ML20114A411
| ML20114A411 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 08/04/1992 |
| From: | Rosen S HOUSTON LIGHTING & POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20114A413 | List: |
| References | |
| ST-HL-AE-4066, NUDOCS 9208110236 | |
| Download: ML20114A411 (180) | |
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The Light company Houston Lighting & Power
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August 04, 1992 ST-HL-AE-4066 File No.: G25 10CFR50.90 10CFR50.91 10CFR50.92 10CFR51 U. S. Nuclear Regulatory Commission Attention: Docunent Control Desk Washington, DC 20585 South Texas Project Units 1 &2 Docket Nos. STN 50-490, STN 50-499 Proposed Amendment to Unit 1 and Unit 2 Technical Specifications and Updated Final Safety Analysis Report Congernina Contai.nInte_nt Volume Pursuant to 10CFR50.90, Houston Lighting & Power Company (HL&P) hereby proposes to amend its Operating Licenses NPF-76 and NPF-80 by incorporating the attached changes to the Technical Specifications and Updated Final Safety Analysis Report, concerning containment volume for the South Texas Project Electric Generating Station (STPEGS) Units 1 & 2. This proposed change incorporates a correction to the containment free volume calculation. The original STP calculation, MC-5281 " Free and Sprayed Volumes Inside Containment" Revision 1, overestimated the containment free volume.
The proposed change reflects the evaluation of the as-built facility. All containment and safety-related systems inside containment will function within design or EQ limits nnd all analysis results are bounded by design and regulatory requirements.
HL&P has reviewed the attached proposed amendment pursuant to 10CFR50.91(a) (1) and determined that it does not involve a significant hazards consideration. The basis for this determination is provided in the attachments. In addition, based on the information containad in this submittal and the NRC Final Environmental Assessment for STPEGS Units 1 and 2, HL&P has concluded that, pursuant to 10CFR51, there are no significant radiological or nonradiological impacts associated wi.th the change avi the proposed license arendment will not have a significant e'tuct on the quality of ohe environment. The STPEGS Nuclear
.. d.ety Review Board has reviewed and approved the proposed changes.
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100049 m 2-oss. m A Subsidiary of flouston Industries incorporated I 9208110236 920804 PDR i
ADOCK 05000498 P ppg O
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llousten Lighting & Power Company South Texas Project Electric Generating $tation ST-HL-AE-4066 File No.: G25 Page 2 I
Upon approval of the proposed change by the Staff, HL&P l requests a 10 day implementation period following the date of issuance ofic e licenso amendment. This will allow adequate time for reproduction and distribution of the change.
In accordance with 10CFR50.91(b), HL&P is providing the State of Texas with a copy of this proposed amendment.
If you should have any questions on this matter, please contact Mr. A. W. Harrison at (512) 972-7298 or me at (512) 972-7138.
S. L. Rosen Vice President, Nuclear Engineering SDP/ag Attachment (s): 1) Significant Hazards Evaluation
- 2) Marked-Up UFSAR Pages
- 3) Marked-Up Technical Specification Pages Tsc\92 025.J01
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. Houston Lighting & Power Company ST-HL-AE-4066 South Texas Project Electric Generating Station File No.:G25 Page 3 cc: Without Attachments Regional Administrator, Region IV Rufus S. Scott Nuclear Regulatory Commission Associate General Counsel 611 Ryan Plaza Drive, Suite 400 Houston Lighting & Power Company Arlington, TX 76011 P. O. Box 61867 Houston, TX 77208 George' Dick, Project Manager U.S. Nuclear Regulatory Commission INPO Washington, DC 20555 Records Center 1100 Circle 75 Parkway J. I. Tapia Atlanta, GA 30339-3064 Senior Resident Inspector c/o U. S. Nuclear Regulatory Dr. Joseph M. Hendrie Commission 50 Bellport Lane P. O. Box 910 Bellport, NY 11713 Bay City, TX 77414 D. K. Lacker J. R. Newman, Esquire Bureau of Radiation Control Newman & Holtzinger, P.C. Texas Department of Health 1615 L Street, N.W. 1100 West 49th Street Washington, DC 20036 Austin, TX 78756-3189 D. E. Ward /T. M. Puckett Central Power and Light Company P. O. Box 2121 Corpus Christi, TX 78403 J. C. Lanier/M. B. Lee City of Austin Electric Utility Department P.O. Box 1088 Austin, TX 7876' K. J. Fiedler/M. T. Hardt City Public Service Board P. O. Box 1771 San Antonio, TX 78296 Revised 10/11/91 L4/NRC/
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter )
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Houston Lighting & Power ) Docket Nos. 50-458-Company, et al., ) 50-499
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South Texas Project )
Units 1 and 2 )
AFFIDAVIT S. L. Rosen, Jr. being duly swo.*:n, hereby deposes and says that he is Vice President, Nuclear Engineering,- of Houston Lighting
& Power Company; that he is duly authorized to sign and file with the Nuclear. Regulatory Commission the proposed revision to the Technical Specifications and Updated Final Safety Analysis Report for the effects of the reduced containment volume; is familiar with the content thereof; and~that the matters set forth therein are true and correct to the best of his knowledge and belief.
S. L. Rosen Vice President, Nuclear Engineering STATE OF TEXAS )
)
COUNTY OF MATAGORDA )
Subscribed and sworn to'before me,-a Notary Public in and for The State of Texas this 96 day of 48u 7 , 1992.
I m h 2m,,d cs %AAU 4 Notery Pub 11c in end ror the ikM;ar;l:rl;;=d LR _
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t-r ATTACHMENT I ST HL AE yo4 0 PAGE I - 0 F _- l 9
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SIGNIFICANT HAZARDS EVALUATION-4 FOR THE EFFECTS OF .
1 REDUCED CONTAINMENT VOLUME i
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ST HL-AEWO6 6 SIGNIFICANT HAZARDS EVA111ATlDll TARLE OF CONTENTS 1.0 Background 1 1.1 Introduction 1 1.2 Proposed Changes 1 2.0 Safety Evaluation 2 2.1 Containment Maximum Temperature 2 b 2.2 Containment Maximum Pressure 3 2.3 Containment Minimum pressure 4 2.4 Containment Subcompartment Analysis 4 2.4.1 Main Steam Line Break Subcompartment Analysis 4 2.4.2 Pressurizer Subcompartment 5 2.4.3 Radioactive Pipe Chase Subcompartment 5 2.4.4 Regenerative Heat Exchanger Subcompartment 6 2.4.5 Residual Heat Removal System Valve Room Subcompartment _6 2.4.6 Steam Generator Loop Compartments 7 2.5 Containment Safety-Related Equipment Qualification 7 2.5.1 Electrical Cable Temperaturc 9 2.5.2 Polar Crane Beam Box Pressurization 9 2.5.3 Polax_ Crane Temperature 9 2.5.4 Containment Air-Lock Seal Temperature 9 2.5.5 Containment Abnormal Temperature 10 2.6 Containment Leakage 10 2.7 Containment Minimum Backptessure for 74CA ECCS Analysis 11 2.8 Containment Hydrogen Generation 11 2.9 Radiological Dose Analyses 12 2.9.1 Puol Handling Accident Inside Containment 12 2.9.2 Control Room (CR), Technical Support Center 13 (TSC), and Offsite LOCA Radiation Doses 2.9.3 Rod Ejection Accident 14 2.9.4 Plant Building Airborne Concentrations 14-2.9.5 Reactor Coolant System Vacuum Design System 15 (RCSVDS) Releases to Containment 2.9.6 Environmental Report Doses 15 2.10 Safety Injection / Containment Spray Operations. 15 3.0 Significant Hazards Determination 16 Appendix A Marked-up UFSAR Pages Appendix B Marked-up Technical Specifications Pages T$C\92-006.001 1
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! 4 ATTACHMENT I ST HL AE-yo c s M BACKGROUND -
- 0F /7 M Introduction The containment is designed to limit the release of . radioactive materials to the environment subsequent to postulated accidents, such that resulting calculated of f site doses are less than the guideline values of 10CFR100.
The capability of the containment to perform this function is ensured by comprehensive design, analysis, and testing that includes consideration of the calculated free containment volume.
as well as the calculated peak containment internal pressure and temperature values. Peak containment internal pressure and temperature values result from analyses of the energy releases of worst case postulated accidents [a double-ended pump suction guillotine (DEPSG) loss of coolent accident (LOCA) for maximum pressure and a double-ended main steam line break (MSLB) for e maximum temperatitre) into the calculated containment free volume.
As part of a probabilistic risk assessment model development effort, calculation MC-5281 "" tee and Sprayed Volumes Inside Containment" Revision 1 was r viewed. The review identified a mathematical error in the containment free volume calculation. Due to the mathematical error, the original calculation overestimated the containment free volume. The containment free volume is reported in the Updated Final Safsty Analysis Report (UFSAR) and the Technical Specifications tit is an input to other design calculations and analyses whose .- .lts are used to establish both design and licensing bases.
Station Problem Report (SPR) 91-0049 was issued to address this condition. A Justification for Continued Operation (JCO)_was also developed. This Safety Evaluation identifies and evaluates the proposed changes resulting from this condition.
M Eronosed Chanceq As the result of revising calculatioq MC-5281, the containpent free.
volume has decreased from 3.56 x 10 ft to 3.41 x 10 ft , with a margin of error of +0.1% and -0.85%. Therefore, the revised containmergt volume value, including the -0.85% margin of error, is 3.38 x 10 , a' reduction of 5.1%. The -0.85% margin of error is applied to the containment pressure analysis in accordance with ANSI /ANS 56.4-1983, " Pressure and Temperature Analysis for Light Water Reactor Containments".
Reanalysis of worst case postulated accident energy releases using the reduced free volume as an input resultad in an inscrease in the values of calculated peak containment internal pressure from 37.5 psig to 40.5 psig and an increase in peak containutent internal atmospheric temperature from 323*r to 328'F<.
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l ATTACHMENT I ST HL AE- 66 These values are presented and used both inm%the.10in wei2 4,90_OF,IY os the Technical Specificatio;.s. In addition, these values are inputs to numerous other analyses whose results are also found in the UFSAR r.nd Technical Specifications.
The proposed changes involve-revision of appropriate portions of UFSAR Sectionc 3.8, 3.11, 5.4, 6.2, 6.4, 6.5, 7.A, 12.2, 15.4, 15.6, 15.7 and Technical Specification Sections 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, B3/4.6.1.2, B3 / 4. 6.1. 4, B3 / 4. 6.1. 6, and 5.2.1, with the above revised values as well as revised calculated margins, setpoints, radiological doses and editorial changes.
Appendix A provides the marked-up pages of necessary UFSAR changes.
Appendix B provides the marked-up pages of necessary Technical Specification changes.
2.d DAFETY EVALUATION Thin safety evaluation was based on a review of numerous calculations and other documents. The ef fect on those calculations or documents which were substantially impacted by the change in containment free volume are described in this safety evaluation in the following topical areas:
- Containment Maximum Temperature o Containment Maximum Pressure o Containment Minimum Pressure
- Containment Subcompartment Analysis
- Containment Safety-Related Equipment Qualification e Containment Leakage o Containment Minimum Backpressure for LOCA ECCS Analysis
- Containment Hydrogen Generation e Radiological Dose Analysis e Safety Injection / Containment Spray Pump Operations 2.3.1 Containmqnt Maximum Temnerature UFSAR Section 6. 2 .1.1. 3 addresses peak calculated containment temperature following a Design Basis Accident (DBA) . The DBA which results in the maximum calculated peak containment temperature is a 1.4 ft* double-ended guillotine rupture (DER) MSLB with minimum containment heat removal system in operation. This DBA is addressed in calculation NC-7007, "MSLB Containment P/T Analysis".
Other MSLBs are analyzed in calculation NC-7047, "MSLB Containment P/T Analysis for Split Breaks".
Previous MSLB analyses used a containment free volume of 6
3.56 x 10 f t" . For the ' reduced containment volume, the VSLB analyses with a reduced free volume of
- 3. 2 0 x 10' f t ,wege . reanalyzed which is about Sg below the revised minimum calculeted volume of 3.38 x lo' f t . The reanalysis of NC-7007 increases the peak containment temperature from 323'F to 328"F.
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The reanalysis of NC-7047 results in a peak containment temperature '
of 311.5'F. Therefore, the results of NC-7047 are bounded by the results of NC-7007.
J UFSAR Table 6.2.1.1-3 reports the containment structure is designed for 286*F based on the p-4k containment vapor temperature of 323*F l as derived in calculati- is MC-574 3, 5746, 5747, 5748. The revision i of NC-7007 (MSLB) incre sos the peak temperature to 328'F. UFSAR Figure 6.2.1.1-27 shows the post-accident tempo.ature above 286'P lasts for less than 100 seconds.
The containment vapor temp 3rature is about 277'F at 6 seconds; then goes up to 328'F at 30 seconds; then goes down to 272*F at 100 seconds. During this period, the film heat transfer coefficient will not be high enough to instantaneously heat up the l
containment structure to the vapor temperature. Therefore, the l containment structural temperature will not exceed 286'F. Hence, the desigr. structural temperature of 286'F remains bounding.
Peak calculated containment temperature is also an input to numerous other analyses which are addressed in this safety I evaluation.
l 2._ul Containment Maximum Pressunee
, Maximum calculated peak containment pressure, as described in UFSAR l Section 6.2.1.1.3, results from a DEPSG LOCA with maximum Safety Injection and minimum containment heat removal. This calculation is documented in NC-7032 " Containment LOCA Pressul.a/ Temperature Analysis".
Previous LOCp analyses had used a containment free volume of
- 3. 56 x 10' f t . For the reduced containment volume, this design basis accidept was reanalyzed with a reduced free volume of 3 . 2 0 x 10' f t , which is abogt g% below the minimum revised calculated volume of 3.38 x 10 ft. This reanalysis resulted in a peak containment pressure of 40.5 psig. The previous peak pressure was 37.5 psig.
The containment design internal maximum pressure is 56.5 psig as identified in UFSAR Table 6.2.1.1-3. The revised peak pressure of 40.5 psig is well below this design pressure. The increase from 37.5 to 40.5 psig reduces the margin of design pressure to post-accident internal peak pressure (shown in UFSAR Table 6.2.1.1-2)-
from 53.6% to 28.3%. However, the margin is still well above the minimum 10% margin stipulated by the acceptance criteria of Standard Review Plan (SRP) 6.2.1.1.A. In addition, the revised long-term containment pressure will continue to be less than 50%
of peak pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as specified in the Safety Evaluation Report (SER) Section 6.2.2.1.1.
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Peak calculated containment pressure is also an input to containment equipment qualification, leak rate testing, and other calculations and analyses which are addressed in this safety evaluation.
- La Containment Minimum Pressurg Containment minimum pressure is determined in calculation NC-703u
" Containment Pressure After an Inadvertent Spray Actuation." The results of the calculation (-3.08 psig)- are presented in UFSAR Table 6.2.1.1-2. The calculation (NC-7036) assumes the initial containment temperature is 120'F and the final temperature is 50*F.
The volume term cancels out in the equation used to calculate the minimum pressure.- Therefore, this calculation is not affected by the containment frae volume change. !!cnce, the containment design minimum pref $3ure of 3.50 psig, reported in UFSAR Table 6.2.1.3-2, and Technical rpecification B3/4.6.1.4 is still applicable.
M Containment S*.1bconDartment Analysig Containment subcompartment analyses inve been performed in several calculations. The results of the analyses are used to determine forces and moments on the equipment and structures of these compartments. The effects of reduced containment volume on containre2. subcompartment analyses are addressed in the subsections that follow.
2.4.1 Main S' eam Line Break Subcompartment Analvs_lg calculation NC-7028 provides the peak differential pressures in containment due to a Main Steam Line Break. Difforential pressures are calculated in thirteen individual subcompartments. Six of these are in direct contact with the main containment volume. For subcompartments which are not in direct contact with the main containment volume there is no impact on the calculated pressures.
For those subcompartments which are in direct contact with the main containment volume, the peak differential pressures are expected j to increase by no more than 6%. The results of the analysis are ,
presented in UFSAk Table 6.2.1.2-13. Table 6.2.1. 2-13 incorporates ;
the highest . calculated differential pressures from - either MSLB l analysis (NC-7025) or Main Feed Water Line Break (FWLB) analysis l
(NC-7048). The results of the MSLB analysis bound the results of :
the FWLB analysis for all subcompartments. Table 6.2.1.2-13 shows l the minimum margin for design pressure is over 120%. Therefore, I the original subcompartment design-difforential pressures remain bounding.
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} ATTACHMENT I ST HL AE 40M PAGE 7 ' 0F /9 SER Section 6.2.1.2 states that substantial design margins exist in the Mai1 Steam and Feedwater Line Subcompartments. As indicated above, the original subcompartment design difforential pressures remain bounding.
2.4.2 Pressurizer Subcomoartment Calculation hC-7016 determines the peak pressures in pressurizer sub,:ompartments due to a spray Line Break. The results of the calculation are presented in UFSAR Tab 1L 6.2.1.2-9. However, Table 6.2.1.2-9 presents results obtained from Rev. O of Calc. NC-7016.
Revision 1 of NC-7016 changed the nodalization scheme. After-including the margins for containment volume reduction, the results from Rev. 1 are lower than those given in Rev. O of NC-7016 and in UFSAR Table 6.2.1.2-9. Table 6.2.1.2-9 states the design pressures are governed by the surge line break.
Calculation NC-7008, Rev. 2 determines the peak pressures in pressurizer subcompartments due to a surge Line Break. A review of NC-7008 shows that the containment volume used in the calculation 1s 3. 3 0 x 10' f t3 . The revised free volume is
- 3. 3 8 x 10' f t3 . Therefore, thu- original calculation provides conservative results for surge line break. UFSAR Table 6.2.1.2-11 gives the results for the surge line break. Since the results of Rev. 1 of NC-7016 are lower then Rev. O, the revised peak differential pressures from Rev. 1 of NC-7016 will continue to be bounded by the design pressures of the surge line break.
The peak forces and moments presented in Rev. 1 of.NC-7016 remain bounding after the effect of reduced containment volume is considered. The peak static forces and moments, not the-dynamic profiles, are used to confirm thu adequacy of the RCS component supports design. Since the peak static forces and moments determined in NC-7016, Rev. 1 remain bounding, the RCS component supports design will continue to be acceptable.
SER Section 6.2.1.2 addresses the Pressurizer Subcompartment and Surge Line Subcompartment Analyses. As indicated above, 'the original subcompartment design differential pressures remain bounding.
2.4.3 Radioactive Pine Chase Subconnarlment Calculation NC-7033 estimates the peak differential pressures in the radioactive pipe chase due to a CVCS letdown line break. The results of the calculation are presented in UFSAR Table 6,5.1.2-17.
Table 6.2.1.2-17 shows the design differential pressure is 1.32 psig. The new calculated pressure is 1.35 psig. The calculation which determines structural loads in the Radioactive Pipe Chase is CC-5414. Calculation CC-5414 neglected the 1.32 psig pressure rise when determining the design loads of the pipe chase region because the differential presaure of 1,32 psig is TOC W 2*066,001 5
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ST HL AEWO66 PAGE 9 0F /7 insignificant. Therefore, increasing the di*ferential pressure from 1.32 psig to 1.35 poig will not impact the structural design calculations.
SER Section 6.2.1.2 states that substantial design margins exist for the Radioactive Pipe Chase Subcompartments. Since calculation CC-5414 is not impacted, the proposed change will not cause a reduction in margin of safety.
Md Recenerative Heat Exchanaer Subpompartment;.
Calculation NC-7034 estimates the peak differential pr.mures in the regenerative heat exchanger subcompartment due to a CVCS letdown line break. As a result of the reduced containment volume, the new differential pressure is expected to increare from 4.6 psig to 4.85 psig. UFSAR Table 6.2.1.2-15 denotes the design pressure as 5.52 psig. The design margin is reduced from 20.0% to 13.8%,
but the original design subcompartment differential pressure remains bounding.
SER Section 6.2.1.2 states that substantial design margins exist for the Regenerative Heat Exchanger Subcompartments. As indicated above, the original subcompartment design differential pressure is bounding.
4 2,4,5 Residual Heat Removal System Valve Room Subcompartment Calculation NC-7035 estimates the peak dif ferential pressures in the RHR valve room subcompartment due to a CVCS letdown line break.
UFSAR Section 6.2.1.2.3.9 discusses RHR valve room subcompartments.
Calculated and design differential pressures for RHR valve room subcompartments are compared in UFSAR Table 6.2.1.2-19. Calculated differential pressures in UFSAR Table 6.2.1.2-19 were obtained from NC-7035, Rev. 2. After considering the effect of the reduced containment volume, the results obtained from NC-7035 do not exceed the design basis.
SER Section 6.2.1.2 states that substantial design margins exist for the RHR Valve Room Subcompartments. As indicated above, the original subcompartment design differential pressures remain bounding.
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ST HL AE40W PAGE 9 0F li 2 94 Steam Generator Loon Comnartmenta Calculation NC-7039 estimates the peak differential-pressures in tne Steam Generator loop compartment due to various primary system line breaks. The results of the calculation are listed in UFSAR Table 6.2.1.2-5. NC-7039 calculates differential pressures in 41 individual subcompartments. Seven of these subcompartments are in direct contact with the main containment volume. Due to the effects of the reduced containment volume, the peak differential' pressures for subcompartments in direct contact with the containment are expected to increase by no more than 74. Table 6.2.1.2-5 shows the minimum margin for -design pressure is over 100%.
Calculation NC-7040 uses the results from Calc. NC-7039 to determine forces and moments on the steam generators and the reactor coolant pumps due to pipe breaks inside the SG Loop Compartments. The reduced containment free volume effects have been evaluated in DCN MC-0119. The transient forces and moments on the SGs and the RCPs f rom NC-7040 are presented in UFSAR Figure 6.2.1.2-21 (30 sheets) and Figure 6.2.1.2-22 (24 sheets). The peak forces and moments presented in Rev. 4 of NC-7040 remain bounding after the effect of reduced containment volume is considered. The peak static rorces and moments, not the dynamic profiles, are used to confirm the adequacy of the RCS component supports design.
Since the peak static forces and moments determined in NC-7040, Rev. 4 remain bounding, the RCS component supports design will continue to be acceptable.
SER Section 6.2.1.2 states that substantial design margins exist for the Steam Generator Compartment Analysis. As indicated above, the origins 1 subcompartment d6aign differential pressures remain bounding.
2,d Containment Safetv-Related Equinment Oualification The effects of the increased accident pressure and temperature are addressed in this section. Radiation effects on qualified equipment were addressed in the UFSAR change submittal for the l extension of fuel burnup, HL&P letter ST-HL-AE-3906, dated October i 30, 1991. That document addressed the increase in radiation doses
! on equipment as the result of revised source terms as well as the reduced containment volume.
As a - result of the reduced free volume, post accident peak containment internal pressure and temperature have been recalculated to be 40.5 psig and 328'F. UFSAR Section 3.11 l addresses equipment qualification and Table 3.11-1 lists the Tscio2-ess. cot 7
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environmental conditions, including containment post accident temperatures and pressures which must be considered for equipment qualification.
Equipment inside containment is quallflod to operate in an accident i environment with pressures and temperatures equal to or higher than ,
, 57 psig* and 340*F*. Since containment equipment have demonstrated l l the ability to perform designated safety functions at pressures and ;
temperatures higher than the recalculated accident peak containment l internal pressure of 40.5 psig and temperature of 328'F, no adverse r effect on system performance is.cxpected. t SER Section 3.11.3 specifies that the acceptance limit for ' the f minimum containment temperature to which safety-related equipment ;
must be qualified to perform during-accident conditions is 323*F. ;
Therefore, the proposed change in containment acciderat temperature j from 323*F to 328'F js a reduction-in the margin of safety from J 17 ' F to 12
- F. This is not considered to be a slanificant reduction ;
in the margin of safety because equipment qualification testa prove ;
that equipment will continue to perform as' designed. Equipment is 7 subjected to qualification -test temperatures for durations l significantly longer than the duration of the - expected peak containment accident temperature. Although the margin-of safety decreases from 17'F to 12
- F -- a s a result - of the increase in '
containment peak ~ temperature from -323 *F to 328'F, the time- 1 temperature equivalency of the qualification test adequately !
exceeds the duration of the increased peak containment accident i temperature.
SER Section 3.11.3 indicates that the acceptance limit . for - the ;
minimum containment pressure to which safety-related equipment must t be qualified to perform during accident conditions is 48.4 psig. ,
Therefore, the proposed change in containment accident pressure ;
from 48.4 psig to 40.5 psig is not a reduction in the margintof !
safety because the margin of safety increases from 8.6. psig to j i
16.5 psig. However, the NRC acceptance value should be ~ revised to
- 40.5 psig to reflect current analysis findings since the existing acceptance value of 48.4 psig is based.on analysis findings which are no longer appropriate.
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- The Hydrogen Recombiners, RCFC motors, Marathon terminal blocks, i Radiation Monitors, and Thermocouple junction boxes are qualified-to lower values because they have no thermal or radiation sensitive ;
materials. The ef fects of the accident environment will not-impede i the safety-related performance-of this equipment. . Justification- i of exemption to these : area. wide requirements are documented in 1 Equipment Qualification Checklist Packages and calculation MC-6201. ;
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PAGE . / / '. 0F' /9 LL1 Electrical cable Temnerattirs --
The maximum temperature of electrical cables inside containment is determined in calculation NC-7063. Maximum temperature occurs during the DBA MSLB. Evaluation of increased temperature.effact associated with the reduced containment free volume results in the following changest Old Value Egv' sed Value 2-conductor cable Tmax = 293*F a # "F 4-conductor cable Tmax = 283*F 2fa*F Conduit Tmax = 281*F 286*F While there is a slight increase in the maximum calculated temperature, the peak cable temperature is below the environmental qualification temperature of 340*F. Therefore, the reduced containment volume does not affect the electrical cable qualification.
2.5.2 Polar Crane Beam Box Pressurization The pressurization of the polar crane beam box is calculated in NC-7015. The containment pressurization rate used in the original calculation is more conservative than the ones obtained in both the revised MSLB and I4CA analyses. Therefore, the results of the original calculation are bounding.
2.S.3 Polar Crane Temoerature The post accident polar crane temperature inside the containment is calculated in NC-7064. The peak equilibrium vapor temperature for the revised LOCA case, af ter initiation of the sprays, is below 260*F. The peak equilibrium vapor temperature for the revised MSLB case, after initiation of the containment sprays, is below 240*F.
For both the LOCA and MSLB cases, the peak equilibrium vapor-temperature, after initiation of thu sprays, is below the design temperature of 286*F. Therefore, the original polar crane design temperature is bounding.
2.5.4 Containment Air-Lock Seal Teperature l The personnel air-lock inflatable seal EQ temperature was calculated in NC-7055. The seal manufacturer established the seal thickness as 0.205 inch miniraum. Calculation NC *2015 assumed a seal thickness of 0.125 inch. Engineering judgement shows that increasing the seal thickness decreases the seal's peak temperature. Hence, the peak temperature for the 0.205 inch thick seal will be lower than the 283.1*F calculated for the 0.125 inch seal. Therefore, the inflatable seal original EQ temperature of 315'F remains bounding.
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i ATTACHMENT 8 ST.EAE-W W PAGE_l') 0F /1 L.L1 Rontainment Abngrmal Temperature The containment abnormal temperature following a loss of offsite power is analyzed in calculation NC-7049, ~ " Containment Heatup Following Fire Outside the RCB", and NC-7056, " Containment Heatup ,
Following Loss of Offsite Power". i The original calculated peak temperature of NC-7049 was 151.3*F at-72 hours after the accident. The effects of reduced containrient free volume result in a new peak temperature of 153'F at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The effect of reduced volume on NC-7049 does not affect either the design criteria of the Containment Building or internal equipment qualification.
The original calculated peak temperature of NC-7056 was 146*F at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The effects of reduced containment free volume result in a new peak temperature of 147.4'F at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The results of NC-7056 are used in calculation MC-5910.
Calculation MC-5910 used 146'F to determine the abnormal temperature rise in the containment. The air outlet temperature calculated in MC-5910 was 116'F. The 116*F temperature is added to the containment normal temperature rise to obtain the containment abnormal peak temperature. Applying the new value of 147.4'F to the applicable equation used in MC-5910 gives the air outlet temperature of 115.97'F. The temperature of 116'F, used in MC-5910 to calculate the containment abnormal temperature, does not change and the results presented in MC-5910 remain bounding.
The maximum containment abnormal temperatures are given in UFSAR Table 3.11-1. The abnormal temperatures given in Table 3.11-1 come from EQ Design Criteria (4E019-NQ-1009). Abnormal . containment temperatures in NQ-1009 are obtained from Calc. MC-5910. Since changes in NC-7056 do not af fect the results of MC-5910, the design abnormal temperatures given in the EQ Design Criteria (NQ-1009-9) and UFSAR Table 3.11-1, remain bounding.
L.5 Containment Leakace Per Technical Specification 3/4.6.1.2, the design containment leak rate (L ) is 0.3 wt %/ day. An Integrated Leak Rate Test (ILRT) is performed at (or greater than) the- calculated peak ' containment internal pressure (designatea P,) to demonstrate .that- the containment leak rate does not exceed this limit. ILRT acceptance criteria is 0.225
). P,,
+3/-0 psig, to wt %/ daymeeting ensure (.75 L imum min pressure tsquirements.Tl1 test pressure used Previous testing performed on Unit 1 at 39. 2 . psig yielded a measured leakage value of 0.123 wt%/ day, which is significantly below the acceptance criteria (45% margin).
isc\92-086,001 10
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ATTACHMENT I ST-HL AE %9 PAGE 13 0F _19 Recent (September, 1991) testing performed on Unit 2 at 43.9 psig yielded a measured leakage value of 0.065 wt%/ day, which is significantly below acceptanca criteria (71% margin).
2.,32 Containment Minimum BackDressure for LOCA ECCS Analysis The decrease in containment free volume affects containment LOCA reflood rate. The reflood rate impacts the fuel peak - clad temperature (PCT). The containment pressure / temperature analysis for LOCA reflood rate was performed by Westinghouse. For this calculation, a low containment pressure is conservative since it >
reduces the core reflood rate and increases PCT. Thejestinghouse analysis used a containment volume of 3. 56 x 10' f t , - which3.41 x 10' f t{s larger than the new containment volume of .
Therefore, the analysis performed by Westinghouse is still conservative and bounding.
L D, Containment Hydrocen Generation UFSAR Section 6.2.5 discusses C. oustible Gas Control in containment. Section 6.2.5 establishes conformance to regulatory requirements for the Combustible Gas Control System.- Containment- '
hydrogen generation is determined in calculation NC-7009. The effect of reduced containment volume on NC-7009 does not violate the regulatory requirements. The hydrogen concentration in the containment will stay below 3.5 volume percent. The Regulatory Guide 1.7 requirement is to stay below 4.0 volume percent.
SRP Section 6.2.5 requires a margin of 0.5 volume percent between the hydrogen concentration llmit (4.0 volume percent) and the hydrogen concentration at which the recombiners actuate. As documented in the revised analysis, which incorporate tLe reduced containment volume, the recombiners will be at full efficiency before the hydrogen concentration reaches 3.5 volume percent. !
Therefore, this meats the requirements of the SRP and there is no reduction in the meroin of safety. .
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- 22. Radioloaical Dose Analyses <
Radiological 7 dose analyses have been performed. in several-calculations. The results of-these analyses are used to determine doses to both onsite and offsite-personnel. They also ensure that regulatory limits are not exceeded.. The effects of reduced containment volume on radiological doses are addressed in the subsections that follow and are based on the conservative assumptions itemized below which are presented in UFSAR Tables 6.4-2,' 12.2.2-1, --
15.4-4, 15.6-10, and 15.7-9. In addition, the '
effects of extended burnup on source terms were previously incorporated per letter ST-HL-AE-3906, dated October 30, 1991.
- Reactor power-level of 4100 MWt versus 3800 MWt, ,
o Containment free volume of 3.2 x 10' f t3 versus 3.56 x- 10' f t3 ?
used for concentrations, o Containment free volume of 3.41 x 10' f t3 versus 3.56 = L 10' f t3 used for leakage, _
e -containment sprayed volume of 79.7% versus 77%, ?
e HVAC flows vary by -54 to +10% and either a high or low flow '
is used, depending on which was most conservative. l 1
2.Jd Puel Handlina Ag.cident Inside Containment .
The onsite and offaite dose consequences due to a fuel handling i accident inside containment are' addressed in calculation NC-6006 +
t and presented in UFSAR Section 15.7.4 and Table '15.7-10. The ef fects of the reduced containment volume have L .) evaluated in revision 5 to this calculation and the comparable results, along with regulatory limits are presented below ,
RESULTS: 3 Beta Skin (rem) Thyro. (rem) Whole' Body (rem) 1 Rev. 4 Rev. 5 Rev. 4 Rev. 5 .Rev. 4 Rev._3 ,
EZB 0.133 0.146 27.3 36.14 0.106 0.113 LPZ 0.039 0.043 7.99 10.56 0.030 0.033 -
(EZB - Exclusion Zone Boundary LPZ - Low Population Zone) l REGULATORY LIMITS: ;
Per SRP 15.7.4, dated July 1981, and SER (Supplement'6)-Appendix ;
Z, the offsite dose consequences are acceptable within the limits ,;
of 10CFR100,.-or 75 rem thyroid and 6 rem whole body. The revised ,
doses are all below these limits.-
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PAGE /9 0F /9 2.9.2 gpntrol Room (CR). Technical Suonort __Qenter (TSC) and offsite LOCA Radiation Doses The onsite - and offsite dose - consequences due to a LOCA are addressed in calculation NO-6013 and presented in UFSAR Section 7 15.6.5 and Tables 6.4-2, II.B.2-2, and 15.6-11. The ef fects of the reduced containment volume have been evaluated in revision 6 to-this calculation and the comparable results, along with regulatory limits are presented below RESULTS:
Beta Skin (rem) Thyroid (remi Whole Body (rem)
Rev. 5 Rev. 6 Rev. 5 Rev. 6 Rev. 5 Rev. 6 EZB 1.16 1.22 126.5 137.0 2.19 2.27 LPZ 0.43 0.47 58.42 66.3 0.68 0 . 7 *,
CR 18.70 21.52 18.21 22.67 2.42 2.43 TSC 21.64 24.55 24.85 28,62 4.74 4.85 (EZB - Exclusion Zone Boundary LPZ - Low Population Zone) t REGULATORY LIMITS:
The following limits are set by 10CFR100 and SER Sections 15.6.5.2.5 and 6.4 (dated April, 1986) for various areas:
Ecla Skin frem) Thyroiti ( rey _1. Wh. ole Body (remi EZB n/a 300 25 LPZ n/a 300 25 Control Room 30 30 5 TSC 30 '0 5
The revised doses all fall below these limits.
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A1TACHMENT I ST HL AE L de PAGE u, Q(0F / 9 2.9.3 Rod Eioction Acclient; The onsite and offsite dose consequences due to a rod ejection accident are addressed in calculation ~ NC-6014 and presented in UFSAR Table 15.4-5. The effects of the reduced containment volume have been evaluated in revision 2 to this calculation and the comparabic results, along with regulatory limits are presented below:
RESULTS:
Containment Leakage Contribution:
Beta Skin (rem) Thyroid (rem) Whole Body (rem)
Rev. 1 Rev. 2 Rev. 1 Rev. 2 Rev. 1 Rev. 2 EZB 0.036 0.040 27.93 35.7 0.11- 0.12 LPZ 0.026 0.029 36.46 48.9 0.069 0.076 Secondary System Release Contribution:
Beta Skin (rem) Thyroid (rem) Whole Body (rem)
Rev. 1 Rev. 2 Rev. 1 Rev. 2 Rev. 1 Rev. 2 EZB 0.0017 0.0017 0.84 1.0G 0.0049 0.0051 LPZ 0.0005 0.0005 0.25 0.29 0.0015 0.0015 REGULATORY LIMITS:
Per SPS 15.4.8, dated July 1981, and SER Section 15.4.8.2 (dated April, 1986), the offsite dose consequences are acceptable within the limits of 10CFR100, or'75 rem thyroid and 6 rem whole body.
The revised doses all fall below these limits.
2.d 1 Plant Buildina Airborne Concentrat,ip@
The onsite radiological dose consequences due to airbone concentrations are addressed-in calculation NC-6032 and presented in UFSAR Table 12.2.2-2. The effects of the reduced containment volume have been evaluated in revision 2 to this calculation cnd the results, along with regulatory limits are presentad below:
RESULTS:
BUILDING MPC FRACTION General MAB 0,0112 MAB Worst Room 0.713 TGB 0.000142 FHB 1
- RCB Refueling 0.967 i
RCB 9.64 (Continuous Purge Allows 4.1 hrs /wk Access) l l
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1 ATTACHMENT I ST-HL PsGE D AE 40Mo /9 OF ._
REGULATORY LIMITS:
All concentrations are within the limits of 10CFR20.103.
Therefora, the concont! rations calculated are acceptable and there is no impact. on the SER.
2.9.5 Reactor Coolant System Vacuum Decas System (RCSVDS)-
Releases to containment Calculation M':-604 2 is used to determine that removal of the reactor head will not cause 10CFR20 limits to be exceeded. The revised calculation (Revision 2) which considered the reduced containment volume only increased the fraction of MPC from 0.021 to 0.022. Thus, the change is insignificant. Therefore, the concentrations calculated are acceptable and there is no impact on '
2.9.6 Environmental Report Doses The onsite and off site radiological dose consequences due to Class 6 and 8 accidents as specified by Reg Guide 4.2 are addressed in calculations NC-6048 and 6050. The effects of the reduced containment volume have been evaluated in Revision 2 of these calculations. The revised doses increase a relatively small amount over previously reported doses, and the changes are not considered significant.
2.10 Safety Injection / Containment Snrav Operations Calculation MC-5521, " Containment Spray System Head Loss and Pump Capacity" demonstrates the adequacy of the CS pumps and piping to provide minimum required water flow - rate. The effects of - the reduced containment volume on pump operation have been evaluated.
The results indicate that the pumps are - capable of providing required flow rates under increased containment pressure conditions.
Calculation MC-6220, "SI/CS Pump NPSH" determines the available not positive suction head . (NPSH ) for the safety injection and containment spray pumps ciuring,the recirculation-phase of recovery ,
following a DBA LOCA. - The 7ffects of the reduced containment ,
volume on NPSH have been eval %Ced. The results indicate that the required NPSH, is not affected by the decrease in containment free volume.
Calculation MC-5645, " Mass Flow Rate of Spray Through Sprayed Regions of Containment" determines the containment spray mass flow i rate within the various regions' of the containment. The effects of the reduced containment volume on flow rate have been evaluated. .
The results of the analysis are reflected in a revision to UFSAR ,
Table 6.2.2-5. The results of this analysis are not used in the design basis. ,
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4 ATTACHMENT l ST HL AE-I PAGE .R OF(C66 /9 L_Q B IONIELCAET__ HA2 ARDS DETgBMINATLQM pursuant to 10CFR50.91, the proposed changes to the UFSAR and the
! Technical Specifications do not involve significant hazards as defined in 10CFR50.92 in that the proposed changes: j 4
(1) do not involve a significant increase in the probability or ?
) consequence.s of an accident previously evaluated.
The containment and safety-related systems inside containment I remain operable as previously analyzed. The changes in the !
containment volume, pressure, and temperature, do not increase or I cause an increase in the likelihood of a DEPSG, MSLB, or any other i DBA. The increase in peak containment post accident pressure [
resulting from the reduced free volume is bounded by the original :
design pressure.
Current containment leakage test results are well within acceptance f' criteria even with an increased test pressure which was greater '
than the recalculated post accident peak containment pressure. In addition, equipment inside containment remains qualified and
- functional in the changed environment and onsite/offsite doses and {
airborne concentrations remain with stipulated limits. In r addition, no changes in either equipment or operator actions are required which would affect either probability or consequences. ,
f (2) The proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. [
l All containment and sTfety-related systems inside containment will ;
function within design or EQ limits and all analysis results are ;
bounded by design and regulatory requirements. No new modes of [
operation are proposed or supposed. No physical plant changes are i necessary or proposed as a result of the changes. The proposed r changes do not adversely affect or alter equipment which is assumed to operate in accident conditions. In addition, no new-operator -
actions are required and no existing actions are either altered or deleted. I (3) The proposed changes do not involve significant reductions in I l
the margin of safety. ;
l i The margin . of safety as defined in . NSAC 125 (Guidelines for .
10CFR50.59 Safety Evaluations, Nuclear Safety Analysis - Center) {
Section 3.8 is the dif ference between a failure point and the i acceptance limit. ;
1 l
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ATTACHMENT / >
ST HL AEf l olk PAGE .. /9 0F > 9 The containment pressure acceptance limit is the containment design pressure of 56.5 psig per SRP and SER Sections 6.2.1. Since the calculated peak containment internal pressure resulting from reduced free volumo (40.5 psig) is below the acceptanco limit, this change does not reduce the margin of safety.
Current leakage rates are significantly below the acceptance limits. The margin between the measured value and the acceptance value was 45% on Unit I and 71% on Unit 2. These margins provide reasonable assurance that actual leak rates will remain within the acceptance criteria- even at the higher pressure resulting from the reduced volume. Therefore, the increase in the calculated peak pressure does not decrease the margin of safety.
Although the proposed change in containment accident peak temperature from 323*F to 328'F results in a reduction in the margin of safety from 17'F to 12*F, it is not considered to be a sionificant reduction. Equipment will continue to perform as designed because the time-temperature equivalency of the qualification test adequately exceeds the duration of the increased peak containment accident temperature.. In addition, revised onsite and offsite radiological doses and airborne concentrations do not exceed previously accepted values.
i Tsctwoes. cat 17
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APPENDIX A MARKED-UP UFSAR PAGES
ATTACHMtNT 2.
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- d. An external pressure of 6,159 lb/ft 8 based on a 415 kip /ft hoop stress resultant from 10 to 45 degrees on the done d
For vertical done tendons:
The vertical dome tendons produce an external pressure of approximately 5,465 lb/ft8 This pressure varies over the surface of the done.
4 Design Basis Accident (DBA) Pressure Loads (P.)
(P.= 56.ser 0 The minimum equivalent static design pressurekis chosen conservatively above the peak pressure occurring as a result 'of a L^A up :: :.nd-4eehdingMDBA (see Section 6.2. K for Containment' pressure response analyses f:: v;.ri M S-LOCAs-) . :T:..-deaigr. presswre 1: P. 55.5 p;ig, c
- 5. Operating and shutdown Thermal Loads (T.)
Operating thermal loads are the most severe thermal conditions for summer and winter operations, normal loads are determined'on the basis of temperature distributions obtained by heat transfer computations. Reference temperature during construction is assumed to be 60'F. ne following temperatures are used in the analysis of the containment structure:
Operating Shutdown Sun Operating Thermal loads (T..) Case Case Notainment inside temperature 120'F 65*F Outside air temperature 95'F 95'F Soil temperature 75*F 75'F Uinter operacing Thermal Loads (T.,,)
Containment inside temperature 120*F 65'F Outside air temperature 25'F 25'F Soil temperature 75'F 75'F
- 6. Test Thermal Leads (Tg )
Thermal los.ds during pressure test, including liner expansion and temperature '
gradient in the vall and dome. The summer and winter operating thermal loads (see item 5 above) are applied as the test thermal loads (Tg) in the design of the Containment. *
- 7. Oserating Pipit.g Imads (R,)
Piping thrust and thermal expansion forces and reactions based on the most eritical steady-state or transient condition during normal operation or
=huciot.rn (Sectir 3.6). ,
- 8. Design Basis Accident Thermal Load (T.)
Additional thermal effects on structure above normal operating loads, resulting from a -LOC #, up cc and kchsing-ehe#DBA.
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- 9. Op: rating Ba;io Earthquako (OBE) 14 d* (E.) PAGE.3 0F W7 leads generated from the 0:. The plant is designed to remain operational
- under the OBE. The OBE loads are based on a maximum free field ground acceleration for the site of 0.05g.
In addition to the structural responses, dynamic soil pressures are applied to ,
the structure. The dynamic soil pressures are calculated by the Mononobe-Okabe Method using the same satsnic accelerations as used to determine the
- structural response (Section 3.7).
- 10. Safe Shutdown Earthquake (SSE) Loads (E,,)
1 Loads generated for the SSE. The structural response and corresponding
( dynamic soil pressures are determined for the SSE based on a maximum free- ,
l field ground acceleration for the sita of 0.10s (Section 3.7).
I l
- 11. Wind Loads (V)
Leads generated by the design basis vind,. Wind loads are calculated based on a design wind velocity of 125 mph (Section 3.3). The appropriate pressure e coefficients used in calculating the design wind pressure are obtained from American Society of civil Engineers (ASCE) 3269, ' Wind Forces on Structures',
for the cylirder and ASCE 4933, ' Wind Loads on Dos.e-cylinder and Dome-Cone Shapes", for ..se dome.
- 12. Tornado Ioads (Ug ) .
Wind, pressure differential and missile loads generated by the design tornado. '
The design pressure tornado load is calculated similarly to the wind load
, using a tornado wind velocity of 360 mph and a gust factor of 1.0 (Section ;
l 3.3).
- 13. External Pressure Load (P,)
l
\
l External pressure load of 3.5 psig resulting from pressure variation either inside or outside the Containment. ,
14 Test Pressure Imad (Pg)
(Pd The test pressure is equal to 1.15 times the DBA pressure l, in accordance with [ -
l Section CC-6210 of the ASME-ACI 359 document. A l 15. ' DEA Thermal Piping Loads (Pg) l Additional pipe reactions and forces above normal operating loads, due to l
thernal effects, occurring as a result of a -LCC/. up te and including d./DBA (Section 3.6).
(
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. ATTACHMENT 1 STPECS LTSAR ST HL-AE 40%
4 PAGE 4- 0F IV 7 3.8.1.3.2.2 Nonservice Load Catecorv This category includes all loading conditions resulting from a system failure and/or those extreme environmental conditions postulated to occur during the life of the plant. !
Also included in this category is the Severe Environmantal Condition. The loads in these conditions occur infrequently in combination with normal -
operating loads. The design probability of occurrence of some of the infrequent loads, such as the OBE, is one during the life of the plant, while one. of other extreme loads, such as tornado and the SSE, are much less than _
that ;
- 1. Severe Environmental Condition This condition considers all the normal operating loads on the structure in '
combination with the loads resulting from an environmental event, such as wind or the OBE, which may occur only infrequently. ;
- 2. Abnormal Condition gg gg g, y a % (Q % t'e ThisDRA.
GLhe\ condition
\ \ ' includes
' chef7 7(eksute \anAtgrgturele_ffects retultitigM i
bcM N Basis /tcuocNT *N e**m%. LOAD (Th, '
- 3. Extreme Environmental Condition This condition includes loads resulting from environmental events which are credible the design but are highly improbable. These events include flood, the $SE, and tornado. '
4 Abnormal / Severe Environmental Condition This condition includes highly infrequent, and severe environmental effects. simultaneous occurrence of abnormal 5.
Abnormal / Extreme Environmental Condition This condition includes pipe rupture loads and direct pressure and jnt .
impingement loads generated by a postulated rupture of high-energy piping .
The condition is the highly improbsble, simultaneous occurrence of a > normal and extreme environmental offects.
A summary of the nonservice load combinations is shown in Table 3.8. l.l.
3.8.1.3.1 LLad_ Combinations on tocalized Areas; t los. .ized at las, such as penetrations,. chell discontinuities, crane girder brackets, tendon and anchorage renes, anU local areas of high thermal gradient, are desigted for '
the same loadiv.g cotbinations as the Containment. In addition, local effects due to geometrical and mechanical discontinuities are considered.
3.8.1.3,4 Effect of Induced Strains on the Liner: Due to the prestressing forces and the DBA temperature effect in conjunction with other loadings, order che steel liner plate is cubjected to compressive stresses. In to prevent instability and excessive deformation in the liner place, continuous stiffeners are provided to anchor the liner to the concrete. The spacings of the stiffecers are determined such that the liner stresses aad strains are in accordar.ce with Section CC-3700 of the ASME ACI 359 docume nt. .
3.8-10 Revision 0
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AT1ACHMENT JA STPECS 6 TSAR ST HL AE- O M OF /'/ 7 PAGE M 3.8.2.1.3.1 El'etrical Penetrationt - Typical electrical penetration is shown on Figure 3.8.1-12. Design details are discussed in Section 8.3.
3.8.2.1.3.2 Pinine Penatrotions - Single barrier piping penetrations are provided for all piping passing through the containment wall. The closure :
of the pipe to the steal liner is accomplished with flued heads, pipe caps, or '
plates buttwelded to the pipe and penetration sleeve. In the case of pipin5 carrying hot fluid, the pipe is insulated. Figure 3.8.1 11 shows a typical high lines,energy line penetration. For single pipe penetration for moderate energy see Figure 3.8.1-10. The MC classification extends from the containment liner to the flued head or cap of the penetration.
3.8.2.1,3,3 Fuel..Iransfer Tube . A fuel transfer penetration is provided for fuel movemen* between the refueling canal in the Containment and the fuel transfer canal n the nib. The penetration consists of a 20 in.
outside diameter stainir. s steel pipe that acts as the transfer tube, and is fitted with a double ga.seted blind flange in the refueling canal and a '
standard gate valve in the fuel transfer canal. The casing stainless steel pipe is provided with expansion bellows and is connected to the Containment steel linar penetration. The transfer tube sleeve assembly is fitted with a test connection which permits local leakage testing of the expansion bellows.
For typical details, see Figure 3.8,1 8.
9Nb 3.8.2.1.4 mainta cone Desien Bases: Containment penrtrations are designed to event o ent integrity during-normal operation of the plant and in the of the Class All Containment penetrations are designed to meet the intent components of the ASME &&FV Code,Section III. Penetrations are designed in accordance with NRC Ceneral Design Criterion (CDC) 53 of -
10CFR50, Appendix A and, in addition, are designed to menc the following considerations:
+
1.
Ability to withstand the maximum design pressure that can occur due to the postulated rupture of any pipe inside the Containment. >
2, Ability to withstand the jet forces associated with the flow from a postulated rupture of the pipe in the penetration and maintain the integrity of the Containment.
- 3. >
Ability to accommodate thermal and mechanical stresses encountered in '
normal operation and other modes of operation and testing.
The anchorages of all penetrations to the Containment vall are designed as Category pipe rupture, I structures and thermal to resist and all forcesloads.
seismic and moments caused by a postulated and welds to the .iner are full penetration welds.The penetration assembly welda 3.8.2.2 Aeolicable Codes. Standards and Specifications.
3.A.2.2.1 Basic code: The basic code for the design, materials, fabrication, testing, and examination of these steel items is the ASME B&PV Code for Nuclear Power Plant Components.Section III, Subsection NE for Class MC Components. .
i 3.8-43 Revision 0
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AUACHMENTJ4 ST.HL AE 40W D ... Dead loads _0F g
_F' AGE.L.
L .. 1.ive 1.; ads F ... Prestress loads Pg .. Test pressure Tg .. Test temperature T .. Thermal conditions effects and loads during startup, normal operating and shutdown I F, .. Piping tions reactions during startup, normal operating and shutdown condi.
a E, .. Leads generated by the OBE
- E , . Loads generated by the SSE l P .. Design sca! M t p m 'Sn$ts Acc'Osur ?Res20RG Lono k T. -
FW W pesogn J r
oasts Au nOeNr &conm. LoAO .
J l R .. Pipe accident reaction Y
.. Equivalent static load on the component generated by the reactions on the broken pipe, jet impingement and missile impact during the DBA.
P,..
Subatmospheric pressure load (external pressure)
The load Table 3.8.2 combination
- 1. utilized in the design of Class MC items is shown in 3.8.2.4 Desien and Analysis Procedurek. The C.i
- MC items are analyzed and designed in accordance with the applicable requirements of ASME Code,Section III, subsection NE.
hatch, personnel airlock, and auxiliary air'ock are performed by a selected ve-dor using appropriate conventional engineering methods.
! 3.8.2.5 structural Accetirance criteria. 'Ihe structural acceptance criteria III of the ASME for Class Code. MC items are in accordance with Article NE.3000 of Sect The design is such that all the stress and strain limita defined in Article NE.3000 are satisfied for pres *ure loads in combination with all mechanical loads atid tl ernal loads.
The requirements of RC 1.57 are co ailed with.
3.8.2.5.1 cene ral ,Grita?.la:
design criteria for Class MC The ASME Code,Section III, Subsection NE 1
limits which vary according to the following factors:l cess are based on establishing st
- 1. Type of stress, such as primary stress secondary stress. and peak stress.
- ,8 45 Revision 0 i \
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!!ydrostatic loads and crane loads (without lifted load) are also created as dead load. '
Superimposed and/or suspended loads whic.h account for piping, cable '
trays, ductwork and miscellaneous equipment distributed throughout floors, are permanent live loads which are considered as equivalent dead loads. All of the dead load components are considered at full value in all loading combinations, including the seismic loading combinations. -
- 2. Live Loads (L) I T1oor occupany loads which account for movable equipment, personnel and ,
maintenance loads including construction loads, are designated as temporary occupancy live loads. Laydown area loada ara designated as .
permar.ent live loads. The live load components designated as temporary occupancy live loads, as defined above, are subject to a 0.25 reduction f actor only when considered in the seismic loading combinations. The liu load components designated as permanent live loads, as defined above, are considered at full value in all loading combinations, including the seismic loading combinations.
- 3. DEA Pressure loads (P.)
The equivalent static design pressure acadings within or across a compartment occurring as a result of a -120.'. up :; :nddeelvding DBA [ti or a rupture of high energy line. ~
4 Operating Thermal leads (T.)
Thermal effects on structures based on the most critical steady. state or transient condition during normal operation or shutdown.
- 5. Operating Piping loads (R )
Piping thrust and thermal expansion forces and reactions based on the most critical steady. state or transient condition during normal operation or shutdown. ,
i
i Additional thermal effects on structures, above norm '. operating loads, resulting from a 1^".'. up :: nd incl dias d # DRA or a rupture of high energy line. X t
- 7. OBE loads (E.)
Ioads generated from the OBE.
- 8. SSE loads (E,,) ;
leads generated from the SSE. .
3.8 55 Revision 0 ,
L
STPEc5 UrSAR ATTACHMENT 2.
(-
Additional pipe reactions and forces, above normal operating loads, due to thermal effects occurring as a result of Mct, q n ad-hehding-the4
- a. DBA or a rupture of high energy line.
)( ,
- 10. Pipe Rupture loads (Y)
Pipe reactions which account for the dynusic effects resulting from postulated [
tvpture of a high. energy pipe. Also included in this rupture loading are direct jet impingement pressure and missile 12 pact effects generated by or durin} the postulated break.
3.8.3.3.2 toad combinations: The design of the Containment internal structures, (except RCS equipment supports, control rod drive hechanism,
[CRDM) lock lugs and embeds, residual heat removal [RHR) pump supports, RitR heat exchanger [10C] supports, and fuel transfer tube supports which are covered in Section 3.8.3,3.2.3) incorporates two general loading categories:
the Service load Category and the Nonservice load Category. F,ach of these categories is divided into several conditions of loading..which are further subdivided into several different load combinations, as described below.
3.8.3.3.2.1 Service Lead catenorv . This category includes all loading conditions encountered during the construction, normal operation, and shutdown periods of the nuclear power plant. The probability of occurrence of these loads is 1.
A summary of the load combinations for the Service Load Category is shown in Tables 3.8.31 and 3.8.3 2 for concrete and steel internal structures, respec.
tively. The concrete and steel internal structures are analyzed and designed to meet the strength requirements for the Service load Category in accordance with the structural acceptance criteria stipulated in Section 3.8.3.3.
3.8.3.3.2.2 Nonservice fend cattggry . This category includes all loading conditions resulting from a system failure and/or those extreme envi.
ronmental conditions postulated to mur during the life of the plant. Also included in this category is the Severe Environmental Condition. The loads in these conditions occur infrequently in combination with normal operating loads. 'The design probability of occurrence of some of the infrequent loads, such as the 03E, is 1 during the life of the plant, while that of other extreme loads, such as the SSE, is much less than 1.
- 1. Severe Environmental Condition -
This condition considers all the normal operating loads on the internal structures in combination with the loads resulting from an environmental event, such as the OBE, which may occur only infrequently.
- 2. Abnomal Condition This condition includes the pressure and temperature effects resulting from the DBA. It may also include pipe rupture loads and direct pressure or jet impingement loads generated by a postulated high. energy (
pipe break accident.
3.8 56 Revision 0 s
, ~ , - - - . , , . - . . . . . , , , _ , . . , . . . ,
f 48tf 3.8.1-Ta (centinued)
INDiete ("4tHTim Fat Dfsfou 4eo Flest awaitSit ce costargutwT 5iT11 vot et ter:
TMS 3 salSEJtr hr4 C M 94 (_o40
.e.,
,,e i s ,
,e_,
3:'=ve:ay* %
ag = Intelet pr,sts ne t I'et prenewe (.1.ts p,3 F
- flnet prestress t (se m t to T ,3 a merest aperatig tosperstwe ' # N I"I f,
,es,4;?,d,,-_e;-o- r. -
weti, wie we g
e . . ,,, ,,s., i.e -
se<e -- we n set
- Torneds temi. (including differentist O g e
- Pipe reactlene daring nerent t operstles er shuth conditlens e t m elsellee) g E,
- Plpe reactlene above normal operating I N II eI $ = ploed good P--e wD r
ca s $i
!? c2 o-e~
- O 6 b.
MC 3
o I
Q tocol toeds are not ceneidered in the overeti enetysts but are teken Inte occar.t in local design.
4 t
f attt 3.5.1-73 (e) (b) (Continued) statst osatTTIS etS4ff
,eieii - ses htE8tM41 l090 s**$ EAT 0 = Deed leed f, = Dnlan eccl eDtge metet t.ecet leeds are feet considered
, (#efest\ererst(ne\le ((,)
In the everett snelyele but are [
but are toten late account in tecet deelen.
L' = tlwe load Pt a fest pressure (=1.t$ p,3 4
fg = Initial prestrees it a fest temperewe (somsel mi to T )
j f = finet prestrees P, = 5estyi enternet pressure (vocuum) f~ 1, e norset operettg tempereture t, e tverating beels eertfuquake I M 5:5 LonD' I. P,
- Desl y eIdent pressure f,= Sefe shutdews eerrWe -
[i u = Wind toed Ug = Tornede toeh (includies elfforentlet pressure and tornado elssiles)
H i
P
- tn en R, = Pipe reectione 4srig tieraal i e #1pe rupture toed Q ;
eperating er shutdown conditlers C
e, = Pipe reectione above norest u =. Flood toed N eterotice toede (f.) ' $
I i'
't 1
. "U C/2 4->4 DN
- e. slyn Conventions are: $' C
]
Stresses and Streine........(*) tonelle........(-) conpreselee k>O Z !
- b. The stresees were obtelned free SPitos computer output. D .
-c z t i o -4 y c. actuel cyttnder bree6e of cercrete reeutt in en atlewebte of 3864 pet. o ,
a n
< 1
- d. The sectlen la asemed crocted i h n concrete atrees le in tonelen. 1 5
-C o e. the esncrete etteusbis strees le 4000 pel booed en concrete test etrenett fg e deos pet. -Q i
i 4 O f.. The elleueble ter strees een be Increened 331/3 percent & rig teet cond!tlen. '
- s. Allouable liner straine sw are beoed on the loweet values free the lowest wetwee free the ainE Code. SeG4n f e olvie sert 2.
' ~
4
\ \
, _ - -.-...._.m... . , , , . . - - - - , , , - _ , - - - . . . - _ - , ., , , _ . , , - , _ - - ~ , , - , , . . , , - , - . . . . _m. _ . - - , - , , ,r_~...,, , ,c.._, , - , _ - , - . - - . , ---- , _
l y.
ATTACHMENT 3 STrEc5 UTSAR ST HL ,AE you PAGE4-- 0F.11CL __
3.11 ENVIRONMENTAL DESIGN OF MECilANICAL AND ELECTRICAL EQUIPMUiT I Safety related mechanical and electrical equipment ie designed to remain functional during and following design basis events. In addition, certain post accident monitoring equipment, is also designed to remain functional l
during or after specified design basis events, or to not fail in a manner which could prevent satisfactory accomplishment of the plant safety functions.
Design basis events consist of normal operation and plant shutdown, loss of offsita power (1DOP) and design basis accidents (DBA).
The following sections provide information to demonstrate acceptable perfor-mance of Non Nuclear Steam Supply System..(NSSS) (i.e., balance of plant)
I equipment as well as NSSS passive mechanical equipment under the specified conditions. Environmental qualification for NSSS equipment is discussed in Section 3.11N.
The programs for preventive maintenance, surveillance and periodic testing have been developed in accordance with Regulatory Guide (RG) 1.33, Rev. 2.
These programs are based on manufacturer recommendations, experience and the results of the project qualification programs. This will ensure that all safetp related equipment in mild and harsh areas will be operable and qualified throughout the life of the plant.
The programs provide for replacement of parte and equipment prior to the end of qualified life.
3.11.1 Equipment Identification and Environmental Conditions A complete list of safety related electrical and mechanical equipment (inclu-ding NSSS passive mechanical equipment) required to be qualified is prot ded in the 10CTR50.49 submittal. A list of all Category 1 and 2 post accident monitoring equipment (in response to RG 1.97, Rev. 2) that is included in the h/otJ4 cg equipment qualification program is provided in Table 7.5197avuormental - opetfed conditions for each area in which the subject equipuent is installed are listed in Table 3.11 1. The conditions are based on the following:
- 1. Normal parameters are thoaevvhich will be maintained during j routine plant operation, shutdown, hot standby, and system testing. The range is based on the limiting conditions of peak outdoor temperature together with equipment design heat loads and minimum outdoor temperature together with no heat loads,
- 2. Abnormal parameters are those which may be caused by such events as loss of nonsafety related heating ventilation, and air con-ditioning (HVAC). The majority of qualified equipment areas are served by safety class HVAC, for which outages due to IDOP are not postulated.
- 3. Accident conditions are those plant conditions resulting from the most limiting pipe failure for that location during which safety. ,
related equipment must operate to mitigate the conseq,2ences of the accident. The length of time that each item of equipment is required to operate in the accident environment following accident i initiation is provided in the 10CFR50.49 submittal.
3.1* 1 Revision 0 l
$TrEc5 UrSAR ATTACHMENT 3 sr.HL AE.yoM PAGE J1__ op 3).)
Class 1E cables, field splices, and terminations for use on the South Texas ,
Project Electric Cenerating Station (STFEC$) with the exception of single '
conductor high temperature silicon insulated cables meet the requirements of >
IEEE 383 1974 as modified by RC 1.131. Single conductor high teanerature silicon insulated cables when used in a class 1E circuit are in.- led in '
conduit only. Polyethylene cables used in safety related applica6 ions are identified in T.tble 3.11 5.
- i 3.11.3 Qualification Test Results Detailed qualification results for electrical and mechanical (including NSSS
- passive mechanical) equipment located in a harsh environment appears in the 10CTR50.49 submittal. Plant specific evaluations are peirformed to ensure that the generic testing performed by vendors encompasses the plant specific environmental conditions. The qualified life of equipment is extended or reduced based on specific plant varis.bles such as environmental parameters, operational cycles, parfonnance characteristics and properties of the materials used in construction of the equipment.
3.11.4 Losa of ventilation The majority of qualified equipment areas are served by safety class HVAC.
These HVAC systems are designed to the single failure criteria and are supplied from the Onsite Standby Power System. Consequently, the normal environ:nental conditions which they provide vill he saintained during all plant modos. However, certain areas of the plant served by safety class HVAC may experience abnorme.1 temperature conditions due to loss of offsite power which would result from switchover to a different cooling medium.
A small amou...: of qualified equipment is in areas served only by nonsafety HVAC. For these areas, the abnormal ranges of environmental conditions are i based on the loss of HVAC.
Table 3.11 1 provides a Ifsting of the worst case environmortal corjitions for various arsas in the plant. These conditions were determined by the criteria listed in Section 3.11.1.
3.11.5 Estimated Chemical and Radiation Environment 3.11.5.1 Chemical Environment. Safety.related systems ::M '--!;;_:d ::
r----.- - - e-f**c t ' :: S th; 22 F:70*we':, ; ::::::, 2nd H nidity ::nd':Sn: '
1hud F Obh 2.11 1 : .d n:tha 1.2. k :ffi:ba, components . _ _ f . .,
gg,g edet:f :y:::me are desiped to perform their functions on long torn contact with S : b _:id andj'~"i _ ;yJi..JJ. solutions recirculated through the p Emergency Core Cooling System and containment Spray System. '" n t e ,* r : n : :p::y follow,v,g '
p:: :ad ; p p:: _r. d.. sit.J in Sv. ilv.. 0.^,.
. Os LOCA, Ag susha*l pH of 4.5 m The Containment atmosphere is maintained below 4 volume. percent hydrogen consistent,e=.
e:nn:1: with = the=acai.a..arecommendations n.c.J of ..RC 1.7,
.u. as discussed in Section 6.2.5. J
- .u-1. -
h h due b de /Miew e? zooo-4ec,o prm bonit ec4t.Then.4% pH g An '
't.o - 9.5 due. 4. odhan of Tsp (%H I adds Na Oil u4tl SAT deleks'en cmIk p 1992.). Spg Aowne, U %l W A4~1a frem o A o.5 p/M . haay dad *tew it 24 A.wo p*g AccoAMAtMt. M 7E(6 325-197Y, Revision 0-i f
f .
i
-. _ . . _ . . - - _ _ , . . . . , . . _ , _ . , , . _ . , _ . _ . , - - - , - , ._ _ _ , . . . . . ~ , . . - , , _ _ _ ,,w....,-... ,_~.m , _
-__ _ _ _ _ _ _ _ _ _ _ . - _ - - - _ - . . . . _ _ _ _ _ _ _ _ _ _ _ - _ _ _ . - , . - . . _ ~ - . . . - . . . . . . - - . - - -.
I I
._ a ,
t
. t i
4 TAB 1.E 3.11-1 ENY1PN M!?TA!. Ct*DITItars 1
- Location Teunerature Pressure R.1stiv.
Mmidity C oulativ. R.dist1.nV' ;
Desame ;
(Envitorusental Normal Rang. Abnerwei Ace t dent'*"* Normel R.ng. Acetdent permal Accid *nF hotet t.*,
Destenster) fear / min.
- F1 f
- F1 Mermal Acciden F fear / min. 11 3
ft1 freds) freds) Tree i
- 1. Reacter containmente t. e ae -
328 gog feeld. CRDPt Stirewd**' 200/65 225/. -4e9- 0.3/pstg -S't pstg 70/0 100 7:10' men 1.2:10" gesume e.=
1 ena
-0.1 pstg -3.1 pets bet.
ein ein 328 155 Ste.= cenerate F 120/65 159/- -49 +0.3/psts H.pstg '0/0 100 2:10' 1.2s10'
(
Compartments g.gma .,
P (pm. 201) mes
-0.1 pstg een 3.1 pstg
.c4 g 5 =tn ein b te U E 329 Pre.svrir.P se..$ G 150/65 197/- 49 +0.3/psig 44 psig 70/0 100 2 10' Enclosura 1.2x10' ;essee g een mx (Re. 206). .33
-0.1 pstg -3.1 psig bete
, ein ein 4.,
l- 1. 1.oCAn c' eelstry eendittens er f.sredfeiel i.e MN 3.II.S. I .
-:_ - :1 : . - i
. :e _ :.: _ :e.
. - - ,. - .. w. .n ., .. . ~ .
..,_: ,._, .._ _: - 2 . .. : ...., .. r - -- -
4
.r :-- ; . ^ :_ . ; ._ ^l.
. = :c -
_, ._ _ . m -
m .
. _ _ " -. ,; m, " .. ._,,m. .m-
- _ _.,__ m ..,___ .. . m _ ...--
, 3 ..-__. _ - - _ _ . -
1 -
a -. ;. . . . -._- --- ,, -- - - --
-..v
...m . _ . . _ - ..
- z <. R i.. . .. 1 . c.,r. ne ti,. w-7..r not.cr.e.4 ..p.sw.. cor -t .p.t.ti
= . _ . . ._. -
., - -- ~- - - a 1:0 *7. , e-.nto.ne c.e .a to e 1.e.N y> >
g-.m
- 3. 6. p.:. Contpe 1 r.41.ei e nr.o.tec. ... . 17 . #ti. .ei.ene ,- t-. . . g- b.e.. m zwi r-A 8 $oZ ,
3mE
- m,
~D 4 O P Li ri pr
. P
i 4 8 i
I i
TABLE 3.11 1 (continued; ENVIR2etEMIAL C0ff0!TIONS !
2.-
3 1stive Cisseletive Radiettend
, imention Teamereture Pressure hetoftv t;;s; i (Environmental Normel Range Abnormel Acetden F ';
Bernel Range Acetdent Norsel Acetdenf*** Endletten i De s t mator) fman/stn. *n f *n Wernal Accident *** f am= / min el iti frede) frada) tyne i
318 4o.5 RoseteF 133/65 142/- .4et +0.3/ pets- -9t ysta 8-5
- 8 l2 j 70/0 100 3.Sa108 4-e=10 gamme Cavity men men bete (Ras. 001, 002) -3.1 pets i
[ -0.1 pelg ein ein 2.5m10" (pg g neutroa ,
{ 315 95 5 r.S Other Areae Insid F 120/65 167/- 4e9 oc.3/ pets .44 peig 70/0 100 3,5m10' 4-410' gamme .
! Secondary Shield sea een 1 l end (Below E1. 19 ft. -0.1 pelg -3.1 pelg bete Rs. 004) ein ein 328 5 '
165/. +es i
other Areas Inside Secondary Shield 120/65 e0.3 pets men een pets 70/0 100 2m10' 1.2x10* gemme and lI ,
(above E1. 19 ft) -0.1 pets -3.1 pets '
kte i
ein ela 5
@$ elle $ g.$ }
Other Arose Outside 120/63 168/- *?? e0.3/polg 44 pelg 70/0 100 3.3x10' 4-+=10 8 5econdary Shield mez games ^ [
men
-0.1 pets -3.1 pelg mg end g j g kee g 3-4
. ela min 328 g.5pets ;j 4
. RHR Finer and Meet 123/65 167/- 4tt +0.3/ pets 70/0 100 -7 108 1.2a108 gamme N Eschenger Roemet og velve Reces (ame. 104, set 0.1 pets met
-3.1 pets a=d beta g i 109, 110, 303, 304, sin sta 306, 105, 108, 111, i 202, 209, 207)
Tenden Accese Ce11ery 95/30 120 NA etc. etm. 80/23 100 100 100 gamme
. (ame. 011, 013) l.2
- t 9
i ,
i
-t cf, >d e
- i. D -
O*M j r
ser mK> ro -
. e di % hl QmE W m l es
. 2 .
i [ '
I y oc""4 w
'W i- -
M i,
4 t
t
I ,
- I 1
4 TAB 12 3.11 1 (Contismed) "
mitmeimA1. coNomoNs i A -
f Im atinn T-Bolettro C malative Radiatte #
_i.ture Zgeseurs m 'd1tv h-
-(Environmental Nernst Range Abnorme1 " AccidenF Normal Range Aceteent Nermal Deaf = amter) famw/=fe Aceident""" Radiation *
- F) (*F) 1 wl Accident" f anu / min *3 ft) femda) fenda) Tyne i
j 2. Isolatten Yalve cubielse j and Fenetration Arman l
$- Watertight teeme 129/40 129/40 225 ats. 4.5 peig 80/20 100 100 100 l (ame. 001. 002, 003 gemme i
. 004) ),
i t Aux. Feedwater Pim;- 104/30 104/40 335 elightly 5.0 eets 90/29 100 2 d - 100 1.3m10' gamme Cubicles (Ree. 00$. Peettive
{ 006. 007. 008) end t
- - Flatforme. ;
' E1. 21 ft 2 in. (Ams.
101. 102, 103, 104) !
i i- Feedvetor Cubtelee 104/S0 104/36 333 eltshcly S.8 pets l j 80/20 100 100 1.5 10' games ;
(Res. 201. 202, 203 poettive a
204) and Feeduster [
1.ine Area (ame. 301 t9 f i
\'
sa 302, 303, 304) *t
.l 7 steen 1trie Aree n
'r .(p.e, 401, 402, 1M/50 1M/36 335 alightly 'S.8 peig 00/20 100 - 100 3.Ss10s ,,,,, g
'403. 404) peettive g t t -
fame. 301. S02 .1M/30 1M/34 335 slightly 2.5 peig . 90/20 100 100 3.Sa10' gamme 503. SM) Peettive
- 3. Electric.1 a==t11.rv '
La11 ding 4
Electrieel Pene. 104/50 104/50 104 etm. etm. 90/20 to 100 a.Sa10'e - gemme tretion Reen ( 2.
3 (Rm. 001) s
+ Neueren Flus Aglifier Channel A Panel 212 683 has an eeeldent dose of lees than 10' rede (NC9040). p }&
gM 'j l7 m r 3>
ro t i
< M i
w-
" E Zn i O -4
- h C
~
L
{ -
k i .
2 i i
. . . . - . - . . . . . _ _ . . . , . . . _ .. . . , _ . . . . , . - , . . . , . . _ . . . . . _ . _ . . _ _ , ,. . . . . ~ . _ . - . . , _ _ , - - _ , . . _ . - - _ - - . . . - . _ . - . _ . . . . - . _ , _ _ . _ - -
' ! - >; l (, tl .
l p'lI ! Ll { e .rLb >! t:t! sI;'
ji o2 -
,44 c] yWx>r m - n>rm%4 N y 2 f.5O C e t
cm k , %R 4 a 2 7 2 3 g l gg g[
n e .
t t a e e e e , , e e e a e ,
m , , m ,
l r dT E
m m
a g
m m
a g
m, g,,
, g m
m a g m
m a
m a
g m
m a
g m
m a
g ,
s e -
r
&ne Fn1 8 8 s s s s t u
t e= 0 0 0 0 0 0 a r
t dd 1 1 1 1 1 1 e l a m m x 0 e
p l er 3 5 3 5 5 0 0 0 0 3 d- cf 0 0 0 0 0 m e- A 1 8 1 1 1 1 1 1 5 1 1 e w R t e- s
- ve e a i e t P
t r . .
a l) l e
aa t n w md m ra 0 0 0 0 0 0 0 0 0 0 0 e u or 0 0 0 0 0 0 0 0 0 0 0 c C Nf 1 1 1 1 1 1 1 1 1 1 1 a -
j d
t a n
e o 0 0 0 0 0 0 0 t o 0 8
0 8 h e -
dl ie t 8 8 8 8 8 8 8 t 9 cf e,; n 3 ve A o p
n.
+
tf ta u ee e -
i, g1 d lh B
e nt Ra _
n 0
2 0
2 0
2 0
2 5
4 0
2 0
2 0
2 0
2 0
2 6
2 b e
e c
w g
li / / / / / / / / / /
am n/
0 8
0 8
0 8
0 8
5 7 8
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8 0
8 0
8 0
4 0
8 e
r ,
rm a em
) s Bef d e
4 s a g
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t r
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c c a i c m. m m
u.
sa i g a n
m
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t a lare t s
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1 r
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htga . . .
ht gi . . h e
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- 4 e
- 0 f t 1 e fr r o) 4 4 7 4 4 4 0 4 e dF t*
0 1
1 0
1 7 0 0 0 % 8 0 h 1 1 1 1 1 t ef c l A a -
- e e r l 0 0 0 0 5 C 0 0 0 0 0 a a 3 5 5 3 6 S 5 5 5 5 5 m / / / / / / / / / / / s r 4 4 4 4 7 4 4 6 0 4 i o) 0 0 0 0 7 0 0 8 M 8 0 h .
a nF 1 1 1 1 1 1 1 1 t ,_
r b*
u A n r i a n e ei t -
gm e n M
n/ e e
- aw 0 0 0 0 5 0 0 0 0 0 0 m R a 5 3 5 5 4 3 5 5 5 5 5 p
? a / / / / / / / / / / / i lf a
4 4 4 4 7 4 4 6 0 4 u m
0 1
0 0 1
0 7 0 0 8 M 8 0 q r
1 1 1 1 1 1 e o r y N ) n o 3 t w 0 n l) o C 0 t e oD t A t t e m n T l e . . n p a . el N u
a es hs ) e) i) m FC lR so 8 F9 uA o b o v E( i e 0 0 q8 o n. a n e ut g m0 n0 E0 R wB C t e e e 0 o e l) ar l aet nm o , t . l r d . l l t e b) a2 et r4
) ie) tt5 S6 0
t7 e0 e .
em a) e0 m) e3 t5 u1 a) c1 e
- m nt C0 ca o0 us0 y0 b0 iR g1 s1 h0 i0 r a es 0 nr d0 t y0 r r(
t e
m ma nm ot rws e .
at re t n rm i
r . e rgm.
S e . rt e.
t e t e i
sm* eo t
cm h0 c
t .
is A0 C .
S .
e t
r1 e . t h
e o r.
ra eR ne em i nR eR i a: l o wR A VR m mmuaem t p e w
t e ie vB F( Er C( Fi( B( D( ER s( N( A( E(
- 9 w
n E
(
a p
y . **" r g ok sW0 "
i g
g
..y w -
p w
M
' , iij!:t li iIi j i} ,!i fll ,t ; ' ;1! } j ii<' :,i4 .i* i,j '. iis
4 i L
1 i
I TASIX 3.11 1 (Continued)
.! . m ine m m u.cw e m e s I L f Relative Cimalttive Radiation W (
j facation Tammeratura Fransare h idtte *--
(Entrorumental Normal Eenge Abnormal Accident
- Normel Emnge Ace 1 dent Wernal Ace 1 dent " Radlettom i'
Deetenator) t-. /= t n . **1 f*r1 hemel Acciden 8 f- wfate ti ft) frada) fenda) Tyr,e Cable Spreading Roen 104/30 104/30 104 ets. ete. 80/20 80 100 1.3 10s ,, ,
(Re. 102) ['
t
- ETAC Reen 104/30 104/50 1M men, atu. 80/20 90 100 300 gemme j (ps, 1045)
Electrical pene. 104/50 1M/50 1% ata. sta. 80/20 00 100 S.la10' ge m
, tretion Aree (mm. 201)
- ?
! a. lay meen s0/50 30/50 so .11shely e0.125* s s0/20 s0 100 100 se===
j (mm. 202) yesitive centrol mesa 7s/72 7s/50 7s olishely +0.125 es s0/20 s0 100 1c0 s-
+
(Rs. 203) yeettive i
! Seemite Eng. Office 78/72 78/50 78 elishelp 80/20 80 100 100 i oo.123*ws. gemme e w (mm. 203s) yesttive H ;
j . "J
'C e11ghtly to , 100 f
l L
Kitches' (En. 205C) 7s/72 7s/50 78 yeettive
+0.125* vs. 80/20 300 gamma Q'
m i O
MVAC Roen 104/30
~
104/50, 104 elightly e0.123* vs. 80/20 90 100 SCO ge m
&r
( -(tm. 206) yeeltive ;
j m ac m.es 104/50 104/50 105 erm. etm. e0/20 e0 100 500 gemme 82
- (mm. 2Ms) I
} Switchgoer Room 86/30 86/50 104 ate, ata. 90/20 00 100 130 gemme
{ (mm. 212)
- Distribution Seen 104/30 104/30 los atu. etm. 80/20 00 10C 130 gamme %
l (ame. 213. 2158)
~
o cn i Settery teen 77/65 77/43 77 ellshtly slightly 73/43 90 100 130 ge m
{2 h; !
j- (mm. 214) negattwo nogettve-3 m %. . i S. i O
h.:
o l
i "$
- i t
u -
1 ( "
. N 4
=
- 1 g , ., -.
.,._..,.w.. _ . ,, ,,.m. , ,,.m, ,g,,yg . . , _ , - ,
f 4
j TABLE 3.11 1 (Continued) l ENVIR0!EffXTAL CQtEITIONS 2- i i Relative j
C-stative Radiation /8 s 1m atfen Tamsara*ura Pressure hat di tv P- - -
t
_ (Environmental Normal Range Abnormal Acelden P 18ernal Range Ace 1 dent Iternet Accident " Radiaties -
, Destenater) f === /= f n . *n (*n Isermat Ac e t A=== ** (man / min t) fa) fradal frada) Tyne 4 {
! Coeruter Roen 72/68 72/68 72 elightly +0.125*eg 60/89 60 100 130 games (Rs. 215) positive 2
- ^
Eteetrical 104/50 104/50 104 atm. ets. 80/20 80 100 7:10'" samme renetratten f ]
! Area (mm. 301) j -
i
, Cable Spreading 104/50 104/50 104 atu. ate. 80/20 SO 100 4108 gamma Room (Rm. 302) [
+
3
, NVAC Reen 104/50 104/50 104 alightly +0.125*ws. 80/20 80 100 500 gamme ;
l (Rs. 307) poettive ,
i corrider 78/72 78/50 78 men. atz. 80/20 80 100 130 gamme !
(Rs. 300) i 104/50 104 Cerrider 104/50 stm. ats. 80/20 80 100 3:108 gamme j
g (mm. 3OSA) y
- e p
o
- switchgear Rose 86/50 86/50 104 sts. stm. 80/20 80 100 100 game !
e
. -(Re. 318) '
i m '
- l. Distributten Reen 104/5d 104/50 104 ats. sta. 80/20 80 100 1.3a10' gamme b I Area (Ra. 319) t 1
feeter Cenerecer 104/50 104/50 104 etm. eta. 80/20 80 100 1.3m10s ,,,,, ;
see Rees (mm. 320) '
Settery seen 77/65 77/65 77 elightly slightly 75/45 80 100 1.3a108
! (Ra. 321) nogettve negative gamma 2- -
l !
- i. Fewer Cabinete 104/30 104/50 104 etm. eta. 80/20 80 100 4a10' gamme -
(mm. 323) ,
4
- Iseutron Flus Amplifier Qwest C Panel Z1.P-696 has en accident daoe of lose than 10' rede (18C9040). --
2 !
1 T Co d 1 4 >-
c3 .-4H--e -
l 3 m2> '
e i FO i
=C 1 !
m '-4 iC ,
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W :
)
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+
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4
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i f [s i ff:F *ft t[ tt![T I [.t! *l[> i (;i j, ,, !i! .si* i lt [ ,. ,
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m m a 8 e 8 e a a a a a -
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'0 1
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=
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- cf e, c
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8 7 0 1
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1 4
0 1 1 4
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cf c
A e l 0 3 2 0 0 9 0 0 0
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/ / / / / -
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tf 3
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7 B
8 5
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3
/
4 2
/
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0 1
0 1
0 1
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- a r W -
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- l u m o t s5 w
- a o a .
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. e e e t r (m a -
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_ ar t a b) a1 r1 R) 3 n) e0 u) a2 Ae r m5 m) m) i m n ne e2 o4 o7 l o e=
C0 4
e0 e0 e1 R4 V1 eu oO o0 o0 p p
m t4 g4 4 dt R5 R5 revm.
pS i m u a e l c a t n == p r . C . l . eu C . C . c.
a of ra ea ma.
oR os A m bs aR t r V
A s A s A m t e e r( tR Va ut VR Vm h ie C( s( f I ( C( O3 N (R f i( f I ( p y vD '
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L M
TAB 12 S.11-1 (Cos'.isesed) r IN71RGGE2tTAL CGQft10lf5 i L 1 Relettve Questetive Radiation P fu etimi T- ---- ra *ure Freesure Numidity *-- --
{
(Environmental Normel Range Abeormat Acciden F Normel Range Aceldoet Normal Acelden M Radiation Deelenater) f===#ein. *F) (*F1 flernal Acefama # famw/mte et foi fe= Jai fredel Tyne HVAC Room 104/30 104/30 104 etm. etm. 80/20 80 100 1.3a10" gemme -
(Res. 308. 309) I Radiation Bootter 104/30' 104/30 104 etm. etm. 80/20 00 100 1.3x10" geesee (Rs. 310)
- 4. Fuel Bandling Buildina HVAC supply IM/65 117/62 120 e11gM2y ellshtly 80/62 100 los 1.2x10" .gemen subeysteen) neset1= nosettw (Rs. 002)
HvaC Roe. 120/63 120/62 120 allshely slightly 10 8 (Rm. 003) nogetive nogettve s0/20 30 100 se=en l2, Eccs Cubietes 104/63 120 8.9 104/62 alightly elightly 80/20 80 10' + da10' gamme (Ras. 004, 003, 006) negative , negative w , ,
if H g Focirculotten Yelve 104/63 104/62 12f slightly elightly 80/20 90 10 8 -9u108 gamme 9
s Roses (Sam. 007 negative nogetive C 008. 009) '
Sprey Additive 104/63 120/62 120 alightly slightly 30/20 100 10s 2.1x10* E Tank Roses nogettre megettve gamme l2 (see. 007A. 000A.
0094)
NVAC Reen 104/65 ellglatly 89 107/62 120 e11gM1y 108 t:fm10' '
80/20 80 gamme
- (Re. 010) nogetive ansetive NTAC Cerben 104/63 120/62 120 " olightly slightly 90/20 100 108 100 gamme Filter Room . segitive negative l2""
(Ra. 106) -
. gw>l -
a mr>d m 70 3.
- '!u ns 0 ,,,
T.
o i U
o@Cfg T
j c
N l .
L j m--
p
=n ,
I TAB 12 3.11 1 (Continued)
ImROIETITAI. CDtf2fTICf'S r 1 L f.meetion Relettre cumeletive Radiationif8 [
Tameere*ure Presente h fditv N-r i (Environneetal Normal Range Almersal Acciden F Normal Range Accident Normal Acciden F Bedtetten De st mater) fame / min *F) f*F) 16amal Ace f dentaso=(ma n / min ti ft) frede) Tyne frada)
Spent Fuel Feel 104/65 120/62 120 alightly elsghtly 40/20 80 2:10s 100 8***e huPe nogettve negative l2 (Ree. 106. 107) storage seen 104/65 119/62 120 etishely slightly 80/20 100 10' 100 gamme (Rs. 108) negative negative New Fuel Storage 104/63 109/62 120 alightly slightly 80/20 100 6 10s 100 gemas (P_e. 203) negative nogetive IfvAC Rees - 104/65- 112/62 120 alightly elightly e0/20 100 10 8 100 gemee (Rs. 204} nogettve nogetive rietform 104/65 112/62 120 alightly e11ghtly 80/20 100 aos 100 88"=a (Rs.-203) negative negative ,
spent N 1 Feel um 104/63 119/62 120 alightly slightly 80/20 100 6:108 100 gesene w
Rees (Re. 206) regative nogetive y *
"U U
- Spent Fuel Peel Ita 104/65 124/62 120 elightly slightly 80/20 100 4:10' 100 ge_e e Rees (Rs. 2C7) negative nogetive V '
g '
Ceek Mend 11og Area 104/65 109/62 120 e11ghtly slightly 80/20 100 los 1.8 10' gamme '
(mm. 303) nogetive negative 7)
- we-Accideet 23/63 90/62 120 alightly e11ghtly 40/20 100 3.5m10'ee+ 1.3310s ,,,,, e sempling Room negative nogettre (En. 305) et.A 1- .,m. ..n.1e ... 1 1 _. e.y e. to e n 1 esv1.e_ent e,10s te.e. ;
gm> ed m ^c >
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! T4BLE 3.11 1 (Contimed) 1 ENVIRfEEtENTA*. CD S ITI!Et3 imention' meistive ComelativeRediettee/*
Tamarature Prea m Isu.edttv * - - -
i (Envis _ - al Normal tange Abnormal . Accident **'" Normel fange Accident Normal 1
w t - e.c) r .r t. *r1 f*r1 1.
AccideetM Radietten 1 Acet h erneur.r. u- tm) trad=1 _trad.) *vne j Migh Activity Spent 104/50 120/4 125 0.3 pelg
- Seeta Storage Tank elightly 80/20 100 2 10' 100 game voegettve 1: (R=. 054)
Roseter nake-up 104/30 104/44 125 e11shely Water fump Cubielee 1.2 pets 90/20 100 2x108 100 se==
' nogettve i (Re. 062) l 3
Refeeling veter Storage Tank Rose 104/50 111/44 130 alightly nogetive 2 4 pelg 40/20 100 Estos 1.3m108 gesme g' i (mm. 063) 1 i Non Radteactive 104/30 133/44 125 elightly 0.3 pelg Fipe Qwee (tm. 064) 30/20 100 108 3. Sales g,,,
negative
- Electrieet Equip. 1%/$0 131/44 135 elightly 1.1 peig 80/20 100 - 100
!~ most Rees (mm. M5) 1.3x108 gasse negative to m
} Essential citiler 115/50 11$/44 120 1.1 petg *4 *
- 4. CO: Pump Reen altshtly SG/20 100 108 130 game *#
negative j *
(Res. 067. 067E. -
c !
- 9 3
~ m7n . . ,
g Cerrider 104/50 126/44 170 1.1 pelg
- elightly 80/20 100 108 100 genes (Rme.-067A. 0675) n*se tl'e !
f corridor 106/30 123/44 195 1.6 peig ellahtly 80/20 100 ' 10' 100 gemme l (Ra. 067C) nemetive
~
f< C m ider 104/30 126/44 170 alightly 1.1 pets 98/20 100 los 1.3nts g,,,,
(mm. 0679) nogettve I j 1 TMM t 2 w -e o=
m I >H t
+
.f
'M 3m
- e
- < EmK
- r m 1
_~ en
.C 2-o g -t a C n? A. s .
p3 2~
Q j
i l=
. _ . . . _ . - , _ _ , . . , - , - . - - . _ ,_ ....m. , ._._,_.,..-__.-._,m, - _ _ _ _ m,. - _ - , _ . , _ _ - , _ _ . - . . . . . _ _ . . _ , - _ - . . . . _ ~ - .
I 4'
)
j' s
TAs12 3.11-1 ( m tinued)
INC.N6L.C3hf**10tr'.
i fu meten Erlative Osamlottwo R 41stie #
Tame rature tree <-se beidfew " ---
(Environnesstel Eternel Range Abnormel Acciden P permal Range Ace 1 dent Normal Deat-a=*ar) f=nerafn *M f *n Bernal Aceteen P Radtetion
, acefa=n F f=..r h 31 fat frede) fraa.) Tese __
valv, Reen 104/50 106/44 125 eltshely 0.3 petg 80/20 100 6 198 100 gamme (Re. 048E) nogetive Chareeel Aboerber $3/30 103/44 123 milth tIf 0.3 Fe15 80/20 100 10' 1.3m1C8 geese Reem (Rm. 06tF) . negative Caerd see Cabiele 85/30 95/14 123 alightly 0.1 peig-
- (Rs. 06988) 80/20 100 10' 100 gemme negerive t
4 Na0 Remeval Skid 104/50 109/44 123 . elightly 0,3 pets 90/20 100 10' e
100 gem Room (Ra. 068Q cogetive i
pump Rose 104/$0 114/44 120 alightly 0.3 pets (Ra. 072) 90/20 10G 6:30' 100 gamme nogetive serie Acid Tonk 104/30 114/44 170 eltaktly 0.3 pets teen (Ra. Sf6) 80/20 100 2x108 100 gen
{ nesetive
! tm valve Reen 104/30 105/44 140 alightly 0.3 pets
! 90/20 100 Su1G1 100 gamme N (Rs. 079) nogettve i $. --
l, ' % -velve Room 104/30 134/44 140 alightly 9.3 peig 80/20 100 10' 100 (Rs. 0794). noget1*e Seeme Velve Rose 104/30 144/44 240 elightly 1.5 peig 80/20 108 10' (mm. 079s) nogetive 100 gamme
')
velve Re.e 104/30 128/44 i'4 elightly 0.3 pelg 40/26 100 10' 100 gamme (mm. 079C) noget1**
cooling water 104/30 113/44 125 e11 # 1y 0.3 patg 8A/20 100 10" 100 geese N.et asehenser v oet1'*
(mm. 106) g Vs >
a mr> v1 x TO uwI k-a
- O zg o
" h, s
. Y .
{
TA312 3.11 1 (Centinued) t 13nr110%2NTAT. CONDITIONS L
Relative C m alatin Radiatie M Imation Temperamre h aamura thanidi ty *-----
n (Environmental Normal Range Abnormal Accide # Normal Range Accident E>rmal Accide # Radtetlea Deat - ree) taastata *n (*n mmt aceta==em tm w/ min. si (t) trede) fr-d.) Tm Non Radlemettve 104/50 119/44 120 alightly 0.3 pets 80/20 100 108 130 gamme tiping (Rm. 1044) negative Non Reeseactive 104/30 112/44 195 elightly 2.4 pets 80. 3 100 10' 100 gamme Flying (Rs. 107) swgettvo i
4 i Redlemetive Pipe 1M/50 131/44 220 stightly 2.4 pelg 90/10 100 10' 1.3 10' gemme j Fen. Area (Re. 104) megative l2 Corrider 104/50 114/46 220 sitehtly 2.4 pets 90/20 100 108 1.3a19'
, (Rm. 10tA) negative gesman g{
] Non Redlesetive Pipe 104/50 117/44 160 alightly 2.4 pets 80/20 100 108 100 gamme
[
[
t Chase (Rs. 1088) megettve f
, Radteactive- 104/50 124/44 250 eltshely 2.6 pets 90/20 100 10' 2.6 10' Pipe Chases gamme i nogetive j '(Rm. 100C) ,
e4 m
' C Radiomettve ripe 104/30 117/44 175 ettehely 2.4 pets 90/20 100 10' 2.6a10' gamme g 4e Chases (Ra. 1000) megative
- i Electrieel Chase 1 % /50 111/44 160 slightly 2.4 peig 80/20 100 los 100 gamme 71 g j- (Ra. 10st) eeseelve '
oryw .t. Campeeter 104/30 111/44 155 elishtir 0.3 Pets e0/20 100 6:108 100 (Rm. 109) so- ,
1j mesettv.
' Redtometive rap. 104/50 121/44 170 ettshely 2.4 Pets a0/20 10e 102 100 se me Chem mesative
} (Res. 110, 1104) !
1
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i m=> ,
70I !
- o -
'J>E
~m o *m i
k SZ Q"
o n
ON C
M 4
i
_.. 4.. . - _ , . .-~,........-..,e . , . . , - -. -~. - --.-- , - - . _ - - - - - . . - ~ , , - - - . - . _. . - ~ . - . . . . ~ . , . . . _ .
x
]
i TA312 3.11-1 (Continued) rwmmmmmr_ conomess 3 .
j
- Beletive Chasialat1== Radietten@
Incarian Tammerature Pressure h=Mf ty h amme (Emironmental Normal Sange Abneteel Accide # permal Sacge Accident Bernal AccidenC"" Ba ttatlee mat. tors r . r t . _ *n tan netw
- aceta- M r .r-tn et te) 3 tenda) tr a.1 trea 1Asa storage Tank & 1M/50 1M/u 120 elightly 0.8 yets 90/20 100 3e10* 100 4
Pump Rose (Ra. 112) asume eegettwo MVAC Rose Y$/72 104/50 120 0.3 peig
( alightly 90/20 100 108 130 ge -
(Rm. 2065) megettwo Non Radiomettwo 104/$0 123/u 123 elightly 0.3 pets 80/20 100 los i
F1pe Penettetten 9.7210' geau nogettwo !
(Re. 215) 1 Radeeste Centrol 78/30 113/4 125 e11ghtly 0.3 yets 90/20 100 108 100 gamme Room (Rm. 217) negative
.Statrosy 78/50*** 104/u 123 elightly 0.3 yets 90/20 100 IJe 100 gamme
} (Re. 217A) aegettve m etrieel Chase 2A.
, 104/50 115/u .?$ elightly 0.3 pelg 80/20 100 10' 100 gamme 4 w ase. (mm. 217s) n*sative ,
. e og O
! ne11.sy 7s/50 112/u 125 e11ghtly 0.3 pets 8 i a0/20 100 10 130 g.mme t2 2
4o (am. 21s) negative -
fi Corrider. 104/50 117/u 123 elightly 0.3 peig 80/20 100 108 lofr see se y'
(mm. 21sc) meset1** ~
4 corrider 104/50 120/u 140 alightly 0.3 pets 80/20 10c 10 8 100 se=se (tm. 2101) meget1**
1, i j sypese Transfer 73/72 84/u 120 alightly 0.3 pelg 80/20 100 100 100 ge m
- Area (2m 219A) esgative
- t. Cycs valve Cnhiele 104/30 1M/u 125 elightly 0.3 yeig 30/20 100 10' 100 geese 1, ,
.(mm. 225) eegative j
eee mre te no stAc or egelpment in this eree; hrerere, W tisperatures listed are beoed spee N edjeeeet ruem teureratures. . ly _f w ;gms mv~
I * >
j mTh..
ra
$2 2
7 2
':' M
]
4 m ! Ow 4E D .N (l -
~ .
L -
1 N ,
j j
._ _ _ _ _ , . . . _ _ . _ . . . . - ~ _ ~ . . _ . . . . -._ m_ .m.m. - - - .m_...
1
[ :.
?.
" I E:
i j
TAB 12 3.11 1 (Contineed) '
, l mmacnunt. cmemmis u
j bene f c.e Relative h 1stive Radiation /8 l T-enhare Preneure E 'Ater *--- -
- j. (Enetreremental Bernal Range Almemel Acetdee F Normel Range Accidmet permal
[
Deaf ==cer) (-= /mi n *n Aeelde # Radletten !
l tan h ramt acetA- M tm.=/mtm. at tal fr*Aa1 fr=Aa1 Tyne 4
j Counting Room 79/64 120/44 125 eltshely 0.3 pets h 90/20 100 los 100 ge-o I (Rm. 231) posttive i
j ' velame control Tank 104/50 116/44 120 0.3 pelg alightly 80/20 100 10' 100 ge-e
$; (mm. 233) nogettve "
l
- . Red. Chem. 1mb and 75/72 113/44 125 elightly 0.3 pets 80/20 100 6a108 t Sample Room 100 game negative j-(Ba. 234) ,
I' semple Room 75/72 113/44 125 eltshely 3 3 pois_
. 30/20 100 3:10* 100 geese I
'(mm. 235) negative i .
j Soren Analyrer 75/72 97/44 120 alightly j
0.3 reig 80/20 100 10' 100 gems (Re. 2354) nogettve 3- ,
l: valve Area 104/50 116/44 120 e11ghtly 0.3 pets 4
.w .(Rs. 234) 80/20 100 6 108 100 gamme m eegative
+4 :
- w >
! velve Reen 104/30 116/44 120 alightly 0.3 petg 40/20 100 10' 100 geama
- 4 (Re. 138K) negative , -
12 5
>= wE j
}' - valve Aree 104/30 116/44 120 alightly 0.3 pets 80/20 100 10' 100 %
gemme 2
a (mm. 23eR) mesettve y 4'
r11ter Area 104/30 111/44 120 s11ghtly 0.3 reig s 80/20 100 Sale 100 ge-a l
} (Rm. 243) megative a ,
i 2 .,
~ Filter Aree 194/30 112/44 140 alightly 0.3 pelg 80/20 100 2x108 100 gemme
} (ame. 243A, 243L) segetive 3 .
L ;
i th b i
6]rn m
14 zw i
)
(- 1
= %"E t i
- O
.C z i O -*
OQ
. n m
h t l Q -
3 0
l1 .
}
. ~
4
+ A. 4 . _ ... ~. , _ ,- _ _ . , ~ _ ~ - __ _ . _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ . _ . _ _ _ _ _ . _ ._ . _ _ . . . . .
TA312 3.11-1 (Continued) 1.a v i nc- ,_mx!_ ctsriTicus Reistive CumulativeRadiatten[
locathm Tameerature h eggure Ibefdity hamme (Environmental Normel Range Aeneraal Acetden f 18ernal Raege Accident Normal ACCiden F Radiatler ha t enater' femm/ min. *n f*n Nermal Accident" f asut/mf m ti it) frada) fradm L J e Mimed sel 104/S0 116/44 123 elightly 0.3 pets 80/20 100 2x10* 100 gemme Dominera11rer nogettw {2 s
Cubieles (Rze 244A-T)
RfAC Equipment Reen 104/S0 94/44 120 ettghtly 2.4 peig 20/20 100 lo s 100 games (Rs. 324) stegative corrider 104/50 121/h 125 elightly 2.4 pets 90/20 100 lo s g3, ,,,,,
(Ra. 324A) negative CCnf Serge Tank 304/50 114/44 125 altshely 2.4 pett 80/20 100 10' 130 Room (Rm. 3245) negative gamma 2 Decentaminatten 81/50 107/44 123 sitghtly 0.5 pelg 80/20 100 2x10s 130 gewa (Rae. 323, 325A) negettwo
. w Fersennel Ratch
~
g Area (Ra. 326) 104/50 112/44 123 elightly negative 0.3 pets 80/20 100 108 1.7x10a g,,,, $
N C
4 N
valve Operating Area 104/30 118/44 120 alightly 0.3 peig 80/20 100 - 10' 1.*n10s g,,,, g (Rm. 327) segative >
IP!AC Been 104/$0 104/44 125 elightly 1.4 pelg 80/20 100 los 1.3a10s g,,,, I2 ,
(mm. 3276) n*5etive l '
4 Nydroges & Radletten ' 104/30 104/44 125 eltshely 0.3 peig 80/20 100 10 8 1.7 108 gamme Monitors (Rm. 33)* negative
- Hydrogen eral radiatten sentters are shielded and espeeed to se occident enwirermeet of 2.33108 toda. g en yw -
>i I g r.1 T d' ro '
}r$~E et E"
?
NQ 1
c _,
C3 e e
h o Thb a !
~
g -
C jQ m
- J 4 l L
1
... I r
- l
! I t-j' TAD 12 3.11-1 (Contimmed)
{
. t j ENTIEM MITISHS f 2 t Selative QaealetiveReetette[8 i tu metan Tasmearenar. Freemeare h 'Atew 4 (Environneetal Normal Range Abneraal AccidenF Normal Eme=e Acetemet Bernal Aceldent " Sedtetion
] Seat m eer) f===/=t= *M f *M Nor=af Aer' h # f=== M = ti f*) frade) f rm&)
f
! Yve. J ceneentrate storage 106/50 116/S4 140 e11shely 0.3 peig 80/20 100 10' 100 gamme i Tank (93. 331) negattw
- corrider . 104/50 109/64 140 elightly 0.3 yeig 80/20 100 Gr108 100 !
j- (Pm. 331A) negative gemene
[
Redweste Mining 1M/50 111/64 125 alightly 0.3 peig 80/20 100 10' 100 gamme Tank (Rm. 334) nogettve t i Teltmo control Tank
' 104/50 107/44 125 alightly 0.3 peig 80/20 100 5:10s 130 gemme [
operating Esen megative l' (Ra. 335)
} 6. Diesel-Cenerater Building t
r
- Engine Room, operat= .106/50 120/42 120 alightly alightly 100 100 100 100 gammme w Ing (Dee. 001. 002, poettive peettive e.q i
- g- 003) *e
. H C'
O u
stairwell (ame. 006, 104/50 120/62 120 sea. sta. 100/20 100
- 100 1e0 games 3 [
005. 006) g l
} Intake Filter Seems . 104/50 120/42 120 elf g%tly elightly 100/2C 100 100 100 ge m p' I j' (Res. 101, 102, 103) pee 7tive peettive t Diesel 011 Storage ~ 104/50 113/29 120 alightly elightly 100/20 100 100 100 m
- Room (ame. 107,'108, megettve nogettve ,
109) stetreelt (ame. 110, 106/30 10^/29 120 alightly slightly 100/20 300 100 100 gamme k 111. 112): nogettve negattie ~~
i j -=
rm-
. 7* c) -
Mhh.
r 4
4
- o 'd$X t
i.
e
)mI
- m ez
' " b 8
O 0 Oob T
l W i 4
4 r 1
f j i t
4
_i i
i
- i. .
I- t l
i .
1 TAS12 3.11 1 (Continued)
. i t
- IINIRCMEMNTAL CONDIT!(Mf1 p J
Selectwo OssulettwoRadiation[8 j h tien T rahmen Pra __-- h='Atew *- -
a :
(Environmental Normel Range. Abnormel Acc ident*" Wormel Range Aceteent Formel Acciden F Radietten .
, Deaf.n. car) f=== rate. *r1 f*r1 hr==1 Acefd==e""f===/=fa a) ft) frad=1 fraam) Tone i
Intake Air 104/53 104/29 120 alightly slightly I2 100/20 100 100 100 geene 1 (Roo. 207, 208, 209) mgetive negative I i
<- 1
{ '7. h rbirne."meme=eme i Bu11dina !
i
]' .Generet Arose 110/50 147/43 54 eligfitly EA 80/20 .M 100 1% 5*m**
2 positive
. S. Klocallamoeine 4
au11dinge
- . Ecsts 104/34 . 104/34 104/34 elightly s11ghtly 80/20 80 100 100 j (Fump will be starting negati e wgative ge -
l21 4
.at 50 *F)
A-N.
. pa - ad I'
Y .
N M D
.i. '
o
't _
th > '
, 7 NA Not applicable m3 i ro p
Nb3
~PE t
w i a op 4 4
f al
? d'.
i M
.% I
- . m i
Y
- 4 t'
j ~
b' 1
~
, ., ....m_ . . . . _ . * ~ , . ...,.,n.._,._,... .
...,nw.,y,,-..,,._r_,:_ _ __..._,r_.,-.... . __
. . . , ~ . , - . ., , , , . , . . , . - . . , - _ . . , _ , . . . . - - . ,, ._y, ,. ,, , ,___,-,
r- ATTACHMENT >
STFECS UFSAR S?*HL AE t/of. (,
u M h 0F J Z 5.4.4.2 Desien Descrietion. The flow restrictor consists of seven inconel venturi inserts which are inserted into the holes in an integral steam outlet low alloy steel forging. The inserts are arranged with one venturi at the centerline of the outlet nozzle and the other six equally spaced around it. After insertion into the low alloy steel forging holes, the Inconel venturi nozzles ore welded to the Inconel cladding on the inner surface of the forging (Figure 5.4 5). .
5.4.4.3 Desien tvaluation. he flow restrictor design has been sufficiently analyzed to assure its structural adequacy. The equivalent throat diameter of the 50 outlet is 16 inches, and the resultant pressure drop through the restrictor at 100 percent steam f1 is approximately 3.4 psig.
This is based on a design flow rate of 4.24 m Ib/hr. Materials of construction and manufacturing of the flow estr ctor are in accordance with Section III of the ASME Code. 6 A Su persed p + t he 4 6" 10 5.4.4.4 Tests and Insceetions. Since the restrictor is not a part of the steam system boundary, no tests and inspections beyond those during fabrication are anticipated.
5.4.5 Main Steam Line Isolation System Refer to Section 10.3.2.
5.4.6 Reactor Core Isolation Cooling System Not applicable to STPECS.
5.4.7 Residual Heat Removal System The RHRS transfers heat from the RCS to the Component Cooling Vater System (CCVS) to reduce the temperature of the reactor coolant to the cold shutdown temperature at a controlled rate during the second part of normal plant cooldown and maintains this temperature until the plant is started up again.
Parts of the RHRC also serve as parts of the Safety Injection System (SIS) during the injection and recirculation phases of a IDCA (Section 6.3).
The RHRS also is used to transfer refueling water from the refueling cavity to the refueling water storage tank (RUST) after the refueling operations are completed.
5.4.7.1 besten Basen. RHRS design parameters are listed in ..
Table 5.4 7.
The RHRS is placed in operation approximately four hours after reactor shut-down when the temperature and pressure of the RCS are approximately 350'F and 350 psig, respectively. ' Assuming that three 10ts and three pumps are in service and that each lut is supplied with CCU at design flow and temperature, the RHRS is designed to reduce the temperature of the reactor coolant from 350'F to 150'F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The time required under these conditions to reduce reactor coolant temperature from 350*F to 212'F is 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. The heat load handled by the RHRS during the cooldown transient includes residual and decay heat .' rom the core and heat from a single operating RCP. The design 5.4 22 Revision 0
iiTACHMENT',(
sTPECS UFSAR ST.HL.AE 4% b PAGE??'t 0F lbf7.
TABLE OF* CONTENTS CHAITER 6 ENCINEERED SAFETY FEATURES lection Title Eggt 6.1 ENCINEERED SAFE 1 ' FEATURES MATERIALS 6.1 1 6.1.1 Metallic Materiais 6.1 1 6.1.1.1 Materials Salection and Fabrication 6.1 1 6.1.1.2 cvaposition Compatibility, and Stability of containment spray Coolants 6.1 2 6.1.2 Organic Materials 6.1 3 6.1.2.1 Protective coatings 6.1 3 6.1.2.2 Cable Insulation 6.1 4 6.1.2.3 Oils and Creases 6.1 4 6.1.2.4 Decomposition Products 6.1 5 6.1.3 Post Accident Chemistry 6.1 5 6.2 CONTAINMENT SYSTEMS 6.2 1 <
6.2.1 Containment Functional Design 6.2 1 6.2.1.1 Containment Structure 6.2 1 6.2.1.1.1 Design Bases 6.2 1 6.2.1.1 1.1 Postulated Accident Conditions 6.2 1 6.2.1.1 1.2 Mass and Energy Release 6.2 2 6.2.1.a.1.3 Effects of EST Systems on Energy Removal 6.2 2 6.2.1.1.1.4 Effects of EST Systems on Pressure Reduction 6.2 3 6.2.1.1.1.5 Containment Leakage Rate Bases 6.2 3 6.2.1.1.1.6 Bases for Analysis of MinLmum containment Pressure 6.2 3 6.2.1.1.2 Design Features 6.2 3 6.2.1.1.2.1 Protection from the Dynamic Effects of Postulated Accidents 6.2 3 6.2.1.1.2.2 Codes and Standards 6.2 3
- 6.2.1.1.2.3 Protection Against External Prest to Loads 6.2 3 l 6.2.1.1.2.4 Potential Water Tapw Inside the L ntainment 6.2-3 6.2.1.1.2.5 Containment Cooling and yeatilation_ Systems 6.2 4 6.2.1.1.3 Design Evaluation 1 kd TeWerature. 6.2 4 i 6.2.1.1.3.1 Containment Paak Prescur LAnalysis A 6.2 4 l 6.2.1.1.3.2 Long Term containment Perforsance 6.2 7 I
6.2.1.1.3.3 Accident Chronology 6.2 8 6.2.1.1.3.4 Energy Balance ., 6.2 8 6.2.1.1.3.5 Functional Capability of Containment Normal Ventilation Systema 6.2 8 6.2.1.1.3.6 Protection Against severe External Loading 6.2 8 l 6.2.1.1.3.7 Post Accident Containment Monitoring 6.2-8 6.2.1.1.3.8 Equipment Qualificatiott 6.2 8 6.2.1.2 Coctainment Subcompartments 6.2 9 6.2.1.2.1 Design Bases 6.2 9 6.2.1.2.2 Design Features 6.2 9 6.2.1.2.2.1 Reactor Cavity 6.2 9 6.2.1.2.2.2 Steam Generator Compartments 6.2 9 6.2.1.2.2.3 Pressurizer Compartment 6.2 10 Tc 6-1 Revision 0 1
1 y e-- r.- + - -- --. -
STPECS UTSAR ATTACHMENT 2.
usT or TAntts ST HL AE-1 PAGE 7t/yo4GOF147 CHAPTER 6 T ble Title f.s.Lg 6.1-1 Engin6ered Safety Features Materials 6.1-6 6.1 2 Not Used 6.1 9 i
6.1 3 Protective Coatings on Vestinghouse Supplied Equipment Inside co mainment 6.1 10 6.1-4 Coating Schedule for Surfaces Inside Containment (Excluding NSSS Scope of Supply) 6.1 11 6.2.1.1 1 Containment Design Accidents 6.2-60 6.2.1.1 2 DBA Calculated Pressures for Containment b.2 62 6.2.1,1 3 containment Data 6.2 63 6.2.1.1 4 Summary of Calculated Containment Pressure and Temperatures (IDCA) 6.2 64 6.2.1.1 5 Engineerad Safety 'estures System Information 6.2 65 6.2.1.1 6 Listing .f Passive (Structural) Heat Sinks 6.2 68 5.2.1. b 7 Modeling of Structural Heat sinks for 1DCA MSI.B Analyses 6.2-72 6.2.1.1 8 Thernophysical Properties Structural Heat Sinks for IDCA and MSLB Analysis 6.2-74 6.2.1.1-9 Surface Heat Transfer coefficients 6.2-75 6.2.1.1-10 Accident Chronology 6.2 76 6.2.1.1 11 Comparative Results: Summary er Results of Containment Pressure and Temperature Analysis for the Spectrum of Postulated 14CA Accidents 6.2 79 6.2.1.1 12 containment Mass strl Energy Balance DEPSC Break with Maximum safety Injection 6.2-80 6.2.1.1-13 containment Energy Distribution 6.2-84 6.2.1.1 14 Summary of the Calculated Containment Pressure and Temperature for MSLB 6.2 85 6.2.1.1 15 >.ccident Chronology 6.2-86 6.2.1.2-1 Short-Term Mass and Energy Release Rates for Subcompartment Analyses 6.2-87 6.2.1.2-2 Not Used 6.2-135 6.2.1.2 3 Not Used 6.2 135 6.2.1.2-4 Not Used 6.2-135 6.2.1.2 5 Stear; Cenerator Loop Con.partment '
Pressure Temperature Analysis 6.2 136 6.2.1.2 6 Steam Generator Loop Compartment Pressure - '
Temperature Analysis - Junction Properties 6.2-138 6.2.1.2 7 Steam Generator Loop Compartment Analysis Force Coefficient (Area Projections) for Steam Generator 6.2-141 6.2.1.2 8 Steam Co terator Loop Compartment t.nalysis Force Coefficients (Area Projection) on Recctor Coolant Pump 6.2 142 6.2.1.2 9 Pressurizar Subcompartoont Analysis Spray Line Break -
6.2-143 6.2.1.2-10 7 12 bewmparrenc Junecien-Description 6.2-144 TC 6-8 Revision 1 i
'l i
STFECS UFSAR ATTACHMENT L ST-HL AE-Lo(c(c PAGE _3 r; 0F l%/ ?
_ LIST OF FIGURES (Continued)
CHAPTER 6 Figure Reference Number Title Number ,
6.2.1.1 17 LOCA-2 Condensing Heat Transfer COEF 6.2.1.1-18 Reactor Decay Heat 6.2.1.1 19 Fan Cooler Heat Removal Rate (Btu /hr) 6.2.1.1 20 Fan Cooler Heat Removal Rate (Btu /hr) 6.2.1.1-21 RHR Heat Exchanger Heat Removal Rate (Btu /hr) 6.2.1.1-22 RRR H6at Exchanger Heat Removal Rate (Btu /hr) .
6.2.1.1 23 steet Sink Heat Removal Rate (Btu /hr) 6.2.1.1 24 Heat Sink Heat Removal Rate (Bru/hr) 6.2.1.1 25 Containment Pressure 6.2.1.1 26 ,
gt h p Tempe qcure Nof U5ely 6.2.1.1-27 Containment Temperature 6.2.1.1 28 -
Condensing Heat Transfer Coefficient 6.2.1.2 1 Not Used 6.2.1.2-2 Not Used 6.2.1.2-3 Steam Cenerator Loop Compa_tment Noda Arrangement El. (-)11'33" - (Sheet 1 of 8) 6C18 9-N 05001 2*
El. (-)2'-0" - (Sheet 2 of o) 6C18 9 N 03002
- El.19'O " - (Shent 3 of B) 6C18 9-N 05003-2* ..
- El. (-)37'3" - (Sheet 4 of 8)- 6C18 9 N-05004 2*
El. 52'0" - (Sheet 5 of 8) 6C18 9-N 05005-1*
El. 68'0" - (Sheet 6 of 8) 6C18-9-N 05006-2*,
Section AA - (Sheet 7 of 8) 6C18-9-N-05007-3*
Section BB - (Sheet 8 of 8) 6C18 9-N-05008-2*
t 6.2.1.2-4 Pressurizer Compartment Nodes Between:
, El. 69' 9" & 85'-0* - (Sheet 1 of 3) 6C18 9 N-05006 4*
El. 46' 7" 6 69' 9" - (Sheet 2 of 3) 6C18 9-N 05005-2*
(Sheet 3 of 3) 6C18 9-N 05008 4* i
- These are reference drawings only and are not to be revised.
l TC 6-14 Revision 0
STPEc5 UFSAR ATTACHMENT L
. ST HL AE-t431.
PAGE % ' 0F _ /W7 LIST OF FICURES-(Continued)
CHAPTER 6 Figure Refarence j Number 1111g Number i 6.2.2 13. HVAC Reactor containment Building Plan E1. (-)11' 3* Area 13 .
.5V14 9 V 00083 12 6.2.2 14 HVAC Reactor Containment Building I Enlarged Plant and Section RCFC 5V14 9 V 00090 3 ,
i 6.2.4-1 Containment Penetrations f (3heets 1 - 100) ,
6.2.4 2 Arrangement of Sump Piping f fj 6.2.5 1 Schematic Electric Recombiner System .;
i 6.2.5-2 Electric Hydrogen Recombiner 6.2.5 3 -,
Hydrogen u m _ ..---, V/O vs.(Time After 1DCAi a J-:P Recombiners) 6.2.5 4 Hg ...-n vfv v.. IL., ah.f IDCA
-( D ye ) - (141.w 'w i n d . gg gp 6.' 55 Integrated Hydrogen Release Following DBA LOCA
~
6.2.5 6 -Puc.-Ac4'nc.; ( tC.C, contxttwalft-
^ =ra**t" : Treneient- g gg
, 6.2.6-1 P&ID Containment lank Rate Test j (CIAT) Pressurization System SC56-0 F 05058 6 -t 6.3 1 P&ID Safety Injection System ;
Unit 1.(Sheet 1 of 2) 5N12-9-F 05013#1-11 ;
Unit 2 (Sheet 2 of 2)- $N12-9 F-0501382 j 6.3-2 P&ID Safety Injection Systen
Unit 1 (Sheet 1 of 2)- 5N12 9 F 05014#1-11 Unit 2. (Sheet 2 of *) 5N12-9 F 05014#2 10
}
6.3-3 P&ID Safety Injection Systen Unit 1 (Sheet 1 of 2) $N12 9 F 05015#1-12 Unit 2 '(Sheet 2 of 2) $N12 9-F 05015#2-11 1
6.3-4 P&ID Safety Injection System [
Unit 1 (Sheet 1 of 2) 5N12-9 F-05016*1 8 Unit 2 (Sheet 2 of 2) 5N12 9 F-05016*2-9 r
't TC 6 18 Revision.1 ,
b
-t l
\ .
- ~ _ - . . . . . - . ~ , . . _ - . _ . _ ~ . _ . - . . - . . - . . , . . _ . . . . , , - - - . . - . - . - _ , - . . . . . . . _ - . ~ -
1 .
- ~ STPECS UTSAR _
- ATTACHMENT 2_.
6.2 CoNTADMENT SYSTDiS ST*HL-AE 40(o(s PAGE 3 7 0F197 6.2.1 Cont (inment Functional Design 6.2.1.1 Containment structure.
6.2.1.1.1 Desien Basag: The Containment design basis is to limit the release of radioactive materials, subsequent to postulated accidents, such that resulting calculated offsite doses are less than the guideline values of 10CTR100. In ordar to meet this requirerent, a design (maximum) Containment leakage rate has been defined in conjunction with performance requirements placed on other Engineered Safety Feature (EST) systems.
The capability of the Containment structure to maintain leak-tight integrity and to provide a predictable environment for operation 'of ESF systems is ensured by a comprehensive design, analysis, and testing propeam that includes consideration of: '
- l. Peak Containment pressure and temperature associated with the most severe postulated accident coincident with the Safe Shutdown Earthquake (SSE).
- 2. Maximum external pressure to which the Containment may be subjected as a result of inadvertent Containment systems' operations that potentially reduce Containment interr.al pressure below outside atraospheric pressure.
6.2.1.1.1.1 Postulated Accident Conditions - The spectrum of postulated accidents considered in determining Containment design peak pressure (and temperature), subcompartment peak preuure, and external pressure are summarized in Table 6.2.1.1-1. The spectrur: of breaks used in the Emergency Core Cooling System (ECCS) analysis for minimum Containment backpressure is defined in Section 6.2.1.5. For postulated subcompartment pipe break accidents, a discussion of break locations is given in Section 3.6.2. As discussed in Reference 3.6-14 and Section 3.6.2.'l.1.1,- item a, reactor coolant loop (RCL) ruptures and the associated dynamic effects are not included in the design bases. Subcompartment analyses are ba' sed on RCL branch pipe. breaks or secondary system pip. breaka. Containment pressure and temperature design is based on nonnechanistic double-ended guillotine breaks.
For Containment structure and subcompartment peak pressure analysis, it is assumed that each accident can occur concurrent with a loss of offsite power (LOOP) and the most limiting single active failure. No two accidents are assumed to occur simultaneously or consecutively.
For each of the categories of Containment peak pressure, subcompartment peak pressure, Containmant external pressure, and Containment minimum pressure, the Design Basis Accident (DBA) is defined as the most severe of the spectrum of accidents postulated for_each case.J e\DBg. al ulagedyreysu(es( marg 4rg t 105ER.T*
eq'JegngalulagdydQet(gnyrecsge a N_e es d asis for the\matgik fot\ A '
. n ;aihmeht a e nverh in\Tahle 4.2hl.k2. Containment design paramecers are given in Table'6.2.1.1-3.
The DBA calculated pressures and margins between ;
calculated and design pressure _ values for various subcompartment analyses are presented in Tables 6.2.1.2-5, 6.2.1.2-9, 6.2.1.2-11, 6.2.1.2-13, 6.2.1.2-15, ,
6.2.1.24 17 and 6.2.1.2-19.
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ATTACHMENT 2_
ST HL AE 404(c PA6E 6. .Z- I PAGE 32__0F f 4 7 -
Ih! SERT A_- . _ _ . . , ,- __
- THE t1 AXIMLN CONTAINMEMT PemK P/tESSUAE DBA DeSCRiP71oNj MHMENT.DE5l6N PRESSURE CALCOLATED 3 PEAM PRESSURE
~
AND t4M%lN SETWEEM THC CAL 40 LATED PEAM AWD DastsN PRE 65vRE_ VALUES ARE GiNEN IN .*TASt E " 6'.2. l. I - 2 .
PAGE 6.Z-9 Ik! SERT B E8CH QOADRANT- coNTAINtNh A ' SC, HAS A VsHT PATH To THE' CC HTW)HMENT AT! THE top OF THE S Cq CoMPARTnENr.
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ATTACHMENT 2 ;
STPECS UTSAR ST hl-AE yo g u PAGE M -OF g 6.2.1.1.2.5 containuent cooline and ventilation systems - During normal reactor operation, the Containment atriosphere is maintained below 120*F by continuous operation of the RCFC System. This system is des.:ribed in detail in Section 6.2.2.2.
6.2.1.1.3 hpien Evaluation:
Ad W Teryera tu(O 6.2.1.1.3.1 Containnent Peak Press eSAnalysis#"TW~tliiF event of a postulated IDCA, main steam line break (MSQjorn feedwater line break, the release of coolant from the rupture area vill cause the high temperature, high pressure fluid to flash to stears. This release of mass and energy raises the temperature and pressure of the Containment atmosphere. The severity of the resulting temperature and pressure peaks developed depends upon the nature, location, and size of the postulated rupture.
In order to establish the controlling rupture for Containment design, a spectrum of primary and secondary breaks, described in Table 6.2.1.1-1, was analyzed to deteruine sech break's significance in selecting a containment design basis.
A Add T A'5 ERT~ 6-I Loss of Coolant Accident - Containment Transient The initial conditions within the Containment prior te accident initiation are given in Table 6.2.1.1-3. The Containment is assumed to be at ambient pressure, design maximum inside and maximum outside temperature (consistent with the data of Section 2.3) to minimize heat transfer during the postulated accident.
For the purpose of the containment peak pressure and temperature analysis, the Safety Injection System (SIS) and the CHRSs (i.e. , CSS and RCFC System) were assumed to operate in the mode that maximizes the containment peak pressure, as shown in Table 6.2.1.1-5. The South Texas Project Electric Cenerating Station (STPECS) analysis indicater that for a DEPSC break with frothing, maximum SI flow results in a slightly higher peak Containment pressure than minimum SI flow.
For the CHRSs, minimum system capacity is conservative for calculating Containment peak pressures. Therefore,.the CHRSs were assumed to be affected by the most restrictive single active failure, which has been determined to be the loss of one DC train coupled with one RCFC unit already out for maintenance. The results of these analyses indicate that a sustained loss of one safety related electrical distribution train (i.e., DG) vill minimize ESF response and maximize accident Containment pressures.
The Containment heat sink data used in the IDCA and liSI.B accident analyses, except the minimum containment backpressure analysis, are described in Tables 6.2.1.1-6 through 6.2.1.1 8. Table 6.2.1.1-6 is a detailed list of the geometry of each heat sink. Table 6.2.1.1-7 describes the resulting simplified heat sink models used Nr computer input. Node spacing used for concrete, steel, and steel-lined c.oncrete heat sinks is fine enough to ensure an accurate representation of the thermal gradient in each slab. A 0.0042 in.
air gap is assumed to exist between the Containment steel liner and concrete vall for peak pressure calculations. It is further assumed that heat is transferred only by conduction across the air gap. ,
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UFSAR Section- 6.2.1.1.3.1, Page 6.2-4 ST-HL AE t{0 b PAGE U N 0F I4 '7 INSERT 6-1 The analysis for 10.48 ft8 DEPSG with Maximum SI and Minimum CHRS (IDCA-2 } was performed with a containment free volume of
- 3. 2 0 X 10' f t3 . All the other LOCA analyses were performed with a containment volume of 3.56 X 10' f t3 . The calculated minimum containment volume (including uncertainties) is 3. 38 X 10' f t 3. "
The analyses for (1) 1.4 ft8 DER, 102% Power, Minimum CifRS MSLB and (2) 1.4 ft8 DER, Hot Shutdown, Minimum CHRS MSLB were performed with a containment free volume of 3.20 X 10' ft . 3 All the other MSLB analyses were performed with a containment volume of 3.56 y lo' f t 3. The calculated minimum +
containment volume (including uncertainties) is 3.38 X 10 6ft .l "
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building internals. The geometry of the structure may be a slab, cylindrical I or spherical. Only the slab geometry was used for this evaluation considering the large surface to thickness ratios in the design and conssivative area approximations. The heat transfer rate at a boundary is equal to a heat transfer' coefficient times the difference between the surface temperature and the bulk temperature. The heat transfer coefficient can be selected to be a variety of functions, such as a constant value, a function of time or temperature, or the Tagani and Uchida correlation. The boundary temperature may be the Containment vapor temperature (or saturation temperature at partial steam pressure for super-heat conditions), the containuent liquid temperature, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cyclic outside temperature, or a constant.
The CSS is explicicly represented in the model. The spray water is taken from an external source (the refueling water storage tank [ WST)) during the injection phase or from the liquid region of the Containment (recirculation phass) and added directly to the Containment vapor space. For two spray trains operating, Containment spray is assumed to be initiated at 82.6 seconds after a IDCA concurrent with a 140P. The delay times account for DC start, sequencing, instrument delay, and system fill.
Pipe break locations, break areas, peak pressures and temperatures, times of peak pressure, and total energy released up to the end of blowdown to the Containment are summarized in Table 6.2.1.14 for each 14CA analyzed. Based upon the results presented in this table, the DEPSG break with maximum SI and minimu:n ContainmentTea emoval was ident as the Containment DBA with a peak pressure o -3h6- psig, which is abou 44 purcent below the design p essure of 56.5 sig, i Figures 6.2.1.1 16 an
.1-17 provide ots of the Containment condensing heat ersnsfer coefficiens as a function of time for the DEPSC with minimum SI and DEPSC with maximum SI cases, respectively.
Main steam Line Break - Containment Transta n Calculation of Containment pressure and temperature transients is accomplished by use of the computer code COPATTA (Ref. 6.2.1.1 1). COPATTA has been previously described in this section. For the KSt.B analysis, the discharge mass and energy flow rates are discussed in Section 6.2.1.4 O.g The calculated s Containment pressure and temperature results are given in Table
% 6.2.1. -14 for the most severe MSLB cases. .The highest calculated temperature is4
- occurring for a double-ended rupture where no entrainment has been in d in the mass and energy releases (section 6.2.1.4). This is conservative since the expected Containment response to a blowdown, including entrainment effects, would result in much lower temperatures than this, at or near the saturated conditions. Figures 6.2.1.1-25 and 6.2.1.1-27 give plota
-of- the pressure and kemperature for the most limiting temperature case. An accident chronology Yor the design basis MSI.B (1.4 f t 2 DER MSI.B) is shown in Table 6.2.1.1-15. <
A significant heat removal source is structural heat sink. Provision is made l 1 in the Containment pressure transient analysis for heat transfer through, and heat storage in, both interior and exterior valls. Every wall is divided into a large number of nodes. For each node, a conservation of energy 6.2-6 Revision 1 I
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ATTACHMENT '2 STFECS UFSAR ST HL-AE-yoG G PAGE _142,__0F /u 7 equation expressed in finite difference form accounts for transient conduction into and out of the node and temperature rise of the node. Table 6.2.1.1 7 is a summary of the Containment structural heat sinks used in the analysis.
Thermophysical properties of these heat sinks are listed in Table 6.2.1.18.
'[hese are the seme heat sink data as those used in the LOCA analysis. Surface heat transfer coefficient for IDCA and HSLB accidents used in the simulation j are given in Table 6.2.1.1 9.
The condensing heat transfer coefficient to the Containment structure is calculated by COPATTA. The condensing heat transfer correlation used in the MSLB analysis is the Uchida correlation (described in Reference 6.2.1.1-4).
For saturated or superheated conditions in the Containment atmosphere, the CCPATIA code uses the temperature difference between the vapor region saturation temperature and the heat sink surface temperature for the condensing heat transfer driving potential. Should the heat sink surface torporature exceed the vapor region saturation temperature, the driving potential used in the calculations is the d!fference between the vapor region temperature and the heat sink surface temperature. Since no condensation can occur under these conditions, the heat transfer coefficient used by COPATTA is 2 Bru/f t 8 hr.'F which corresponds to 's convective heat transfer mechanism.
6.2.1.1.3.2 bne Term contalment Performanes, ~ The long term results of the most severe cases for the primary and secondary side breaks were evaluated to verify the ability of the CHRS to maintain the Containment below the design conditions. These evaluations were Lased upon conservatively assumed performance of the ESFs. The CHRS operating mode was assumed to be minimal based on one DG failed and one RCFC unit out for maintenance. Thus, only two of three CHRS trains minus one RCFC. unit were functional. For the most severe DEPSG case, realistic hot leg recirculation was assumed at 13- -
hours. For all other LOCAs, the more conservative evaluation of cold leg l injection and operation of the CSS to 10 seconds 8
was assun.ed. The heat generation rate due to decay after reactor shutdown is provided in Figure
! 6.2.1.1 18.
The Containment pressure / time responses for the pump suction leg cases cut to 10' seconds (11.6 days) are shown on Figures 6.2.1.1-4 and 6.2.1.1-5. The Containment pressure / time response for the hot leg, cold leg, and smaller pump l suction leg IDCAs to 1,000 seconds and most severe MSLB out to 2,000 seconds l are shown on Figures 6.2.1.1 6 through 6.2.1.1-9 and 6.2.1.1 25. The DBA y 2 8 3 0 l (DEPSG break) long term analysis shews that by(432 seconds, the Containment pressure is reduced below 50 percent of the peak calculated Containment pressure. The maximum pressure of psig occurn at 82.6 seconds in the DBA. The Containment atmosphere nd sump watar temperature versus time are ..
l also shown on Figures 6.2.1.1-10 through 6.2.1.1-15 for the LEPSC and double-I ended hot and cold leg accidents. 40 5 l
l The energy removal rates :for the Containment i'an coolers, the RHR heat exchangers, and the Containment passive heat sinks for the DEPSC break with minimum SI and DEPSG break with maximum SI as a function of time are shown in Figures 6.2.1.1-19 through 6.2.1.1-24 The energy distrib2tions in the .
Containment versus time are shown on Figure 6.2.1.1-1A and 6.2.1.1-1B for the DBA.
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ATTACHMENT 2.
STPECS UFSAR 37.g[.AE LOMo
. PAGE M 0F N'7 pressure and temperaturu profiles have been used- for qualification of the equipment.
6.2.1.2 fontainment subcow artments.
6.2.1.2.1 Desien Bases: Subcompartments within the Containment.
principally the reactor cavity, tl - SC comparaments, the pressurizer compartment, the surge line comparasent, the main steam line compartment, and the feedwater line compartment are designed to withstand the transient differential pressures and jet impingement forces of a postulated pipe break.
Vencing of these chambers is employed to keep the differential pressures within structural limits. In addition, neither pipe whip nor forces transmitted through component supports threaten the integrity of the subcompartments of the containment structure, i The spectrum of pipe breaks ar.alyzed for each subcompartment-is listed in '
Table 6.2.1.1-1. The characteristics of the. pipe ruptures were determined in accordance with the methods and criteria of Section- 3.6.2. - As discussed in Reference 3.6 14 and Section 3.6.2.1.1.1,- item a, RCL ruptures and the
- associated dynamic effects are not included in the design bases. The accident-that results in the maximum differential pressure across - the walls of the respective comparament is designated 'as the subcompartment design basis. 3 Calculated differential pressures are compared to the design pressure values '
used in the structural design of subcompartment walls and equipment to ensure that peak calculated values are less than design values. These design and calculated pressure differentials - are presented under esch subcomprtment section below. >
6.2.1.2.2 Desien Features:
6.2.1.2.2.1 Reactor Csvity - The reactor cavity is a heavily reinforced concrete structure that parforms the dual function of providing reactor vessel support and radiation shielding. It is described.in Section 3.8.3.1 and is shown in the general arrangemer.t drawings of Section 1.2. t.t the elevation of
- the primary. piping nozzles, the reactor vessel is surrounded by an inspection i coroid. No pipe ruptures are postulated in the reactor cavity or inspection ;
toroid. '
6.2.1.2.2.2 Steam Generate . Co n srte nts - The steam generator (SG) subcompartments are shown in Figure 6.2.1.2 3. ' The SC and its supports have ;
been described in Section 3.8.3.1 and the general arrangement of the 70 nd i associated structural arrangement-is presentid in Section 1.2.- The general ,
arrangement drawings presented in Section 1.2 have beeh used to define nodal boundaries of Figure 6.2.1.2-3., The SC subcompartments consist of the entire g ggggg- g free volume between the primary shield and the secondary shield _ walls and from t e s en aYe n o op a !
.J In addition to the above vent path, two more' vent paths vent the -
break nodes to the Containment. These are (a)-the eight peretration paths ;
theti lead the hot and cold leg pipes to the reactor cavity { kith h_ venth ateA M\] '
CAM YbD and -(b) the six heating, ventilation, and air-conditioning (HVAC) ,
vents between the SC compartments above El.19 ft and subpedestal region below (
El.16 ft{dith M\ coral traa Mf V% \t%22 SC compartments A and D, and 3 and C are directly connected together vnilel and b, and C and D are connected via a 6.2-9 Revision 0 l
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PAGE _ V4 0F _ /W7 nodal boundaries be taken at significant flow restriction . De addition of arbitrary nodal boundaries would violate this requirement and could lead to y erroneous results, nis, and the guidelines and recommendstions of Section 3.2. of NUREC-0609, have been followed in the nodalization. In light of this, and since COPDA is an NRC approved program, no sensitivitv studies were performed. This is consistent with Section 3.2.1 of NUREC-0609. The node end junction parameters for the SC loop compartments are given on Tables 6.2.1.2-5 and 6.2.1.2 6.
The flow from one node to the other was calculated using the homogeneous equilibrium model option for the analysis. The peak pressures for each subcompartment are listed in Table 6.2.1.2-5.
The pressure differential given on Table 6'.1.1.2 5 is generally evale-ted The with respect to node 41, the Containment volume, axcepc where specified.
pressure time' histories' for all cases are presented in nodes close to the traak in Figure 6.2.1.2-20. These t. odes are in the 50 compartment in which the breaks occur.
Force coefficients on the SC and reactor coolant pump (RCP) have been evaluated to help facilitate determination of forces and. moments due to the pressures generated by the breaks. Torce evefficients represent the projections of the SC and P.CP on three autually perpendicdar planas selected far this purpose. These coefficients are pretented in Table 6.2.1.2-7 and 6.2.1'.2 8 for these two components.
The forces and moments plot versus time for the SC and RCP are presented on Figures 6.2.1.2-21 and 6.2.1.2 22 for the break cases identified on the ,
figures.
6.2.1.2.3.4 Pressurimer subcom artment - The pressurizar subcocpartment design pressure is established by a double ended break in the pressurizer spray line at the side of the pressurizer. This break location i': in the most restrictive location and results in the maximum pressuo and equipment load. '
The noding of the pressurizer subcompartment is sho*rn cra Fi ra K U .? A R a node #-junction dingtsa.provided on Figure 6,2.1.2- Od. . 4 h, dim
-parameters-ere-.previded in Lbl.a 0,2 1. M ud 0." .1 ^ 10 - Plo3s of s (my [ne 6teaP. ,
cmalaul-aTT6 pressure are given on Figure 6.2.1.2 23, and calculated an a peak pressure are compared in Table 6,2.1.2-9. Mass and energy release races are provided in Table 6.2.1.2 1.
6.2.1.2 %5 suree 1.ine subcom artment - Surge line subcompartments are shown on Figure 6.2.1.2 5, with a node and junction diagram given on Figure 6.2.1.2-13. The model shown is for breaks in the pressurizer skirt area and '
in the vestibule. Node and junction parameters are provided in Tables 6.2.1.2-11 and 6.2.1.2 12. The insulation around the bottom of the pressurizer and piping has been assumed to remain in place. This will be conservative from the standpoint of volume and vent areas. Curves of calculated and design pressures are provided on Figure 6.2.1.2 24 and calculated and design pressures are compared in Table 6.2.1.2 11. Mass and energy release rates are provided in Table 6.2.1.2-1.
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i ATTActiMENT 1 STFECS UTSAR AE-WM l PAGE M 0F M '
- 2. Failure of Main Teodwater Pump Trip No credit is taken for feedvater pump trip and coastdown in calculating main feedvater addition prior to feedvater line isolation. Th.e re fore , l this failure has no effect or. th+ results presented.
- 3. Failure of a M. sin Steam Line Isolation Valve Tailure of an MSIV is assumed to increase the volume of steam piping which empties into the containment by 5.570 efe . B is case is included in the analysis.
The effects of this failure on calculated containment pressures and temperatnres were compared with the effects of the failure of one
- Containment spray train. With respect to the' maximum. Containment presswe, calculations showed that the adverse effects of a maic, staan
. lins isolation valve failure were considerably less than that of one Containment spray train failure. With respect to the maximum containment temperature, no significant difference was found between the two failures.
- 4 Failure 'of one Contairment Heat Removal Train
) The worst single failure following a'steaa line break is the failure of one of the three redundant CHRS trains. The spray actuatiot is assumed to occur 69.1 seconds following the time at which the Containment pressure reaches 12.0 psig. The fan cooler actuation is assumed to occur 30.0 seconds after the containment pressure reaches its setpoint.
4 Add INSERTG2 +. the end of this para 3 mph 6.2.1.4.6 This sectica is not used.
. 6.2.1.4.7 Sherr Term Steam tire and Feedvater Line Breaks: no REIAP 5/ MOD 1 (Ref. 6.2.1.2 6) computer progrsn was used for the secondary system pipe break analysis. The initial conditions for the main steam system are taken to be at full power (100 percent) operating conditions (1,100 psia and 556.6'T). Some of the majec assumptions made in the analysis, all of them conservatively maximizing the blevdown, are the following:
- 1. The postulated high energy line doubla-ended rupture is assu:ned to reach its maximum open area (vich each pipa and discharging th oupi a break area _ equal to the internal cross sectional area of the pipe) within ore millisecond of break initiation.
- 2. During the course of the transient, the SC pressure and temperature are assumed to remain constant at the initial conditions.
- 3. -
All four MSIVs remain vide open for the duration of the transient.
Since the simulation is carried out to one second only, the transient will have endeu before the MSIVs receive the signal to close.
4 De sink volume.is maintained at a constant pressure and temperature of 14.7 paia and 120*T, respectively, with the noncondensible. air quality pegged at 0.98.
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UFSAR Section 6.2.1.4.5, Subsection (4), Page 6.2 INSERT 6-2 "The fan cooler setpoint is at Initiation of Safety Injection signal. The LOCA and MSLB analyses conservatively used a fan cooler setpoint at Containment HI-1 signal (5.5 psig)."
I
STPECS UFSAR ATTACHMENT k ST-HL-AE yo4 4 PAGE:H2 OF f 15.6.5) that considers various break sizes, break locations, and MooF~~.LQ --
d discharge coefficients for the double ended cold leg guillotines that affect the mass and energy released to the Containment. This effect is considered for each case analyzed. During re/.11, the mass and energy released to the Containment is assumed to be zero, which minimizes the Containment pressure.
During reflood, the effect of steam / water mixing between the SI water and the steam flowing through the RCS intact loops reduces the available energy released to the Containment vapor space and therefore tends to minimize Containment pressure.
6.2.1.5.2 Initial Contairment Internal Conditions: The following initial values were used in the analysis:
Containment pressure l': .? psia Containment temperature 90'F Refueling water storage tank temperature 50'F Service water temperature 33'T l Outside temperature 29'T Relative humidity 99 percent The Containment initial conditions of 90*F and 14. 7 psia are representatively low values anticipated during' normal full-power operation.
6.2.1.5.3 containment Volume: The maximum free volume used in the analysis is 3.56 x 10'3 f t . Add INSW 6-3 b"e 4 6.2.1.5.4 Active Heat Sinks: The CSS and the RCFCs operate to remove heat from the Containment.
Pertinent data used in the analysis for these systems are presented in Table 6.2.1.5-3.
Figure 6.2.1.5-2 presents heat removal rate versus Containment temperature (in the post IDCA environment) per RCFC unit. Conservatively high RCFC System heat removal rates were calculated by using the minimum expected component cooling water temperature.
The sump temperature was not used in the analysis because the maximum peak cladding temperature occurs prior to initiation of the recirculation phase for the CSS. In addition, heat transfer between the su=p water and the Containment vapor space was not considered in the analysis.
l 6.2.1.5.5 Steam / Vater Mixine: Water spillage rates from the broken- -
l loop accumulator are datermined as part of the core reflooding calculation and i are included in the Containment code calculational model (COCO) .
l 6.2.1.5.6 Passive Heat Sinks: The passive heat sinks used in the analysis, along with their thermophysical properties, are given in Table 6.2.1.5 4 l
6.2.1.5.7 Heat Transfer to Passive Heat sinks: The condensing heat transfer coefficients used for heat transfer to the steel Containment structures is given on Figure 6.2.1.5-3 for the limiting break. The Containment pressure transient for the limiting break is shown on Figure 6.2.1.5 1.
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UFSAR Section 6.2.1.5.3, Page 6.2-26 -
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d INSERT 6-3 The calculated containment free volume used in the analysis is larger than the actual calculated value in accordance with ANSI /ANS-56.4-1983 Standards. The larger containment free volume decreases the IDCA rt: flood rate which increases the peak clad temperature, s
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. ST.HL AE 40 p_M PAGE.J4 4 0
- 2. A major portion of the recirculated air is returned to the RCFC through the return air risers, which are located se the polar crane rail level.
The rising air with the action of the spray provides adequate mixing of the Containment atmosphere above the operating floor.
DEitTr.]
o 1 bko\ C thina' t h dk a f re 1 e Cs in pe at n ra o 18 530 ) 1 qvi ig a ra.ed o Co et up n Na o th 8 a e n en e emp en o 8 5 c (arb a ay 1 e. K%
- 2. 1 ontheopya'n f1 e on 1 en s in 1 ei O c,o wh h 3 ,4 , O ns de d r e a e v 1 el 0 n.
nthna s s eo .
i
- 3. re io 3 u 00 f is\sp ye in ty- o o 28 5 ,o .
da ve ed 'o 4 1 co a e e, Npace es t o ca R ge a o
o a to a 3 vo um ch ge- p .
hour.
Nas e ir flows d rag.takgttiy,hou .
( fg0 . ne mp rth p n f be ' o is t r in o ni s thh1 p to th 1 o da a e v 11 us to re s f e en i e s co a e v 1 ne di e
- n. (s e 9 . an E. 0 che r e .
8 i c r a ra d. s egtgn w e 6
\ 3k00 vytta 1 9 p et a s a d to eg n ve.
h i s d a to s a ro h*e o o r is s . 30 t-e o gne f a ompar e i th ul ac ev n . 8 -
- 5. 0 3, f hi ,\00 t 3
s O n. anh (- 11 t- i 7 ,
e r s na is t i e r e o id e sp .
d th s o st 1 va d s f a ov '
o w re ru gr ti Ap o y 7 ee t a RC El. 6 ft .
a n uta d se ar ow i 1 es th qu t s gg n El 2 c. \ \ _
Figures 6.2.2-13 and 6.2.2 14 show the RCFC and relsted ducevork. Figures 6.2.2-6 through 6.2.2-12 illustrate the CSS spray covertge at various elevations in the Containment. Table 6.2.2' 5 provides estimates of the spray mass flow rates in the individual regions identified above.
6.2.2.3.5 Purm Net Positive Suction Head Recuirements: The minimum available not positive suction head (NPSH) for the CSS pumps is such that an adequate margin is maintained between the required and the available NPSH for both the injection and recirculation phases , ensuring the proper operation of the CSS . Recirculation operation gives the limiting NPSH requirements for the CSS pumps.
- Figures are based on the RCFC runnin5-6.2-36 Revision 0 I
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} I ATTACHMENT 1 STPECS UFSAR ST HL AE-4o(,lo PAGE __5ct OF N 7 _
6.2.5.2 System Desien. The Containment Hydrogen Control System is designed so that all components are of seismic Category I design. All components located within the Containment are capable of withstanding the environmental conditions in the Containment following a LOCA, as described in Section 3.11.
The hydrogen recombiners are located inside the Containment. The recombiners are supplied power from separate buses in the Engineered Safety Features Power Supply System. The dynamic effects of pipe rupture are considered in the ,
design and location of Containment Hydrogen Control System components.
6.2.5.2.1 Princioles of Ooeration: The Containment Hydrogen Control System is based on the following concepts:
- 1. Following a postulated IDCA, the hydrogen concentration in the Containment is assumed to build up due to metsifwater reaction, adiolysis, and corrosion. The hydrogen concentration is observed vie Rep lnte wH b he Containment hydrogen monitoring instrum sntation, and if the pgg p oncentration approaches the RC 1.7 control limit of 4 percent by vo; ume j th: :::pcint s ad i. 3.', yercent by ni a C 7 the hydrogen recombiners are manually activated.
- 2. As a backup to the Hydrogen Recombiner Subsyste.s. a purge system is available for use to force the hydrogen-air mixture out of the Containment through the Supplementary containment Purge Subsystem. The air thus removed is replaced with air, via the Supplementary Containment Purge Subsystem Su ly,_thus reducing the overall hvdrnren -
y concentration. INSERT 6-5 at ne end of W5 f a'a 9fh 6.2.5.2.2 Hydroren Monitorine: The hydrogen concentration will ts continuously monitored following a iDCA and displayed in t!'e control room.
The Containment Hydrogen Monitoring System is described in Section 7.6.5.
6,2.5.2.3 containment Mixine: One of the functions of the containmunt mixing subsystem (RCFC Subsystem) is to provide homogenuous m'.xing of the Containment atmosphere to assure that pockets of hydrogen do not occur. Tha system is SC 2, seismic Category I. It is completely redundant, with duplicate piping, equipment, and instrumentation located in opposite quadrants of the Containment for maximum functional independence.
The RCFC Subsystem as described in Section 6.2.? has six 33 1/3 percent-
- capacity fans which take suction from the upper levels of the Concainment.
The containment atmosphere is then forced through the fan coslers and ,_
discharges into the lower portion of the Containment within the secondary
~
shield wall, circulating around the reactor vessel, SCs, and pres;turiter, and -
rising in the center of the Containment to the upper and dome levels of the Containment.
Subcompartments, such as the RHR HXs and pumps, pressurizer, reactor coolant drain tank, etc. , are designed to avoid pocketing and the resultant buildup of hydrogen above the 4 percent' limit.
6.2.5.2.4 Electric Hydroren Recoqh1Mrg: The electric hydrogen recombiners are designed in accordance with RGs 1.7, 1.26, 1.28, 1.2 s 6.2-49 Revision
\ sNs
t i
! r-----.__-
ATTACHMENT C o ST HL AE +JL 6 UFSAR Section 6.2.5.2.1, Page 6.2-49 L PAGE .51_._ og fej 7_ ,
i INSERT J-4 The setpoint used in the analysis is 3.25 percent by volume.
INSERT 6-5 However, the analysis takes credit for the hydrogen recombiners when the containment hydrogen concentration reaches 3.5 volume percent, at which time the recombiners are assumed to operate at full efficiency. This gives the hydrogen recombiners about four day,s,to reach full efficiency from the time the recombiners are actuated at 3.25 volume J percent. The hydrogen concentration Watt peaktwhen the recombiners reach full efficiency.
1 b
S t
4
1 AUACHMENT l
- STPECS UTSAR gy.ggE. Qo(o (p -
PAGE 5'L .0F --
i contains an isolation transformer plus a silicon control rectifier (SCR) controller to regulate power into the n - ..e r . This equipment is not exposed to the post lDCA environment since it is in the Electrical Auxiliary ;
Building. To control the recombination process, the correct power input to I bring the recombiner above the threshold temperature for recombination is set ,
on the controller. The correct power required for reccabination depends upon l Containment atmosphere conditions and will be determined when recombiner operation is required. For equipment test and periodic checkout, a i thermocouple readout instrument is also provided on the main control board for monitoring temperatures in the recombiner. ;
Reference 6.2.51 provides a description of the testing ci' a full scale prototype electric hydrogen recombiner.
~
. 6.2.5.2.5 containment Hydroren Purrinr: A function of the 4 Supplementary Containment Purge Subsystem is r:o act as a backup to ths Hydrogen Recombiner Subsystem. If purging is necessary, dilution of the containment atmosphere is achieved throu:;h the use of the Supplementary Containmenc Purge Subsystem, described in Section 9.4.5.2.1.
For hydrogen dilution following a 1DCA, the system is opuated as follows:
- 1. At a Containment hydrogen concentration of 3.$ percent by volu:ne, ont of tvo 5,000 fc3/ min redundant exhaust fans is starr.ed, taking suction from the Containment atmosphere when Contaiment '
pressure h near atmospherie.
- 2. Makeup dilut!on tir is supplied by the Suppleneatary Containment Purge Supply Subsystem. One of the two supply fans is started to provide the
, dilution air flow into the Containment. The flow rate is adjusted by an air-operated flow control damper.
- 3. The Supplementary Containment Purge Subsystem meets the requirements of RG 1.7.
6.2.5.3 Desien Evaluation.
6.2.5.3.1 Hydrocen Recombiners: Prediction of hydrogen ganeration following a LOCA shows that although hydrogen production rate decreases with time af ter the accident, . increase in total hydrogen accumulation could exceed k
Win 4 percent by volume, and coner ensures ye necessary to prevent hydrogen 6acep h tien to this level. The hydrogen recombiner provides the means to TSSERT b7 prevent unsafe levels of hydrogen concentration from being reached in the Containment following a 14CA. A prediction of the hyhogen generation is cantained on Figures 6.2.5-3 through 6.2.5-5. Figures 6.2.5 3 and 6.2.5 4 ]- .
show the hydrogen concentration in the containment versus time following IDCA -
and the effect of one or two recombiners starting:27.25 days after the lhCA. l Each recombiner in assumed to operate at a minimum design flow rate of 100 scfm. _ _
6.2.5.3.2 Containment Material Reouirements: In order to ensure that hydrogen generated from secondary sources such as corrosion is kept 'co a minimum, aluminum materials are not used for Containment structural purposes, and their use is minimized for other purposes.
6.2-51 Revision [
, - ,-e
~
ATTACHMENT /
UFSAR Section 6.2.5.3.1, Page 6.2-51 ST HL-AE yo06 PAGE 5 7 0F /V 7 INSERT 6-6 The hydrogen recombiners provides the means to' prevent unsafe levels of hydrogen in the containment following a LOCA.
Figure 6.2.5-3 predicts the containment hydrogen concentration versus time following a IDCA. Figure 6.2.5-3 also shows the oflects of one med two hydrogen recombiners working at. full ef ficiency af ter 31.4 ._.tys following a IDCA. The analysis assumes that cach recombiner. operates at a minimum design ;
flowrate of 100 scfm. Figure 6.2.5-5 shows the hydrogen generation rate versus time following a IDCA.
e
! ATTACHMENT /
STFECS UTSAR ST HL-AEfo66 JGE .5.L OF /V 6.2.5.3.3 Hydreren Analysis: Hydrogen is generated within ~the l containment by various mechanisms, as described below.
- 1. Radir lytic Hydrogen Genetation Water is decomposed into free hydrogen and oxygen by the absorption of energy emitted by fission products containad in the fuel and fission t products fntimately mixed v.ith the LOCA water. The quantity of hydrogen produced by radiolysis is a function of both the energy of ionizing radiation absorbed by the thCA water and the not hydrogen radf olysis yield, C(H;), pertaining to the particular physical chemical state of the feradiated water.
' Studies indicate that the net hydrogen yield fror the radiolysis of pure water is 0.44 0.45 molecule per 100 eV of absorbed energy when the l
gaseous radiolysis products are continuously purged from the water. In the presence of reactive solutes and water in the absence of gas purging of the solution, significat:c recombinatica of the products of radiolysis can occur, thocePf reducing the net hydrogen yield. Hovaver, in accordance with RC 1.7, a value of 0.5 moleevle/100 eV has been assumed for the net yield of hydrogen from radiolysis of all LDCA water.
The assumptions given in RC 1.7 were used to determine the fission-product distribution after the accident. This distribution is assumed to occur instantaneously aftcr the ace dant, and hydrogen production is j
assumed to begin immediately. Fifty percent of the halogens and one
- percent of the solids in the core are assumed to be released from the fuel and intimately mixed with the water in the sump. All noble gas activity is released from the fuel and is present in the Contatraent atmosphere. Table 6.2.5 3 gives a su:us:ary of the assumptions made in
! the analysis. . The analysis was based on the equations given in Appendix
! A to NRC Standard Review Plan Section 6.2.5.
, 2. Zirconium Vater Reaction l One of the major sources of hydrogen ic: mediately following a LOCA is due te, metal-water reaction. The extent of the metal-water reaction depends j strongly on the course of events assumed for the 1DCA and the l effectiveness of the ECCSs. The extent of metal water reaction is evaluated in accordance with the assumptions of'RC 1.7.
ll Zirconium reacts with steam according to the reaction:
3 Zr + 2QO A Zr02 + 2H2 Thus, for each mole of zirconium that reacts, 2 moles of free hydrogen l
i are produced,f j
! l; The NRC model assumes a 5 percent zirconium water reaction, however, I
based on NRC Branch Technical Posit ta (BTP) CSB 6-2, a 1.5' percent zirconium water reaction is used for analysis of this plant. The
, hydrogen generated is assumed to be released to the Containment l
atmosphere. For the estimated 50,780 pounds of zirconium metal in the ]
active portion of the core, this amounts to 762 pounds of-etreo i reacting. The total hydrogen generation n chie e m e is then l
from zirconium-waiev reaction, I ' 6.2-52 ITlon 4As
k STPECS UFSAR ATTACHMENT I EEFERENCES PAGE._55_0F /d _
Section 6.2t l 6.2.1.1 1 Bechtel Power Corporation Computer Code: COPATTA User's Guide, Volume 1. " Practical User's Guide', Volume II, ' Theoretical User's Guide" 1974, i 6.2.1.1 2 Bechtel Power Corporation. Topical Report No. BN TOP-3, Rev. 4,
' Performance and sizing of Dry Pressure Containments', March 1983.
l ~t S
6.2.1.1-3 G1husMev-rhack, D. b C.,eC Review k. of Heat Transfer coefficients for Condensine Steam in a Containment Buildine Followine a loss of Coolant Acciden. IN 1388, September 1970.
6.2.1.1 4 Uchida, H. , A. Cgama, and Y. Togo, " Evaluation of Post-Incident Cooling Systems of Light Vater Power Reactors", Proceedines of the Third International Conference on the Peaceful J.gn of Atomic Enerry, Volume 13. S e = sion 3.9, United Nations ,
Geneva (1964).
6.2.1.1-5 Tagami, Takashi, " Interim Report on Safe. Assessments and Facilities Establishment Project Japan for Period Ending June 1965 (No. 1)c.
6.2.1,2-1 Bechtel Power Corporation, 'COPDA Compartment Pressure Design l
Analysis", (Bechtel Computer Code), 1973.
S.2.1.2-2 Bechtel Power Corporation, "Subcompartment Pressure and I Temperature Transient Analysis", Topical Report No. BN TOP-4, (Rev. 1), October 1977.
6.2.1.2 3 Crane Co. , "Flov of Fluids", Technical Paper No. 410, 1969.
6.2.1.2-4 Idel' Chik,1.E. , " Handbook of Hydraulic Resistance Coeffeients of Local Resistance and of Friction", AEC-TR-6630, 1966.
6.2.1.2 5 Shepard, R.M., H.W. Massie, R.H. Mark and P.J. Doherty, "Vestinghouse. Mass and Energy Release Data for Containment l
Design *, VCAP 8264-P-A, Proprietary (June 1975) and VCAP-l 8312 A Revision 1, Nonproprietary (June 1975).
l l 6.2.1.2-6 REIAP 5/ MOD 1 Code manual Volume 1: System Models and Numerical Methods, NUREC/CR-1826, ECC-2070, 1980.
6.2.1.2 7 American Nuclear Standard, " Design Basis for Protection of Light Water Nuclear Power Plant Against Effects of Postulated Pipe i Rupture, ANSI /ANS-58.2 1980.
l l
l 6.2-58 Revision 0 l
~
strEcs UrSAR
' TABLE 6. 2.1.1 2 :
M CHMEhf.1 -'
ST-HL-AE lD M LIMITINfb DB A ' CAL.C.ut.Atro PeAw 7%wssvRes 44EhD_ Sp a 34 #4 d
.= :. na a-cJizD 51.n u e15 ~ ~.
_I >
' pne5S U RE FDR ' CON *TAINMENT -
WA< MAX / MIN Design Basis Design Calculated Parameter Accident % - W1= e*J 1 . Marrin imwgNAu PeakyPressure Double-ended Pump Suction Cuillotine 56.5 psig M sig y M e.g- 2 f.3 ~
Break with Maximum Safety Injectiori and Minimua Con-tainment Heat Removal. *
.Pr '.X '
Exte nal .
MsN Inadvertent Opera - (-)3.5.psig (-)3.08psigf 12.04~
Pressure tion of the Con.'
tainment Spray Systea ,
r 6.2-62 Revision 0
.i l: ;
1
!~
i
.4
.'.....a,-.--.-,.- - ..----- - - - - ~ ---~- -- ~ ~ - --~~-~ ~~~~~~~~~~' ~ " " ~ ~ " ~ ' ' ' ~ ~~ ~ ~'~ '
- ._._ . _ _. - . _ _ . _ _ _ _ . _ . ___ _ . .. . _ _ = _ _ - . _ . _ ._. _.
-[;. .
- STPECS UTSAR
. ATTACHMENT 12-TABLE 6.2.1.1-3 ST-HL AE-t/04#
PAGE <? 0F M7 CONTAINMENT DAT6-I. Gsneral information A. . Internal Design Pressure: 56.5 psig-B. External Design Pressure:
. (-)325 psigi ;
C. Structural Design Temperature:
8 D. Free. Volume: 3 6'
O' +3.4tatoff! 9 O##- - 0. 5 0 x 1[' f f- 2.ionoft-E. Design leak Rate: ~
0.3% per cay II. Initial Conditions A. Reactor Coolant System .
(At design overpower of.130% and at normal liquid levels)
- 1. Reactor Power Level: -3893 Mvt (includes 17 Mvt RCP power) -
- 2. Nominal SG Outlet Coolant Temperature: 560.0*F
- 3. Nominal SC Inlet Coolant Temperature:
' . 625.1'F 4
Reactor Coolant Mass: 573,260 lba
- 5. Liquid Plus Steam Energy *: 345.9 x 108 Beu B. Containment ,
- 1. Pressure:
0.3 psig -'15 psia
- 2. Temperature:
120'F
- 3. Relative Humidity: 70%=
4 Essential Cooling Water Temperature: 110'F-
- 5. Refueling Water Temperature: >
104*F l
- 6. Outside Temperature:
95'F-C. Stored Water (as applicable)- ,
- 1. - Refueling Water Storage Tank: 60,306 ft 3-fe,ect
- 2. . All Accumulators (safety injection . 3,600 ft' tanks):
A*4 FRROR BAKO APPLiss To - Tuis CALC.JLateo NEY . FRS6 . \/c4.ME VMLUE. !
we - espeg BAgo Ac.coums - FoR Tut -. V8 Reus , Misc.et.L.AMEOO3 ' Coeoo me,5 ,
(sov SPreerw ety c4ctutareo. .
All energies are relative to.32'F -
- i. **
6.2 ! -Revision-0 l
o I-i 7
wr+=. w w - w T w " "i n.va-+r -v r --+, w +
, v e Www ww w, --sm r* = e v - g = -q, w.t-r yo-~ w,w e -, ewe-e-v -v r W t erwe, y -m -ge e w . e v =s ye g y-m *w y - y-- - ' Me
.. m . . . . - . . - - . . ...m._ .. _ _.m .. ~ ._ . . . . _ _ _ . . . _ _ _ . = . _ _ . ~ ~ . . _ . . _ _ , _ .e.... . ~ . . . . . < . . . . . ~
4... m.
l j- - . , .
i i '
g i TASLE 6.2.1.1-4 SUMMAnf or CAttutAtto CouTAtentul Patssuet AND tretsar>nts (tota 3 i i l . i ener w noteeeed ,
1 Peek Peak flee of- to Careelruent !
Pressure Temperature Post Pressure tsp to End gf
. Pipe great Aree and Type - (cola) f*F) feeconds) 4 tendo m (10" Stu)
I !
+
1 37.3T . 30T.0 - 82.6 373.36 10.48'ft2 DEPSC, Min. SI, Min. cu t 1
rm,SA 3:31 h.. -30h5- 82.4 373.36 1 10.48 ft 2 etPsc, Max. 11, Min. C e s -t t
36.79 ' 282.0 39.3 375.45
.- 9.18 ft 2. DENL, Man. s!,' Min, cHas 2 ,,, u,,, 3 ,,3 a.2, ,,2 3.,,, ,
I ,,et , ,,,, , , , , , ,, e. ,
4 6.29 <t 2 - ' o.6 otPse, Man. si, Min. ens 36.77 ' sos.s s2.s 372.co 3 o
e2.6 3es.or -
l . .e - 3 et 2- Pss, Men. si. Min. ens 36.o4 29s.o 7 4 m
l-
" 3- . !
.. i
?
i 5
mww EPd !
> h Add TNSERT '6-7 here *@E W>m 3 4mic *m ,
h !
'E Wh Sek c r.
i- \ -
2 f
v O
1 (
- f. O I
kv ..
.i i '
_________m_.__ _ _ . _ _ _ _ _ _ _ _ _ _ _ + , ~m
I ATTACHMENT.2 ST-HL AE-404,lc UFSAR Table 6.2.1.1-4, Page 6.2-64 PAGE A 9 0F / 47_ i INSERT 6-7
" Note: The analysis for 10.48 ft' DEPSG with Maximum SI and Minimum CHRS was performed with a containmet,t free volume of 3.20 X 10' f t 3. All the other IDCA analyses were performed with a containment volume of 3. 56 X 10' f t3 . The calculated minimum containment volume (including uncertainties) is
- 3. 3 8 X 10' f t3. "
l s
r r e e , + = +
s TABLE 6.2.1.1 8.
ATTACHMENT ST HL-AE- Li ot,c *.L; THERM 0 PHYSICAL PROPERTIES, j PAGE Q 0F.,- j y 7 - j or STPrCTURAL HFET SINF.y FOR IDCA AND MSLB ANALYSIS
( .:.
t N rmal Volumetric Haat- l I
Conductivity . Capacity Material (B ru /hr. f t. *34-- i (Etu/ft8.*M- +
i
,e?s- O.375 49.9: ;
- Amercote 90 (organic) ';
i Dimetcote 6 0.633 21.67 (inorganic)
Nutech Paint 0.1258 28.29 ;
i Air 0.0174 0.0101 Carbon Steel 25.0- 54.0-I Concrete 0.~ 8 30.0
- Stainless Steel 9.4 54.0 r i
F i
l-e i
es l
_ % 3
- h i
f
..A .!
6.2-74 'Revisior P
r l- . . . _ . . . . _ _ , . - - . . . . . _ - . . . . . . . . . . , , - - . . . - _ , - - - , . - . - - , . . . , _ _ _ . . . ~ . _ _ , _ .
t
k
! STPECS UTSAR ATI ACHMENT -7 TABLE 6.2.1.1-10 ST-HL-AE 4%(o PAGE 61 0F /2/ '7 ACCIDENT CMRONOLOGY A. Most Severe Pump Suetion Break Break Type: Double-ended Guillotine Break with Max SI, Min CHRS Time (Seconds) Event 0 Break 16.5 Accumulator injection begins 21.0 Peak Containment pressure during blowdown 25.0 ECCS injection begins
~
-3 38.o Beginning of fan cooler operation
- 82. Beginning of Containment spray injection I
82.6 Peak Containment pressure 137.0 End of core reflood 1,216.0 Beginningofrecirculatjon f- /veduced to .
-3,025.0 2930.0 i
V containmentpressure(isA50perce of pea value (sed 3e pressv<e, t
4,008.4 End of steam generator energy release 46,800.0 Hot leg recirculation initiated (13 hrs) 4
.~
I 6.2-74 Revision 0
, ,, e ,-o r
~ . , - . . - - . . . - _ . - - - - . - . . _ . ~ ~ - . . . - . . . - - - .. - ~ - _ - - . . - .. .,
i t
4 p
taett 6.2.1.1-11 i r
=
COMPesattyt etsults;_._.Stmoet F st3 tits F_ttyta; mort Pottsune aso trartsarvet poetests fee tat SMCteum F resretarti tMa ACCT 9tets l 4 f f W
i I
2 3 4 3 4 1 ,
accl& 4 how outtien PS wet les told les PS PS seed toestlen ,
(es) etc 0.6 tit setSt '
6
{ Deele-mbd CEE DEC j .:.
Sreek type
- *- guillet 4w f
(DEC; 3.99 ft2 10.48 ft 2 10.46 ft 2 v.13 ft 2 8.2% ftI. 4.26 ft 2 seest stre swa one m em mogr Safety lajectle, ein
- ia pin ein elm l ein ein
' Cente6few=q seet 9,= newel System (twes)
-9P:9- 36.S 30.5 34.8 34.0 ,
Ped preeaure, pelg 37.4 4 i
82.6 82.6 39J 16.05 82.6 82,6 g 1 .tfae to peet pressure, oee o '
33.I. 296.6 305.4 295.0 307.0 -30*+ 252.0 '
ra i 7 eeek teveesture. *r w
- 87.6 82.6 39.3 82.6 82.4 - 82.6 @
Tlee to peak te v*r( % . set $ l tewegy relee:,d te T=e*etwa et tsee of peet p ws - 44&.37 AM.46 i 4 % .82 3&c.31
! 106 see 451.1 - 452.48
! teeryr etnerted br pesolve '
- best side y time of ped $7.59
-eer36- 62.67 19.93 79.96 75.73 pree9we.W 9te 82.69 4'
. i T tw9rlavePee Wim 287.60 269.9 291.13 299.98 m (n >
3t2.30 tsee of pad pressure,1 See 296.9 N MTd m
' tare,y h. e eeter et iof. 33 69.1 99.96 96.96 %.pg =
tlae to pr ; ' N9sure, M Stif 99.10 48- 77.4 0 m!!::
b * /"
j t e - w by tentel e f 4 's eqp t time of 3. N 0.9 3.06 2.72 op
$h !
Stw 3.21 -4,@9- 9.9 m 9 sw W.1 C A. .
e 4- - ,
o se, w t- vreele g
[- ) spreys out t tie = et peek e., s.e e.e q o.e e.e .
i l ,r m or, 1 st.
i 2
V a
j / '
'
- AM INscERT 6.-g here i.
! 4 -
l
UFSAR Tabic 6 2.1.11, Page 6.2-79
!--: AT1 ACHMENT A '
ST HL AE*'iA It /'/ q .
PAGE .121-. _0F INSERT 6-8 t
Not e : The analysis for 10.40 ft' DEPSG with Maximum SI and Minimum CHRS was performed with a containment free volume of 3.20 X 106 ft. 3 All the other thCA analysen were performed with a containment [
volume 'of 3.56 X 10' f t3 . The calculated minimum containment volumo (including uncertaintion) is 3.38 X 10 8 ft." 3 I
h I
i I
t i
h
?
I k
h i
h r
4 l - -
i; i Ih' Ik ;' ,7ii ;: f tI[t [{ ! lI ll:f: i)!lli' r! f n t;!jL!Ifi;6 l' i y4sEm =a c2N -
- h$4A
>4yoQT nCgr t-o>g '
oDb::,Y r
s i
e et e t
e ee rt s 2 8 t u 9 7 4 2 0 2 eO 7 7 3 5 2 9 yr*J 2 8 t et 1 6 3 0 2 3 1 4 9 c 2 7 4 2 2
Dc e7 2 s 3
9 1
2 2 6
4 8 3 9 3, 6 1s8 3 1 1 f
oit e
gde Cdre ree W37 is 9 3 8 6 9 7 8 2 0 0 0 t. 7 9
8 9 0 3 8 0 8
1 I f t 6 8, 6 1 1 1 7 -
gc 9 e 3 1, 4 1, 3 - 1, ee 6 -
G SSM -
1 t t 1 1
. g 7 foee
- d t
) t, 4 7 0 4 9 6 1 8
7 0 3 0 3 7
t T a
r 9ttB' u eO r
t s 0
3 1
1 9
0 1
0 1 3
2 1
3 2
5 2 0 0 0
1 3
5 ce C
_ )
d m
t a
g a
t l
u W3 n B r u
r v t t T i 6 t C t ee n otu e
r a
s r e us 5 7 0 2 0 0 C
( t e
$ ks6 s 8 5 7 0 0 9 5 7
0 0 0 3 0 8 Z ae 6 5 2 0 8 2 0 2
1 ou e t
t A
- er2 PP8 9
2 9 8 7 4
1 6 4
- - 7 4
- t t t t -
t 1
t a, n e o t 1 i s t n a 2
i f e S t
6 t t E
s u st t u t e ni t
e e s t s f oe m t t E od e 4 7 1 0 2 9 4 0 0 0 0 0 2 i s a us r S t f u
r te5 0
8 1
7 3 5 0 6 8
2 1
3 7
0 0 0 0 0 6
8 3
CS G fS2 2 3 3 3 P
E D
ne
- o e d e 3 6 0 0 9 9 9 0 0 0 0 0 8 uk s oe 9 4 5 0 2 2 9 0 0 0 0 0 2 3
t e1 5 5 1 3 1 t -
SP2 2 3 3 3 m
2, 9 0 0 0 9 e t
e l e t e i s 9
2 1
0 0
0 0 3, 2 1
9 2
1 0
0 0
O 0 0 0 y1 2
.H n 7 IO e
- g. 8 9 e g
n 6
- i. C - i n
1.
p in m mm t
R u
T C
B t
ue o ;s i
n T sa e.
e oe rt r ye y y o%'ter m o R r
e e t
e s y g
f s y b bp b r
b t c E e y. r ds d do d d d _s y S t t s g e e e e e e ev;e t s ne n
e
- e w
e r e n e My eea r d'
.de d c ds v
o o v
oeo e s N m
u m
u , n ay ai m e me m ee l
e n I r yr e d E s m t
r t
r e o E. t e yi y r r rde e e t c
t e
t sr ep gp re g 2 t
ii t e J t t r5 t t ew n
o ne t.e e e
t e
it n n s e n
e n e o
e e
e o
t o J C C m s Y. I E .E E p N p t A
e MI" n g<eEo=
- _
,7i 3ji}<1i ,i ) i II ! ltib $l
.]
l 1
ATTA0tiMENT 1 UFSAR Table 6.2.1.3-12, Page 6.2-82 E F M '/ )
INSERT 6-9 The results presented in this Table are for a containment free volume of 3.56 X 10 6 ft. The revised 3
analysis for a containment volume of l 3.20 X 10 6 ft 3 did not contain sufficient information to update this Table. Section 6.2.1.1.3.1 discusses containment volume used in LOCA analyses."
7
)
5 i
, STPECS UFSAR KITACHMENT 2.
TABl.E 6.2.1.1 13 SI*HL AEdK> (c (c PAGE .11_ 0F l'/ 7
[g TAINMENT ENEPCY DISTRYBli10N Host Severe Temperature Secondary System Break Break type: 1.4 ft8 Double. ended Rupture at 102 Percent Power (Min. CHRS) l l
Enerry (10s Etui ;
Prior At Peak 888 End of "3 i
- - Ene 1,g.,);tn}; _ Pressure Calculation I hiergry
% \ e w dc.wn 4Energyrej)easedto 0 -29 & 368 ra nr9 634.I4-l I V~
U
" M nmeht (See Table 6.2.1.4 2)
Energy Content of RCB *th911. 6 ami 213.O 494138 3l Atmosphere 888 D Id D Energ'y Content of RCB 0 -4,4 l l S. "2 -1& 9 'P01 6 Internal Struccc es (88 I/
% I V~ l Energy Content of 0 4 SD, C 4Hr4 763. 6 Recirculation Water R_/ d (r. ump)
Energy Removed by Reactor 0 --Eh4- 13 3 -11 I l'37 Containment Fan Coolers R/ V Energy Removed by 0 -0,0 11. ~3 -MM I08,7 Containment Spray Systen O I'N /
--->- Adel INSERT 6-10 here.
- 1. Based on 32'T Datum
""e4LcoJ29'LQatm
/g 3. At 409.75 ;;co+.4s 9 66 ser.ond S
- 4. At 2000 seconds
~6 2 84 Revision 1
_ - . - - _ - - . _ . . _ _ _ _ _ _ _ _ . . .. .- - ~ . . . - - - . . - - - _ _ _ _ _ . - . - - _
I l
UrSAR Table 6.2.1.3+13, Page 6.2-84 A11 ACHMENT k STHl.AEll040 PAGE V/- . OF /l/ 'l !
~
INSERT 6-10 t
" Note: The analysis for 1.4 f t' DER at 102% Power with Minimum >
- CHRS was performed with a containment free volume of
- 3. 2 0 X 10' f t3 . The calculated minimum containment volume (including uncertainties) is 3.38 X 10 6 ft." 3 1
s I
STFECS UTSAR --
f ATTACHMENT 1 ST HL AE-g TAB 1.E 6.2.1.1 14 PAGE (d_4066 O F Lt/ -],
StHvARY OF THE CALCU1ATED CONTAINME?TT PRES $URE AND TEMPJfXIURE FOR MSLB f N Break Feak Press.gred esigi Peel Tenerature ('F)
/
1.4 ft8 DLR, 102% Power, -Md- @6. l 4H- 3 N 8 Min QIR$ j 1, 1.4 ft8 DER, 70% Power, 24,9 322
,f
- Min CHRS - 7 ,,,,
1.4 ft8 DER, Hot Shutdown, 4%.4-29,I CR 3 R '7 Min CHRS O ' _
O.86 ft$ Split, 102% 24.5 306.5 Power, MSIV failure Mu CHRS 0.908 ft$ Split. 70% 24.4 306.0 Power, MSIV failure Max CHRS 0.942 ft2 Split, 30t 24.5 304.3 Power, MS!V failure Max CHM 0.4 ff/ Split, Hot . 22.9 287.1 Shutd;wn, Min CHRS
.=# %,
x 4 A dd IN S E RT 6-11 here.
(m. t
'"Xis .
4 p s,
'.2 85 ,[ Revision 4d t -
UFSAR Table 6.2.1.1-14, Page 6.2-85 " ATTACHMENT ?>
ST.HL.AE l ffo op m y L_ PAGE L3_fo --
INSERT 6-11
" Note: The analyses for (1) 1.4 f t' DER, 102% Power, Minimum CHRS and (2) 1.4 ft8 DER, Hot Shutdown, Minimum CIIRS were performed with a containment free volume of 3.20 X 106 ft 3. All the other MSLB analyses were performed with a containment volume of 3.56 X 10' f t 3. The calculated ,
minimum containment volume (including uncertainties) is 3.38 X 106 ft." 3 t
I e- =v- i-- sv* .g v w u-pc.,cy-r '-a
---Tw-- -
c- -1rr - ,+-v--r- T T T *'T *=r--
i STPEGS UTSAR ATTACHMENT *2 TABLE 6.2.1.1 15 ST HL-At to46 ACCIDDTT 01ROf40 LOGY Des 1 & n Basis MSLB (1.4 ft8 DER 102% Power Min CHRS) 4 Tine (S eendri Event l 0 Break
- 1 Main feedvater and main steam isolation setpoints reached 3 H1 1 setpoint pressure reached 8 Main feedwater isolation conplete 10 Main steam isolation complete
/
fd0 3 39
' Peak 4e*Pmtv<e of 328 : 'F t . < 'ac k e d Fan cooler operation begins k
g(di ut.3seepointpressurereachy
-e n w .p.r a r.-. m a n = ..a
+ u
-i* d65 reek pressure or 40-psia .adesched.
g
-~
ggray systems lae5 1 n delinring now t Containment 1.800 AW pumps secured a NJ IN SER T 6-17 here.
i ,
6.2 86 Revision-03
- x. d'
, ~ . - . - . . - , . - , . , - . . , , . + ,
I UPSAR Table 6.2.1.1-15, Page 6.2-86 l ATIACHUU!T ST Hl-AE.406 2f PAGE 2L. DF 1El . ,
INSERT 6-17
" Note: The analysis for 1.4 ft* DER at 102% Power with Minimum CHRS was performed Vith a containment free volume of
- 3. 2 0 X 10' f t3 . T)e calculated minimum containment volume (including uncertaintics) is
)
- 3. 3 8 X 10' f t3 . "
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, of Wt 7 6,2,1,t 3,5 3,2,3,2,5,8 3,2,1,5,5,5,5,3,3,5,5 5,2,2,5,2,1,2,2 1,1,'4,5,4,4,4, 12224644431 222444424455001 0T337767000111 1 333 en a
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Tiant GEeEnat0R LOOP CSEPa8Teent petssumE-ftsetesTture amatists secas,erteent sedet seeceiptian'*3 i
Det t y* sorgin [
ties to Peek sede volume Wol Pook Pressure (b)
. se.
f fft i Dfffm . afet (no?al Pressure (sect 5 %se Ce effference *ressure Pe* rent i
4041 7.50 e100 40, 30,450.00 0.3 0.30 1 ,
i
-i, ^ , ^ j 49('d) 42 2,1 4.28
- S. $6 8 X IO "
, et i e '
H 4
i g -
4 we e
i.
os e
w . g w ,
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e C
3* Cn 3=
H --1 ,
cm m22 -
- .TO QPZ -
- e. Inittet cerufittene for ett nodos are 9denticets e g
- 120*F, Preeeure = 14.7 pela and retettwe Iwasidity = 100%. W. m =m
_C CZ
- b. flisee are peak differenttet preeeures tietyven the rede eruf the contelruent (node 41) escapt eeere speelf fed.
- p. a riO '
9 j p
( e. Brook Casees -
< I ,
t i . a in. ur,e, - no.e . is
' 3*
a 2 = 16 In. surge llee, break node = 14 s'
3 e 12 In. mar /st tirw, break node = 12 1 7 2
a sede 41 s i- O to contelneent and nede 42 to reacter twlty, the denlyt- f i .\ d, of botti are ewe based en etene sewester too, ces,orteent Pt.enetrose.
j ,
y
- - i i
4 J
I
1 sTrr.cs UFSAR ATTACHMENT a j 51 kL AE 40%,
TAB 1.E 6.2.1.2 9 d4GE ~74 _0F _/di 7 i
( ~
PRESSURIIul SUBCOMPARTMENT ANALYST $ ,,
' SPRAY LINE RRFAE Feak Time to Fe'ak
' Nat Differential Differential Design Design Volume Pressure Fressure Press are* Margin (see) (t) j Node fft') (naid) (esid)
-1 Si tt? . ' ^ -
---?.1$ 0.iV
-e- l 693.50 edt 813 4r30 C. C _i
-92 659.60 arte G.76 ArH4 o. 25 * !
4- 3 663.06 4t44.T.51 4N O.' A9 j
+t 703.44 6,M 5.6 4 GriHW O, 2 6 -
+5 1,101.25 3,0*5 78 4414'o.23 >
t 1,007.41 -h43 5. 4 8 4 val O. A1 ,
+7 1,007.41- vve 5. 3 S A34 o. M 4
4- 8 1,078.95 fv0 6; YY GrH4- O.24
-le9 N x IO S
$dM N/A C5 -1t 10 ,128.59 -
O.40 3 35 t
t
- The design pressure is governed by nurge line break. ,:
( Initial conditions for all nodes are identical .. Temp.
- 120'F,=
Press. - 14.7 psia, and relative humidity - 50%,
6.2 143 :Revisiongh i
._.,,,,,,_,...m_ .,....,.._,,,,,..0;_,__-,.~._,._,...-.-.,,._ , . . . , _ . ..,w, -,-...,,,-..,,.,,,r....%,.
i sTFECS UTSAR ATTACHMENT 4 ST.HL.AE-I PAGE_76 b GG_. OF M '
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' 6.2 144- Revision 0
4 1
i :
STPECS UFSAR -
l
} ATTACHM T& a .
- l. TABLE 6.2.1.2 13 ST HL AE.
- PAgg g b
/ j OF _///-7 ;
j \ -
j MAIN str>.M LINE AND TEEDt'ATER LINE SURCCMFARMENTS ANALYSIS f
}
I reak Time to Peak l Not Differential Differential Design Design i l- ,
Pressure Pressure Margin Volume Pressure d (ft 8 ) (esid) (see) (esid) (s) ;
H2de 6,030.42 12.97 0.019 30.15 132.4 l 1
[
15.00 **
, 2 16,982.55 +: M l. 3 7 0.055 3 5,332.80 13.65 0.017 30.15 120.9 fI .
4,4+ l . 58 *15.00 **
4 19,518.67 0.053 -
1 i'
5 6,766.11 6.45 0.029 14.25 120.9 1
.15.00 **
j 6 15,748,10 1,63 0.044 ;
l 7 5,958.95 6.44 0,026 14.25 121.3 9
0.044 15.00- - -**
8
. 8 19,100.75 -1.89 %
9 35,385.71 ih40 0 95 0.050 25.50 J
25.50 **
- 10 2,975.70 0,94 1.04 0.050 1.% '
11 -s . x 10' - . .
i 12 50,653.39 0.03 0.08 negligible N.A.
- 13 48,905.96 0.05 0.09 negligible N.A.
1
- 14 58,770.56 0.04 0.09 - neglisible N.A.
l- .
n.
i i
1 .
f 6
,, -t
- Large mar 15 n exists.
(
Initial conditions for all rodes are identical. Temp. - 120'F. -
Press. - 14.7 psia, and relative humidity - 506 6.2 147 ,
Revision,D J
. . , . , , ~ , . , , , , , , , , - - - - . . , + . - . . . , , , . . . - .
,_,_,,.,.m_,-r ,.,....,,,m-.-,.....,+
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RE2uteAflVE ut4T fERASEft wwwNmui prEAL DESCBtPTtes*
i i .
i
- s Seelp seelen Post Preeeure mergin Differentlet Preneure ,
vety fft I tanfal - tantal 31 4 sede ^ M Totlan M 6F\ 7.52 M i 3.
sortheast omdrent of El. 37 ft 3 In., 2.2sa.8 ,
1 t j between esimeth 4* and 279* and between secondary arid centelnment units.
\ l t
i (tous to. 307) i 3.38 .
-3.46 a 100 5/A i betence ef teetter Centelament Sulldftg
~
2 g
E. -
.e. . U l i*
5 l s
- . . . - e
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=
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TABLE 6.2.1.2-17 RADIOACTIVE PIPE CHASE SUBCOMPARTMENT NODAL DESCRIPTIONW Net Free Peak.Fressure Design Differential Pressure Margin Node' Descrintion Volp)
(ft unstA (nsirl (t)
- 1. Space between floors 5,331 4,43-1. 3 T 3r 1. 35- none at Els. 29 ft and 37 ft-3 in between the secondary shield and the 9
~ Containment and between g radial shield walls, g
'n e*
- 2. Space directly below 5.763 N/A N/A 7, node 1 on E1. 29 ft between azimuth 6 and 43+ (Room No. 210SE) 3.37 6
- 3. Balance of RCB W x 10 N/A N/A p nm
_i Ehs 5M5
- = -
wo5
- 3. 1 %D E i.. Initial conditions all nodes are 1.'entical: Temp. - 120-F, 1
$ Fress. - 14.7 psia, and humidity - 254. j
)%a .,
6 l
'4 V
j E
(
1
, i i
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- TABI.E'6.2.1.2-19
~
RHRIA VALVE ROOM SUBCONFARTMENT
?--
n I
a 4
Net Free Feek Pressure Design ,
\' ~
Vol Differential Pressure Margin <
(omini (dais) (t)
Node - Descriotion' (f 1 ,
t 1.877 1.5 1.5 191* ;
1 Northeast quadrant at ,
I E1.19. ft adjacent to ,
i SIS Accumulator tank ] -
(Room no. 209). T i j i
- j. 3.38 g '
2 Balance of RCB . -3.56 x 10 6 N/A i
! e. - es t
- p 3 N
{
U
- i j; -
i i j .
was.
- ~
$NP go I .
z >
rn K
-* t- nm 1
D3 ,
YF -
[
.\; .
Design margin is the result of initial design basis turntng out to be greater than final q I-
{ '
Edesign basis..
O
[!
v
r %
STPEcs Ur$AR ^ TACNM bif D ST.HL AE Lloq y PAGE Q 9p ,,
TABLE 6.2.1.3 16 'Q pf.BESSUR12ATION DIERCY REl#ASE C4tCU1ATIONS FOR DEISG VITH max 1Htm SkrTCUARDS (IDC Proken 1&29_l:2%D ibrarion Use the Fest reflood Table to 272 seconds calculated centainment. pressute at 272 seconds - 32.47 psig Mditional fraccion to be removed from the broken loop steam generator (56.5 32.47) / 56.5 - 0.425 Time required to equilibrate - (0.425) x (25,400,000) / (79,127) - 136.4 seconds Broken Leon Deeressurization Time - 3.600 secondr Energy to be Remoced - 25,400,000 x (1.0 - 0.425) - 14.605 x 10' Btu Heat Removal Rato - 14.605 x 10' / 3,600 - 4,056.9 Btu /aec Mass Boil-off Rate - 4,056.9/944.07 - 4.3 lbm/see Energy Addition Rate - 4.3 x 1,164.76 - 5,008.0 Bru/see Duration 408.4 to 4,008.4 seconds t
4 AM INSERT 6-13 Here 6.2 175 Revision 0 i
1 1
ATTACHMENT %
S UFSAR Table 6.2.1.1-16, Page 6.2-175 l r;T.10. AE '/O/_;
JjE ff_01: 6 4 'I'2 INSERT 6-13
- The results prenanted in this Table are for a containment free volume of 3.56 X 10 6 ft I. The revised analysis for a containment volume of 3.20 X 10 6 ft3 did not contain sufficient information to update this Table. Section 6.2.1.1.3.1 discusaus containment volume used .in IDCA analyses."
!1
ST E S WSAR ^
ATTACHMEA TR S T.HL.AE. ' O (, f, 'f
^
TABLI 6.2.1.3 16 (Continued) & 0FJ'/7 DIfffSStfRIZATION ENERGY PILEASE CALCU1ATIONS FOR DEPSC VITIM.XTMb5S ATECUAPl)S fn gq u oon Ecuilibration itse the Post.reflood Table to 947 seconds calculated contattunent pressure at 947 seconds - 24.42 psig Additional fraction to be reroved frota the irtact loop stears generator
($6.$ . 24.42) / $6.5 - 0.568 Time required to equ111 brace - (0.568) x (98,800,000)/(71,415 x 3) - 261.84-seconds Intnet toen Det.ressurization Titre - 2.799.56 seconds _
Energy to be Re.soved - 98,600.000 v. (1.0 0.$68) - 42.6 x 10' Btu Heat Removal :tata - 42.6 x 10'/2,799.$6 - 15.24$.3 Btu /soe Mass Boil.off Rate - 15.245.3/949.3 - 16.06 lba/sec Erergy Addition Rate = 2.6,06 x 1,162.0 - 18,661.0 Btu /see Duration - 1,208.84 to 4,00f.4 seconds
.~
l 6.2 176 Revision 0 i
l
STFECS UTSAR ATTACHMENT E ST HL AF TABl.E 6.2.1, bl6 (Continued) PAGE .2.,40
> 0F.(,0>M 2 DEPRESSlfRIZATION ENERCY REttASE CALCULATIONS FOR DEPSC VInl>4AXI- SAFECUARDS
- (thCA.2 V
/
I Su:gatry of the Relgig.es Between 0.0 and 272.0 seconds the Post reflood Table rhould be used.
Betwo,n 272.0 and 408.4 seconds the Post reflood Releases 'are:
j' Hass Rate (vapor) - 217.01 lbs/sec Mass Rate (liquid) - 1,519.84 lba/see Energy Rate (vspor) - 255,582 Btu /sec 2
Energy Rate (liquid) - 416,299 Stu/sec-Between 327.4 and 947 seconds the following releases should be added to those in the Post reflood Table Mass Rate (vapor) - 4.3 lba/sec Energy Rate (vapor) - 5,008.0 Btu /see Between 947 and 1,115.5 seconds the Post reflood Releases are:
Mass Rate (vapor) - 102.1 lba/sec Mass Rate (liquid) - 1,612.79/lbm/sec Energy Rate (vepor) - 1.465 x 10 5Btu /sec Energy Rate (liquid) - 4.417 x 10 Btu 5
/see Between 1,216.0 and 4,008.4 seennds the releases are a function of the decay heat ans the d= pressurization releases below Mass Rate - 102.1 lba/sec Energy Rate - 120,508 Beu/see From 4,008.4 seconds the releases are only a function of decay heat.
.~
d 6.2 177 Revision 0 t
4 4
. -
- sTrres UrsAt ATTACHMENT 1 i -
ST HL AE 40G G !
ust.E 6.2.2 5 PAGE JUl 0F ._/.!/.2 "
j EsmMnm or cM SPR4V MA$$ FM RATE fpR IN DWipudL. ErstoNS
, &ggE,MA:: 2. nTza-1, i
A The spray mass fit j rates for the various regions are as follows:
D Contairasent Dome Area 29.440 lb/ min i'
- 2) from operating floor (E1. 68 ft) 29,840 lb/ min to the springline (E1.1.53 ft)
- 3) Inside the secondary shield vall below E1. 19 ft 304 lb/ min i ;
- 4) between E1.19 ft and E1.q?ro '
'68 ftInside the
-1.101'1b/ min secondary shield w including the refuelin i ut id. C.. . m ' :, g,cavityd.i;1d 11 .
ovrstoe Tw seco.4044y 5HwuswAos ., i i 5)QEl. 52 ft to E1. 68 ft c,,7to -674te,1b/ min 6)E1.19 ft to E1. 52 ft c3E1. (-)2 ft to 11. 19 ft s,E 9 o + cit 9-Ib/ min el)selowE1.(-)2ft 2, son 4 M lb/ min
,, ggo W 1b/ain P
W 4
6.2 224 Revision 0-
[- , , , - , - , - - - . - - - - - - - - ~ ~ ~ ~ ~ ~ ~ ~ " ~ ~ ~ ~ ' " ~ ~
_ . . .- -- _ . - . _- - -. . . = . _
l ATTACHMENT 2.
' sTrtch Ur$AR ST HL AE tD6(o PAGE .RS_0F _.!N 7 TABl.E 6.2.5 4
(. {1RfT PAPAMETEP.S FOR CAlfU1ATINC POST ACff BENT a
ifYDROCEN CENEPATION l Humber of Loops 4 Thorwal Power Rating 4 @ MVT l Containment Free Volume x 108 fts Maximuu Normal containsent Terperature 120'T Weiz.ht Zirconium Cladding, 30.780 lbs Hydrogen irom RCS 1.169 sef
(
mr 6.2 236 Revision)f}
{
t_ . .
._-. - -_ . ~.. ... . = - . . - - - - - . . - -. .- _ - . . - - ~_ -. - ..
p ,
^
5HMEN Q --
6 ,
i I TAB 12 6.2.
I Post-LOCA--~
""" " ' r menosioN MATES USED IN - T - _
g CONTAINttENT HYDROCEN CENrnATION ANALYSIS '
9 d Temperature . Certosion Rates
('M (ib.molas/fts hr) ,
'. Nne Altmalnum 120 1.744 x 10-s 2.871 x 10** ,
150 1.744 x 10*8 1.141 x 10*'
175 3.907 x 10*8 3.603 x 10*'
9.952 x 10*s 1,13g x go o 200 . .
. 225 2.535 x 10-' 3.594 x 10**
- 250 6.457 x 10** . 1.135 x 10*.s 260 9.386 x 10*'- 1.798 x 10 s f,g4g x go s 270 1.364 x 10-s -
275 1,645 x 10 3 __3.5_85 x 10-s t .133 300 4.189 x 10-8 1- He x 10*'
3.l O 6.090 x 10~I 1 798f x 104 4
4 5
)
. , . t 4' .
A 6.2 239 Revision /
,e---- vm, ,- , n r ,n.,- c - - , , , , - , - - - , , e,--4.-a m-e-,mn,,,,4 . .m.,-- , .,.,,w4g ,,_ng,
ATTACHMENT 2.
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(330t saw!1)
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% ENERGY INVENTORY Figure 6.2.1.1 1 A $ Revision 0
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UNITS 1 & -2 POST ACCIDENT (LOCA) CONTAINMENT TEMPERATURE TRANSIENT I I
Figure 8.2.5-6 Revision 0 i
._ . . _ . _ - - . . . . . , , . , _ _ . . . . _ . - . . _ . - _ - . _ . . - - ~ - , - - _ _ _ . _ . _ . - - . - . - _ . - _ . . . , - . - - . . . _ _ .
STPECS UFSAR r ATTACHMENT %
ST HL.AE yoG. 4 TABLE 6.4 2 L PAGE llfa_. 0F /V"/-- ;
CONTROL ROM DO3E ANALYSIS AsJAul ens.
Containment leak *6e assumptions 0.36 (0 24 hrs)
(Bued on A CsMwtaf fret. L/s/v=e. of 3,4) y M.4 ) 0.156 (130 days) 3 ES fons 8,280 cu s /hr Pressurization askeup air inflow parameters:
flow rate 4 ts/ min filter efficiency
- 98.5% inorganic, 2 98.5% organic, 994 particulate control room envelope clean up air (recirculation) parameters!
filtered flow rate ts/ min (rectret.lation air) filter efficiency 954 inorganic, 95% organic, 996 particulate envelope free volume 274,080 fts l2 envelope unfiltered inleakage 10 fts/ min Meteorological dispersion factors (including wind speed and direction allowances):
Containment ESF leakage and laakage Case Purge Case 0 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.06x10*8 sec/ns 1.29x10'8 sec/m8 8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.03x10 see/as 8.55x10s ,,ef,s 1 4 days 4.45x10 sec/ms 5.42x10'8 sec/ms 4 30 days 1.91x10 sec/m e 2.32x10*8 sec/ms occupancy assunsptions:
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, control room 1004 1 4 days, control room envelope 606 4 30 days, control rova envelope 404 Breathing rate of operator 3.47 x 10 ma j,,e
- 1765 cfm is filtered through makeup and recirculation filters: 235 efs is filcered through makeup filters only. Effective filter efficiency for 2 2000 cfm is given above.
6.4 12 Revision 2
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i TACHMENT .2 a .
0 TABLE 6.4 2 (Continued) [A 'f 1
\ I CONTROL ROOM DOSE. ANALYSIS g
l f Resulta
- Whole Body Skin i Operator dose. 0 30 day period (rom)
- gid C---
v ,)ata c y 4
Contaitunent leakage M . 6,3 M l.H & :Al.6 M
- ESF leakage 1 58 .&r+tt (,.(,4.,67x10 8 a.i h46x104 2 ,
! Cemtainment purging 0.06f +r4Nr3 6.1 W10** q .4 . A.2x10
l Direct dose from Containnent ... 6 07 444 ...
l
! Direct dose from cloud of ... D.h") 4.46 ...
l 1 released fission products g ;
j Iodine filter loading ... 2.ftx10*8 ...
J i 4 Total = 3,A .6 ') 44,46- -26 5 Mrft 2.4/ l2 i
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ami minimize the release' of radioactive iodine to h environment. Thfa *se section describes b iodine removal capability. ef & Css, m analysis of. 'l
& radiological consequenceu of the Imss.of Coolant Accident (iDCA) is given f in section 15.6.5 "
6.5.2.1 Dasten maans. . m design bases of h Css. fot ruoving iedtne from 6 Containment atmosphere ares,' , - .
, 3.
- 1. CDC 41, as related to Containment ' atmosphere"eledup.
- e- ,. O i. g .
- 2. CDC 42 'aA related to inspection of Containment, atmosphere cleanup systems.
- w.,';yyt.i.
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4 ith the single failure of m
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- 5. The Css is designed to obtain adequate severage of the Containeout volume in order to limit (in conjunction with. ohr safeguards systems)-
- the offsite thyroid deses to a limit less than that established by
- 10CTR100, using the assumptions in 30 1.4. y 6
m spray nozzles are designed to minimize the possibility of clogging and to ' produce droplet sizes effective for lodine absorption.
( )
bo*h Mi A ' 'I PP"' 8i"" i' I Y .'t' 3
- 7. The pH of the3 i njection spray isf,1;. " .
.= e 7.; ; *0.0. The equilibrium pH of the Containment sump is '" "' - -^
'I.o to 9 3 -
J.. i. , the *JJt H.= '
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6.5.2.2 system Danien. - m C5s design la discussed in detail in .
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9
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sumps, it " :-- t'.rt , ": ;;;y n"' tin 2:' '"") M12: nt=: x; -
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h number of nozzles and the nessle spacing on each header is given in section 6.2.2. A schematic of the huders illustrating h nossie orientations is given on Figure 6.2.2 3. A f t :-';-* - ' *- ;rg "i-i -
, - -- -- r--' iil 1** ;--; 1-** 0112A- .
e top'al es une the Conta sent Artion un in abi 6.5. ; t}ig spra unt gral overs approxinhttiy ertent @ ksadar[given of\the "Itted f .-
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4 as h in Section 6. 2.3.4.t ftge va uno6 'fhe regions covered by the Containment spray s aMreMed
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' ATTACHMENT 1 sTrtcS UTSAR L_ ST PAGE HL 339AE 0F_40(c c, /// Z TAB 12 II.B.2-2 i
MSI, Acciorrr RADIATION IIvrt.5/D05E9 l
f j Continuous occupancy Areas: 30 day Domes (Remi Ca-= Beta Thyroid 5 .2 2. # 27..'70 f Control Room 2.4/ I+:et
- 4. rs Technical Support Center 4.8 2W 28 bx )i <
, ,.-- .)-
Infrequent Access Areas: ;
UFSAR Figure Dose Rate (RAfr)
Reference Area Time after accident 1 1 hr i day 1 wk 1 month 4
12.3.1 36 Post. accident .75 4.5 x 10*8 1.1 x 10a 6 x 10
( ,j sample station 12.3.1 27 Health Physics 6 x 10-s 3. 6 x 10*' 9 x 10'8 4.8 x 10*8 counting room 12.3.1 27 Radwaste count. 3.1 x 10 3 1.8 x 10*8 4.6 x 10 2.4 x 10's ing room l
12.3.1 28 Plant vent 4.74 .28 7.1 x 10'8 3.8 x 10*8 radiation
- monitor i
12.3.1 25 Auxiliary shut. 8 x 10 4.0 x 10'8 1.2 x 10*8 6.4 x 10
down panel i
7.A.II,B.2 7 Revision 2 a
. = - - ._. _ _ - . . - .- =
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Sittc5 UrSAR
~ ATTACHMENT A ST HL AE yo4 (,
PAGE. G O t"
- The CVCS and the SrpCCS vere designed to purify reactor TUU1wnt-th. OFw -
roug g.- .
exchangers af ter reactor shutdown and cooldovn. This ensures that the effect of activity spikes will not significantly contribute to the contatiment airborne activity during refueling operations. The contribution to airborne activity due to reactor vessel head removal is considered negligible as the RCS Vacuum Degassing System (Section 11.3) vill remove this activity prior to head removal. The detailed listing.of the expected airborne isotopic concentrations in typical accessible regions is presented in Table 12.2.2 2.
The final design of the plant ensures that the expect ,,4rgorne isotopic concentrations in the typical accessible regions are the maximum permissible concentration for the critical organ for tb appropriate isotope for occupational workers, as adjusted on the basis of expected occupancy in the regions (i.e., access to contairment is expected only 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per week during normal operation).
12.2.2.1 tiedel for Calculatine Airborne Concentrationg.
Plant areas with airborne radioactivity are characterized by a constant leak rate of a radioactive source at a constant source strength with a constant ex.haust rate of the contaminant, this leads to a peak or equilibrium airborne concentration of the radioisotope in the regions as calculated by the following equation:
C,(t) - (LR), A, (PT), (i.e%') (Eq. 12.2.2 1)
(VA,i)
(- where:
(1Jt), - leak or evaporation rate of the ith radioisotope in g/sec, in the applicable region, and A. - activity concentration of the ith leaking or evaporating radioisotope in pC1/g (PT), - partition factor or the fraction of the leaking activity that is airborne for the ith radioisotope A,, - total removal rate constant for the ith radioisotope in see '
from the applicable region
- ( A., + A.)
(i., and A. are the removal rate constanta in sec*' due to radioactive decay and the exhaust from the applicable region respectively for the ith radioisor. ope) i t - time interval between the start of the leak and the time at l which the concentration is evaluated in seconds V - free volume of the region in which the leak occurs in em' 12.2-5 Revision 0 i
, ~ . - , , , - - ,
STFECS UTSAR N
ATTACHMENT;l -I TAB 1.E 12.2.2 1 ST HL.AE.yg4 E' OF N "/
f12AMEIERS AND ftS5UMPTIONS U1G i DR CAlnftATING AIPl.0 PRE RADIOACTIVITY
- 1. Reactor Containment Building (RCB) g _. -
- a. Reactor.en coolaw .<n,t a.c.tivity concentrations are listed in Table 11.1 7 was a.a as T.eie si.e. g .c
.dj i+ed fer . p,wer fpven .4 Aios Wt.
- b. Release ~ rates from RCS:N '
Noble Cas 14/ day Iodines .0016/ day Particulates .00014/ day h1
- c. Containment volume is 1v68 x 10* f t'
- d. Purge rates:
4150 Normal operation - continuous 5000 scfm Prior and during refueling Jo,,ooo scfm
- e. During operation, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per week are assumed to be needed for Containment access.
- f. Refueling source term is due to evaporation of the fuel pool.
(Area - 1827 f t', 120'F, Air flow = 75 ft/ min Evaporation Rate Assumed 1.25 gal / min
- g. Refueling source tern:
1 pCi/,ga Trititus, 5.9 x 10-8 pcl/6m - 1 131 Tritium PT-1, 1 131 FF .01
- 2. Mechanical Auxiliary Building (MAB)
.-_.-. _ - _ _ _ _- m. --
- a. are(listed
, a p. in T4.ble 11.1-7 Re va,cactor coolant activitysi.e.
e_m_ecie,._t..g.sTe concentrations.ed r ore e digi ,,,c ye.,,l of 4toeMWt,
- b. MAB total leakage is assumed to be 160 lb/ day. All is assumed to be primary coolant,
- c. Nuclide Partition factors:
Noble Gas 1.0 Halogen .0075 Particulate .001 0
12.2 29 Revision 0
I STFLCS UFSAR -~
ATTACHMENT 7 TAB 12 12.2.2 1 (Continued) E- ST HL A0F.t/Oz,f,/v]
(.sp. , i PAPAMETERS AND ASSlMPTIONS U$1D IQR CALCU1ATING AIRBORNE RADIDACTIVITY
- d. Two cases are analyzed using the above cowson assumptions and the following case specific assumptions:
- 1) Ceneral Areat Volume - 45 x 10* ft' INAC Flow = scfm f
- 11) Vorst Pctential Room: Volume - M9-448 743 M INAC Flow - ^^ ::!=Jo s c [m (14akage for this case is taken to be 4o the 160 lb/ day)
- 3. Turbine Generator Building (TCB)
- a. Secondary activity is taken from Table 11.1 7 is eJji.sf ed 4*r
- Power
- b. Leakage is taken to be 1,700 lb/hr of secondary steas levet of 4880 MW'
- c. A partition factor of 1 was assumed for all nuclides '
- d. TCB volume is 3.01 x 10' f t' -
- e. TCB INAC flow rate was assumed to be 855,^00 aefs %
9 i t , t.go i 4. Fuel liardling Building (nib)
I a. Source of activity is the fuel pool where the following sources are assumed:
1 pC1/sm Tritium, 5.9 x 10 pCL/gm I.131
- b. Partition factors are:
Tritium 1.0 1 131 0.01
- c. Fuel Pool evaporation rate is assumed to be 1.02 gal / min (Area - 1,482 f t',120'F, Air flow - 75 f t/ min)
- d. ntB volume is 1.26 x 10' f t'
- e. DIB INAC flow rate is MM%O scfm 26,94*)
i 12.2 30 Revision 0
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12.2 31 Revision 0
7
\ RCB Noneal RCD Refuelinig w.B Oaseral Worst MAB TGB FHB MPC Canc. - Cone. Area Come- Roose Come. Come- Come. .
NtJCI.!DE (microCi/ent) (nueroCVed) (mmroci/ml) (meereci/ml) (rescreci/ad) (reseroCVest) (mscreCVal) 5E.06 6.3tE46 4.68E46 - - 3.59L10 5.08E.06 H-3 Kr43ra 6.0E 06 5. ORE 47 - 9.56E-10 6.16E48 1.3 tE-14 -
Kr-15 12 4 7.57E48 - 5.16Lil 3.23E49 7.23E-16 -
Kr47 1.0E46 1.25EM - 5.72Lf0 3.96E4 8.1 6 15 - -
Kr48 1.0E-06 7.01E47 - 1.80E49 1.IIE47 2 49E 14 - 2 Xe.13 tM 2.0E45 2.25E47 - 1.58 & l0 9.88E49 2.1 5 15 -
Xe-133M IE45 11406 - 9.14L10 5.73E48 I.2GE-14 -
Xe-333 IE45 6.09E.05 - 4.42F48 2.77E46 5.9 tE-I3 -
4 Xe-135 4.0E46 2.19E46 - 2.8CE49 l 77E47 3.72E-14 -
Ar 41 22 46 2.62E47 - - - - - y O
Br83 3.0E49 f.84Lil - 3.9513 2 605-11 8.88E.16 -
9E49 3.56E49 2.75LIO 1.8 E 11 1.19E49 1.99E-13 2.99E-10 N 1 131 5.39E-10 6.5tE-14 3-132 2.0E47 3.68E-10 - 8.14E-12 -
4 I-133 3E48 4.09 & O9 - 2.90E-ll 1.835 #9 1.05E-13 -
y I-134 5247 7.8IE-fl - 3.568-12 2.56610 3.56E-15 -
g 1-135 1.0547 f.37E-09 - 1.54Lil 9.85L10 6.59E-I4 - y- 1 I.1 5 13 8.IIE-16 5.0514 6.6518 - gn Rb46 7E48 -
1 2 -08 3.40E-II 2JEI3 1.4 5 11 1.6 E 15 -
Co-IM --
Co-136 2.0E47 1.81E-11 - 1.76E-13 7_90012 9.04Ll6 -
g _ l Co-137 1E 08 2.47E-11 - 1.69E-13 1.05E-Il 1.3 E 15 - -
Cr-52 2.0E 06 2.59L12 - . l.79E-14 1.12L12 1.3 E 16 - d 2.84LIS 1J8513 3.195-17 N Me-54 4.0E.08 4.l E l3 -
9 2 07 2.16E-12 - 3.47E-14 9.22E-13 1.095-I6 -
Fe-55 N Fe 59 5E48 1.3 E 12 - 917E-15 5.80E-13 7.84Ll7 -
Co-58 5.0E48 2.1511 - I.4 5 13 922E-12 1.1 5 15 -
m Cs4 92 49 2.78512 - 1.90E-14 1.19E-12 1.45Ll4 -
3r49 3 OE.08 4.76E-13 - 3 2 5 15 2.04E-13 3.25LI7 -
I
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j_ 5r-90 fE49 1.36E-14 - 9.2517 5.806-15 7.84E-19 -
o 6.84E-15 4.34E-13 2.4617 - 4 in St-91 3.0E47 5.48E-13 -
g Y-90 1.0E47 1.50L15 - 1.16E-17 714E-15 4.346t9 - m ":: > -
Y-91M 2.0E45 5.80E-14 - 3.64E-15 2.64E-13 I.89E-17 -
- US U 5.9 5 16 3.69E-14 4.8 5 18 M Y Y-91 3.0E 08 8.60LI4 -
ul I 3.53536 2.24E-14 1.62E-18 - 2' l T-93 IE47 2.89E-14 - i Zr-95 3fE48 Ll5E-14 - 5,58 & I6 3.49E-14 4.82E-18 -
g, d Nb-95 1.0E47 4.7 E I4 - 4.6E16 2.90E-14 4.88518 -
p 6.6EIS N Mo 99 2.0E-07 I.04E-10 - 7.99ttl3 5.00E-Il -
I.CE45 '7J3E-Il Sl?E-13 3DE-Il 1.39E-14 -
N Te-99M Re-103 . 8OE48 53E14 - 4.IIL:6 2.575-14 3.2518 -
6E49 f.36E-14 - 9.27E-17 5.80E-15 7.84E-19 -
Rs.106 9.6 037 1.lE42 Oli ', 4E44 1.0 MPC Fractice
15.6.5.3.1.1 Activity Release to Centairsment - ne offsite M doses OF _f--Q-- ---- lj resulting from a hypothetical accident releasing core activity ha.ve been analyzed. Activity releases of these magnitudes have a considerably lower probability than thatc associated with gap activity releases. For th analysis of this hypothetical determine the initial ase, the assumptions outlined in Ec 1.4 were used to activity release. Thus, a total of 100 percent of the core noble gas inventory and 50 percent of the core iodine inventory is assumed to be immediately available for leakage from the Contairement. The total core inventory is S ven in Table 15.A 1.
e Of the iodine activity released to the Containment, it is assumed that 95.5 percent is in the elemental form, 2 percent is in the organic or methyl iodine form, and 2.5 percent is in particulats form.
15.6.5.3.1.2 contaire nt Model Par == tare - The quantity of activity released through leakage from the Containment was calculated with a two. volume model of the Containment to represent sprayed and unsprayed regions of the Contairement. This modal is discussed in Appandix 15.3. -
The Containment leak rate to the atmosphere used in the analysis is the design basis leak rate indicated in the Technical Specifications. For the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the leak rate is assumed to be 0.30 percent per
- 2 day, while for the remainder of the 30 day period the leak rate is assumed to be 0.15 percent per day.
to the environment. This Containment leakage is assumed to leak directly agaer ishotahfr4e volume b che bnta nhnh h): An\cabkla\edsthe\155 7] l j
1111bc ft%r rart of this volume ns covered by the containment spray, while aone is not. The major portion of tb unspray d _ volume is within the 2
secondary shield wall blev the operatine noorJ QaanMdilcubw as appwxuaanen (40,0Q0 fV. V,g rged(vo(umggnasJ The transfer rate between the sprayed and unsprayed regions is assumed to be limited to the forced convectLv induced by the Rasetor contairueent Fan Cooler (RCFC) units. The number of units asstmeed in operation and the total mi.xing flow are presented in Table 15.6 10.- This assumed minimum flow rate conservatively neglects the effects of natural convection, steam condensation, and diffusion, although these effects are expected to enhance the mixing rate between the sprayed and unsprayed volumes. The majority of the RCFC air supply, except a small portion discharged to the done, is discharget. co tb space within the secondary shield wall, where it is relieved to the balance of the containment volume through tb vent areas. Th RCFC units are described more fully in Sections 6.2.2 and 6.2.5.
For fishion products other than iodine, the only removal processes considered are radioactive decay and leakage. Iodine is assumed to be removed by radioactive decay and leakage, plateoucqd The ,
effectiveness of the Containment sprayAfor %goremoval by the CSS. of the iodine the Containae,Jn atmosphere and the modeL used to determine the lodine removal pfficiencyrare discussed in Section 6.5.2. Only the elemental and particulate t
d iodine forms are assumed to be effectively removed by the spray. 4-epeey-remove 1w.>te-+f-14re-he"1--is-aeeumek until-the 4idm. Aemeneel--ledine-le-g re h:ed by-a '--ar a# 1U Mter dia-tin , th; dementel-epray r-----I race-le-eer " " '- -- T b;:v;- :1 w el-plet::= d yee & ule e removal-by-spray - e ee d ""- %
- H 2
d- ~n*aminatica factor.--($ :" 1^^ *-
15.6 1L_ Revision 2 '
THE ToTAt_ FREE Vo ME 4Ho SWED WWE @ M C N N W W BE h c%go mo ar,w,sooacsen su settias (..z.j --
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. ATTACHMENT R ST HL-AE yo(c G TNnFRT FOR PAGE 15t6 1A PAGE./.JJa. 0F N7 A spray removal rate of 20 hr" is assumed until the airborne alonental iodine is reduced by a factor of 60. After this time, the olomental spray removal rate is assumed to be zero. The dop,osition removal rate for elemental lodine is assumed to bo 4.5 hr until a DF of 100 is reached. No additional credit is taken for deposition after a DF of 100 is reachod. For part.4.culate iodino, a spray rewval rate of 6.9 hr is aqsumod until a IT of 50 is reached and it is then reduced to 0.7 hr* until e 1)F of 1000 in roached. Mter this time, the particulato spray 2 coval rate is annumod to bo zoro.
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ATTACHMENT 'l ST HL AE Ilotelo I Table 15.4 4 PAGE Lll0F //17 k' PARAMETERS USED IN R0D FJECTION ACCIDENT ANAINSIS Pa rarre t e rs Core thermal power. MVt 4,8 00 4r600 Tuel defects prior to accident 1.0t Sc tube leak prior to and during accident 1.0 gal / min Primary coolant concentrations Table 15.A 2 Secondary coolant concentrations Table 15.A 5 equivalent to Tech. Spec. limit)
Primary coolant mass. Ibs 576,000 Assumed gap inventory Table 15.A.- MOk d s* L ,y e b.<ypl ruel failed by accident - 10% of fuel ro s n ore Fuel melted by accident 0.25% of core
. Release to Containment (available j} for leakage) from fuel failed 1006 of gap inventory of noble gases and iodines from fuel melted 100% of noble gases and i 25% of iodines 1
Containment free volume, ft* 3.? 4,+6 x 10
~m_
Containment leak rate, t per day (6* sect en [(.bh ment ree vah., e of 3Alal0# 4h w . - v
! 0 24 hrs. 0.30 1 30 days 0.15 Activity released to primary coolant from fuel failed 100% of gap inventory of noble gases and iodines from fuel malted 100% of noble gases and l
50% of iodines Time between accident and equalization of primary and secondary system
(" pressures, see 1.250 15.4 43 Revision 0 t
,. STFECS UTSAR I.16 T2 -I Table 15.4 5 7 k[/ DOSES RESULTING FROM ROD FJECTION .ACCJDENT _
j Containment Leakage Contribution EZL . Stance
- Oct he doses T yroid, ren 3.5*l htt x 10 1 hole body 6amma, ram 4 2 4rt x 10*1 Sktn beta, ren 4.0 4,4 x 10*8 LFZ distance
- accident duration doses .
Thyroid, rea 4.99 4rt6 x 101
- Whole body gamma, rem 7,6 4 79 x Wa Skin beta, ren 2.9 4 ,6 x 1o*8 W
Secondary System Releans Contribution EZB distance. 0 2 hr doses Thyroid, rem 1. 0 *i : 1 ^~ 1 Whole body gamma, rem 5.09 -'rea x 10**
Skin beta, rea 7 x 10' LPZ distance accident duration C doses Thyroid. res 2.9 fr6 x 10'1 Whole body gamsa, tem 1,471,46 x 10'8 Skin beta, ren 5,oy + red x 10**
- Exclusion Zone Boundary distance te 1,430 meters.
.("; Outer boundary s! Iow Population Zore is 4,800 meters from plant..
15.4 45 Revision 0-
______________________u_______._ _ . -J
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ST HL 33 0FAE yo y,M reachedr-G epceye are easider d -fiveefve-only--in-the-sprepOQttnt-ob '
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15.6.5.3.1.3 Centainment laAh*re Domen - Doses resulting from activity leakage from the Containment have been calculated using the models presented in Appendix 15.B. The thyroid, whole body gamma and skin beta doses are presented in Table 15.611 for t.he EZB distance of'1,430 meters and the outer boundary of the 1.PZ at 4,800 s,eters.
15.6.5.3.2 ISF leakare Contribution: A potential source of fission product leakage following a 1DCA is the leakage from Engineered Safety Features (EST) components which are located in the Fuel llandling Building (IRB). This leakage may be postulated to occur during the recirculation phase for long term core cooling and containment cooling by sprays. The water contained in the Containment sumps is used after the injection phase and is recirculated by the ECCS pumps and the Containment spray pumps.
15.6.5.3.2.1 Fission Product Source Teng . Since most of the radiciodina released during the IDCA would be retained by the Containment sump water, due to operation of the CSS and the ECCS, it is conservatively assumed that 50 percent of the core iodine inventory is introduced to the sump water ,
ta be recirculated through the external piping systems.
Because noble gases are assumed to be available for leakage from the containment atsoephere and are not readily entrained in water, the noble gases are not assumed to be part of the source term for this contribution to the total IDCA dose.
15.6.5.3.2.2 LeahAgp_fssumDti us - The amount of water in the containment sumps at the start of recirculation is the total of the RCS water and the water added due to operation of the engineered safeguards, i.e., the ECCS and CSS. This amount has been calculated to be 512,494 gallons. This value is conservatively low to maximize iodine concentration in the sump water.
The ECCS recirculation piping and components external to the Containment are designed in accordance with applicable codes and are described in Section 6.3.
The CSS is described in Sections 6.2.2 and 6.5.2.
The maximum potential recirculation loop leakage is tabulated in Table 15.6 12. Each recirculation subsystem includes a high. head safety injection (ID(SI) pump, a low head safety injection (1J131) p*mp, a residual heat exchanger, the Containment sump, and associated piping and valves. Thus three separate subsystems are provided for recirculation, as well as for injection, each of which is adequate for long term cooling.
Since three redundant subsystems are available durug recirculation, leakaga for any component in any subsystem can be terminated by shutting down the IRSI and ID(S! pump associated with that subsystes and by closing the appropriate pump suction and discharge isolation valves.
l Marimum potential recirculation laakages are indicated in Table 15.6 12. The l 1eakage rate assumed for dose calculation purposes is conservatively twice the leakage rate given in Table 15.6-12.
15.6-15 Revision 2
1 i
STFECS UFSAR ATTACHMENT Q ST HL AE.Ho(,4 n PAGE./No op 74/7 TABLE 15.6 10 PARAMETERS USED IN ANALYSIS OF ft)SS OF-C001)Mr ACCIDENT OFFSITE DOSES Parameter Core thermal power, MVt 4,100 2 contairusent model 2 volume (sprey and unsprayed)
Activity released to Containment and available for leakage Noble gases 100t core activity-Table 15.A 1 W, * -
Iodines 50t core activity ^ .
Table 15.A 1 4 Form of iodine activity -
Elemental 95.5%
organic 2.04 Particulate 2.5%
containment free volume, ft8 Total 3,"Ao 4,46 x 108 Unsprayed _
50 _4,40 x to s 2 r -
Containment leakage rate, t per da (Easelen A pabiergf_ fq, Md% of 4.4lxleht 0 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0,30 1 30 days 0.15-Number of RCFC units operating 3 of 6 Tsa Mixing rate between sprayed and li,i4 ^775 }
unsprayed region, ft*/ min Removal coefficients _ _
W "
Eles tal y), 1 18.8
\ org ic, r'7sp O.
Ia
? icu te, 2 6 3- 2
- - I E men 1 p1 e ou , hr'1 spray -4 2 l-k spraye - 0. 6
^-
Assumed lodine decontamination factor (DF)
Elemental (spray stops at a DF of ) 100 Organic .-
l2 Farticulate 100 l ' '.LJMt'Y 1 w
[$; 15.6 33 Revision 2
T4 -
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STPEGS UFSAR AlTACHMENT R ~M ST HL AE 4ag TABLE 15.6 11 A E @R OF Juv DOSE RESULTING F1 TOM IARGE BREAK IDSS OF COOLANT ACCf DENT Pareaeter containment t#akare Dgn a Exclusion Zone Bounda 2 hr thyroid, rams 1.l 4 h40 x 108 whole body gamma, rems 2 2.(, ih+e- 2 skin beta, rems t.Al 1 r44-tov Population Zon 30 days thyroid, rems 6'A.6+ F)re's 2 whole body Samma, rems 1,*3/ 4,4 x 10'1 4
skin beta, rems 4,7 47G x 10 FSF Lenkar* Doses Exclusion Zone Boundary 0 2 hr thyroid, rems ,
g ,g4. 2.54 x 10**
whole body gamma, rems J g,33 -8ve x 10 2 skin beta, rems g -4. M x 10
lov Population Zone 0 30 days thytold, rems 3, 73 he x 104 ,
whole body gamma, rems g3 hti x 10 2 skin beta, rems
- g arm x 10 i Containment Purrine Doses f
Exclusion Zone Boundary 0 2 hr thyroid, rems gq A A bNM 3-whole body gamma, rems ( g ,o b44 x 101 1 2 skin beta, rems 3,4 4,44 x 10'8 low Population Zonr 0 30 days thyroid, rems A,40 3,44-2 whole body gamma, rems k 1,>3 4.44.x 10 4 skin beta, rems I, 4,9 erk x 10
Total Dosta j Exclusion Zone Boundary 0 2 hr thyroid, rems n,y &,66 x 108 whole bcdy gamma, rems a,3 TM skin beta, rems 2 l,22 1-it l
15,6 35 Revision 2
! 1
STFECS UPSAR A & HT S7.yt,4 UT PA(;g 3 y
'if
' TABLE 15.6 11 (Cont >.
DOSE RESULTING FROM 1ARCE EPEAK IDSS.0F.0001 ANT ACCIDENT l
Low Population Zone 0 30 days 6 '
thyroid, rems G,43 GrNr x 10 1 2 whole body Samaa, rems o, y3 GrW '
skin beta, rems g,p er49-l l
1 L
1
- Exclusion Zone Boundary is at 1,430 m. Outer boundary of low Population Zone is at 4,800 m.
15.6 36 Revision 2
STFECS UPSAR " gg ,
Purge Isolation monitors are given in section 11.5. Isolet a N Containment is desuibed in Section 6.2.4, which discusses the valves, mode o .-
operation, closure time, and other information. t *<
The Fuel Handling Building (ntB) Ventilation System is described in Section 9.4.2. Should a fuel handling accident occur in the ntB, the spant fuel pool ventilation monitors are capable of identifying that the activity release has taken place, diverting the building exhaust flow through the carbon filter units, and starting the booster fans. The spent fuel pool ventilation monitors are discussed in Section 11.5.
ne design basis fuel handitos accident is elassified as an ANh: Condition IV event, a limiting fault.
A block disgram summarizing various protection sequences for safety actions required to mitigate the consequences of this event is provided in Figures 15.0 28 and 15.0 29.
15.7.4.2 Analysis Assumotions. The assumptions postulated in tha calculation of the radiological consequences of a fuel handling accident in the DIB or the RC5 are consistent with the assumptions of Regulatory Guide (RC) 1.25,
- 1. Fission Product Inventories The 21scharged fuel assembly with the peak inventory is the assembly assweed to be dropped. The assembly inventory is determined assuming maximum full .
power operation at the end of core life immediately preceding shutdown. A (
)
decay period of 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> is applied. '
The gap mooel discussed in RC 1.25 is used to determine the fuel clad gap activities. Rus.10 percent of the total assembly todines and noble. gases, except 30 percent of the str 85 p The total assembly and fuel clad gap activities at tMWorTeIclior c +a% w I shutdown and the assumed time of the accident are given in Table 15.7-7.
- 2. Analysis of Consequences An analysis of a postulated fuel handling accident in both the ntB and the RCB is performed. The parameters used for the analysis are listed in Table 15.7 9.
The assumptions for the conservative RC 1.25 evaluation are
- a. The accident occurs 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> after plant shutdown. Radioactive decay of tne fission product inventory during this interval is taken into account.
- b. All the rods in one fuel assembly rupture, plus an additional 50 fuel rods assumed to be damaged by the dropped fuel assembly,
- c. The assembly damaged is the highest powered assembly in the core. (
The values for the individual fission product inventories in the (
15.7-4 Revision 0 l
1
_ _ _ _ _ _ __ _ _ _ _ _ _b
i ATTACliMENT 4 i
sTrtes UrsAa ST lit. AE-40%
PAGE _! u 5 0.' J3/ '7 damaged assembly were esiculated based upon the total core -
CMHvities s., M1eviith my a Fa'dfalWTa~clor of 13Weade t--t MO) e
~:m- ~
- d. The minimus water depth bliitveen the top of the damaged fuel rods I and the spent fuel pool surface or the refueling water surface is 23 ft.
- e. All of the gap activity in the damaged rods is released to the I refuellng water or the spent fuel pool and is assumed to consist of 10 percent of the noble gases other than Kr 85, 30 percent of the Kr 85, and 10 percent of the total radioactive iodin rods at the time of the accident. g yg,g g g, g. m y,,
- f. The noble gases released to the spent fuel pool or refuellig water '"
f"#
are released at ground level,to the environment. Credit is i assumed for isolation of the Containment. W"Ma'$
1
- g. Tha todine gap inventory is cos}wsed of 99.75 percent inorganic j species and 0.a.5 percent organic species.
- h. The spent fuel pool and refueling water decontamination factors (DFs) for iodine are taken as 100 in accordance with RG 1.25.
- 1. All iodine escaping from the spent fuel pool is exhausted over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period at ground level to the environment. The iodines are exhausted directly to the anvironment until the ntB isolation dampers close, diverting the exhaust through the
/r-, charcoal filters of the ntB Exhaust Air Subsystem. The nl8
( ! Exhaust Air Subsystem is described in Section 9.4.2.
- j. The charcoal filter efficiency is assumed to be 90 percent for inorganic iodine and 70 percent for organic lodine, according to RC 1.25.
- k. The iodines escaping from the refueling water pool in the RCB are sxhausted until the Containment is automatically isolated. A ground level release with no filtration is assumed.
- 1. The O to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> accident dispersion factors given in Table 15.B 1 are applicable.
15.7.4.3 Radiological Consecuence1 The thyroid and whole body doses at the EZB and the 1.PZ for the design basis fuel handling accidents are presented in Table 15.710, for accidents occurring in the MLB and the RCB.
15.7.5 Spent Fuel Cask Drop Accident In accordance with 10Cnt71, the spent fusi shipping cask is designed to sustain a free fall in air of 30 f t onto an unyielding surface followed by a specified puncture, fire, and immersion in water with the release of no more l than a specified small quantity of radioactivity. The design of the spent
! fr 4 handling equipment limits the postulated fall of a spent fuel shipping 8
'i issa than 30 ft. as described in Section 9.1.
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PAGE ./LI (n 0F' IV7-e ., TA31.E 15.7 9 FARAMITERS IfSED FOR PUEL HANDLING ACCIDIM1 Parameters
, ' tower level, Nut 4,100 Tims between plant shutdown and accident, hr 42 RCA normal purge isolation valvt dosure time, esc 60 _ _ _
Radial peaking factor- 1.65 ( Fw S) 1.l(certs - .d)
Activity in assembly gap at time of accident Table 15.7-7
- RCB dilution volume 3 '
(t of containment free volume)
Damage to fuel assembly, rods ruptured- 314 Form of iodine activity released to pool solution:
Inorganic iodine 99.75n
(
g,) ; Crganic iodine- 0.25%
Minimum water depth between top of 23 damage rods and pool surface, ft Decontamination factor in pool water:
Inorganic iodine 133 Organic iodine 1 Noble. gases 1 Filter efficiencies of FHB Exhaust Air Subsystem:-
Elemental iodine 906 Organic iodine - 704 FHB Exhaust isolation damper closure time, sec 20 DIB Exhaust isolation damper leakage, cfm 100 NIB d11ution volume, ft' 50,000
.W -
W bop oc+ivity cad'uswd a br _ Kooo (wD/rro dM )e bu, K/
7)-' . T-13 ) c.sf um e d % b e 12 prc en+ o f a ssembly och.Ty for %
o c,c .'cl e n+ inscd o conturowen+.
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( PAGE - /L/ OF f TABLE 15.7 10 DOSES RESULTING FROM FUEL HANDLING ACCIDENTS 4-t i
Accident Occurrine Inside Containment Exclusion zone boundary (1,430 m):
-Thyroid, rea 3(, . l +t-7 ars +
i Whole body gamma, rea l, Q ' h4N t, lo j Skin beta, rea g,44 h%. t low population zone (4,800 m):
l Thyroid, rem 'g,gh 7.99 1
Whole body gamma, ren Skin beta, res 3.3 3.1 x 10*8 3.9 x 10-2 l 4a3
)
i I
,' Accident Occurrina In nlB ,
1 l Exclusion zone bour.tary i
! Thyroid, rea 28.4 !.
- Whole body gamma, rem -2.73 x 10'1 2
. . Skin beta, rea 3.58 x 10'1 ,
Low population zone
! Thyroid, ren- 8.30 j Whole body gamma, tem 8.0 x 10 2 2 Skin ra ;a, ren '1.05 x 10'1 i
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