENTS A ccidawit fgggg Time Mae ) 2~ Stariuo of an inactive Reactor Cooient Looo inmedon of pump startuo 0 at an incenset Temperature Power reaches P 8 trip seipemt 2. 8 Rode begin to drop 3.3 Minimum DNW4 occurs 4.0
~.
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FNAL SMETYM M ' UNITS 1 and 2 Starse of an inesow gg Neester Ceedent Pung Mouss 15.1.6 4
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TIME (SEC) SEQUOYAH FIPeAL SAFETY ANALYSIS REPORT UMTS 1 and 2 honup of an homene Roemor Coceant Punp Heat Pts a Time; Prenewteer Pressure vs. Time Figure 15.2.6-3
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e SQN 15.3.4 Canalete 1.oss of Forced Reactor Cooljat Flow 15.3.4.1 Identification of Causes and Accident Descrietton. 4 A cceplete loss of forced reactor coolant flow any result from a simultaneous loss of electrical supplies to all reactc,r coolant pumps. If the reactor is at power at the time of the accident the immediate effect of loss of coolant flow is a rapid lacrease in the coolant temperature. This increase could result in 015 with subsequent fuel damage if the reactor were tot tripped promptly. The following provide necessary protection against a loss of coolant flow accident: 't j
- 1. Undervoltage or underfrequency on reactor coolant pump power supply busses.
- 2. Low reactor coolant,1000 flow.
The reactor trip on reactor coolant pump bus undervoltage is provided to protect against conditions which can cause a loss of vo tage to all " reactor coolant pumps, l.a., station blackout. This function is blocked below approximately _10 percent power (Permissive 7).. The reactor _ trip on reactor coolant pump underfrequency is provided to open the. reactor coolant pump breakers and trip the reactor for an . underfrequency condition, resulting from frequency disturbances on the- . major power grid. The trip disengages the reactor coolant. pumps from the power grid so that the pumps' kinetic energy is available for full coastdown. The reactor trip on low primary coolant 1000 flow is provided to protect against loss-of flow conditions which affect only one reactor coolant loop.- It also serves as a backup to the undervoitage and'underfrequency L- trips. This function is generated by two out of thre; zicu flow signals o - per . reactor coolant loop. Ahove approstaately 35 percevrt. power I (Permissive 8). low flow la any loop will- actuate a reactor trip. . 8etween approstettely 10 percent power and 35 percent power-(Permissive 7 i and Permissive 8), low flow la any two loops will actuate a reactor trip. l Normal power for the reactor coolant pumps is-supplied through busses l-free a transformer connected to the generator. . Each pump is on a __ _
- separate bus. Ingen generator trip occurs, the busses are automatically transferrog to a transformer supplied from external power lines, and' the pumps will contless-to supply coolant flow to the core. following any.
turbine trip, where there are no electrical faults which require tripping' the generator from the network ' the generator remains connected to the network for approstaately 30 seconds.. The reactor-coolant pumps roasta-connected to the generator thus ensuring full flow for 30 seconds after m the reactor _ trip before any transfer is made.
/
O 15.3 3 0116F/COC4- , l- n = ~ - - - - - " - - - - - ~~^ ~ " ~ " ~
) 1 SON.4 1
15.3.4.2 M2 ef I"n and consmounes 1 Memed of Anshraig 1 i The complete opersoon. less of flow transeent has been analyzed for a loss of f our pumos with four locos in The transient is anatyred by three diottal computer codes the calculated flows. the nucteer power ,
. trenasent andis us me of reactor tne based on !
temperature transsents. The FACTRAN code Afs' a- syseem sm"we and ' trenaient based on the nucesar power and now from LO used to cusuiets the heatl flux is then . flux from FACTRAN and the flow from eLOFTRAN. , on the host ' T WL% The DNW transiente presented represent aton is used the minimum of the t for DNOR ecal or themble ceil, I The method of arietysis and the assumptions made re0erding inhieff ope reactivity coefficiente are idemical to those discussed in section 15 2 emc ng conditione and loss of supply to e5 pumps st power, a reactor trip is ., eptactueesd that nsw ow by -alth temperature for conservatism.underfrequency. An additional excepoon la n t a overage {6tti W [g 15.3.41 through 18._C - ,
= cf events la shown hee on Table 15.3.41 for the cases c=d
" I h redo. as;a 2.=L v.he fv z, of N eme for each of == m. . t fka=* 6 L == not ehenno0, an'w d DNS e -
3 C~.~. cwra for each of the cases le not less then= wie on d . i9 15.3.4.3 Canchamiens the DNOR does not decreens belowrcedthe reactorasfety cooient flow,enefysis no cied demsos or reisese of flesien products a the nessear Caetentl3 . Syeese 5.0 Anaeyens has shown that for frecuency decoy reens less ihan MHs/s determewd that, for the weret cose, uea/second, . theser Mont oriewhichdecay rues i s 4 I ts.3 s a
.-,---v---. - . , _
. 1 SON 4
- 2. W 1he reacter is in automeele control mor's, wnhdrewel'of a aineis rod cluster control assembly we result in the immotstry of the other rod cluster comrol assembeles in the comroming bank.
The trensieRt we then proceed in the same manner as Case 1 desenbed above. For such casas as above e irtp win ultimatody ensue, ahhoug*1 noi sufncitattfy fast in ad cases to prevent a mirumum DNS rs'% irt me core of less than the safety analvass limrt. ., 15.3.6.3 Conefusions ' For the esse of one rod cluster control assembly fuNy wnhdrawn, with the reactor in the automatic hf (g or the manual control mode and initially operating at fut power wtth Bank D st the insertion an upper bound of the numtw of fuel rods exponenone DNM 1.3 is 5 percent of the total fuel f rods 5 the core. { For both cases discussed, the indicators and alarme mentioned would function to alert the operator to the malfunction before DNS could occur. For cose 2 discussed above, the inserDon limit alarms (low and low 4cw elarms) would also serve in this regard. i 15.3.7 References 1. Meyer, P. E., "NOTRUMP, A Nodal Transsent Smed Broek and General Network Code", WCAP 10040 A, August 1985.
- 2. Lee, H., Rupprocht, S. D., Schwartz, W. R., Tauche, W. D., **/.' ' ese SenaB Broek ECCS Evoluston Model Using the NOTRUMP Code", WCAP 10001-A, August 1988.
- 3. W. A. Bessee, C. L Case, A. C. Spencer, *LOCTRA R2 Propem Laes of Cooient Transsent '
Analysis," WCAP 7838, January 1972.
- 4. S. Altomare and R. F. Beny, "The TLRTLS 24.0 DINumion Depletion Code,' WCAP 7758.
I September 1971.
- 5. R. F. Barry, ' LEOPARD A Spectrum Dependent Non Spedal Depletion Code for the 18M 7094,* WCAP 3286 26, September 1983. o 0'
I
- 6. F. M. Gordelon, "Cateutetion of Mow Caesadown After Lane of Reactor Cooient Pump (PHOENIX Codeh* WCAP 7989, Sepeamter 1972.
- 7. -
Sumett, T. W. T., et aL, *LOFmAN Code Desertpden,' WCAP 7907 P A reuprietaryL -- WCAP 7907 A T .Oc 'n i, rApre 1984). hf ht I S. C. "- f_, ""*'"*", A ".c.
. "/-Gode for-Thermal Trenelems in tmh FustMods." WCW Woh but A [
-900er desse 99PS:-
- 9. F. M. Susdelen, mLaL, 'LOCTA4V Progrem: Laes et Cooient Transient Analysis."
WCAP a30s. INon ProprietaryL WCAP4301 (Proprietary) June,1974.
.1 f
cn l 15.3 12 l.
, - ., ,. y,- , y -# -,w,yw-w , - - - . v.,
FSAA 15.3.4 - Ceaplete Less of Forced Reactor Coolant Flow Inserts for reanalysis due to increased FAH insert A : Change Reference 8 to the following:
- 8. G.,
T..nsients in a U0Hi grove, H.
'FACTRAN 2 Fuel Rod,' - A Fortran IV Code for Thermal WCAP-7908-A, December 1989.
Insert B: the limiting case analyzed.The calculated sequence of events is trip occurring on a bus underfrequency condition.This Figures case corresp for the Complete loss of Flow analysis.15.3.4-1 through 1 Included in these nuclear power, and DNBR, each asThe a minimum function of time DNBR is not less than the safety analysis limit. l l l l l l b v t v
. 1 SON 4 ;
I TABLE 15.3.41 TIME SEQUENCE OF EVENTS FOR CONDmON lif EVENT : 5 Accident gyjgg y,m tw; Complete Less of Forced Reactor Coolant Flow All loops operstme. all pumps c down Coastdown begins g 0
- Rod Modon begins - 1.5 Mirumum ON8R occurs p c1ct ' Uk F,(Iow'.^b e
64 L
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TABLE 15.3.4-1 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS. Accident Event Time (sec.) Complete Loss of Forced Reactor-Coolant Flow All loops operating, all pumps coasting down all operating pumps-begin 0.0 to coastdown at 5 Hz/sec frequency decay rate Reactor underfrequency tr'ip 0.84 setpoint reached Rods begin to drop 1.44 - Minimum DNBR occurs 3.6 , s F 9 e W *- 70-
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l I l l t 4 m 12 - F I 3 N 8 ' If $ 3 W W 44 h 2 0, 0 1 2 3 4 S e 7 g g to TIME (SEC) l t I SEQUOYAH P1NAL SAFETY ANALYSS FIEPORT UMT51 and 2 campesas Lane er w Asessor Coodart Fknr l Host Phat m ime (Hut chamet l Figure 18,3.4 ac N
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e,. . L u.o,. l .. s.. . J i.e TICURE 13.3.4-3 Complate' Loss of Yorced Reactor Coolant Ft.ow DNBA Yersus Tinte ( 79
F a 4 2.6 il I 22 l
- 2. ,
t8 t8" s ' t.4 t.2 0 1 2 3 4 S e 7 s e to , TIME (SEC)
, SEQUOYAH FNAL SAFETY ANALYS8 REPORT UMTS 1 and 2 conwsne Laes of Paees meessor comment Nn OpWLvs.ftne Figure 15.3.4 3 W
SON 4 Figure 15.2.4 3 shows the informemen avedaNe to the operster en the eere power bued en the Nusdeer System for the Eg Xe case. As shown instantanoeus in nusteer power from 100% to 7.5% (< 5 - :J.2r en standard 30 essend 1. Fesm 7.5% the used unel the precursor leetepes been desisted. Fesm the pc.nt shown, an it der haff INe is . For the caos Xe, the NIS stab 6e reeeng en source range is aoNoved very rapedy, nunuses -
- ' to 21. minutes for the 54 Xe case.
i.. i Seocence of events Tabies are ettached Tables 15.2.41,15.2.4 21. These
* $ show that for betn cases > 15.T A of operator action
,sf acceptance crnens for this e is met, in addhion to the High is avedemie. Therefore the
$hutdown Alarm, there is V also the High Pressurtser Trio and alarm avedeble. In order to a very large total 6ducen volume is . The only means of acT.- :t.'.4 this large prosaurtter to to anew the fWng. A4 shown, however, Ihh results in a High Presourtaer La very 4 ur,y in e . m
/ /\
t two 44 arms would provies the overster en adeouste act of Indesdens that a bcron dii% ,! event was in progress and aies atow sdequete time for operator correedve accen. f 15.2.5 N.inali a== of FW " m Cantant Plow 15.2.5.1 !/- r - - ' ; af A- - ard _A - " : E-"'- A partial lose of seelant flew accident ces result from a meshenical or electrical feture h a reester oestant pump, or from a fault in the power suppfy to the pump. W the reester is at power at the tune of the assident. the immedste effect of nose of osoient flow is a rapid ingeese in sie easient temperature. This Inoresse sould readt in DNS with auhseguent fuel damage If the resseeir is not * ' tripped premedy. The neesenery preenoden assinet a perdal lese of easient flow sosident is provided by the low primary easient flew reasser wie which is accusend by two out of vues low flow alsnais h any reacter cosient loop. Above appromimetsfy 35% power (Permissive al, low flow h any loop we actuete a rescuw sto essenen appremimeisey 10% power (Perminehe 7) and the power level sw c: .J , to Permieshe 8 low flow h any two loops we aceween a reester vtp A reessor trip anonal from the pump underveltage reisy le provided as an ansiolpesary alytel which serves se a l backup
" to the low fisw sienst. It funedens eenentiesy leendessy to the low flow wie as that above
. w-":% 7 an undervalesse reiey wie sipisi from any two pumps we actuees a reester win.
l' I l I \ . 15.2 17 7f - c__ _ _ . . _ _
. 1 SCN S i
Normal power for the purnos is susoised ttwovsh buses connected to the generator and each pumo is susp6ed fmm a diferent bus. When a ponerator tne occure, the pumpe are automeccacy transferred to a bus swolled from estemel power lines, and the pumps wd continue to supph coolant flow e De core. Feaowing any turtlene tnp where there are no electrical faults whsch recuere tnpoing te generator from the network. the generator ramasne connected to the network for sporoxi. . 4 30 seconds. The reactor cooient pumpe remain connected to the generator thus ensunng full flow for approximately 30 seconds after the reactor tne before any transfer is made. 15.2.5.2 Anahraia of Effects and Car,tenuances Marbod of M Parnel le as of now involving loss of two seos wnh four loops in operation hos been snelyzed. TNs transient na aWysed by three eingetal computer codes. Mret, the LOMMAN Code (Reference 4) is used to calculate the loop and cece flow dunne the trenaient. the ene of reactor trip beood on l the calculated flows, the nucteer poww trenesent, ane the primary system pressure and temperature transients. The FACTRAN Code (Reference 3) le een used w ceicunste the heet flux tranasent bened on the nucosar power and flow from LOFTRAN. Finegy, the THINC Code is used to ' calculate the deserture frum nuc6este boigne retto (DNOR) during the transient, %eed on the heet flux frota FACTRAN and the flow from LOFTRAN. The WRS 1 sermission le used for DNM calculecon. The DNM trenaients presented represent the enethnum of the typecal or thimble cas. i Tvnical initial Candtione l l Inmes operedng condmons eseumed are me most adverse weih respect to the derpin to DNS, i.e., ' monimum siemey suas power level, mir*num samedy ese pressure, and meatmum steady stem - , cooient everage temperature. See Subsection 15.1.2 9er esoisnetton of inieel condWons. In addWon to the initled average temperature oenedon in Seeeetten 41.2,1.5*F wee added to $e t' inittel eversee temperature for conservedam. "O -" M :x 5:; n- "-e 5 - 'T rp
._.-_...a m_
-- ., .-- _ _ m m. _
__m a__- _
>-- u -- _ m 15.2 18 p
SON-4 i I Panctivhv Commcients 1 A conservative 6y large absolute value of the Doposer.only power ce f"wd is used (See Table A 15 M.1.22). The total integrated Dopow reactmry from 0 to 100% power is assumed to be 0.016 The lowest absolute magnitude of the rrhe-i temperature coefficient (0.0 )$iWI in assumed since this resuits in the maximum hot. spot heat flux dunne the inrtial part of the transient I when the minimum DN8R is reached. I Aow Coastdown oh, F-The flow coastdown atWysis is based on a momentum balance around asch reactor cooknt loop and across the reactor core. This momentum balance is combined whh the comwturty equenon a pomo momemum balance and the pump charsctentucs and le based on high estimates of system ' prosaure leases. E9End!E Wg
- The calculated sequence of events is shown on Tab 6e 15.21 for the case
( 4k2.5-1 ;;.#42E e'e 74 '::; :::nf;J . te w . ?; -- p anatyred. Mowee _1. , ?; . - - ': _ b (:n ::C;=r"? =_:::rdhOrdMn^2 nic 2;n:'.dr^ _ _ _ . gg The minimum DNOR 5 r -' t n::: le not less then the asfory analyele Amit. lg 15.2.5.3 Candaisna ' The anatysle shows that the DNM we not decrease below the asfoty ansfysisAnit et any $me l7 durhe the trenesent. Thus there we be no claddne damage and no resesse of*flamon products to the ".weter Cocient Syisem. [9
.2. 6 T- ..me Of Am .-- "_ P ^ = (*- ' n I -
15.2.8. - "-- ='
. of <*- -- and A= " ; t- - "
Y , gg if the plant is operedng loop due to the pressure pump out of w ' , . e le reveres flow through the inactive [ vessel. The said leg temperature in mi n.cew ioop is idemums = vm of m ese inspe nho ,s. car oore inist n, temperature). W the reester at power, le a tempereews drop aeross the steem (tAL generoser in the and, wth the reverse flow, les temperesure of the inacWw loop is lower resciar oere inist temperesure. A procedures require that the unit be brought m a imod of ines then of fus power to storene a puno in an inscWve loop in order a artne the inocen leep het les temperature ekmr to the esse intet tempereews. Stortino of an idle reactor costant pump whhout l il.2.a
SON.4 1
- 6. Turtme Lead . Turtnne load wee assumed conetont unt8 the sleeve hydroube govemor drwee the throule vehe w6de open. Then turtune load drepe as steam pressure drope.
- 7. Reactor Tets + Reactor Trip was artisted by low pressuruer pressure assumed at a conservedveh low value of 1775 pais.
Results The transeent resoonee is shown in Rouros 15.2.141 and 15.2.14- 2. Nucosar power starts decreenme vnmediately due to boron insect,on but steem flow does not secrease untM 15 seconds into the treneiert weien the turtnne thrtmie vetvo goes wide open. The trummatch betwoon load and ! nuclear power causee 7,,,, pressurtaer water level, and pressurtser pressure to drop. The low pressure trip set poitu le reached at 64 seconde end rode stort movine into the t. ore at 64 seconds. After wic, pressurse and temperetures slowty rios since the turtWne le tripped and the toector is producing some pcvw due to deleyed neutron fissione and decey heet. 15.2.14.3 Cmdinan NrY Rautta of the anatysis show that spurious nefety Wection with or w6thout knmedlete ructor trip a i presents no hesord to the imogrtty of the Reactor Cosient System. N.VINUN DNS ratio is never less than the initief vetus. Thus there we be no cladene damage and no reisese k' A of fisason products to the reacter cooient system. ! If the reactor does not trip :.T.T:%i, the low prosaure reactor trip we be b. This tripe . the turtene and preveme asem cooldown viersary ugeedne recovery from the incident. -
. l.
15.2.15 Refemnase
- 1. W. C. Gensieff 'An Evoluoden of Andoisesed Operadonal Trenalems in Westinghouse Pressurtmed Water Reactore," WCAP 7488, May 1971.
- 2. D. H. Risher, Jr., R. F. herry, 'TWWeQ.I A khd646menolonel Neutron IQnence Computer Code,' WCAP 797H A (^. g s i 'rd, WCAP 8028 A CA 'c_A January 1975.
' t pNt 4 (0. C. l; g "7A.,. ".',A 7 ^4forh i7._' M%T'"/,"WCA." G.
g- g( . .-,.. 4 Bumett. T. W. T., et ei., 't.0FTRAN Code Descripelen", WCAP 7907 7 A (Propriete% WCAP. 7907 A pien,,Propriseeryl, Aart 1984. g a 9
-e I*
15.2 44 9N
i FSAA 15.2.5 - Partial Loss of Forced Reactor Coolant Flow
-Inserts for reanalysis due to increased FAH Insert A : Change Reference 3 to the following:
- 3. Hargrove, H. G., 'FACTRAN - A Fortran-IV Code for Thermal Transients in a 00 2 Fuel Rod," WCAP-7908-A,. December 1989.
Insert B: Figures 15.2.5-1 through 15.2.5-3 show the resulting transient conditions for the 2/4 Partial Loss of Flow analysis. Included in there figures are total RCS flow, f aulted loop flow, average and hot c.hannel heat flux, nuclear power, and DNBR, each as a function of time. t l l . {
/
b
q l SON 8 ' TABLE 15.21 (Sheet 21 (Contmuod) But REQUENcf 0F EVENTS FOR CONDmON 11 EVENTS A ccident ! Exatn ima f 6.1 contromed aeron Oil
- 1. Guring 3 refusane stem, Dauden begine o
Operster of dilution: margin to refusiing
- e Mii prestuded (by administrative seneretel g
stame -
>1140 ,
- 2. DEudon During FuE Pmwer operadon
/ ,
3-
- a. Automode
. Meestor hof comret shusseum 2520 lI QU cauden basine O 3
Reeeear vtp eseeint vueshed h ,fer ever semperseure WT <120 1
, A i
snu m me, = nei n g euelen sendnued efter ste6 > 2400 Perdel Loss of Pereed nosew coseens rim. 3 As loose operedne, two pumps coasdne l$ down - Coastdown beslne 0 3 to. w ,ses . v. i.4/ Red.down .e. 2.4f i num of.in e s.t t 4 h
i sou-s -
+
t fl#UK4 W itl+ 1W TeurWva6- '
/
/
/
,/
/ ,
/
/
/
s.. 4 8.# '
/
/
~ ~ -
l
- ~~ ~ e/
. , _ ~ ~c.re n' /
b .. -
~~~
l .. " k ,/
.- Laos Fisw ;
..o / -
w -,
,4-, -;
/ .
3 t
, / -
/
/
M GURE 15.2.3-1 . Partial Loss of Forced Raattor Coelaat Flow - 4ees and Loop Flow versus Tine ' kQ:k.t"NtSM 4 a
14 m I,,. 8 e
'N. .,
a 6' a ' N 4 f 5 I 2 oc 1 o. O ' 2 3 4 I 5 6 7 g 9 to 14 1, 2 -
- 5 1.
S. l 8" d
- 4' a
" 2" 00 1 2 3 a 5 e 7 8 9 10 TIME (SEC)
SEQUOYAH FNAL SAFETY ANALYSN5 REPORT UMTS I and 2 Pwed tw at Ferood mensaar codern mow Roement Vessel arW Laep Row ve, Time Figure 15.2.5-1 l
?7
son-s ; Rf9tA<.6 W tTt+ TM RLLout4L l.8 1 .. > a yu . / . g .6 t . . -
.8<
. . . . . . . , . . i.
4 FIGURE 15.2.5-2a Partial Loss of Forced Reacter Coelaat Flow
.Isat Flus versus Time t
Tl
..,.;,., . . - . _ , .,e, . . , , _.. , , , . . .. .,_x.._r, r. ,a-
t 4 m 1*$ s ,. e u .a 5 m
.4 3
- 9. # 1 2 3 4 3 g ; g g, g TIME (SEC)
O e SECOOYAH . MNAL SAFETY ANALYSNS REPORT UNrTS 1 and 2 Penw toes at Forood Asamer Cooiere new Heat Phat ve. Th (Het W Figure 15.2.!Ma i
e t
?
?
SQN-8 t t 9 i f i [g# lN h M VlM b s.4 - i 14 > f I e L
,. r -
5 m
- m. ~ f
- g .6 .
3 w
** I g :
.e'n 4-
.4'* f i
8. t 8 8 8 4 6 6 F G 9 to t iet - sette i - I l I-. I. l-FIGRE 15.2.5-2b
,' complete Lees of Forced Isacter Coelaat Flow -
Rest Flus versus fine " t l
l t i I 1 4 j i 12 - I .
'^
8
- l i
$a b
l t n - y a 6 6
- 4, ,
2" }
,i O. 0 1 2 3 4 5 6 7 3 3 to TIME (SEC) l t
SEQUOYAH j MNAL SAPMMW.YS8REPO97 i UMTS 1 and 2 Penal Laos of Fwood Reester Centert Fkur Hess Pha m Time W Channet Agure is.2.5 ab t i _ ,_ ___ _ , .-- -- - -- - - ~ = - ^ ~ ~'~ "' ~'" ~
SQN-8 , Qgpwg W tT1> %(i Foetwi W 0 i.. l .
/ ,
I. ' 3 i. V
= .e -
er f* ' I
.e -
l t .
- e. _.
i i e . t . r e e se vint stati syiw TIGURI 13.2.h-2c # Partial Loss of Forced teactor Coolant flow Nuclear Power versua Time 9b -
I. 1 I,o . 5 , , e H,
=
g :. . 2' O. 1 2 3 4 5 6 7 g 9 10 TIME (stc) seovens PNAL SAPETY ANALYSIS REPORT UMTB 1 and 2 Pared Laes er Pomed meneem cooient mw ! masser penera nne j Plgure 18.2.2 ! r
SQN-8 i EPwg WitM T4C R u u; W c, i I 4.6 t.e " 88 f 10
- 46 *
.4 <>
a.a
- 8 8 8 8 6 6 e e , g, tlag egggs TIGURZ 15.2.5-3 Partial Loss of Torced Reactor Coolant Flow DNBR versus Time i .
4 28 i f.e 1.3 1. l.8 1.6 1 1.4 o h h h h g g TIME (SEC) sEQUOYAH FNAL SAF4TY ANALYSIS REPORT LK4TS 1 and 2 - Pared Laos of Forood Roemer Cooiart Row Opst m Time Figure 15.2.54
sons Lockad Mmar Beautta Transeerst values of pressunter pressure, rtector vessel new coastdown, tweiser power, and hot channer, host flux are shown in Fouros 15.4.41 trveuch 15.4.4 3. - Mammum Reactor Coolant System pressure, mammum clad tempesture md amount of airconium. 1 water reactum are centenned in Talde 15.4.4.1. Fours 15.4.4-4 shows the cted temperature trarwent for the weret case. 15.4.4.3 Canctuaiene I 1. Since Wte peak RCS pressure reached during any of die Wensients is less than that wNch would coues stresses to exceed the fauned coneden spots limns, the inteenty of the primary ocelent system is not e't OM. 2. Since the peak cied surface temperature ceiculated ftK the het sh?t durin0 the worst transient remains w-t'Ci ess than l the reguietory lanit and the amount of Zlroonium-weser reaction is emes, the core wiB remain in piece and imect with no conecouensel 4 less of core ocean 0 case.2ty.
$ 15.4.5 % Mandna Amandant ~
'f 1s.4.5.1 insndpenden af Caunan and Amident Qamerimean Q
k ofW,e a .e.e n.tu,e funof =, eied ode in . asse,,iny . spite - e.,**t,s eThe socident is denn controis and physical emiseelens imposed on fuel honene coeredene. As refuenne seeredens are
$ conducted in accordance wlWt prescrbed procedures under drect survetence of a supervisor.
h 15.4.5.2 A- ' i: af r_ - - and C_ - --- - t 5 M the snelysis and eeneoguences of die possuieted fuel hanen0 accident, refer a $sheecuen ' 15.5.8. 15.4.8 P mm of a cameras med e_: ^- '_ ' V ' ; med M Cantral Asammthe Bandard 15.4.6.1 Idensabannian af Cp and AarMant Qamerianian
** *ccident is dWhen as .e mechanical feaurs of eenew rod machen ,,,seo,, %
resuhine in to edecelen of a red cluster P 15.4 42 t _wr-.
- n. ._m__r__.___ .% . .. v, e.- - e,.y-e--+w,-.c. v.,,,-m-3 v e w _. .n-yv,--,,,3-e-,y+-r,,v..- v%,o-m-,--.-+--,,--ww .een ye r.we -e im.ww ,-
l S@ control assembly and dr,1ve shaf t. failure is a rapid reactivity insertior together with anThe consequence of t power distribution, possibly leading to localized fuel rodcore adverse amage. d Deston Precauttons and Protection Certain features in the Sequoyah kuclear Plant Dressurl2ed are to lla.' intended to preclude the consequences the if the Cossibility accident were of a water reactor rod ejection to occur , or accident sound. .onservative mechanical design of the rod housings, toThese includ nuclear design which lessent the potential power. and a uster ejection control assemblies and alnlaltes the nuacer of asseacti Mechanleal Desten _ The aschanical design Is discussed in Section 4.2. quality control procedures intended to preclude therodpossibility ofMech cluster control asseely drive mechanisa housing failure sufficient to allow listed below: a rod cluster cc1 trol assembly to be rapidly ejected from core 1. assembled and shop tested at 4100 pst.Each full length contr 2. The mechanisa housings are lodividually hydrotested as they are attached to the of during the hydrotest head adapters the completed in the reactor vessel head, and checked RC5. 3. Stress levels in the mechanism are not affected by anticipated sys transtants at power, or by the thermal movement of the coolant icops. Moments laduc14 by the design earthquake can be accepted withinCode, ASME the allowable Section !!!, forprimary working stress Class I components. range specified by the 4 The lengthlatch mechanism of forged fype-304 stainlesshousing steel.and rod travel housing are each a sin excellent notch encountered. toughness at all temperatures which will beThis material exhibits A sigelficant mergin of strength in the elastic range together vith th large energy absorption capability in the plastic rangenal gives e additio assurahce that gross fatture of the housing will not occur. The joints latch sechanism housing and rod travel housing, are reinforced by canopy type rod welds. periodic inspections of these (and other) welds.Aestnistrative regulattoms require 15.4-43 Ol m /COC4 j h
SQN Nuclear Deslan I (von if a rupture of a rod cluster control assembly drive mechanism housing is postulated, the operation of a Clant uttitatng chemical thin is sucn that Inn,rently iteited. the severity of an ejecttd rod cluster control assembly is In general, the reactor is ocerated with the ece cluster control changes caused by core depletion ar.3 tenon transients are coNensated by boron changes. Further. the location and grouping of control rod tanks are selected during the nuclear casign to lessen the severity of a rod cluster control assembly ejection accident. should a rod cluster control assembly be eject from its normal posit'aTherefore, during could be high expected powertooperation, occur. only a minor reactivity excursion, at worst, However, it may be occasionally desirable to operate with larger than normal insertions. For this reason, a rod insertion llett is defined as a function of power level. Operation with the rod cluster control assemblies power distribution. above this limit guaranteGs shutdown capability and acceptable is continuously indicated The position of all roos-in the control rod cluter control assembites An alarm will occur if a bank of rod cluster control assemblies approaches its lasertion Itatt or if one assembly deviates free its bank. There are low and low-low level insertton monitors with visual and audio signals. OS tating instructions ' require boration at low level alara and emergency elion at the low-low alars. gg, tor Prot 1ction - The reactor protection in the event of a rod ejection accident has been described in Reference 26. The protection for this accident is provided by the power range high rautron flux trip (hlgh anst low setting) and high rate of neutron flus increase trip. These protection functions are described in detail in Section 7.2. Effects on Adjacent liousinos Disregarding the remote possibility of the occurrence of a rod cluster control assembly mechanism housing failure, investigations have shown that fat)ure of a housing due to either longitudinal or circuarfarential cracking 11 not expected to cause damage to adjacent housings leading to incrcased severity of the initial accident. LimitinoCritir,j,g Due to the entremely 104 probability of a rod cluster control assembly ejection accident. Itettcd fuel damage is considered an acceptable consequence. Comprehensive studies of the threshold of fuel failure and of the threshold of significant converston of the fuel thermal energy to mechanical energy, have been carried out as part of the SPERT project by 15.4.44 0117F/COCa N
l SC+8 i the Idaho Numiner Carpereden (Reference 27). Extenerve tests of U0, arconnan c6ad fusi rnds representesw of thdru in Preamrtsed Water Reacter type cores how demonstreted fedure ttveeheide in the rence of 440 se 257 ceuem. However, other rods s of a oe trey efferent desson how emnibited feture.s les low as 225 sougm. These ree A4 eMer eierencency from the TREAT (Reference 28) resuhs, wNch indicted that this ihresheid decreases by ateut 10% whh fuel g , i . bumup. The cied fedure mechanism appears as be enshing for aero kette mis and bntile freetwo for irradseted rods. Alee moortatt is the certversson region of thermal to mechanical energy . ratio becomes mergnety detecteWe abow 300 copyn for unsmunseed rods and 200 caugm for irre6sted rede: catastrophic feture, flerpo fusi dispersal, large pressure rieel even for irrodseted reds, old not oecur below 300 caugm. j in view of the above esperimental renuns, conservethe orteerte are aspeed to ensure that there bene er no poeshery of fuel esserealin the oestant, grees lettee demersion, or severe shcmk waves. These artnerte are: 1. Aversos caugLn forfuel peastfuel. irremeted errthalpy et the hot spot below 225 ce8% for unirredeted fuse and 200 2. Peak reacter cooient prosaure less then that which wowW ceues stresses to exceed the fouis eenmeen sessa amers.
- 3. .
Fuel mettbts we he Emiend to less then 10% of the fuel volume et the hot em Werese fusi petet enthalpy is below the Emho of artnerten (1) above. 15.4.s.2 Anahana at EM-= and caa=== heathed of Anahrels The ana6yele of the RCCA edecelon sooldent is performed in two stages, Aret an everage eers nuedeer power venaient esiculosen and then a het spot heet venefer estouissen. The eversos c concWilen is performed usine epistel neueren Idneeles meshede se determine the everage power seneriske weih ihne instueno the nrteus essel oore feedback efloses, l.a., Despier reaceh4ty and mostreser remetweey, sneeley erW temmeresure wenslense h the het eget are then dessemine muhtefvene the eversos esse energy emnereden by the het channed fosser and perfernene a fusi trenaient
- :M heet vensfer easedesen. The power deMbuelen estaulated wthout faase=* is Ti aneumed as perelet Wiroughout the renaient.
A deteRed desuesten of the museed of anefynde een be found h Asfemnes 29. t o % '* l00 15.4 45 4
- - _ , . - . . < . , - - . . . - , , , , ~ _ - . - . - - - , - - ., -
SQN.4 averace Core Anal Ysis The spatial kinetics comouter code. TWINKLE (Refer e 30), is used for # the averate core transient analysis. This code ut crots sections q generated by LIOPARD (Refs, ence 31) to solve the two group neutron diffusion tneory kinetic ecuations in one, two or three spatial dimensions (rectangular coordinatos) for sin delayed neutron groups and up to 2000 spatial points. The computer code inclueet a detailed mJltiregion, transient fuel-clad-coolant heat tt ansfer model for calculation pointwise Doppler and moderator * 'tek effects. In this analysis, tae code is used as a one ., local axial kinetics code since it allows a more realil *le represantation of the special effects of axial moderator feedbac= and rod cluster control asseecly movement and the elimination of arlal feedback weighting factors. However, since the radial dimension is missing, it is still necer ary to escloy very conservative methods (described below) of calculating .he ejected rod worth and hot channel factor, Further description nf TWINKLE appears in Subsection 15.1.9. Mot Soot Analysis The_ average core energy addition, calculated as described above. is sultiplied by the approcriate het enannel factors, and the hot spot analysis is performed usthg the detailed fuel and clad transient heat transfer computer code, FACTRAN (Reference 25). This computer code c;'c.ulates the transient temperature distribution in a cross section of.a anal clad 00 fuel rod, and the heat flux at the se: face o' the rod, using as input ti s nuclear power versus time and the local coolant (nnditions. The zirconf un-water reaction is explicitly represented, and all natorial properties are represented as functions of temperature. A Darabolle radial power generation is used within the fuel rod. FACTRAN uses the Otttus-lioelter or Jens-Lottes correlation to determine the " S Mat transfer before 04, and the Bishop-Sandburg-Tong cor a;a r (Reference 32) to detursine the file bolling coefficient aftsi W . Tb DNS heat flux is not calculated, lastaad the code is forced into Oess by specifying a conservative DN8 heat flux. The gap heat transfer coefficient can be calculated by the codel however, it is adjusted in order to force the full power steady state tPleerature distributton to agree with that pref:-ted by design fuel heat transfer codes prassetly used try Westinghouse. l Fot full power cases, the design initial hot channel factor (F.7) is i input to the code. The hot ct'anel factor during the transient is I Assumeri to lacrease from the .dy state design value to the maximum trantient wslue in 0.1 seconds, and remain at the masteue for the duratiott the translent. This is conservative, since detailed spatial kinetics W elt show that the hot channel factor decreases shortly after l the nucl power peak due to power flattening caused by preferential ! feedbac( 1 the hot channel (Reference 29). Further eescription of i FAC1 RAM an ears in Subsection 15.1.9. l l l 15.a 46 0117F/COC4
/0 I i
i l l SCW System Overnressure analys t s Because safety 11alts for fuel there is little likelihood of fuel dispersal into the coolant. damage the specified earlier ar pressure surge may therefore be calculated on the basis of conventional neat transfer free the fuel and promet heat generation in the coolant. The pressure surge is calculated by first performing the fuel heat transfer calculation to determine the average and hot soc
- heat flux versus time. -
Using this heat flux data, a THINC calculation is conducted M determine the volume surge. plant transient computer code. Finally, the volume surge is staulated in This code calculates the pressure transle ,' taking into account fluid transport in the systen, heat transfer and to the pressure steam relief generators, and the action of the pressurtzer spray vsives. No credit is taten for the possible prer*ure reduction caused by the assumed failure of the control rod pressure housing. Calev14 tion of Basic Parameters Input parameters for the analysis are conservatively selected on the basis of calculated values for this type of core. The more important pa;'ameters are discussed below. Table 15.4.5-1 presents the parameters used in this analysis. [11tcted Rod Worths and Hot Channel Factors The values for ejected rod worths and hot channel factors are calculated using a synthesis of one dimensional and two dimensional calculations. Standard nuclear design codes are used in the analysts. no credit is taken for the flux flattening effects of reactivity featack. The calculation is performed for the maalaus allowed bank insertion at a given power level, as deterstned by the rod insertion limits. Adverse Xenon distributions and part length rod positions are considered in the calculations. The total translent hot channel fa'ctors F,T, is then obtained by combining the axial and radial factors. Appropriate margins are added to the results to allow for calculational uncertalattes; tacitsding an allowance for nuclear power peaking due to fuel doestftcation. Reactivit? Feedback Nelghtino factors The largest temperature rises, and hence the largest reactivity feedbacks occur in channels where the power is higher than averags. Since the weight of a region is depencent on flux, these regions have high weights. This means that the reactivity feedback is larger than that indicated by a stacle str.gle channel analysts. Physics calculations were carried out for temperature changes with a flat temperature distribution. 15.4-47 Ol17F/C")C4
/04
- _ _ _ _ _._.._ ._. _ . _ _ . _ _ . _ _ . _ _ _ _ ~ _._. _ _ . _
l= SON 4 and whh a large number of asiel and re&al temperature estnkmons. " ::-Ay c compared and affectrw wwonong factors determined. These <;% feedbecks for the approprises fluz shape, in spedal this analye bnetics method is empdoyed, amiel ;.i;L4 is not used. In ademon, no wese se o. to the tran the moderator feedboek. A conservative re&al weeghtne factor is ent asphed fuei temperature to otrteen an effective fuel temperature as a functon of time accountn or the i three ihmenssonal aneWis (Reference 29).masaan0 spoorf emension. T , Maa mar and r'-
' - t"- <- L The critical baron oensentradons et the I jWE nucteer code in order to attain moderator density coefitcient curven, cor'toered apphed to these toresults.
actual design condsfons for the plant. As docussed above, no weigh stoody stone compueer code with a Doppler ;ct. 4 fac larger than 1.0 (assnudmeesh 1.3L bast asn make the ' predictions before aisetion. This weser.eing factor we incrosse urn ashddent co discussed *abow. The trenaient ;C.-+ factor used in the ano6veis le presem 1. Danm m i N at = ._. ". C -.. A - W.~. e of the affecew detsyed neutron fraction y M typiosp vised values of 0. t r-.rus of lue and 0.50% se and of afe for sw fina cyces. The easident le se erected red when he wenn is neesfy aquel to or sresear ihon A, se, in in aero power . Is . order to bogenbe of atow for0.4ss cyete and fusure fuelofevoies, A, at and cyoses. peselmisde esemetes were used in the a Trio ."-- % W 1- l3 including the effect of one sewek rees. These .elues .
. e g
of rod rnotion enouned 0.s seconds after Wie high n assumed for the trip to sepelet of 0.2 essends for the t ;..r r.i channel to produce a signet. 0.15 i
/43 15.4 48
4 SON-5 breaker to egen and 0.15 seconds for the cost to runsaae the rods. The curve of rod inseroon versue time whleh was used is shown in Figure 15.1.51. The time to fue inserton assumed together with Wie 0.5 second delay overosometes the twne for enerwncant ineernon of shutdown reactivtry into the core. This is peruculerty wnportant conservatism for. hot fue power accidents. BAAkit!A The values of the parameters used in the analysis, as we5 as the results of the analysis, are presented in Tahis 15.4.61 and discussed below, aanmna of cveis. Fun Power Control bek D was assumed to be inserted to ha insertion amit. The weret e6ected rod worth and hot channel factor were 0.20% AKE and 7.11 respective 6y. The peak hot spot fusi center temperature reached the begrinen0 of life melt temperature of 4900'F. However, meinne was 8 restncted to less then 10% of the pellet.
% of cuena. Zare N=
l M this condition control bank D was assumed to %e fuoy inserted and C was at its insertion amAL The worst a6ected rod is located in control bank D and has a worth of 0.75% AkA and a het " channel factor of 14.06. b -- End af Cved4 4 F Power . . Control bank D was aneumed to be inserted to its insertion Endt. The a6ected rod wor 1h and hotl fo channel factors were 0.21% AkA and 7.88 respecWwely. The peak het spot fuel temperature saceeded the and of 80s meet tempensture of 4400*F. Howower, malains was resenceed to less ig then 10% of the penet. The verteden in most senaeresse with humus is discussed in Paragraph +
-4.4.1.2.
End of Cvels. Zaro Power The e6ected red worth and het channel facier for this esse were stamined aneuming centrol bank D to be fury inserted and bank C at ise inseriten amit. The resuhs were 0.37% Akand 28.0 resosednv. The peak fusi eeneer tempereews was 4381*F. These 50L eore power resuhe are 9 from an analysis for noode 2 eparaden, which is more asildne ese edade 3 eparaden. EoL A summary of te seems presented above is alwen in Tahis 15.4.81. - The nusteer . anh soon fuel and eled lampereews ransients for the weret sees in tenne of fuel mest hal power) are presemed k.j(pges 15.4.&1 and 15.4.8 2. The same transients for the worst casegi iri term clad tomew, 6)L sore power) are presented in,fleures 15.4.&3 and 15.4.64.
/oY l 15.4 49 i
= - .- -.--.- . .- . .-
1
- i I
i l' SONS 1 i r- I EtBalon%DM h ! It is assumed that fleeson prochacts are reisesed from the gaps of as rods having e DNOR of less tam the safety anahsas timet. In as cases consedered, less then 10% of the rods entered DNS based on a detesled 3 demonssonal THINC analysis. Ahhough twmted fuel mehmg at the hot spot l 3' I was prod.cted for the full power cases, in precoce metong e not expertod since the anahsis conservativeh assumed that the het spots before athi after specoon were es;Mdsci. Pressurd Surne A deteeled e=le=a=*==i of the pressure surge for an t$sedort worst I deter et SOL het fun power, indcoton thet the peak pressure does not exceed that wNch wouhl cause stress to exceed the fouhod coneson stress limhs meterance 291. 3lnce the severfty of the presem anahsis does not ' exceed this ' worst case' anetyeis, the accident for this plant we not renun in an exceserve I pressure rise or further damage to the RCS. Lattice Deformatiana i A large tempennure Orespont wE exist in the region of the het eget Since the fuel rods are free to move in the verocal eBracelen. efferentiel exponeson between asperses rods cannot preshre i distortion. However, the tempersture gradieme seress ins 9vidual rods may produce a fdros innene. l to bow the mHooint of the rods toward the het apet. Phyoles amiculations indeste that the not L l resuet of this wouhl to a nossove reactMey ineerden. In prestice, no elenancent bowing is . , anticipated, ainee the seruseuref rigisty of Wie core is more then suffleient es withstand the forces produced. Boene in the hot spot region would produce a not floot away from that region. However, the host from the fuel is reisesed to the woner reindve#y dowW, and it is sensidered inconceevesca inst oress now we he aufnoient to produce sieninoent leasse forces. Even if maselve and rapid beEng, sufflaient to deenet the leales, is hypochseinesy possuhread, the large void freedon in the het spot roeien wouM 3roduos e reduser m the soul oore moderseer en fuel ratio, and a large reduction in Wils reelo at the itet spot. The not sNest wouhl therefore to a nesseNe feecaneck. It een he eenchaded that no eeneelvehis mechanism selais for a not positive feedback reeufeing from laules deformoden. . In fact. a ames negative f sh ,* may result. The 4 effect w::;..;r; ignored h the analyses. 15.4.6.3 CGDEMME Even on a pessimissic tesis, the analyses indeste that sie desertied fuel and cied Emits are not L emoeeded, it is eenniudad that there is no alonger of oudden fuel depersal inen the cootent. 56nce the peak prosaure does not neced that which wouki cause seresses en exceed the fouhad condmon seress Betles, it iseeneluded that them is no danger
-eI*
p* N
/05 15.4 50
.-N
SON-8 l of further consequerttled damage to the primary system. The anWyses hc.ve demonstrated that upper limit h % product resase as a resuit of a number of fuel rods entenno DNB amounts to 10%. , 15.4.7 Refrenens
- 1. ' Accootancs Crtters for Emergency Core Coolwv1 Systems for Ught Water Coosed Nuctsar Power Reactors.* 10 CFR 50.48 and Apperdx K of 10 CFR 50. Federal Repstar, Volume 39, Number 3, January 4,1974.
- 2. 'The 1981 Verason of the Wi.G#.euse ECCS Evaluation Model Using BASH', WCAP. y 11524-A, Revies:n 2 (Non$roprietary), March 1987.
- 3. Wwn#.euse Doctrk Cecisei'.er., 'E.T.eeye.cy Core Cooang Performance *, June 1, 1971 SYwi.c#,cuse NES riW,.i.cy).
- 4. James C. Hessen, et al., *Laborstory Simulations of C3edding - Steam P.eections Fosowbg Loss of Coolant Mc6 dent in Water Cooled Power Reactors' ANL 7009.
- 5. J. M. Hetmen, ' Fuel DeneHkstion Ew' Tw.ud Results and Model for Reactor AW&ri, WCAP-8219, October,1873.
- 6. Deleted try Amencknent 8 6
- 7. De6eted by Amendment 8
- 8. F. R. h% 'Seeem Weser Crfdcal Flow frtirn High Praasurs Systern,' Handford Laboratories, HW-80635, January,1964.
9.- F. H. Moody, ' Maximum Flow Raes of Single Ce.Tw.er.t Two-Phase Mixture *, Paper No. 64-HT 35, and ASME Puhaemelon.
- 10. F. F. Cadek, et sL, PWR FLECHT (FuB Length Emergency Cors Heat Transfed, Final j Report,* WCAP 7648, Apre,1971.
l
- 11. De6eesd by Amendraene S. O l
- 12. L Beher, Jr., arme Just, J. C., *Squdles of Metal Wseer Rosedons at High Temperature,'
ANL 4548,1982.
- 13. Canns8detod Edeen Company of New York. Indien Point Unit No. 2 Final Safety Analysis flapert, Supplemerita 12 and 13, U. S. Atomic Energy CeT Taeelwi Dacket Number 50-247,
/O&
15.4 51
SON-4
- 14. C T T c. r--, Edmon Company, 7.aon Staton Final Safety Armiyas Rep .
Appends 1972, 14E, U. S.Atome Energy Commesson Docket Numbers y,50,295 an . 15.(19301. Dettus. F. W. and L. M. K. Boseter, Urnversrty of Califomie .,2,433 , (Berkeley), Pubia E
- 16. Jens, W. H., and P. A. Lottes, ' Analysis of Hest Transfer, Bumout, ensity Pres Data for High Pressure Water," USAEC Report ANL-4427 (IH11. ,
- 17. Macbeth, R. V., 'twenout A.selysis, Pt. 2, The Seeis turHut Curve,' O. K . -
Pt. 4, "Aposession of Local Constions Hypothems Tubes and Rectangular Channeis,' AEEW4 267 (1943),
- 18. Douges, R. S., and W. M. Roheensow, FBm Bonne on the Inalde of Vertic Upward Flow of Pluid at Low Quentnies. M(T Report 907S 26.
- 19. 245 D. (Mey M. 1966). McOgot, L W. Ormond and H. C. Perbins, , , .
- 20. W. H. McAdam. Host Transmission, W t"ll: 3rd faption,1964, p.172.
21.Syeesm,"J. M. Geets, WCAP 7900, " MARVEL June 1972. A Olglsel Computer Code for Transient Anafys 22.page F. S. 134.Moody, Transactions of the ASME, Joumel of Host Transfer, Feb , r l 23.(PHEONIX F. M. Bentaton, 'Cokselen Code),* WCAP 7900, September 1972. of Flow '-nr. Afear Loss of Reester Cootent Pu ( 24.7907 Bumett, A "1. N., T. W. T.,
"1 ri, Apre et 1984. sL, 'LOFTMAN Code Ma* WCAP 7907 P A (Proprie , .
C h(gInt
..g..,,..._..... ...
., - ..... .. - _ _ n.o..,,.
-tu A
-. .. - m
,c 26.Reacter,*
T. W. T, temma, Apre,196s. 5 CAP-730s, 'Moneser Pressesien Syseem Dhareity h waseinghoues P 27.Nucteer T. G. Corporation Taustus,#fed., 1370," June,1970. Annual Report Seert Propoet October 1964 Septemb
- 29. R. C. Lametesten, and F. J. Teste. ' Studies in TREAT of Zirosioy-2 Cod UO Fuel Elements,' ANL 7225, January June 1966, p.177, November 1966.
l
/01 15.4 52 1
_- w
_.----.--n. .- h b an b QwYO~
. r .u m -.
' _ _ _ ^-,
..__-..-_m. _
. -_ 'r
, ; - "n " - . . . _ _ __.. .
- _ C. ' '/.T " " M " ^. .* . _ . , 0 _ n, G i.
- 30. D. H. fueher, A., R. F. serfy, *TWHKLE ,
nencs Comouest uary 1975. A Mu 31.7094,*R. F.WCAP marry, 3289 28, "t.20PAAD September 1983. A Soectruen b 1.. Non Soedel Depteden l e for the IBM-
- 32. A. A. thehoo. R. O. Seneers, and L 5. Tens. .
- 33. *Weednghouse ECCS "=" i. Model E.~.i," WCAP 4339 S d l W. and 2meden, T. A., Juh 1974, WCAP4338 tarosstesaryl, June 19
- 34. Serdelen F. M., EL.M., *LOCTA N Program:
8306, June 1974, WCAP 4301 (Propr6storyl, June 1974.Lano ef Comient Tre 35, aardeten. F. W. suL,
- SATAN-IV progranu Cn.
3 of f.mer ef-Comient,* WCAP-8304, June ,ue 1974. 4F_^"C-H IPre xe.y, n. 0 .uc, C e eden Med.1er C.,. non.e Afier A Laseof4astant N (WREPL000 cadeh' WCAP4171, June 1974. - WCAP 4170 (P reertsesry), June 1974.
- 37. Heich T., and *~ -__ /., M.,,uy*Long % .Tenn le
_ ode 4 c4
- 34. Delseed by
- n i_. 8.
I
- 38. Deissed by ^r_ - 2._. 8. ..
40 J. J. 00eunno, F. C. Anderson, R. L .. , toiner and n L W Facters for Power and Test nosesor Stena," TD14444 March 1981ese , 4i. w. x.an,=, .uk, C.rar* ee v= a e en sensener.de r s b AseMant 1989. by Canestunene vendne for n em. *ie M.as. nsMneen Ptent ' WCA
- 72. November i
{- 1 l l 1 \ ,$ l . 15.4-53
-' + --
. - -. ~ - _. .- . - _ . - - = . . .
FSAA 15.4.6 - Rupture of a control Rod Drive Mechaniso-Houseing (Rod Cluster Control Assembly F.jection)
-Inserts for reanalysis due to increased Fg.
Insert A : Change Reference 25 to the following:
- 25. Hargrove, H. G., "FACTRAN - A Fortran !Y Code for Thermal Transients in a UO 2 Fuel Rod," WCAP-7908-A, December 1989.
Insert B: Change Reference 29 to the following:
- 29. D. H. Risher, Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Xinetics Methods," WCAP-7588, Revision I-A, January, 1975.
l
/o9
l { SQN 9
- 42. Laen Reguissary Guhte 1.7, Comrol of Combustitdo Gas Concemretione in Commi of Comient Accidem. I'
- 43. for W.JufyB. Caeret, *0N4. Nuedeer Sefety Research and Cw ':;.u.; Program BMoomNy Rep
. August 1964, *0RNL TM 2364, Nov.1964,
- 44. W. E, Cottres. *0RNL Nucteer Safety Research and Os.Ou.; Program 8H4cmNy Repor for September October,1964, "ORNL TM2425, p. 53, January 1968.
- 45. W. D. Reicher, M.M*--
Contecimems," J. See, TM and L F. Moone,
- Poet 4.0CA Hydro 9en Genersdon in PWR
+=s 10. 42fb427. ti t71L 4
- 48. Techpadaav H. E. 2)ttel,10. and42"T.'13 H.(1971L Row *Radetion and Thermal Statety of Sprey Solutions," thala 4
- 47. A. O. Amen, 'The Radetion Chemistry of Weser and Aqueous Soludons, *T.'.ew,, N. J., Van Nostrand,1961. .
- 48. NRC D. A.Report Powers and R. O. Meyer, "Cloddng Sweane and Rupture Models for LOCA Anatye NUREG-0630, Apre 1980.
- 49. Neednghouse ECCS Evoluodon Model,1941 Version," WCAP 9229'(Proprietary Versid WCAP 9221 (Non proprietary Version), February 1942.
- 50. Branch Technical Poettien, (SIS 6.2, 'Coneral of Combustade Games Concemretion in Comebmont Fotowing a LOCA").
t
- 51. "Seouoyah Unite 1 and 2 Seesm Generoser Tuto Piussine LOCA Senettivity Anstyeis,"
Weednghoues, Lateer TVA 43 850, Nowmber 3,1983.
- u. oeisted y An e. .
I
- 53. L E. Erin, et aL, 'Senmary Report Process Protoedo Syream Eagle 21 Upgrado, RTD Se, NSLS, MSS. EAM and TTO knotomemadan Seguoyah Unite 1 and 2,* WCAP 12904 (proprietary), W12844 (Nen. Proprietary), March 1900.
eT*
//O 15.4 . . _ _ _ - - _ _ _ _ _ . .. .-
.-- j] l SON 8 , I l TABLE 15.4.1 12 (Sheet 3) I (Contmuod) - TIME SEQUENCE OF EVENTS FOR CONDITION IV EVENTS , Acendent inC1 Time fSeel
- a. End of RCCA emeted 0 Oycie, Zero Power Reactor trip setposit reached 0.16 (High Neutmn Rux, Ngh sordngl Rods begin to drop 0.66 Peak cied average temperature S
reached 0.73 Peak fuel center temperature reached 2.49 4g,49, - L. Avo n wstw Ll 4
, pt
F5AR 15.4.6 - Rupture of a Control Rod Drive Mechanisa i Houseing:(Rod cluster Control Assembly Ejection) o { insert for reanalysis due to increased F g Insert C : Add the following into Table 15.4.1-12 (Sheet 3)- h it
- b. End of RCCA ejected Cycle, Full Power 0 Reactor trip setpoint reached 0.05 (High Neutron Flux, high setting)
Rods begin to drop 0.55 Peak clad average temperature reached 2.36 ~ Peak fuel center temperature 3.99 reached * '
- c. Beginning of RCCA ejected l Cycle, Full Power 0
-Reactor trip setpoint reached O.'05 (High Neutron' Flux, high setting)
Roas begin to drop 0.55 L 1 Peak clad. average temperature -2.29 reached Peak fuel center temperature 4.36 reached l-
+
e see-S IAGLE 15.4.6-1 PMMEltat M 85TM R$ W IM M EMSIES Casital , h1 le Mt Iles in Lifa Seelseilsus Deelsudste
- Ssud End
'Pomer Level 182 pct 4 pct let pct 8 pct EJocted red esorth, ERet/t .38 .75 .21 e.9r sel.yed sienstrast fractlen, E .M .95 .44 8.45 teseech reectivity eselghtlas 1.3 2.4 1.6 3.43 trip seestlwity, ER&/k 4.8 2.0 4.0 2.s f,before real electlen. -e:Se 2.43. - --
-e.se-2. (>2. .-
f, ofter red electlen F.19 14.85 7.88 as.e y sweer of operational pamps 4 - J 4 4 U man. f w l pellet meerage r- t , . e, .e -esse sf12.1 3,w -4 afoS6 sru
. e . l .es. te . e. se wrs um -.e 4871 4wi
- e. fot .t. red e r r, ii -see- 181 m -* n7 wi i 3 .
k
i
,- l 1
I
~
09 BCO We M o n
/
a.s -
=
I. g...
- 8. <
.s< ,
4 e
/
/ 4 TIME g
ISitt
.. ., s. ,
FIGURE 15.4.6 Nuclear Power versus Time for.V55. Act. EFF
. ( /j'
- _ .' ____.__ ~ _ _ - . , ,
- _ , _ . ~ , ,
a p. . . - a x-9
- 3. .
j i 2.5<-
- 2. -
e -
$ i.s 1 -
.5.-
. 00 1 2 3 4 S 6 7 8 9 10 TIE (SEC). SEQUOYAH FINAL SAFETY ANALYSS REPORT UMTS 1 and 2 Rod Chaser Ceed AssemWy Elmdon
'wumiser Powern.Two scL, e Mgure 15,4.s.1 .
//$
I
p
= SQN-8 C kt.p (t., N & k okSNi Ag i
- 4 sese.
sese. . Melt ;= _4s e _ _ oo_-, r , [ . -
# , Fuel Center Tamp k'eeen.
i Fuel Ag Temperature
# e sees. ,-
f Cl Temperature' iebe. - -
- e. i.
[.- s. .. s. e. - v. ' e. s. se. f tant estes w TIGtTRE 15.4.6-2
- Tuel. and Clad Temperature versus
' Time for 755. BCL. EZP
_//h
+s , -- --
6000 5000 , MeItiog 4400'F
----- _ _ _ _ ~
w Fuel Center Temperature - i
$ 4000<
S. i w
$ 3000-Fuel Avg Temperaturs I
5 2000" t Clad Temperature 10M < -
\
0 0 1 2 3 4 5 6 7 8 = TM (SEC)
?
4., SEQUOYAH FNAL SAFETY ANALYSIS REPORT UMTS 1 and 2 mod thener connes Assemtwy E>cean Pust and Camd Temporere vs. Time (ECt., M'P) Figure 15.4.INE
//7
e d SQN-8 48 3 , le8 ** a a. 4 k y se e y g j , le*8** l di 1 l t
- I'
al* 8. 8.5 4. 2. 6 3. 3, g .,
- at esac N 9/t*
4.i nGITRI 15.4.6-3 Nuclear Power versus Time for V58. EOL, RIP 8! 9
SQN-8 l i i l ' s.... Melting 4800 0F l l _ , l r i w " 1 l- *L. 4eee. ' ruel Centar T,mp i W >f s p., , / 3889. " EU,) j 7 i . 8 l i
! ci u r ,,,.,, ,
l 1989. ' w 8.
- g. 3. 3. 3. 4 l. t.. F. 8. 9' l*
tant osace w . l i I l v - TICU12.15.4. 44 .,,, Tual and Clad Temperature ~ versus Time for V5R. ECL EZP E
- . . , . . ~ ,
3 SECL-91451, Rev. 2 - j-10.1.2 LOCA 4
-f
/20
.. - .-.- . .. =. .. ~_ . - - ~ ... . . .. .-.
SON 5 1 S.4 CONDfTION IV . UMmNo FALAT1_ because Wteir consequences woved Wude eted the p radioeceve meternel. These are the most drasoc wtuch0 must amounts be desagned of ag rectosent limrong desagn esses. Cendroon IV faults are not and . to c%Hthus e flSrion P environment resutong in an undue nok to putWe hseKA and w**y n. ;m of gwe 10 Crm Part 100. A sangee Conernon IV fault is not to causenea ve490s conseq of furetions of systems needed System (ECCS) and the containment. to cooe with the fault including those of the Em been cissa. fwd is1 thss category: For the purposes of tNe report de fanovnne feutta ha 1. Mejor rupture of pipes contabing reactor v coolant us to and 'iciud e rusnure of the tatoest nice in the Reactor Cooient System aoen of coo . 2. Ma$er secondary system pipe ruptures.
- 3. Steem generator tube rupture.
I 4 senese reactor coolant pumo nocud rotor.
- 5. Fuel hendEng accident. l I
6. Rupture of a control red .ThieT. housing trod cluster cenenal -- : TI;y - The analysis of thyred and whene body deses, reautene from uct evente les beste for these censuisdone are presensed in Cha Resort aise inctuees the escuemen of systems . s4a a
' ::t ;y oenetsudne is tenene Asman product leakages from the esmeinment fenowens a Cancelen N oesunenes .
15.4.1 M6 " = r*- 'm Ex-. Mme >% n - of e ^ m A- > m j The ana#vais saecrfied tv 10 CPR 50.44 'Aeospanas CAserie for Emerg for Ught Water Nusdeer cooient acodont anefyele%;Messeers' 5_ le presented M 9 tis secdon. The reeufte of the loss 9 l l Crrtene. The desanoden of the verteus aspects 44, and 49. of the LOCA adeh the Aeosocence en i I l l l l l
'l 1
15.4 1
/2 V 1
.. - - . - . - - -. -- . - . ~ . _ - . . - . - - - . .- ~~
SONG The ECC$gpven when ooersteg dunne the inject >on mode with me most severe sangee falure, e desgned to ineet me Acceptance Cntena (11. t 5.4.1.1.2 Mamed of Nrmal Anatvnis Descrictens of the vanous esoects of the LOCA anaeys4a are provided in Referoness 2 and 49. Then cocuments coscree me maeor ononomene moceied. the interfaces among the comovter
- codes, and me features of the codes wNch serve to mectam comokance wnh me accectance entena of 10 CRF 50.46.
The analysis of a large break LOCA trans4ent is divided into three phases: Blowdown, Reful, and Reflood. A senes of comewter codes has been deveisped to analyas the tronaaent based on m specrfic phenomens whch govern each phase. During the blowdown perden, the SATAN V1 code (Reference 35) is used to calculate the RCS preeeure, enmelpy, danalty, and mese and energy flows in the pnmary system, as we8 as the heat transfer between the primary and secondary system. At the end of me blowdown, information en the stem of the system is transferred to the WREPLOCO code (Meterence 38) wheh performa me ceiculsdon of the refR ser6ed a bottom of core 1900) recovery tone. Once the vesses has refRed to the beF@ of she care, me reflood porton of me trenesent begins. The SASH code moference 2)is used to ceiculets the therma 6-hydrauhe seculation of Ine MCS for th3 reflood phase. informason concemeg me core boundary conditlens is taken from at of the above codes and input to the LOCSAAT code (fWerence 2) for me swooes of celeuleting the core fuel red termal - reasones for the entre trenesent. From the bouridery sendleiens, LOCSART esmouses the inds conettens and heet trenefer coefneient for me fue longe of me fuel red by empioving mechanisse models accreertete to the actual flow and heet transfer regimes. Cenaarvative assumotons ensure that the fusi rods modeled in me ceiculation represent the hotteet rods in me enes core. 15.4.1.1.3 Cantainment Anahrain The contammen aressure anefysis is performed wiet the LOTIC.2 (37) oods. The trenesent Ii pressure computed by me LOTIC code can be amored in me SASM code for me surpose of ! comovene the renees presented in Figure 4.14 15.y: The sentamment dem used in me eenesinmen 1-i determns the ICCS besleressure are presented in Tables 15.4.14 and 18.414.' The mass and energy reisese reens used for me sentenment backeressure ceiculation as a funenen of nme surw9e blowdowev$re given in Tebis 18.4.1-4, T , 15.4.1.1.4 Hanen af Laren gresk Samemme Calculesiens of said les double-ended guiartne pies breaks were performed over a range of Moody discharge egeftedents (C,1, for a plant asmeier in design te me Sequoyah Units, to identify me case wmch oneduees the highest peak sted tomooreture. Per that analysis, coisuissions were performed for discharge coefnements of 0.4,0.6, and 0.8. This esesorum of breaks wee performed assumeng , the avadaddity of ordy nonsnum aefety egocean flow casocity, in essertence wgh the sangde fasiure entena of 10 CPR 50, Ascenda K. A break discharge esoffissent of 0.4 was found to result in the hegnest peak clad temperature. Saeed on these resuRs. mis dasher $e coefhesent was cheaan to De anatyred as the lwrwang break size for Seouoveh Units 1 and 2. This caos was found to result m the kmeteg peak clad temperature of 400+487, wench is be6ow the 2200*F lima of 10 CFM 50.46 ! t Jtc49 l t5.4 3
SON 4 k 15.4.1.1.3 N at cenr-.-mt MM 44 sat 6 Teosomes the % of pur1Pne on tha calculated post 40CA Seevoyah u e, a contamme cWeu6seen was Aret performed to ottan the amount of mess whch ? earts throuc sets of purge he sunne the inroel pomon of a postulated 1.OCA trarwent. Purge ene h is e to De avaastne. In addroon, the ame to vecch the necessary to generate the S.I. segnal are conservativWy assossed as 15 seconds total
. Thus. now tnrough the postviated inroe double-ended part of fuey open coid leg break.avadable purge lines was evWuated from 0 0 to 5 5
. seconds for e
Y e I" 15.44
/.13
_ _ _ _ _ _ _ _ - - - - - - - - - - - - - - _ _ - _ _ _ _ _ - - = _ - - _ - - - - - - _ - _ - . ----- _
SON-4 1 The sedendeten employed the N> node TM0 comeuter cade model wk:n is desenbod in so 6.2.1.3.4. The>24 inch purge suppN lhes are connected to Vohnnes 34,37, and 25; pur omhouse Bruee ese eennected to 3e arid 28. Penedde comannesene of susefy lines and annaust open to sw esmesehe.e were coneadored. Esen of mese purgs enee le reoresemed try a now pain of crees osedert eroe eeuel to 2.844 ft' and a tetel flow reentence few eowel to and eart losa, three ftAv operi Iwtterfty vefves and a debris screeni. The meet conservaeve two pers of 24wwh on.rge and supoh lines were assumed to be esos in this calculosen. In seemon, two 12 inch lines connected to TMO node 29 were medeied as open. In a computation for ECC3 performance, the prestaat imoect on contal.vnent pressure occurs for the purge coes of maidmum air masa lesa. wfunn is beoed toen the two 12 inch lines bemg coen and irwehes three open purge Enes s the lower compartment (TMO elemente 34,36, and 37) and one purse Ine open in me useer comparenant tegemer wten a sold les brsek in TMO Vohnee t. A total of 2820 pounde of air are calculeted to be lost in mis sees. The meatmum aar less case is tne Eneelne case because any steam lost arough purgme h an ECCS besterseaure elevation would omerw4ee be celeutened as centense h me les bed. Therefore, any senem lost through surgmg is utelmeisey of no eeneceuence in me comsinment pressure determineelen, when any air less direcoy reduees esteutooed pressure. Yo incorporene the Tut >esseuisend resuste, the wines compress.en peak of es LOTIC sees was aqueted to consider me mese nest throuWi purgies. The corrected LOTIC sonesinmerw pressure thus renects the lose of mese through surging dunne the first few seconde et me LOCA venaient. The imoest of the reduced comsinment preseure en ECCS sortermance le inesuded in the d i peak % temperenne gp 4. seeing the plant Technical Spesseneden pealung tasear en T mie resust permies purging,of me Seousyah conselnment durew normes esereden to be conducted enough swee asse of purse tue. i a-- 0 15.4 7 J
/24'
SQN-9 TABLE 15.4.1 1 LARGE BREAK Results CD = 0.6 DECLG Peak cladding temp ('F) 2169
- Peak cladding location (ft) 7.0 Local Zr/H2O reaction (max) 6.79 Local Zr/H2O location (ft) 7.0 Hot rod burst time (seconds) 57.13 Hot rod burst location (ft) 5.75 Core-wide Zr/H 2O reaction (%) <l.0 Calculation Assumotions Core power (MWt),102 percent of 3411 Peak linear power (kw/ft),102 percent of 13.067 Total Peaking factor (at license rating) 2.40 Hot channel enthalpy rise peaking factor 1.62 Accumulator water volume (cold leg delivered)(ft') 3 @ 1050 per accumulator Steam generator tube plugging level 10 percent, uniform
- This value is applicable until such time as the' standard fuel with inconel grids is removed from the core. At that time, this value will be 2069'F.
/a 5
SQN9 TABLE 15.4.13 BACXPRESSURE TRANSENT USED IN ANALYSIS i
-l r= (=) pr-- sm i M.3 17.25 38.7 16.56 42.0 '
16.1 50.0 15.56 55.0 15.41 - 57.1 15.36 . 64.0 15.29 85.0 14.97 100.0 14.66 10e.5 14.54 - 150.1 15.74 l 200.0 13,9g 250.0 15.M L l i i. h y - , - - . + ., , . - , , . , . . .,w
g SQN4 TABLE 15.4.16 (het 1) MASS AND ENERGY RELEASE RA17.s C = 0.6 TIME $ j (sec) (lb/sec) (BTU /sec) 2.00E+00 5.5819E+04 2.9663E+07 4.00E+00 3.76385+04 2.05285+07 6.00E+00 2.9473E+04 1.6673E+07-8.00E+00 2.40093+04 1.4579E+07 1.00E+01 2.05983+04 l.'2424E+07 1.20E+01 1.8380E+04 1.0753E+0i 1.24E+01 1.7432E+04 1.0253E+07 1.40E+01 1.5121E+04 9.1177E+06 1.50E+01 1.35485+04 8.3830E+06 1.60E+01 1.21485+04 7.70595+06 1.70E+01 1.07795+04 7.0083E+06 1.80E+01 8.90095+03 6.07575+06 1.90E+01 7.6423E+03 5.38585+06 2.00E+01 6.5315E+03 4.7371E+06' 2.10E+01 5.6643E+03 -4.2114E+06 2.20E+01 4.7930E+03 3.5873E+06 2.30E+01 3.6109E+03 '2.5890E+06 - 2.40E+01 3.4061E+03 2.2302E+06 2.50E+01 4.7350E+03 2.35393+06 2.60E+01- 5.1390E+03 2.28185+06 2.70E+01 5.06373+03 2.0867E+06 129-
SQN.9 TABLE 15.4.14 (Shast 2) ,
, (ContumeO TIME $
(see) __( lb/sec) e (BTU /sec) 2.60E+01 5.01775+03 1.9106E+06 2.90E+01 4.82753<03 1.6756E+06 3.00E+01 4.6839E+03 1.45763+06 3.10E+01 4.0353E+0) 1.1095E+06 3.20E+01 6.0525E+03 1.5903E+06 3.4OE+01 2.9313N+03 6,8024E+05 3.60E+01 6.4982E+03 1.1822E+06 3.80E+01 1.02295+03 5.93493+04 4.00E+01 1.0070E+03 5.4401E+04 - 4.20E+01 9.9170E+02 5.74885+04 4.40E+01 9.7660E+02 5.65873+04 4.60E+01 9.6240E+02 5.5740E+04 ' 5.00E+01 9.3620E+02 5.41771+04-5.40E+01 9.1190E+02 5.2727E+04 5.795+01 8.8975E+02 5.1463E+04 5.86E+01 8.9061E+02 5.63695+04~ 6.373+01 8.41595+02 7.64073+04 7.283e01 2 -.1933 E+02 5-4335E+04 9.63E+01 2.27075+03 2.9896E+05-1.08E+02 2.1768E+03 2.8944E+05-1.55E+02 5.6515E+02 2.4903E+05 1.963+02 5. 7374E+02 2.3885E+05 , 2.38E+02 6.1940E+02 ~
~
2.5198E+05 . 3.16E+02 6.3191E+02 ! 2.4965E+05
/2 f a y- .a , ,.-.c --
.yr.. .,m, + y 4--, -'wan ---.-
sque TABLE 15.4.17 LARGE BREAX TIME SEQUENCE OF EVENis DECL Ce = 0,'6 r_a.3 START h uip sl# O' i S.I. signal 27 46 injection (CL) g,*, End of blowdows
- Bonom of core recovery pg Ai *ePry (CL) 100.47
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, Comparth, nt Pressure 0 i
u Figure 15.4.1-1 _ _) 6 i I l
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N __ l 4000. y . nI h x-gy-y g-2"* g e r
-4000.
am 5.0 10.0 15.0 (k=> 210 30.0 hl.0 NE M SEQUOYAH NUCLEAR PLANT UNITS 1 & 2 Final Safety Analysis Report Double Ended Cold leg Guillotine Break, C D=0.6 Cors Flowrate Figure 15.4.1-3 tD
asoa. I N w
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- o. l a s.o ino is.o no no me no TIME (see)
SEQUOYAH NUCLEAR PLANT UNITS-1 & 2 Final Safety Analysis Report. Double Ended Cold Leg' Guillotine Break, C 0.6 Cold Leg. Accumulator Flowrate D Figure 15.4.1-4
-/35
m. So,
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SEQUOYAH NUCLEAR PLANT UNITS 1 & 2 Final Safety Analysis Report Double Ended Cold Leg Guillotine Break, CD 0.6 Core Pressure Drop Figure 15.4.1-5 i
/31
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as.s -ano as.o riu s M 1 SEQUOYAH NUCLEAR PLANT UNITS 1 & 2. i Final Safety Analysis Report Double Ended Cold Leg Guillotine' Break..C 0.6 D Break Mass Flowrate , Figure 15.4.1 ,
' I
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. .- _ -_._ ____.__2.._=..__.__.. _ . . - - . _ . _ . _ _ . - _ . _ _ - _
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as.o 30.o :is.0 Tians(see SEQUOYAH NUCLEAR PLANT. UNITS 1 & 2 Final Safety Analysis Report Double Ended Cold Leg Guillotine Break, C 0.6 , Break Energy Flowrate D Figure 15.4.1-7
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-.;___=____. :
_ __ _ . _ . . _ - - _ ... _ _ _ . _ _ z_ _ __
1, e 1.2 1.0 ' i o.s 15 a. g . o.4 o.2 a 0, 5.0 10.0 18.8 age as age 38.0 Tuss (e.g 1 A SEQUOYM NUCLEAR PLANT UNITS 1 & 2 Final Safety Analysis Report Double Ended Cold leg Guillotine Break, C 0.6 0 Normalized Core Power Figure 15.4.1 ' I323 ._. . _ _ _ -_ . . _ _ . . . . . . _ . _ _ . . . _ . . _ . _ _ - . . _ - ~ , _ , . . . - .,.._._:,.._-.-._-_-.
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16
- 14 fI l je 4 d 10 - '
CR JW' 0 303 iluE (S C) SEQUOYAH NUCLEAR PLANT UNITS 1 & 2 Final Safety Analysis Report Double Ended Cold Leg Guillotine Break, C 0.6 D Core ind,Downcomer Liquid Levels Figure 15.4.1-9 156
-- ------.--..-._.n__ _ _ _ _ _ __ _ _ _ , _ _ _ _
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I 9 - 8 u 7 u l'z 6 i C z 5 E 4
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o o 3 r L1 2 1 0 0 25- 50 75 100 125 150 175 200 225 250 TIME (SEC) 1 SEQUOYAH-NUCLEAR PLANT UNITS 1 & 2 Final Safety Analysis Report Double Ended Cold' Leg Guillotine Break, C D=0.6 Core Inlet Fluid Velocity l Figure 15.4.1-10 l l 3 'l
I
\
l 3% i I 2000 AccuuutAton j 1750 1500 l l e,1250 l uJ k g 1000 - 9 1 . 750 ' puuren a 500 250 0 0 50- -100 150 200 250 TIME afTER BOC (sec) t i i SEQUDYAH NUCLEAR PLANT UNITS 1 & 2 Final Safety Analysis Report c Double Ended Cold Leg Guillotine Break, C 0 0.6 Accueulator and Pumped SI Flowrate Figure 15.4.1 ,
~
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50 40 30 -
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' SEQUOYAN NUCLEAR PUNT UNITS 1 & 2 Final Safety Analysis Report l Double Ended Cold Leg Guillotine Break, C0=0.6 Fluid Mass Flux figure 15.4.1-12 IVI l
l 103 Y od I i W Ls.
% \
S 102 . -. g r
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L. e u - g _,,,,,,,g 10 1 _ ,,_._ 6 , z i G ' QC b-a LJ I ' 100 0 50 100 150 200 250 300 350 TIME (S) SEQUOYAH NUCLEAR PLANT :. NITS 1 & 2 Final Safety Analysis Report Double Ended Cold Leg Guillotine Break, CD 0.6 Rod Heat Transfer, Coefficient Figura 15.4.1-13 I42.
2200. 2000. DN 1800. ! b w 1600. ' g- . E g 1400. t W l
- 1200.
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f I 800.O. 50. 100. 150. 200. 250. 300. 350. TIME (S) l l l t SEQUDYM NUCLEAR PLMT UNITS 1 & 2 Fins) Safety Analysis Report Double Ended Cold Leg Guillotine Break, C 0.6 CTad Peak Temperature D
- 444 Figure 15.4.1-14 l
\ 1800 7 , i 1703 - l
' I 150J. '/
l t [ 1503.
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1103. f N/ l -- 1003. 900. O. 50. 100 TIME[ 150. (S) 200. 250. 300. 350. SEQUOYAH NUCLEAR PLANT UNITS 1 & 2 Fins! Safety Analysts Report Double Ended Cold Leg Guillotine Break, C =0.6 Clad Temperature at the Burst Elevation D ; Figure 15,4.1-15 1 4i
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itMc iscci SEQUOYAH NUCLEAR PLANT UNITS 1 & 2 Final Safety Analysis Report Double Ended Cold Leg Guillotine Break, C 0.6 Upper Compartment $tructural Heat Removal D Rate Figure 15.4.1-16 14 S
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i SEQUOYAH Final NUCLFAR PLANT UNITS 1 & 2 sfety Analysis Report Double Ended Cold Leg Guillotine Break, C 0 6 _- Heat ssmoval by IC Orain D . Figure 15.4.1-20 - gi 1 f !
, _ _ _ _ _ - ______.__ -- ----~- - -
SECL-91-$5J, Rev. 2 10.1.3 Giapter 4 1 1
. 1 l
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SQN-6 i TABLE 4.1-1 (Sheet 1) REACTOR DESIGN COMPARISON TA8LE SEQUOYAH UNITS 1 & 2 REFERENCE PLANT 17:17 FUEL ASSEM8LY 17:17 FUEL ASSEMBLY ' WITH DENSIFICA WITH DENS!FICATION i _ THERMAL AND HYORAULIC DESIGN PARAMETERS EFFECT EFFECTS -
- 1. Reactor Core Heat Ouput, MWt 3411 3411
- 2. Reactor Core Heat Output, Stu/hr- 11,641.7 x 10*
3. 11,641.7 x 10' Heat Generated in Fuel, % 97.4 97.4 4 System Pressure, Nominal, psia 2250 2250
- 5. System Pressure. Min. Steady State, os1a 2200
- 6. 2220 Minimum DN8R for Design Transteats 30 Coolant Flow
>l.30 !
M 6.!- ONS Correlation .
* '" ,, with modifled "L" (M-3 with modif f ec
' spacer factor spacer factor
- 7. Total Thermal Flow Rate, Ib/hr 138.0 x 10*
- 8. Effective Flow Rate for Heat 132.7 x 10' Transfer. Ib/hr 127.7 x 10* 126.7 x"10'
- 9. Effective Flow Area for Heat Transfer, ft' 51.1 51.1
- 10. Average Velectty Along Fuel ,
Rods, ft/sec 15.6 ' 15.7
- 11. Average Mass velocity, Ib/hr-ft' 2.50 x 10* 2.48 x 10' Coolant Temperature. 'F
- 12. Nominal-Inlet 544,7 552.5
- 13. Average Rise in Vessel 63.1 14 Average Rtse in Core 64.2 67.4 64.9
> 15. Average in Core
- 582.2
- 16. Average in Vessel 585.9 +
578.2 544.7 Heat-Transfer -
- 17. Active Heat Transfer, Surface Area, ft' 59,700 59,700
- 18. Average Heat Flux, 8tu/hr-ft' 189,800 189,800
- 19. Maxtous Heat Flux for Normal Operation, Stu/hr-ft' M' WITI80 474,500' "
- 20. Average Thermal output, kw/ft 5.44
- 21. Maxime Thermal Output for 5.44 Normal: Operation, kw/ft- 20.;'" 13.0 13. 6 ' "
- 22. Peak Linear Power for Determination of Protection Setpoints, kw/ft 18. 0 ' " 14.0
- 23. Heat Flux Not Channel Factor, F, 4,497- 2 46 2.50
't? "!: 't /; :; ;;;;;?:td r!th tt: nh M r. - ?:M (b) This Ilmit is associated with the value of F, - 2.50 (c) See Subparagraph 4.3.2.2.6 M'6W m e:!p '
l1
,. l$( 0264F/COCa l-l~ i l , . . . . _ _ - _ _ _ _ _ . = . . . _ . . . . . _ . . _ - . . . . _ . - - _ . . , _ . . - - _ . . . . _ . c.- , - - _ _ _
_ _____ _ _ m. _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ _ _ . _ _ _ . _ i SQL6 including not the above factors. provided the assumed error in oeeration does cont' ave constant. for a period which is long comeared to the renon time The calculation results shown on Figure a.3.2 23 welch are greater than 18 Wft result from transtonts which would proceed without operator interventton for greater than 0.25 hour and would result in violation of the control rod insertion llatts in the Technical Specifications. Imbalance penalties are required for overpower protection.Since th It should be
-l' noted that a reactor overpower accident is not assumed to occur coincident with an Independent operator error. Additional detailed ,
discussion of these analyses is presented in Reference 7 and 30. Q Analyses channel factors of possibleF, and operating F* for peak power shapes show that the appropriate het local power density and for
- DNS analysts at full power are the values given in Tabue 4.3.2 2 and addressed in the SQN Technic 5 ffications.
The maximum allowable F, e shown in the sgu Technica Specifications. be increased with decreasing power, as Increasing F., with decreasing radial powerpower shape is chepers tied by the DNS protection setpoints and allows ~ s with red invertion to the lasertion limits, as described in Section 4 .3. The allowance for increased F" peraltted is F*.. . (1 + 0.3 (1-p)). This becomes a design . basis criterton which is used for est'ablishing acceptable control rod , patterns and control bank sequencing. 1.lkevise, fuel loadf ag patterns for each The vorstcycle yalues areof selected with consideration of this design critetton. F*,. for pos:Ible red confIgerat1ons occurf tng in normel operation are used in verifying that this criterton is set. Typical radical factors and radial power distrikettons are shown in , Figure 4.3.2-4 through 4.3.2-11. The worst values generally occur when i the rods are asNNd to be at their lasertion limits, lialntenance of i axial offset cem>rol estabil'shes red positions dich are above the l allowed rod lasertion lletts, thus troviding increasing margin to the
- F.. Criterton. Section 3.2 of Apference 8 discusses the determination of F... Thesa 11stts are taken as tapet to the thermal hydraulic design basis, as described in Section 4.4.3.2.1, 6 When a situation is possible in normal operation uhtch could result la local pouer destitles la escess of those assumed as the precondition for a sutsegoest hypothetical accident, but which would not itself cause fuel fatture, ahleistrative controls and alares are provided for returntag the core to a safe condttion. These alares are descrtH la detall in Chapter 7.0.
4.3.2.2.7 traerimenta) Verification of Power Distributtan Analysts t' i This subject is discussed in depth in Reference 2. A summary of this report is given here. In a measurement of peak local power density. F, with the moveable detector systes described in Subsection 7.7.1 and 4.4.5. the following uncertainties have to be considered. l l%
- 4.3 13 0047F/COC4
SQN-4 4,4 TW-^ AND wvDRAuuc OtsscN 4.4.1 h The overwl ottoceve of the thermal and hydreuisc desgn of the reactor core is to p neat trsesfer wascM is comeathis wtth the heet senersoon sistnbuten removal by the Reactor Coolant System or the Emer0enCY Co's Cooling System iwnen a assures that the foeoweg performance and Safety cntena reewarements are met;
).
Fuel damage' is not expected during normal eseresen and operemonal transients (Con any trenesent conecons ansang from faults of moderste frequency (Cenefoon 111. It is not pesadde. however, to erectude a very emed numter of red fedures. These wd be unthm tne casetery of the sient cisenus system and are eenesseem w4h tale sient essagn bases.
- 2. The reacter con be brought to a este state fenpwing a Ceneden til event omh enfy a amedi fresser, of fuel rods damaged' although aufhanent fuel damage ment secur as erectues resumeoen of opereden waneut canaderatie eutage eme.
- 3. The reactor con be brought to e aefe state and the core can be kast subertdcal wnh accostable heet trenefer geometry fo8eweg tronasems anang from Canelelen IV evems.
In order is sedsty the aheve crtierte the fotowing design bases have been estabashed for the thermal and hvermute semen of the reerter ears. 4.4.1.1 Qqearture frern heteate ReEna Deelen heels sente There wie he et least a 96% prehobety ther desertas from numisses beans (DNW we not occur on the amielne fuel rods durtne normal esereglen and operedenal senaients and any trenannt conedens artsing from fouhe of modersec frecuency (Censten I and N ewems# at 90% confidence novel. Hiseerteney, Wes artnerten has been senservedvety est by aftertne as she potewing thermal
- denen bases: there must be at leset a 95% seatsbety that sie minimum desernas from nucinate boena rede (DNWD af Was Raidne gewer red thstic Consten 1 and I swome is grosser than or I
y seuel as shMWem
- - - - - a, Smit # ^ Z e. --_ __- _ 2, " % ZT. M ^- % i .i ,
~ -- -
A -g _7_ gx u _ _ -
,, - - m g f g r'"{__
gg ,
'l Discussion H6esertenAy, eds Ofm Smit hee been 1.30 for Wiesenghoues assecedens. In this asonescen. the w961 esentegen flaterense SSI is _ 2,3 W the signinannt snorevement in the s.curery 1 of the artdesi heet Aus gradeden Iry using sto M1 servetecten insemed of seewious DNS correissions, a CPIM tatt af 1.17 As asoEssbie for the t 7af 7 Seandard fusi asemmidy (Meference
- 84) and for the VANTAGE SH Nel weernisly (Aefsne* Afl.
9 e** 4.4 1
-,------man,, me,.,,..v,,,,,-,,,,-,n ,s v.,e- , , , - , , -r,,-,4,,-w,e my r p n-g,4 , :p,,p --e n,-- , m n.-p g g qp n ,-4 p w ,4 g -p.,y -,p w.w- w , es-- y , e
Insert A The design limit DNBR is set at ~
- ne-:himble cell. Plant spectfi 1.22 for the typical cell and'i.21 f:r CNB penalties and allowance and analysis of the plant for flexib'?.ity in the design:nargin to accomodate
- hor re analyses to a ONBR limit value of 1.38.is provided 'cy performing'ane sa. Operacten fety b
t
- A id p +)' (6 L
-4
. j L
a l
.+
l l 5 l i i_,_,_ , _ - . , . , , . . . . , - - - - - - - - " ~ ~ " " " ' ~ ' ~ " ' ~ ' ~ ~ ~
SQN-4 l [g5C'I _ i M C a 4, ----+$= ene# vets 2w=_aerCPv8Rn- nmet of 1.38. Companne t / ty provernine deserturn from nucteets tW secourts heet tnwwfw e assured twtw een me fuel etad and the reactor coedant, theretry prevermne esed Maxwnum Nel rod surface samage as a ree M 6amage as used hers is denned as penctracon of ttw Messen product bam eted). e DS a.s. t . .\ i'
- _ _ _ _ _ _ _ , _ _ . - - - ~ ~ ~ " ---
$ntert b i
'"J!.:
g au f /f / 4Pt' The design method used to MINI-Revised Tharm.a1 Damign see the P uru DOIR dasign basis is the consarvative ap of e (R.afaranca /9/) (Rafarance 101) . plication Xavised which 1.s a nucinar paaxing factors In the
-RTDP met 2VdTharmal Design Procedura statistically with the DNBA dasign fuel fabrication p,arameters are ceabinedunca limit such that probability (with 95 percentcorrelation there is uneartaintiesat laart a to define the 95 percent when the calculated confidence) design limit. minimum DIGR that Esta vill not occur c' are for the nuclear The uneartainties enthalpy ha*. *included in ths MINI-RTDP n enthalpy rise engineering "M1 factor, F(N, N) the TEINC-IV and transient hot-channel factor, F(I, 3) ;
parameters are codes. Since and the plant safety considered the uncartainties in these anmiysas are in performed determining the design DNER value, us the uneartaintias for thase parametars. input valuas without design limit value is M /.XJ y4s*ForT'M lication, the DKER In addition to the considerations above, a s ea/ / l/ /** A h *4 + has been limit valueconsidared in the present analysis.pecific of 1.38 has plant allevance In particular a DNBR plant. The diffarance between been used in the safety analyses,for the analysas and the design the tNER val used in the safety upacific E3tB margin Isr3R value (1.34 vs. .31) provides plant to offset penalties the red how penalty and othar DNB that may occur. EAis flexibility la the design, operation or analysis of the plantDNB margin may . For condJticas outside the correlation (refer to section e of parameters for the WR.5- 1 used with .4.3.3.1), graatar thana tafER 1000 c.euslation psia. lim t of 1.30 forthe W-3 correlation pressure equal to or is psia), For low pressure applications (500-1000 the W-3 t3rER serralation limit is 1.45 (Rafarance 103). t Gb
t SOM 4.4.2.2.4 W H= TeenaferCz r =.s The W red sh heet .: boding are presortted in Sunweregreen 4.4.2.8.1.3wfor ch during outcooled forced conv 4.4.2.2.5 Fuel Cind T& .Tamtures The outer surface of the fuel rod at the het spot coerstes at a temperature of approm for steady state opersson at rated power throughout oore Efe due to the onset o Ir% (IWeinrung<pf afel, this temperature is that of the cied metal outer surfees, ourtne ooersson over ifu We of me core. me bulldus of saidse and crud e causes me cied awface temperature a heresse. A3ewence le made in me fuel osmer meet evolunden for mis tempersture the. Since the marmeHmbrause doelen heels amite DNS. odeouste heet trenefer le provided between the fusi sind and me rencour see,.dnt as met me more thermal aussut le not emised my seneider6dene of me cied temperature. . 4.4.2.2.6 T. _ = m n er F -s 72x_; The tessi heet fhm het channel facter, P., le esAned by me raele of the enesimum se core everse heet fka es elocussed h Subsere0resh 4.3.2.2.1. wie demon value F. hr normal l Inrtudne fuel Mg Bcealen effects. Subserneresh 18.4.1.1.7 desuases me F. value used h LOCA e@. l l l i l I l i l l l l l i
/61 4.48
G.0 nsa resums M e peak local poww of if6 kW/ft at fus poww **woons. The peak linse power for oeterm4Nrdon of proteccon setpoents a 21.1 kWnt. Th >:ss ygne atmoerature at trss kWat l must De below the 00 meet temowsturt * *r tne ktsome i "v s. swuonne asowaness for uncertainees. The fuel tamperature oss>cn basa e oiscusu r , Ar.eevon 4.4.1.2 and rssurts in a maxwnum 6Aowab&e caecuit'ed centerina tempertture of 477."'t A The posk finear power for preventon of contortine melt s > 21.1 kW/ft. The canter %ne tamoorsture at the peak linear ocwer rasurung from ovwpower trans,ents/cverpower arrora (assorning a marunum oyarpower of 118%) is besow that recured to produce tvWong. 4.4.2.3 CrWal Hast P6 Ratie or Decarmre h h Mw Ram and Wrino Techno6ev The r:2wmum CN8Ra for the rated power, and anocioemd trenssent condloona tre given in Table 4.4.21. The minimum DNOR in the ktwting flow channes wit be downstream of the peak heat flus locaton (hot lect) due to the increased downstroem enthalpy nee. DNER's are calculated oy using the correktion end defir9tione deecr%ed in .he foeowing suboersynons 4.4.2.3.1 and 4.4.2.3.2. The TH4NC4Vfstoutw code (dlacussed in Suboaragraon 4.4.3.4.11 le used to determne the flow distnbution in es core end the local conditione in the hot channot for use in the DNS correlat6on. The use of het channed factors le dLacussed in Suboersgrace 4.4.3.2.1 (nucteer hot channel factors) and in Suboeregraph 4.4.2.3.4 (engwisonng hot channes factors). 4.4..'i 5.1 Denarture from Nuciasta BoEno Technoloov TM WRS-1 DNS comdedon le sop 5 cable to VANTAGE SH fuel since, from a DNS perspective, the l VANTAGE SH assemedy la *1ue#y idendcal to me 17:17 incer si R Grid doenon. As documented in M.i.r.cs 87, me use of the VMS 1 DNS correlation wtth a 95/96 Emit DN9R of 1.17 is acoEcaele to the VANTAGE 5H fuel assembey. I For condh6one outa:de the range of aspecabety of me VMS.1, the W-3 comdaSon le used. , The W-3 corvolabon, and sevefu modficetiene of it, have seen used in Westinghouse CHF caeculetona. The W-3 wee erleinety developed from amole tube esta. (Refonance 391 but was emsci modned to aesty to the 0.422 inch 0.0. ved T ytd. (Reference 42) and 't." and. (Reference 341 as wsE as me 0.374 inch 0.0., (Reference 84,401 red hunde data. These mod 6ficatione to the W 3 cerratedon have been demoneerseed to les adequets for reactor rod bundle desegn. For the W 3 conslesen, me 9645 SnWt DN8R le 1.30 et eyeens pressures greater then or sous to i 1000 est Per low preneurs essecst6on (5001000 peo, me 95/95 Imit DNWIis 1.44 (Reference ' 92). !
.a e
/5?
4.4-9
SCN-6 4.a.3.2.1
% clear - Enthaley else %e-channel r actoi . rt witn N fuel rods and height H.Gtvan the local y, power density q' (kW/f t) at a 2 in a ::re i FI. = Mt rod tower O ' rage roc power _ . grJo
- t ro . vo . 2 )dr_ ( 4. a- 19) 1I 1"q' (x,y,z)dz N all !
rods ! the way in which F . Is used in the Out calculation is !aportant The location of afntmus DNSA decends on the antal profile of DNSA depends on the enthalpy rise toathat point. S e-and the valu tically, the maximum for ainlaus CNSA. value of the rc4 integral is used to identify Ja most Itkely rod An antal power profile is obtained unich when normalized flun along the to list tas-design ting rod.value of Fl., recreates. the antal heat the same axial profile with rod average powers which are typtcal ofT distributions found in hot assecoltes. profiles can be combined with worst case radta) distributtons forIn this mann reference DNS calculations. an integral and 1: It should be noted again that F*,. ' t s used as such in the DNS calculations. Local heat fluxes are attained by using hot channel and adjacent channel expl':1t power shapes shapes throughout which take into account variations in hortaontal power the core. radial power shapes is discussed in Aeference 52.The sensitivity of the T181NC. For operation at a fraction P of full is 91VJfLbyh Wr, the design FI used FI. 1 (1 .; 0.3 (1-7)) (4.4 20) I The perat 14 of F1. t included la the DW prots.ction setootnts sad allows radial power: shape changes wtth rod lasertton to tr* Insertton Itatts, thus alleutng greater flesibility in the nuclear design . 6 4.4.3.2.2 Amtal Neat Flu 011tribut1ans As discussed le Paragraph 4.3.2.1 the natal heat flus distributton can vary as a reself of red action, power change, or due to spati41 nonen transients dich any occur in the estal direction. Consequently, it~ 15 necessary to seasure the a:141 power labalance by means of the es-core nuclear detectors (as discussed in Subparagraph 4.3.1.2.7) and protect the core from excessive ast41 power-tasalance. The Reactor Trip System provides auteettic reduction of the trie-settsotat in the Overtemperature
'T channels on escessive antal power tasalance; that is, when an extremely large antal offset corresponds to an antal shape which desfgn as1al shape.
l I
/ S 9' ^
4 ,,,, ..........
~ - ~.--_ -_ - - - -. - __ - -_-
I l M twa Who "m m e s "~~ (Reverenes 801 in 18 arts -21PLM red bunees ONS teste have been eendusNed b expertenced reducten mere In sewer of aser* than enes
, ininmees einese tese. red DNS in the mesteut hundes sevyemet for ahert tumeu ee surfecs of the red. W these arid ""'m tosse, no eateadverse bodene en me eM^'s rod or a:W other red in the hunde se a consecuence of esersting in D 4.4.3.9 Inamv ***u .
er Rumps of Wri '-' Sumi "' r . ._ A M decussion W WWorlo00ine
- noted thet the resulting energy redenes .,iq isfuel not ages:1d to a rede.
4.4.3.10 Pun # nad anwa _ a =m t'- . pie. ' - - rueeser ears. The eMees of fusi m assemWy ble semWy er ensemel to me ages seuse lonel reduettene in emelant . sem 'ha flow flow.hand. The me roerner, and how for along the flow seroom the reducalen s eneuence me fuel red behavior. The eMeses of easient re eneAsw whlen vne hisekag core perfennenes are deserwuned dose are usuesy used to augment anehtiset teele auch as osmauser beeh by anehtlesi. and egertmen e emerhaemel propeme almaar se see that the eredened DNM is dependent ... and liefonenesupon she les
- 44) ahaiwe m ty and mass seiseety.
fuel assemkh en a suhahannel hesia, ages en DNM regardsee witham the 63, it le shown that for a fuel assemMy ahntu to che e age essws. In Astorence WasainghtA N answesely stedete the flow deathnien wdWalm the fuel assemWy when th Pus recovery of the Asw wee found to ensur about 30 inestas saked.downes the reester esereelne et the namhet fud power senddene t speamed e the hieskage. Whhh T an inerosas si endesley aid dearesse in mese weissity in ete tener sortene w.uu not f.ma u m -,seded,s. .e w ups of the fust va m a7-From a rowiew of the egen lleeresse it la eenshaded sist Goes hieshese asen leales seres
- incanta. A. Olseuse, gJ, Wiederence 411. show suk v eeky le sooroeshed een. Per !
roeuses wens aise found ter 2 and 3 este esmeiseek . Skater hische I 3 l l l l \ i- 40-0 4.4 31
.,-m_ _ , - - . , ,, ,.. - ,.
Roforoncos 101 3. Ray, *EDrI Rsvised Thermal WCAP-L3178-P, WestitthouS* Design Procedure (MINI RTDP)'" Elect.rit" Corporation (March 1989) , ,* 3 ', A. J. Friedland, and s. Ray, Procedure ,
- atavised Thermal Design Lattar, A.
WCAF-11397 (Proprietary), (February 1987) and C. Thadani (UENRC) (Westinghouse), to W. J.
" Acceptance for Referancl.q Johnson Topical of Licensin Report WCAP-11397, Revised Tharsal Design Procedure, g" (January isss).
103 USNRC 1986 Istter,dated (Westinghouse), C. R. Berlinger (USMRC) to E. P. Rahe, Jr. June is, ites, entitled: noduction in Fuel Assembly R m est for Maximum Red Bow Penalty. E 4 T.J.mit for Calculation of
/G/
SON 4 TAR.I 4,4,21 (Sheet 1) RfACTOM DESIGN COMPAR110N TAat r Sequoyah Urdts Reference mant 1 & 217 x 17 17 x17 Whh T>rnal and HydrsAe Denian Prametart Whh Derwfication Dennifiennen Reactor Cors Host Output. WWt11 3411 3411 Reactor Cars Host Ovaput. 57Uhr 11.641.7xlv 11,641.7x17 Heat Generated in Fwel, % 57.4 57.4 System Pressure. Nominal, pois 2250 2250 System Pressure. Mirumum Sesedy Stres, pole 2220 2220 ! Mirumum DNER et Nominal Conoblona Typical Mow Channel M j.y,J 2.04 TNmede (Cold Wes) Mow Channel Minimum DNOR for Design Trarments MA* q 1.71
>1.38 >1.30 l DNS Com2ction VMS1 'L' (W-3 with modified soeoer (faceers Cooient Row Total Thermal Mew Rees. RWr 134.0x108 132.7x1&
EffecWve Mow Rees for Heat Transfer, RAr 127.7s1 7 126.7x17 EffecWwe Mew Area for Heat Transfer. Pt* 51.1tSTEM, 51.1 51.3 (V40 - Average Vetecety Along Puel Rods, ft/sec 15.448706, 15.7 15.8 (VSD4 Avere0e Mesa Veloelty,82r ft8 2.50s1F (87Di 2.48a10' Comeent Touwerseurs 2.44 10*(Y560 Nomined ledge, W 544.7 552.5 Awarege Nasle vessel,4 43.1 64.2 Average Mas is Cere, W 57.8 84.8 Average in Ccre. WB 582.2 585.9 Average in Vesest,4 ~ 578.2 544.7
/62
TABLE 4.4.21 (Sheet 2) (Cormnued) REACTOR OfS!CN COMPARISON TAaLF Seoucyah Units Reference Mant 1 & 217 a 17 17 x 17 Whh Thannal and Hydraune D* Parameters VAth Darw 6catb6 Dendentie Active Heat Transfer, Surface Arse. Ft* St.700 Anrage Heat Rux, BTUMr fta 59.700 jgg,gon igg,gon Mammum Heat Mux, for normal openrdon BTUmr ft' 440,300" 440,300* Aversee Thermal output, kWM s.44 3.44 Mammum Thermal oviput, for 3 normal operadon kWM 12.6*/#l Peek Uneer Pont for den of protecton setpoims, kWM 21.1" 18.0* , Pressure Cros" Across care, paa 23.4 .t. 2.3 25.7 22.s Across vessel, incsuding norde nel 44.88 A 4.s 4s.1 2 4.5 (a) This amit is associsted with the value of F. = %g.w0 tbl Sesed on best estimate reactor flow rues as hw in Seedon 5.1. (c) See Suboeragraph 4.3.2.2.4.
' ~
(d) ' TIN / lN1* ^N! E't Y CL' e
?
e' '
/62;
l l i V SQN
\
- 39. !
, L. 5. Tong, Review ' tolling 5ertes, Crist s and Critical Neat Flut,' AIC Critical T!4-25887,1972. 40.
- c. W. Mill, F. E. Motley. F. F. Cadet. A. N. Wenael. *(ffect of 17 3 17 Fuel Assemely Geometry on Dus.' WCAP-4296-P-A (Westinghouse-Proprietary) and WCAP-8297 (non-Proerietary), February. I976).
- 41. N. Cheleser, J. Welsman and L. 5. Tong. "Subchannel Th of Bod Bundle Cores," WCAP-7015. Revtsten 1. January, 1949. ermal Analysts !
42. F, t. Notley and F. F. Cadek, *Ces Tests Results for new Ntataf Vane Grids (R) "WCAP-7895-P.A, (Proorlotary), and WCAP-7958-A. (non-Proprietary). January, 1975. l a3. Letter fras J; F. Stolt (NRC) to C. Elchelstager (Nesttnghouse). I ! Sue,1ect: Staff tvaluation of WCAP-7958, WCAP-4054. WCAP-4547, and WCAP-4762, Aar11 19, 1978. ad. L. 5. Tong, ' Prediction of Deearture from aucleate Solitag for aa j Antally non-catform Weat Flu Otstributton." J. ' Auct. Energy 21 241-2a4 (1967). ! 45. A. T. Laney and F. J. Noody. 'The Thermal Nydraulics of a let ting Water teatier.' American nudlear Settety,1977, t
- 44. F. F. Cadet, F. E. Notlep and D. P. Oostatcts. *tPfect of Antal Seat 1og as laterchamael 'herest Mt:1ag vith The a Nt 1at Vane Gr18.* !
i WCAP-7941-L. June.1772, (Nesttaghouse Prodrietary), and NCAP-7959, Octater,1972.
- 47. D. 5. home. C. M. Aayle. "Crossflew Mistas totwees Parallel Flow Channels Burlag tell' ag. Part !! Nesserement of Flow and Enthaley in Two Parallel Chamaels." get,-371. part 2. Decenter. -1967.
- 44. D. 5. Reus. C. M. Angle. 't.rossfiew N1 stas tetween Parallel Flow Channels Burtag teillag. Part !!! Effect of Spacers en N1 stag Between Tue Chamaels.* mt.-371 part 3. January.1969.
- 49. J. N. Seneales-Santale and P. GriMith. "Tue phase Flow Nixtog 1a had Sundle $stehamaels.* ASNt Paper 72 40VNt-19.
- 50. F. E. lentley. A. N. Mensel F. F. Cadet. *7he (Meet of 17 a 17 Fuel Asseely teamstry ea laterchamael Thermal Ntstag." ICAP-4299. Marca, 1974.
- 51. F. F. Cadet. *!aterchanne1 Theres1 N1:Iag vtth Nia1ng Yane Grids.*
WCAP-7647-L. May.1971. (No:tIagneuse Preerietary), and ICAP-7753. Septeaser.1971.
- 52. L. t. Nectretter. "Aeolicattom of the TitINC N Program to Pte De s t ge " Z^ i - . ---:^ 2 - "*1 t--*
, WCM- $95'0 N., i. # ~ T ,PA.Maky @M,
\ .
/V{
a.4::34 0048F/CDC' a
m _ . _ _ ~ _ _ _ - - _ _ _ - . - . _ _ . . _ _ _._ _ _ . . _ - . . _ . - __ _ 50s-4 13. F. W. Dittus and L. M. c.- noelter. ' West Transfer in auto Aadiators of the Tubular Type.' Calf F. Univ. Puellcation in Eng., 2. as.13, 443 441 (1930). 54 J. Helsaan. " West Transfer to Mater Flowing' Parallel to Tune Bundles.* Mucl. Sci, tag., 6. 78-79 (1959).
- 55. J. A. 5. Than W. M. Malker. T. A. Fallca and G. F. 5. aet sing,
= tolling in Proc. Instn. Mech. Enges. Sub-coolet Mater During Flouve Nested Tutes or Annu11,* -
180. Pt. C. 226-48 (1955-44). a
,) Tests of the San Onofre teactor Model,'
57. G. Mettront. " Studies of the Connecticut-Yankee Nyerasile Model,' NY0-1250-1 June. IN5. I 54. I. E. Idel'chtk, "Nanecok of Nyeraulle Rest stance. AIC-TR-4430 1940. 59. L. F. Moody. " Friction Factors For Ploe Flow." Transmission of tan American Society of Mechanical Engineers. M. 471-444-(1944). 60. l G. W. Maurer. 'A Method of Predictlag Steady State totilag Vaser Fractions la Reactor Caelaat Chassels.* 1e4P5-87-19. pp. 59-70, June, 1944. ' i 61. P. Griffith J. A. Clart and W. N. Rohsenew ' Weld volumes In i - Satcooled teillag Systems.' Asm Paper _ me. 58-87-19.
- 62. R. W. Sourlag. " Physical Model, based .on tuttle Detachment, and -
Calculatten of Steas Voldage la te Suecceled Region of a Nested Chassel
- 1518 - 1 0 . Decommer. 1982. j 63.
L. E.ras Itochreiter. for N. Chelemer and P.' T. Chu. 'TN!K IV Aa Iseroved
-eydraulls Analysis of God Bundle Ceres.'
s -Msg.4 i N 64 F. D. Carter. *!alet Ortf tstas of Open MS Ceres." MCAP-900s. f ,
.lassary. ISS. (Westinghouse Preprietary), and NCAP-7834. January.
1971.
- 65. J. Sheftheet. " Application of the TRINC Progras to PW Design."
NCRP-73ES-L. August. IN9. (leastinghouse Preprietary), and l ICRP-7838. January. 1971.
~
- 64. [. N. IIsvendstern and R. O. Sameerg. -*Slagte Phase I.acal tolltag and Sulk totilag Pressure Dree Correlations.' NCAP-2850. Aert!.
1964. (Mest1aghouse Preerletary) and CAP-7914. - June. I972. 1 4 % 33 0044F/CCCa
- /(M
. . _ _ _ _ _ _ . _ . m_ . _ ~ _ _ _ _ _ _ _ _ _ - , ~ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _
SECL-91-451, Rev. 2 10.2 Tech Spec Mark-Ups (@
N i 2.1 SAFETY llMITS BASES y.1 REACTOR CORE possible cladding perforation which would result in the re products to the reactor coolant. by restricting fuel operation to within the nucleate boiling regime whe heat slightlytransfer above thecoef coolant ficient is large saturation and the cladding surface temperature temperature. s e i result in excessive claddinOperation above the upper boundary of the nucleate b from nucleate boiling (DNB)g temperatures because of the onset of departure coefficient. DNB is not a directly measurable parameter during opera therefore THERMAL POWER and Reactor Coolant Temperature and Pressure related to DNB through the WRB-1 correlation ~and the V-3 correlation for conditions outside the range of WRB-1 correlation. been developed to predict uniform and non-uniform heat flux distributions. the DNB flux and the location of DNB R142 for axiallyT The local DNB heat flux ratio,
' core location to the local heat flux, is indicative of the ma i Dcs n The DNB design basis is as follows: .
g probability that the minimum DNBR of the l_imitino rod durino Condition I an g ,L *foeino 11 ovents is creater useu t we WRE-1 vi than or eoual to the LONM limit of the DNO co oni ANBR limit is established,W 3 cur i cid.ivn in tM5 applic6 tion). Th; corr:lation R142 g band cn- th; :ntir; exlicatie excerimental data setsah that there is a ; En3 will not occur when the minimum DNBR is at the DNBR limit. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER Reactor Coolant System pressure and average temperature for which the minim , DNBR is no less than the safety analysis DNBR limit, or the average enthal the vessel exit is equal to the enthalpy of saturated liquid . Thecurvesarebasedonanenthalpyhotchannelfactor,Fh,specifiedin the Core Operating Limit Re R159 1.55 for axial power shape. port (COLR) and a reference cosine with a peak of Fh at reduced power based An on allowance is included for an increase in the expression: F g=F [1+ PFAH (1-P)] where P a THERMAL POWER , R159 RAlt0 IHERMAL POWER F = COLR, and theFfg limit at RATED THERMAL POWER (RTP) specified in the PF AH = the power factor multiplier for Fh specified in the COLR. SEQUOYAH - UNIT f B 2-1 Amendment No. 19, 114, 13d, 155 1
/M y; " -
POWER DISTRIBUTION LIMITS Dig's BASES - DNB'R Fuel rod bowing reduces the value of DNB ratio. Margin has been retained between the DNBR value used in the safety analysis M1.30){ and the FWRfMI 4 R142 k m lat m limit l(1.17)!to completely offset the rod bow penalty. ! The applicable value of rod bow penalty is referenced in the FSAR. Ri$ Margin in excess of the rod bow penalty is available for plant design R142 flexibility. The hot channel factor F[(z) is measured periodically and increased by a cycle and height dependent power factor, W(z', to provide assurance that the l limit on the hot channel factor, F (z), is met. W(z) accounts for the effects 9 of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(z) R15 function is specified in the COLR. ^ l 3/4.2.4 QUADRANT POWER TILT RATIO The vadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and pericdically during power operation. The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F nis reinstated by reducing ' the power by 3 percent from RATED THERMAL POWER f5r each percent of tilt in excess of 1.0, 3/4.2.5 DNB PARAMETERS The limits on the ONB related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The Ifmits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety R142 , analysis DNBR limit throughout each analyzed transient. The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their ; limits following load changes and other expected transient operation. l
, /ST SEQUOYAH - UNIT 1 8 3/4 2-4 Amendment No. 19, 138, 155 00i k i. I '
a
-t r, 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR C0RE possible cladding perforation which products to the reactor coolant.
would eating of the fuel result and in thT e release of fission by restricting fuel operation to within the nucl slightly above the coolant eate boilingsaturation regime where the g surface temperature is tem result in excessive cladding temperatures ng regime couldbecause from coe nucleate f ficient. boiling (DNB) and the resultantonset sharp rd of departure therefore THERMAL POWER and Reactor Coolant on and Temper related to DN8 through the WRB-1 correlation and the W-3ure and Press conditions outside the range of WRB-1 correlation.correlation for i been developed to predict the DNB flux and the locatiThe ONB correlations R130 uniform and non uniform heat flux distributions on of DNB for axially 3 ONBR, defined as the ratio of the heat flux that wouldThe local DNB hea n core location to the local heat flux, is indicative rgin of to the macause DNB a DNB. Dhh The DN8 design basis is as follows: , DNPA probability that the minimum DNBR of the limitino rod dthere must be at least JI events Lindt,$t&c u:cd (the is WB creater W3 than
~
- or eaual to the fm"a limit of thuring Condition I a R130 r M NBR limit is establishe nb ::d r th:Orr:hti r, h thb sc;1icat4ed.e ONO corr + aticea eg::rf eTh!errelatid c r,t i r: : 014cj$1:
The -)suchthatthereisa95percentprobabilitywith95percentconfid ta4-data-set-l h5T will not occur when the minimum DNBR isence . at that the DNBDNBR lim Reactor Coolant System pressure and average R, tempe
' ator DNBR is no less than the safety analysis ONBR limit r which the t.inimum the vessel exit is equal to the enthalpy of saturated quid. li, or the average enthalp R130
! Tne curves are based on an enthalpy hot channel N factor
,Fg , specified in )
1.55 N for axial power shape.the Core Operating Limit Report (COLR) R146 a F 3g at reduced power based on the expression:An allowance i I R21 N F 3H
- 1 AH O~
where P = THERMAL POWER , RATED THERMA POWER P F R14e COLR,= the Fh limit at RATED THERNAL POWER (RTP) spec and PF 3g = the power factor multiplier for Fh specified in the COLR . SEQUOYAH - UNIT 2 '- B 2-1 Amendment No. 21, 104, 130, 146
/(8 March 30, 1992
v U i l POWER DISTRIBUTION LIMITS Dedi n BASES DN s Fuel rod bowing reduces the value of DNB ratio. Margin has been retained
.between the DNBR value used in the safety analysis i(1.38) land theMFil R130 icorr m ticM iimitI(1.17)}to completely offset the rod bow' penalty.
The applicable value of rod bow penalty is referenced in the FSAR. R146 Margin in excess of the rod bow penalty is available for plant design R130 flexibility. The hot channel factor FQM(z) is measured periodically and increased by R21 a cycle and height dependent power factor W(z), to provide assurance that the limit on the hot channel facter, F (2), is met. W(z) accounts for the effects 9 of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(z) function is specified in the COLR. R146l 3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. Radial power distribJtion measurements are made during startup testing and periodically during power operation. R2 The two hour time allowance for operation with a tilt condition greater than 1.02 but less than.1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F q si reinstated by reducing the power by 3 percent from RATED THERMAL POWER for each percent of tilt in excess of 1.0. 3/4.2.5 DNB PARAMETERS The limits on the DNS related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the-transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety K130 analysis DNBR limit throughout each analyzed transient. ' The 12 hour periodic surveillance of these parameters through instrument R21 readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expec%d transient operation. 170, ' s l SEQUOYAH - UNIT 2 B 3/4 2-4 Admendment No. 21, 130, 146 March 30. 1992
SECL-91-451, Rev. 2 - 10.3 COLR Markups S f
2.2 cot,a ros smooona cs:T7 1, crets g (3/4 1.3.5)Shutaavn 2ed Insertion limit 2.2.1 (Specification 3/4..1 3 5) The shutdeva below: rods shall'be withdrawn to a position Cycle Burnue as defined f M 1.lC21 1 2,000 Stans withem
> 2,000 to < 14.000 1 226 to 1 233 1 14,000 1 222 2.3 to 1 231 1 224 (3/4 1.3.6]Centrol Erd t.1 = l t fmeerties to i 231 (specification 3/4.1 .. 3 5) 2.3 1 shows is Figure 1.The control rod baala shall b 2.4 g rium Differ.anca (specificatlem
[3/4 2 1) 2.1) 3/4e on as limited 2.4.1 2.5 The auial flum difference (h , (3/4.2 2) East Fluz Rat Channel FD)fact limits are provided i n Figure 1. er . Pg(3) (Specificatica 3/4 2
. 2)
RTP F 0 r g( z) 1 ,,,,,,,,,
- r( 2)
P for P > 0 5 , RTP
'O Fg( 2 ) 1 ,,,,,,,,,* E(2) 0.5 for P 1 0.5 where P = RA2ED . . . . .
. PONSI vunr1mnfTEEm(Lt Pcast 2.5.1 2.5.2 F*
O #b 1(3) is provided la 71 T ure 3 Page 2 of 10 1038U
/71 O
_ .-_--- - ~ -
C01.2 FCI StgtfoyA3 gygg 3, gv M 1.$.3 Fipres 4 through 7 Note that the W(2) values repired by TS SR verssa core interpolation. height for all cycle buraups through e ermine wf2) thT e use of three pelat 2.6 Fuelaar rnehain,218e Not channel -M I (3/4.2.3) 7,
- F P AR (S ecification 3/4.2.3)
N RTP fag i F43 * (1+PF4g * (1.p)) Turew where P , a u. Poetra gg g 2.6.1 2.6.2 y , Q FFAg = 0.3 I l l l-i Page 3 of 10 1038U l
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l. 1038U
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w 7 C01.4 FCI SEQUOTAR CTN!! 2, CTC1.E [ " 2.2 Shutdawn sed fanartien timit (3/4.1.3.5) (Specification 3/4 1.3.$) 2.2.1 The equal to 225 steps withdrawn. shutdown rods shall be wit.hdrawn an er to 2.3 { _ trol sod insertien Limit (Specification 3/4.1.3.6) (3/4 1 J.6] 2.3.1 The shown control rod1. banks shall be limited in physi la Figure cal insertion as 2.4 Arial Flum (3/4.2.1] Differamen_ (Sper,1fication 3/4.2.1) 2.4.1 The salal fluz difference (AFD) 2.5 limita are provided in Figure 2. Rest Flur Not chanaal facter (3/4.2.2) . Fg(1) (Specification 3/4.2.2) ETP . Fg(2) i
- E(2)
P for P > 0.5 RTP Fg(2) 1 _* E(2) 0.5 for P 1 0.5 Tuen a c. me where P = RATED TEREDGI. POWER 2.5.1 rTP 2. Fg = }446 I 2.5.2 K(3) is provided la Figure 3. 2.5.3 Note Figures that4 the N(3) 4. through values required by TS $3 4.. 2 2. 2 are provided in versas interpelatloa.sere height for all cycle buramps through the use of th ree point-Page 2 of 11 i 1038U l
/ 79'
1 7 Cot.a roa st000YM (nt!T 2. CTGI.R{ 2.6 N Muelaar rathniev Rina net ci2nnal Fa-ter (3/4.2.3)
. rg (gP*Cificatloa 3/4. 2. 3 )
~
N RTP F M i F.ig
* (!*PFg * (1.P))
MFEXLL PCWER vtere P , m g m m g g 2.6.1 877 rg .d /' 62 2.6.2 , PFg = 0.3
- k t
- l. ,;
Page 3 of Il l- 1 l l i-l 1038U
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